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#REDIRECT [[2CAN121202, Arkansas Nuclear One Unit 2: License Amendment Request to Adopt NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)]]
| number = ML12353A041
| issue date = 12/17/2012
| title = Arkansas Nuclear One Unit 2: License Amendment Request to Adopt NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)
| author name = Schwarz C J
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000368
| license number = NPF-006
| contact person =
| case reference number = 2CAN121202
| document type = Letter, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 716
}}
 
=Text=
{{#Wiki_filter:2CAN121202
 
December 17, 2012
 
U.S. Nuclear Regulatory Commission
 
Attn: Document Control Desk Washington, DC  20555-0001
 
==SUBJECT:==
License Amendment Request to Adopt NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)
 
Arkansas Nuclear One - Unit 2
 
Docket No. 50-368 License No. NPF-6
 
==Dear Sir or Madam:==
 
In accordance with 10 CFR 50.90, Entergy Operations, Inc. (Entergy) proposes to amend Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2). This License Amendment Request (LAR) requests Nuclear Regulatory Commission (NRC) review and approval for adoption of a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205, "Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants."  The LAR follows Nuclear Energy Institute (NEI) 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c)."  This submittal describes the methodology used to demonstrate compliance with, and transition to, National Fire Protection Association (NFPA) 805, and includes regulatory evaluations, probabilistic risk assessment (PRA), change evaluations, proposed modifications for non-compliances, and supporting attachments.
 
The transition includes the following high level activities: 1) a new fire safe shutdown analysis, 2) a new fire PRA, and 3) completion of activities required to transition the licensing basis to
 
10 CFR 50.48(c).
 
A series of reviews and observation meetings occurred as part of the transition process. These served to increase communication between the NRC and transitioning licensees, develop transition lesson learned reports from observation visits, improve the NFPA 805 Regulatory Guide and Inspection Procedures, gain insights on the Enforcement Discretion Policy, and develop a LAR template.
 
In addition to the Pilot Plant Process, NEI established the NFPA 805 Task Force, to ensure successful implementation of RG 1.205. The NFPA 805 Task force provided the interface between the pilot plants, the nuclear industry, and the NRC. The NFPA 805 Task Force, working with the NRC, developed a Frequently Asked Questions (FAQ) process for obtaining Entergy Operations, Inc. 1448 S.R. 333 Russellville, AR  72802 Tel  479-858-3110 Christopher J. Schwarz Vice President - Operations Arkansas Nuclear One
 
2CAN121202 Page 2 of 3
 
clarifications to RG 1.205, NEI 04-02, and NFPA 805. This process is discussed in the enclosed NFPA 805 Transition Report for ANO-2, Section 3.4. Attachment H of the report provides the FAQs that ANO-2 used to support transition to NFPA 805.
 
contains the ANO-2 NFPA 805 Transition Report with supporting attachments. The report provides the required technical and r egulatory assessments to enable the NRC to begin the review and approval of the new licensing basis.
 
Enclosures 2 and 3 contain the marked-up and re-typed pages, respectively, of the Operating License and Technical Specifications (TS).
Section 5.5 of Enclosure 1 contains the ANO-2 proposed implementation schedule for transitioning to the new fire protection licensing basis. The proposed modifications and implementation actions in Tables S-1 and S-2 of Attachment S provide Entergy's commitments in support of the NFPA 805 transitioning process. Enclosure 4 contains the summary of the new commitments associated with this request.
 
An update to the ANO-2 Safety Analysis Report (SAR) will be performed and submitted in accordance with 10 CFR 50.71(e). The station Fire Hazards Analysis (FHA), which is considered part of the SAR and common to both ANO units, will be revised as necessary and submitted consistent with the submittal of the ANO, Unit 1 (ANO-1) SAR, in accordance with 10 CFR 50.71(e). Because these submittals are controlled by regulation, no new commitment related to these submittals is proposed in this letter.
 
Should you have any questions concerning this letter, or require additional information, please contact Stephenie Pyle at 479-858-4704.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on December 17, 2012.
 
Sincerely, ORIGINAL SIGNED BY JAIME H. MCCOY FOR CHRISTOPHER J. SCHWARZ
 
CJS/dbb 
 
==Enclosures:==
: 1. NFPA 805 Transition Report
: 2. Proposed Operating License and Technical Specification Changes (mark-up)
: 3. Revised Operating License and Technical Specification Pages 4. List of Regulatory Commitments
 
2CAN121202 Page 3 of 3
 
cc: Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard  Arlington, TX  76011-4511
 
NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR  72847
 
U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8B1 One White Flint North 11555 Rockville Pike
 
Rockville, MD 20852
 
Mr. Bernard R. Bevill
 
Arkansas Department of Health    Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205
 
Entergy Operations, Inc. Arkansas Nuclear One - Unit 2 Enclosure 1 to 2CAN121202 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition
 
Transition Report
 
December 17, 2012
 
Arkansas Nuclear One - Unit 2 NFPA 805 Transition Report Enclosure 1 to 2CAN121202 Page i TABLE OF CONTENTS Executive Summary....................................................................................................................v  Acronym List..............................................................................................................................vi
 
==1.0 INTRODUCTION==
................................................................................................................1  1.1 Background..............................................................................................................1 1.1.1 NFPA 805 - Requirements and Guidance..................................................1 1.1.2 Transition to 10 CFR 50.48(c).....................................................................2 1.2 Purpose....................................................................................................................
3  2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM...........................................4 2.1 Current Fire Protection Licensing Basis...................................................................4 2.2 NRC Acceptance of the Fire Protection Licensing Basis.........................................7 3.0 TRANSITION PROCESS.................................................................................................23 3.1 Background............................................................................................................23 3.2 NFPA 805 Process.................................................................................................23 3.3 NEI 04 NFPA 805 Transition Process............................................................25 3.4 NFPA 805 Asked Questions (FAQs)......................................................................26 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS........................................................27 4.1 Fundamental Fire Protection Program and Design Elements................................27 4.1.1 Overview of Evaluation Process...............................................................27 4.1.2 Results of the Evaluation Process............................................................29 4.1.3 Definition of Power Block and Plant..........................................................30 4.2 Nuclear Safety Performance Criteria......................................................................30 4.2.1 Nuclear Safety Capability Assessment Methodology................................30 4.2.2 Existing Engineering Equivalency Evaluation Transition..........................41 4.2.3 Licensing Action Transition.......................................................................42 4.2.4 Fire Area Transition...................................................................................44 Arkansas Nuclear One - Unit 2 NFPA 805 Transition Report Enclosure 1 to 2CAN121202 Page ii 4.3 Non-Power Operational Modes..............................................................................47 4.3.1 Overview of Evaluation Process...............................................................47 4.3.2 Results of the Evaluation Process............................................................49 4.4 Radioactive Release Performance Criteria............................................................50 4.4.1 Overview of Evaluation Process...............................................................50 4.4.2 Results of the Evaluation Process............................................................51 4.5 Fire PRA and Performance-Based Approaches.....................................................51 4.5.1 Fire PRA Development and Assessment..................................................51 4.5.2 Performance-Based Approaches..............................................................54 4.6 Monitoring Program................................................................................................57 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program.......58 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program....................58 4.7 Program Documentation, Configuration Control, and Quality Assurance..............64 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805........................................................................64 4.7.2 Compliance with Configuration Control Requirements in Section 2.2.9 and 2.7.2 of NFPA 805....................................................67 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805.....70 4.8 Summary of Results...............................................................................................72 4.8.1 Results of the Fire Area Review................................................................72 4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase..............................................................................73 4.8.3 Supplemental Information - Other Licensee Specific Issues....................73
 
Arkansas Nuclear One - Unit 2 NFPA 805 Transition Report Enclosure 1 to 2CAN121202 Page iii
 
==5.0 REGULATORY EVALUATION==
........................................................................................80 5.1 Introduction - 10 CFR 50.48..................................................................................80 5.2 Regulatory Topics..................................................................................................84 5.2.1 License Condition Changes......................................................................84 5.2.2 Technical Specifications............................................................................84 5.2.3 Orders and Exemptions............................................................................84 5.3 Regulatory Evaluations..........................................................................................84 5.3.1 No Significant Hazards Consideration......................................................84 5.3.2 Environmental Consideration....................................................................85 5.4 Revision to SAR.....................................................................................................85 5.5 Transition Implementation Schedule......................................................................85
 
==6.0 REFERENCES==
.................................................................................................................
86 ATTACHMENTS.........................................................................................................................91  A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements...............................................................................................................A
-1  B. NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review.........................................................................................................B-1
 
C. NEI 04-02 Table B Fire Area Transition......................................................................C-1
 
D. NEI 04-02 Non-Power Operational Modes Transition.......................................................D-1
 
E. NEI 04-02 Radioactive Release Transition........................................................................E-1
 
F. Fire-Induced Multiple Spurious Operations Resolution......................................................F-1 G. Recovery Actions Transition..............................................................................................G-1
 
H. NFPA 805 Frequently Asked Question Summary Table...................................................H-1
 
I. Definition of Power Block.....................................................................................................I-1  J. Fire Modeling Verification and Validation (V&V).................................................................J-1
 
K. Existing Licensing Action Transition..................................................................................K-1 Arkansas Nuclear One - Unit 2 NFPA 805 Transition Report Enclosure 1 to 2CAN121202 Page iv L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))......................L-1
 
M. License Condition Changes...............................................................................................M-1
 
N. Technical Specification Changes......................................................................................N-1
 
O. Orders and Exemptions.....................................................................................................O-1  P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)........................................................P-1
 
Q. No Significant Hazards Evaluations..................................................................................Q-1
 
R. Environmental Considerations Evaluation.........................................................................R-1 S. Plant Modifications and Items to be Completed During Implementation...........................S-1
 
T. Clarification of Prior NRC Approvals..................................................................................T-1
 
U. Internal Events PRA Quality..............................................................................................U-1
 
V. Fire PRA Quality................................................................................................................V-1 W. Fire PRA Insights..............................................................................................................W-1 Arkansas Nuclear One - Unit 2 Executive Summary Enclosure 1 to 2CAN121202 Page v Executive Summary Entergy Operations, Inc. (Entergy) will transition the Arkansas Nuclear One, Unit 2 (ANO-2) fire protection program to a new Risk-Informed, Performance-Based (RI-PB) alternative per 10 CFR 50.48(c), which incorporates by reference NFPA 805. The licensing basis per 10 CFR 50, Appendix R, will be superseded.
 
In letter dated November 2, 2005 (0CAN110502, ML053140128), Entergy informed the NRC of the intent to transition ANO, Unit 1 (ANO-1) and ANO-2 to the 2001 Edition of NFPA 805.
The transition process consisted of a review and update of ANO-2 documentation, including the development of a Fire Probabilistic Risk Assessment (PRA) using NUREG/CR-6850 as guidance. This Transition Report summarizes the transition process and results. This Transition Report contains information:
Required by 10 CFR 50.48(c). Recommended by guidance document Nuclear Energy Institute (NEI) 04-02, Revision 2, and appropriate Frequently Asked Questions (FAQs). Recommended by guidance document Regulatory Guide (RG) 1.205, Revision 1.
 
Section 4 of the Transition Report provides a summary of compliance with the following
 
NFPA 805 requirements:
Fundamental Fire Protection Program Elements and Minimum Design Requirements  Nuclear Safety Performance Criteria, including:
o Non-Power Operational Modes o Fire Risk Evaluations  Radioactive Release Performance Criteria  Monitoring Program  Program Documentation, Configuration Control, and Quality Assurance Section 5 of the Transition Report provides regulatory evaluations and associated attachments, including:
Changes to License Condition  Changes to Technical Specifications, Orders, and Exemptions,  Determination of No Significant Hazards and evaluation of Environmental Considerations.
 
The attachments to the Transition Report include detail to support the transition process and results.
Attachment H contains the approved Frequently A sked Questions (FAQs) not yet incorporated into the endorsed revision of NEI 04-02 that were utilized by ANO-2 in the preparation of the License Amendment Request. These FAQs have been used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805. The methodologies associated with these FAQs have been included in the Transition Report for Nuclear Regulatory Commission
 
approval.
Arkansas Nuclear One - Unit 2 Acronym List Enclosure 1 to 2CAN121202 Page vi Acronym List AACDG Alternate AC Diesel Generator EDG Emergency Diesel Generator ADS Automatic Depressurization System EEEE Existing Engi neering Equivalency Evaluation ACW Auxiliary Cooling Water EFW Emergency Feedwater ADV Atmospheric Dump Valve EOOS Equipment Out of Service AFW Auxiliary Feedwater EOP Emergency Operating Procedure AHJ Authority Having Jurisdiction ERO Emergency Response Organization ANO-1 Arkansas Nuclear One, Unit 1 ESF Engineered Safety Features ANO-2 Arkansas Nuclear One, Unit 2 ESFAS Engineered Safety Features Actuation System ANS American Nuclear Society F&O Findings and Observations AOP Abnormal Operating Procedure FAQ Frequently Asked Question APCSB Auxiliary & Power Conversion Systems Branch FHA Fire Hazards Analysis AP&L Arkansas Power and Light FM Factory Mutual Research Corporation ASEP Accident Sequence Evaluation Program FP Fire Protection ASME American Society of Mechanical Engineers FPP Fire Protection Plan ATWS Anticipated Transient Without Scram FPRA Fire Probabilistic Risk Assessment BAMT Boric Acid Makeup Tank FR Federal Register BE Basic Event FRE Fire Risk Evaluation BHEP Basic Human Error Probability GE General Electric BTP Branch Technical Position GET General Employee Training BWR Boiling Water Reactor GDC General Design Criterion BWROG Boiling Water Reactor Owners Group GL Generic Letter CCDP Conditional Core Damage Probability HELB High Energy Line Break CCW Component Cooling Water HEP Human Error Probability CDF Core Damage Frequency HFE Human Failure Event CE Combustion Engineering HGL Hot Gas Layer CEA Control Element Assembly HPSI High Pressure Safety Injection CLB Current Licensing Basis HRA Human Reliability Analysis CPC Core Protection Calculator HRE High Risk Evolution CRD Control Rod Drive HRS High Risk Significant CST Condensate Storage Tank HVAC Heating, Ventilation, and Air Conditioning CT Current Transformer ICM Interim Compensatory Measures CVCS Chemical and Volume Control System IEF Initiating Event Frequency DBA Design Basis Accident IF Ignition Frequency DBD Design Basis Document IFA Internal Flood Analysis DID Defense-in-Depth IN Information Notice DHR Decay Heat Removal IPEEE Individual Plant Examination of External EventsEC Emergency Class ISLOCA Interfacing System Loss of Coolant Accident ECCS Emergency Core Cooling System KSF Key Safety Function ECP Emergency Cooling Pond LA Licensing Action Arkansas Nuclear One - Unit 2 Acronym List Enclosure 1 to 2CAN121202 Page vii Acronym List (continued)
LAR Licensing Amendment Request RA Recovery Action LCO Limiting Condition for Operation QCST Quality Condensate Storage Tank LER Licensee Event Report RCA Radiological Controlled Area LERF Large Early Release Frequency RCP Reactor Coolant Pump LOOP Loss of Offsite Power RCS Reactor Coolant System LPSI Low Pressure Safety Injection RG Regulatory Guide LRA Low Risk Significant RI-PB Risk-Informed, Performance-Based MCA Multi-Compartment Analysis RIS Regulatory Issue Summary MCC Motor Control Center RHR Residual Heat Removal MCR Main Control Room RP Radiation Protection MOV Motor Operated Valve RPS Reactor Protective System MSIS Main Steam Isolation Signal RPV Reactor Pressure Vessel MSIV Main Steam Isolation Valve RWT Refueling Water Tank MSPI Mitigating System Performance Index RWST Refueling Water Storage Tank MSO Multiple Spurious Operation SAR Safety Analysis Report MSSV Main Steam Safety Valve SDBCS Steam Dump and Bypass Control System NEC National Electric Code SDC Shutdown Cooling NEI Nuclear Energy Institute SDM Shutdown Margin NFF Non-power Compartment Fire Ignition Frequency SER Safety Evaluation Report NFPA National Fire Protection Association SFP Spent Fuel Pool NPP Nuclear Power Plant SG Steam Generator NPO Non-Power Operations SI Safety Injection NRC Nuclear Regulatory Commission SLCS Standby Liquid Control System NSCA Nuclear Safety Capability Assessment SMA Seismic Margin Analysis OL Operating License SOPP Shutdown Operations Protection Plan OMA Operator Manual Action SORV Solenoid Operated Relief Valve OSHA Occupational Safety and Health Administration SPDS Safety Parameter Display System P&ID Piping and Instrument Diagram SR Supporting Requirement PCRS Paperless Condition Reporting System SRV Safety Relief Valve PCS Primary Control Station SSA Safe Shutdown Analysis PDMS Plant Data Management System SSC Structure, System, or Component POS Plant Operational State SSCA Safe Shutdown Capability Assessment PRA Probabilistic Risk Assessment SSE Safe Shutdown Earthquake PRM Plant Response Model SSEL Safe Shutdown Equipment List PSF Performance Shaping Factor SSLD Safe Shutdown Logic Diagram PWR Pressurized Water Reactor STM System Training Manual PWROG Pressurized Water Reactor Owners Group SW Service Water QA Quality Assurance SWGR Switchgear QCST Quality Condensate Storage Tank T-H Thermal-Hydraulic Arkansas Nuclear One - Unit 2 Acronym List Enclosure 1 to 2CAN121202 Page viii Acronym List (continued)
THERP Technique for Human Error Rate Prediction UL Underwriter's Laboratories, Incorporated TR Transition Report V&V Ve rification an d Validation TRM Technical Requirements Manual VFDR Variances From Deterministic Requirements TS Technical Specification ZOI Zone of Influence TSC Technical Support Center 
 
Arkansas Nuclear One - Unit 2 1.0 Introduction Enclosure 1 to 2CAN121202 Page 1
 
==1.0 INTRODUCTION==
 
The Nuclear Regulatory Commission (NRC) has promulgated an alternative rule for fire protection requirements at nuclear power plants, 10 CFR 50.48(c), National Fire Protection Association (NFPA) Standard 805 (NFPA 805). Entergy Operations, Inc. (Entergy) is implementing the Nuclear Energy Institute (NEI) methodology NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," to transition Arkansas Nuclear One, Unit 2 (ANO-2) 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 ANO-2 complies with the new requirements.
 
1.1 Background
 
1.1.1 NFPA 805 - Requirements and Guidance
 
On July 16, 2004, the NRC amended 10 CFR 50.48, "Fire Protection," to add a new subsection, 10 CFR 50.48(c), which establishes new Risk-Informed, Performance-Based (RI-PB) fire protection requirements. 10 CFR 50.48(c) incorporates by reference, with exceptions, the National Fire Protection Association's NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants - 2001 Edition," as a voluntary alternative to 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.
 
As stated in 10 CFR 50.48(c)(3)(i), any licensee's adoption of a RI-PB program that complies with the rule is voluntary. This rule may be adopted as an acceptable alternative method for complying with either 10 CFR 50.48(b) for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979, or 10 CFR 50.48(f) 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 1 Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.
Arkansas Nuclear One - Unit 2 1.0 Introduction Enclosure 1 to 2CAN121202 Page 2 A depiction of the primary document relationships is shown in Figure 1-1:
 
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
 
Entergy submitted a letter of intent to the NRC on November 2, 2005 (0CAN110502, ML053140128), for ANO-2 to adopt NFPA 805 in accordance with 10 CFR 50.48(c). By letter dated December 22, 2008 (0CNA120805, ML083500404), the NRC granted an enforcement discretion period based, in part, on the date in which pilot plant submittals were approved by the NRC. By letter dated July 28, 2011 (0CNA071107, ML112030193), the NRC extended the enforcement discretion period for ANO-2 based on a commitment by Entergy to submit the letter of transition to NFPA 805 no later than March 30, 2012.
code o f federal regulations 1 1 0 0  C C F F R R  5 5 0 0..4 4 8 8 ((c c))  N N a a t t i i o o n n a a l l  F F i i r r e e  P P r r o o t t e e c c t t i i o o n n  A A s s s s o o c c i i a a t t i i o o n n  S S t t a a n n d d a a r r d d  N N F F P P A A  8 8 0 0 5 5  1 1 9 9 8 8 5 5  LITTERA SCRIPTA MANET R R G G  1 1..2 2 0 0 5 5  P P e e r r f f o o r r m m a a n n c c e e--B B a a s s e e d d  S S t t a a n n d d a a r r d d  f f o o r r  F F P P  f f o o r r  L L i i g g h h t t  W W a a t t e e r r  R R e e a a c c t t o o r r  E E l l e e c c t t r r i i c c  G G e e n n e e r r a a t t i i n n g g  P P l l a a n n t t s s  N N F F P P A A  8 8 0 0 5 5  2 2 0 0 0 0 1 1  e e d d.. N N E E I I  0 0 4 4--0 0 2 2  R R I I--P P B B  F F O O R R  E E X X I I S S T T I I N N G G  L L I I G G H H T T--W W A A T T E E R R  N N U U C C L L E E A A R R  P P O O W W E E R R  P P L L A A N N T T S S  G G U U I I D D A A N N C C E E  F F O O R R  I I M M P P L L E E M M E E N N T T I I N N G G  A A  R R I I--P P B B  F F P P  P P R R O O G G R R A A M M  U U N N D D E E R R  1 1 0 0  C C F F R R  5 5 0 0..4 4 8 8 ((C C))  E E n n d d o o r r s s e e m m e e n n t t I I n n c c o o r r p p o o r r a a t t i i o o n n  b b y y  R R e e f f e e r r e e n n c c e e G G u u i i d d a a n n c c e e  R R e e q q u u i i r r e e m m e e n n t t s s  N N U U C C L L E E A A R R  E E N N E E R R G G Y Y  I I N N S S T T I I T T U U T T E E Arkansas Nuclear One - Unit 2 1.0 Introduction Enclosure 1 to 2CAN121202 Page 3 By letter dated March 27, 2012, Entergy submitted a License Amendment Request (LAR) to adopt NFPA-805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)" (2CAN031201) (ML 12087A113). However, by letter dated September 7, 2012 (2CNA091201) (ML 12208A196), the NRC denied acceptance of the March 27, 2012, LAR and in addition, denied a request for extension of enforcement discretion.
In accordance with NRC Enforcement Policy, the enforcement discretion may be regained following re-submittal and NRC acceptance of the LAR. The enforcement discretion period would then continue until NRC approval of the LAR is completed.
 
1.1.2.2 Transition Process
 
The transition to NFPA 805 includes the following high level activities:
 
A new Nuclear Safety Capability Assessment (NSCA);  A new Fire Probabilistic Risk Assessment (PRA) using NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, as guidance;  Completion of activities required to transition the pre-transition Licensing Basis to 10 CFR 50.48(c) as specified in NEI 04-02 and RG 1.205.
1.2 Purpose
 
The purpose of the Transition Report is as follows:
: 1) Describe the process implemented to transition the current fire protection program to comply with the additional requirements of 10 CFR 50.48(c); 2) Summarize the results of the transition process; 3) Explain the bases for conclusions that the fire protection program complies with 10 CFR 50.48(c) requirements; 4) Describe the new fire protection licensing basis; and 5) Describe the configuration management processes used to manage post-transition changes to the station and the Fire Protection Program, and resulting impact on the Licensing Basis.
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 4 2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM 2.1 Current Fire Protection Licensing Basis ANO-2 was licensed to operate on July 18, 1978. The fire protection program at ANO-2 is based on the NRC requirements, as well as the requirements of state and other federal agencies, and insurance carriers. With regard to NRC criteria, the ANO-2 fire protection program addresses the guidelines of Appendix A to the Auxiliary and Power Conversion Systems Branch (APCSB) Technical Position 9.5-1 (APCSB 9.5-1). Various aspects of the fire protection program are detailed, as required, to show conformance with the guidelines or to demonstrate the equivalency of alternative approaches, and the following license condition:
 
ANO-2 license condition 2.C.(3)(b), as approved by letter dated March 31, 1992 (0CNA039215),
states:
(b) Fire Protection EOI shall implement and maintain in effect all provisions of the approved fire protection program as described in Amendment 9A to the Safety Analysis Report and as approved in the Safety Evaluation dated March 31, 1992, subject to the following provision:
The licensee 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, ANO-2 license condition 2.C.(3)(e) states:
(e) Arkansas Power & Light (AP&L) 1 shall complete the following modifications by the indicated dates in accordance with the staff's findings as set forth in the fire protection evaluation report, NUREG-0223 "Fire Protection Safety Evaluation Report."  Implementation Dates for Proposed Modifications Applicable Section of NUREG-0223 Date      3.1  Portable Radio Communication Equipment March 31, 1979      3.2  Separation of Power Cables in Manholes
* 3.3  Protection from Water Spray
* 3.4  Protection of Redundant Cables in the MCC Room (2096-M)
December 30, 1978      3.5  Protection of Redundant Cables in the Hallway - Elevation 372 (2109-U)
  *, **      3.6  Protection of Redundant Cables in the Cable Spreading Room (2098-L)
* Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 5 Applicable Section of NUREG-0223 Date      3.7  Protection of Redundant Cables in the Switchgear Room (2100-Z)
* 3.8  Protection of Redundant Cables in the Electrical Equipment Room (2091-BB)
September 30, 1978      3.9  Protection of Redundant Cables in the Lower South Piping Penetration Room
 
(2111-T)    September 30, 1978      3.10  Protection of Safe Shutdown Cables in the Upper South Piping Penetration Room
 
(2084-DD)
 
September 30, 1978      3.11  Protection of Redundant Reactor Protection System Cables (2136-I)
  *, **      3.12  Fire Dampers  September 30, 1978      3.13  Portable Extinguisher for the Control Room (2199-J)  November 15, 1978      3.14  Smoke Detectors  *, **      3.15  Manual Hose Stations (2055-JJ, 2084-DD, Containment, Elev. 317' of Auxiliary Building)
  *, **      3.16  Portable Smoke Exhaust Equipment  December 1, 1978      3.17  Emergency Lighting  December 1, 1978      3.18  Reactor Coolant Pump Oil Collection System
* 3.19  Control of Fire Doors  March 31, 1979      3.20  Administrative Control Changes  December 1, 1978 (Numbers in parentheses refer to fire zone designations in the AP&L fire hazards analysis.)
1 AP&L is the predecessor to Entergy Arkansas, Inc.
* Prior to startup following the first regularly scheduled refueling outage. ** Technical Specifications covering these items should be proposed not later than 90 days prior to implementation.
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 6 See Attachment M and Enclosures 2 and 3 for proposed changes to license conditions 2.C.(3)(b) and 2.C.(3)(e).
 
By letter dated July 18, 2007, new license condition 2.C.(10) was added to address Section B.5.b. of the February 25, 2002, Interim Compensatory Measures (ICM)
Order (EA-02-026) and related NRC guidance, associated with loss of large areas of the plant due to explosions or fire, including those that an aircraft impact might create. This license condition will be maintained with the ANO-2 transition to NFPA 805.
 
(10) Mitigation Strategies
 
The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:
 
(i) Fire fighting response strategy with the following elements: 1. Pre-defined coordinated fire response strategy and guidance
: 2. Assessment of mutual aid fire fighting assets
: 3. Designated staging areas for equipment and materials
: 4. Command and control 5. Training of response personnel
 
(ii) Operations to mitigate fuel damage considering the following: 1. Protection and use of personnel assets
: 2. Communications
: 3. Minimizing fire spread 4. Procedures for implementing integrated fire response strategy 5. Identification of readily-available pre-staged equipment
: 6. Training on integrated fire response strategy
: 7. Spent fuel pool mitigation measures (iii) Actions to minimize release to include consideration of: 1. Water spray scrubbing
: 2. Dose to onsite responders In addition to the above, ANO-2 Technical Specifications require the following:
 
6.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:
: c. Fire Protection Program implementation
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 7 2.2 NRC Acceptance of the ANO-2 Fire Protection Licensing Basis
 
In 1977-78, AP&L, hereafter to be referred to as Entergy, which manages ANO, conducted a fire hazards analysis study to meet the criteria of APCSB 9.5-1 for ANO-2. The results of this study were submitted to the NRC on February 28, 1978 (0CAN027802), and subsequently, the ANO-2
 
fire protection program was documented in NUREG-0223 dated August 30, 1978 (2CNA087812).
 
The 20 commitments delineated in license condition 2.C.(3)(e) of Section 2.1 above have been satisfied (or exemptions received), are historical, and will be deleted from the operating license during the transition to NFPA 805. A discussion of these commitments follows:
 
3.1 "Portable radio communication equipment will be provided, and available for fire brigade use."
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203): 
"The NRC inspectors found that portable radios were stored on the first floor of the licensee's administrative building. The portable radios were in a locker which included built-in electrical charging outlets. The portable radios had separate microphones which could be clipped to the user's collar and throat microphones for use under respiratory equipment. Licensee representatives stated that all members of the fire brigade received instructions in the use of the portable radios as part of "Emergency Response" training and that the portable radios were used on some of the drills. The portable radio frequency assigned for fire brigade use was monitored in the control rooms. There were no violations or deviations identified."
3.2 "Redundant power cables for service water pumps and fuel transfer pumps will be separated by a barrier where redundant cables are in a common manhole in the yard
 
area."  This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203): 
"Manholes were not inspected by the NRC inspector because the licensee did not consider that they should be entered during mode 1 (power) operation, which both units were in. From records, it was established that Unit 2 barriers were installed-"  "-no violations or deviations were identified."
The remaining information in the above inspection report excerpt is related to ANO, Unit 1 (ANO-1) and, therefore, is not included here.
 
At the time, the open issue related to the fuel transfer pumps for Emergency Diesel Generators (EDGs) was that the redundant power cables for the pumps were routed together. This did not meet the concern addressed in Item 3.2, nor did it meet Appendix R requirements, which were issued later. To address this concern, Entergy modified the fuel transfer system to allow cross-connect capability between units such that an EDG could be supplied with fuel oil from any available fuel transfer pump. This emergency capability is addressed in station procedures to this day.
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 8 With regard to Service Water (SW) pump cable separation in common manholes, an exemption was granted by the NRC in letter dated March 22, 1983 (0CNA038328).
This exemption letter is discussed in further detail following the discussion of Item 3.20 of this section.
3.3 "A few vital AC panels and safety-related motor control centers will be provided with drip and spray shields where this equipment is in an area protected with an automatic water suppression system."
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Drip/spray shields have been installed for the protection of AC Panels 2RS1, 2RS2, 2RS3, and 2RS4. Motor Control Center 2B61 was not inspected because of access limitations but the work required was documented. Motor Control Center 2B51 was inspected and a drip/spray shield had been installed; however, because of the location of a spray head on the water spray system, this panel could be subject to flooding. This was discussed with a licensee representative, and the licensee has agreed to investigate the adequacy of the panel seals and/or relocation of the spray head. This is an open item (50-368/8212-04). There were no violations or deviations identified."
In letter dated July 25, 1985 (0CNA078522), the NRC stated:  "Open Item 50-368/8212-04:  This item was open pending determination by the licensee that Motor Control Center 2B51 would not be subject to flooding because of the location of the water suppression system spray head. The NRC inspector determined that the licensee has installed a deflector spray shield between the spray head and the panel.
It has also been confirmed that the panel accesses are all rubber gasketed. This item is closed."
3.4 "The red division tray in the vicinity of the green division motor control center will be provided with a fire protective insulation, and the green division cables in the vicinity of the red tray will be sprayed with a flame retardant coating." (2096-M 2B63)
In letter dated May 12, 1980 (2CAN058009), Entergy stated: 
"The green division cable tray in this MCC Room is completely covered with a thin galvanized sheet metal flame shield. Where cables penetrate this flame shield, they and the hole which they pass through are completely covered with carboline. Where the red tray crosses over the green tray, the green cables have been entirely covered over with carboline."
3.5 "To protect redundant safe shutdown cables in the auxiliary building hallway - elevation 372 feet, either a deluge system actuated by heat and smoke detectors and coating of cables where redundant cables are in proximity will be provided, or all cables will be coated and smoke detectors and a wet pipe sprinkler system installed." (2109-U)
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):
"The NRC inspectors found that the licensee had installed heat and smoke detectors which actuated a deluge system and had coated redundant cables. Although the hallways are separate spaces, licensee action was similar for both. The installed deluge system appeared to meet National Fire Code requirements. The licensee was reviewing the adequacy of action taken as described in paragraph 2 of this report.
Pending completion of the licensee's review, the NRC inspectors noted the following:
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 9
: a. The barrier material used was not yet accepted as a "3-hour barrier."  b. There were several terminal boxes that were not covered by barrier material, although the conduits on both sides of them were covered. Examples are terminal boxes 561, 346, and 345.
: c. Conduit above door 57 (Unit 1) appeared to be covered with barrier material over only half of its length.
Resolution of these specific items is considered to be an open item pending completion of the licensee's review under 10 CFR 50, Appendix R, Item III.G (50-313/8215-04; 50-368/8212-03). No violations or deviations were identified."
In letter dated July 25, 1985 (0CNA078522), the NRC stated:  "Specific item resolution was inspected and it was found that the installation of the barrier material had been significantly redefined. The separation criteria requires only a "1-hour barrier" and this has been installed on conduits with cables of concern and is the Hemyc Barrier System. Terminal boxes in the conduit runs were covered. The conduit above Door 57 was determined not to contain any cables of concern. Adequacy of the review and determination of the cables of concern will be addressed and inspected during the inspection for Appendix R compliance. This item is closed."
3.6 "The existing water spray system in the cable spreading room will be evaluated to the criteria of NFPA-15, 1977, and upgraded as required to meet this code. Flame retardant barriers or cable coating will be provided in areas where redundant safe shutdown cabling are in proximity to each other." (2098-L)
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "The water spray system in the cable spreading room meets the criteria of NFPA-15, 1977. Flame retardant barriers and cable coating have been provided in areas where redundant safety shutdown cabling is in proximity to each other. There were no violations or deviations identified."
3.7 "Flame retardant barriers or cable coating will be provided in the switchgear rooms where redundant cables are in proximity to each other." (2100-Z South)
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Flame retardant barriers and cable coating have been provided in the two switchgear rooms where redundant cables are in proximity to each other. There were no violations or deviations identified."
3.8 "A fire retardant board will be added to cover two cable trays where redundant diesel generator cables are in proximity to each other." (2091-BB North Electrical Equipment Room)  This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Fire retardant board has been installed to cover the two cable trays that contain redundant diesel generator cables that are in proximity to each other. The NRC inspector found on examination that the boards had been subsequently damaged and no longer meet the installation criteria. This condition was pointed out to a licensee representative and is an open item pending repair or replacement of the boards (50-368/8212-05)."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 10 In letter dated July 25, 1985 (0CNA078522), the NRC stated:  "Open Item 50-368/8212-05:  This item was open pending repair or replacement of damaged fire retardant board installed on a cable tray in the electrical equipment room. After the Appendix R, evaluation it was found that separation was met and that it was not necessary to take credit for the board. This item is closed."
3.9 "A fire retardant board barrier will be installed over the nonsafety-related trays that present combustible pathways between redundant diesel generator cables." (2111-T Lower South Electrical Penetration Room)
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "The redundant diesel generator cables have been protected by installation of fire retardant boards. There were no violations or deviations identified."
3.10 "A flame retardant coating will be applied to one conduit containing cable associated with a diesel generator." (2084-DD Upper South Piping Penetration Room)
 
Fire Zone 2084-DD was shown to be in compliance with 10 CFR 50, Appendix R, in Entergy letter dated July 1, 1982 (0CAN078202), "Results of Appendix R Compliance Review."  The ANO Fire Hazards Analysis (FHA) states:  "A flame retardant coating has been applied to one conduit containing cable associated with a diesel generator as per 0CAN067803 and Unit 2 SER (NUREG-0223), Item 5.4, August 1978."
3.11 "Fire detection devices and a water spray system will be provided in the corridor at elevation 386 feet of the auxiliary building to protect the four channels of the reactor protection system indication." (2136-I Health Physics Room)
 
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Fire detection devices and a water spray system have been installed and provide protection for the four channels of the reactor protection system. There were no violations or deviations identified."
3.12 "Additional fire dampers are being installed in ventilation duct penetrations of fire barriers."
 
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Properly rated fire dampers have been installed in ventilation duct penetrations of fire barriers. There were no violations or deviations identified."
3.13 "A portable water or halon extinguisher will be provided in or adjacent to the control room." (2199-J)
 
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "The NRC inspectors found that a small (10 pound) halon (1211) fire extinguisher was permanently mounted in each control room. No violations or deviations were identified."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 11 3.14 "a. Heat and smoke collectors will be provided over the smoke detectors in the containment building. b. Smoke detectors will be provided in each control room cabinet which contains safe shutdown equipment. c. Additional smoke detectors will be provided as listed in the applicant's letter, dated July 7, 1982, such that detectors are provided in all safety-related areas containing combustibles. d. Smoke detectors will be provided as listed in the applicant's letter, dated July 7, 1978, in various safety-related areas which contain no combustibles, but which contain redundant safe shutdown cabling in conduit. e. Non UL or FM listed detectors in safety-related areas will be replaced with approved devices."  With the exception of smoke detectors in the Containment Building (referred to as Item 'a' in the original NRC Safety Evaluation Report or SER), this requirement was
 
documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Item a. could not be inspected at this time and will be an open item pending conditions suitable for inspection (50-368/8212-06). Item b. is considered satisfactory since smoke detectors have been installed in each control room cabinet which contains safe shutdown equipment. Items c., d., and e. were inspected on a sampling basis and found satisfactory. There were no violations or deviations identified."
In letter dated June 10, 1985 (0CNA068503), the NRC stated:  "(Closed) Open Item 368/8212-06:  Smoke and heat detectors in containment. The NRC inspector toured the Unit 2 containment and verified that heat and smoke collectors were installed over the heat and smoke detectors in the containment, as required by the Unit 2 Fire Protection Safety Evaluation Report."
3.15 "Manual hose stations accessible to the lower south piping penetration area will be provided. The 75-foot length of hose at the hose station serving the upper south piping penetration area will be replaced with a 100-foot length of hose. Manual hose stations will be provided in the containment building." (2055-JJ Lower South Piping Penetration Room, 2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room, 2032-K Containment South, 2033-K Containment North, and Elevation 317' of the Auxiliary Building)
 
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "By record review and interview, the NRC inspectors established that hose reel stations were installed in containment (2HR-43 through 2HR-55) and that the 75-foot hose length had been replaced by a 100-foot length. The NRC inspectors reviewed DCP 80-2002. This DCP installed three hose reel stations, two on elevation 317' of the auxiliary building and one on elevation 335' of the same building."
3.16 "Portable smoke exhaust units with flexible ductwork will be provided so that three units are available to each ANO-1 and ANO-2."
 
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203): 
"The NRC inspectors checked three fire carts, one located for Unit 1, one for Unit 2, and one on the turbine deck between the two units. Each of the unit fire carts had two 110 volt portable blowers and flexible ductwork sections. The fire cart, which was common to the two units, had three of the blowers and flexible ductwork pieces. No violations or deviations were identified."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 12 3.17 "Fixed emergency lights will be provided in the control room independent of existing normal and emergency lighting. Portable hand held sealed beam lanterns will be provided for fire brigade use."
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203): 
"The NRC inspectors found that the licensee had installed emergency lighting as specified. Licensee representatives stated that this lighting was part of the installation made to meet the requirements of 10 CFR 50, Appendix R, Item III.J. The NRC inspectors had noted similar lighting installations in other plant areas. The fire carts described in paragraph 7 of this report were also found to contain portable hand held, sealed beam lanterns. No violations or deviations were identified."
3.18 "The reactor coolant pump oil collection system will be upgraded to provide collection capability at all potential leakage points."
 
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "Neither oil collection system was actually viewed, because both units were in mode 1. The NRC inspectors did review drawings and photographs of the installation made. It was noted that the design for both units had been reviewed by NRR. R. W. Reid letter of May 11, 1980, and R. A. Clark letter of November 5, 1980, are applicable to Units 1 and 2, respectively. The NRC inspectors had no questions in this area of the inspection. No violations or deviations were identified."
 
On July 1, 1982 (0CAN078202), Entergy submitted the results of its ANO-1 and ANO-2 Appendix R review. On August 15, 1984 (0CAN088404), Entergy requested an exemption from the requirements that a Reactor Coolant Pump (RCP) oil collection system be seismically qualified and capable of containing the oil from all RCP motors.
This exemption was approved in NRC letter dated October 26, 1988 (2CNA108802):
"The licensee stated in a letter dated August 15, 1984 that the reactor coolant pump lube oil systems are qualified to remain functional during and after an SSE. Therefore the following guidance of Generic Letter 86-10, "Implementation of Fire Protection Requirements," applies:
Where the RCP lube oil system is capable of withstanding the safe shutdown earthquake (SSE), the analysis should assume that only random oil leaks from the joints could occur during the lifetime of the plant. The oil collection system, therefore, should be designed to safely channel the quantity of oil from one pump to a vented closed container. Under this set of circumstances, the oil collection system would not have to be seismically designed.
The existing oil collection system is designed to safely channel the quantity of oil from one pump to a vented closed container, and so conforms with the above staff guidance. On this basis the staff concludes that the licensee's alternate design of the oil collection system provides an equivalent level of fire safety to that achieved by compliance with Section III.0.
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 13 The special circumstances of 10 CFR 50.12 apply in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. In this case the design of the reactor coolant pump lubricating systems and the oil collection systems meets certain criteria previously determined by the staff to be acceptable for assuring adequate fire safety. Thus the underlying purpose of the rule would be satisfied without requiring the oil collection system to be seismically qualified and capable of holding the oil contained in all of the reactor coolant pumps."
3.19 "Fire doors which separate redundant safe shutdown equipment or which separate safe shutdown equipment from large oil hazards will either be locked or provided with electrical supervision to alarm if opened."
 
This requirement was discussed in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):
  "The NRC inspectors found that many fire doors had installed electrical supervision, but, except in those cases wherein the fire door also happened to be a security door, the electrical supervision was not activated; i.e., the alarm did not work. It was also found that the only fire doors that were locked were those that were also security doors. Specifically, the fire doors that separated the diesel generator rooms for both Unit 1 and Unit 2 (doors 39 and 259, respectively) were neither locked nor did the installed alarms operate. Since the Table 3 SER requirements are incorporated into the licenses for both units, failure either to have doors 39 and 259 locked or to have operating supervisory alarms on them is an apparent violation."
In letter dated May 6, 1983 (0CNA058307), the NRC stated: 
"(Closed) Violation (50-313/8215-01; 50-368/8212-01). This violation was the result of the failure to have fire doors between spaces with redundant safe shutdown equipment either locked or electrically supervised. The NRC inspector found that these fire doors were now electrically supervised and would cause an alarm if left in an open position."
3.20 "Procedures are being developed or changed to incorporate controls over combustible materials and ignition sources, fire brigade staffing and training, fire fighting procedures, quality assurance provisions, and definition of fire protection duties and
 
responsibilities."
This requirement was documented in NRC Inspection Report 50-313/82-15 50-368/82-12 dated August 6, 1982 (0CNA088203):  "The NRC inspectors reviewed the licensee procedures listed below. These procedures addressed the requirements for administrative procedures related to fire protection and prevention.
1015.07, "Fire Brigade Organization and Responsibilities," Rev 2, November 30, 1981. 1023.20, "Fire Plan/Fire Brigade Training," Revision 2, February 23, 1982 1053.01, "Control of Combustibles," Revision 1, January 15, 1982 1053.02, "Control of Ignition Sources," Revision 1, February 21, 1981 1053.03, "Safety and Fire Prevention Inspection," Revision 1, May 13, 1981 1903.22, "Fire or Explosion," (from Emergency Plan), Revision 3, April 28, 1982 1903.41, "Duties of the Emergency Fire Team," Revision 3, December 8, 1981 There were no violations or deviations identified."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 14 On November 19, 1980, the NRC published the Fire Protection Rule, 10 CFR 50.48 and its guidance for implementation of that rule, Appendix R to 10 CFR 50. The effective date of the regulation was February 19, 1981. On July 1, 1982, Entergy submitted the results of its Appendix R compliance review and specific exemption requests (0CAN078202). Supplemental information and clarification of exemption requests were submitted on November 11, 1982 (0CAN118210). The following exemptions were approved in the staff's SER (0CNA038328) dated March 22, 1983. The NRC basis for acceptab ility of each exemption is also included below.
ANO-2  An exemption from 10 CFR 50, Appendix R Section III.G.2, Intake Structure, Below El. 354':  Exemption to requirement for automatic fire suppression. "This zone consists of the service water pump intake bays; therefore, the water level in the intake bays precludes the possible accumulation of transient combustible materials as anticipated in other plant areas. Because the likelihood of an exposure fire is low, these alternative features compensate for the required suppression system and provide a level of fire protection equivalent to that required by Section III.G of Appendix R. Therefore, the exemption is granted."  An exemption from 10 CFR 50, Appendix R Section III.G.2, Intake Structure, El. 354':  Exemption to requirement for automatic fire suppression system. "The in-situ combustible loading at this elevation consists primarily of cable insulation. This fire area is provided with portable fire extinguishers, manual hose stations and a smoke detection system. In the event of an exposure fire involving transient combustible materials, there will be a time lag between the ignition of the fire, detection and alarm, and the fire brigade response. The proposed configuration of cables and one-hour rated fire barrier will provide protection against the thermal flux of an exposure fire for a sufficient period of time to enable the fire brigade to respond and extinguish a fire prior to damage of both trains. The level of existing protection, in conjunction with the modifications in the Intake Structure, El. 354', provides a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R. Therefore, the exemption is granted."  An exemption from 10 CFR 50, Appendix R Section III.G.2, Intake Structure, El. 366':  Exemption to requirement for 20-foot separation and automatic fire suppression system. "Due to the low in-situ combustible loading in the intake structure, the partial width missile barriers in conjunction with the automatic sprinkler system will provide adequate assurance that one train of service water pumps will be maintained free of fire damage. The resulting fire protection for the Intake Structure, El. 366' provides a level of protection equivalent to the technical requirements of Section III.G of Appendix R.
Therefore, the exemption is granted."  An exemption from 10 CFR 50, Appendix R Section III.G.2, Yard Area Manholes 2MH01E, 2MH02E and 2MH03E:  Exemption to requirements for 20-foot separation, one-hour fire barrier, detection, and automatic fire suppression system. "Filling the manholes with sand or vermiculite will prevent a fire from occurring in the manholes and, therefore, an adequate level of fire protection will be provided equivalent to Section III.G of Appendix R. Therefore, the exemption is granted."
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 15 In letter dated August 15, 1984 (0CAN088404), Entergy provide the NRC with the following information:  "These manholes are filled with sand to prevent propagation of fire from damaging redundant trains of service water cabling. During the fourth refueling outage (2R4), which is scheduled to commence in mid-1985, the "swing" service water pumps will be provided with a separate cable leading to the redundant switchgear of the opposite division, i.e., for ANO-1, power for the "swing" pump will be directly available from the "green" 4160V bus independent of 1MH04 and 1MH06, and for ANO-2, power for the "swing" pump will be available from the "red" 4160V bus independent of 2MH01E, 2MH02E, and 2MH03E. With the completion of those modifications, this area will meet Appendix R, and the sand will no longer be needed."  An exemption from 10 CFR 50, Appendix R Section III.G.2, Pump Room, Fire Zone 2024JJ:  Exemption to requirement for fire barrier with three-hour rating. "The combustible loading in this area is substantially lower than needed for a fire of three-hour duration. An exposure fire would therefore be of limited severity and duration. The design of the watertight door provides a degree of inherent fire protection. It is our experience that typical watertight doors provide equivalent protection as three-hour rated fire doors. This combination of features provides reasonable assurance that a fire in this area will be detected and extinguished before the redundant emergency feedwater pump is damaged. Thus, the level of existing protection for this area provides a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R. Therefore, the exemption is granted."  An exemption from 10 CFR 50, Appendix R Section III.G.2, Containment Building, Fire Zones 2032K and 2033K:  Exemption to requirement for 20-foot separation with no intervening combustibles or fire hazards. "Due to the restricted access to this area, the probability of an exposure fire from the accumulation of transient combustibles and which could potentially bypass the cable tray fire stops is unlikely. Therefore, the fire stops installed in the intervening cable trays will provide reasonable assurance that one train will be maintained free of fire damage.
Because the amount of in-situ combustibles is low in this area and early warning detection provided, a fire in this area which could damage both RHR letdown power cable conduits is unlikely. Further, it is unlikely that fire damage would occur in the specific manner required. This combination of conditions provides reasonable assurance that a spurious operation of both valves is not likely to occur. The RCS pressure instrumentation is separated by 17 feet. A concrete missile barrier provides a partial radiant energy shield between the redundant instrumentation channels. Because of the unlikely nature of significant fire occurrence inside containment and the mitigating effects of the intervening missile barrier, there is reasonable assurance that one train will be maintained free of fire damage. The level of protection inside containment provides a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R. Therefore, the exemption is granted."  An exemption from 10 CFR 50, Appendix R Section III.G.2, Tank Rooms, Pump Rooms and Corridors, Fire Zone 2040JJ:  Exemption to requirement for automatic fire suppression system. "Because of the low in-situ combustible loading in this area, intended installation of one-hour barriers and intervening walls, and the detection system, in the event of an exposure fire, redundant cables will be protected for a sufficient period to enable the fire Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 16 brigade to respond and extinguish a fire prior to damage of both trains. The protection in the charging pump area in conjunction with the proposed one-hour barriers provides a level of fire protection equivalent to Section III.G. Therefore, the exemption is granted."
In letter dated August 15, 1984 (0CAN088404), Entergy provide the NRC with the following information:  "Two of three charging pump cables have been enclosed in a one-hour barrier and one train of the charging pump corridor area cable trays has been protected with a one-hour barrier for a sufficient length to permit 20 feet of horizontal separation between unprotected redundant cables."
An exemption from 10 CFR 50, Appendix R Section III.G.2, Lower South Piping Penetration Area, Fire Zone 2055JJ:  Exemption to requirement for fire barrier with three-hour rating. "In this area, an open doorway would violate the three-hour barrier and would provide direct communication between two fire areas containing redundant trains of safe shutdown related cable. The three-hour rated fire door with special latch for overpressure release which will be installed, is adequate for the fire hazard in these areas. The protection provided for the piping penetration area provides a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R. Therefore, the exemption is granted."
The new door (Door 210) installation was completed in December 1983. In letter dated August 30, 1985 (0CAN088508), Entergy stated that an analysis supporting removal of the latching mechanism had been performed from a fire perspective and that further analysis, relating to High Energy Line Breaks (HELB), would be performed in the future:  "An analysis has been performed to evaluate the suitability of a modified latch (i.e. no latch) on door 210 which separates Zone 2055-JJ and Zone 2040-DD. Based upon the existing fire protection features and calculated fire duration in the area, it was determined that the modified latch on the door would not adversely effect the margin of safety to the public, nor would it increase the likelihood of a fire spreading through the area. An exemption for this door as a three-hour barrier was previously granted in 0CNA038328. A reevaluation of the necessity for a modified latch on the door to satisfy HELB criteria is presently being performed. Based upon these results the previously referenced exemption may no longer be necessary."
An exemption from 10 CFR 50, Appendix R Section III.G.2, Corridor, El. 372', Fire Zone 2109U:  Exemption to requirement for 20-foot separation or one-hour fire barrier. Note that this exemption is no longer required for ANO-2. "The existing protection for this area in conjunction with the proposed one-hour fire rated barrier and flame resistant coating of intervening cable trays provides a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R.
Therefore, the exemption is granted."
In letter dated July 27, 1983 (2CAN078313), Entergy provide the NRC with the following information:  "As addressed in our letter (0CAN038330) to you dated March 29, 1983, fire zones 34Y, 40Y, and 73W of Unit 1 and 2109U of Unit 2 were identified as requiring one hour fire barrier installation which was to be completed by April 1, 1983."  "During the period April 1-22, 1983, the subject fire zones were all deemed adequately protected. As we expedited our construction efforts and completed the required one hour barrier installation by April 22, 1983, we feel a good faith and best effort approach was presented by our completion of this project in advance of the requested exemption date of May 1, 1983."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 17 An exemption from 10 CFR 50, Appendix R Section III.G.3, Upper North Piping Penetration Area, Fire Zone 2081HH:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."  An exemption from 10 CFR 50, Appendix R Se ction III.G.3, Motor Control Center, Fire Zone 2096M:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."  An exemption from 10 CFR 50, Appendix R Section III.G.3, Electrical Equipment Room, El. 368', Fire Zone 2091BB:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."  An exemption from 10 CFR 50, Appendix R Section III.G.3, Corridor Area, Fire Zone 2107N:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 18 An exemption from 10 CFR 50, Appendix R Section III.G.3, Pipeway and Equipment Access Way, Fire Zone 2223KK:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."  An exemption from 10 CFR 50, Appendix R Section III.G.3, Pump Room, Fire Zone 2106R:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."  An exemption from 10 CFR 50, Appendix R Section III.G.3, Core Protection Calculator Panel, Fire Zone 2150C:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."  An exemption from 10 CFR 50, Appendix R Section III.G.3, Health Physics Area, Fire Zone 2136I:  Exemption to requirements for fixed fire suppression system. "All of the above fire zones represent a similar configuration, i.e., combustible loading is light, there is alternate shutdown capability, and manual fire suppression equipment is available. The low combustible loading in these areas ensures that safety-related equipment in adjacent areas will not be threatened. The installation of a fixed fire suppression system in all these areas and detection for Fire Zones 2081HH and 2136I will not significantly increase the level of fire protection in these areas. The existing fire protection in conjunction with alternate shutdown capability in the above areas provides a level of fire protection equivalent to the technical requirements of Section III.G.3 of Appendix R. Therefore, the exemptions are granted."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 19 In letter dated May 13, 1983 (0CNA058316), the NRC accepted the alternate shutdown capability and methods for ANO-1 and ANO-2 in the event of a fire in the control room, cable spreading room, or other critical area. From the aforementioned SER:
"The goals of reactivity control, inventory control, decay heat removal, and pressure control are met. The goal of process monitoring is only partially met in that no means of monitoring source range radioactive flux is provided. The goal of adequate support systems has been met. Based on our review, we conclude that the design proposed for Arkansas Nuclear Plant, Units 1 and 2 meets the requirements of Appendix R to 10 CFR Part 50 Items III.G.3 and III.L with respect to safe shutdown in the event of a fire, except where an exemption to the 72 hour shutdown requirement has been granted to ANO-1 and with the following exception:
: 1. A source range flux monitoring capability electrically independent of the control room should be provided at the safety parameter display system for both units."
This requirement was documented in NRC Inspection Report 50-313/87-14 50-368/87-14 dated September 30, 1987 (0CNA098716):  "Areas examined during the inspection included implementation of and compliance to the safe shutdown requirements of 10 CFR 50, Appendix R."  "The following process monitoring instrumentation is available in the control room and on the SPDS "Alternate Shutdown" display:  Source Range Flux."
During the period following the initial Appendix R submittal date, requirements continued to evolve, resulting in further Appendix R reanalysis. Based on this reanalysis, additional exemptions were requested in letters dated August 15, 1984 (0CAN088404), August 30, 1985 (0CAN088508), and October 29, 1987 (0CAN108710). The following exemptions were approved in the NRC's SER (2CNA108802) dated October 26, 1988. The NRC basis for acceptability of each exemption is also included below. An exemption from 10 CFR 50, Appendix R Section III.G.2.b due to a lack of 20 feet of separation free of intervening combustible materials between the redundant diesel generator exhaust fan outlets (Fire Area B, Fire Zone 2114-I). "The special circumstances of 10 CFR 50.12 apply in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. In this case, the light combustible loading, the non-credible path necessary for the fire to spread to the redundant safe shutdown equipment, and the installation of 3-hour fire rated doors committed to by the licensee, all provide assurance that the redundant train will be adequately protected. Thus the underlying purpose of the rule would be satisfied without requiring spatial separation of the exhaust fan outlets."
This requirement was documented in NRC letter dated October 26, 1988 (2CNA108802):  "If a fire were to occur in or near one of the exhaust fans, it would be expected to develop slowly with initial low heat release and slow temperature rise. The lack of a roof over Fire Zone 2114-I would preclude any accumulation of hot gases over this equipment. Further, in order for the fire to seriously affect the redundant equipment, it would have to spread over and down into the room below, which is not considered credible. Additionally the licensee has completed the installation of 3-hour rated fire doors between redundant trains of equipment. Therefore, the possibility of a single fire in one of these fire zones damaging redundant equipment becomes extremely unlikely, despite the horizontal separation distance of less than 20 feet between redundant trains. On this basis, the staff concludes that the licensee's alternative fire protection configuration provides an equivalent level of fire safety to that achieved by compliance with Section III.G.2.b."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 20 An exemption from 10 CFR 50, Appendix R Section III.G.3 due to a lack of a fixed fire suppression system in the control room and printer room (Fire Area G, Zone 2199-G). "The special circumstances of 10 CFR 50.12 apply in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. The rule is meant to ensure that a fire in the control room or printer room would not prevent a safe plant shutdown. The existing fire protection features provide reasonable assurance that the ability to achieve safe shutdown of the plant is maintained. Thus the underlying purpose of the rule would be satisfied without requiring a fixed fire suppression system."  An exemption from 10 CFR 50, Appendix R Section III.G.2.b due to the lack of an automatic fire suppression system in the upper and lower south piping penetration rooms (Fire Area EE, Zone 2084-DD and 2055-JJ). "The special circumstances of 10 CFR 50.12 apply in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. In this case the moderate fire loading, the spatial separation free of intervening combustibles between redundant safe shutdown related valves, the capability of the fire brigade to respond quickly once a fire is detected by the automatic fire detection system all provide assurance that redundant safe shutdown components will be adequately protected. Thus the underlying purpose of the rule would be satisfied without requiring automatic suppression systems in the upper and lower south piping penetration rooms."  An exemption from 10 CFR 50, Appendix R Section III.G.2.a due to a lack of a complete three-hour fire-rated barrier between redundant level transmitters for the safety grade Quality Condensate Storage Tank (QCST) (Yard Area). "The special circumstances of 10 CFR 50.12 apply in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. In this case the absence of significant in-situ fire hazards, and the physical location and arrangement of the equipment provide assurance that the redundant level indication equipment would be adequately protected until the fire was brought under control by the fire brigade. Thus the underlying purpose of the rule would be satisfied without requiring a 3-hour fire-rated barrier between the redundant QCST level transmitters."  An exemption from 10 CFR 50, Appendix R Section III.O due to a lack of a reactor coolant pump oil collection system that is designed to withstand a Safe Shutdown Earthquake (SSE) and sized to hold the oil from all RCPs.
On August 15, 1984 (0CAN088404), Entergy requested an exemption from the requirements that an RCP oil collection system be seismically qualified and capable of containing the oil from all RCP motors.
This exemption was approved in NRC letter dated October 26, 1988 (2CNA108802):
"The licensee stated in a letter dated August 15, 1984 that the reactor coolant pump lube oil systems are qualified to remain functional during and after an SSE. Therefore the following guidance of Generic Letter 86-10, "Implementation of Fire Protection Requirements," applies:
Where the RCP lube oil system is capable of withstanding the safe shutdown earthquake (SSE), the analysis should assume that only random oil leaks from the joints could occur during the lifetime of the plant. The oil collection system, therefore, should be designed to safely channel the quantity of oil from one pump to a vented closed container. Under this set of circumstances, the oil collection system would not have to be seismically designed.
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 21 The existing oil collection system is designed to safely channel the quantity of oil from one pump to a vented closed container, and so conforms with the above staff guidance.
On this basis the staff concludes that the licensee's alternate design of the oil collection system provides an equivalent level of fire safety to that achieved by compliance with Section III.0. The special circumstances of 10 CFR 50.12 apply in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule. In this case the design of the reactor coolant pump lubricating systems and the oil collection systems meets certain criteria previously determined by the staff to be acceptable for assuring adequate fire safety. Thus the underlying purpose of the rule would be satisfied without requiring the oil collection system to be seismically qualified and capable of holding the oil contained in all of the reactor coolant pumps."  An exemption from 10 CFR 50, Appendix R Section III.G.J due to a lack of eight-hour battery powered emergency lighting units in the access paths to the intake structure and diesel fuel storage vaults. "The licensee requested an exemption from Section III.J due to a lack of 8-hour battery powered emergency lighting units in the access paths to the intake structure, and diesel fuel storage vaults which are areas required to be manned for safe shutdown. Because these locations are essentially identical to locations involved in an exemption from Section III.J granted for Unit 1, there is no need for an exemption for Unit 2."
Note that the ANO-1 exemption was provided in NRC letter dated October 26, 1988 (1CNA108806).
 
On December 23, 1996 (2CAN129612), Entergy requested exemption from the requirements of Section III.O of 10 CFR 50 Appendix R for ANO-2, which requires that a RCP oil collection system be provided to hold the oil from all RCPs, including fill lines. By letter dated June 14, 1997 (2CNA069701), the NRC approved exemption of the RCP oil fill lines from the oil collection system requirements of A ppendix R. From the aforementioned SER:
"On the basis of the enclosed safety evaluation, the U. S. Nuclear Regulatory Commission (NRC) staff concluded that the design of the oil filling system and the level of protection provided by the licensee through the use of certain compensatory measures during oil fill operations provides reasonable assurance that a lube oil fire will not occur. The staff also concluded that a worst-case postulated fire, due to not having a lube oil collection system for the RCP lube oil fill lines, would be of limited magnitude and extent. In addition, such a fire would not cause significant damage in the containment building and would not prevent the operators from achieving and maintaining safe shutdown conditions. The staff concluded, therefore, that the lack of an oil collection system for the RCP lube oil fill lines is an acceptable exemption from the technical requirements of 10 CFR Part 50, Appendix R, Section III.O.
 
Therefore, contingent on the use of the compensatory measures that are itemized in the licensee's December 23, 1996, exemption request, the NRC staff has concluded that the licensee's proposed use of the remote oil addition system will not present an undue risk to public health and safety and is consistent with the common defense and security. The NRC staff has determined that there are special circumstances present, as specified in 10 CFR 50.12(a)(2), in that application of 10 CFR Part 50, Appendix R, Section 111.0 is not necessary in order to achieve the underlying purpose of this regulation."
 
Arkansas Nuclear One - Unit 2 2.0 Overview of Existing Fire Protection Program Enclosure 1 to 2CAN121202 Page 22 By letter dated October 8, 1997 (2CAN109703), as supplemented by letter dated February 25, 1999 (2CAN029905), Entergy requested an exemption from 10 CFR 50, Appendix R, Section III.G.2, for lack of an automatic suppression system and detection system for the area below Elevation 354' of the ANO-2 intake structure. The NRC approved this exemption in letter dated October 1, 1999 (2CNA109902). From the aforementioned SER:
"Power and control cables for the sluice gates are also located in the SW intake bays. Sluice gate valves 2CV1470-1, 2CV1472-5, and 2CV1474-2 are normally open, which corresponds to the safe shutdown position. The redundant control cables are separated horizontally by approximately 8 feet. As stated previously, the time critical function of the SW system is to provide cooling to the diesel generators. The licensee stated that if a fire were to cause the sluice gates to spuriously close, adequate time would be available before the SW was required to manually realign any affected component. The in-situ combustibles in "A" SW intake bay and the administratively allowed quantity of transient combustibles (5 pounds) do not pose a credible fire threat to the SW pump cables. In the staff's view, a fire involving transient combustibles in excess of the administratively allowed quantity is the only type of fire that could damage redundant SW pump power cables. The licensee has addressed this threat by protecting both the red train power supply cable for SW pump 2P4A and the green train power supply cable for swing SW pump 2P4B with 1-hour fire-rated barriers, by embedding the red train power supply cable for SW swing pump 2P4B in concrete, and by administratively requiring the presence of craft personnel or a fire watch, if the administrative transient combustible limit is exceeded. A fire involving transient combustibles could be extinguished by the craft personnel or the fire watch during its incipient stage. In the event the fire grows beyond the incipient stage before it is extinguished, the craft personnel or the fire watch could summon the plant fire brigade. In addition, the smoke and hot gases would be directed upwards into the higher elevations of the intake structure, which are equipped with an automatic fire detection system. Therefore, in the event that a fire in the intake bay is not discovered by the craft personnel or the fire watch, it would be detected by the automatic fire detection system and the plant fire brigade would be dispatched. If the fire exposes the redundant conduits, the 1-hour fire-rated barriers and the concrete embedding, with an equivalent 1-hour fire rating, would provide fire resistive protection, with margin, for the expected fire hazards and, therefore, provide reasonable assurance that the power cables would not be damaged before the fire either burns itself out or is extinguished by the craft personnel, the fire watch, or the fire brigade. On this basis, the staff concludes that the existing fire protection design features, coupled with the administrative controls, provide reasonable assurance that a fire in the "A" SW intake bay would not damage the redundant SW pump power cables and, therefore, would not adversely affect the ability to achieve and maintain post-fire safe shutdown. The staff also concludes that the installation of fire detectors and an automatic fire suppression system in the area below the 354-foot elevation of the ANO-2 intake structure would not result in a significant increase in the level of fire safety for the redundant SW pumps. Additional details concerning the exemption are provided in the staff's Safety Evaluation dated October 1, 1999. For the forgoing reasons, the NRC staff has determined that there is a low probability of occurrence for a fire event in the ANO-2 intake structure below the 354-foot elevation. This low probability of occurrence combined with the lack of combustible material, administrative controls, and the fire protection features provided, as stated in the licensee's submittals, is sufficient to reasonably ensure adequate protection for redundant equipment in the SW system, such that there is reasonable assurance that at least one means of achieving and maintaining safe shutdown conditions will remain available during and after any postulated fire. Therefore, the addition of fire detectors and an automatic fire suppression system is not necessary to achieve the underlying purpose of Appendix R, Section III.G.2.c."
 
Arkansas Nuclear One - Unit 2 3.0 Transition Process Enclosure 1 to 2CAN121202 Page 23 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 SER; 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, ANO-2 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 applied to the requirements for which deterministic compliance could not be shown.
 
Arkansas Nuclear One - Unit 2 3.0 Transition Process Enclosure 1 to 2CAN121202 Page 24 Figure 3-1 NFPA 805 Process [NEI 04-02 Figure 3-1 based on Figure 2-2 of NFPA 805]
2 
 
2 Note: 10 CFR 50.48(c) does not incorporate by reference Life Safety and Plant Damage/Business Interruption goals, objectives and criteria. See 10 CFR 50.48(c) for specific exceptions to the incorporation by reference of NFPA 805.
Establish fundamental fire protection elements    (Chapter 3) Identify fire hazards Identify performance criteria to be examined (Chapter 1) Identify structures, systems, or components (SSCs) in each fire area to which the performance criteria applies Deterministic Approach Maintain compliance with existing plant license basis (10 CFR 50 App. R, Approved Exemptions, Engineering Evaluations) Deterministic Basis Verify deterministic requirements are met Existing Engineering Equivalency Evaluations NFPA 805 Section 2.2(f) Performance-Based Approach Evaluate ability to satisfy performance requirements (Chapter 4) Performance Basis Define fire scenarios and fi re design basis for each fire area being considered.
Evaluate using, e.g.,
* Fire modeling to quantify the fire risk and margin of safety
* PSA to examine impact on overall plant risk NFPA 805 Section 2.2(g) NFPA 805 Section 2.2(e) NFPA 805 Section 2.2(d) NFPA 805 Section 2.2(c) NFPA 805 Section 2.2(b) NFPA 805 Section 2.2(a) Nuclear safety, Life safety, Property damage/business interruption, Radiation Release Risk-Informed Change Evaluation Evaluate risk impact of changes to the approved design changes NFPA 805 Section 2.2(h)
Is change acceptable? Documentation and configuration control NFPA 805 Section 2.2(j) NFPA 805 Section 2.2(i)
Establish monitoring program Design Basis Documents, Fire Hazards Analysis, Nuclear Safety Capability assessment, Support engineering calculations, Probabilistic safety analysis, Risk-informed change evaluations No Yes Feedback Examples Evaluate compliance to performance criteria Arkansas Nuclear One - Unit 2 3.0 Transition Process Enclosure 1 to 2CAN121202 Page 25 3.3 NEI 04 NFPA 805 Transition Process
 
NFPA 805 contains technical processes and requirements for a RI-PB fire protection program. NEI 04-02 was developed to provide guidance on the overall process (programmatic, technical, and licensing) for transitioning from a traditional fire protection licensing basis to a new RI-PB method based upon NFPA 805, as shown in Figure 3-2.
 
Section 4.0 of NEI 04-02 describes the detailed process for assessing a fire protection program for compliance with NFPA 805, as shown in Figure 3-2.
Figure 3-2 Transition Process (Simplified) [based on NEI 04-02 Figure 4-1]
 
FP Fundamentals Review and Confirmation Nuclear Safety Review and Confirmation Transition Report Sect. 4.1 Transition Report Sect. 4.2 Transition Report Sect. 4.4 Transition Report Sect. 4.3 Transition Report Sect. 4.5 Transition Report Sect. 4.6 Transition Report Sect. 4.7, 5 Transition Report Sect. 4.8, 5 FP Fundamentals Assessment Identify outliers / VFDRs Identify outliers / VFDRs Radioactive Release Assessment Perform Engineering Analyses Non-power Operational Mode Assessment Nuclear Safety Analyses Use PB Approach if needed (Fire Modeling or Fire Risk Evaluations) Submit for NRC approval if needed Confirm / Establish Adequate Documentation / Quality and Configuration Controls Verify / Establish Monitoring Program Regulatory Submittal and Approval Arkansas Nuclear One - Unit 2 3.0 Transition Process Enclosure 1 to 2CAN121202 Page 26 3.4 NFPA 805 Asked Questions (FAQs)
 
The NRC has worked with NEI and two Pilot Plants (Oconee Nuclear Station and Harris Nuclear Plant) to define the licensing process for transitioning to a new licensing basis under 10 CFR 50.48(c) and NFPA 805. Both the NRC and the industry recognized the need for additional clarifications to the guidance provided in RG 1.205, NEI 04-02, and the requirements of NFPA 805. The NFPA 805 FAQ process was jointly developed by NEI and NRC to facilitate timely clarifications of NRC positions. This pr ocess 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 utilized by ANO-2 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.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 27 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 meet 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 ANO-2 Fire Protection Program to the requirements of NFPA 805 Chapter 3 was performed and documented in Engineering Change (EC)-6960. The EC used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 (see Figure 4-1).
 
Each section and subsection of NFPA 805 Chapter 3 was reviewed against the current fire protection program. Upon completion of the activities associated with the review, the following compliance statement(s) was used:  Complies - For those sections/subsections determined to meet the specific requirements of NFPA 805. Complies with clarification - For those sections/subsections determined to meet the requirements of NFPA 805 with clarification. Complies by previous NRC approval - For thos e sections/subsections where the specific NFPA 805 Chapter 3 requirements are not met but previous NRC approval of the configuration exists. Complies with use of Existing Engineering Equivalency Evaluations (EEEEs) - For those sections/subsections determined to be equivalent to the NFPA 805 Chapter 3 requirements as documented by engineering analysis. Submit for NRC Approval - For those sections/subsections for which approval is sought in this LAR submittal in accordance with 10 CFR 50.48(c)(2)(vii). A summary of the bases of acceptability is provided (see Attachment L for details).
In some cases multiple compliance statements have been assigned to a specific NFPA 805 Chapter 3 section/subsection. Where this is the case, each compliance/compliance basis statement clearly references the corresponding requirement of NFPA 805 Chapter 3.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 28 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. Existing Fundamental Fire Protection Program and Design Element Meets the NFPA 805 Chapter 3 Requirement NFPA 805 Chapter 3 Deviation Has Previous Approval Existing Engineering Equivalency Evaluation? Does Not Meet Chapter 3 nor is there Previous Approval Enter 'Complies' in Compliance Statement Field In the Compliance Basis Field provide:
* No additional clarification In Reference Document Field provide:
* Document References that demonstrate compliance Document any Open Items found during Review Enter 'Complies with Clarification' in Compliance Statement Field In the Compliance Basis Field provide:
* Provide details on clarification In Reference Document Field provide:
* Document References that demonstrate compliance Document any Open Items found during Review Enter 'Complies via Previous Approval' in Compliance Statement Field In the Compliance Basis Field provide verbatim excerpt from:
* Approval document In Reference Document Field provide:
* Licensing Document References Document any Open Items found during Review Enter 'Complies with use of existing Engineering Equivalency Evaluation' in Compliance Statement Field In the Compliance Basis Field provide:
* Summary of bases for Engineering Evaluation In Reference Document Field provide:
* Document References that demonstrate compliance Document any Open Items found during Review Note 1 Note 2 Yes Yes Yes No No No Choose One Note 1:  If the excerpt that provided the formal approval does not contain sufficient details of the previous approval, provide an excerpt(s) of licensing submittals regarding the issue for which previous approval is being claimed. Place the excerpt of the submittals before the excerpt of the formal approvals in the Compliance Basis field, if necessary. Note 2:  Existing Engineering Equivalency Evaluations (previously known as Generic Letter 86-10 evaluations, exemptions, and deviations) were performed for fire protection design variances such as fire protection system designs and fire barrier component deviations from the specific fire protection deterministic requirements. Section 2.2.7 of NFPA 805 allows EEEE that clearly demonstrates an equivalent level of fire protection compared to the deterministic requirements to be transitioned. Compliance Indeterminate Further Action Required Is a License Amendment Required? Enter 'Further Action Required' in Compliance Statement Field In the Compliance Basis Field provide:
* List of Actions to be taken In Reference Document Field provide:
* Corrective Actions, as appropriate
* Document References Document any Open Items found during Review Enter 'License Amendment Required' in Compliance Statement Field In the Compliance Basis Field provide:
* Summary of bases for license amendment In Reference Document Field provide:
* Corrective actions, as appropriate
* Document References Document any Open Items found during Review Note 3 Note 3:  Further Action Required indicates an interim position used during the process of completing the B-1 Table.  'Further Action Required' entries should be resolved prior to submitting the LAR. If they are not, then a confirmatory activity should be added to the LAR submittal.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 29 4.1.2 Results of the Evaluation Process 4.1.2.1 NFPA 805 Chapter 3 Requirements Met or Previously Approved by the NRC Attachment A contains the NEI 04-02 Table B-1, Transition of Fundamental Fire Protection (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 for ANO-2 either:  Complies directly with the requirements of NFPA 805 Chapter 3,  Complies with clarification with the requirements of NFPA 805 Chapter 3,  Complies through the use of existing engineering equivalency evaluations which are valid and of appropriate quality, or  Complies with a previously NRC approved alternative to NFPA 805 Chapter 3 and therefore the specific requirement of NFPA 805 Chapter 3 is supplanted.
 
4.1.2.2 NFPA 805 Chapter 3 Requirements Requiring Clarification of Prior NRC Approval NFPA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein."  In some cases prior NRC approval of an NFPA 805 Chapter 3 program attribute may be unclear. However, there are no clarifications necessary for ANO-2.
 
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:
3.3.3 - Approval is requested for epoxy floor coverings at ANO that may not meet the NFPA 805 requirements for "interior finish." 3.3.5.1 - Approval is requested for wiring above suspended ceilings that may not comply with code requirements. 3.3.5.2 - Approval is requested for use of schedule 40 PVC for underground and embedded applications. 3.3.12 (1) - Approval is requested for acceptability of oil misting from the reactor coolant pumps/motors. 3.5.3 - Approval is requested for continued use of existing electric fire pump motor and electric fire pump controller that were not UL Listed/Approved for fire pump service at the time of installation. Additionally, approval is requested for not meeting the requirements of NFPA 20 Sections 626a, 626d.e2, and 626d.e5 for the Cummins Diesel Engine controller, since vendor documents do not identify a certification for the batteries and do not identify the discharge rate of the lead acid batteries. 3.5.16 - Approval is requested for use of the fire protection water supply system to supply cooling loads on either unit during outages.
The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements is provided in Attachment L. ANO-2 requests NRC approval of these
 
performance-based methods.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 30 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.
 
A list of plant structures was derived from a review of plant layout drawings and supplemented by plant walk downs in order to provide a complete listing of the structures in the owner controlled area. Each structure was reviewed to determine if it was required to meet the NFPA nuclear safety goal, meet the NFPA radioactive release goal, or be evaluated for other NFPA considerations. The structures identified as meeting the aforementioned guidance for the power block are listed in Attachment I.
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 NSCA Methodology review consists of four processes:  Establishing compliance with NFPA 805 Section 2.4.2  Establishing the Safe and Stable Conditions for the Plant  Establishing Recovery Actions  Evaluating Multiple Spurious Operations The methodology for demonstrating reasonable assurance that a fire during Non-Power Operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3.
 
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 Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 31 (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 existing post-fire Safe Shutdown Analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 1, Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. The methodology is depicted in Figure 4-2 and consisted of the following activities:  Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01, Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to
 
the station. The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section: o Aligns o Aligns with intent o Not in Alignment o Not in Alignment, but Prior NRC Approval o Not in Alignment, but no adverse consequences The comparison of the ANO-2 existing post-fire SSA methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in CALC-ANO2-FP-09-00032.
 
In addition, a review of NEI 00-01, Revision 2, (ML091770265) Chapter 3, was conducted to identify the substantive changes from NEI 00-01, Revision 1 that are applicable to an NFPA 805 fire protection program. This review was performed and documented in EC-40607, "NEI 00-01, Section 3, Rev. 1 to Rev. 2 Gap Analysis for NFPA 805 LAR."
 
Results from Evaluation Process
 
The method used to perform the existing post-fire SSA with respect to selection of systems and equipment, selection of cables, and identification of the location of equipment and cables, either meets the NRC endorsed guidance from NEI 00-01, Revision 1, Chapter 3 (as supplemented by the gap analysis) directly or met the intent of the endorsed guidance with adequate justification as documented in Attachment B.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 32 Figure 4-2 Summary of Nuclear Safety Methodology Review Process (FAQ 07-0039)
 
SSD - Safe Shutdown
 
Comparison to NEI 00-01 Revision 2
 
An additional review was performed of NEI 00-01, Revision 2, Chapter 3, for specific substantive changes in the guidance from NEI 00-01, Revision 1 that are applicable to an NFPA 805 transition. The results of this review are summarized below:  Post-fire manual operation of rising stem valves in the fire area of concern (NEI 00-01, Section 3.2.1.2)
ANO does not credit Recovery Actions for valves located in the fire affected area. Analysis of open circuits on high voltage (e.g., 4.16 kV) ammeter current transformers (NEI 00-01, Section 3.5.2.1)
The potential for an open circuit on a current transformer (CT) circuit resulting in secondary damage and possibly resulting in the occurrence of an additional fire has been evaluated and is documented in EC15217, "C urrent Transformer (CT) Open Circuit Concerns." In Strict Alignment with NEI 00-01 Guidance? Meets Intent of          NEI 00-01 Guidance?
Can lack of          alignment potentially result in adverse consequences? Has NRC approval been obtained for method? Assemble Documentation Determine and Document Applicability of NEI 00-01 Sections For Applicable NEI 00-01 Sections, perform Comparison of SSD Method vs. NEI 00-01 Document and Address Open Item (Consider entry in Corrective Action Program)
Document Yes Yes No Yes No Yes No No Step 3 Step 2 Step 1 Step 4 Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 33 Analysis of control power for switchgear with respect to breaker coordination (NEI 00-01, Section 3.5.2.4)
The Safe Shutdown Analysis (SSA) does not discuss breaker coordination in detail. Breaker coordination at ANO is addressed in CALC-85-E-0087-24, "Safe Shutdown Cable Analysis," and described in Upper Level Document ULD-0-TOP-12, "ANO Unit 1 and 2 Electrical Protection/Coordination."  Details are addressed in a series of calculations.
4.2.1.2 Safe and Stable Conditions for the Plant
 
Overview of Process
 
The nuclear safety goals, objectives and performance criteria of NFPA 805 allow more flexibility than the previous deterministic programs based on 10 CFR 50 Appendix R and NUREG-0800, Section 9.5-1 (and NEI 00-01, Chapter 3) since NFPA 805 only requires the licensee to maintain the fuel in a safe and stable condition rather than achieve and maintain cold shutdown.
 
NFPA 805, Section 1.6.56, defines Safe and Stable Conditions as follows "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 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." 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 standby plant operating state for the event.
 
Results  Based on the NFPA 805 Nuclear Safety Capability Assessment Methodology (Table B-2), the NFPA 805 licensing basis for ANO-2 is to shutdown the reactor and maintain the reactor in a hot standby condition (defined as Mode 3, Keff < 0.99, RCS temperature  300 &deg;F) following any fire occurring with the reactor operating at power.
This NSCA Methodology evaluation compares the NRC endorsed process in Chapter 3 of NE I 00-01, Revision 1, in accordance with NEI 04-02, Revision 2, requirements.
 
Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses. At-Power analysis, Mode 1 through achieving and maintaining Mode 3. This analysis is discussed in Section 4.2.4. Non-Power analysis, which includes Mode 4 and below. This analysis is discussed in Section 4.3.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 34 Recovery actions (including defense-in-depth recovery actions) are subjected to a feasibility review. This review is conducted in accordance with the NRC endorsed guidance in NEI 04-02, Revision 2.
 
The functions addressed in the NFPA 805 Nuclear Safety Capability Assessment Methodology (Table B-2) are important to post-fire safe shutdown and 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 o Electrical Systems o Emergency Diesel Generator Fuel Oil o Cooling Systems The 'At Power' safe shutdown analysis postulates a single fire occurring at 100% power and provides a listing of conflicts that may impact the assured success path to meet a particular nuclear safety performance goal. The 'At Power' safe and stable strategy includes entry into hot standby (Mode 3) and stops prior to the point of manually initiating a cooldown. Safe and stable conditions in Mode 3 may continue long term as described below.
 
Reactivity Control Adequate shutdown margin (SDM) post-trip is provided by insertion of the Control Element Assemblies (CEAs). The CEAs require no motive force or electrical power to fulfill their safety function to insert into the core. No addition of boric acid solution is required to support post-trip hot standby conditions and a K eff of < 0.99. However, boric acid can be added to the Reactor Coolant System (RCS) to increase the SDM, as needed, either from a Boric Acid Makeup Tank (BAMT) or the Refueling Water Tank (RWT).
Adequate SDM in support of a plant cooldown to cold shutdown (Mode 5) is assured using the BAMTs and RWT aligned to the Charging Pump suction. If necessary, RCS pressure can be reduced to permit borated water makeup from the RWT via one or more High Pressure Safety Injection (HPSI) pumps. To ensure the reactor remains sub-critical with a K eff < 0.99, a proceduralized specified minimum amount of borated water is injected into the RCS prior to and/or during the plant cooldown.
 
Pressure Control Systems
 
RCS pressure is maintained by controlling the rate of charging/makeup to the RCS and/or use of Pressurizer heaters and spray. If Reactor Coolant Pumps (RCPs) are secured, the Charging Pumps provide an auxiliary spray path to support Pressurizer pressure reductions. The two redundant banks of proportional heaters and the auxiliary spray path are safety-related, vital-powered components. Although utilization of the Pressurizer heaters and/or auxiliary spray Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 35 reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions can be made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure. Pressurizer and reactor head vents may also be available, if necessary, to reduce Pressurizer (RCS) pressure. Pressure increases can be made by initiating charging/makeup to maintain Pressurizer level/pressure. Manual control of the related pumps is acceptable.
 
Inventory Control Systems
 
Reactor Coolant System Inventory Inventory makeup to the RCS is only required to account for expected RCS leakage, RCS shrinkage, and RCP seal controlled bleed-off. ANO-2 has design features and procedures to ensure that an adequate source of borated inventory is maintained for RCS inventory control, with regard to long term Mode 3 operations and to support cooldown to Mode 5, utilizing the Chemical and Volume Control System (CVCS). The Charging Pumps, taking suction from either the RWT or BAMT, can be used to control RCS level. If necessary, RCS pressure can be reduced to permit makeup via one or more HPSI pumps. The content of the BAMT(s) is normally sufficient, without use of RWT inventory, to support a cooldown of the RCS to Mode 5, if so desired. The significant volume of borated water available minimizes any concerns with regard to maintaining the fuel in the reactor vessel in a safe and stable condition during or
 
following a fire event.
 
Spent Fuel Pool (SFP) Inventory
 
Makeup to the SFP can be supplied by the BAMTs, the RWT, demineralized water, the Service Water (SW) system from Lake Dardanelle, or from excess RCS water from holdup tanks. Some of these sources are aligned from different plant locations such that a fire event will not prevent makeup to the SFP. The various sources available minimize any concerns for maintaining the fuel in the SFP in a safe configuration during or following a fire event.
Decay Heat Removal Systems Decay heat removal is accomplished using forced or natural circulation via the Steam Generators (SGs) in Modes 3 and 4. Upon entry into Mode 3, Emergency Feedwater (EFW) will automatically start or can be manually placed in service (from the Control Room or locally) and will provide secondary makeup water to the SGs (only one SG is required to remove decay heat from the RCS), with pressure control provided by the Steam Dump and Bypass Control System (SDBCS) or the Main Steam Safety Valves (MSSVs). The SDBCS exhausts steam from the SG to the Main Condenser (if available) or to atmosphere. The EFW system capability is discussed in further detail below.
The SDBCS is used to support cooldown to Mode 5 when desired. The system valves can be
 
operated from the Control Room or locally (valves are located in the secondary plant). When RCS pressure and temperature requirements are met, the Shutdown Cooling (SDC) system is placed in service to continue decay heat removal through Mode 4 (hot shutdown) and into
 
Mode 5. Based on the above, long term safe and stable conditions can be maintained with forced or natural circulation via the SGs. Cooldown to Mode 5 may be performed, if desired, and further long term core cooling established via the SDC system. The SDC system will maintain the fuel Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 36 in the reactor vessel in a safe and stable condition via one of two Low Pressure Safety Injection (LPSI or SDC) pumps and respective SW-cooled heat exchanger. Given appropriate RCS pressure conditions, either Containment Spray pump can also be aligned to serve as a SDC pump. Emergency Feedwater Mode 3 conditions can be maintained via forced or natural circulation supported by steaming from one or both SGs. A qualified Condensate Storage Tank (QCST) provides a source of condensate grade water to the ANO-2 EFW pumps and, as needed, the ANO-1 EFW pumps. The ANO-2 EFW system also has direct access to two non-qualified CSTs. ANO-1 also maintains a non-QCST containing condensate grade water. Valves can be manipulated to transfer water between ANO-1 and ANO-2, if needed. The ANO-2 Technical Specification (TS) requirements establish a minimum volume of available condensate (equivalent to the partial contents of one tank) that ensures ANO-2 can be supplied sufficient EFW to maintain Mode 3 conditions for up to 1 hour and then transition to Mode 4 at a cooldown rate of 75 &deg;F/hr (ANO-2 Safety Analysis Report (SAR) Section 9.2.6.1). However, the other aforementioned tanks normally contain sufficient volume that would support maintaining Mode 3 conditions for a prolonged period of time. In addition, all tanks can receive makeup from the Domestic Water system and the city water supply.
 
Should condensate sources be exhausted, the ANO-2 EFW pump suctions can be aligned to the SW system (Lake Dardanelle) as an indefinite supply of cooling water. The SW system can also be aligned to/from the Emergency Coo ling Pond (ECP) should Lake Dardanelle become unavailable for any reason. Any of these alignments can be manually performed from the Control Room or locally. The ECP is designed to provide the heat sink capability for the SW system for up to 30-days (ANO-2 SAR Section 9.2.5.1). Based on the above, the fuel in the reactor vessel will be maintained in a safe and stable condition during or following a fire event.
 
Process Monitoring The instrumentation selected is based on the guidance of NRC Information Notice (IN) 84-09
 
and NRC Regulatory Guide (RG) 1.189, which identify the minimum monitoring capability considered necessary for a pressurized water reactor (PWR). Instrumentation is powered from buses that provide power directly from station vital batteries or from Emergency Diesel Generators (EDGs). Battery capacity is main tained via battery chargers powered from EDGs (or offsite power, if available).
 
Support Systems Electrical Systems The AC and DC distribution systems are credited in order to meet fire protection performance goals and functions. The safeguards 4160 V buses can either be aligned to the EDGs, the Alternate AC Diesel Generator (AACDG), or available offsite power sources.
 
Emergency Diesel General Fuel Oil A source of fuel oil is required for long term reliance on the EDGs. A non-qualified bulk fuel oil storage tank supplies fuel oil to four underground safety-related EDG storage tanks (one tank per EDG, two EDGs for ANO-1 and two for ANO-2). Each ANO-2 underground storage tank Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 37 supplies a fuel oil day tank associated with the respective EDG. Fuel oil supplies can be cross-connected between ANO-1 and ANO-2, and between EDGs, if needed. The bulk tank also supplies fuel oil to the non-qualified AACDG, sometimes referred to as the station blackout diesel. A or B EDGs  The capacity of one safety related EDG fuel oil tank plus the capacity of the respective fuel oil day tank will support 3.5 days of operation for one EDG during an extended loss of offsite power condition at full rated load. The mission time assumes post-accident conditions with electrical loads significantly greater than those expected to support a fire event with no concurrent design basis accident (DBA).
AACDG The AACDG has a fuel oil day tank to initially supply the AACDG operation pending fuel transfer. Assuming the bulk fuel oil storage tank (described above) is maintained at minimum level, sufficient fuel oil is available for the AACDG to run at full load for a minimum of 4.5 days. The AACDG acts as a backup to one or both ANO units should EDGs fail for any reason.
 
The onsite fuel oil capacity is sufficient to operate the EDGs or AACDG for longer than the time that would be necessary to replenish the onsite supply from outside sources.
Cooling Systems Active heating, ventilation, and air conditioning (HVAC) systems are required for limited plant areas as described in Attachment C (Table B-3).
Based on the above, sufficient support systems will remain available to ensure the fuel in the reactor vessel will be maintained in a safe and stable condition during or following a fire event.
 
Summary  The fire brigade will respond to fire events within the Protected Area boundary in accordance with procedures, thus mitigating the overall im pact of the event. In addition, any fire or explosion onsite affecting Engineered Safety Features (ESF) systems will result in an Emergency Class (EC) declaration of Alert or higher, which requires Emergency Response Organization (ERO) activation. The ERO will assist the Control Room personnel with implementation of the longer term actions necessary to maintain the fuel in a safe and stable configuration. Following plant stabilization in Mode 3 (assuming the fire required a unit shutdown), assessment and repair activities would commence to restore plant equipment or replenish supplies needed to support long term Mode 3 operation, RCS cooldown, or reactor restart. ERO resources will be available to assi st the Control Room in fire damage assessment and establishing multiple success paths.
 
The 'At Power' safe and stable strategy presents no adverse impact on risk due to the following considerations:
 
Procedures exist to address loss of power and other loss of equipment that may result from a fire event  The ERO will be activated for fires that could affect one or more Engineered Safety Features equipment trains to provide site technical support Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 38 Compensatory measures and recovery plans can be developed based on the fire damage scenario
 
The transition for ANO-2 to a new NFPA 805 fire protection licensing basis under 10 CFR 50.48(c) per NEI 04-02 requires that the licensee perform an engineering analysis to assess the impact of fires occurring in all operational modes, including non-power operations (NPO). The 'Non-Power' analysis strategy is intended to prevent fires from occurring. For all non-power modes, the equipment required to demonstrate key safety functions are identified using a pinch-point analysis. The 'Non-Power' safe and stable strategy includes cooldown initiating from hot standby (Mode 3), through Modes 4, 5, 6 and defueled (i.e. no-mode) and places SDC in service for long term cooling capability (see Attachment D).
The balance of 'At Power' and 'Non-Power' strategies meets the definition of nuclear safety goal of NFPA 805, Section 1.3.1, in that "reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintain the fuel in a safe and stable condition."
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 operator manual actions (OMAs) as recovery actions in the LAR (RG 1.205, Regulatory Position 2.21 and NEI-04-02, Section 4.6). As a minimum, NEI 04-02 suggests that the assumptions, criteria, methodology, and overall results be included for the NRC to determine the acceptability of the licensee's methodology.
 
The discussion below provides the methodology used to transition pre-transition OMAs and to determine the population of post-transition recovery actions. This process is based on FAQ 07-0030 (ML110070485) and consists of the following steps:
 
Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition. Step 2: Determine the population of recovery actions that are required to resolve variances from deterministic requirements (VFDRs) (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth). Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path. Step 4: Evaluate the feasibility of the recovery actions.
Step 5: Evaluate the reliability of the recovery actions.
Results
 
The review results are documented in EC-27716, "ANO2 Fire Area Risk Evaluations for Transition to NFPA-805."  Refer to Attachment G for the detailed evaluation process and summary of the results from the process.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 39 4.2.1.4 Evaluation of Multiple Spurious Operations
 
NEI 04-02 suggests that a licensee submit a summary of its approach for addressing potential fire-induced multiple spurious operations (MSOs) for NRC review and approval. As a minimum, NEI 04-02 suggests that the summary contain sufficient information relevant to methods, tools, and acceptance criteria used to enable the NRC to determine the acceptability of the licensee's methodology. The methodology utilized to address MSOs for ANO-2 is summarized below.
 
As part of the NFPA 805 transition project, a review and evaluation of ANO-2 susceptibility to fire-induced MSOs was performed. The process was conducted in accordance with NEI 04-02 and RG 1.205, as supplemented by FAQ 07-0038, Revision 3 (ML110140242). The PWR Generic MSO list in Revision 2 of NEI 00-01, dated May, 2009, was utilized.
 
The approach outlined in Figure 4-3 (based on FAQ 07-0038, Revision 3) is one acceptable method to address fire-induced MSOs. This method used insights from the Fire PRA developed in support of transition to NFPA 805 and consists of the following:  Identifying potential MSOs of concern. Conducting an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Revision 1, Section F.4.2). Updating the Fire PRA model and existing post-fire NSCA to include the MSOs of concern. Evaluating for NFPA 805 Compliance. Documenting Results.
This process is intended to support the transition to a new licensing basis. Post-transition changes would use the RI-PB change process. The post-transition change process for the assessment of a specific MSO would be a simplified version of this process, and may not need the level of detail shown in the following section (e.g., an expert panel may not be necessary to identify and assess a new potential MSO; identification of new potential MSOs may be part of the plant change review process and/or inspection process).
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 40 Figure 4-3 Multiple Spurious Operations - Transition Resolution Process (Based on FAQ 07-0038)
 
Results Refer to Attachment F for the process used by ANO-2 and the results from the process.
 
Identify Potential MSOs of Concern
* NSCA
* Generic List of MSOs
* Self Assessments
* PRA Insights
* Operating Experience Ex pert Panel Identif y and Document MSOs of Concern Update PRA model & NSCA (as appropriate) to include MSOs of concern
* ID equipment
* ID logical relationships
* ID cables
* ID cable routing Ste p 1 Ste p 2 Ste p 3 Ste p 4 Ste p 5 Pursue other resolution o p tionsDocument Results Evaluate for NFPA 805 Compliance Compliant with NFPA 805?
Yes No Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 41 4.2.2 Existing Engineering Equivalency Evaluation (EEEE) 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,  The EEEE is an appropriate use of an engineering equivalency evaluation,  The EEEE is of appropriate quality,  The standard license condition is met,  The EEEE is technically adequate,  The EEEE reflects the plant as-built condition, and  The basis for acceptability of the EEEE remains valid In accordance with the guidance in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02, as clarified by FAQ 07-0054, "Demonstrating Compliance with Chapter 4 of NFPA 805," EEEEs that demonstrate that a fire protection system or feature is "adequate for the hazard" are summarized in the LAR as follows:  If not requesting specific approval for "adequate for the hazard" EEEEs, then the EEEE 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 reference 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 EC-31053, "NFPA 805 Existing Engineering
 
Evaluation Transition."
In accordance with the guidance provided in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02, as clarified by FAQ 07-0054, "Demonstrating Compliance with Chapter 4 of NFPA 805," EEEEs used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in the Attachments A and C as appropriate.
In addition, none of the transitioning EEEEs require NRC approval.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 42 4.2.3 Licensing Action Transition
 
The existing licensing actions (exemption requests and safety evaluations) review was performed in accordance with NEI 04-02. The methodology for the licensing action review included the following:  Determination of the bases for acceptability of the licensing action. Determination that these bases for acceptability are still valid and required for NFPA 805.
Results
 
Attachment K contains the detailed results of the Licensing Action Review. 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).
 
The following licensing actions will be transitioned into the NFPA 805 fire protection program as previously approved (NFPA 805 Section 2.2.7). These licensing actions are considered compliant under 10 CFR 50.48(c). Appendix R Exemption 17, FA - NN, RCP Oil Collection, Not Meeting III.O Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. Appendix R Exemption 19, FA - NN, RCP Oil Fill Line, Not Meeting III.O Criteria, NRC approval letter 2CNA069701 dated 6/14/1997.
 
The following licensing actions are no longer necessary and will not be transitioned into the NFPA 805 fire protection program:  Appendix R Exemption 01, FA - OO, Not Meeting III.G.2 Criteria, NRC approval letter 2CNA109902 dated 10/1/1999. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 02, FA - OO, Not Meeting III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 03, FA - CC, Not Meeting 3-hour Rated Barrier III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because an EEEE has found the current plant configuration to be acceptable. Appendix R Exemption 04, FA - NN, Not Meeting III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 43 Appendix R Exemption 05, FA - DD, Not Meeting III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 06, FA - EE-L (Originally a portion of FA - EE), Not Meeting III.G Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the door of issue has been modified and determined to be acceptable. Appendix R Exemption 07, FA - JJ, Lack of Barrier/Separation III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 08, FA - GG, Not Meeting III.G.3 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 09, FA - HH, Not Meeting III.G.3 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 10, FA - B-3, Not Meeting III.G.3 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 11, FA - B-2, Not Meeting III.G.3 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 12, FA - G, Not Meeting III.G.3 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 13, FA - KK (Originally FA - B), Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. This exemption is no longer required because a fire in this zone has been shown to not result in a loss of offsite power.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 44 Appendix R Exemption 13A, FA - QQ (Originally FA - B), Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. This exemption is no longer required because a fire in this zone has been shown to not result in a loss of offsite power. Appendix R Exemption 14, FA - G, Not Meeting III.G.3 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 15, FA - EE-L (Originally a portion of FA - EE), Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4. Appendix R Exemption 16, FA - YD, Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. This exemption is no longer required because the postulated loss of tank level indication does not cause a fire related need for plant shutdown, due to the Technical Specification required minimum volume maintained during power operation, and remaining
 
deterministically compliant under NFPA 805 criteria. Appendix R Exemption 18, FA - YD, Emergency Lighting, Not Meeting III.J Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. This exemption is no longer required because NFPA 805 does not require 8-hour battery backed emergency lights.
 
Since the above exemptions are either compliant with 10 CFR 50.48(c) or no longer necessary, in accordance with the requirements of 10 CFR 50.48(c)(3)(i), ANO-2 requests that the above exemptions be rescinded or retained as part of the LAR process, as designated in Attachment K. See Attachment O, Orders and Exemptions.
 
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.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 45 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 evaluat ions (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.
Step 3 - VFDR Identification and 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 4.5.2 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.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 46 Figure 4-4 Summary of Fire Area Review [Based on FAQ 07-0054 Revision 1]
 
FA - Fire Area Assemble Documentation Document fulfillment of Nuclear Safety Performance Criteria (B-3 Table) Identify INITIAL Variances from Deterministic Requirements of NFPA 805 Section 4.2.3        (B-3 Table)
Select Approach    NFPA 805 Chapter 4 Document Final Disposition of VFDR Compliance options include: -  Accept as is                                              -  Require FP system/features                      -  Require recovery action                           
-  Require Programmatic Enhancements      -  Require Plant Modifications (B-3 Table) Bring into Compliance with Section 4.2.3 of NFPA 805 Use Fire Modeling to demonstrate compliance with NFPA 805 Section 4.2.4.1 Use Fire Risk Evaluations to demonstrate compliance with NFPA 805 Section 4.2.4.1 NFPA 805 Section 4.2.4.1 Criteria Met? Delta CDF/LERF Acceptable (on a FA basis) & DID and Safety Margin Maintained? NFPA 805 Section 2.4.4 Guidance from RG 1.174 Section 2 & RG 1.205 Section 2.2.4 Yes Yes No Document Required Fire Protection Systems and Features                    (B-3 and LAR Table 4-3) Select another Compliance Option No Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 47 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 NSCA Documents are provided. Licensing Actions - Specific references to exemption requests / deviations / 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 is provided.
Attachment T contains no items for which ANO-2 is requesting concurrence of prior
 
approval. EEEE - Specific references to EEEE that rely on determinations of "adequate for the hazard" that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability is 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.
 
4.3 Non-Power Operational (NPO) Modes
 
4.3.1 Overview of Evaluation Process
 
ANO-2 implemented the process outlined in NEI 04-02 and FAQ 07-0040, Clarification on Non-Power Operations. The goal (as depicted in Figure 4-5) is to ensure that contingency plans are established when the plant is in a 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:
o Review plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and o 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.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 48 Figure 4-5 Review Plant Operational States (POSs), KSFs, Equipment, and Cables, and Identify Pinch Points
 
Review Existing Outage Management Process For the Evaluated Plant Operational States, identify the KSFs for review Is the Required Equipment    for each identified KSF included in SSD Analysis? Are the Cables analyzed and appropriate and complete? Identify equipment, cables, and appropriate information and Enter into Analytical Database Yes Yes No No Identify locations where a Single Fire may damage All Credited Paths for KSF Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 49 Figure 4-6 Manage Pinch Points
 
4.3.2 Results of the Evaluation Process
 
Based on FAQ 07-0040, the plant operating states considered for equipment and cable selection are defined in calculation CALC-09-E-0008-02 "ANO-2 NFPA 805 Non Power Operations Assessment."  Once the applicable plant operating state for non-power operations was defined, the systems necessary to maintain and support each KSF were identified. The associated piping and instrument diagrams (P&IDs) and single line drawings for the required systems were marked-up and annotated to identify the necessary equipment and to develop a CAFTA fault tree. The drawing mark-ups and the fault tree are documented in support calculation CALC-09-E-0008-04 "ANO-2 NFPA 805 NPO Fault Tree and PID Attachments."
The CAFTA fault tree provides all associations for power supplies, supporting equipment, and other equipment dependencies that could fail equipment necessary to NPO.
Equipment        Out of Service (OOS) Fire Protection Defense-in-Depth Actions Fire Protection Defense-in-Depth Actions Fire Protection Defense-in-Depth Actions Determine Fire Area Impact based on NPO Fire Area Assessments Implement Contingency Plan for Specific KSF 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 KSF Equipment Availability Changed? Higher Risk Evolution?
KSF Lost? Yes Yes Yes No No No Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 50 The Plant Data Management System (PDMS) is the cable and raceway software that provides the controlled database for NSCA equipment and associated circuit analysis. In general NPO equipment is a subset of NSCA equipment. Existing equipment was evaluated in PDMS to determine if the circuit analysis was appropriate for NPO. Additional equipment identified as being needed for NPO, but not previously evaluated, was evaluated and added as necessary to PDMS and, where added, flagged accordingly as being only required for NPO. All new circuit analysis was performed in accordance with existing methodologies established at ANO consistent with guidance provided within NEI 00-01.
The pinch-point analysis was performed using ARC software. ARC software extracts the necessary data from PDMS and maps it to the NPO CAFTA fault tree. Each Fire Area for NPO was evaluated to determine which equipment could be rendered unavailable. Equipment which could spuriously operate or fail resulting in the loss of a KSF in a Fire Area was given a compliance strategy (i.e. recovery action) to allow NPO compliance (top gate success). This effectively captured affected equipment necessary to maintain a KSF in any plant area/zone which could be compromised due to a fire. In accordance with FAQ 07-0040, any fire area not in deterministic compliance caused by all of the credited success paths for a given KSF being lost is considered a pinch-point.
The results of each Fire Area assessed for NPO are described in detail in calculation CALC E-0008-02 with slightly less than half the fire areas in deterministic compliance. Availability of systems and equipment for each KSF is identified. Recoveries due to a pinch-point are provided by KSF in tabular form. The presence of detection and suppression systems and any existing procedural controls is indicated. Where Fire Areas of multiple zones are comprised, clarification is provided to illustrate the zones impacted (pinch-point) and those unaffected (deterministic compliance). Insights from CALC-08-E-0016-01 "Fire Probabilistic Risk Assessment Plant Partitioning and Fire Ignition Frequency Development" have been used to provide a risk-informed assessment of any Fire Area determined to be a pinch-point.
Consideration and usage of the following methods to manage risk were applied as applicable to
 
any Fire Area that is a pinch-point:
Prohibition or limitation of hot work in fire zones during periods of increased vulnerability. Limitation of combustible materials in fire zones during periods of increased vulnerability. Pre-emptive actions such as opening breakers or re-aligning of equipment, if hot work is to be performed. Modification to eliminate spurious operation in areas determined to be pinch-points.
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 us ing the methodology contained in CALC-ANO2-FP-08-00001, "NEI 04-02 Table G-1, Radioactive Release Transition Report" (Rev. 2). The objective of the CALC was to ensure fire protection goals, objectives, and criteria were met as they relate to potential radioactive release scenarios. The methodology consists of the
 
following:  A review of ANO-2 and common fire pre-plans (PFP-U2 and PFP-UC) and fire brigade training materials was performed to identify Fire Protection Plan 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.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 51 A review of engineering controls to ensure containment of gaseous and liquid effluents (e.g., smoke and fire fighting agents) was also performed. This review included all plant operating modes (including full power and non-power conditions).
 
4.4.2 Results of the Evaluation Process The radioactive release review determined the fire protection (FP) program is compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205. The site specific review of the direct effects of fire suppression activities on radioactive release is summarized in
 
Attachment E.
 
4.5 Fire PRA and Performance-Based Approaches
 
RI-PB evaluations are an integral element of an NFPA 805 fire protection program. Key parts of RI-PB evaluations include:  A Fire PRA (discussed in Section 4.5.1 and Attachments U, V, and W). NFPA 805 Performance-Based Approaches (discussed in Section 4.5.2).
4.5.1 Fire PRA Development and Assessment
 
In accordance with the guidance in RG 1.205, a Fire PRA (FPRA) model was developed for ANO-2 in compliance with the requirements of Part 4 "Internal Fires at Power Probabilistic Risk Assessment Requirements," of the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS) combined PRA Standard, ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application," (hereafter referred to as Fire PRA Standard). ANO-2 conducted a peer review by independent industry analysts in accordance with RG 1.200 prior to a risk-informed submittal. The resulting fire risk assessment model is used as the analytical tool to perform Fire Risk Evaluations (FREs) during the transition process.
 
Section 4.5.1.1 describes the Internal Events PRA model. Section 4.5.1.2 describes the FPRA model. Section 4.5.1.3 describes the results and resolution of the peer review of the FPRA, and Section 4.5.1.4 describes insights gained from the FPRA.
 
4.5.1.1 Internal Events PRA
 
The ANO-2 base internal events PRA (ANO-2 PSA Level-1 Model 4p02) was the starting point for the FPRA.
 
The ANO-2 PRA has undergone a RG 1.200, Revision 1, Peer Review against the ASME PRA standard requirements by a team of knowledgeable industry (vendor and utility) personnel. The ASME PRA standard contains a total of 327 numbered supporting requirements in nine technical elements and the configuration control element. Eleven of the standard requirements represent deleted requirements (IE-A8, IE-A9, SC-A3, SY-A9, SY-B9, HR-G8, IF-A2, IF-B4, IF-D2, IFE2, and QU-D2) and twenty-six were determined to be not applicable to the ANO-2 PRA. Of the 290 remaining standard requirements, 252 standard requirements, or 87%, were rated as Capability Category II or greater. Only 9% of the standard requirements were rated as not met.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 52 In the course of this review, fifty-nine new Facts and Observations (F&Os) were prepared, including twenty five suggestions and thirty four findings. Most of the findings pertained to documentation issues. However, there were four key findings. Those key findings are discussed below.
: 1. While Entergy did calculate initiating events based on reactor critical years, the frequencies were not adjusted to reflect average plant availability. Applying an adjustment factor to account for average plant availability will result in a slight decrease in Core Damage Frequency (CDF).
Response in relation to NFPA-805 The initiating event frequencies are set to True or False in the quantification of the Fire PRA. Therefore, this issue does not affect the Fire PRA results.
: 2. There appeared to be some mission time errors in the Interfacing System Loss of Coolant Accident (ISLOCA) model. The result was that the ISLOCA contributions to CDF for ANO-2 were several orders of magnitude lower than the industry averages.
While changes to the ISLOCA model would not significantly impact overall CDF, there could be an impact on Large Early Release Frequency (LERF).
Response in relation to NFPA-805 The ANO-2 PRA basic event file was revised to reflect longer mission times for those applicable ISLOCA events. 3. ANO-2 used the Technique for Human Error Rate Prediction (THERP) and Accident Sequence Evaluation Program (ASEP), Basic Human Error Probabilities (BHEPs) to quantify their operator action Human Error Probabilities (HEPs). The problem is that the THERP and ASEP data is presented in terms of median values and error ranges while the CAFTA code requires means and error ranges as input. Entergy did not convert the medians and error ranges to means and error ranges so they effectively biased their HEPs in a non-conservative direction.
Response in relation to NFPA-805 The data used in calculating the HEPs for the ANO-2 human recovery events has been converted to mean values as required in the internal events PRA model. Mean values were also used in the development of HEPs in the Fire PRA developed  for NFPA-805.
: 4. There was a discrepancy between the emergency diesel generator component boundaries defined in PRA-ES-01-003, "Compilation of Generic Reliability Data and Component Boundaries for Probabilistic Safety Assessment" and the component boundaries found in the actual model.
Response in relation to NFPA-805 The internal events PRA model has been revised by setting those events modeled separately from the EDGs, but considered part of the EDG, to a failure probability of zero. This approach retains the detailed modeling that is desired for the EOOS model, but removes the impact of conservatively accounting for these failure probabilities in the quantification of the model.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 53 Also, at the time of the peer review, the ANO-2 Internal Flood Analysis (IFA) had not been completed to the point that it could be reviewed. Entergy used the same IFA methodology for all three of their PWRs with the Waterford-3 plant being the lead plant. The Waterford-3 IFA had been essentially completed at the time of the peer review. Therefore, Entergy requested that the peer review team review the IFA methodology for Waterford-3 to confirm that the methodology met the standard. The Peer Review findings reflect the review of the Waterford-3 IFA. ANO-2 has since completed the ANO-2 IFA in accordance with the methodology and addressed the issues identified during the Waterford-3 review. In addition, a peer review of the ANO-1 IFA has been conducted. The issues identified as part of the ANO-1 IFA were also incorporated as issues to be addressed as part of the ANO-2 IFA. The standard requirements and the status of any findings associated with the IFA are included in Attachment U.
 
The discussions above relate to the most significant issues identified in the Peer Review.
Attachment U provides a detailed assessment of each of the standard requirements reviewed by the Peer Review team.
 
4.5.1.2 Fire PRA
 
The internal events PRA was modified to capture the effects of fire both as an initiator of an event and as a potential failure mode for affected circuits and individual targets. The FPRA was developed primarily through the use of the guidance for FPRA development in NUREG/CR-6850, approved FAQs, and recent EPRI FPRA methodology development efforts.
The FPRA was quantified using the EPRI FRANC software.
 
The FPRA quality and results are discussed in the subsequent sections and in Attachments V and W, respectively.
 
Fire Model Utilization in the Application
 
Fire modeling was performed as part of the Fire PRA development (NFPA 805 Section 4.2.4.2). RG 1.205, Regulatory Position 4.2, and Section 5.1.2 of NEI 04-02, provide guidance to identify fire models that are acceptable to the NRC for plants implementing a risk-informed, performance-based licensing basis.
 
The acceptability of the use of these fire models is included in Attachment J.
 
4.5.1.3 Results of Fire PRA Peer Review
 
The ANO-2 FPRA was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, Part 4. The review was conducted by the Westinghouse Owners Group in June 2009.
 
The results (i.e., standard requirement capability assessments and F&Os) documented in the FPRA peer review report were used to support further development of the FPRA for the NFPA 805 application.
 
The FPRA update addressed the standard requirements assessed recommended improvements (i.e., Not Met or Capability Category I). Completion of recommendations related to standard requirement assessments and finding type F&Os results in a closure of technical gaps to a Capability Category II assessment for the associated standard requirements. Any outstanding findings have been dispositioned for the potential impact on the FPRA and the application. The results of the peer review are summarized in Attachment V.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 54 4.5.1.4 Risk Insights
 
Risk insights were documented as part of the development of the FPRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for FPRA development and is useful in identifying the areas of the plant where fire risk is greatest. A detailed description of significant risk sequences associated with the fire initiating events that contribute above 1% of the calculated fire risk for the plant was prepared for the purposes of gaining these insights and an understanding of the risk significance of MSO combinations. These insights are provided in Table W-1.
 
4.5.2 Performance-Based Approaches
 
NFPA 805 outlines the approaches for performing performance-based analyses. As specified in Section 4.2.4, there are generally two types of analyses performed for the performance-based approach:  Fire Modeling (NFPA 805, Section 4.2.4.1). Fire Risk Evaluation (NFPA 805, Section 4.2.4.2).
4.5.2.1 Fire Modeling Approach
 
The fire modeling approach was not utilized for the transition.
 
4.5.2.2 Fire Risk Approach
 
Overview of Evaluation Process
 
The Fire Risk Evaluations were completed as part of the ANO-2 NFPA 805 transition. These Fire Risk Evaluations were developed using the process described below. This methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. These are summarized in Table 4-1.
 
Table 4-1 Fire Risk Evaluation Guidance Summary Table Document Section(s) Topic NFPA 805 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, Appendix J Change Evaluation, Change Evaluation 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)
 
During the transition to NFPA 805, variances from the deterministic approach in Section 4.2.3 of NFPA 805 were evaluated using a Fire Risk Evaluation per Section 4.2.4.2 of NFPA 805. A Fire Risk Evaluation was performed for each fire area containing VFDRs.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 55 If the Fire Risk Evaluation meets the acceptance criteria, this is confirmation that a success path effectively remains free of fire damage and that the performance-based approach is acceptable per Section 4.2.4.2 of NFPA 805.
The Fire Risk Evaluation process consists of the following steps (Figure 4-7 depicts the Fire Risk Evaluation process used during transition.
This is generally based on FAQ 07-0054, Revision 1):
 
Step 1 - Preparation for the Fire Risk Evaluation. Definition of the Variances from the Deterministic Requirements. The definition of the VFDR includes a description of the problem statement and the section of NFPA 805 that is not met, type of VFDR (e.g., separation issue or degraded fire protection system), and proposed evaluation per applicable NFPA 805 section. Preparatory Evaluation - Fire Risk Evaluation Review. Using the information obtained during the development of Attachment C and the Fire PRA, a review of the VFDR was performed. Depending on the scope and complexity of the VFDR, the reviewers may include the Safe Shutdown/NSCA Engineer, the Fire Protection Engineer, and the Fire PRA Engineer. The purpose and objective of this review was to address the following:
o Review of the Fire PRA modeling treatment of VFDR o Ensure discrepancies were captured and resolved
 
Step 2 - Performed the Fire Risk Evaluation  The Evaluator coordinated as necessary with the Safe Shutdown/NSCA Engineer, Fire Protection Engineer, and Fire PRA Engineer to assess the VFDR using the Fire Risk Evaluation process to perform the following:
o Change in Risk Calculation with consideration for additional risk of recovery actions and required fire protection systems and features due to fire risk.
o Fire area change in risk summary
 
Step 3 - Reviewed the Acceptance Criteria  The acceptance criteria for the Fire Risk Evaluation consist of two parts. One is quantitatively based and the other is qualitatively based. The quantitative figures of merit are CDF and LERF. The qualitative factors are defense-in-depth and safety margin.
o Risk Acceptance Criteria. The transition risk evaluation was measured quantitatively for acceptability using the CDF and LERF criteria from RG 1.174, as clarified in RG 1.205, Regulatory Position 2.2.4.
o Defense-in-Depth. A review of the impact of the change on defense-in-depth was performed, using the guidance from NEI 04-02. NFPA 805 defines defense-in-depth
 
as: - Preventing fires from starting - Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage - Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 56 In general, the defense-in-depth requirement was considered to be satisfied if the proposed change does not result in a substantial imbalance among these elements (or echelons). The review of defense-in-depth was qualitative and addressed each of the elements with respect to the proposed change. Defense-in-depth was performed on a fire
 
area basis. Fire protection features and systems relied upon to ensure defense-in-depth were identified as a result of the assessment of defense-in-depth.
o Safety Margin Assessment. A review of the impact of the change on safety margin was performed. An acceptable set of guidelines for completing the assessment is summarized below. - Codes and standards or their alternatives accepted for use by the NRC are met, and - Safety analysis acceptance criteria in the licensing basis (e.g., SAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty The requirements related to safety margins for the change analysis are described for each of the specific analysis types used in support of the FRE.
FIGURE 4-7 Fire Risk Evaluation Process (NFPA 805 Transition) [Based on FAQ 07-0054, Revision 1)
 
Identification of VFDRs        (From B-3 Tables) Determine how to model the VFDR in the Fire PRA Calculate VFDR Delta CDF and Delta LERF Discuss and Document in Fire PRA and Fire Risk Evaluation Documentation Prepare for Fire Risk Evaluation Perform Fire Risk Evaluation Review of Acceptance Criteria Evaluate Delta CDF and Delta LERF Evaluate the Maintenance of Defense-in-Depth and Safety Margin Discuss and Document in Fire Risk Evaluation Calculation Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 57 Results of Evaluation Process
 
Disposition of VFDRs The ANO-2 SSA and the NFPA 805 transition project activities have identified a number of variances from the deterministic requirements of NFPA 805 Section 4.2.3. These variances were dispositioned using the FRE process.
 
Each variance dispositioned using a FRE was assessed against the FRE acceptance criteria of CDF and LERF; and maintenance of defense-in-depth and safety margin criteria from Section 5.3.5 of NEI 04-02 and RG 1.205. The results of these calculations are summarized in
 
Attachment C.
 
Following completion of transition activities and planned modifications and program changes, the plant will be compliant with 10 CFR 50.48(c).
 
Risk Change Due to NFPA 805 Transition
 
In accordance with the guidance in RG 1.205, Section C.2.2.4, Risk Evaluations, risk increases or decreases for each fire area using FREs and the overall plant should be provided. Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area.
 
RG 1.205, Section C.2.2.4.2 states in part:
 
"The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions). The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk neutral or represent a risk decrease."
The risk increases and decreases are provided in Attachment W.
 
4.6 Monitoring Program
 
NFPA 805, Section 3.2.3(3), requires 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, establishes 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 ANO NFPA 805 Monitoring program.
The monitoring program will be implemented after issuance of the Safety Evaluation, as part of the Fire Protection Program transition to NFPA 805. In order to assess the impact of the transition to NFPA 805 on the current monitoring program, the ANO Fire Protection Program Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 58 documentation, such as the maintenance program processes, Fire Protection Program implementing procedures, and plant change processes will be reviewed. Sections 4.5.3 and 5.2 of NEI 04-02, as clarified in the NRC approved version of FAQ 10-0059, will be used during the review process. The process is described in the following sections.
 
The following scope will be documented appropriately in the ANO NFPA 805 Monitoring
 
Program:
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 ANO will be completed as part of NFPA 805 amendment implementation (See Attachment S).
 
4.6.1 Overview of NFPA 805 Requirements for the NFPA 805 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 establish the NFPA 805 monitoring program and 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 equi pment within existing plant programs  The performance criteria for the availability and reliability of the required structures, systems and components  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 following Safety Evaluation issuance as part of the transition to NFPA 805. The monitoring process will be comprised of four phases.
Phase 1 Scoping Phase 2 Screening using risk criteria Phase 3 Risk target value determination Phase 4 Monitoring implementation The phases of the monitoring process are described as follows and depicted in Figure 4-8. The results of these phases will be documented in the ANO monitoring program evaluation
 
developed during implementation.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 59 Phase 1 Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program:
SSCs required to comply with NFPA 805, specifically:
o Fire protection systems and f eatures required by the NSCA o Fire protection systems and f eatures modeled in the FPRA o Fire protection systems and features required by Chapter 3 of NFPA 805 o NSCA equipment (for the purposes of NFPA 805 Monitoring, "NSCA equipment" includes NSCA equipment, Fire PRA equipment, and NPO equipment) o SSCs relied upon to meet radioactive release criteria  Fire protection programmatic elements  Key assumptions in engineering analyses (specifically analyses performed to demonstrate compliance with the nuclear safety and radioactive release performance criteria)
As a minimum, the fire protection systems and features (required to meet Chapter 3 of NFPA 805 and the NSCA criteria) and SSCs required to meet the radioactive release criteria will be included in the existing inspection and test program, and system/program health program. In addition passive features (barriers, drains, curbs, etc.) that are relied upon to demonstrate compliance with Chapter 4 of NFPA 805 will also be included in the existing inspection and test program, and system/program health program. Once the applicable NFPA 805 radioactive release and passive feature SSCs have been added to the existing inspection and test program as well as system/program health programs, the existing programs will be adequate for routine
 
monitoring of these SSC.
 
Plant specific initiatives may be undertaken to optimize fire protection surveillance and testing practices and frequencies based upon performance in accordance with the guidance in EPRI Technical Report 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide
 
for Fire Protection Systems and Features."
 
Phase 2 Screening Using Risk Criteria The equipment from the Phase 1 scoping will be screened to determine the appropriate level of NFPA 805 monitoring. As a minimum, the SSCs identified in Phase 1 will be part of an inspection and test program, and/or system/program health reporting process. If not included in the current program, the SSC(s) will be added in order to assure that the criteria can be met
 
reliably.
: 1. Fire Protection Systems and Features Those fire protection systems and features identified in Phase 1 are candidates for additional monitoring in the NFPA 805 program commensurate with risk significance.
Compartments smaller than Fire Areas may be used provided the compartments are independent (i.e., share no fire protection SSC). If compartments smaller than Fire Areas are used, the basis will be documented in the NFPA 805 Monitoring Program engineering evaluation.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 60 The Fire PRA is the primary tool used to establish the risk significance criteria and performance bounding guidelines. The screening thresholds used to determine risk significant analysis units will be those that meet the following criteria:
Risk Achievement Worth (RAW) of the monitored parameter  2.0 AND either Core Damage Frequency (CDF) x (RAW)  1.0E-7 per year OR Large Early Release Frequency (LERF) x (RAW)  1.0E-8 per year CDF, LERF, and RAW (monitored parameter) are calculated for each fire area. The "monitored parameter" will be established at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration).
Fire protection systems and features that meet or exceed the criteria identified above will be included in a monitoring program such as the site Maintenance Rule Program described in applicable Maintenance Rule Program procedures:  EN-DC-203, Maintenance Rule Program  EN-DC-204, Maintenance Rule Scope and Basis  EN-DC-207, Maintenance Rule Periodic Assessment  EN-DC-205, Maintenance Rule Monitoring Fire protection functions and SSCs will be classified as high or low risk significant in Maintenance Rule and appropriate performance criteria established. The remaining required fire protection systems and features will be monitored in accordance with existing inspection and test programs and in the existing system/program health program and fire impairment processes and procedures, such as EN-DC-143-02.
: 2. Nuclear Safety Capability Assessment (NSCA) Equipment Required NSCA equipment identified in Phase 1 (except equipment within the scope of Non-Power Operations) will be screened for safety significance using the Fire PRA and the Maintenance Rule Scope and Basis guidelines, which differentiate High Safety Significance (HSS) equipment from Low Safety Significance (LSS) equipment. HSS NSCA equipment not currently monitored in the Maintenance Rule will be included in the Maintenance Rule program. All NSCA equipment not designated as HSS will be considered LSS and not included in the monitoring program beyond normal inspection and test programs, and system/program health reporting processes and procedures.
For NPO modes, attempting to quantitatively measure the effectiveness of fire prevention to manage fire risk during Higher Risk Evolutions is not feasible. Therefore, fire risk management effectiveness will be monitored programmatically similar to combustible material control and other fire prevention program processes. Additional monitoring beyond inspection and test programs or system/program health reporting will not be necessary to assess fire risk management effectiveness during NPO modes.
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 61
: 3. SSC Relied upon for Radioactive Release Criteria
 
Since the evaluations performed to meet the radioactive release performance criteria are qualitative, the SSC relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement. Additionally, since 10 CFR Part 20 limits (which are lower than releases due to core damage and containment breach) for radiological effluents are not being exceeded, equipment relied upon to meet the radioactive release performance criteria is considered inherently low risk. Therefore, additional monitoring beyond inspection and test programs and system/program health reporting is not considered necessary.
: 4. Fire Protection Programmatic Elements
 
Monitoring of programmatic elements is required in order to assess the performance of the fire protection program in meeting the performance criteria. These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements. Programmatic elements include:  Transient combustible control and transient exclusion zones  Hot work control and administrative controls  Impairment and compensatory measures including program compliance  Fire brigade effectiveness Monitoring of programmatic elements is qualitative in nature since the programs are not amenable to the numerical methods used to derive reliability and availability.
Phase 3 Risk Target Value Determination Phase 3 establishes the target values for reliability and availability for the fire protection systems and features that met or exceeded the screening criteria and for the HSS NSCA equipment identified in Phase 2.
 
Target values for reliability and availability for the fire protection systems and features are established at the component level, program level, or functionally through the use of the pseudo-system or the "performance monitoring group" (PMG) concept. The actual action level is determined based on the number of component, program, or functional failures within a sufficient bounding time period (2 to 3 operating cycles).
 
Since the HSS NSCA equipment is identified using Maintenance Rule guidelines, the associated equipment-specific performance criteria will be established as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions.
 
The action level threshold for reliability and availability will be no lower than the fire PRA assumptions. Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against these action levels. The Monitoring Program failure criteria and action level targets will be documented in the NFPA 805 Monitoring Program
 
engineering evaluation.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 62 Fire protection systems and features, NSCA equipment, SSCs required to meet radioactive release criteria, and fire protection program elements that do not meet the screening criteria in Phase 2 will be included in existing inspection and test programs, and system/program health programs. Reliability and availability criteria will not be assigned.
Phase 4 Monitoring Implementation Phase 4 is the implementation of the ANO monitoring program, once the monitoring scope and criteria are established. Monitoring consists of periodically gathering, trending, and evaluating information pertinent to the performance and/or availability of the equipment, and comparing the results with established goals and performance criteria to verify the goals and criteria are being met. Results of monitoring activities will be analyzed in a timely manner to assure that appropriate corrective action is identified and taken. The corrective action process will be used to address performance of fire protection and nuclear safety SSCs that do not meet performance criteria.
 
For fire protection systems and features and NSCA HSS equipment that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, corrective actions will be initiated to identify the negative trend in accordance with the ANO corrective action processes and procedures. A corrective action plan will then be developed to ensure performance returns to the established level. Fire protection health reports, self-assessments, regulator and insurance company (NEIL) reports provide inputs to this monitoring program, as does the corrective action process delineated in procedure EN-LI-102.
 
When applicable, a sensitivity study will be performed to determine the margin below the action level that still provides acceptable fire PRA results to assist in prioritizing corrective actions.
 
A periodic assessment will 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 will be conducted as part of other established assessment activities. Issues that will be addressed include:  For systems with performance criteria, do performance criteria still effectively monitor the functions of the system?  Do the criteria still monitor the effectiveness of the fire protection and NSCA systems?  Have the supporting analyses been revised such that the performance criteria are no longer applicable or new fire protection and NSCA SSCs, programmatic elements and/or functions need to be in scope?  Based on the performance during the assessment period, are there any trends in system performance that should be addressed that are not being addressed?
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 63 Figure 4-8 Post-Transition NFPA 805 Monitoring Program
 
Yes See Figure E-2 for Detail Yes Yes Yes Yes Is performance acceptable?
No No Yes Determine scope or SSCs and Programmatic Elements to Monitor No Phase 1 Scoping Phase 2 Screening Using Risk Criteria Phase 3 Risk Target Value Determination Phase 4 Monitoring Implementation Establish risk significance criteria Establish list of HSS Fire and NSCA SSCs (for each PMG) based on Criteria Establish Functional Failure criteria and Action Levels for each PMG Establish reliability and availability criteria for each PMG Is new SSC level goal required?
Is goal met? Address performance via corrective action process Establish goals Monitor performance goals Address performance via corrective action process Is new SSC level goal required?
SSCs in preventive maintenance & condition monitoring program Perform appropriate maintenance on SSCs Is performance acceptable? Is new SSC level goal required?
No No No Address performance via corrective action process Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 64 Figure 4-9 NFPA 805 Monitoring - Scoping and Screening
 
4.7 Program Documentation, Configuration Control, and Quality Assurance 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805 In accordance with the requirements and guidance in NFPA 805 Section 2.7.1 and NEI 04-02, ANO-2 has documented analyses to support compliance with 10 CFR 50.48(c). The analyses are being performed in accordance with Entergy's processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analyses.
Function Currently in Maintenance Rule? Component currently in FPRA? High Safety Significance of feature by compartment?
No Yes No No No Fire Protection Systems and Features Fire Protection Programmatic Elements Rad Release Engineered Systems and Features FPRA Components NSEL Components NOP Components NSCA Phase 1 - Sco p in g Phase 2 - High Risk Significance?
Included in Maintenance Rule? Yes Yes Use Maintenance Rule for Monitoring Yes Yes Normal System & Program Health Monitoring Process or Outage Risk Management for NPO NFPA 805 Specific Monitoring Process establish targets for reliability/unavailability in Phase 3 No *Fully describe process used*
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 65 Analyses, as defined by NFPA 805 Section 2.4, performed to demonstrate compliance with 10 CFR 50.48(c) will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy. Note these analyses do not include items such as periodic tests, hot work permits, fire impairments, etc.
 
The Fire Protection Design Basis Document described in Section 2.7.1.2 of NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 will be created as part of transition to 10 CFR 50.48(c) to ensure program implementation following receipt of the safety evaluation. Appropriate cross references will be established to supporting documents as required by Entergy processes (see Attachment S). Figure 4-10 provides an example of the post-transition documents and their relationships.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 66 Figure 4-10 NFPA 805 Planned Post-Transition Documents and Relationships
 
Non-Power Equipment and Data PRA Equipment and Data NSEL Comp Cables NFPA 805 DOCUMENTS NSCA Databases NSCA CALCULATION NSCA SUPPORTING INFO Comp & Cable Method/Results FA Assessment Method/Results SSA Drawings MSO and OMA Treatments Coordination Calculations / MHF Manual Action Feasibility T-H Calculations B-2 Table B-3 Table Plant DBDs that support NSCA Non-Power Mode NSCA Treatment Non-Power Operations Calculations NFPA 805 FIRE RISK EVALUATIONS Fire Risk Evaluation Calculation(s) FHA DATABASE DATA Ignition Sources &
Scenarios FP System and Features Data Inventory of Hazards B-1 Table              Detailed Data FHA SUPPORT DOCUMENTATION FP Systems Code Compliance Evaluations FP Drawings FP System and Feature DBDs Engineering Equivalency Evaluations Radioactive Release Review Calculation Fire Pre-Plans Revised License Condition Revised UFSAR FIRE SAFETY ANALYSIS (DBD)
* On a Fire Area Basis
* On a Generic Basis
* Fire Area Description
* FHA Database Information
* Nuclear Safety Performance Criteria Compliance Summary (NEI 04-02 B-3 Table Results)
* Non-Power Evaluation Results Summary
* Radioactive Release Summary
* B-1 Table Results
* Radioactive Release (Training)
* Monitoring Program Fire PRA Bold text indicates new NFPA 805 documents
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 67 4.7.2 Compliance with Configuration Control Requirements in Section 2.2.9 and 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 Entergy 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), A.2.4.4, D.5 Change Evaluation NEI 04-02 5.3, Appendix B, Appendix I, Appendix J Change Evaluation, Change Evaluation Forms (Appendix I) RG 1.205 C.2.2.4, C.3.1, C.3.2, C.4.3 Risk Evaluation, Standard License Condition, Change Evaluation Process, FPRA
 
The Plant Change Evaluation Process consists of the following 4 steps and is depicted in Figure 4-11:  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 Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 68 bases. This screening process is modeled after the NEI 02-03 process. This process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).
The characteristics of an acceptable screening process that meets the "assessment of the acceptability of risk" requirement of Section 2.4.4 of NFPA 805 are:  The quality of the screen is sufficient to ensure that potentially greater than minimal risk increases receive detailed risk assessments appropriate to the level of risk. The screening process must be documented and be available for inspection by the NRC. The screening process does not pose undue evaluation or maintenance burden.
If any of the above is not met, proceed to the Risk Evaluation step.
 
Risk Evaluation The screening is followed by engineering evaluati ons 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 CDF (change in core damage frequency) and LERF (change in large early release frequency) criteria from the license condition. The proposed changes are also assessed to ensure consistency with the defense-in-depth philosophy and that sufficient safety margins were maintained.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 69 Figure 4-11 Plant Change Evaluation [NEI 04-02 Figure 5-1] Note references in Figure refer to NEI 04-02 Sections
 
Identify / Define Change Complies with    Chap 3 or previously approved Alternative?
Involve NFPA Chapter 3 Rqmnt? Defining the Change (5.3.2)
License Amendment Request NOT Required License Amendment Request Preliminary Risk Screening (5.3.3) Yes Yes No No No Is the change trivial? B Yes Does change impact Risk greater than minimally?
B Yes No Yes Initial Evaluation Screens?  (5.3.4.1 & 2)
C No Detailed Evaluation (5.3.4.3) CDF & LERF OK? B C No DID and Safety Margin (SM) OK? No Yes Yes Change Not Acceptable Document Conclusion Fire PRA Capability Category Assessment PRA Capability Category Assessment Risk Evaluation (5.3.4) Acceptance Criteria (5.3.5)
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 70 The ANO-2 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 FPP license basis Reviews will be utilized to maintain configuration control of the FPP documents. The configuration control procedures which govern the various ANO-2 documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements (see Attachment S).
 
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 (see Attachment S).
The process for capturing the impact of proposed changes to the plant on the FPP 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 FPP 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 FPP impacts will be sent to qualified individuals (Fire Protection, Safe Shutdown/NSCA, PRA) to ascertain the program impacts, if any. If FPP 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, or  Performance-Based Approach:  Utilize the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the ANO-2 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 to 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.
4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805
 
Fire Protection Program Quality During the transition to 10 CFR 50.48(c), ANO-2 performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805. ANO-2 will maintain the existing Fire Protection Quality Assurance program.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 71 Fire PRA Quality
 
Configuration control of the FPRA model will be maintained by integrating the FPRA model into the existing processes used to ensure configuration control of the internal events PRA model. This process complies with Section 5 of the ASME Standard for PRA Quality and ensures that Entergy maintains an as-built, as-operated PRA model of the plant. The process has been peer reviewed. Quality assurance of the FPRA is assured via the same processes applied to the
 
internal events model.
 
This process follows the guidance outlined in RG 1.174, which requires the use of qualified individuals, procedures that require calculations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. Although the entire scope of the formal 10 CFR 50, Appendix B, program is not applied to the PRA models or processes in general, often parts of the program are applied as a convenient method of complying with the requirements of RG 1.174. For example, the procedure which addresses independent review of calculations for 10 CFR 50, Appendix B, is applied to the PRA model calculations, as well.
 
With respect to quality assurance (QA) requirements for independent reviews of calculations and evaluations, those existing requirements for FPP documents will remain unchanged. Entergy specifically requires that the calculations and evaluations in support of the NFPA 805
 
LAR, exclusive of the FPRA, be performed within the scope of the QA program, which requires independent review as defined by Entergy procedures. As recommended by NUREG/CR-6850, the sources of uncertainty in the FPRA were identified and analyzed for sensitivity in support of the transition to NFPA 805.
The removal of conservatism inherent in the FPRA remains a long-term goal; nevertheless, the FPRA results were deemed sufficient for evaluating the risk associated with this application.
While Entergy continues to strive toward a more "realistic" estimate of fire risk, use of mean values continues to be the best estimate of fire risk. During the FRE process, the uncertainty and sensitivity associated with specific FPRA parameters were considerations in the evaluation of the change in risk relative to the applicable acceptance thresholds.
 
Specific Requirements of NFPA 805, Section 2.7.3
 
NFPA 805, Section 2.7.3.1 - Review Analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) are performed in accordance with Entergy procedures that require independent review.
 
NFPA 805, Section 2.7.3.2 - Verification and Validation Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805.
 
NFPA 805, Section 2.7.3.3 - Limitations of Use
 
Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) are used and were used appropriately as required by Section 2.7.3.3 of
 
NFPA 805.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 72 NFPA 805, Section 2.7.3.4 - Qualification of Users
 
Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.
 
During the transition to 10 CFR 50.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g., fire modeling) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805 Section 2.7.3.4.
 
Post-transition, for personnel performing fire modeling or FPRA development and evaluation, Entergy will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805, Section 2.7.3.4, to perform assigned work (see 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 FPRA 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 Zone:  Fire Area/Zone Identifier.
 
== Description:==
Fire Area/Zone Description. NFPA 805 Regulatory Basis:  Post-transition NFPA 805 Chapter 4 compliance basis (Note:  Compliance is determined on a Fire Area basis; therefore, a compliance basis is not provided for individual fire zones.)  Required Suppression/Detection:  Detection/suppression is required in the Fire Area based on NFPA 805 Chapter 4 compliance. The information is provided on a zone basis. The basis for the requirement of the fire protection system is designated as follows: S - Separation Criteria: Systems required for Chapter 4 Separation Criteria (NFPA 805, Section 4.2.3) E - EEEE: Systems required for acceptability of Existing Engineering Equivalency Evaluations (EEEEs) (NFPA 805, Section 2.2.7) L - LA Criteria: NRC approved Licensing Action (LA) (i.e., Exemptions/
Deviations/Safety Evaluations) (NFPA 805, Section 2.2.7) R - Risk Criteria: Systems required to meet the Risk Criteria for the Performance-Based Approach (NFPA 805, Section 4.2.4)
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 73 D - Defense-in-depth Systems required to maintain adequate balance of Defense-in- Criteria: Depth for a Performance-Based Approach (NFPA 805, Section 4.2.4)
Attachment W contains the results of the FREs, additional risk of recovery actions, and the change in risk on a fire area basis.
 
4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase
 
Planned modifications, program, procedure, and evaluation changes and upgrades to comply with NFPA 805 are described in Attachment S. Attachment S contains two tables. Table S-1 identifies plant modifications required to be completed and Table S-2 identifies programs, procedures, and document changes and upgrades to be completed. 
 
The Plant Change Evaluation Process will be implemented using Entergy fleet procedures EN-DC-115, Engineering Change Process, and EN-DC-128, Fire Protection Impact Reviews.
EN-DC-115, which is used to evaluate proposed plant changes, includes steps to ensure proposed activities such as changes in pump motors, cabling, transient combustibles, fire loading, and ignition sources are evaluated for impact to the FPRA. Based on the results of the EN-DC-115 impact review, additional reviews may be required in accordance with EN-DC-128. Guidance will be provided in EN-DC-128 to define changes in relationship to NFPA 805 Chapter 2 or 3 requirements, changes in Radioactive Release performance criteria, NSCA capability in NPO modes, power operations credited NSCA SSCs, changes in combustible loading, changes that may adversely impact fire areas with suppression or detection systems, etc. Based on the type of change, EN-DC-115 will also include a preliminary risk screening that will be followed by either a qualitative or quantitative review. The qualitative and quantitative reviews will include questions to determine if NRC approval is required prior to making the plant
 
change.
 
The FPRA model will represent the as-built, as-operated and maintained plant following completion of the risk related modifications that are implemented in support of those modifications identified in Attachment S. Following installation of modifications and the as-built installation details, additional refinements surrounding the modification may need to be incorporated into the Fire PRA model. However, these changes are not expected to be significant. See Implementation Item 9 in Table S-2 of Attachment S.
 
4.8.3 Supplemental Information - Other Licensee Specific Issues
 
4.8.3.1 The overall CDF shown in Table W-2 (6.47E-05) reflects the fire risk for ANO-2 post transition to NFPA-805. The total risk associated with the current plant will be addressed by the continuing compensatory measures currently applied for the existing Appendix R open items. In general, the reduced number of feasible operator manual actions (due to plant modifications completed since 2005), and the on-going mixture of roving and continuous fire watches will be maintained until the proposed modifications are implemented to remove the need for these interim compensatory measures. As noted in Table W-2 the primary contributor to the overall plant risk is Fire Area G, where the Alternate Shutdown measures for Appendix R are concentrated. Existing plant procedures for post-fire safe shutdown focus on this area. Operations personnel are currently trained and able to promptly respond to any postulated fire in this critical area, now well defined by the NFPA 805 risk analysis.
 
Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 74 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D) Required Detection System (S, L, E, R, D) Required Fire Protection Feature (S, L, E, R, D) Required Fire Protection Feature and System Details 2MH01E  Between Aux Bldg and Intake Structure 2MH01E Concrete Manhole East 4.2.4.2 None None N/A  2MH01W  Between Aux Bldg and Intake Structure  2MH01W Concrete Manhole West 4.2.3.2 None None N/A  2MH02E  Between Aux Bldg and Intake Structure  2MH02E Concrete Manhole East 4.2.4.2 None None N/A  2MH02W  Between Aux Bldg and Intake Structure  2MH02W Concrete Manhole West 4.2.3.2 None None N/A  2MH03E  Between Aux Bldg and Intake Structure  2MH03E Concrete Manhole East 4.2.4.2 None None N/A  2MH03W  Between Aux Bldg and Intake Structure  2MH03W Concrete Manhole West 4.2.3.2 None None N/A AA  "B" HPSI, LPSI and Containment Spray Pump Room and Gallery  2007-LL "B" HPSI, LPSI and Containment Spray Pump Room and Gallery 4.2.4.2 E E, R, D N/A Detection and Partial Suppression AAC  Alternate AC Diesel  2MH12 Manhole near SBO diesel 4.2.3.2 None None N/A  SBOD Alternate AC Diesel 4.2.3.2 N/R N/R N/A ADMIN  Administration Building  ADMIN Administration Building 4.2.3.2 None N/R N/A B-2  Unit 2 General Plant Multiple Elevations  2045-XX Turbine Lube Oil Storage Tank Room 4.2.4.2 N/R None N/A  2078-QQ Heat Exchanger Equipment Room 4.2.4.2 None None N/A  2092-PP Chiller Water System Equipment Room 4.2.4.2 None None N/A  2147-A Chemical Storage Room 4.2.4.2 E None N/A Suppression  2148-A Corridor 4.2.4.2 None None N/A  2151-A Fuel Handling Room (El. 404) 4.2.4.2 None E, R, D N/A Detection Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 75 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D) Required Detection System (S, L, E, R, D) Required Fire Protection Feature (S, L, E, R, D) Required Fire Protection Feature and System Details  2152-D Computer Room 4.2.4.2 None E, R, D N/A Detection  2153-A Ventilation Equipment Room 4.2.4.2 None E, R, D N/A Partial Detection  2155-A Steam Pipe Room 4.2.4.2 None None N/A  2156-A Containment Purge Air Equipment Room 4.2.4.2 None R, D N/A Detection  2172-ZZ Storage and Shop Room 4.2.4.2 N/R None N/A  2177-YY Neutralizer Tank Room 4.2.4.2 None None N/A  2178-AAA Lube Oil Reservoir 4.2.4.2 R None N/A Partial Suppression  2200-MM Turbine Building 4.2.4.2 E E, R, D N/A Partial Suppression and Detection  2201-B ANO-2 Operations Support Facility 4.2.4.2 None R, D N/A Detection  2223-KK Pipeway Equipment Access Room (Aux. Bldg. Extension) 4.2.4.2 None E, R, D N/A Detection  2225-WW Regenerative Waste Pump & Tank Room 4.2.4.2 None None N/A  2229-SS Storage Room 2232 4.2.4.2 E None N/A Partial Suppression  2230-RR Drum Filling Room 4.2.4.2 None None N/A  2231-TT Plant Heating Boiler Room 4.2.4.2 N/R None N/A  2242-OO H&V Mechanical Equipment Room, AO Shack, Lab & Storage Room 4.2.4.2 N/R R, D N/A Partial Detection  2243-NN Chemistry Lab, Kitchen & Offices 4.2.4.2 N/R None N/A  2261-UU Plant Heating Boiler Day Tank 4.2.4.2 N/R None N/A B-3  North Penetration Areas  2091-BB North Electrical Equipment Room 4.2.4.2 None E, R, D N/A Detection  2112-BB Lower North Electrical Penetration Room 4.2.4.2 E, R E, R, D N/A Suppression and Detection  2183-J Upper North Electrical Penetration Room 4.2.4.2 E, R E, R, D N/A Suppression and Detection B-4  CEDM Room  2154-E CEDM Equipment Room 4.2.4.2 None E, R, D N/A Detection Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 76 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D) Required Detection System (S, L, E, R, D) Required Fire Protection Feature (S, L, E, R, D) Required Fire Protection Feature and System Details B-5  North and South Aux Bldg Stair  2149-B Stairwell No. 2001 4.2.3.2 None None N/A  2158-F Stairwell No. 2055 4.2.3.2 None None N/A B-6  Aux Bldg General Access Area, A & C Pump Rooms  2006-LL General Access Room 4.2.4.2 E E, R, D N/A Detection and Partial Suppression  2010-LL "C" HPSI Pump Room 4.2.4.2 None E, R, D N/A Detection  2011-LL Tendon Gallery Access 4.2.4.2 None E, R, D N/A Detection  2014-LL "A" HPSI, LPSI, & Containment Spray Pump Room 4.2.4.2 None E, R, D N/A Detection CC  Emergency Feedwater Pump Room (Turbine Driven)  2024-JJ Emergency Feedwater Pump Room (Turbine Driven) 4.2.3.2 None E N/A Detection DD  Unit 2 General Area 335' Elevation  2019-JJ Boric Acid Condensate Tank Room 4.2.4.2 None None N/A  2032-JJ Spent Resin Storage Tank Room 4.2.4.2 None None N/A  2040-JJ Corridor 4.2.4.2 None E, R, D N/A Partial Detection  2068-DD Hot Machine Shop 4.2.4.2 None E, R, D N/A Detection EE-L  South Piping Penetration Rooms  2055-JJ Lower South Piping Penetration Room 4.2.4.2 None E, R, D N/A Detection  2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room 4.2.4.2 None E, R, D N/A Partial Detection EE-U  Lower South Electrical Penetration  2111-T Lower South Electrical Penetration Room 4.2.4.2 E E, R, D N/A Suppression and Detection FF  Emergency Feedwater Pump Room (Motor Driven)  2025-JJ Emergency Feedwater Pump Room (Motor Driven) 4.2.3.2 None E N/A Detection Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 77 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D) Required Detection System (S, L, E, R, D) Required Fire Protection Feature (S, L, E, R, D) Required Fire Protection Feature and System Details G  Unit 2 Alternate Shutdown Areas  97-R Cable Spreading Room 4.2.4.2 R, D R, D N/A Detection and Partial Suppression  129-F Control Room 4.2.4.2 R, D R, D N/A Detection and Partial Suppression  2098-C CPC Room 4.2.4.2 E, R, D E, R, D N/A Detection and Suppression  2098-L Cable Spreading Room 4.2.4.2 E, D E, R, D N/A Detection and Suppression  2119-H CR Printer Room 4.2.4.2 None E, R, D N/A Detection  2136-I Health Physics Corridor 4.2.4.2 E, R, D E, R, D N/A Detection and Partial Suppression  2137-I USEP Room, Decon, Hot Instrument Shop4.2.4.2 E, R, D E, R, D N/A Detection and Suppression  2150-C Old CPC Room 4.2.4.2 None E, R, D N/A Detection  2199-G Unit 2 Control Room 4.2.4.2 None E, R, D N/A Detection GG  Unit 2 North Electrical and Piping Penetration Area  2076-HH Electrical Equipment Room 4.2.4.2 None E, R, D N/A Detection  2081-HH Upper North and Lower North Piping Penetration Room 4.2.4.2 None E, R, D N/A Detection HH  Unit 2 General Area 354' Elevation  2063-DD Sample Room 4.2.4.2 None None N/A  2072-R Upper Volume Control Tank Room, Lower Tank and Pump Room 4.2.4.2 None E, R, D N/A Detection  2073-DD Access Room, Pump Room, Tank Room (2B62 & Resin Addition Room) 4.2.4.2 E, R E, R, D N/A Detection and Partial Suppression  2096-M Motor Control Center (2B63) 4.2.4.2 None E, R, D N/A Detection  2106-R Degasifier Vacuum Pump Room 4.2.4.2 None E, R, D N/A Detection  2107-N Corridor (North of Stairway 2001) 4.2.4.2 None E, R, D N/A Detection Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 78 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D) Required Detection System (S, L, E, R, D) Required Fire Protection Feature (S, L, E, R, D) Required Fire Protection Feature and System Details II  North Switchgear Room  2101-AA North Switchgear Room 4.2.4.2 None R, D N/A Detection JJ  Corridor  2109-U Corridor 4.2.4.2 E, R, D E, R, D N/A Detection and Partial Suppression K  Tank Rooms  16-Y Clean Waste Receiver Tank Room 4.2.3.2 None None N/A  2020-JJ Boron Holdup Tank Vault 4.2.3.2 None None N/A KK  Unit 2 South Emergency Diesel Generator and Boric Acid Makeup Tank Rooms  2093-P South Emergency Diesel Generator Room 4.2.4.2 E, R E, R, D N/A Detection and Partial Suppression  2114-I EDG Air Intake Room 4.2.4.2 None E, R, D N/A Detection  2115-I Boric Acid Makeup Tank Room 4.2.4.2 None E, R, D N/A Detection L  Diesel Fuel Storage Vault Area  TKVLT Diesel Fuel Storage Vault 4.2.3.2 N/R N/R N/A MM  West Battery and DC Equipment Rooms  2099-W West D.C. Equipment Room 4.2.4.2 None E, R, D N/A Detection  2103-V West Battery Room 4.2.4.2 None R, D N/A Detection NN  Unit 2 Containment Building  2032-K Containment Building South Side 4.2.4.2 E E, R, D N/A Partial Suppression and Detection  2033-K Containment Building North Side 4.2.4.2 E E, R, D N/A Partial Suppression and Detection OO  Unit 2 Intake Structure  INTAKE Intake Structure (Unit 2) 4.2.4.2 N/R D N/A Detection Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements Enclosure 1 to 2CAN121202 Page 79 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D) Required Detection System (S, L, E, R, D) Required Fire Protection Feature (S, L, E, R, D) Required Fire Protection Feature and System Details QQ  North Emergency Diesel  2094-Q North Emergency Diesel Generator Room 4.2.3.2 E E N/A Suppression and Detection  2114-I EDG Air Intake Room 4.2.3.2 None E N/A Detection SS  South Switchgear and East DC Equipment and Battery Rooms  2097-X East D.C. Equipment Room 4.2.4.2 None E, R, D N/A Detection  2100-Z South Switchgear Room 4.2.4.2 None R, D N/A Detection  2102-Y East Battery Room 4.2.4.2 None E, R, D N/A Detection TT  Electrical Equipment (2B9/2B10) Room  2108-S Electrical Equipment (2B9/2B10) Room 4.2.4.2 None E, R, D N/A Detection YD  Miscellaneous Yard Locations  YARD Miscellaneous Yard Locations 4.2.3.2 N/R N/R N/A  Legend: Fire Protection Features are features required to meet NFPA 805 Chapter 3 requirements. S - Credited Separation Criteria is derived from PRA in Attachment C - Table B-3 VFDRs. E - EEEE Criteria:  Credited Systems/Features are derived from Attachment A - Table B-1 and/or Attachment C - Table B-3. L - NRC approved Licensing Action is derived from Attachment K and/or Attachment A - Table B-1 VFDRs. R - Risk Criteria is derived from PRA in Attachment C - Table B-3. D - Defense-In-Depth Criteria is derived from PRA in Attachment C - Table B-3. N/R - System is operational in fire area, however it is Not Required. None - Fire protection feature is not present in the fire zone.
 
Arkansas Nuclear One - Unit 2 5.0 Regulatory Evaluation Enclosure 1 to 2CAN121202 Page 80
 
==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 ANO-2 does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, General Design Criterion (GDC) 3, Fire Protection. The NRC addressed the overall adequacy of the regulations during the promulgation of 10 CFR 50.48(c) (Reference Federal Register (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 CFR 50.48(c)(3)(ii) provides, in pertinent parts, that a licensee intending to adopt the new rule must, among other things, "modify the fire protection plan required by paragraph (a) of that section to reflect the licensee's decision to comply with NFPA 805."
Therefore, to the extent that the contents of the existing 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.
Arkansas Nuclear One - Unit 2 5.0 Regulatory Evaluation Enclosure 1 to 2CAN121202 Page 81 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 ANO-2 FP program and applicable industry and ANO-2 documents that address the topic.
 
10 CFR 50.48(a) Table 5-1 10 CFR 50.48(a) - Applicability/Compliance Reference 10 CFR 50.48(a) Section(s) Applicability/Compliance Reference (1) Each holder of an operating license issued under this 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: See below (i) Describe the overall fire protection program for the facility; NFPA 805 Section 3.2 Attachment A (ii) Identify the various positions within the licensee's organization that are responsible for the program; NFPA 805 Section 3.2.2 Attachment A (iii) State the authorities that are delegated to each of these positions to implement those responsibilities; and NFPA 805 Section 3.2.2 Attachment A (iv) Outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage. NFPA 805 Section 2.7 and Chapters 3 and 4 Attachments A and C (2) The plan must also describe specific features necessary to implement the program described in paragraph (a)(1) of this section such as: See below (i) Administrative controls and personnel requirements for fire prevention and manual fire suppression activities;  NFPA 805 Sections 3.3.1 and 3.4 Attachment A (ii) Automatic and manually operated fire detection and suppression systems; and NFPA 805 Sections 3.5 through 3.10 and Chapter 4 Attachments A and C (iii) The means to limit fire damage to structures, systems, or components important to safety so that the capability to shut down the plant safely is ensured. NFPA 805 Section 3.3 and Chapter 4 Attachment A (3) The licensee shall retain the fire protection plan and each change to the plan as a record until the Commission terminates the reactor license. The licensee shall retain each superseded revision of the procedures for 3 years from the date it was superseded. NFPA 805 Section 2.7.1.1 requires that documentation (Analyses, as defined by NFPA 805, Section 2.4, performed to demonstrate compliance with this standard) be maintained for the life of the plant. OP-1003.014, Revision 6, "ANO Fire Protection Program" and EN-AD-103, Revision 12, "Document Control and Records Management Programs," address the scope and retention standards for ANO-2. (4) Each applicant for a design approval, design 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. Not applicable. ANO-2 is licensed under 10 CFR 50.
 
Arkansas Nuclear One - Unit 2 5.0 Regulatory Evaluation Enclosure 1 to 2CAN121202 Page 82 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 safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. NFPA 805 Chapters 3 and 4 Attachments A and C Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. NFPA 805 Sections 3.3.2, 3.3.3, 3.3.4, 3.11.4 Attachment A Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. NFPA 805 Chapters 3 and 4 Attachments A and C Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components NFPA 805 Sections 3.4 through 3.10 and 4.2.1 Attachment C 10 CFR 50.48(c) Table 5-3 10 CFR 50.48(c) - Applicability/Compliance Reference 10 CFR 50.48(c) Section(s) Applicability/Compliance Reference (1) Approval of incorporation by reference. National Fire Protection Association (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51. General Information. NFPA 805 2001 edition is the edition used. (2) Exceptions, modifications, and supplementation of NFPA 805. As used in this section, references to NFPA 805 are to the 2001 Edition, with the following exceptions, modifications, and supplementation:  General Information. NFPA 805 2001 edition is the edition used. (i) Life Safety Goal, Objectives, and Criteria. The Life Safety Goal, Objectives, and Criteria of Chapter 1 are not endorsed. The Life Safety Goal, 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 Plant Damage/Business Interruption Goal, Objectives, and Criteria of Chapter 1 are not endorsed. The Plant Damage/Business Interruption Goal, Objectives, 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 criteria of Sections 1.5.1(b) and (c), a high-pressure charging/injection pump coupled with the Pressurizer Emergency Core Cooling System (ECCS) vent valves as the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for pressurized-water reactors (PWRs) is not permitted. Feed and bleed is not utilized as the sole fire-protected safe shutdown methodology. (iv) Uncertainty analysis. An uncertainty analysis performed in accordance with Section 2.7.3.5 is not required to support deterministic approach calculations. Uncertainty analysis was not performed for deterministic methodology.
Arkansas Nuclear One - Unit 2 5.0 Regulatory Evaluation Enclosure 1 to 2CAN121202 Page 83 Table 5-3 10 CFR 50.48(c) - Applicability/Compliance Reference 10 CFR 50.48(c) Section(s) Applicability/Compliance Reference (v) Existing cables. In lieu of installing cables meeting flame propagation tests as required by Section 3.3.5.3, a flame-retardant coating may be applied to the electric cables, or an automatic fixed fire suppression system may be installed to provide an equivalent level of protection. In addition, the italicized exception to Section 3.3.5.3 is not endorsed. Electrical cable construction complies with a flame propagation test that was found acceptable to the NRC as documented in Attachment A. (vi) Water supply and distribution. The italicized exception to Section 3.6.4 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 must submit a request for a license amendment in accordance with paragraph (c)(2)(vii) of this section. ANO-2 complies by previous NRC approval. See Attachment A. (vii) Performance-based methods. Notwithstanding the prohibition in Section 3.1 against the use of performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under &sect; 50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach; (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). The use of performance-based methods for NFPA 805 Chapter 3 is requested. See Attachment L.
(3) Compliance with NFPA 805. See below (i) A licensee may maintain a fire protection program that complies with NFPA 805 as an alternative to complying with paragraph (b) of this section for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979. The licensee shall submit a request to comply with NFPA 805 in the form of an application for license amendment under &sect; 50.90. The application must identify any orders and license conditions that must be revised or superseded, and contain any necessary revisions to the plant's technical 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. The LAR was submitted in accordance with 10 CFR 50.90. The LAR included applicable license conditions, orders, technical specifications/bases that needed to be revised and/or superseded. (ii) The licensee shall complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan required by paragraph (a) of this section to reflect the licensee's decision to comply with NFPA 805, before changing its fire protection program or nuclear power plant as permitted by NFPA 805. The LAR and transition report summarize the evaluations and analyses performed in accordance with Chapter 2 of NFPA 805.
Arkansas Nuclear One - Unit 2 5.0 Regulatory Evaluation Enclosure 1 to 2CAN121202 Page 84 Table 5-3 10 CFR 50.48(c) - Applicability/Compliance Reference 10 CFR 50.48(c) Section(s) Applicability/Compliance Reference (4) Risk-informed or performance-based alternatives to compliance with NFPA 805. A licensee may submit a request to use risk-informed or performance-based alternatives to compliance with NFPA 805. The request must be in the form of an application for license amendment under &sect; 50.90 of this chapter. The Director of the Office of Nuclear Reactor Regulation, or designee of the Director, may approve the application if the Director or 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).
No risk-informed or performance-based alternatives to compliance with NFPA 805 (per 10 CFR 50.48(c)(4)) were utilized.
5.2 Regulatory Topics
 
5.2.1 License Condition Changes
 
The current ANO-2 fire protection license conditions 2.C.(3)(b) and 2.C.(3)(e) are being replaced consistent with the standard license condition in Regulatory Position 3.1 of RG 1.205, as shown in Attachment M.
 
5.2.2 Technical Specifications
 
ANO-2 conducted a review of the Technical Specifications (TSs) and TS Bases to determine which TSs and/or TS Bases are required to be revised, deleted, or superseded. ANO-2 determined that the changes to the TSs and applicable justification listed in Attachment N are adequate for the ANO-2 adoption of the new FP licensing basis.
 
5.2.3 Orders and Exemptions
 
A review was conducted of the ANO-2 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 O.
 
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 Arkansas Nuclear One - Unit 2 5.0 Regulatory Evaluation Enclosure 1 to 2CAN121202 Page 85 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. ANO-2 has evaluated the proposed amendment and determined that it involves no significant hazards consideration.
5.3.2 Environmental Consideration
 
Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c). That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.
 
5.4 Revision to the SAR
 
After the approval of the LAR and in accordance with 10 CFR 50.71(e), the ANO-2 SAR will be revised. The format and content will be consistent with NEI 04-02, as addressed in FAQ 12-0062.
 
5.5 Transition Implementation Schedule
 
The following schedule for transitioning ANO-2 to the new FP licensing basis requires NRC approval of the LAR in accordance with the following schedule (see Enclosure 4):  Implementation of new NFPA 805 FP program provide in Attachment S, Table S-2, which includes procedure changes, process updates, and training of affected plant personnel, will occur six months following SER issuance. Modifications required to support and complete the ANO-2 transition to NFPA 805 as provided in Attachment S, Table S-1, will be completed prior to startup from the second ANO-2 refueling outage following SER issuance. Appropriate compensatory measures will be maintained until modifications are complete.
 
Arkansas Nuclear One - Unit 2 6.0 References Enclosure 1 to 2CAN121202 Page 86
 
==6.0 REFERENCES==
 
The following references were used in the development of the Transition Report (TR). Additional references may be found in the NEI 04-02 Tables in the various Attachments.
: 1. 10 CFR 50.48, Fire Protection, 65 FR 38190, June 20, 2000; [69 FR 33550, June 16, 2004; 72 FR 49495, Aug. 28, 2007]
: 2. National Fire Protection Associated (NFPA) Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)
: 3. 0CAN110502, Entergy letter to the NRC dated November 2, 2005, Letter of Intent to Adopt NFPA 805 - Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants, 2001 Edition
[ADAMS Accession No. ML053140128]
: 4. NUREG/CR-6850, EPRI / NRC-RES Fire PRA Methodology for Nuclear Power Facilities , September 2005
: 5. NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2, Nuclear Energy Institute, Washington, DC, April 2008 [ADAMS Accession No. ML081130188]
: 6. Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, Revision 1, December 2009
: 7. 10 CFR 50.82, Termination of License ,  [61 FR 39301, July 29, 1996, as amended at 62 FR 39091, July 21, 1997; 68 FR 19727, Apr. 22, 2003]
: 8. 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 [45 FR 76611, Nov. 19, 1980; 46 FR 44735, Sept. 8, 1981, as amended at 53 FR 19251, May 27, 1988; 65 FR 38191, June 20, 2000; 77 FR 39907, Jul. 6, 2012]
: 9. 0CNA120805, NRC letter to Entergy dated December 22, 2008, Arkansas Nuclear One, Units 1 and 2 - Evaluation of the Request for an Extension of Enforcement Discretion in Accordance With the Interim Enforcement Policy for Fire Protection Issues During Transition to National Fire Protection Standard NFPA 805 , [ADAMS Accession No.
ML083500404]
: 10. 2CAN031201, Entergy letter to the NRC dated March 27, 2012, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) [ADAMS Accession No. ML12087A113]
: 11. 2CNA091201, NRC letter to Entergy dated September 7, 2012, Arkansas Nuclear One, Unit 2 - Non-Acceptance of Request to Adopt NFPA 805, "Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants,"2001 Edition (TAC NO. ME8282), [ADAMS Accession No. ML12208A196]
: 12. 0CNA071107, NRC letter to Entergy dated July 28, 2011, Arkansas Nuclear one, Units 1 and 2 - Commitment to Submit a License Amendment Request to Transition to 10CFR 50.4(c), National Fire Protection Association Standard NFPA 805, and Request to Extend Enforcement Discretion, { ADAMS Accession No. ML112030193]
 
Arkansas Nuclear One - Unit 2 6.0 References Enclosure 1 to 2CAN121202 Page 87
: 13. Appendix A to APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976 (August 23, 1976)
: 14. 0CAN027802, NRC letter to Entergy dated February 28, 1978, Fire Protection Safety Evaluation Report, NUREG-0223
: 15. 0CNA039215, NRC letter to Entergy dated March 31, 1992, Issuance of Amendment Nos. 158 and 132 to Facility Operating License Nos. DPR-51 and NPF Arkansas Nuclear One, Units 1 and 2
: 16. NUREG-0223, 2CNA087812, NRC letter to Entergy dated August 30, 1978, Issuance of Fire Protection Safety Evaluation Report for Arkansas Nuclear One, Unit 2
: 17. 0CNA088203, NRC letter to Entergy dated August 6, 1982, Results of Inspection Conducted June 21-25, 1982
: 18. 0CNA038328, NRC letter to Entergy dated March 22, 1983, Issuance of Exemptions to Certain Requirements of Appendix R to 10 CFR 50
: 19. 0CNA078522, NRC letter to Entergy dated July 25, 1985, Results of Inspection Conducted May 20-24, 1985
: 20. 2CAN058009, Entergy letter to the NRC dated May 12, 1980, Fire Protection Items 3.4, 3.8, 3.9 and 3.18
: 21. NFPA-15, Standard for Water Spray Fixed Systems for Fire Protection, May 1977
: 22. 0CAN078202, Entergy letter to the NRC dated July 1, 1982, Results of Appendix R Compliance Review
: 23. 0CAN067803, Entergy letter to the NRC dated June 15, 1978, Fire Protection
: 24. 0CNA068503, NRC letter to Entergy dated June 10, 1985, Results of Inspections April 1-20, 1985
: 25. 0CAN088404, Entergy letter to the NRC dated August 15, 1984, Results of Reanalysis Against NRC Clarification/Interpretation of Appendix R to 10CFR50
: 26. 2CNA108802, NRC letter to Entergy dated October 26, 1988, Evaluation of Exemptions from the Technical Requirements of Appendix R to 10 CFR Part 50
: 27. Generic Letter 86-10, Implementation of Fire Protection Requirements (April 24, 1986)
: 28. 0CNA058307, NRC letter to Entergy dated May 6, 1983, Results of Inspection April 18-22, 1983  29. 0CAN118210, Entergy letter to the NRC dated November 11, 1982, Request for Additional Information to Appendix R Compliance Submittal
: 30. 0CAN088508, Entergy letter to the NRC dated August 30, 1985, Results of Reanalysis Against NRC Clarification / Interpretation of Appendix R to 10CFR50 - Supplemental Information
: 31. 2CAN078313. Entergy letter to the NRC dated July 27, 1983, Fire Protection Open Items
 
Arkansas Nuclear One - Unit 2 6.0 References Enclosure 1 to 2CAN121202 Page 88
: 32. 0CNA058316, NRC letter to Entergy dated May 13, 1983, Safety Evaluation Regarding Safe Shutdown Capability in the Event of a Fire
: 33. 0CNA098716, NRC letter to Entergy dated September 30, 1987, NRC Inspection Report
: 34. 0CAN108710, Entergy letter to the NRC dated October 29, 1987, Request for Exemption to Section III.G.2 of Appendix R
: 35. 2CAN129612, Entergy letter to the NRC dated December 23, 1996, 10 CFR 50 Appendix R Exemption Request - Reactor Coolant Pump Lube Oil System
: 36. 2CNA069701, NRC letter to Entergy dated June 6, 1997, Exemptions to Certain Requirements of 10CFR Part 50, Appendix R Section III.0 Arkansas Nuclear One, Unit 2
: 37. 2CAN109703, Entergy letter to the NRC dated October 8, 1997, 10CFR50 Appendix R Exemption Request from Section III.G.2 for the Intake Structure
: 38. 2CAN029905, Entergy letter to the NRC dated February 25, 1999, 10CFR50 Appendix R Exemption Request from Section III.G.2 for the Intake Structure
: 39. 2CNA109902, NRC letter to Entergy dated October 1, 1999, Issuance of Exemption from the Requirements of 10 CFR 50, Appendix R, Section III.G.2.c for Arkansas Nuclear One, Unit No. 2
: 40. NRC letter to NEI dated July 12, 2006, Process for Frequently Asked Questions for Title 10 of the Code of Federal Regulations, Part 50.48(c) Transitions, [ADAMS Accession No.
ML06166010]
: 41. Regulatory Issue Summary (RIS) 2007-19, Process for Communicating Clarifications of Staff Positions Provided in Regulatory Guide 1.205 Concerning Issues Identified During the Pilot Application of National Fire Protection Association Standard 805 (August 20, 2007), [ADAMS Accession No. ML071590227]
: 42. Engineering Change (EC)-6960, NFPA 805 Transition Fundamental Elements Table B-1 , Revision 0
: 43. NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection , 1969 Edition, Rev. 0
: 44. NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, January 2005, Rev. 1
: 45. CALC-ANO2-FP-09-00032, ANO-2 Transition NSCA Methodology , Rev. 1  46. FAQ 07-0039, Lessons Learned - NEI 04-02 B-2 and B-3 Tables, Rev. 2, NRC Closure Memo [ADAMS Accession No. ML091320068]
: 47. EC-40607, NEI 00-01Section 3, Rev. 1 to Rev. 2 Gap Analysis for NFPA 805 LAR
: 48. EC-15217, Current Transformer (CT) Open Circuit Concerns
: 49. CALC-85-E-0087-24, Safe Shutdown Cable Analysis
: 50. Upper Level Document ULD-0-TOP-12, ANO Unit 1 and 2 Electrical Protection/Coordination
 
Arkansas Nuclear One - Unit 2 6.0 References Enclosure 1 to 2CAN121202 Page 89
: 51. NUREG-0800, Section 9.5.1, Fire Protection Program, Rev. 5  52. ANO-2 Safety Analysis Report (SAR) Section 9.2.6.1, Condensate Storage and Transfer System Design Basis, Amendment 24
: 53. Information Notice (IN) 84-09, Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10 CFR 50, Appendix R), Rev. 1
: 54. RG 1.189, Fire Protection for Operating Nuclear Power Plants , Rev. 2  55. FAQ 07-0030, Establishing Recovery Actions, NRC Closure Memo [ADAMS Accession No. ML110070485]
: 56. EC-27716, ANO2 Fire Area Risk Evaluations for Transition to NFPA-805
: 57. FAQ 07-0038, Lessons Learned on Multiple Spurious Operations Closure Memo
[ADAMS Accession No. ML110140242]
: 58. FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, Revision 1, [ADAM Accession No. ML110140183]
: 59. FAQ 10-0059, NFPA 805 Monitoring, Rev. 5, NRC Closure Memo [ADAMS Accession No.
ML120750108]
: 60. EN-LI-102, Corrective Action Process , Rev. 20
: 61. EC-31053, NFPA 805 Existing Engineering Evaluation Transition , Rev. 0  62. FAQ 07-0040, NON-Power Operations Clarifications Closure Memo [ADAMS Accession No. ML082200528]
: 63. CALC-09-E-0008-02, ANO-2 NFPA 805 Non-Power Operations Assessment, Rev. 0  64. CALC-09-E-0008-04, ANO-2 NFPA 804 NPO Fault Tree and PID Attachments , Rev. 0  65. CALC-08-E-0016-01, Fire Probabilistic Risk Assessment Plant Partitioning and Fire Ignition Frequency Development, Rev. 0  66. CALC-ANO2-FP-08-00001, NEI 04-02 Table G-1, Radioactive Release Transition Report, Rev. 2  67. PFP-U2, ANO-2 Pre-Fire Plan, Rev. 11
: 68. PFP-UC, Support Facilities Pre-Fire Plan, Rev. 13
: 69. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, March 2009 [ADAMS Accession No.
ML090410014] [[RG 1.200, Revision 1, January 2007 - ADAMS Accession No.
ML070240001; Clarification to RG 1.200, Revision 1, July 2007 - ADAMS Accession No.
ML071940235; Draft RG 1.200, Revision 1, was issued as DG-1161, September 2006 -
ADAMS Accession No. ML062480134; RG 1.200, Revision 0, February 2004 - ADAMS Accession No. ML040630078; RG 1.200, Revision 0, was issued for trial use with SRP Chapter 19.1 - ADAMS Accession No. ML040630300; Draft RG 1.200, Revision 0, was issued as DG-1122, November 2002 - ADAMS Accession No. ML023360076]]
 
Arkansas Nuclear One - Unit 2 6.0 References Enclosure 1 to 2CAN121202 Page 90
: 70. ASME/ANS RA-Sa-2009, "Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers and the American Nuclear Society, La Grange Park, IL, 2009
: 71. PRA-ES-01-003, Compilation of Generic Reliability Data and Component Boundaries for Probabilistic Safety Assessment , Rev. 0  72. NUREG/CR-6850, "EPRI/NRC-RES, Fire PRA Methodology for Nuclear Power Facilities," Volumes 1 and 2, U. S. Nuclear Regulatory Commission, Washington, DC, September
 
2005 [ADAMS Accession Nos. ML052580075 (Volume 1) and ML052580118 (Volume 2)]
: 73. EN-DC-204, Maintenance Rule Scope and Basis, Rev. 2  74. EN-DC-205, Maintenance Rule Monitoring , Rev. 4  75. EN-DC-207, Maintenance Rule Periodic Assessment, Rev. 2  76. EN-DC-203, Maintenance Rule Program, Rev. 1  77. EN-DC-143, Program and Component Health Reports Supplemental Guidance, Rev. 2
: 78. EPRI TR 1006756, Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features - Not included in portal
: 79. RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, U. S. Nuclear Regulatory Commission, Washington, DC, May 2011 [ADAMS Accession No.
ML023240437]
: 80. NEI 02-03, Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program, June 2003
: 81. Voluntary Adoption of NFPA 805 as RI-PB Alternative, FR Notice 69 FR 33536 [ML041300697, June 16, 2004]
: 82. FAQ 07-0032, Clarification of 10 CFR 50.48(c), 10 CFR 50.48(a) and GDC 3 Clarification, Rev. 2, NRC Closure Memo [ADAMS Accession No. ML081400292]
: 83. OP-1003.014, ANO Fire Protection Program, Rev. 6 
: 84. EN-AD-103, Document Control and Records Management , Rev. 13
: 85. FAQ-12-0062, UFSAR Content, Rev. 1 Arkansas Nuclear One - Unit 2 Attachments Enclosure 1 to 2CAN121202 Page 91 ATTACHMENTS Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-1 A. NEI 04-02 Table B Transition of Fundamental FP Program & Design Elements NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.1* General This chapter contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features. These fire protection program elements and minimum design requirements shall not be subject to the performance-based methods permitted elsewhere in this standard. Previously approved alternatives from the fundamental protection program attributes of this chapter by the authority having jurisdiction (AHJ) take precedence over the requirements contained herein. N/A General statement, no technical requirements. See sub-sections for specific compliance statements and references.
3.2 Fire Protection Plan N/A N/A Section header, no technical requirements. See sub-sections for specific compliance statements and references.
3.2.1 Intent A site-wide fire protection plan shall be established. This plan shall document management policy and program direction and shall define the responsibilities of those individuals responsible for the plan's implementation. This section establishes the criteria for an integrated combination of components, procedures, and personnel to implement all fire protection program activities. Complies The site-wide fire protection plan is delineated in EN-DC-330 and OP-1003.014. The Fire Protection Program identifies the plant and corporate management positions responsible for implementing the Fire Protection Program and assigns their responsibilities and authorities. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 2.0 provides scope 3.2.2* Management Policy Direction and Responsibility A policy document shall be prepared that defines management authority and responsibilities and establishes the general policy for the site fire protection program. Complies Management responsibilities and authorities are delineated in EN-DC-330 and OP-1003.014. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.0 - Responsibilities 3.2.2.1* Management Policy on Senior Management The policy document shall designate the senior management position with immediate authority and responsibility for the fire protection program. Complies EN-DC-330 delineates responsibilities and authorities to plant and corporate management positions for implementing the Fire Protection Program and assigns ultimate responsibility of the ANO Fire Protection Program to the Site Vice President. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.0 - Responsibilities Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-2 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.2.2.2* Management Policy on Daily Administration The policy document shall designate a position responsible for the daily administration and coordination of the fire protection program and its implementation. Complies The ANO FPP delineates the responsibilities for administration of the current fire protection program across several organizations such as Engineering, Operations, Nuclear Oversight, Training, Maintenance, etc. The Director, Engineering (site) has responsibility to coordinate implementation to ensure compliance. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.9 - Supervisor Fire Protection 3.2.2.3* Management Policy on Interfaces The policy document shall define the fire protection interfaces with other organizations and assign responsibilities for the coordination of activities. In addition, this policy document shall identify the various plant positions having the authority for implementing the various areas of the fire protection program. Complies The ANO FPP assigns responsibilities and authorities among the organizations for implementing the fire protection program. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.0 - Responsibilities 3.2.2.4* Management Policy on AHJ The policy document shall identify the appropriate AHJ for the various areas of the fire protection program. Complies EN-DC-330 and OP-1003.014 defines the NRC as the AHJ for areas involving nuclear safety. EN-DC-330, Fire Protection Program, Rev. 1, Section 3.0 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 4.4 3.2.3* Procedures Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established: Complies General statement. See subsections for specific compliance statements and references. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-3 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.2.3 Procedures (1)* Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program. Complies Procedures are established for inspection, testing and maintenance of fire protection systems as identified in the ANO FPP. Surveillance frequencies are outlined in the ANO FPP and may be modified in accordance with the methodology in EPRI Report TR-1006756, Fire Protection Equipment Surveillance Optimization and Maintenance Guide. EN-DC-330, Fire Protection Program, Rev. 1, Section 5.3 [3] EPRI TR-1006756, Fire Protection Equipment Surveillance Optimization and Maintenance Guide, Rev. July, 2003 OP-1003.014, ANO Fire Protection Program, Rev. 6 OP-1104.032, Fire Protection Systems, Rev. 69 OP-2104.032, Unit 2 Fire Protection System Operations, Rev. 32 3.2.3 Procedures (2)* Compensatory actions implemented when fire protection systems and other systems credited by the fire protection program and this standard cannot perform their intended function and limits on impairment duration. Complies Compensatory actions are implemented as required by EN-DC-330 and as identified in the ANO-2 Technical Requirements Manual (TRM). Compensatory measures for any new systems required by the transition to 805 will be maintained in accordance with the requirements for similar systems in the TRM. ANO-2 TRM, Rev. 50, Sections 3.3.6 and 3.7.1 through 3.7.5 EN-DC-330, Fire Protection Program, Rev. 1, Section 5.3 [4]
3.2.3 Procedures (3)* Reviews of fire protection program - related performance and trends. Complies Program performance including system monitoring and trending along with program health reports are implemented in accordance with administrative control procedures. Implementation Item - The monitoring program required by NFPA 805 will include a process that monitors and trends the FPP based on specific goals established to measure effectiveness. This will be done prior to the implementation date. See implementation item in Attachment S. EN-DC-329, Engineering Programs Control and Oversight, Rev. 4 EN-DC-330, Fire Protection Program, Rev. 1, Section 4.14OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.0 - Responsibilities OP-1104.032, Fire Protection Systems, Rev. 69 OP-2104.032, Unit 2 Fire Protection System Operations, Rev. 32 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-4 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.2.3 Procedures (4) Reviews of physical plant modifications and procedure changes for impact on the fire protection program. Complies Plant modifications and procedure changes are reviewed for impact on the FPP as described in EN-DC-128. EN-DC-128, Fire Protection Impact Reviews, Rev. 5, Section 1.0 - Purpose 3.2.3 Procedures (5) Long-term maintenance and configuration of the fire protection program. Complies Long-term maintenance and configuration of the fire protection program are established by the ANO FPP procedure. OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.0 - ResponsibilitiesEN-DC-128, Fire Protection Impact Reviews, Rev. 5 3.2.3 Procedures (6) Emergency response procedures for the plant industrial fire brigade. Complies Emergency response procedures for the fire brigade are detailed in OP-1015.007, Fire Brigade Organization and Responsibilities. OP-1015.007, Fire Brigade Organization and Responsibilities, Rev. 25, Section 1.0, Purpose 3.3 Prevention A fire prevention program with the goal of preventing a fire from starting shall be established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following: (1) Prevention of fires and fire spread by controls on operational activities (2) Design controls that restrict the use of combustible materials The design control requirements listed in the remainder of this section shall be provided as described. Complies The ANO fire prevention program is established and implemented as detailed in the FPP. It includes controls on operational activities and design controls that restrict the use of combustible materials. See following subsections for additional specific compliance statements and references. EN-DC-128, Fire Protection Impact Reviews, Rev. 5, Section 5.4 - Fire Protection Program Review OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 6.0 - Fire Protection Program EN-DC-161, Control of Combustibles, Rev. 7 3.3.1 Fire Prevention for Operational Activities The fire prevention program activities shall consist of the necessary elements to address the control of ignition sources and the use of transient combustible materials during all aspects of plant operations. The fire prevention program shall focus on the human and programmatic elements necessary to prevent fires from starting or, should a fire start, to keep the fire as small as possible. Complies Control of ignition sources (EN-DC-127) and transient combustible materials (EN-DC-161) are established and implemented as detailed in the FPP. See following subsections for additional specific compliance statements and references. EN-DC-127, Control of Hot Work and Ignition Sources, Rev. 11, Section 1.0 - Purpose EN-DC-161, Control of Combustibles, Rev. 7, Section 1.0 - Purpose OP-1003.005, Fire Prevention
 
Inspection, Rev. 13, Section 1.0 - Purpose Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-5 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.1.1 General Fire Prevention Activities The fire prevention activities shall include but not be limited to the following program elements: Complies The ANO fire prevention program is established and implemented as detailed in the ANO FPP. See subsections for specific compliance statements and references. Entergy has developed multiple directives to address fire prevention. These directives address, at a minimum, the FPP elements identified in this section. Upon review of the elements listed below, the NFPA 805 code requirements are satisfied and no other additional elements were evaluated. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6 3.3.1.1 General Fire Prevention Activities (1) Training on fire safety information for all employees and contractors including, as a minimum, familiarization with plant fire prevention procedures, fire reporting, and plant emergency alarms. Complies General fire safety training for employees and contractors is covered during initial site indoctrination and annual re-qualification in General Employee Training (GET). FCBT-GET-PATSS, Plant Access Training, Rev. 17, Section 6 3.3.1.1 General Fire Prevention Activities (2)* Documented plant inspections including provisions for corrective actions for conditions where unanalyzed fire hazards are identified. Complies Periodic plant inspections are scheduled, conducted and documented as required by OP-1003.005. Corrective actions are initiated for conditions that decrease the effectiveness of the FPP. OP-1003.005, Fire Prevention Inspection, Rev. 13, Section 1.0 - Purpose 3.3.1.1 General Fire Prevention Activities (3)* Administrative controls addressing the review of plant modifications and maintenance to ensure that both fire hazards and the impact on plant fire protection systems and features are minimized. Complies Administrative controls requiring the fire protection review of plant modifications and maintenance are covered in EN-DC-128 and EN-MA-101. EN-DC-128, Fire Protection Impact Reviews, Rev. 5, Section 1.0 - Purpose EN-MA-101, Fundamentals of Maintenance, Rev. 12 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-6 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.1.2* Control of Combustible Materials Procedures for the control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements: Complies The ANO fire prevention program is established and implemented as detailed in the ANO FPP. Procedures include, but are not limited to, elements 3.3.1.2 (1) through (6). See subsections for specific compliance statements and references. EN-DC-330, Fire Protection Program, Rev. 1 OP-1003.014, ANO Fire Protection Program, Rev. 6 3.3.1.2 Control of Combustible Materials (1)* Wood used within the power block shall be listed pressure-impregnated or coated with a listed fire-retardant application. Exception:  Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 cm) or larger shall not be required to be fire-retardant treated. Complies EN-DC-161 states that lumber used in areas within the scope of this procedure should be treated with a pressure impregnated fire retardant chemical. If pressure impregnated wood is not available, obtain Fire Protection Staff approval prior to using wood treated with surface applied chemicals. Heavy wood members with a cross sectional area greater than or equal to 6" x 6" are NOT required to be treated with a fire retardant. EN-DC-161, Control of Combustibles, Rev. 7, Sections 5.3 [2] & [3]
3.3.1.2 Control of Combustible Materials (2)
Plastic sheeting materials used in the power block shall be fire-retardant types that have passed NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, large-scale tests, or equivalent. Complies EN-DC-161 states that plastic film and fabrics used as sheeting material for protective floor coatings or temporary enclosures shall be approved self-extinguishing fire retardant plastic sheeting (NFPA 701, UL Standard 214, or equivalent standard). EN-DC-161, Control of Combustibles, Rev. 7, Section 5.3 [4] & Attachment 9.13 3.3.1.2 Control of Combustible Materials (3) Waste, debris, scrap, packing materials, or other combustibles shall be removed from an area immediately following the completion of work or at the end of the shift, whichever comes first. Complies Combustibles are controlled by Procedure EN-DC-161. Section 5.2 [6] states that waste, debris, scrap, oil spills, or other combustibles resulting from the work activity should be removed promptly following completion of the work or at the end of each shift, whichever comes first. EN-DC-161, Control of Combustibles, Rev. 7, Section 5.2 [6] EN-MA-101, Fundamentals of Maintenance, Rev. 12, Section 5.17 OP-1000.018, Housekeeping, Rev. 29 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-7 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.1.2 Control of Combustible Materials (4)* Combustible storage or staging areas shall be designated, and limits shall be established on the types and quantities of stored materials. Complies Combustible storage or staging areas are designated and limits established on the types and quantities of stored materials in accordance with EN-DC-161. EN-DC-161, Control of Combustibles, Rev. 7, Section 5.6 and Attachment 9.1 3.3.1.2 Control of Combustible Materials (5)* Controls on use and storage of flammable and combustible liquids sha ll be in accordance with NFPA 30, Flammable and Combustible Liquids Code, or other applicable NFPA standards. Complies with use of EEEEs This requirement was evaluated by NFPA 30 Code Compliance Evaluation. Per FAQ 06-0020, the following guidance applies as to which NFPA standards referenced in Chapter 3 are applicable:  "Where used in NFPA 805, Chapter 3, the term, 'applicable NFPA Standards' is considered to be equivalent to those NFPA standards identified in the current license basis (CLB) for procedures and systems in the FPP that are transitioning to NFPA 805." No other NFPA standards were determined to be applicable. See implementation item in Attachment S. CALC-ANOC-FP-09-00007, ANO Code Compliance Report for NFPA 30 2000 Edition, Rev. 0 EN-DC-161, Control of Combustibles, Rev. 7, Section 5.4 3.3.1.2 Control of Combustible Materials (6)* Controls on use and storage of flammable gases shall be in accordance with applicable NFPA standards. Complies Specific administrative directives have been developed for use and control of flammable gases in accordance with NFPA 55 and Occupational Safety and Health Administration (OSHA). No other NFPA standards were determined to be applicable (FAQ 06-0020). EN-DC-161, Control of Combustibles, Rev. 7, Section 5.5 EN-IS-109, Compressed Gas Cylinder Handling and Storage, Rev. 7, Section 1.0 - Purpose 3.3.1.3 Control of Ignition Sources Control of Ignition Sources. N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-8 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.1.3.1* Control of Ignition Sources Code Requirements A hot work safety procedure shall be developed, implemented, and periodically updated as necessary in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations. Complies Hot work is controlled through administrative procedures in accordance with NFPA 51B. Notwithstanding the above, there are cases where sprinkler systems are purposely defeated, especially those that are electronically activated (such as those which automatically actuate via a smoke detector) prior to performing hot work in the respective area. In such cases, an operator may be assigned responsibility to unisolate the system should a fire occur in the area while hot work activities are ongoing. Other measures may be established to compensate for these sprinkler types, such as establishing a fire watch with a fire extinguisher. With such measures in place, the ability to respond to a fire in the respective area is not significantly impaired. Although rare, procedures do permit hot work under such conditions with specific management approval, provided suppression capability is made available. Given these controls, ANO complies with the intent of the NFPA 51B requirement. Compliance with NFPA 241 is addressed through compliance with NFPA 51B. NFPA 241, 2000 Edition, as referenced by NFPA 805, 2001 with respect to hot work states: "Responsibility for hot work operations and fire prevention precautions, including permits and fire watches, shall be in accordance with NFPA 51B." CALC-ANOC-FP-08-00011, ANO Code Compliance Report for NFPA 51B 1999 Edition, Rev. 0 EN-DC-127, Control of Hot Work and Ignition Sources, Rev. 11, Sections 1.0 &
2.0 [1] (a) 3.3.1.3.2 Control of Ignition Sources on Smoking Limitations Smoking and other possible sources of ignition shall be restricted to properly designated and supervised safe areas of the plant. Complies Smoking is required to be prohibited in certain areas by administrative controls. ANO policy prohibits smoking inside buildings. OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 6.2.2 OP-1000.018, Housekeeping, Rev. 29 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-9 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.1.3.3 Control of Ignition Sources for Leak Testing Open flames or combustion-generated smoke shall not be permitted for leak or air flow testing. Complies The use of open flames or combustion smoke as a testing medium is prohibited by EN-DC-127. EN-DC-127, Control of Hot Work and Ignition Sources, Rev. 11, Section 5.2 [25]
3.3.1.3.4* Control of Ignition Sources on Portable Heaters Plant administrative procedure shall control the use of portable electrical heaters in the plant. Portable fuel-fired heaters shall not be permitted in plant areas containing equipment important to nuclear safety or where there is a potential for radiological releases resulting from a fire. Complies Portable fuel-fired heaters are not permitted in plant areas containing equipment important to nuclear safety or where there is a potential for radiological releases resulting from a fire per EN-DC-127. EN-DC-127, Control of Hot Work and Ignition Sources, Rev. 11, Sections 5.1 [4], 5.2 [8] & Attachment 9.3 3.3.2 Structural Walls, floors, and components required to maintain structural integrity shall be of noncombustible construction, as defined in NFPA 220, Standard on Types of Building Construction. Complies Plant buildings are metal and concrete construction with fire walls and/or shield walls to isolate critical areas or equipment. Structural components consist of structural steel or reinforced concrete. In general, areas housing safety-related systems, equipment, and components are of concrete or masonry construction. ANO-2 SAR, Unit 2 Safety Analysis Report, Rev. 24, Sections 9.5.1.1 A. and 9.5.1.3.1 EN-DC-128, Fire Protection Impact Reviews, Rev. 5 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sections 4.11, 5.0, and 8.0
 
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, Life Safety Code, requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requirements for Class I interior floor finishes.
Submit for NRC approval ANO 2 SER, NUREG-0223, Sections 4.11, 5.0 and 8.0 include general statements about basic wall, floor, and ceiling structures having adequate resistance to prevent the spread of an unsu ppressed fire. Section 4.11 includes the following text: "The applicant's fire hazards analysis concludes that the basic wall, floor and ceiling structures bounding each fire zone have adequate resistance to prevent the spread of an unsuppressed fire through the barrier." Section 5.0 - Evaluation of Specific Plant Areas, provides details of various areas throughout the plant. Section 8.0 - Conclusions, notes that "Upon implementation and NRC verification of the applicant's proposed modifications summarized in Section 3.1, we find that the requirements of General Design Criterion 3 will be fully satisfied..." Coatings at ANO are maintained per SPEC-ANO-A-2436 and SPEC-ANO-A-2437. Epoxy floor coverings at ANO may not meet the NFPA 805 requirements for "interior finish" and are an exception to the interior finish requirement. ANO requests formal NRC approval of this exception. See Attachment L for further details on the request for NRC approval for interior finishes. EN-DC-128, Fire Protection Impact Reviews, Rev. 5 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sections 4.11, 5.0, and 8.0 SPEC-ANO-A-2436, Furnishing, Delivery and Application of Field Painting Outside of Containment, Rev. 3 SPEC-ANO-A-2437, Furnishing, Delivery and Application of Field Painting Inside of Containment, Rev. 1 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-11 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.4 Insulation Materials Thermal insulation materials, radiation shielding materials, ventilation duct materials, and soundproofing materials shall be noncombustible or limited combustible. Complies Procedure EN-DC-115, Engineering Change Development, specifically addresses fire protection program impact resulting from addition of flammable materials. OP-1003.014, ANO Fire Protection Program, states in Section 6.2.1.D that "Materials used in the plant shall be non-combustible or approved by Fire Protection Engineering." EN-DC-115, Engineering Change Development, Rev. 13, Attachments 9.3 and 9.4 EN-DC-128, Fire Protection Impact Reviews, Rev. 5 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 6.2.1.D 3.3.5 Electrical N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
3.3.5.1 Electrical Wiring Above Suspended Ceiling Limitations Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays with solid metal top and bottom covers.
Submit for NRC approval Wiring above suspended ceilings is addressed in approved Modifications procedures and are minimized. ANO has wiring above suspended ceilings that may not comply with the requirements in this code section. See Attachment L of the Transition Report for further details on the request for NRC approval for existing wiring above suspended ceilings. OP-6030.109, Installation of Electrical Cable and Wire, Rev. 6, Sections 9.1.5 and 9.29.1.D OP-6030.112, Installation of Raceway Systems, Rev. 6, Section 9.1.7 3.3.5.2 Electrical Raceway Construction Limits Only metal tray and metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components.
Submit for NRC approval Installation of raceway systems is addressed in approved procedures. Cable tray and conduit material is primarily of substantial metal construction. However, use of Schedule 40 PVC is allowed by procedure for underground and embedded applications per NFPA 70, National Electric Code. See Attachment L for further details on the request for NRC approval for use of PVC for embedded and underground applications. E-2059, Sh. 10, Conduit & Cable Tray Notes & Details, Rev. 4, Item IV.1.i E-2059, Sh. 5, Conduit & Cable Tray Notes & Details, Rev. 9, Items II.4 & II.5 NFPA 70, National Electric Code, Rev. 2008 Edition, Article 352 OP-6030.112, Installation of Raceway Systems, Rev. 6, Sections 9.1.5, 9.2.10, and 9.4.6 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-12 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.5.3* Electrical Cable Flame Propagation Limits Electric cable construction shall comply with a flame propagation test as acceptable to the AHJ. Complies by previous NRC approval NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, Section 4.8, states that "The cables in the plant were required to pass IPCEA standard S-19-81 flame tests. Additionally, completed cables were given a seven-minute flame test... The flame tests show that the cabling does not burn vigorously in the configurations used in the test. We find that retest to the IEEE Standard 383 procedure and criteria would not provide information that would change any of our recommendations or conclusions. Accordingly, we find the electrical cables used at the Arkansas Unit 2 plant acceptable." Cable specifications were revised to meet the requirements of IEEE 383-1974, IEEE 323-1974 and applicable IPCEA standards. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.8 SPEC-ANO-E-2425, 5000 Volt and 8000 Volt Cable, Rev. 1 SPEC-APL-E-2412, 600 Volt Single and Multiconductor EPM Power and Control Cable, Rev. 1 SPEC-APL-E-2413, Instrument and Special Cable, Rev. 2 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-13 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.6 Roofs Metal roof deck construction shall be designed and installed so the roofing system will not sustain a self-propagating fire on the underside of the deck when the deck is heated by a fire inside the building. Roof coverings shall be Class A as determined by tests described in NFPA 256, Standard Methods of Fire Tests of Roof Coverings. Complies by previous NRC approval NFPA 256 was not an original design requirement for the plant or referenced in BTP 9.5-1 or a condition in previous NRC SERs. However, original metal roof deck construction conformed to Underwriters Laboratories Class A roof covering materials and Underwriters Laboratories Metal Deck Assemblies Barriers Construction No. 1 (Ref. 0CAN097705). The requirement to "be listed by Underwriters' Laboratories, Inc., as suitable components for Class A construction" is contained in Technical Specification 6600-A-2, 6600-A-2002 and 6600-A-2023. These specs address roofing requirements for ANO 1 and ANO-2. 0CAN097705, Fire Protection, Additional Answers to Staff Questions, 9/21/1977, Answer to Item 5 ANO-2 SAR, Unit 2 Safety Analysis Report, Rev. 24, Section 9.5.1.3.1 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.11, Fire Barriers SPEC-6600-A-002, Tech Spec for Built-up Roofing, Roof Insulation and Vapor Barrier, Rev. 2 SPEC-6600-A-2002, Tech Spec for Built-up Roofing, Roof Insulation and Vapor Barrier, Rev. 1 SPEC-6600-A-2023, Tech Spec for Elastomeric Roofing, Rev. 1 3.3.7 Bulk Flammable Gas Storage Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing systems, equipment, or components important to nuclear safety. Complies Bulk compressed or cryogenic flammable gas storage is not permitted inside structures, housing systems, equipment, or components important to nuclear safety. EN-IS-109, Compressed Gas Cylinder Handling and Storage, Rev. 7, Section 5.1 [4]
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-14 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.7.1 Bulk Flammable Gas Location Requirements Storage of flammable gas shall be located outdoors, or in separate detached buildings, so that a fire or explosion will not adversely impact systems, equipment, or components important to nuclear safety. NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be followed for hydrogen storage. Complies with use of EEEEs This requirement was evaluated by NFPA 50A Code Compliance Evaluation. This evaluation identified modifications required for the Hydrogen Gas Bottle Room to meet code requirements. See implementation item in Attachment S. CALC-ANOC-FP-09-00008, ANO Code Compliance Report for NFPA 50A 1973 Edition, Rev. 1 EN-DC-115, Engineering Change Development, Rev. 13, Attachments 9.3 and 9.4 EN-DC-161, Control of Combustibles, Rev. 7, Sections 1.0 [2], 5.1 [2], 5.5 EN-IS-109, Compressed Gas Cylinder Handling and Storage, Rev. 7, Section 1.0 3.3.7.2 Bulk Flammable Gas Container Restrictions Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is not pointed at buildings. Complies NFPA 10 defines "High-Pressure Cylinder":  For the purposes of this standard, high-pressure cylinders (and cartridges) are those containing nitrogen, compressed air, carbon dioxide, or other gases at a pressure higher than 500 psi (3447 kPa) at 70&deg;F (21&deg;C). EN-IS-109 requires outdoor high-pressure flammable gas storage containers located so that the long axis is not pointed at buildings. EN-IS-109, Compressed Gas Cylinder Handling and Storage, Rev. 7, Section 5.1 [5]
3.3.7.3 Bulk Flammable Gas Cylinder Limitations Flammable gas storage cylinders not required for normal operation shall be isolated from the system. Complies Flammable gas cylinders that are not in use are isolated by plant procedures. EN-DC-161, Control of Combustibles, Rev. 7, Section 5.5 [1] (a) & (b) 3.3.8 Bulk Storage of Flammable and Combustible Liquids Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing systems, equipment, or components important to nuclear safety. As a minimum, storage and use shall comply with NFPA 30, Flammable and Combustible Liquids Code. Complies with use of EEEEs This requirement was evaluated by NFPA 30 Code Compliance Evaluation.
CALC-ANOC-FP-09-00007, ANO Code Compliance Report for NFPA 30 2000 Edition, Rev. 0 EN-DC-161, Control of Combustibles, Rev. 7, Sections 1.0 [2], 5.1 [2]
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-15 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.9* Transformers Where provided, transformer oil collection basins and drain paths shall be periodically inspected to ensure that they are free of debris and capable of performing their design function. Complies ANO procedures include requirements for inspecting transformer oil collection basins and drain paths. OP-1003.005, Fire Prevention Inspection, Rev. 13, 6.1.3.K OP-1015.033, ANO Switchyard and Transformer Yard Controls, Rev. 18, Section 5.6 3.3.10* Hot Pipes and Surfaces Combustible liquids, including high flashpoint lubricating oils, shall be kept from coming in contact with hot pipes and surfaces, including insulated pipes and surfaces. Administrative controls shall require the prompt cleanup of oil on insulation. Complies Administrative procedure EN-DC-161 addresses the use of combustible liquids around hot pipes and surfaces. EN-DC-161, Control of Combustibles, Rev. 7, Section 5.4[1](c) 3.3.11 Electrical Equipment Adequate clearance, free of combustible material, shall be maintained around energized electrical equipment. Complies ANO procedures designate storage areas for combustible materials, none of which are around energized electrical equipment. Energized electrical components are maintained free from adjacent combustible material per OP-1003.005, Fire Prevention Inspection. EN-DC-161, Control of Combustibles, Rev. 7 OP-1003.005, Fire Prevention Inspection, Rev. 13, Section 6.1.3.L 3.3.12* Reactor Coolant Pumps For facilities with non-inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply. Complies by previous NRC approval Exemption Granted by NRC as documented in 2CNA069701 and 2CNA108802. 2CNA108802, includes the following discussion:  "The licensee requested approval of exemptions from the technical requirements of Section III.O of Appendix R to 10 CFR Part 50 to the extent that it requires the reactor coolant pump (RCP) oil collection system to be sized to hold the contents of the entire lube oil system for all pumps and to be designed to withstand a safe shutdown earthquake (SSE). (continued) 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997, Section 5.0 of Evaluation 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Section 6.0 of Safety Evaluation FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 45 (continued)
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-16 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.12 (continued)
Based on the above evaluation, the licensee's alternative design of the oil collection system provides an equivalent level of safety to that achieved by compliance with Section III.O of Appendix R. Therefore, the licensee's request for exemption should be approved." 2CNA069701 includes the following discussion:  "Therefore, contingent on the use of the compensatory measures that are itemized in the licensee's exemption request dated December 23, 1996, the staff concluded that the design of the oil filling system and the level of protection provided during oil fill operations provide reasonable assurance that a lube oil fire will not occur. The staff also concluded that in the event of a worst-case postulated fire, due to not having a lube oil collection system for the reactor coolant pump lube oil fill lines, it would be of limited magnitude and extent. In addition, such a fire would not cause significant damage in the containment building and would not prevent the operators from achieving and maintaining safe shutdown conditions. The licensee's request for exemption should, therefore, be granted." OP-1504.001 provides instructions to inspect the lube oil collection system for operability and integrity on all four of the Reactor Coolant Pumps for both ANO-1 and ANO-2. OP-1504.001, Visual Inspection of the Unit 1 & 2 RCP Oil Collection System, Rev. 9 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-17 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.12 Reactor Coolant Pumps (1) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oil from all potential pressurized and non-pressurized leakage sites in each reactor coolant pump oil system.
Submit for NRC Approval Exemption granted by NRC as documented in 2CNA069701 and 2CNA108802. See Section 3.3.12. The RCP oil collection systems are designed and sized to collect and contain oil from potentially pressurized and unpressurized leakage areas in seismic event resulting in failure of the lubrication system. See Attachment L of the Transition Report for further details on the request for NRC approval for evaluation of oil misting from the reactor coolant pumps/motors. 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988 3.3.12 Reactor Coolant Pumps (2) Leakage shall be collected and drained to a vented closed container that can hold the inventory of the reactor coolant pump lubricating oil system. Complies by previous NRC approval Exemption Granted by NRC as documented in 2CNA069701 and 2CNA108802. See Section 3.3.12. 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988 3.3.12 Reactor Coolant Pumps (3) A flame arrestor is required in the vent if the flash point characteristics of the oil present the hazard of a fire flashback. Complies by previous NRC approval Exemption Granted by NRC as documented in 2CNA069701 and 2CNA108802. See Section 3.3.12. 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-18 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.3.12 Reactor Coolant Pumps (4) Leakage points on a reactor coolant pump motor to be protected shall include, but not be limited to, the lift pump and piping, overflow lines, oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps. Complies by previous NRC approval Exemption Granted by NRC as documented in 2CNA069701 and 2CNA108802. See Section 3.3.12. 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988 3.3.12 Reactor Coolant Pumps (5) The collection basin drain line to the collection tank shall be large enough to accommodate the largest potential oil leak such that oil leakage does not overflow the basin. Complies by previous NRC approval Exemption Granted by NRC as documented in 2CNA069701 and 2CNA108802. See Section 3.3.12. 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988 3.4 Industrial Fire Brigade N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
3.4.1 On-Site Fire-Fighting Capability All of the following requirements shall apply N/A General statement. See subsections for any specific compliance statements and references.
3.4.1 On-Site Fire-Fighting Capability (a) A fully staffed, trained, and equipped fire-fighting force shall be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty and shall conform with the following NFPA standards as applicable: Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. The on-site fire brigade is appropriately staffed, trained and equipped and complies with NFPA 600. NFPA 1500 and NFPA 1582 do not apply at ANO. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 OP-1015.007, Fire Brigade Organization and Responsibilities, Rev. 25, Sections 4.1 - 4.3 and 6.1.2 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-19 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.1 On-Site Fire-Fighting Capability (a)(1) NFPA 600, Standard on Industrial Fire Brigades (interior structural fire fighting). Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. The on-site fire brigade is appropriately staffed, trained and equipped and complies with NFPA 600. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 OP-1015.007, Fire Brigade Organization and Responsibilities, Rev. 25, Sections 4.1 - 4.3 and 6.1.2 3.4.1 On-Site Fire-Fighting Capability (a)(2) NFPA 1500, Standard on Fire Department Occupational Safety and Health Program. N/A Not applicable to ANO.
3.4.1 On-Site Fire-Fighting Capability (a)(3) NFPA 1582, Standard on Medical Requirements for Fire Fighters and Information for Fire Department Physicians. N/A Not applicable to ANO.
3.4.1 On-Site Fire-Fighting Capability (b)* Industrial fire brigade members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required. Complies A fully staffed, trained, and equipped fire-fighting force is available at all times to control and extinguish all fires on site. This force is required to have a minimum complement of five persons on duty and conforms with the applicable NFPA standards of this element. OP-1015.007, Fire Brigade Organization and Responsibilities, Rev. 25, Section 5.0 3.4.1 On-Site Fire-Fighting Capability (c) During every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance. Exception:  Sufficient training and knowledge shall be permitted to be provided by an operations advisor dedicated to industrial fire brigade support criteria. Complies The Fire Brigade Leader (Unit 2) and three other members (Unit 1) are from the Operations Department. Fire Brigade members respond to the fire under the direction of the Operations Shift Manager. The Operations Shift Manager is not dedicated to fire brigade support. OP-1015.007, Fire Brigade Organization and Responsibilities, Rev. 25, Sections 4.1 - 4.3 and 6.1 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-20 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.1 On-Site Fire-Fighting Capability (d)* The industrial fire brigade shall be notified immediately upon verification of a fire. Complies OP-2203.034 requires an action to notify the fire brigade to respond in the event a fire. OP-2203.034, Fire or Explosion, Rev. 13, Step 10 3.4.1 On-Site Fire-Fighting Capability (e) Each industrial fire brigade member shall pass an annual physical examination to determine that he or she can perform the strenuous activity required during manual fire-fighting operations. The physical examination shall determine the ability of each member to use respiratory protection equipment. Complies EN-RP-501 and EN-NS-112 requires each member of the fire brigade to maintain a current annual physical that ensure the member is capable of performing strenuous activities and the ability to use respiratory protection equipment. EN-NS-112, Medical Program, Rev. 9, Section 5.6[3] EN-RP-501, Respiratory Protection Program, Rev. 4, Section 5.3, Medical Surveillance 3.4.2* Pre-Fire Plans Current and detailed pre-fire plans shall be available to the industrial fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1.5. Complies Pre-fire Plans are provided for both safety related and non-safety related areas of the facility. PFP-U2, ANO Prefire Plan (Unit 2), Rev. 11, Section 1.2 PFP-UC, ANO Prefire Plan (Common), Rev. 13, Section 1.2 3.4.2.1* Pre-Fire Plans Contents The plans shall detail the fire area configuration and fire hazards to be encountered in the fire area, along with any nuclear safety components, and fire protection systems and features that are present. Complies Pre-fire Plans detail the fire area configuration, fire hazards to be encountered and the fire protection systems and features present. PFP-U2, ANO Prefire Plan (Unit 2), Rev. 11, Sections 1.2 & 1.4 PFP-UC, ANO Prefire Plan (Common), Rev. 13, Sections 1.2 & 1.3 3.4.2.2 Pre-Fire Plans Updates Pre-fire plans shall be reviewed and updated as necessary. Complies Pre-fire Plans shall be reviewed and updated as required by OP-1003.013. OP-1003.013, Control of Prefire Plans, Rev. 1, Section 5.0 OP-1003.014, ANO Fire Protection Program, Rev. 6, Section 5.8.3 3.4.2.3* Pre-Fire Plans Locations Pre-fire plans shall be available in the control room and made available to the plant industrial fire brigade. Complies Controlled copies of the pre-fire plans are readily available for use by the Fire Brigade Leader, Control Room Supervisor and Shift Manager per OP-1003.013. OP-1003.013, Control of Prefire Plans, Rev. 1, Section 6.0 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-21 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.2.4* Pre-Fire Plans Coordination Needs Pre-fire plans shall address coordination with other plant groups during fire emergencies. Complies Pre-fire Plans require the Fire Brigade Leader to maintain contact with the Shift Manager who interfaces with other plant groups. OP-2203.034, Fire or Explosion, Rev. 13, Various Steps PFP-U2, ANO Prefire Plan (Unit 2), Rev. 11, Section 1.4 PFP-UC, ANO Prefire Plan (Common), Rev. 13, Section 1.3 3.4.3 Training and Drills Industrial fire brigade members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply. (1) Plant industrial fire brigade members shall receive training consistent with the requirements contained in NFPA 600, Standard on Industrial Fire Brigades, or NFPA 1500, Standard on Fire Department Occupational Safety and Health Program, as appropriate. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. NFPA 1500 does not apply since ANO uses a Fire Brigade and is not a Fire Department. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (a) (2) Industrial fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity and health physics considerations, to ensure that each member is thoroughly familiar with the steps to be taken in the event of a fire. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-22 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.3 Training and Drills (a) (3) A written program shall detail the industrial fire brigade training program. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. The training program is detailed in OP-1063.020, "Fire Brigade Training Program." CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (a) (4) Written records that include but are not limited to initial industrial fire brigade classroom and hands-on training, refresher training, special training schools attended, drill attendance records, and leadership training for industrial fire brigades shall be maintained for each industrial fire brigade member. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (b) Training for Non-Industrial Fire Brigade Personnel. Plant personnel who respond with the industrial fire brigade shall be trained as to their responsibilities, potential hazards to be encountered, and interfacing with the industrial fire brigade. Complies Non-Industrial Fire Brigade personnel at ANO do not respond with the Industrial Fire Brigade to a fire. Instructions are given for reporting fires. FCBT-GET-PATSS, Plant Access Training, Rev. 17, Section 6 3.4.3 Training and Drills (c)* Drills. All of the following requirements shall apply. (1) Drills shall be conducted quarterly for each shift to test the response capability of the industrial fire brigade. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. This requirement is addressed specifically by EN-TQ-125, Fire Brigade Drills. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-23 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.3 Training and Drills (c) (2) Industrial fire brigade drills shall be developed to test and challenge industrial fire brigade response, including brigade performance as a team, proper use of equipment, effective use of pre-fire plans, and coordination with other groups. These drills shall evaluate the industrial fire brigade's abilities to react, respond, and demonstrate proper fire-fighting techniques to control and extinguish the fire and smoke conditions being simulated by the drill scenario. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (c) (3) Industrial fire brigade drills shall be conducted in various plant areas, especially in those areas identified to be essential to plant operation and to contain significant fire hazards. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (c) (4) Drill records shall be maintained detailing the drill scenario, industrial fire brigade member response, and ability of the industrial fire brigade to perform as a team. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. Drill records are specifically addressed in EN-TQ-125, Fire Brigade Drills. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 3.4.3 Training and Drills (c) (5) A critique shall be held and documented after each drill. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation. A critique is specifically required by EN-TQ-125, Fire Brigade Drills. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 EN-TQ-125, Fire Brigade Drills, Rev. 1 OP-1063.020, Fire Brigade Training Program, Rev. 16 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-24 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.4 Fire-Fighting Equipment Protective clothing, respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with the applicable NFPA standards. Complies This requirement was evaluated by NFPA 600 Code Compliance Evaluation and is documented in OP-1003.005. CALC-ANOC-FP-08-00005, ANO Code Compliance Report for NFPA 600 2000 Edition, Rev. 1 OP-1003.005, Fire Prevention Inspection, Rev. 13, Section 5.4.5 and 5.4.6 OP-1015.007, Fire Brigade Organization and Responsibilities, Rev. 25, Section 8.0 and Supplement 1 3.4.5 Off-site Fire Department Interface N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
3.4.5.1 Mutual Aid Agreement Off-site fire authorities s hall be offered a plan for their interface during fires and related emergencies on site. Complies The London Fire Department, by Letter of Agreement, agrees to provide personnel and equipment as required to assist the ANO Fire Brigade in extinguishing fires located at the ANO site, includes both inside and outside the protected area. ANO Emergency Plan, Rev. 35, Section A, Item 2.4 & App.1, Item 18 3.4.5.2* Site-Specific Training Fire fighters from the o ff-site fire authorities who are expected to respond to a fire at the plant shall be offered site-specific training and shall be invited to participate in a drill at least annually. Complies At least once per year a drill is conducted with invited participation of the London Fire Department. ANO Emergency Plan, Rev. 35, Section N, Item 2.2 EN-TQ-125, Fire Brigade Drills, Rev. 1, Section 5.3 3.4.5.3* Security and Radiation Protection Plant security and radiation protection plans shall address off-site fire authority response. Complies Plant security and radiation protection plans address off-site fire response. Response by security and radiation protection (RP) is initiated by OP-2203.034. ANO Emergency Plan, Rev. 35, Section 2.4.1.A OP-1043.002, Access Control, Rev. 69, Step 6.3.9 OP-1203.048, Security Event, Rev. 16, Controlled OP-2203.034, Fire or Explosion, Rev. 13, Step 15 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-25 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.4.6* Communications An effective emergency communications capability shall be provided for the industrial fire brigade. Complies Use of plant radios is described in OP-2303.034, Fire or Explosion. ANO Emergency Plan, Rev. 35, Section 3.1(g), Att. F and H NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.7 OP-1903.062, Communications System Operating Procedure, Rev. 25OP-2203.034, Fire or Explosion, Rev. 13 3.5 Water Supply N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
3.5.1 Water Supply Flow Code Requirements A fire protection water supply of adequate reliability, quantity, and duration shall be provided by one of the two following methods. (a) Provide a fire protection water supply of not less than two separate 300,000 gal (1,135,500 L) supplies. (b) Calculate the fire flow rate for 2 hours. This fire flow rate shall be based on 500 gpm (1892.5 L/min) for manual hose streams plus the largest design demand of any sprinkler or fixed water spray system(s) in the power block as determined in accordance with NFPA 13, Standard for the Installation of Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service. Complies by previous NRC approval ANO-2 Fire Protection SER (NUREG-0223), Section 4.3.1.1, includes the following text: "Fire water for Unit 2 is supplied by two fire pumps located in the Unit 1 intake structure. These pumps are shared by Unit 1 and Unit 2. The two fire pumps take suction from separate water bays which are normally supplied from Dardanelle Reservoi r through intake screens. The service water bays can also be supplied from the emergency cooling water pond which is the ultimate heat sink. The ultimate heat sink would not be degraded by fire water supply requirements. We find that the fire water supply system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." (continued)
NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ions 4.3.1.1 and 4.3.1.2 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-26 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.5.1 (continued)  NUREG-0223 (SER), Section 4.3.1.2 includes the following text: "...Either of the two fire pumps has sufficient capacity to supply the maximum sprinkler demand with adequate reserve available for fire hoses."  3.5.2* Water Supply Tank Code Requirements The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection. Exception No. 1:  Water storage tanks shall not be required when fire pumps are able to take suction from a large body of water (such as a lake), provided each fire pump has its own suction and both suctions and pumps are adequately separated. Exception No. 2:  Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service. Complies by previous NRC approval Exception No. 1 applies. ANO-2 Fire Protection SER (NUREG-0223), Section 4.3.1.1, includes the following text: "Fire water for Unit 2 is supplied by two fire pumps located in the Unit 1 intake structure. These pumps are shared by Unit 1 and Unit 2. The two fire pumps take suction from separate water bays which are normally supplied from Dardanelle Reservoi r through intake screens. The service water bays can also be supplied from the emergency cooling water pond which is the ultimate heat sink. The ultimate heat sink would not be degraded by fire water supply requirements. We find that the fire water supply system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.1 3.5.3* Water Supply Pump Code Requirements Fire pumps, designed and installed in accordance with NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of the largest pump or pump power source.
Submit for NRC approval This requirement was evaluated by NFPA 20 Code Compliance Evaluation. See Attachment L of the Transition Report for further details regarding the request for NRC approval associated with the electric fire pump. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation
 
Report for ANO-2, 8/30/1978, Sect ion 4.3.1.2 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-27 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.5.4 Water Supply Pump Diversity and Redundancy At least one diesel engine-driven fire pump or two more seismic Category I Class IE electric motor-driven fire pumps connected to redundant Class IE emergency power buses capable of providing 100 percent of the required flow rate and pressure shall be provided. Complies This requirement was evaluated by NFPA 20 Code Compliance Evaluation. ANO has a diesel engine-driven fire pump capable of providing 100 percent of the required flow rate and pressure per NUREG-0223. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.2 3.5.5 Water Supply Pump Separation Requirements Each pump and its driver and controls shall be separated from the remaining fire pumps and from the rest of the plant by rated fire barriers. Complies This requirement was evaluated by NFPA 20 Code Compliance Evaluation. Per NUREG-0223, the pump and its driver and controls are separated from the remaining fire pumps and from the rest of the plant by rated fire barriers. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.2 3.5.6 Water Supply Pump Start/Stop Requirements Fire pumps shall be provided with automatic start and manual stop only. Complies Fire pumps start automatically when the pressure in the fire main drops. Fire pump shutdown is accomplished by manual means only. See code compliance report for NFPA 20. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.2 3.5.7 Water Supply Pump Connection Requirements Individual fire pump connections to the yard fire main loop shall be provided and separated with sectionalizing valves between connections. Complies Fire pump connections to the yard fire main loop are provided and separated with sectionalizing valves between connections per NUREG-0223. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-28 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.5.8 Water Supply Pressure Maintenance Limitations A method of automatic pressure maintenance of the fire protection water system shall be provided independent of the fire pumps. Complies An automatic electric jockey pump maintains pressure on the fire water piping system. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.2 3.5.9 Water Supply Pump Operation Notification Means shall be provided to immediately notify the control room, or other suitable constantly attended location, of operation of fire pumps. Complies Fire pump operation is annunciated in the Unit 1 Control Room. Note that the ANO-1 and ANO-2 Control Rooms are a common area, separated by a glass door. Any fire alarm received on equipment common to both units can be immediately communicated to Operators of the other unit. CALC-ANOC-FP-09-00006, ANO Code Compliance Report for NFPA 20 1969 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.2 OP-1203.009, Fire Protection System Annunciator Corrective Action, Rev. 27, Annunciator K-12 3.5.10 Water Supply Yard Main Code Requirements An underground yard fire main loop, designed and installed in accordance with NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be installed to furnish anticipated water requirements. Complies with use of EEEEs This requirement was evaluated by NFPA 24 Code Compliance Evaluation.
CALC-ANOC-FP-09-00015, ANO Code Compliance Report for NFPA 24 1995 Edition, Rev. 1 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-29  NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.5.11 Water Supply Yard Main Maintenance Issues Means shall be provided to isolate portions of the yard fire main loop for maintenance or repair without simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stations provided for manual backup. Sprinkler systems and manual hose station standpipes shall be connected to the plant fire protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems. Complies by previous NRC approval ANO-2 Fire Protection SER (NUREG-0223), Section 4.3.1.3, includes the following text: "Each of the two fire pumps has a separate discharge into the 12-inch underground fire loop which encircles both Unit 1 and Unit 2. Valving is arranged so that a single break in the discharge piping will not remove both fire pumps from service. All yard fire hydrants, fixed water suppression systems, and interior fire hose stations are supplied by the fire loop. Sectionalizing valves are provided on the loop to allow isolation of various sections for maintenance or repair. For certain areas inside the plant, both automatic suppression systems and fire hose stations are supplied by a common piping system, so that both primary and backup protection would be lost by closure of a single control valve or by a single break. For such piping system failures, hoses could be run from outside hydrants to provide protection during the short interval while repairs are being made. Such alternative protection is required to be provided by the technical specifications... We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." Note that the aforementioned fire specifications have since been relocated to the ANO-2 TRM. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-30 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.5.12 Water Supply Compatible Thread Connections Threads compatible with those used by local fire departments shall be provided on all hydrants, hose couplings, and standpipe risers. Exception:  Fire departments shall be permitted to be provided with adapters that allow interconnection between plant equipment and the fire department equipment if adequate training and procedures are provided. Complies ANO-2 Fire Protection SER (NUREG-0223), Section 4.3.1.3, includes the following text: "The hydrant hose threads are compatible with those of the local fire department.... We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 3.5.13 Water Supply Header Options Headers fed from each end shall be permitted inside buildings to supply both sprinkler and standpipe systems, provided steel piping and fittings meeting the requirements of ANSI B31.1, Code for Power Piping, are used for the headers (up to and including the first valve) supplying the sprinkler systems where such headers are part of the seismically analyzed hose standpipe system. Where provided, such headers shall be considered an extension of the yard main system. Each sprinkler and standpipe system shall be equipped with an outside screw and yoke (OS&Y) gate valve or other approved shutoff valve. Complies by previous NRC approval NUREG-0223, Unit 2 Fire Protection SER, accepts this system. Section 4.3.1.3 summary states: "We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 ULD-0-SYS-09, ANO Fire Protection System, Rev. 4, Section 4.8, Codes and Standards 3.5.14* Water Supply Control Valve Supervision All fire protection water supply and fire suppression system control valves shall be under a periodic inspection program and shall be supervised by one of the following methods. (a) Electrical supervision with audible and visual signals in the main control room or other suitable constantly attended location. (b) Locking valves in their normal position. Keys shall be made available only to authorized personnel. (continued) Complies by previous NRC approval NUREG-0223 (SER), Section 4.3.1.3 includes the following text: "Some fire water system control valves are electrically supervised; others are not electrically supervised, including those on the underground fire loop and at the fire pumps. The facility's technical specifications require a periodic check of the position of those valves which are not locked, sealed, electrically supervised, or otherwise secured in position to assure that valves are maintained in the open position...                          (continued)
NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 OP-2104.032, Unit 2 Fire Protection System Operations, Rev. 32, Section 2.0 and Attachment A Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-31 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.5.14 (continued) (c) Sealing valves in their normal positions. This option shall be utilized only where valves are located within fenced areas or under the direct control of the owner/operator. We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." Note that the aforementioned fire specifications have since been relocated to the ANO-2 TRM.
3.5.15 Water Supply Hydrant Code Requirements Hydrants shall be installed approximately every 250 ft (76 m) apart on the yard main system. A hose house equipped with hose and combination nozzle and other auxiliary equipment specified in NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be provided at intervals of not more than 1000 ft (305 m) along the yard main system. Exception:  Mobile means of providing hose and associated equipment, such as hose carts or trucks, shall be permitted in lieu of hose houses. Where provided, such mobile equipment shall be equivalent to the equipment supplied by three hose houses. Complies by previous NRC approval NUREG-0223, Unit 2 Fire Protection SER, accepts this system with deviations. Section 4.3.1.3 summary states: "We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." CALC-ANOC-FP-09-00015, ANO Code Compliance Report for NFPA 24 1995 Edition, Rev. 1 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 3.5.16* Water Supply Dedicated Limits The fire protection water supply system shall be dedicated for fire protection use only. Exception No. 1:  Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis. Exception No. 2:  Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section. Submit for NRC approval System may be used with a temporary pump to supply cooling loads on either unit during outages. See Attachment L of the Transition Report for further details on the request for NRC approval for use of the fire protection water supply system for non-fire protection purposes. OP-1104.032, Fire Protection Systems, Rev. 69, Sections 2.2 and 17.0, and Attachment E Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-32 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.6 Standpipe and Hose Stations N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
3.6.1 Standpipe and Hose Stations Code Requirements For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems. Complies with use of EEEEs This requirement was evaluated by NFPA 14 Code Compliance Evaluation. See Implementation Item in Attachment S. CALC-ANOC-FP-09-00005, ANO Code Compliance Report for NFPA 14 1983 Edition, Rev. 2 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.4 3.6.2 Standpipe and Hose Stations Capability Limitations A capability shall be provided to ensure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel. Complies This requirement was evaluated by NFPA 14 Code Compliance Evaluation. CALC-ANOC-FP-09-00005, ANO Code Compliance Report for NFPA 14 1983 Edition, Rev. 2 3.6.3 Standpipe and Hose Stations Nozzle Restrictions The proper type of hose nozzle to be supplied to each power block area shall be based on the area fire hazards. The usual combination spray/straight stream nozzle shall not be used in areas where the straight stream can cause unacceptable damage or present an electrical hazard to fire-fighting personnel. Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltage shock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flow from full open to full closed. Complies NUREG-0223 Section 4.3.1.4 states: "Nozzles on the hose lines are of the adjustable spray type; in areas of potential electrical fires, they are of a type rated for this service." CALC-ANOC-FP-09-00005, ANO Code Compliance Report for NFPA 14 1983 Edition, Rev. 2 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.4 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-33 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.6.4 Standpipe and Hose Stations Earthquake Provisions Provisions shall be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE). Complies by previous NRC approval NUREG-0223, Fire Protection SER for ANO-2, states in Section 4.3.1.3: "Portions of the fire water system that are within engineered safety features rooms are supported in accordance with seismic Category I requirements; all other portions are designed to meet seismic Category II criteria." NUREG-0223 states in the Section 4.3.1.3 summary: "We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.2 3.6.5 Standpipe and Hose Stations Seismic Connection Limitations Where the seismic required hose stations are cross-connected to essential seismic non-fire protection water supply systems, the fire flow shall not degrade the essential water system requirement. N/A Not applicable at ANO. Hose stations are not cross-connected to essential non-fire protection water supply systems. 3.7 Fire Extinguishers Where provided, fire extinguishers of the appropriate number, size, and type shall be provided in accordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted to be positioned outside of fire areas due to radiological conditions. Complies with use of EEEEs This requirement was evaluated by NFPA 10 Code Compliance Evaluation. See Implementation Item in Attachment S. CALC-ANOC-FP-09-00009, ANO Code Compliance Report for NFPA 10 1998 Edition, Rev. 0 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, 4.3.3 3.8 Fire Alarm and Detection Systems N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-34 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.8.1 Fire Alarm Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code
. Alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble signals to the control room or other constantly attended location from which required notifications and response can be initiated. Personnel assigned to the proprietary alarm station shall be permitted to have other duties. The following fire-related signals shall be transmitted: Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. The NRC accepted the fire detection and signaling system in the ANO-2 SER subject to stated modifications. 0CNA088203 documents that the modifications were implemented. 0CNA088203, NRC Inspection Report, 8/6/1982, Item 31 CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.2 3.8.1 Fire Alarm (1) Actuation of any fire detection device. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 (continued)
 
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(continued)
CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 3.8.1 Fire Alarm (2) Actuation of any fixed fire suppression system. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-36 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.8.1 Fire Alarm (3) Actuation of any manual fire alarm station. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 3.8.1 Fire Alarm (4) Starting of any fire pump. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance
 
Report for NFPA 72A 1985 Edition, Rev. 0 (continued)
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-37 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.8.1 Fire Alarm (4)
(continued)
CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 3.8.1 Fire Alarm (5) Actuation of any fire protection supervisory device. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 3.8.1 Fire Alarm (6) Indication of alarm system trouble condition. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 (continued)
 
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(continued)
CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 3.8.1.1 Fire Alarm Communication Requirements Means shall be provided to allow a person observing a fire at any location in the plant to quickly and reliably communicate to the control room or other suitable constantly attended location. Complies Communication of a fire emergency is provided through the use of the plant paging system, the intra-plant telephone system, and radio communication equipment. ANO Emergency Plan, Rev. 35, Section H, 2.0 - Communication Systems 3.8.1.2 Fire Alarm Prompt Notification Limits Means shall be provided to promptly notify the following of any fire emergency in such a way as to allow them to determine an appropriate course of action: Complies Notification of a fire emergency to all affected personnel is provided by the referenced implementing procedures. The primary line of notification to plant personnel and fire brigade would be through the use of the plant paging system, which is strategically located throughout the plant site. Also, the intra-plant telephone system would be used. This system allows direct dialing between all plant telephones. Additionally, various radio-based equipment can be used. ANO Emergency Plan, Rev. 35, Section H, 2.0 - Communication Systems OP-2203.034, Fire or Explosion, Rev. 13, Steps 10, 12, and 15 3.8.1.2 Fire Alarm Prompt Notification Limits (1) General site population in all occupied areas. Complies The primary line of notification to plant personnel and fire brigade would be through the use of the plant paging system, which is strategically located throughout the plant site. ANO Emergency Plan, Rev. 35 OP-2203.034, Fire or Explosion, Rev. 13, Step 12 3.8.1.2 Fire Alarm Prompt Notification Limits (2) Members of the industrial fire brigade and other groups supporting fire emergency response. Complies The primary line of notification to plant personnel and fire brigade would be through the use of the plant paging system, which is strategically located throughout the plant site. Also, the intra-plant telephone system and radios would be used. ANO Emergency Plan, Rev. 35 OP-2203.034, Fire or Explosion, Rev. 13, Steps 10 and 15 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-39 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.8.1.2 Fire Alarm Prompt Notification Limits (3) Off-site fire emergency response agencies. Two independent means shall be available (e.g., telephone and radio) for notification of off-site emergency services. Complies Two independent means are available (telephone and radio) for notification of off-site emergency services. ANO Emergency Plan, Rev. 35 OP-2203.034, Fire or Explosion, Rev. 13, Step 15 OP-1903.062, Communications System Operating Procedure, Rev. 25 3.8.2 Detection If automatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its applicable appendixes. Complies with use of EEEEs This requirement was evaluated by NFPA 72 Code Compliance Evaluation. CALC-ANO2-FP-09-00001, ANO Code Compliance Report for NFPA 72E 1974 Edition, Rev. 1 CALC-ANO2-FP-09-00002, ANO Code Compliance Report for NFPA 72A 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00003, ANO Code Compliance Report for NFPA 72D 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00021, ANO Code Compliance Report for NFPA 72A 1985 Edition, Rev. 0 CALC-ANO2-FP-09-00022, ANO Code Compliance Report for NFPA 72E 1984 Edition, Rev. 1 3.9 Automatic and Manual Water-Based Fire Suppression Systems N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-40 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.9.1* Fire Suppression System Code Requirements If an automatic or manual water-based fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be installed in accordance with the appropriate NFPA standards including the following: Complies with use of EEEEs This requirement was evaluated by NFPA 13 Code Compliance Evaluation and NFPA 15 Code Compliance Evaluation. Compliance is addressed below.
3.9.1 Fire Suppression System Code Requirements (1) NFPA 13, Standard for the Installation of Sprinkler Systems. Complies with use of EEEEs This requirement was evaluated by NFPA 13 Code Compliance Evaluation. See Implementation Item in Attachment S. CALC-ANO2-FP-09-00005, ANO Code Compliance Report for NFPA 13 1974 Edition, Rev. 10 CALC-ANO2-FP-09-00006, ANO Code Compliance Report for NFPA 13 1975 Edition, Rev. 1 CALC-ANO2-FP-09-00007, ANO Code Compliance Report for NFPA 13 1980 Edition, Rev. 1 CALC-ANO2-FP-09-00008, ANO Code Compliance Report for NFPA 13 1983 Edition, Rev. 1 CALC-ANOC-FP-08-00013, ANO Code Compliance Report for NFPA 13 1994 Edition, Rev. 1 3.9.1 Fire Suppression System Code Requirements (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. Complies with use of EEEEs This requirement was evaluated by NFPA 15 Code Compliance Evaluation. See Implementation Item in Attachment S. CALC-ANO2-FP-09-00004, ANO Code Compliance Report for NFPA 15 1977 Edition, Rev. 0 CALC-ANO2-FP-09-00020, ANO Code Compliance Report for NFPA 15 1982 Edition, Rev. 0 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-41 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.9.1 Fire Suppression System Code Requirements (3) NFPA 750, Standard on Water Mist Fire Protection Systems. N/A Systems addressed by NFPA 750 are not utilized at ANO.
3.9.1 Fire Suppression System Code Requirements (4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems. N/A Systems addressed by NFPA 16 are not utilized at ANO.
3.9.2 Fire Suppression System Flow Alarm Each system shall be equipped with a water flow alarm. Complies Water flow is alarmed by the system and/or fire pump start alarm. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.2 OP-2203.009, Fire Protection System Annunciator Corrective Action, Rev. 23, Panel 2C343 3.9.3 Fire Suppression System Alarm Locations All alarms from fire suppression systems shall annunciate in the control room or other suitable constantly attended location. Complies Water flow is alarmed by the system and/or fire pump start alarm. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.2 OP-2203.009, Fire Protection System Annunciator Corrective Action, Rev. 23, Panel 2C343 3.9.4 Fire Suppression System Diesel Pump Sprinkler Protection Diesel-driven fire pumps shall be protected by automatic sprinklers. Complies A wet pipe sprinkler system provides coverage within the Diesel Fire Pump Room. FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Fire Area INTAKE, Intake Structure, Section 32.1 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-42 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.9.5 Fire Suppression System Shutoff Controls Each system shall be equipped with an OS&Y gate valve or other approved shutoff valve. Complies This requirement was evaluated by NFPA 24 Code Compliance Evaluation. NUREG-0223, Section 4.3.1.3 states in part: "Sectionalizing valves are located on the loop to allow isolation of various sections for maintenance or repair." Also, "We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." CALC-ANOC-FP-09-00015, ANO Code Compliance Report for NFPA 24 1995 Edition, Rev. 1 NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 3.9.6 Fire Suppression System Valve Supervision All valves controlling water-based fire suppression systems required to meet the performance or deterministic requirements of Chapter 4 shall be supervised as described in 3.5.14. Complies by previous NRC approval NUREG-0223 (SER), Section 4.3.1.3 includes the following text: "Some fire water system control valves are electrically supervised; others are not electrically supervised, including those on the underground fire loop and at the fire pumps. The facility's technical specifications require a periodic check of the position of those valves which are not locked, sealed, electrically supervised, or otherwise secured in position to assure that valves are maintained in the open position... We find that the fire water piping system conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Sect ion 4.3.1.3 3.10 Gaseous Fire Suppression Systems N/A N/A Section header, no technical requirements. See subsections for specific compliance statements and references.
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-43 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.10.1 Gaseous Suppression System Code Requirements If an automatic total flooding and local application gaseous fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance with the following applicable NFPA codes: Complies This requirement was evaluated by NFPA 12A Code Compliance Evaluation. No other type of fixed gaseous suppression system is relied upon for safety related areas at ANO. CALC-85-D-2075-07, Calculation of Weight of Halon 1301 Required for CPC Room, Rev. 0, Complete CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 6.2.5 TDS398X0020, Halon 1301 System for CPC Control Room, Rev. 0, Complete 3.10.1 Gaseous Suppression System Code Requirements (1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems. N/A No fixed gaseous suppression systems other than Halon are relied upon for safety related areas at ANO.
3.10.1 Gaseous Suppression System Code Requirements (2) NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems. Complies This requirement was evaluated by NFPA 12A Code Compliance Evaluation. CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 3.10.1 Gaseous Suppression System Code Requirements (3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems. N/A No fixed gaseous suppression systems other than Halon are relied upon for safety related areas at ANO.
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-44 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.10.2 Gaseous Suppression System Alarm Location Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other constantly attended location identified. Complies Halon system only. This requirement was evaluated by NFPA 12A Code Compliance Evaluation. CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 24.3 OP-2203.009, Fire Protection System Annunciator Corrective Action, Rev. 23 3.10.3 Gaseous Suppression System Ventilation Limitations Ventilation system design shall take into account prevention from over-pressurization during agent injection, adequate sealing to prevent loss of agent, and confinement of radioactive contaminants. Complies The Halon design calculations have confirmed that requirements to prevent over-pressurization are satisfied. CALC-85-D-2075-07, Calculation of Weight of Halon 1301 Required for CPC Room, Rev. 0 CALC-95-E-0085-01, Halon Concentration in Unit 1 and Unit 2 Control Room Areas, Rev. 0 CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 SPEC-6600-M-2079, Specification for Halon 1301 System, Rev. 0 3.10.4* Gaseous Suppression System Single Failure Limits In any area required to be protected by both primary and backup gaseous fire suppression systems, a single active failure or a crack in any pipe in the fire suppression system shall not impair both the primary and backup fire suppression capability. N/A ANO-2 does not use a backup gaseous suppression system.
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-45 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.10.5 Gaseous Suppression System Disarming Controls Provisions for locally disarming automatic gaseous suppression systems shall be secured and under strict administrative control. Complies This requirement was evaluated by NFPA 12A Code Compliance Evaluation. ANO-2 TRM, ANO-2 Technical Requirements Manual, Rev. 50, Section 3.7.3, CPC Room Halon CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 3.10.6* Gaseous Suppression System CO 2 Limitations Total flooding carbon dioxide systems shall not be used in normally occupied areas. N/A ANO-2 does not use total flooding carbon dioxide systems in normally occupied areas.
3.10.7 Gaseous Suppression System CO 2 Warnings Automatic total flooding carbon dioxide systems shall be equipped with an audible pre-discharge alarm and discharge delay sufficient to permit egress of personnel. The carbon dioxide system shall be provided with an odorizer. N/A ANO-2 does not use total flooding carbon dioxide systems.
3.10.8 Gaseous Suppression System CO 2 Required Disarming Positive mechanical means shall be provided to lock out total flooding carbon dioxide systems during work in the protected space. N/A ANO-2 does not use total flooding carbon dioxide systems.
3.10.9 Gaseous Suppression System Cooling Considerations The possibility of secondary thermal shock (cooling) damage shall be considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide. Complies Halon 1301 has minimal cooling effect. CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 6.2.5 EN-DC-128, Fire Protection Impact Reviews, Rev. 5, Attachment 9.4, Step 1.2.m Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-46 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.10.10 Gaseous Suppression System Decomposition Issues Particular attention shall be given to corrosive characteristics of agent decomposition products on safety systems. Complies No corrosive characteristics from agent decomposition products have been identified. CALC-ANO2-FP-09-00019, ANO Code Compliance Report for NFPA 12A 1985 Edition, Rev. 0 3.11 Passive Fire Protection Features This section shall be used to determine the design and installation requirements for passive protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire. N/A General statement, no technical requirements. See subsections for specific compliance statements.
3.11.1 Building Separation Each major building within the power block shall be separated from the others by barriers having a designated fire resistance rating of 3 hours or by open space of at least 50 ft (15.2 m) or space that meets the requirements of NFPA 80A, Recommended Practice for Protection of Buildings from Exterior Fire Exposures. Exception:  Where a performance-based analysis determines the adequacy of building separation, the requirements of 3.11.1 shall not apply. Complies with use of EEEEs This requirement was evaluated by NFPA 80A Code Compliance Evaluation. CALC-ANOC-FP-08-00009, ANO Code Compliance Report for NFPA 80A 1996 Edition, Rev. 1 FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 6.4.5, Fire Barriers, Seals & Penetrations 3.11.2 Fire Barriers Fire barriers required by Chapter 4 shall include a specific fire-resistance rating. Fire barriers shall be designed and installed to meet the specific fire resistance rating using assemblies qualified by fire tests. The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of Fire Endurance of Building Construction and Materials, or ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Materials. Complies with use of EEEEs This requirement was evaluated by EC-1956, ANO-1 & 2 Fire Area/Fire Zone Compliance. NUREG-0223 Section 4.11 states: "We find that the fire barriers meet the objectives outlined in section 2.2 of this report and are therefore acceptable." (continued) EC-1956, ANO-1 & 2 Redefined Fire Areas / Fire Zones, Rev. 0 FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 6.4.5, Fire Barriers, Seals & Penetrations (continued)
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-47  NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.11.2 Fire Barriers (continued)
See Implementation Item in Attachment S. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.11 EC-1956, ANO-1 & 2 Redefined Fire Areas/Fire Zones, Rev. 0 3.11.3* Fire Barrier Penetrations* Penetrations in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire dampers having a fire resistance rating consistent with the designated fire resistance rating of the barrier as determined by the performance requirements established by Chapter 4. (See 3.11.3.4 for penetration seals for through penetration fire stops.) Passive fire protection devices such as doors and dampers shall conform with the following NFPA standards, as applicable: (see subsections) Exception:  Where fire area boundaries are not wall-to-wall, floor-to-ceiling boundaries with all penetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall be required to assess the adequacy of fire barrier forming the fire boundary to determine if the barrier will withstand the fire effects of the hazards in the area. Openings in fire barriers shall be permitted to be protected by other means as acceptable to the AHJ. Complies with use of EEEEs ANO complies with clarification in regards to NFPA 101 in that the features referenced in NFPA 101 are documented in the NFPA 80 and 90A Code Compliance Evaluations. NFPA 101 Section 8.2.3.2.1(a), with regards to rated fire door assemblies, refers to NFPA 80. NFPA 101 Section 9.2.1, with regards to rated fire dampers, refers to NFPA 90A. CALC-ANOC-FP-08-00006, ANO Code Compliance Report for NFPA 80 1999 Edition, Rev. 0 CALC-ANOC-FP-08-00010, ANO Code Compliance Report for NFPA 90A 1999 Edition, Rev. 0 FHA, ANO-1 & 2 Fire Hazards Analysis, Rev. 15, Section 6.4.5, Fire Barriers, Seals & Penetrations NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.9 3.11.3* Fire Barrier Penetrations (1) NFPA 80, Standard for Fire Doors and Fire Windows. Complies with use of EEEEs ANO complies with clarification in regards to NFPA 101 in that the features referenced in NFPA 101 are documented in the NFPA 80 and 90A Code Compliance Evaluations. NFPA 101 Section 8.2.3.2.1(a), with regards to rated fire door assemblies, refers to NFPA 80. CALC-ANOC-FP-08-00006, ANO Code Compliance Report for NFPA 80 1999 Edition, Rev. 0 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-48 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.11.3* Fire Barrier Penetrations (2) NFPA 90A, Standard for the Installation of Air-Conditioning and Ventilating Systems. Complies with use of EEEEs ANO complies with clarification in regards to NFPA 101 in that the features referenced in NFPA 101 are documented in the NFPA 80 and 90A Code Compliance Evaluations. NFPA 101 Section 9.2.1, with regards to rated fire dampers, refers to NFPA 90A. CALC-ANOC-FP-08-00010, ANO Code Compliance Report for NFPA 90A 1999 Edition, Rev. 0 3.11.3* Fire Barrier Penetrations (3) NFPA 101, Life Safety Code. Complies with use of EEEEs ANO complies with clarification in regards to NFPA 101 in that the features referenced in NFPA 101 are documented in the NFPA 80 and 90A Code Compliance Evaluations. NFPA 101 Section 8.2.3.2.1(a), with regards to rated fire door assemblies, refers to NFPA 80. NFPA 101 Section 9.2.1, with regards to rated fire dampers, refers to NFPA 90A. CALC-ANOC-FP-08-00006, ANO Code Compliance Report for NFPA 80 1999 Edition, Rev. 0 CALC-ANOC-FP-08-00010, ANO Code Compliance Report for NFPA 90A 1999 Edition, Rev. 0 3.11.4* Through Penetration Fire Stops Through penetration fire stops for penetrations such as pipes, conduits, bus ducts, cables, wires, pneumatic tubes and ducts, and similar building service equipment that pass through fire barriers shall be protected as follows. Complies with use of EEEEs Through Penetration Fire Stops were approved initially by the NRC as documented in NUREG-0223 Section 4.9. Subsequent EEEE's have been documented in Table B-3 by fire area. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.9 3.11.4* Through Penetration Fire Stops (a) The annular space between the penetrating item and the through opening in the fire barrier shall be filled with a qualified fire-resistive penetration seal assembly capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period. Complies with use of EEEEs Through Penetration Fire Stops were approved initially by the NRC as documented in NUREG-0223 Section 4.9. Subsequent EEEE's have been documented in Table B-3 by fire area. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.9
 
Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-49  NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.11.4* Through Penetration Fire Stops (b) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier through opening fire stop and shall be permitted to be installed on either side of the barrier in a location that is as close to the barrier as possible. Exception:  Openings inside conduit 4 in. (10.2 cm) or less in diameter shall be sealed at the fire barrier with a fire-rated internal seal unless the conduit extends greater than 5 ft (1.5 m) on each side of the fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. (5.1 cm) shall constitute an acceptable smoke and hot gas seal in this application. Complies with use of EEEEs Through Penetration Fire Stops were approved initially by the NRC as documented in NUREG-0223 Section 4.9. Subsequent EEEE's have been documented in Table B-3 by fire area. NUREG-0223, Fire Protection Safety Evaluation Report for ANO-2, 8/30/1978, Section 4.9 Arkansas Nuclear One - Unit 2 Att. A - NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure 1 to 2CAN121202 Page A-50 NFPA 805 Chapter 3 Section Requirements / Guidance Compliance Statement Compliance Basis Reference Document 3.11.5* Electrical Raceway Fire Barrier Systems (ERFBS) ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area. ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Area."  The ERFBS needs to adequately address the design requirements and limitations of supports and intervening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size, fill, and type shall be demonstrated. Exception No. 1:  When the temperatures inside the fire barrier system exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86-10, "Fire Endurance Acceptance Test Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Training Within the Same Fi re Area," Supplement 1, functionality of the cable at these elevated temperatures shall be demonstrated. Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems Du ring and After Fire Endurance Test Exposure." Exception No. 2:  ERFBS systems employed prior to the issuance of Generic Letter 86-10, Supplement 1, are acceptable providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at the time of acceptance. N/A ANO does not credit Electrical Raceway Fire Barrier Systems.
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-1 B. NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
 
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the mal-operation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliabili ty of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance 3  Deterministic This section discusses a generic deterministic methodology and criteria that licensees can use to perform a post-fire Methodology safe shutdown analysis to address regulatory requirements. The plant-specific analysis approved by NRC is reflecte d 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 Sectio ns III.G.1, III.G.2 and III.G.3 apply to equipment and cables required for achieving and maintaining safe shutdown in any fire area. Although equipment and cables for fire detection and suppression systems, communications systems and 8-hour emergency lighting systems are important features, this guidance document does not address them.
Additional information is provided in Appendix B to this document.
 
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns The Appendix R Analysis at ANO was revalidated based on guidance provided in NEI-00-01, Revision 1. This is documented in various calculations.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, All CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, All Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-2 NEI 00-01 Ref NEI 00-01 Guidance 3.1  [A, Intro] Safe This section discusses the identification of systems available and necessary to perform the required safe shutdown Shutdown Systems functions. It also provides information on the process for combining these systems into safe shutdown paths. Path Development 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.1.b requires that repairs to system s and equipment necessary to achieve and maintain cold shutdown be completed within 72 hours. It is the intent of the NRC that requirements related to the use of manual operator actions will be addressed in a forthcoming rulemaking. [Refer to hard copy of NEI 00-01 for Figure 3-1]
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The systems to accomplish functional requirements to ensure performance goals are met have been developed in Safe Shutdown Equipment List (SSEL) and Safe Shutdown Capability Assessment (SSCA) calculation.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 5.2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-3 NEI 00-01 Ref NEI 00-01 Guidance 3.1  [B, Goals] Safe The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and compo nents Shutdown Systems remains free of fire damage for a single fire in any single plant fire area. This goal is accomplished by det ermining Path Development those functions important to achieve and maintain hot shutdown. Safe shutdown systems are selected so that th e capability to perform these required functions is a part of each safe shutdown path. The functions important to post-fire safe shutdown generally include, but are not limited to the following:
Reactivity control Pressure control systems Inventory control systems Decay heat removal systems Process monitoring Support systems  Electrical systems  Cooling systems
 
These functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown, which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment. If these functions are preserved, then the plant will be safe because the fuel, the reactor, and the primary containment will not be damaged. By assuring that this equipment is not damaged and remains functional, the protection of the health and safety of the public is assured.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent The systems required to ensure performance goals are met have been developed in the SSEL and SSCA. The selection of components and system is based on achieving the performance goals in each fire area. There is no specific shutdown path.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 5.3
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-4 NEI 00-01 Ref NEI 00-01 Guidance 3.1  [C, Spurious Operations] In addition to the above listed functions, Generic Letter (GL) 81-12 specifies consideration of associated circuits Safe Shutdown Systems with the potential for spurious equipment operation and/or loss of power source, and the common enclosure and Path Development failures. Spurious operations/actuations can affect the accomplishment of the post-fire safe shutdown fun ctions listed above. Typical examples of the effects of the spurious operations of concern are the following:
A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability  A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdown path.
Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment. Common power source and common enclosure concerns could also affect these and must be addressed.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns The cable selection and circuit analysis for 10 CFR 50, Appendix R, safe shutdown components for ANO considers spurious operation due to associated circuits, common power supplies, and for common enclosures.
A special subset of components considered for spurious operation involves reactor coolant pressure boundary components whose spurious operation can lead to an unacceptable loss of reactor pressure vessel / RCS inventory via an interfacing system loss of coolant accident. These components are defined as high/low pressure interface valves and are subject to more stringent circuit analysis. This high/low pressure interface boundary valve definition is in alignment with those in NEI 00-01, NEI 00-01 Appendix C. and FAQ 06-0006 to NEI 04-02, but is limited to those components potentially subject to ISLOCAs in excess of makeup capability.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.3.6 CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 7, 8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-5 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1  Criteria / Assumptions The following criteria and assumptions may be considered when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL)
Methodology, Rev. 1, Section 4 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.1  [GE BWR Paths] [BWR] General Electric (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.
Applicability Comments  Not Applicable ANO-2 is a Combustion Engineering (CE) 2-loop, U-tube SG PWR with a dry ambient pressure containment and Operating License (OL) NPF-6 issued 09/01/1978.
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-6 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.2  [SRVs / LP Systems] [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 (Low Pressure Injection, Reactor Building Spray) 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 a SER dated Dec. 12, 2000.
 
Applicability Comments  Not Applicable ANO-2 is a CE 2-loop, U-tube SG PWR with a dry ambient pressure containment and OL NPF-6 issued 09/01/1978.
NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.3  [Pressurizer Heaters] [PWR] GL 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/c harging pumps). Hot shutdown conditions can be maintained via natural circulation of the RCS through the SGs. The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Fire Safe Shutdown credited the use of one SG using a natural circulation cooldown, with no credit for pressurizer heaters, RCPs tripped, charging available, and EFW aligned to a SG. Cooldown accomplished using atmospheric dump valves (ADVs) and controlling SG level.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL)
Methodology, Rev. 1, Section 6
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-7 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.4  [Alternative Shutdown  The classification of shutdown capability as alternative shutdown is made independent of the selection of systems Capability] used for shutdown. Alternat ive shutdown capability is determined based on an inability to assure the availability of a redundant safe shutdown path. Compliance to the separation requirements of Sections III.G.1 and III.G.2 may be supplemented by the use of manual actions to the extent allowed by the regulations and the licensing basis of the plant, repairs (cold shutdown only), exemptions, deviations, GL 86-10, fire hazards analyses, or fire protection design change evaluations, as appropriate. These may also be used in conjunction with alternative shutdown capability.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Determination of alternate shutdown capability is determined when all process monitoring and control function are no longer available from the Main Control Room due to a fire.
Credit for alternate shutdown capability is only taken for Fire Area G.
Comments  Alternate shutdown capability is monitored in the Technical Support Center (TSC) using the Safety Parameter Display System (SPDS). Operators are used throughout the plant to achieve shutdown.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.34 OP-2203.014, Alternate Shutdown, Rev. 25, 7/6/2011, All
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-8 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.5  [Initial Conditions] At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and available for post-fire safe shutdown. Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation (LCO), etc. in progress. The units are assumed to be operating at full power under normal conditions and normal lineups.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All safe shutdown systems are assumed available. The majority of equipment relied upon for safe shutdown is safety related and/or governed by TS or TRM operability/functionality requirements.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.2.4 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.6  [Other Events in  No SAR accidents or other design basis events (e.g. LOCA, earthquake), single failures Conjunction with Fire] or non-fire induced transients need be considered in conjunction with the fire.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns No other external events, accidents, or failures unrelated to the fire are assumed to occur concurrently with the postulated fire or any subsequent activities to achieve cold shutdown conditions.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 5.0 1) CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.1.1 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-9 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.7  [Offsite Power] For the case of redundant shutdown, offsite power may be credited 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 (LOOP) if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a LOOP. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for 72 hours.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Except where demonstrated by analysis to remain available post-fire, offsite power may or may not be available when addressing fire effects. Thus, it cannot be assumed that a loss of offsite power (LOOP) will occur and cause components to fail to their safe shutdown position.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 5.0 3)
CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.2.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-10 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.8  [Safety-Related Post-fire safe shutdown systems and components are not required to be safety-related. Equipment]
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Safe Shutdown Equipment (SSE) may or may not be safety-related.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.3.14 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.9 [72 Hour Coping] The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/tri
: p. Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis. At least one train can be repaired or made operable within 72 hours using onsite capability to achieve cold shutdown.
Applicability Comments  Applicable For alternate shutdown Appendix R, III.L, requires that cold shutdown be achieved in 72 hours. Appendix R, III.G.1.
b, requires that systems necessary to achieve and maintain cold shutdown can be repaired and be operable in 72 hours.
 
Alignment Statement Alignment Basis Aligns In the current Appendix R Licensing Basis, systems necessary to achieve and maintain cold shutdown (CSD) from either the control room or emergency control station(s) can be repaired within 72 hours, and systems necessary to maintain hot shutdown must be available to achieve and maintain hot shutdown until repairs are complete and cold shutdown can be initiated. For alternate shutdown, systems that are required to achieve cold shutdown within 72 hours must be made available. Transition to CSD is not required for Safe and Stable under a NFPA 805 Licensing Basis and repair of non-risk significant equipment for CSD will be addressed post transition in site procedures.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Section 2 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 4.2.2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-11 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.10  [Manual / Automatic Manual initiation from the main control room or emergency control stations of systems required to achieve and Initiation of Systems] maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC; automati c initiation of systems selected for safe shutdown is not required, but may be included as an option.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Only manual initiation of systems required to achieve and maintain safe shutdown is credited. Automatic operation of specific components within these systems is credited where appropriate (such as minimum flow valves). In general, automatic initiation of systems required to achieve and maintain safe shutdown (i.e., Safety Injection Actuation Signal initiation) is not credited unless the initiation signals are shown to be free of fire damage. However, fire induced automatic initiation signals are evaluated for the possibility of spurious component operation and their subsequent adverse impact on safe shutdown.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Section 3.1 CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.2.6
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-12 NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.11  [Multiple Affected Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve an d maintain safe Units] shutdown for each affected unit must be demonstrated.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent ANO-1 and ANO-2 do not share equipment required to meet performance goals for control of reactivity, inventory, pressure, and decay heat removal.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachments 4 through 72 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2  Shutdown Functions The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown. For additional information on BWR system selection, refer to GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths for the BWR."
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 5.2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-13 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.1  Reactivity Control [BWR] Control Rod Drive System The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability. Manual scram/reactor trip is credited. The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor.
 
[PWR] Makeup/Charging A method for ensuring that adequate shutdown margin is maintained is provided by ensuring borated water is utilized for RCS makeup/charging.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Adequate SDM for cold shutdown is assured using the BAMT and RWT aligned to the charging pump suction or, should the RWT dump to Containment Building sump, by aligning HPSI pump after RCS pressure has been lowered.
 
Comments  To ensure the reactor remains subcritical with a K eff < 0.99, a specified minimum amount of borated water from the BAMT is defined by plant procedures to be injected into the RCS.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Table 2 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 6.1.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-14 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.2  Pressure Control The systems discussed in this section are examples of systems that can be used for pressure control. This does not Systems restrict the use of other systems for this purpose.
[BWR] Safety Relief Valves (SRVs)
The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These are operated manually. Automatic initiation of the Automatic Depressurization System is not a required function.
 
[PWR]  Makeup/Charging RCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions are made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure. Pressure increases are made by initiating charging/makeup to maintain pressurizer level/pressure. Manual control of the related pumps is acceptable.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns RCS pressure will lower slowly as the pressurizer cools due to ambient losses wi th no pressurizer heaters in operation. Additionally, depressurization methods credited are the auxiliary spray, pressurizer high point vents, or Emergency Core Cooling System (ECCS) vents.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Table 2 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, Sections 6.1.4 and 6.1.5
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-15 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.3  Inventory Control [BWR] Systems selected for the inventory control function should be capable of supplying sufficient reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.
 
[PWR] Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown. Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The charging pumps, aligned to either the RWT or BAMT, can be used to control RCS level. If the RWT dumps to the containment sump directly due to spurious valve operation, the HPSI pump can be aligned, provided RCS pressure has been lowered.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Table 2 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 6.1.3
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-16 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.4  Decay Heat Removal [BWR] Systems selected for the decay heat removal (DHR) 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 reactor building (suppression pool).
Removing sufficient decay heat from the reactor to achieve cold shutdown.
 
[PWR] Systems selected for the decay heat removal function(s) should be capable of:
Removing sufficient decay heat from the reactor to reach hot shutdown conditions. Typically, this entails utilizing natural circulation in lieu of forced circulation via the RCPs and controlling steam release via the ADVs. Removing sufficient decay heat from the reactor to reach cold shutdown conditions.
 
This does not restrict the use of other systems.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Decay heat removal is accomplished using natural circulation cooldown. It entails aligning EFW to at least one SG with initial pressure control provided by the MSSVs to achieve Hot Standby. To achieve cold shutdown, the ADVs are used until the RCS pressure and temperature requirements are met to align the SDC system.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Table 2 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, Section 6.1.6, 6.1.7
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-17 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.5 Process Monitoring The process monitoring function is provided for all safe shutdown paths. NRC Information Notice (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 III.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 III.G.2). In general , process monitoring instruments similar to those lis ted below are needed to successfully use existing operating procedures (including Abnormal Operating Procedures).
BWR Reactor coolant level and pressure Diagnostic instrumentation for safe shutdown systems Suppression pool level and temperature Level indication for tanks needed for safe shutdown Emergency or isolation condenser level PWR Reactor coolant temperature (hot leg / cold leg) Level indication for tanks needed for safe shutdown Pressurizer pressure and level Steam generator level and pressure Neutron flux monitoring (source range) Diagnostic instrumentation for safe shutdown systems
 
The specific instruments required may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs. prescriptive), and systems and paths selected for safe shutdown.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The instrumentation selected is based on the guidance of NRC IN 84-09 and NRC RG 1.189, which identify the minimum monitoring capability considered necessary for a PWR.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 6.1.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-18 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6  Support Systems [Blank Heading - No specific guidance]
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document References provided in subsequent subsections.
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-19 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.1  Electrical Systems AC Distribution System Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E busses either directly from the busses or through step-down transformers/load centers/distribution panels for 600, 480 or 120 VAC loads. For redundant safe shutdown performed in accordance with the requirements of Appendix R, Section III.G.1 and 2, power may be supplied from either offsite power sources or the Emergency Diesel Generator (EDG) depending on which has been demonstrated to be free of fire damage. No credit should be taken for a fire causing a LOOP. Refer to Section 3.1.1.7.
DC Distribution System Typically, the 125 VDC distribution system supplies DC control power to various 125 VDC control panels including switchgear breaker controls. The 125 VDC distribution panels may also supply power to the 120 VAC 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 EDGs to become operational. Once the EDGs are operational, the 125 VDC distribution system can be powered from the EDGs via the battery chargers.
[BWR] Certain plants are also designed with a 250 VDC Distribution System that supplies power to Reactor Core Isolation Cooling and/or High Pressure Coolant Injection equipment.
The DC control centers may also supply power to various small horsepower Appendix R safe shutdown system valves and pumps. If the DC system is relied upon to support safe shutdown without battery chargers being available, it must be verified that sufficient battery capacity exists to support the necessary loads for sufficient time (either until power is restored, or the loads are no longer required to operate).
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The AC and DC distribution systems are credited in order to meet performance goals and functions. The safeguards 4kV buses can either be aligned to the EDGs or available offsite power.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 6.1.9.1 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-20 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 [B]  Cooling Systems Heating, Ventilation, and Air Conditioning (HVAC) Systems HVAC Systems may be required to assure that safe shutdown equipment remains within its operating temperature range, as specified in manufacturer's literature or demonstrated by suitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents).
 
HVAC systems may be required to support safe shutdown system operation, based on plant-specific configurations. Typical uses include:
 
Main control room, cable spreading room, relay room ECCS pump compartments Diesel generator rooms Switchgear rooms Plant-specific evaluations are necessary to determine which HVAC systems are essential to safe shutdown equipment operation.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns An active HVAC system is required for EDGs and Control Room. Passive dampers have been included to account for their different position based on seasonal changes.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Table 2 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 6.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-21 NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 [A] Cooling Systems Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations. Typical uses include:
 
RHR (residual heat removal) / SDC (shutdown cooling) / DHR Heat Exchanger cooling water Safe shutdown pump cooling (seal coolers, oil coolers) Diesel generator cooling HVAC system cooling water Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Cooling water systems are credited to support EDGs, SDC, and HVAC systems.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1, Table 2 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 6.1.9.2 NEI 00-01 Ref NEI 00-01 Guidance 3.1.3  Methodology for Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and Shutdown System developing the shutdown paths. The following methodology may be used to define the safe shutdown systems and Selection paths for an Appendix R analysis (refer to hard copy of NEI 00-01 for Figure 3-2).
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns with intent The selection of components and system is based on achieving the performance goals in each fire area. There is no specific shutdown path, both red and green train SSE can be used to accomplish performance goals.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 5.3 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-22 NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.1  Identify Safe Review available documentation to obtain an understanding of the available plant systems and the functions required Shutdown to achieve and maintain safe shutdown. Documents such as the following may be reviewed: Functions Operating Procedures (Normal, Emergency, Abnormal)
System descriptions Fire Hazard Analysis (FHA)
Single-line electrical diagrams Piping and Instrumentation Diagrams (P&IDs) [BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR" Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Various documents were reviewed to determine equipment required to achieve and maintain safe shutdown. The Operation department was consulted in selection process.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 1 Table 2 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, Sections 3.3 through 3.7
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-23 NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.2  Identify Combinations  Given the criteria/assumptions defined in Section 3.1.1, identify the available combinations of systems capable of of Systems that achieving the safe shutdown functions of reactivity control, pressure control, inventory control, decay heat Satisfy Each Safe removal, process monitoring, and support systems such as electrical and cooling systems (refer to Section 3.1
.2). Shutdown Function This selection process does not restrict the use of other systems. In addition to achieving the required saf e shutdown functions, consider spurious operations and power supply issues that could impact the required safe shutdown function.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns The available combination of systems/components capable of achieving and maintaining the safe shutdown functions are depicted on the fault trees.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, All CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sections 6 and 7
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-24 NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.3  Define Combinations  Select combinations of systems with the capability of performing all of the required safe shutdown functions and of Systems for Each designate this set of systems as a safe shutdown path. In many cases, safe shutdown paths may be defined o n a Safe Shutdown Path divisional basis since the availability of electrical power and other support systems must be demonstrated f or each path.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent The logical relationship of safe shutdown equipment is established in the fault trees and illustrates how th is equipment functions together to achieve and maintain safe shutdown.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.4 NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.4  Assign Shutdown Assign a path designation to each combination of systems. The path will serve to document the combina tion of Paths to Each systems relied upon for safe shutdown in each fire of area. Refer to Attachment 1 to this document (NEI 00-01) f or an Combination of example of a table illustrati ng how to document the various combinations of systems for selected shutdown paths.
Systems  Applicability Comments  Applicable Attachment refers to BWROG guidance on post-fire safe shutdown.
Alignment Statement Alignment Basis Aligns with intent Path designation are not used. The logical relationship of safe shutdown equipment is established in the fault trees and illustrates how this equipment functions together to achieve and maintain safe shutdown.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.4 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-25 NEI 00-01 Ref NEI 00-01 Guidance 3.2  Safe Shutdown The previous section described the methodology for selecting the systems and paths necessary to achieve and Equipment Selection maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix R function. The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the same safe shutdown path as that system. The list of safe shutdown equipment will then form the basis for identifying the cables necessary for the operation or that can cause the mal-operation of the safe shutdown systems.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The SSEL is used as input into the development of the Safe Shutdown Fault Trees. The SSEL is also used to identify those components requiring post-fire safe shutdown circuit analysis.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 4.1
 
NEI 00-01 Ref NEI 00-01 Guidance 3.2.1 Criteria / Assumptions Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 4.2 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-26 NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.1  [Primary / Safe shutdown equipment can be divided into two categories. Equipment may be categorized as (1) primary Secondary components or (2) secondary components. Typically, the following types of equipment are considered to be primary Components] components:
Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc. All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder), power supplies, or other electrical components that support operation of primary components (i.e., EDGs, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).
 
Secondary components are typically items found within the circuitry for a primary component. These provide a supporting role to the overall circuit function. Some secondary components may provide an isolation function or a signal to a primary component via either an interlock or input signal processor. Examples of secondary components include flow switches, pressure switches, temperature switches, level switches, temperature elements, speed elements, transmitters, converters, controllers, transducers, signal conditioners, hand switches, relays, fuses and various instrumentation devices.
Determine which equipment should be included in the SSEL. As an option, include se condary components with a primary component(s) that would be affected by fire damage to the secondary componen
: t. By doing this, the SSEL can be kept to a manageable size and the equipment included on the SSEL can be readily related to required post-fire safe shutdown systems and functions.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns In general, for electrical supervised equipment, the following criteria were used. Where a component received an actuation signal, it may have been determined to add the component generating a spurious signal as a primary component to aid in circuit analysis. Additionally, manual and check valves that are in the flow path, or act as a boundary valve, are included as primary components.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.2 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-27 NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.2  [Fire Damage to Assume that exposure fire damage to manual valves and piping does not adversely impact their ability t o perform Mechanical their pressure boundary or safe shutdown function (heat sensitive piping materials, including tubing with brazed or Components (not soldered joints, are not included in this assumption). Fire damage should be evaluated with respect to the abi lity electrically supervised)] to manually open or close the valve should this be necessary as a part of the post-fire safe shutdown scenario.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Mechanical Components susceptible to fire damage (brazed or soldered instrument lines, instrument tubing for credited instruments, etc.) are identified and evaluated on a fire area basis. This may take the form of a standalone evaluation or may be incorporated into the SSEL.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 5.0 8)
CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.2.l NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.3  [Manual Valve Positions] Assume that manual valves are in their normal position as shown on P&IDs or in the plant operating procedures.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The normal operating position of the component is determined by a review of operating procedures.
Comments  The normal operating position of the component is entered (for components which have more than one normal position, e.g., a pump which may or may not be operating during normal operation, both positions shall be entered). These positions are determined by a review of operating procedures.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.3.j Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-28 NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.4  [Check Valves] Assume that a check valve closes in the direction of potential flow diversion and seats properly with sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used for inventory control, decay heat removal, equipment cooling or other related safe shutdown functions.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Properly oriented check valves are assumed to provide adequate isolation when credited for boundary isolation.
Comments  Properly oriented check valves are reviewed to ensure, as system operating parameters change, that system pressure ensuring properly oriented check valve is acting as credited isolation, does not open to introduce undesired fluids or gases into safe shutdown system (i.e., gas binding of pumps, boration dilution event).
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.2.b
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-29 NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.5  [Instrument Failures] Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and f low transmitters) are assumed to fail upscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired signal to the control circuit.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Damage to instrument cables is assumed to fail the instrument in the least desirable state. That is, the instrument could fail high, low, or in some intermediate condition.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Re
: v. 1, Section 6.1.5.2 NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.6  [Spurious Identify equipment that could spuriously operate or mal-operate and impact the performance of equipment on a Components] required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Circuit analysis was performed on each electrically supervised potentially spurious component. Included in analysis was whether cable fault mechanism was due to either an intra- or inter-cable hot short to consider Bin 1 of NRC Regulatory Issue Summary (RIS) 2004-03 identification.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6.1.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-30 NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.7  [Instrument Tubing] Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a result of fire. Determine and consider the fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Instrument tubing was evaluated to determine effect of exposure fire in addition to cable failure effects.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Sect. 5.0 subsections 7) & 8):  Attachment 8.39
 
NEI 00-01 Ref NEI 00-01 Guidance 3.2.2  Methodology for Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown equipment. Equipment Selection Use the following methodology to select the safe shutdown equipment for a post-fire safe shutdown analysis (refer to hard copy of NEI 00-01 for Figure 3-3).
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 5.3
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-31 NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.1  Identify the System  Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each shutdown Flow Path for Each path. Refer to Attachment 2 fo r an example of an annotated P&ID illustrating this concept. Shutdown Path
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Highlighted flow path of P&ID was performed.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-32 NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.2  Identify the Equipment  Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to assure that in Each Safe Shutdown all equipment in each system's flow path has been identified. Assure that any equipment that could spuriously System Flow Path operate and adversely affect the desired system function(s) is also identified. If additional systems are ide ntified Including Equipment which are necessary for the operation of the safe shutdown system under review, include these as systems That May Spuriously required for safe shutdown. Designate these new systems with the same safe shutdown path as the primary sa fe Operate and Affect shutdown system under review (refer to Figure 3-1). System Operation Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent The SSEL identifies the minimum set of plant equipment that is required to demonstrate the plant's ability t o achieve and maintain post-fire safe shutdown for all applicable areas of the plant. To develop the list, a thorough review of plant documents, including P&IDs, System Training Manuals, Normal and Abnormal Operating Procedures, and the SAR was conducted. The SSEL is the result of an iterative process including component selection, circuit analysis, and area compliance assessments. Safe shutdown paths are not designated; method to achieve and maintain safe shutdown is shown on fault trees for each fire area.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Me thodology, Rev. 1, Section 4.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-33 NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.3  Develop a List of Safe Prepare a table listing the equipment identified for each system and the shutdown path that it supports. Identify Shutdown Equipment any valves or other equipment that could spuriously operate and impact the operation of that safe shutdown and Assign the system. Assign the safe shutdown path for the affected system to this equipment. During the cable selection Corresponding System phase, identify additional equipment required to support the safe shutdown function of the path (e.g., electrical and Safe Shutdown distribution system equipment). Include this additional equipment in the safe shutdown equipment list. Path(s) Designation Attachment 3 to this document provides an example of a SSEL.
The SSEL identifies the list of equipment wit hin to Each the plant considered for safe shutdown and it documents various equipment-related attributes used in the analysis.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent The SSEL is controlled in the Plant Data Management System (PDMS). A shutdown path is not used, but instead a method to achieve and maintain safe shutdown is shown on a fault tree available for each fire area.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL)
Methodology, Rev. 1, Section 7
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-34 NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.4  Identify Equipment Collect additional equipment-related information necessary for performing the post-fire safe shutdown analysis for Information Required the equipment. In order to facilitate the analysis, tabulate this data for each piece of equipment on the SSEL. for the Safe Shutdown Refer to Attachment 3 to this document for an example of a SSEL. Examples of related equipment data shou ld Analysis include the equipment type, equipment description, safe shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment. Other information such as the following may be useful in performing the safe shutdown analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concern, and spurious operation concern.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent Similar information was obtained for each safe shutdown component. No safe shutdown paths are designated, but the method to achieve and maintain safe shutdown is shown on a fault tree for each fire area.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, All
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Enclosure 1 to 2CAN121202 Page B-35 NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.5  Identify Dependencies In the process of defining equipment and cables for safe shutdown, identify additional supporting equipment such Between Equipment, as electrical power and interlocked equipment. As an aid in assessing identified impacts to safe shutdown, Supporting Equipment, consider modeling the dependency between equipment within each safe shutdown path either in a relational Safe Shutdown database or in the form of a Safe Shutdown Logic Diagram (SSLD). Attachment 4 provides an example of a Systems and Safe SSLD that may be developed to document these relationships. Shutdown Paths
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns with intent In the process of defining equipment and cables for safe shutdown, identification of additional supporting equipment such as electrical power and interlocked equipment was completed and is contained within the SSEL in PDMS. A safe shutdown path is not identified for each safe shutdown component. A fault tree is used to identify dependency between components.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.37 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL)
Methodology, Rev. 1, Sections 4.3.1.4, 4.3.1.6, and 5.3.4
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-39 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the mal-operation of the equipment identif ied in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance cr iteria.  (a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection dev ice (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is th at the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.
NEI 00-01 Ref NEI 00-01 Guidance 3.3  Safe Shutdown Cable This section provides industry guidance on the recommended methodology and criteria for selecting safe shutdown Selection and Location cables and determining their potential impact on equipment required for achieving and maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the proper operation or that could cause the mal-operation of safe shutdown equipment are identified and that these cables are properly related to the safe shutdown equipment whose functionality they could affect. Through this cable-to-equipment relationship, cables become part of the safe shutdown path assigned to the equipment affected by the cable.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns This is introductory guidance information and contains no specific guidance.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, All Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-40 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1  Criteria / Assumptions To identify an impact to safe shutdown equipment based on cable routing, the equipment must have cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe shutdown system equipment.
Consider the following criteria when selecting cables that impact safe shutdown equipment:
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-41 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.1  [Cable Selection] The list of cables whose failure could impact the operation of a piece of safe shutdown equipment includes more than those cables connected to the equipment. The relationship between cable and affected equipment is based on a review of the electrical or elementary wiring diagrams. To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate the power, control, instrumentation, interlock, and equipment status indication cables related to the equipment. Consider reviewing additional schematic diagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of the equipment to operate as required in support of post-fire safe shutdown. As an option, consider applying the screening criteria from Section 3.5 as a part of this section. For an example of this see Section 3.3.1.4.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All cables including those from interlocks, instruments, and power supplies that could potentially adversely impact the desired operation of a safe shutdown component are listed. This includes cables external to the component control circuit if any cable fault could adversely impact the required state of the component unless the cable(s) are included with another safe shutdown component. Primary scheme cables for each safe shutdown component are listed and any reasons to exclude that cable is documented. The required drawings to perform and verify the cable selection and circuit analysis include the P&ID showing the component, the schematic, and others as required.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev.
1, Section 6.1.5, 6.1.12
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-42 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.2  [Cables Affecting  In cases where the failure (including spurious actuations) of a single cable could impact more than one piece of Multiple Components] safe shutdown equipment, include the cable with each piece of safe shutdown equipment.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All cables including those from interlocks, instruments, and power supplies that could potentially adversely impact the desired operation of a safe shutdown component are listed. Circuit analysis is completed on a component level; where a cable may affect several safe shutdown components, these cables are assigned to those safe shutdown component's circuit analysis.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6.1.5
 
NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.3  [Isolation Devices] Electrical devices such as relays, switches and signal resistor units are considered to be acceptable isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Circuit isolation devices credit coordinated fusing, normally open operator controlled contacts, and other isolation devices as noted in PDMS, to determine if cables are not required and will not impact a safe shutdown component and prevent its safe shutdown function.
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6.1.6 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-43 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.4  [Identify "Not Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e., annunciator circuits, Required" Cables] space heater circuits and computer input circuits) unless some reliance on these circuits is necessary. Howe ver, they must be isolated from the component's control scheme in such a way that a cable fault would not impact the performance of the circuit.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Cable that is listed on main schemes are screened out as not required, if they will not affect the safe shutdown function of a safe shutdown component. These cables are typically for motor space heater, testing, annunciator, or computer inputs.
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6.1.6
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-44 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.5  [Identification of For each circuit requiring power to perform its safe shutdown function, identify the cable supplying power to each safe Power Supplies] shutdown and/or required interlock component. Initially, identify only the power cables from the immediate ups tream power source for these interlocked circuits and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distribution system to capture the remaining equipment from the electrical power distribution system necessary to support delivery of power from either the offsite power source or the emergency diesel generators (i.e., onsite power source) to the safe shutdown equipment. Add this equipment to the safe shutdown equipment list. Evaluate the power cables for this additional equipment for associated circuits concerns.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent Power supply cable selection shall typically end at the closest electrical isolation device for the componen t identified in the SSEL. For instance, power supply cables to a motor control center (MCC) will not be listed for a motor operated valve, only the power supply cable from the MCC to the valve will be listed. The MCC would be identified as a safe shutdown component in the SSEL and a separate circuit analysis is performed for the MCC.
The circuit analysis for MCC or switchgear (SWGR) shall include the following as appropriate:
 
Feed power cable Feed circuit breaker control circuit 125 VDC control power Spurious actuation of the undervoltage coils Non-safe shutdown load power cables
 
Control power cables may be listed against the switchgear (when appropriate), and the switchgear availability is linked to the individual component's availability in the safe shutdown fault trees; therefore, control power cables need not be included with the circuits selected for each individual load breaker powered by that switchgear.
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-45 Comments  Breaker coordination assures that the protective device nearest the fault operates prior to operation of upstream devices. On switchgear and/or load centers where breaker coordination relies on relays, coordination may fail if control power or breaker control cables are lost; therefore, load power cables are assigned to switchgear as required so the analyst may verify that breaker control is not lost by ensuring breaker control cables are not impacted and that control power is available to trip breakers, thus ensuring proper coordination. This review may take place in the fire area compliance document or may be documented in circuit selection/analysis by revising circuit analysis to add cables to SSE as required.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sections 6.1.5.3 & 7.1 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.6  [ESFAS Initiation] The automatic initiation logics for the credited post-fire safe shutdown systems are not required to support safe shutdown. Each system can be controlled manually by operator actuation in the main control room or emergency control station. If operator actions outside the main control room are necessary, those actions must conform to the regulatory requirements on manual actions. However, if not pr otected from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely affect any post-fire safe shutdown system function.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Instruments which do not provide a credited control function, but whose spurious operation could adversely affect safe shutdown are considered to be required safe shutdown components. Examples include instrumentation involved in the initiation of the ESFAS automatic control logics. The population of cables that are involved with the automatic initiation logics are identified in the safe shutdown analysis, and the instrumentation is depicted on the fault trees Any manual action to recovery components due to actuation signal are evaluated for feasibility.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, All CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Re
: v. 1, Section 6.1.5.2 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-46 NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.7  [Circuit Coordination] Cabling for the electrical distribution system is a concern for those breakers that feed associated circuits and are not fully coordinated with upstream breakers. With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component. For example, the power feed to a pump is necessary to support the pump. Similarly, the power feed from the load center to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cables discussed above would also support the power supply. For example, the power feed to the pump discussed above would support the bus from which it is fed because, for the case of a common power source analysis, the concern is the loss of the upstream power source and not the connected load. Similarly, the cable feeding the MCC from the load center would also be necessary to support the load center.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Breaker coordination is ensured by reviewing the time current curves from the plant's coordination study to ensure coordination. Coordination assures that the protective device nearest the fault operates prior to operation of upstream devices. The means of assuring circuit protection and coordination is provided in a series of calculations. These calculations demonstrate that the Class 1E and non-Class 1E power supplies credited for safe shutdown compliance do have adequate coordination.
 
Comments  Molded case breakers less than 600V do not require a separate power source to ensure protective features remain available. Breakers for 480V load centers and medium voltage SWGR (4,160V & 6,900V) require DC control power for the protective relaying necessary to assure coordination. This control power is a required input in the fault sub-trees associated with the availability of the aforementioned distribution equipment. The non-safe shutdown loads energized from switchgear that rely on DC control power for relay and metering circuits are deemed required circuits (see comments for alignment statement NEI 00-01 Section 3.3.1.5.).
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 7.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-47 NEI 00-01 Ref NEI 00-01 Guidance 3.3.2  Associated Circuit Appendix R, Section III.G.2, requires that separation f eatures be provided for equipment and cables, including Cables associated non-safety circuits that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve hot shutdown. The three types of associated circuits were identified in Reference 6.1.5 and further clarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference 6.1.6. They are as follows:
 
Spurious actuations  Common power source Common enclosure.
 
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 7.0
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-48 NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [A]  Associated Circuit  Safe shutdown system spurious actuation concerns can result from fire damage to a cable whose failure could Cables - Cables cause the spurious actuation/mal-operation of equipment whose operation could affect safe shutdown. These Whose Failure May cables are identified in Section 3.3.3 together with the remaining safe shutdown cables required to support c ontrol Cause Spurious and operation of the equipment. Actuations Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-49 NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [B]  Associated Circuit  The concern for the common power source associated circuits is the loss of a safe shutdown power source due to Cables - Common inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on a non-safe shutdown load Power Source circuit supplied from the safe shutdown power source, a lack of coordination between the upstream supply Cables breaker/fuse feeding the safe shutdown power source and the load breaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdown bus. This would result in the loss of power to the safe shutdown equipment supplied from that power source preventing the safe shutdown equipment from performing its required safe shutdown function. Identify these cables together with the remaining safe shutdown cables required to support control and operation of the equipment. Refer to Section 3.5.2.4 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Breaker coordination is ensured by reviewing the time current curves from the plant's coordination study to ensure coordination. Coordination assures that the protective device nearest the fault operates prior to operation of upstream devices. The means of assuring circuit protection and coordination is provided in a series of calculations. These calculations demonstrate that the Class 1E and non-Class 1E power supplies credited for safe shutdown compliance do have adequate coordination.
Comments  Molded case breakers less than 600V do not require a separate power source to ensure protective features remain available. Breakers for 480V load centers and medium voltage SWGR (4,160V & 6,900V) require DC control power for the protective relaying necessary to assure coordination. This control power is a required input in the fault sub-trees associated with the availability of the aforementioned distribution equipment. The non-safe shutdown loads energized from switchgear that rely on DC control power for relay and metering circuits are deemed required circuits (see comments for alignment statement NEI 00-01 Section 3.3.1.5.).
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 7.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-50 NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [C]  Associated Circuit The concern with common enclosure associated circuits is fire damage to a cable whose failure could propagate Cables - Common to other safe shutdown cables in the same enclosure either because the circuit is not properly protected by an Enclosure Cables isolation device (breaker/fuse) such that a fire-induced fault could result in ignition along its length, or b y the fire propagating along the cable and into an adjacent fire area. This fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e., multiple fires).
Refer to Section 3.5.2.5 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The electrical circuit design for ANO provides proper circuit protection in the form of circuit breakers, fuses, and other devices that are designed to isolate cable faults before the cable ignition temperature is reached. Adequate electrical circuit protection and cable sizing were included as part of the original plant electrical design and are maintained as part of the design change process. Fire rated barrier and penetration seal designs used at ANO preclude the propagation of fire from one fire area to the next to alleviate fire propagation concerns.
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 7, 8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-51 NEI 00-01 Ref NEI 00-01 Guidance 3.3.3  Methodology for Cable  Refer to Figure 3-4 for a flow chart illustrating the various steps involved in selecting the cables necessary for Selection and Location performing a post-fire safe shutdown analysis. Use the following methodology to define the cables requi red for safe shutdown including cables that may cause associated circuits concerns for a post-fire safe shutdown analysis (refer to hard copy of NEI 00-01 for Figure 3-4).
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-52 NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.1  Identify Circuits For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams Required for the including the following documentation to identify the circuits (power, control, instrumentation) required for Operation of the Safe operation or whose failure may impact the operation of each piece of equipment: Shutdown Equipment Single-line electrical diagrams Elementary wiring diagrams Electrical connection diagrams Instrument loop diagrams
 
For electrical power distribution equipment such as power supplies, identify any circuits whose failure may cause a coordination concern for the bus under evaluation.
 
If power is required for the equipment, include the closest upstream power distribution source on the safe shutdown equipment list. Through the iterative process described in Figures 3-2 and 3-3, include the additional upstream power sources up to either the offsite or the emergency power source.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All cables, including those from interlocks, instruments, and power supplies, that could potentially adversely impact the desired operation of a safe shutdown component are listed. This includes cables external to the component control circuit, if any cable fault could adversely impact the required state of the component, unless the cable(s) are included with another safe shutdown component. Primary scheme cables for each safe shutdown component are listed and any reasons to exclude that cable is documented. The required drawings to perform and verify the cable selection and circuit analysis include the P&ID showing the component, the schematic, and others as required.
In some special cases, circuit analysis was completed based on components being skid mounted.
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-53 Safe shutdown components which have support systems that are not modeled/credited in the analysis do not have cables identified (i.e., instrument air not credited and loss of air position was not the same as loss of power, power supplied was not diesel backed and it was determined not to credit non-diesel backed power supplies in order to minimize component selection). These components are assumed to fail in every fire area. These components will always require an operator action to perform their cred ited safe shutdown function and are termed "Always Fail."  This ensures that the required manual action is captured, since it will require the analyst to take action to recover the affected flow path. The AF comp-type designation is used when the plant current configuration cannot be credited without an operator action.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6.1.5, 6.1.12, 6.3 PDMS, Plant Data Management System CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology for ANO-2, Rev. 1, Section 7 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-54 NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.2  Identify Interlocked  In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, cables and Circuits and Cables equipment. Assign to the equipment any cables for interlocked circuits that can affect the equipment. Whose Spurious Operation or While investigating the interlocked circuits, additional equipment or power sources may be discovered. Include Mal-operation Could these interlocked equipment or power sources in the safe shutdown equipment list (refer to Figure 3-3) if t hey can Affect Shutdown impact the operation of the equipment under consideration.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns See alignment basis for previous NEI 00-01 section 3.3.3.1.
Circuit analysis includes any interlock generated from schemes separate from the primary component.
 
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL)
Methodology, Rev. 1, Section 7 CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev.
1, Section 6.1.5, 6.1.12 PDMS, Plant Data Management System Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-55 NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.3  Assign Cables to the Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that may result Safe Shutdown in mal-operation of each piece of safe shutdown equipment. Equipment Tabulate the list of cables potentially affecting each piece of equipment in a relational database including the respective drawing numbers, their revision and any interlocks that are investigated to determine their impact on the operation of the equipment. In certain cases, the same cable may support multiple pieces of equipment. Relate the cables to each piece of equipment, but not necessarily to each supporting secondary component.
If adequate coordination does not exist for a particular circuit, relate the power cable to the power source. This will ensure that the power source is identified as affected equipment in the fire areas where the cable may be damaged. Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns All cables that support or could adversely affect the ability to achieve and maintain post fire safe shutdown have been identified using the methodology defined within CALC-85-E-0087-24. The cables and safe shutdown components with which they are associated have been entered into the PDMS.
Reference Document CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL)
Methodology, Rev. 1, Section 7 CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-56 NEI 00-01 Ref NEI 00-01 Guidance 3.5  Circuit Analysis and This section on circuit analysis provides information on the potential impact of fire on circuits use d to monitor, Evaluation control and power safe shutdown equipment. Applying the circuit analysis criteria will lead to an understanding of how fire damage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section should be used in conjunction with Section 3.4, to evaluate the potential fire-induced impacts that require mitigation.
 
Appendix R, Section III.G.2, identifies the fire-induced circuit failure types that are to be evaluated for impact from exposure fires on safe shutdown equipment. Section III.G.2 of Appendix R requires consideration of hot shorts, shorts-to-ground and open circuits.
 
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.2 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1  Criteria / Assumptions Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations. Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.0 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-57 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.1  [Circuit Failure Consider the following circuit failure types on each conductor of each unprotected safe shutdown cable to determine Types and Impact] the potential impact of a fire on the safe shutdown equipment associated with that conductor.
 
A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or from some other external source resulting in a compatible but undesired impressed voltage or signal on a specific conductor. A hot short may cause a spurious operation of safe shutdown equipment.
 
An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit may prevent the ability to control or power the affected equipment. An open circuit may also result in a change of state for normally energized equipment.  (e.g. [for BWRs] loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an open circuit will result in the closure of the MSIVs). Note that RIS 2004-03 indicates that open circuits, as an initial mode of cable failures, are considered to be of very low likelihood. The risk-informed inspection process will focus on failures with relatively high probabilities.
A short-to-ground may result from a fire-induced breakdown of a cable insulation system, resulting in the potential on the conductor being applied to ground potential. A short-to-ground may have all of the same effects as an open circuit and, in addition, a short-to-ground may also cause an impact to the control circuit or power train of which it is a part.
Consider the three types of circuit failures identified above to occur individually on each conductor of each safe shutdown cable on the required safe shutdown path in the fire area.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns with intent The Safe Shutdown circuit analysis considers cable faults as follows:
All AC grounded circuits must consider any and all shorts, hot shorts, shorts- to-ground, and open circuits. All DC grounded and ungrounded circuits must consider any and all shorts, hot shorts, shorts-to-ground, and open circuits. All ungrounded circuits (both AC and DC) will be analyzed as if the circuit is grounded. This process accounts for the possibility of the circuit experiencing a ground fault as result of the fire. Three phase AC hot short in the proper sequence to cause spurious operation is not considered credible except for high-low pressure interface components.
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-58  For ungrounded DC circuits, two hot shorts of the proper polarity (without grounding) causing spurious operation is not considered credible except for high-low pressure interface components. Only manual initiation of systems required to achieve and maintain safe shutdown is credited. Automatic operation of specific components within these systems is credited where appropriate (such as minimum flow valves). In general, automatic initiation of systems required to achieve and maintain safe shutdown (i.e., Safety Injection Actuation Signal initiation) is not credited unless the initiation signals are shown to be free of fire damage. However, fire induced automatic initiation signals are evaluated for the possibility of spurious component operation and their subsequent adverse impact on safe shutdown. The required cable selection for spurious operation components shall identify the minimum population of cables that could cause the component to spuriously operate. This criterion conservatively assumes other cables of the appropriate polarity and potential are routed in the same raceway with the selected cable(s). For multiple conductor cables, all potential fault consequences due to any combination of hot shorts (inter or intra), shorts-to-ground, or open circuits should be considered. The effect of a cable fault is only seen in fire areas where the cable is routed and recovery of the component, if required, is justified on a fire area basis.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-59 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.2  [Circuit Contacts Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal mode/position of the safe and Operational shutdown equipment as shown on the schematic drawings. The analyst must consider the position of the safe Modes] shutdown equipment for each specific shutdown scenario when determining the impact that fire damage to a particular circuit may have on the operation of the safe shutdown equipment.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All circuits contacts are assumed to be normal position, handswitches either in auto, maintain open, maintain closed, or other position as determined from documents. Spurious signal from relay and instrument contacts are modeled while selecting cables from interlocks. Relay and instrument contacts are assumed to go to position that could provide permissive signal or actuate; if monitored parameter or interlocked device changes at a point during shutdown (i.e. temperature switch starts fan, level/pressure switch changes suction source, breaker contact closes to align another breaker), it is assumed to be in the worst position for cable fault and required shutdown position.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sect ion 4.3.7, 6.1.5.
1, 6.1.6, 6.1.9
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-60 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.3  [Duration of Circuit Assume that circuit failure types resulting in spurious operations exist until action has been taken to isolate the given Failures] circuit from the fire area, or other actions have been taken to negate the effects of circuit failure that is causing the spurious actuation. The fire is not assumed to eventually clear the circuit fault. Note that RIS 2004-03 indicates that fire-induced hot shorts typically self-mitigate after a limited period of time.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent As shown in fire area compliance calculation and manual action feasibility, circuit failure is either mitigated by operator action in control room or by manual actions. No credit is taken for fault clearing on a component and then being operable. Credit is only taken for reactor trip occurring should a fault occur on a Reactor Protective System (RPS) cable. Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, All CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.2.2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-61 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.4  [Cable Failure When both trains are in the same fire area outside of primary containment, all cables that do not meet the separation Configurations] requirements of Section III.G.2 are as sumed to fail in their worst case configuration.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All equipment and cables are assumed to be damaged in a fire area. Credit is taken for Appendix R, III.G.2, compliance and applicable exemptions in those fire areas outside of primary containment.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 3.1, 5.0 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5  [A, Circuit Failure The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to ide ntify any potential Risk Assessment combinations of spurious operations with higher risk significance. Bin 1 failures should also be the focus of the Guidance] analysis; however, NRC has indicated that other types of failures required by the regulations for analysis should not be disregarded even if in Bin 2 or 3. If Bin 1 changes in subsequent revisions of RIS 2004-03, the guidelines in the revised RIS should be followed.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Intra- or inter-cable hot shorts were identified to assist in utilizing information in NRC Regulatory Issue Summary (RIS) 2004-03 in a risk-based evaluation on a fire area basis of spurious actuations and to aid in the transition to a risk-informed, performance-based fire protection licensing basis as outlined in NFPA-805. No cable interactions identified in RIS 2004-03 are used to exclude analyzed cable faults.
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 6.1.8 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-62 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5  [B, Cable Failure For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the same cable is the Modes] most common mode of failure. This is often referred to as "intra-cable shorting."  It is reasonable to assume that given damage, more than one conductor-to-conductor short will occur in a given cable. A second primary mode of cable failure is conductor-to-conductor shorting between separate cables, commonly referred to as "inter-cable shorting."  Inter-cable shorting is less likely than intra-cable shorting. Consistent with the current knowledge of fire-induced cable failures, the following configurations should be considered:
 
A. For any individual multi-conductor cable (thermoset or thermoplastic), any and all potential spurious actuations that may result from intra-cable shorting, including any possible combination of conductors within the cable, may be postulated to occur concurrently regardless of number. However, as a practical matter, the number of combinations of potential hot shorts increases rapidly with the number of conductors within a given cable. For example, a multi-conductor cable with three conductors (3C) has 3 possible combinations of two (including desired combinations), while a five conductor cable (5C) has 10 possible combinations of two (including desired combinations), and a seven conductor cable (7C) has 21 possible combinations of two (including desired combinations). To facilitate an inspection that consider s most of the risk present ed by postulated hot shorts within a multi-conductor cable, inspectors should consider only a few (three or four) of the most critical postulated combinations.
B. For any thermoplastic cable, any and all potential spurious actuations that may result from intra-cable and inter-cable shorting with other thermoplastic cables, including any possible combination of conductors within or between the cables, may be postulated to occur concurrently regardless of number (the consideration of thermoset cable inter-cable shorts is deferred pending additional research).
 
C. For cases involving the potential damage of more than one multi-conductor cable, a maximum of two cables should be assumed to be damaged concurrently. The spurious actuations should be evaluated as previously described. The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research.
D. For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and target conductors are each located within the same multi-conductor cable.
E. Instrumentation Circuits. Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as required power and control circuits. There is one case where an instrument circuit could potentially be considered an associated circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause mal-operation (e.g., unwanted start/stop/reposition signal) of systems necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-63  Applicability Comments  Applicable RIS 2004-03, Revision 1 wording:
 
Consistent with the current knowledge of fire-induced cable failures, the following configurations will be considered for power, control, and instrumentation circuits whose fire-induced failure could prevent operation of safe-shutdown equipment or through mal-operation cause a flow diversion, loss of coolant, or other scenarios that could significantly impact the ability to achieve and maintain hot shutdown:
A. For any individual multi-conductor cable (thermoset or thermoplastic), failure that may result from intra-cable shorting, of any possible combination of conductors within the cable may be postulated to occur concurrently regardless of number. For cases involving the potential damage of more than one multi-conductor cable, assume a maximum of two cables to be damaged. Inspectors should consider only a few (three or four) of the postulated combinations whose failure is likely to significantly impact t he ability to achieve and maintain hot shutdown.
B. For any two thermoplastic cables, failures of any combination of conductors that may result from inter-cable shorting (i.e., between two cables) may be postulated to occur concurrently. Inspectors should consider only a few (three or four) of the postulated combinations whose failure is likely to significantly impact the ability to achieve and maintain hot shutdown.
 
C. For cases involving direct current (DC) control circuits, consider the potential spurious operation due to failures of the control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus). Consider potential spurious actuations when the source and target conductors are each located in the same multi-conductor cable.
D. The decay heat removal system isolation valves at high-pressure/low-pressure interfaces may be subject to three-phase, proper-polarity hot short cable failures. Although this failure is unlikely, it could cause the opening of these valves which would pressurize the low-pressure portion of the decay heat removal system piping outside of containment with the reactor coolant at or near normal reactor operating pressure. These three-phase power cables (either thermoset or thermoplastic jacketed) will be inspected to ensure that they are not subject to three-phase hot shorts that could cause the decay heat removal valves to spuriously open.
Alignment Statement Alignment Basis Aligns with intent The majority of cable at ANO are thermoset. No additional evaluation of cable failure modes was considered in circuitry analysis other than whether spurious actuation was caused by inter- or intra-cable hot short.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, All Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-64 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2  Types of Circuit Appendix R requires that nuclear power plants must be designed to prevent exposure fires from defeating the ability Failures to achieve and maintain post-fire safe shutdown. Fire damage to circuits that provide control and power to equipment on the required safe shutdown path and any other equipment whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of fire damage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, it must be assured that one redundant train of equipment capable of achieving hot shutdown is free of fire damage for fires in every plant location. To provide this assurance, Appendix R requires that equipment and circuits required for safe shutdown be free of fire damage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect to the electrical distribution system, the issue of breaker coordination must also be addressed.
This section will discuss specific examples of each of the following types of circuit failures:
Open circuit Short-to-ground Hot short.
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, All
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-65 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.1  Circuit Failures This section provides guidance for addressing the effects of an open circuit for safe shutdown equipm ent. An open Due to an Open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit wil l typically Circuit prevent the ability to control or power the affected equipment. An open circuit can also result in a change of state f or normally energized equipment. For example, a loss of power to the MSIV solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV.
 
NOTE:  The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis. Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits:  Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment. In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment. Evaluate this to determine if equipment fails safe. Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage. Figure 3.5.2-1 shows an open circuit on a grounded control circuit (refer to hard copy of NEI 00-01 for Figure 3.5.2-1). Open circuit No. 1: An open circuit at location No. 1 will prevent operation of the subject equipment. Open circuit No. 2: An open circuit at location No. 2 will prevent opening/starting of the subject equipment, but will not impact the ability to close/stop the equipment.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All grounded and ungrounded circuits consider open circuits as a fire induced failure mechanism.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Secti on 4.2.1, 4.2.2, 4.2.3, 4.2.8 CALC-ANOC-FP-09-00014, Current Transformer (CT) Open Circuit Concerns, Rev. 0, 8/3/2009, All Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-66 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2  Circuit Failures This section provides guidance for addressing the effects of a short-to-ground on circuits for safe shutdown Due to equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system resulting in the potential on Short-to-Ground the conductor being applied to ground potential. A short-to-ground can cause a loss of power to or control of required [A, General] safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist.
 
Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related to shorts-to-ground:  A short to ground in a power or a control circuit may result in tripping one or more isolation devices (i.e. breaker/fuse) and causing a loss of power to or control of required safe shutdown equipment. In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to an interlocked relay or other device that may cause one or more spurious operations.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All grounded and ungrounded circuits consider any and all shorts to ground.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sections 4.2.1, 4.2.2, and 4.2.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-67 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2  Circuit Failures Short-to-Ground on Grounded Circuits. Typically, in the case of a grounded circuit, a short-to-ground on any Due to part of the circuit would present a concern for tripping the circuit isolation device thereby causing a loss of control Short-to-Ground power.
[B, Grounded Circuits] Figure 3.5.2-2 illustrates how a short-to-ground fault may impact a grounded circuit (refer to hard copy of NEI 00-01, Revision 1 for Figure 3.5.2-2).
 
Short-to-ground No. 1:
 
A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit. This will result an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected.
Short-to-ground No. 2:
A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns All grounded circuits consider any and all shorts to ground.
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sections 4.2.1, 4.2.2, and 4.2.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-68 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2 Circuit Failures Due to  Short-to-Ground on Ungrounded Circuits a Short-to-Ground In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the circuit ma y not [C, Ungrounded result in tripping the circuit isolation device. Another short-to-ground on the circuit or another circuit from the Circuits] same source would need to exist to cause a loss of control power to the circuit.
Figure 3.5.2-3 illustrates how a short to ground fault may impact an ungrounded circuit (refer to hard copy of NEI 00-01, Revision 1 for Figure 3.5.2-3).
Short-to-ground No. 1:
 
A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit if short-to-ground No. 3 also exists either within the same circuit or on any other circuit fed from the same power source. This will result in an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected.
Short-to-ground No. 2:
 
A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns All ungrounded circuits (AC and DC) are conservatively considered to be grounded to simplify analysis. This process accounts for the possibility of the circuit experiencing a simultaneous fault and ground as the result of fire induced damage.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.2.3
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-69 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3  Circuit Failures Due This section provides guidance for analyzing the effects of a hot short on circuits for required safe shutdown to a Hot Short equipment. A hot short is defined as a fire-induced insulation breakdown between conductors of the same cable, [A, General] a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner.
 
Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:
A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a conductor-to-conductor hot short or an internal hot short.
 
A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation of equipment. This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns All grounded circuits and ungrounded circuits consider any and all hot shorts. If the hot short results in a spurious actuation, the circuit failure is reviewed to determine if it is the result of an intra- or inter-cable hot short.
A three phase AC hot short in the proper sequence to cause spurious operation is not considered credible except for high-low pressure interface components.
For ungrounded DC circuits, two hot shorts of the proper polarity (without grounding) causing spurious operation is not considered credible except for high-low pressure interface components.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 4.2.1, 4.2.2, 4.2.
3, 4.2.8, 6.1.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-70 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3  Circuit Failures Due to  A Hot Short on Grounded Circuits. A short-to-ground is another failure mode for a grounded control circuit. a Hot Short A short-to-ground as described above would result in de-energizing the circuit. This would further reduce the [B, Grounded Circuits] likelihood for the circuit to change the state of the equipment either from a control switch or due to a hot short. Nevertheless, a hot short still needs to be considered. Figure 3.5.2-4 shows a typical grounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section. In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted. The following discussion describes how to address the impact of these individual cable faults on the operation of the equipment controlled by this circuit (refer to hard copy of NEI 00-01, Revision 1 for Figure 3.5.2-4).
Hot short No. 1:
 
A hot short at this location would energize the close relay and result in the undesired closure of a motor-operated valve. Hot short No. 2:
 
A hot short at this location would energize the open relay and result in the undesired opening of a motor-operated valve.
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns All grounded circuits consider any and all hot shorts.
 
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sections 4.2.1, 4.2.2, and 4.2.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis Enclosure 1 to 2CAN121202 Page B-71 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit Failures Due to  A Hot Short on Ungrounded Circuits. In the case of an ungrounded circuit, a single hot short may be sufficient to a Hot Short cause a spurious operation. A single hot short can cause a spurious operation if the hot short comes from a [C, Ungrounded circuit from the positive leg of the same ungrounded source as the affected circuit. Circuits] In reviewing each of these cases, the common denominator is that in every case, the conductor in the circuit between the control switch and the start/stop coil must be involved.
Figure 3.5.2-5 depicted below shows a typical ungrounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section.
In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted. The discussion provided below describes how to address the impact of these cable faults on the operation of the equipment controlled by this circuit (refer to hard copy of NEI 00-01, Rev. 1 for Figure 3.5.2-5).
Hot short No. 1:
A hot short at this location from the same control power source would energize the close relay and result in the undesired closure of a motor operated valve.
Hot short No. 2:
A hot short at this location from the same control power source would energize the open relay and result in the undesired opening of a motor operated valve.
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns All ungrounded circuits consider any and all hot shorts. All ungrounded circuits (both AC and DC) are analyzed as if the circuit is grounded. This process accounts for the possibility of the circuit experiencing a ground fault as result of the fire.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Sections 4.2.3 and 6.1.8
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location Enclosure 1 to 2CAN121202 Page B-74 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be identified.
 
NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.4  Identify Routing Identify the routing for each cable including all raceway and cable endpoints. Typically, this information is obtained of Cables from joining the list of safe shutdown cables with an existing cable and raceway database.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The PDMS relates safe shutdown cables to route points (i.e., conduit, junction boxes, tray, equipment). The route points are associated to a fire zone based on conduit and tray drawing. The fire zones are associated with a fire area. These relationships allow determination of cables and equipment impacted on a fire area basis.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 6.1 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.3.10 CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location Enclosure 1 to 2CAN121202 Page B-75 NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.5  Identify Location  Identify the fire area location of each raceway and cable endpoint identified in the previous step and join this of Raceway and information with the cable routing data. In addition, identify the location of field-routed cable by fire area. This Cables by Fire produces a database containing all of the cables requiring fire area analysis, their locations by fire area, and their Area raceway.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The PDMS relates safe shutdown cables to route points (i.e., conduit, junction boxes, tray, equipment). The route points are associated to a fire zone based on conduit and tray drawing. The fire zones are associated with a fire area. These relationships allow determination of cables and equipment impacted on a fire area basis.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 6.1 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 4.3.10 CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, Section 2
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location Enclosure 1 to 2CAN121202 Page B-76 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.4  Circuit Failures Due The evaluation of associated circuits of a common power source consists of verifying proper coordination to Inadequate Circuit between the supply breaker/fuse and the load breakers/fuses for power sources that are required for safe Coordination shutdown. The concern is that, for fire damage to a single power cable, lack of coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdown power source that is required to provide power to safe shutdown equipment.
 
For the example shown in Figure 3.5.2-6, the circuit powered from load breaker 4 supplies power to a non-safe shutdown pump. This circuit is damaged by fire in the same fire area as the circuit providing power to from the Train B bus to the Train B pump, which is redundant to the Train A pump.
 
To assure safe shutdown for a fire in this fire area, the damage to the non-safe shutdown pump powered from load breaker 4 of the Train A bus cannot impact the availability of the Train A pump, which is redundant to the Train B pump. To assure that there is no impact to this Train A pump due to the associated circuits' common power source breaker coordination issue, load breaker 4 must be fully coordinated with the feeder breaker to the Train A bus (refer to hard copy of NEI 00-01, Revision 1 for Figure 3.5.2-6).
A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, the time-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must be considered to ensure that coordination is demonstrated at the maximum fault level.
The methodology for identifying potential associated circuits of a common power source and evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows:  Identify the power sources required to supply power to safe shutdown equipment. For each power source, identify the breaker/fuse ratings, types, trip settings and coordination characteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus. For each power source, demonstrate proper circuit coordination using acceptable industry methods.
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location Enclosure 1 to 2CAN121202 Page B-77  For power sources not properly coordinated, tabulate by fire area the routing of cables whose breaker/fuse is not properly coordinated with the supply breaker/fuse. Evaluate the potential for disabling power to the bus in each of the fire areas in which the associated circuit cables of concern are routed and the power source is required for safe shutdown. Prepare a list of the following information for each fire area:  Cables of concern. Affected common power source and its path. Raceway in which the cable is enclosed. Sequence of the raceway in the cable route. Fire zone/area in which the raceway is located. For fire zones/areas in which the power source is disabled, the effects are mitigated by appropriate methods. Develop analyzed safe shutdown circuit dispositions for the associated circuit of concern cables routed in an area of the same path as required by the power source. Evaluate adequate separation based upon the criteria in Appendix R, NRC staff guidance, and plant licensing bases.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent Breaker coordination assures that the protective device nearest the fault operates prior to operation of ups tream devices. The means of assuring circuit protection and coordination is provided in a series of calculations. These calculations demonstrate that the Class 1E and non-Class 1E power supplies credited for safe shutdown compliance have adequate coordination. On switchgear and/or load centers where breaker coordination relies on relays, coordination may fail if control power or breaker control cables are lost; therefore, load power cables are assigned to switchgear as required so analyst may verify that breaker control is not lost by ensuring that the breaker control cables are not impacted and that control power is available to trip the breaker thus ensuring proper coordination. The review may take place in the fire area compliance document or may be documented in circuit selection/analysis by revising circuit analysis to add cable to a safe shutdown component as required.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev.
1, Section 6.1.5.3, 7.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location Enclosure 1 to 2CAN121202 Page B-78 NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.5  Circuit Failures Due The common enclosure associated circuit concern deals with the possibility of causing secondary failures due to to Common Enclosure fire damage to a circuit either whose isolation device fails to isolate the cable fault or protect the faulted cable Concerns from reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.
The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns The electrical circuit design for ANO provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before the cable ignition temperature is reached. Adequate electrical circuit protection and cable sizing were included as part of the original plant electrical design and are maintained as part of the design change process. Fire rated barrier and penetration seal designs used at ANO preclude the propagation of fire from one fire area to the next to alleviate fire propagation concerns.
Reference Document CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev.
1, Section 7.1, 8.0
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-79 Fire Area Assessment. An engineering analysis shall be performed in accordance with the requirements of Section 2.3 for each fi re area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5 (see Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic)).
 
NEI 00-01 Ref NEI 00-01 Guidance 3.4  Fire Area Assessment By determining the location of each component and cable by fire area and using the cable to equipment relationships and Compliance described above, the affected safe shutdown equipment in each fire area can be determined. Using the list of Assessment affected equipment in each fire area, the impacts to safe shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area can be determined. The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis and evaluation criteria contained in Section 3.5 of this document.
 
Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniques available for individually mitigating the effects of each of the potential impacts.
 
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-80 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1  Criteria / Assumptions The following criteria and assumptions apply when performing fire area compliance assessment to m itigate the consequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area.
 
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 5.0 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.1  [Number of Assume only one fire in any single fire area at a time. Postulated Fires]
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns The fundamental basis for the analysis is that a single fire occurs in a single plant area at a time.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 3.1, 5.0
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-81 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.2  [Damage to Assume that the fire may affect all unprotected cables and equipment within the fire area. This assumes th at Unprotected neither the fire size nor the fire intensity is known. This is conservative and bounds the exposure fire that is Equipment and required by the regulation.
Cables]  Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns A basic assumption of the methodology is that there will be fire damage to equipment and cables located within a fire area. The size and intensity of the fire required causing this equipment damage is not determined. Rather, fire damage is assumed to occur regardless of the level of combustibles in the area, the ignition temperatures of any combustible materials, the lack of an ignition source, or the presence of automatic or manual fire suppression and detection capability.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 3.1, Attachment 8.1 through 8.36
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-82 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.3  [Assess Impacts to Address all cable and equipment impacts affecting the required safe shutdown path in the fire area.
All potential Required Components] impacts within the fire area must be addressed. The focus of this section is to determine and assess the potential impacts to the required safe shutdown path selected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properly protected.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent The Appendix R compliance ARC and the safe shutdown fault tree does not require that all affected components be addressed. Components are addressed until the fault tree shows a method to achieve and maintain safe shutdown (i.e., recovery of top gate COMPLIANCE).
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 3.1, Attachment 8.1 through 8.36 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.4  [Manual Actions] Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in accordance with NRC requirements.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns with intent The process defined in FAQ 07-0030 was used to determine recovery actions.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, All CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 6.2.3, Attachment 8.1 through 8.36 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-83 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.5  [Repairs] Where appropriate to achieve and maintain cold shutdown within 72 hours, use repairs to equipment required i n support of post fire shutdown.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The 72 hour requirement from NEI 00-01 is only applicable to the 10 CFR 50 Appendix R licensing bases. NFPA 805 does not require a plant to transition to cold shutdown within 72 hours, but instead requires licensees to provide reasonable assurance to achieve and maintain the fuel in a safe and stable condition. For ANO-2, the required end state of "safe and stable" under NFPA 805 will be met when the plant is in a stable hot shutdown configuration.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Comm on Results, Rev. 4, Section 3.3.3, 5.10, 6.2.10, 7.2.10 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 6.2.3, Attachment 8.1 through 8.36
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-84 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.6  [Assess Compliance Appendix R compliance requires that one train of systems necessary to achieve and maintain hot shutdown with Deterministic conditions from either the control room or emergency with control station(s) is free of fire damage (III.G.1.a). Criteria] When cables or equipment, including associated circuits, are within the same fire area outside primary containment and separation does not already exist, provide one of the following means of separation for the required safe shutdown path(s):  Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (III.G.2.a)  Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.b). Enclosure of cable and equipment and associated non-safety circuits of one redundant train within a fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.c). For fire areas inside non-inerted containments, the following additional options are also available:  Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fi re hazards (III.G.2.d);  Installation of fire detectors and an automatic fire suppression system in the fire area (III.G.2.e); or  Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield (III.G.2.f). Use exemptions, deviations and licensing change processes to satisfy the requirements mentioned above and to demonstrate equivalency depending upon the plant's license requirements.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns The Appendix R criteria are used to determine compliance strategies on a fire area basis.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Section 4.2 Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-85 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.7  [Consider Additional Consider selecting other equipment that can perform the same safe shutdown function as the impacted Equipment] equipment. In addressing this situation, each equipment impact, including spurious operations, is to be addressed in accordance with regulatory requirements and the NPP's current licensing basis.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns The fire area compliance methodology reviews availability of non-directly affected components to achieve and maintain safe shutdown. If additional equipment that may not be impacted in the fire area was identified, these components were added to SSEL, a circuit analysis completed and basic event included in the safe shutdown fault tree. This is part of the iterative process in performing a safe shutdown analysis.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev.
9, Section 6.2.3, Figure 6.3 NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.8  [Consider Instrument Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent effects on Tubing Effects instrument readings or signals associated with the protected safe shutdown path in evaluating post-fire safe shutdown capability. This can be done systematically or via procedures such as Emergency Oper ating Procedures.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns Mechanical components susceptible to fire damage (brazed or soldered instrument lines, instrument tubing for credited instruments, etc.) are identified and evaluated on a fire area basis.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.39 CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Meth odology, Rev. 1, Section 5.3.2.l Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-86 NEI 00-01 Ref NEI 00-01 Guidance 3.4.2  Methodology for Fire Refer to Figure 3-5 for a flowc hart illustrating the various steps involved in performing a fire area assessment. Area Assessment Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R compliance (refer to hard copy of NEI 00-01 for Figure 3-5).
 
Applicability Comments  Applicable Specific guidance is in subsequent subsections.
 
Alignment Statement Alignment Basis Aligns Specific guidance is in subsequent subsections.
 
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.1  Identify the Affected Identify the safe shutdown cables, equipment and systems located in each fire area that may be potentially Equipment by damaged by the fire. Provide this information in a report format. The report may be sorted by fire area and by Fire Area system in order to understand the impact to each safe shutdown path within each fire area (see Attachment 5 for an example of an Affected Equipment Report).
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The ARC relates the information in PDMS to the safe shutdown fault trees by basic events. Reports can be generated on a fire area basis that are sorted alpha-numerically. In addition, the user interface provided in ARC allows the user to assess the impact on safe shutdown fault tree.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All CALC-85-E-0087-23, Safe Shutdown Equipment List (SSEL) Methodology, Rev. 1, All CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Rev. 1, All Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-87 NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.2  Determine the Based on a review of the systems, equipment and cables within each fire area, determine which shutdown p aths Shutdown Paths Least are either unaffected or least impacted by a postulated fire within the fire area. Typically, the safe shutdown path with the least number of cables and equipment in the fire area would be selected as the required safe shutdown path. Consider the circuit Impacted By a Fire failure criteria and the possible mitigating strategies, however, in selecting the required safe shutdown pa th in a in Each Fire Area particular fire area. Review support systems as a part of this assessment since their availability will be important to the ability to achieve and maintain safe shutdown. For example, impacts to the electric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safe shutdown. By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to the frontline systems that will require this power to support their operation.
Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Identify all equipment not in the safe shutdown path whose spurious operation or mal-operation could affect the shutdown function. Include these cables in the shutdown function list. For each of the safe shutdown cables (located in the fire area) that are part of the required safe shutdown path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path.
 
When evaluating the safe shutdown mode for a particular piece of equipment, it is important to consider the equipment's position for the specific safe shutdown scenario for the full duration of the shutdown scenario. It is possible for a piece of equipment to be in two different states depending on the shutdown scenario or the stage of shutdown within a particular shutdown scenario. Document information related to the normal and shutdown positions of equipment on the safe shutdown equipment list.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The ARC software relates the safe shutdown fault tree to the information in PDMS to visually present the direct fire losses as well as the interactions of cable/equipment associated with ESF actuation system (ESFAS), power, cooling water, and HVAC. This allows the analyst the ability to quickly ascertain what component/functions should be recovered for compliance.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-88 NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.3  Determine Safe Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine the Shutdown Equipment equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire area, and what Impacts those possible impacts are.
 
Applicability Comments  Applicable None
 
Alignment Statement Alignment Basis Aligns Compliance strategies are provided for interactions requiring recovery. These compliance strategies include Appendix R criteria as well as additional circuit analysis as necessary.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-89 NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.4  Develop a Compliance The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Strategy or Disposition Figure 1-2): to Mitigate the Effects Due to Fire Damage - Provide a qualified 3-fire rated barrier. to Each Required Component or Cable - Provide a 1-hour fire rated barrier with automatic suppression and detection. - Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate that there are no intervening combustibles within the 20 foot separation distance. - Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability. - Provide a procedural action in accordance with regulatory requirements. - Perform a cold shutdown repair in accordance with regulatory requirements. - Identify other equipment not affected by the fire capable of performing the same safe shutdown function. - Develop exemptions, deviations, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensing change process. Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R, Section III.G.2.d, e and f.
Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The ARC software relates safe shutdown fault tree to the information in PDMS to visually present the direct fire losses (equipment/cabling) as well as indirect losses due to interactions of ESFAS, power, cooling water, and HVAC. This allows the analyst the ability to quickly ascertain what component/functions should be recovered for compliance. Compliance strategies for the mitigation of fire induced failures are assigned to the losses as necessary to demonstrate the ability to achieve and maintain safe shutdown.
VFDRs are identified. Mitigating strategies to address the VFDRs in a performance based FRE will be developed and documented for transition to NFPA 805. The safe shutdown success paths were analyzed and potential impacts identified. These potential impacts were resolved by specifying one or more of the options listed above such that the least impacted safe shutdown success path could be identified.
 
Credit for existing features and exemptions is taken wherever possible and procedural (recovery) action specified as a last resort.
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-90  Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.1
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-91 NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.5  Document the  Assign compliance strategy statements or codes to components or cables to identify the justification or mitigating Compliance Strategy actions proposed for achieving safe shutdown. The justification should address the cumulative effect of th e Or Disposition actions relied upon by the licensee to mitigate a fire in the area. Provide each piece of safe shutdown equipme nt, Determined to Mitigate equipment not in the path whose spurious operation or mal-operation could affect safe shutdown, and/or c able for the Effects Due to Fire the required safe shutdown path with a specific compliance strategy or disposition. Refer to Attachment 6 for an Damage to Each example of a Fire Area Assessment Report documenting each cable disposition. Required Component or Cable  Applicability Comments  Applicable None Alignment Statement Alignment Basis Aligns The ARC software relates safe shutdown fault tree to the information in PDMS to visually present the direct fire losses (equipment/cabling) as well as indirect losses due to interactions of ESFAS, power, cooling water, and HVAC. This allows the analyst the ability to quickly ascertain what component/functions should be recovered for compliance. Compliance strategies for the mitigation of fire-induced failures are assigned to the losses as necessary to demonstrate the ability to achieve and maintain safe shutdown.
Reference Document CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, All
 
Arkansas Nuclear One - Unit 2 Att. B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment Enclosure 1 to 2CAN121202 Page B-92 NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5  [C, Likelihood of  Determination of the potential consequence of the damaged associated circuits is based on the examination of Undesired specific NPP (P&IDs) and review of components that could prevent operation or cause mal-operation such as flow Consequences] diversions, loss of coolant, or other scenarios that could significantly impair the NPP's ability to achieve and maintain hot shutdown. When considering the potential consequence of such failures, the [analyst] should also consider the time at which the prevented operation or mal-operation occurs. Failures that impede hot shutdown within the first hour of the fire tend to be most risk significant in a first-order evaluation. Consideration of cold shutdown circuits is deferred pending additional research.
Applicability Comments  Applicable RIS 2004-03, Revision 1, wording:
The potential consequences of the damaged circuits are determined by examining plant specific system documentation and by reviewing components that could fail to operate, prevent operation, or cause mal-operation, such as flow diversion, loss of coolant, or other scenarios that could significantly impair the nuclear power plant's (NPP's) ability to achieve and maintain hot shutdown. When considering the potential consequence of such failures, the inspector will also consider the time at which the prevented operation or mal-operation occurs. Failures that impede hot shutdown within the earliest stages of the fire are significant in a first-order evaluation.
Alignment Statement Alignment Basis Aligns with intent A multi-spurious operation expert panel was assembled to determine scenarios that could significantly impair the ability to achieve and maintain hot standby.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Section 5.9.1
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-1 C. NEI 04-02 Table B Fire Area Transition Fire Area ID:  2MH01E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2MH01E Concrete Manhole East 
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from boric acid makeup tanks (BAMTs) prior to switching to refueling water tank (RWT).
: 2. Inventory Control Letdown isolated and reactor coolant pumps (RCPs) secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. 
: 3. Pressure Control Reactor coolant system (RCS) vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using main steam safety valves (MSSVs), if atmospheric dump valves are not immediately available. Emergency feedwater (EFW) pump 2P-7B feeding steam generator (SG) A and SG-B from condensate storage with service water (SW) as a backup.
5a. Vital Auxiliaries (Electrical) Engineered safety feature (ESF) 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red train Control Room heating, ventilation, and air conditioning (HVAC). 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from the Safety Parameter Display System (SPDS) is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 4 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.2
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-2  Fire Area ID:  2MH01E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. No automatic suppression is installed in this area and only electrical cables are present. Electrical manholes have been periodically subject to weather related flooding and with no adverse short-term consequences. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations
 
Engineering Evaluation ID: CALC-85-E-0053-11 Combustible Loading Calculation for Penetrations in Manholes Summary: Purpose:  Evaluate the adequacy of the 3-hour barrier between the east and west sides of manholes with two penetrations that are not sealed. Basis for Acceptability:  Based upon transient control, inaccessibility, and installed caps or location of the sleeves, the penetrations are acceptable as is.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2MH01E Concrete Manhole East No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-3  Fire Area ID:  2MH01E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-24 Title: ANO-2 Fire Area East Manholes Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New Auxiliary feedwater (AFW) source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The Fire PRA quantification does not credit any suppression or detection systems for Fire Area 2MH01E. Fire Area 2MH01E has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence. CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (conditional core damage probability or "CCDP") were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional defense-in-depth (DID) methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-4  Fire Area ID:  2MH01E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: 2MH01E-01 VFDR: Fire damage to control cables in the area may impact the SW swing pump 2P-4B if pump 2P-4A is out of service. Offsite power is available eliminating the immediate need for SW cooling of the emergency diesel generator (EDG), but SW is required as a long term source of feedwater.
a) Spurious operation of the motor operated disconnect 2A-5 may result in a loss of power to 2P-4B.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
End of Fire Area 2MH01E
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-5  Fire Area ID:  2MH02E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2MH02E Concrete Manhole East Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 6 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.4 Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-6  Fire Area ID:  2MH02E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. No automatic suppression is installed in this area and only electrical cables are present. Electrical manholes have been periodically subject to weather related flooding and with no adverse short term consequences. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions
 
Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-85-E-0053-11 Combustible Loading Calculation for Penetrations in Manholes Summary: Purpose:  Evaluate the adequacy of the 3-hour barrier between the east and west sides of manholes with two penetrations that are not sealed. Basis for Acceptability:  Based upon transient control, inaccessibility, and installed caps or location of the sleeves, the penetrations are acceptable as is.
 
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2MH02E Concrete Manhole East No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-7  Fire Area ID:  2MH02E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-24 Title: ANO-2 Fire Area East Manholes Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The fire PRA quantification does not credit any suppression or detection systems for Fire Area 2MH02E. Fire Area 2MH02E has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence
. CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-8  Fire Area ID:  2MH02E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: 2MH02E-01 VFDR Fire damage to control cables in the area may impact the SW swing pump 2P-4B, if pump 2P-4A is out of service. Offsite power i s available eliminating the immediate need for SW cooling of the EDG, but SW is required as a long term source of feedwater.
a) Spurious operation of the motor operated disconnect 2A-5 may result in a loss of power to 2P-4B.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
End of Fire Area 2MH02E
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-9  Fire Area ID:  2MH03E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2MH03E Concrete Manhole East Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to onsite EDGs. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 8 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.6
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-10  Fire Area ID:  2MH03E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. No automatic suppression is installed in this area and only electrical cables are present. Electrical manholes have been periodically subject to weather related flooding and with no adverse short term consequences. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-85-E-0053-11 Combustible Loading Calculation for Penetrations in Manholes Summary: Purpose:  Evaluate the adequacy of the 3-hour barrier between the east and west sides of manholes with two penetrations that are not sealed. Basis for Acceptability:  Based upon transient control, inaccessibility, and installed caps or location of the sleeves, the penetrations are acceptable as is.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2MH03E Concrete Manhole East No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-11  Fire Area ID:  2MH03E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-24 Title: ANO-2 Fire Area East Manholes Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The Fire PRA quantification does not credit any suppression or detection systems for Fire Area 2MH03E. Fire Area 2MH03E has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence
. CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-12  Fire Area ID:  2MH03E Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: 2MH03E-01 VFDR Fire damage to control cables in the area may impact the SW swing pump 2P-4B if pump 2P-4A is out of service. Offsite power is available eliminating the immediate need for SW cooling of the EDG, but SW is required as a long term source of feedwater.
a) Spurious operation of the motor operated disconnect 2A-5 may result in a loss of power to 2P-4B.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
End of Fire Area 2MH03E
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-13  Fire Area ID:  2MH01W Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2MH01W Concrete Manhole West Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 5 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.3
 
Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. No automatic suppression is installed in this area and only electrical cables are present. Electrical manholes have been periodically subject to weather related flooding and with no adverse short term consequences. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-14  Fire Area ID:  2MH01W Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-85-E-0053-11 Combustible Loading Calculation for Penetrations in Manholes Summary: Purpose:  Evaluate the adequacy of the 3-hour barrier between the east and west sides of manholes with two penetrations that are not sealed. Basis for Acceptability:  Based upon transient control, inaccessibility, and installed caps or location of the sleeves, the penetrations are acceptable as is.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2MH01W Concrete Manhole West No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area 2MH01W Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-15  Fire Area ID:  2MH02W Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2MH02W Concrete Manhole West
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 7 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.5
 
Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. No automatic suppression is installed in this area and only electrical cables are present. Electrical manholes have been periodically subject to weather related flooding and with no adverse short term consequences. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-16  Fire Area ID:  2MH02W Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-85-E-0053-11 Combustible Loading Calculation for Penetrations in Manholes Summary: Purpose:  Evaluate the adequacy of the 3-hour barrier between the east and west sides of manholes with two penetrations that are not sealed. Basis for Acceptability:  Based upon transient control, inaccessibility, and installed caps or location of the sleeves, the penetrations are acceptable as is.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2MH02W Concrete Manhole West No No No No No No No No No No No No
 
P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
 
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area 2MH02W
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-17 Fire Area ID:  2MH03W Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2MH03W Concrete Manhole West
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 9 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.7
 
Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. No automatic suppression is installed in this area and only electrical cables are present. Electrical manholes have been periodically subject to weather related flooding and with no adverse short term consequences. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-18  Fire Area ID:  2MH03W Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-85-E-0053-11 Combustible Loading Calculation for Penetrations in Manholes Summary: Purpose:  Evaluate the adequacy of the 3-hour barrier between the east and west sides of manholes with two penetrations that are not sealed. Basis for Acceptability:  Based upon transient control, inaccessibility, and installed caps or location of the sleeves, the penetrations are acceptable as is.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2MH03W Concrete Manhole West No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
 
VFDRs This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area 2MH03W Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-19  Fire Area ID:  AA - B HPSI, LPSI, and Containment Spray Pump Rooms and Gallery Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2007-LL "B" HPSI, LPSI, and Containment Spray Pump Room and Gallery
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to onsite EDGs. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 10 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.8
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-20  Fire Area ID:  AA - B HPSI, LPSI, and Containment Spray Pump Rooms and Gallery Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. This area is located at the lowest point in the auxiliary building and is separated from the only other fire area at this elevation by water-tight doors and barriers. Ponding depth is much less than the lowest elevation of non-credited shutdown equipment in the area. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-112 Penetration Seal Analysis For Penetrations 2007-04-0007, -0011, -0026, and 2030-01-0002 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3) hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
Engineering Evaluation ID: CALC-89-R-2002-64 Penetration Seal Analysis For Penetration 2031-01-0001 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
 
Engineering Evaluation ID: CALC-89-R-2002-97 Penetration Seal Analysis For Penetration 2038-01-0016 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection systems, and the fire brigade's response and is adequate for the hazards.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-21 Fire Area ID:  AA - B HPSI, LPSI, and Containment Spray Pump Rooms and Gallery Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
Engineering Evaluation ID: CALC-ANO2-FP-09-00023 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area AA Summary: Purpose:  This evaluation is to evaluate and document the acceptability of a Unit 2 penetration seal in Fire Area AA to be used in a 3-hour rated fire area boundary.
Basis for Acceptability:  The installed penetration has less penetrating items and mass and the amount of free space is acceptable per A-2600 and thus the installed configuration is bounded by a 3-hour fire test noted on A-2600, Detail 8.
Engineering Evaluation ID: CALC-ANO2-FP-09-00024 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area DD Summary: Purpose:  This evaluation is to evaluate and document the acceptability of a Unit 2 penetration seal in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00036 Fire Protection Engineering Evaluation of Penetration Seals in U-2 Fire Area AA, Part 2 Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-22  Fire Area ID:  AA - B HPSI, LPSI, and Containment Spray Pump Rooms and Gallery Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-07-00003 Water Tight Fire Doors Evaluation Units 1 & 2 86-10 Evaluation Summary: Purpose:  10 CFR 50, Appendix R, requires fire barriers separating redundant trains of safe shutdown equipment to be separated by rated fire barriers that includes all opening, i.e. fire doors, fire dampers, penetration seals, etc. However, th ere are locations in the plant that are susceptible to flooding that are equipped with water tight doors that are not UL Listed or FM Approved to be used in a 3-hour rated barrier. Thus, this evaluation will determine if the water tight fire doors are acceptab le to be used in the 3-hour rated fire barriers at ANO by comparison with UL Listed 3-hour fire doors. The guidance provided in Generic Letter 86-10 will be utilized for the evaluation. Basis for Acceptability:  The water tight doors installed at ANO in 3-hour rated fire barriers have been determined to be acceptable for use based on the hazards in the areas. Although these doors are not 3-hour rated fire doors, they will provide the protection needed in the areas they are used.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2007-LL "B" HPSI, LPSI, and Containment Spray Pump Room and Gallery P Yes No No No No Yes Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-23  Fire Area ID:  AA - B HPSI, LPSI, and Containment Spray Pump Rooms and Gallery Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-23 Title: ANO-2 Fire Area AA Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area AA was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area AA has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-24  Fire Area ID:  AA - B HPSI, LPSI, and Containment Spray Pump Rooms and Gallery Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: AA-01  VFDR Fire damage to cables in the area may impact SW functions. SW is the long-term source of feedwater to the SGs via its connecti on to EFW system once condensate is depleted. Offsite power is avai lable eliminati ng the immedi ate need for SW c ooling of the EDG.
a) Spurious closure of sluice gate 2CV-1470-1, if 2P-4A is the operable pump. Spurious operation of this sluice gate could result in a loss of SW loop 1.
b) Spurious closure of sluice gate 2CV-1472-5, if 2P-4B is the operable pump. Spurious operation of this sluice gate could result in a loss of SW loop 1.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-1470-1. b) No further action is required for 2CV-1472-5.
End of Fire Area AA
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-25  Fire Area ID:  AAC Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description SBOD Alternate AC Diesel 2MH 12 Manhole near SBO Diesel
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 11 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.9
 
Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. Fire suppression activities would only impact the station blackout diesel generator. Plant equipment in other areas is isolated from effect of fire in this fire area. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-26  Fire Area ID:  AAC Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations
 
Engineering Evaluation ID: No engineering evaluations are applicable to this fire area. Summary: N/A Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET SBOD Alternate AC Diesel Yes Yes No No No No No No No No No No 2MH 12 Manhole near SBO Diesel No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
 
VFDRs This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area AAC Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-27  Fire Area ID:  ADMIN - Administration Building Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description ADMIN Administration Building
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 38 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.10 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. The Administration Building is only credited for alternate shutdown when the Technical Support Center is used to monitor shutdown in the event the Control Room is evacuated. Plant equipment in other areas is isolated from effect of fire in this fire area. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-28  Fire Area ID:  ADMIN - Administration Building Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: No engineering evaluations are applicable to this fire area. Summary: N/A  Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET ADMIN Administration Building  No Yes No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area ADMIN
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-29  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2045-XX Turbine Lube Oil Storage Tank Room 2078-QQ Heat Exchanger Equipment Room 2092-PP Chiller Water System Equipment Room 2147-A Chemical Storage Room 2148-A Corridor 2151-A Fuel Handling Room (El. 404) 2152-D Computer Room 2153-A Ventilation Equipment Room 2155-A Steam Pipe Room 2156-A Containment Purge Air Equipment Room 2172-ZZ Storage and Shop Room 2177-YY Neutralizer Tank Room 2178-AAA Lube Oil Reservoir 2200-MM Turbine Building 2201-B Operations Support Facility 2223-KK Pipeway Equipment Access Room (Aux. Bldg. Extension) 2225-WW Regenerative Waste Pump & Tank Room 2229-SS Storage Room 2232 2230-RR Drum Filling Room 2231-TT Plant Heating Boiler Room 2242-OO H&V Mechanical Equipment Room, AO Shack, Lab & Storage Room 2243-NN Chemistry Lab, Kitchen & Offices 2261-UU Plant Heating Boiler Day Tank
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-30  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36B or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-B from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV 2A-4 aligned to onsite EDG. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 12 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.27
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-31  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. The large open area of the turbine building minimizes any ponding concerns and it is expected that water from suppression activities would ultimately migrate to the turbine building basement where no equipment needed to maintain safe and stable conditions is located. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions
 
Licensing Action: Appendix R, Exemption 11, FA - B-2, Not Meeting III.G.3 Criteria , NRC approval letter 0C NA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations
 
Engineering Evaluation ID: CALC-89-R-2002-01 Penetration Seal Analysis for Penetration 2098-07-0094 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area as well as superior penetration seal construction as compared to the detail, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-02 Penetration Seal Analysis for Penetration 2098-07-0098 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-32  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-03 Penetration Seal Analysis for Penetration 2098-07-0097 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-04 Penetration Seal Analysis for Penetration 2073-05-0140 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-05 Penetration Seal Analysis for Penetration 2104-08-0133 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-08 Penetration Seal Analysis for Penetration 2098-07-0096 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-33  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-09 Penetration Seal Analysis for Penetration 2073-05-0135 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-10 Penetration Seal Analysis for Penetration 2073-05-0152 Summary: Purpose: Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary since the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-106 Penetration Seal Analysis For Penetration 2104-07-0121 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low and moderate fire durations, available smoke detection system, fire suppression on one side and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
 
Engineering Evaluation ID: CALC-89-R-2002-109 Penetration Seal Analysis For Penetration 2153-01-0048 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-34  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-30 Penetration Seal Analysis for Penetration 2128-02-0006 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Considering that the seal 1) meets or exceeds all the specified requirements except for the type/number of penetrating items, 2) the deviating penetrants are small and will transfer a relatively small amount of heat, and 3) the favorable conditions associated with the physical location of the seal, this seal is judged to be acceptable for thi s specific application.
 
Engineering Evaluation ID: CALC-89-R-2002-34 Penetration Seal Analysis for Penetration 2081-02-0155 Summary: Purpose: Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary since the seal deviates from the tested configuration. Basis for Acceptability:  Considering that 1) the seal has inherent features that add to its fire resistivity, 2) the location of each seal is favorable with respect to the hazards in the area and 3) detection and manual suppression are available to mitigate the consequences of a fire, the modified seal is judged to be acceptable for this specific application.
 
Engineering Evaluation ID: CALC-89-R-2002-60 Penetration Seal Analysis for Penetration 2147-01-0017 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and automatic fire detection to protect against th e hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-89 Penetration Seal Analysis For Penetration 2075-01-0028 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based as the differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-35  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-94 Penetration Seal Analysis For Penetration 2153-01-0116 Summary: Purpose: Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sinc e it deviates from the tested configuration. Basis for Acceptability: The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and manual fire suppression systems, and the differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Engineering Evaluation ID: CALC-ANO2-FP-06-00004 Penetration Seal FB-2104-07-0007 86-10 Evaluation Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller. A smaller seal is advantageous since it would have less exposure area to the fire.
Engineering Evaluation ID: CALC-ANO2-FP-09-00017 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area B-2 Summary: Purpose:  This evaluation is to document the acceptability of unit 2 penetration seals in Fire Area B-2 to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The combustible loading, the smoke detection and the response by the fire brigade to suppress a fire in the incipient stage on either side. Thus, the penetration seals are considered to be adequate for the hazards in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00024 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area DD Summary: Purpose:  This evaluation is to evaluate and document the acceptability of a Unit 2 penetration seal in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-36  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00028 Engineering Evaluation of Penetration Seals in Fire Area HH Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetration seals in Fire Area HH to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00034 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area FF Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area FF to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetration seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection system (near side only) and the response by the fire brigade to suppress the fire in the incipient stage.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00035 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area GG Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area GG to be used in a 3-hour rated fire area boundary.
Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 & 2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Building penetrations to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The low probability of a fire starting in the areas of the penetrations, the installed smoke detectio n and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual firefighting equipment for those areas in the unit's Auxiliary Buildings.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-37  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-09-00011 Fire Protection Engineering Evaluation of Units 1 & 2 Aux Bldg Elevator Doors Summary: Purpose:  Evaluate elevator doors as part of the 3-hour fire boundary.
Unit 2: 335' 2040-JJ Fire Area DD 354' 2073-DD Fire Area HH 386' 2136-I Fire Area G 404' 2151-A Fire Area B-2 The elevator doors were previously evaluated in calculation 85-E-0053-04 however the calculation will be superseded by this evaluation as part of the NFPA 805 transition project.
Basis for Acceptability:  Based on the low and moderate (Fire Zone 2136-I only) combustible loading, the availability of the smoke detection systems (and suppression system in Fire Zone 67-U) and the availability of the fire brigade with manual firefighting equipment, the elevator door are considered to be adequate for the hazards in the area and acceptable for the 3-hour rated fire barriers.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-38  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2045-XX Turbine Lube Oil Storage Tank Room P No No No No No No No No No No No 2078-QQ Heat Exchanger Equipment Room No No No No No No No No No No No No 2092-PP Chiller Water System Equipment Room No No No No No No No No No No No No 2147-A Chemical Storage Room Yes No No No No No Yes No No No No No 2148-A Corridor No No No No No No No No No No No No 2151-A Fuel Handling Room (El. 404) No Yes No No No No No Yes No Yes No Yes 2152-D Computer Room No Yes No No No No No Yes No Yes No Yes 2153-A Ventilation Equipment Room No P No No No No No Yes No Yes No Yes 2155-A Steam Pipe Room No No No No No No No No No No No No 2156-A Containment Purge Air Equipment Room No Yes No No No No No No No Yes No Yes 2172-ZZ Storage and Shop Room P No No No No No No No No No No No 2177-YY Neutralizer Tank Room No No No No No No No No No No No No 2178-AAA Lube Oil Reservoir P No No No No No No No Yes No No No 2200-MM Turbine Building P P No No No No Yes Yes No Yes No Yes 2201-B Operations Support Facility No Yes No No No No No No No Yes No Yes 2223-KK Pipe-way Equipment Access Room (Aux. Bldg. Extension) No Yes No No No No No Yes No Yes No Yes 2225-WW Regenerative Waste Pump & Tank Room No No No No No No No No No No No No 2229-SS Storage Room 2232 P No No No No No Yes No No No No No 2230-RR Drum Filling Room No No No No No No No No No No No No 2231-TT Plant Heating Boiler Room P No No No No No No No No No No No 2242-OO H&V Mechanical Equipment Room, AO Shack, Lab & Storage Room Yes P No No No No No No No Yes No Yes 2243-NN Chemistry Lab, Kitchen & Offices P No No No No No No No No No No No 2261-UU Plant Heating Boiler Day Tank P No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-39  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-21 Title: ANO-2 Fire Area B-2 Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions The following equipment is recovered in the post transition baseline case:  2A-309 switchgear breaker  2P-32A reactor coolant pump  2P-32B reactor coolant pump  2P-32C reactor coolant pump  2P-32D reactor coolant pump Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area B-2 was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Taking credit for suppression in Fire Zone 2178-AA (CALC-PRA-A2-05-010), Fire Area B-2 has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-40  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None  VFDRs VFDR ID: B2-01  VFDR Fire damage to control and power cables in the area may impact Hot Standby Decay Heat Removal functions resulting in the following:
a) Loss of MSIV 2CV-1060-2 isolation capability to the credited SG B. b) Loss of MSIV 2CV-1010-1 isolation capability to the non-credited SG A.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-1060-2. b) No further action is required for 2CV-1010-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-41  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: B2-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Loss of power from 2B52 or spurious opening of valve 2CV-5630-1 could result in RWT drain-down to containment sump in conjunction with spurious opening of valve 2CV-5649-1 Containment Sump Recirculation valve. b) Disables the remote start permissive relay to the low suction pressure for Charging Pumps 2P-36B and C(G) preventing starting the pumps from the Control Room. c) Spurious closure or loss of open capability for valves 2CV-4920-1 and 2CV-4921-1 resulting in isolation of BAM supply to Charging pumps. d) Loss of DC control circuit power to Pressurizer Heater bank #1 resulting in loss of trip capability from the Control Room.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-5630-1 and 2CV-5649-1. b) No further action is required as 2P-36-B and 2P-36C are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. c) No further action is required as 2CV-4920-1 and 4921-1 are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. d) Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-42  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: B2-03  VFDR: Fire damage to DC power cables could result in a loss of the Control Room trip capability of the RCPs 2P-32A through D. Securi ng the pumps assures normal pressurizer spray is secured and prevents potential RCP seal damage. The 2H1 and 2H2 buses are located in the Turbine Building Fire Zone 2200-MM which results in loss of control power and circuits that could defeat the ability to stop the RCPs from the Control Room. Also, the potential exists for control cable damage that could result in spurious starting of any RCP. Fire impacted cables are in the area of the non-safety switchgear (or below) and fire exposure could restrict access to the RCP breakers. Fire affecting the control circuits for 2A-1, 2A-2 and 2X03 may disable the automatic power transfer from 2X02 Unit Auxiliary Transformer to 2X03 Startup Transformer #3.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with modification of the DC control power circuit to switchgear 2A-1, 2A-2, 2H-1, 2H-2, and a recover y action credited to secure offsite power.
 
VFDR ID: B2-04  VFDR: Fire damage to cables in the area could result in failure of the EDG Fuel Transfer Pumps (2P-16A/B). Given the need to operate EDG #2, the capability to transfer fuel from the ANO-1 EDG Fuel Transfer Pumps (P-16A/B) will be required.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-43  Fire Area ID: B Turbine Building & General Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: B2-05  VFDR: AAC Generator control cables for 2A-410 trip and close functions are routed in various trays in Fire Zone 2200-MM as is the redundant 2A-310 control cable. Both cables must fail in order to close their respective tie breaker.
Loss of this function could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Additional circuit analysis of the tie breakers 2A-310 and 2A-410 demonstrates that spurious closing of both breakers is not credible due to interlocks and the required sequence of events.
 
VFDR ID: B2-06  VFDR: Fire damage to control cables could result in the potential spurious start of the AFW Pump 2P-75 and Condensate Pumps 2P-2A, 2P-2B, 2P-2C, and 2P-2D and render Control Room trip capability unavailable.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
 
VFDR ID: B2-07  VFDR: Fire damage to control cables in the area may impact the SW swing pump 2P-4B if pump 2P-4C is out of service. Offsite power ma y not be available in this fire area and SW is required for EDG jacket water cooling and as a long term source of feedwater.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
End of Fire Area B-2 Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-44  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2091-BB North Electrical Penetration Room 2112-BB Lower North Electrical Penetration Room 2183-J Upper North Electrical Penetration Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. Variance from the determinist ic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 13 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.28
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-45  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. Fire zone 2091-BB in this fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station from Fire Area B-2 (Turbine Building). Any water from manual suppression in 2091-BB is expected to migrate to the large open area of the turbine building where ponding would not be a concern. The remaining zones have automatic suppression but have minimal ponding levels. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions
 
Licensing Action: Appendix R, Exemption 10, FA - B-3, Not Meeting III.G.3 Criteria , NRC approval letter 0C NA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-36 Penetration Seal Analysis For Penetration 2091-01-0087 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is an equivalent in heat transmission.
Engineering Evaluation ID: CALC-ANO2-FP-06-00006 Piping Chases in Room 2076 Going to Room 2091 86-10 Evaluation Summary: Purpose:  Evaluate two piping chases to determine if they provide an acceptable fire barrier between Fire Areas GG and B-3. Basis for Acceptability:  The installed penetrations are equivalent to the fire tested configurations of electrical conduit penetrations and fire doors. The access panels were verified to meet the structural integrity and heat transfer minimum standards for fire rated doors Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3-hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual fire fighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-46  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00014 Fire Protection Engineering Evaluation for Unit 2 Fireflex Seals Summary: Purpose:  This evaluation is to evaluate Unit 2 penetration seals utilizing Fireflex as a sealing material and document their acceptability to be used in a 3-hour rated fire area boundary based on the hazards in the area. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection systems, and the fire brigade's response and are adequate for the hazards.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00027 Fire Protection Engineering Report of Penetration Seals in Fire Area B-3 Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetration seals in Fire Area B-3 to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is adequate for the hazards based on the combustible loading, smoke detection systems, suppression system on the top side (Fire Zone 2112-BB) and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Building penetrations to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual fire fighting equipment for those areas in the unit's Auxiliary Buildings.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-47  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2091-BB North Electrical Penetration Room No Yes No No No No No Yes No Yes No Yes 2112-BB Lower North Electrical Penetration Room Yes Yes No No No No Yes Yes Yes Yes No Yes 2183-J Upper North Electrical Penetration Room Yes Yes No No No No Yes Yes Yes Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-48  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-22 Title: ANO-2 Fire Area B-3 Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions The following equipment is recovered in the post transition baseline case:  2CV-0789-1 EFW pump condensate suction valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. The following  modification is area specific and credited to reduce risk in this fire area:  2CV-4698-1 circuit modified to prevent spurious opening and resolve IN 92-18 issue. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. The modification to 2CV-4698-1 will prevent spurious operation.
Additional Fire Area Considerations The detection system located in Fire Area B-3 was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Taking credit for suppression in Fire Zones 2112-BB and 2183-J (CALC-PRA-A2-05-010), Fire Area B-3 has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights"
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-49  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-50  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: B3-01  VFDR: Fire damage to control and power cables in the area may impact EFW water functions resulting in the following:
a) Spurious closure of valve 2CV-1075-1 (IN 92-18) which feeds the credited SG B. b) Spurious opening of valve 2CV-1052 (IN 92-18) to the atmospheric dump to SG B. c) Loss of MSIV 2CV-1060-2 isolation capability to the credited SG B. d) Spurious closure of 2CV-0789-1 causes isolation of CST supply to EFW Pump 2P-7B. e) Loss of MSIV 2CV-1010-1 and associated MSIV bypass 2CV-1040-1 (IN 92-18) isolation capability to SG A. f) Failure to open or maintain open 2CV-0716-1 from SW system to pump 2P-7B upon depletion of condensate.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-1075-1. b) No further action is required for 2CV-1052. c) No further action is required for 2CV-1060-2. d) Recovery action for 2CV-0789-1. e) No further action is required for 2CV-1010-1 and 2CV-1040-1. f) No further action is required for 2CV-0716-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-51  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: B3-02  VFDR: Fire damage to control cables for 2P-32C (breaker 2H-22) and 2P-32D (breaker 2H-12) can result in a spurious re-start of 2P-32C and 2P-32D, respectively. Securing the pumps assures normal pressurizer spray is secured and prevents potential RCP seal damage.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Additional circuit analysis on the RCP pump motors concluded that fire damage to the RCP cables would not prevent the Control Room from tripping the pumps for a fire in Fire Area B-3. Since the Control Room will retain the ability to trip the RCPs afte r fire in Fire Area B-3 this is not a VFDR, and the associated basic events have been excluded from the model in the Post-Transition Baseline Case.
VFDR ID: B3-03  VFDR: Fire damage to cables in the area could result in spurious operation of 2CV-1530-1 (SW Loop 1 to component cooling water (CCW) Heat Exchangers) requiring isolation of credited SW loop to prevent pump run-out. This valve is an IN 92-18 concern and spurio us operation can result in failure of the valve in an undesired position with the inability to reposition.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-52  Fire Area ID: B North Penetration Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Open items Item ID: B3-HVAC
 
== Description:==
Analysis is required to determine post fire room heat-up in locations requiring operator manual actions given a loss of HVAC. The result of this analysis will be used to evaluate availability of equipment required for post fire shutdown and habitability for feasibility of operator manual actions based on room temperature.
Disposition: The analysis and evaluation for post fire room heat-up is performed in EC-35075. In Fire Area B-3 excessive temperatures have been identified in Fire Zone 2091-BB following a loss of ventilation. A modification is proposed to modify the existing redundant r oom ventilation to prevent a single fire in other fire areas from impacting both exhaust fans. This modification will assure that either 2VEF-63 or 2VEF-64 will remain functional and maintain room temperature at an acceptable level (see Attachment S).
Status: Closed  Corrective Action Ref.: CR-ANO-C-2006-0048 CA-55
 
End of Fire Area B-3
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-53  Fire Area ID: B CEDM Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2154-E CEDM Equipment Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B or 2P-36C available with gravity feed from BAMTs, normal injection path isolated, and manual alignment to RCS using HPSI header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 14 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.29
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-54  Fire Area ID: B CEDM Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station from Fire Area B-2 (Turbine Building). Any excess water from suppression is expected to migrate to the large open area of the turbine building where ponding would not be a concern. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-124 Penetration Seal Analysis For Penetration 2154-01-0039 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on available smoke detection system and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and th e installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
 
Engineering Evaluation ID: CALC-89-R-2002-62 Penetration Seal Analysis For Penetration 2154-01-0078 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will h ave no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-55  Fire Area ID: B CEDM Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-63 Penetration Seal Analysis For Penetration 2154-01-0081 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will h ave no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2154-E CEDM Equipment Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-56  Fire Area ID: B CEDM Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-05 Title: ANO-2 Fire Area B-4 Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions The following equipment is recovered in the post transition baseline case:  2P-32A Reactor Coolant Pump  2P-32B Reactor Coolant Pump  2P-32C Reactor Coolant Pump  2P-32D Reactor Coolant Pump Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. The following  modification is area specific and credited to reduce risk in this fire area:  2C70, 2C71, 2C72, 2C73, 2C75*, 2C80, and 2C409* panels will have incipient detection installed (*CALC-PRA-A2-05-010). IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area B-4 was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area B-4 has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-57  Fire Area ID: B CEDM Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-58  Fire Area ID: B CEDM Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs VFDR ID: B4-01  VFDR: Fire damage to control cables for RCPs 2P-32A (breaker 2H-11), 2P-32B (breaker 2H-21), 2P-32C (breaker 2H-22) and 2P-32D (breaker 2H-12) can result in a spurious re-start of 2P-32A, 2P-32B, 2P-32C and 2P-32D, respectively. Securing the pumps assur es normal pressurizer spray is secured and prevents potential RCP seal damage.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR ID: B4-02  VFDR: Fire damage to control cables associated with 2CV-4824-2 (IN 92-18) can result in a spurious failure of the valve in the open position. This results in a diversion path from the primary inventory control path to auxiliary pressurizer spray using the charging syst em. Makeup using the charging pump is through the high pressure safety injection (HPSI) header by manually opening 2CVC-115 and closing 2CV-4840-2 from Control Room to isolate auxiliary spray.
Loss of these functions could challenge the Inventory and Pressure Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: No further action is required as 2CV-4824-2 is a component in the chemical and volume control system (CVCS) system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.
 
End of Fire Area B-4
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-59  Fire Area ID: B North and South Aux Building Stair Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2149-B Stairwell No. 2001 2158-F Stairwell No. 2055
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 15 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.30
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-60  Fire Area ID: B North and South Aux Building Stair Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. The stairwells do not contain automatic suppression and firefighting activities are limited to manual methods. The lack of combustibles in the stairwells also limits suppression needs and it is expected that ponding will be minimized and have no impact on adjoining areas. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-120 Penetration Seal Analysis For Penetration 2066-03-0018 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00035 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area GG Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area GG to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-61  Fire Area ID: B North and South Aux Building Stair Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2149-B Stairwell No. 2001 No No No No No No No No No No No No 2158-F Stairwell No. 2055 No No No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
 
VFDRs This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area B-5
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-62  Fire Area ID: B Auxiliary Building Access Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2006-LL General Access Room 2010-LL "C" HPSI Pump Room 2011-LL Tendon Gallery Access 2014-LL "A" HPSI, LPSI & Containment Spray Pump Room Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A and 2P-4C available to feed SW headers 1 and 2. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 16 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.31 Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-63  Fire Area ID: B Auxiliary Building Access Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. This Area is located at the lowest point in the auxiliary building and is separated from the only other fire area at this elevation by water-tight doors and barriers. Ponding depth is much less than the lowest elevation of non-credited shutdown equipment in the area. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-84-D-2002-01 Fire Resistant Hatch Covers Summary: Purpose: Design fire resistant hatch covers for installation in opening located in Rooms 2040 (Elev. 335') and 2069 (Elev. 354'
).Basis for Acceptability:  Hatches are designed to withstand the effects of a 3 hour fire from below and a 1 hour fire from abov e which is adequate for the hazard.
Engineering Evaluation ID: CALC-89-R-2002-112 Penetration Seal Analysis For Penetrations 2007-04-0007, -0011, -0026, and 2030-01-0002 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
 
Engineering Evaluation ID: CALC-89-R-2002-71 Penetration Seal Analysis For Penetration 2055-01-0198 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will h ave no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-64  Fire Area ID: B Auxiliary Building Access Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00023 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area AA Summary: Purpose:  This evaluation is to evaluate and document the acceptability of an ANO-2 penetration seal in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability: The installed penetration has less penetrating items and mass and the amount of free space is acceptable per A-2600 and thus the installed configuration is bounded by a 3-hour fire test noted on A-2600, Detail 8.
Engineering Evaluation ID: CALC-ANO2-FP-09-00024 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area DD Summary: Purpose:  This evaluation is to evaluate and document the acceptability of an ANO-2 penetration seal in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00025 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area EE Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area EE to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on factors such as combustible loading, smoke detection systems, suppression systems and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00036 Fire Protection Engineering Evaluation of Penetration Seals in U-2 Fire Area AA, Part 2 Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-65  Fire Area ID: B Auxiliary Building Access Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-07-00003 Water Tight Fire Doors Evaluation Units 1 & 2 86-10 Evaluation Summary: Purpose:  10 CFR 50, Appendix R, requires fire barriers separating redundant trains of safe shutdown equipment to be separated by rated fire barriers that includes all opening, i.e. fire doors, fire dampers, penetration seals, etc. However, th ere are locations in the plant that are susceptible to flooding that are equipped with water tight doors that are not UL Listed or FM Approved to be used in a 3-hour rated barrier. Thus, this evaluation will determine if the water tight fire doors are acceptab le to be used in the 3-hour rated fire barriers at ANO by comparison with UL Listed 3-hour fire doors. The guidance provided in Generic Letter 86-10 will be utilized for the evaluation. Basis for Acceptability:  The water tight doors installed at ANO in 3-hour rated fire barriers have been determined to be acceptable for use based on the hazards in the areas. Although these doors are not 3-hour rated fire doors, they will provide the protection needed in the areas they are used.
Engineering Evaluation ID: CALC-ANOC-FP-09-00012 Engineering Evaluation for Units 1 & 2 Partial Suppression Systems Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains that are protected by a 1-hour fire wrap. Basis for Acceptability:  The areas of concern (1-hour fire wrap) are adequately covered by the suppression systems and thus a fire would not damage both trains and prevent them from performing their designed functions.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2006-LL General Access Room P Yes No No No No Yes Yes No Yes No Yes 2010-LL "C" HPSI Pump Room No Yes No No No No No Yes No Yes No Yes 2011-LL Tendon Gallery Access No Yes No No No No No Yes No Yes No Yes 2014-LL "A" HPSI, LPSI & Containment Spray Pump Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-66  Fire Area ID: B Auxiliary Building Access Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-06 Title: ANO-2 Fire Area B-6 Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area B-6 was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area B-6 has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-67  Fire Area ID: B Auxiliary Building Access Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: B6-01  VFDR: Fire damage to control and power cables in the area may impact SW functions. SW is the long-term source of feedwater to the SG s via its connection to EFW system once condensate is depleted. Offsite power is available elimi nating the immedi ate need for SW cooling of the EDG.
a) Fire damage to power cables in the area may impact SW functions resulting in a loss of all three SW pumps (2P-4A, swing pump 2P-4B(R) & (G), and 2P-4C). Currently an electrical raceway fire barrier system (wrap) is credited for both 2P-4A and 2P-4C and is identified for potential removal based upon the FRE.
b) Loss of valve 2CV-1425-1 prevents isolation of Auxiliary Cooling Water needed to control SW pump run-out if only one pump is available.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required for components associated with SW (vital auxiliaries).
End of Fire Area B-6
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-68  Fire Area ID: CC - Emergency Feedwater Pump Room (Turbine Driven) Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2024-JJ EFW Pump Room (Turbine Driven)
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 17 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.11 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station from Fire Area DD. This room is lower than fire area and separated from redundant equipment by a watertight door. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-69  Fire Area ID: CC - Emergency Feedwater Pump Room (Turbine Driven) Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: Appendix R, Exemptio n 03, FA - CC, Not Meeting 3-hour Rated Barrier III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because an EEEE has found the current plant configuration to be acceptable.
Engineering Evaluations Engineering Evaluation ID: CALC-ANO2-FP-09-00017 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area B-2 Summary: Purpose:  This evaluation is to document the acceptability of unit 2 penetration seals in Fire Area B-2 to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The combustible loading, the smoke detection and the response by the fire brigade to suppress a fire in the incipient stage on either side. Thus, the penetration seals are considered to be adequate for the hazards in the area.
Engineering Evaluation ID: CALC-ANOC-FP-07-00003 Water Tight Fire Doors Evaluation Units 1 & 2 86-10 Evaluation Summary: Purpose:  10 CFR 50, Appendix R, requires fire barriers separating redundant trains of safe shutdown equipment to be separated by rated fire barriers that includes all opening, i.e. fire doors, fire dampers, penetration seals, etc. However, th ere are locations in the plant that are susceptible to flooding that are equipped with water tight doors that are not UL Listed or FM Approved to be used in a 3-hour rated barrier. Thus, this evaluation will determine if the water tight fire doors are acceptab le to be used in the 3-hour rated fire barriers at ANO by comparison with UL Listed 3-hour fire doors. The guidance provided in Generic Letter 86-10 will be utilized for the evaluation. Basis for Acceptability:  The water tight doors installed at ANO in 3-hour rated fire barriers have been determined to be acceptable for use based on the hazards in the areas. Although these doors are not 3-hour rated fire doors, they will provide the protection needed in the areas they are used.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-70  Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2024-JJ EFW Pump Room (Turbine Driven) No Yes No No No No No Yes No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area CC
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-71  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2019-JJ Boric Acid Condensate Tank Room 2032-JJ Spent Resin Storage Tank Room 2040-JJ Corridor 2068-DD Hot Machine Shop
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36B or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. Variance from the determinist ic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 18 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.12
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-72  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area no automatic suppression system and firefighting activities are from controlled manual methods using hose station(s). Ponding is minimal due to the large area and the expected propagation path does not impact equipment in other areas. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: Appendix R, Exemption 05, FA  DD, Not Meeting III.G.2 Criteria, NRC approval letter 0C NA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations Engineering Evaluation ID: CALC-84-D-2002-01 Fire Resistant Hatch Covers Summary: Purpose: Design fire resistant hatch covers for installation in opening located in rooms 2040 (Elev. 335') and 2069 (Elev. 354'
).Basis for Acceptability:  Hatches are designed to withstand the effects of a 3 hour fire from below and a 1-hour fire from above which is adequate for the hazard.
 
Engineering Evaluation ID: CALC-89-R-2002-112 Penetration Seal Analysis For Penetrations 2007-04-0007, -0011, -0026, and 2030-01-0002 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
 
Engineering Evaluation ID: CALC-89-R-2002-127 Penetration Seal Analysis For Penetration 2055-04-0059 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on available smoke detection system and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items with equivalent exposed surface area.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-73 Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-129 Penetration Seal Analysis for Penetration 2048-05-012 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Based on the similarity of tested and installed penetrating items, the availability of fire extinguis hers and detection to protect against the hazards of the area as well as the limited combustible loading, it is considered that this penetration seal offers equivalent protection to the surrounding hazards and therefore the deviation from the detail is negligible and considered acceptable.
Engineering Evaluation ID: CALC-89-R-2002-21 Penetration Seal Analysis for Penetration 2017-05-0063 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and detection to protect against the hazards of th e area, as well as the limited combustible loading, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-23 Penetration Seal Analysis for Penetration 2055-07-0173 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression to protect against the hazards of the area, as wel l as limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-28 Penetration Seal Analysis for Penetration 2074-01-0001 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  The installed configuration significantly differs from the approved detail. However, the differences can be justified by the low combustible loading, the available detection and suppression capabilities and by comparative analysis to other test configurations. Based on the hazards in the area and considering the above factors, the installed configuration is deemed to be capable of maintaining a 3 hour barrier.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-74  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-31 Penetration Seal Analysis for Penetration 2048-06-0040, 41, 42 Summary: Purpose:  Evaluate penetration seals to determine if it is acceptable to utilize them in a three (3)-hour rated fire boundary since the seals deviate from the tested configuration. Basis for Acceptability:  Considering that the seals 1) meet or exceed all the specified requirements except for the size of penetrating items, 2) the seal depth is over twice that of the specified detail and 3) the favorable conditions associated with the physical location of the seals, these seals are judged to be acceptable for this specific application.
Engineering Evaluation ID: CALC-89-R-2002-43 Penetration Seal Analysis for Penetration 2017-05-0077 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Since the referenced test substantiates the ability of a large diameter mechanical penetrant through a sleeve with a zero annulus and sealed with 10" of silicone foam to successfully withstand a 38-minute fire endurance test and the fact that the referenced penetration is located in a fire zone with a fire duration of less than thirty (38) minutes, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-45 Penetration Seal Analysis for Penetration 2016-05-0002 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  The seals utilized at ANO that consist of sleeves with pipes utilizing an annulus that is three (3) inches or less, ten (10) inches of silicone foam with one (1) inch of damming material and located in a barrier with fire duration of less than forty-three (43) minutes are considered acceptable.
Engineering Evaluation ID: CALC-89-R-2002-46 Penetration Seal Analysis For Penetrations 2055-02-0079 & -0097 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection on both sides of the barrier, and the response by the fir e brigade to suppress a fire in the incipient stage on either side the differences identified will have no effect on the seal sin ce the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-75  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-47 Penetration Seal Analysis For Penetrations 2055-05-0021, -0023, -0025, -0028, -0030, & -0043 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3) hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The lack of damming material on one side does not degrade the installed configuration. The installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller.
A smaller seal is advantageous since it would have less exposure area to the fire and the installed configuration is sealed the full depth of the barrier ~ 18".
 
Engineering Evaluation ID: CALC-89-R-2002-53 Penetration Seal Analysis For Penetration 2058-01-0009 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection on both sides of the barrier, and the response by the fir e brigade to suppress a fire in the incipient stage on either side the differences identified will have no affect on the seal sin ce the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Engineering Evaluation ID: CALC-89-R-2002-64 Penetration Seal Analysis For Penetration 2031-01-0001 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-73 Penetration Seal Analysis For Penetration 2055-02-0137 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the low fire duration, smoke detection on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will have no effect on the sea l since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-76  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-89 Penetration Seal Analysis For Penetration 2075-01-0028 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based as the differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
 
Engineering Evaluation ID: CALC-89-R-2002-97 Penetration Seal Analysis For Penetration 2038-01-0016 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection systems and the fire brigade's response, and is adequate for the hazards.
 
Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
Engineering Evaluation ID: CALC-ANO2-FP-06-00002 Penetration Seal FB-2055-05-0005 86-10 Evaluation Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller. A smaller seal is advantageous since it would have less exposure area to the fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-77  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3 -hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual fire fighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
Engineering Evaluation ID: CALC-ANO2-FP-09-00014 Fire Protection Engineering Evaluation for Unit 2 Fireflex Seals Summary: Purpose:  This evaluation is to evaluate Unit 2 penetration seals utilizing Fireflex as a sealing material and document their acceptability to be used in a 3-hour rated fire area boundary based on the hazards in the area. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection systems, and the fire brigade's response and are adequate for the hazards.
Engineering Evaluation ID: CALC-ANO2-FP-09-00017 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area B-2 Summary: Purpose:  This evaluation is to document the acceptability of unit 2 penetration seals in Fire Area B-2 to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  Based on the combustible loading, the smoke detection and the response by the fire brigade to suppress a fire in the incipient stage on either side the penetration seals are considered to be adequate for the hazards in th e area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00024 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area DD Summary: Purpose:  This evaluation is to evaluate and document the acceptability of an ANO-2 penetration seal in Fire Area AA to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-78  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00025 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area EE Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area EE to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on factors such as combustible loading, smoke detection systems, suppression systems and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00028 Engineering Evaluation of Penetration Seals in Fire Area HH Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetration seals in Fire Area HH to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00035 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area GG Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area GG to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANOC-FP-07-00003 Water Tight Fire Doors Evaluation Units 1 & 2 86-10 Evaluation Summary: Purpose:  10 CFR 50, Appendix R, requires fire barriers separating redundant trains of safe shutdown equipment to be separated by rated fire barriers that includes all opening, i.e. fire doors, fire dampers, penetration seals, etc. However, th ere are locations in the plant that are susceptible to flooding that are equipped with water tight doors that are not UL Listed or FM Approved to be used in a 3-hour rated barrier. Thus, this evaluation will determine if the water tight fire doors are acceptab le to be used in the 3-hour rated fire barriers at ANO by comparison with UL Listed 3-hour fire doors. The guidance provided in Generic Letter 86-10 will be utilized for the evaluation. Basis for Acceptability:  The water tight doors installed at ANO in 3-hour rated fire barriers have been determined to be acceptable for use based on the hazards in the areas. Although these doors are not 3-hour rated fire doors, they will provide the protection needed in the areas they are used.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-79 Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-09-00003 Units 1 & 2 HELB Doors Fire Protection Engineering Evaluation Summary: Purpose:  This evaluation is to document in an engineering report form the fire protection engineering evaluation for the ANO-1 and ANO-2 fire doors that are also classified as high energy line break (HELB) doors with modified or missing door latches. Basis for Acceptability:  The room is protected by a smoke detection system that alarms in the Control Room and the prompt response by the fire brigade with access to manual firefighting equipment would prevent any fire (in the unlikely event one does occur) from building sufficient pressure to open the door.
Engineering Evaluation ID: CALC-ANOC-FP-09-00011 Fire Protection Engineering Evaluation of Units 1 & 2 Aux Bldg Elevator Doors Summary: Purpose:  Evaluate elevator doors as part of the 3-hour fire boundary.
Unit 2:  335' 2040-JJ Fire Area DD 354' 2073-DD Fire Area HH 386' 2136-I Fire Area G 404' 2151-A Fire Area B-2 The elevator doors were previously evaluated in Calculation 85-E-0053-04 however the calculation will be superseded by this evaluation as part of the NFPA 805 transition project.
Basis for Acceptability:  Based on the low and moderate (Fire Zone 2136-I only) combustible loading, the availability of the smoke detection systems (and suppression system in Fire Zone 67-U) and the availability of the fire brigade with manual fire fighting equipment, the elevator doors are considered to be adequate for the hazards in the area and acceptable for the 3-hour rated fire barriers.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2019-JJ Boric Acid Condensate Tank Room No No No No No No No No No No No No 2032-JJ Spent Resin Storage Tank Room No No No No No No No No No No No No 2040-JJ Corridor No P No No No No No Yes No Yes No Yes 2068-DD Hot Machine Shop No Yes No No No No No Yes No Yes No Yes
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-80  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary
 
FRE Calculation: CALC-09-E-0008-07 Title: ANO-2 Fire Area DD Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area DD was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area DD has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-81  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards and are specifically documented within the FRE calculation.
Comments: None    VFDRs  VFDR ID: DD-01  VFDR: Fire damage to power and control cables in the area may impact EFW functions resulting in the following:
a) Failure to open 2CV-0716-1 resulting in loss of SW supply to EFW pump 2P-7B. SW is needed for a long-term source of feedwater upon depletion of condensate.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-82  Fire Area ID: DD - Unit 2 General Area 335' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: DD-02  VFDR: Fire damage to cables in the area may impact SW functions. SW is the long-term source of feedwater to the SGs via its connecti on to EFW system once condensate is depleted. Offsite power is avai lable eliminati ng the immedi ate need for SW c ooling of the EDG.
a) Spurious closure of 2CV-1400-1 (IN 92-18) will result in a loss of loop 1 SW loop to ESF loads. b) Spurious closure of 2CV-1481-1 (IN 92-18) and 2CV-1460 will result in loss of the SW return path for loop 1 to Lake Dardanelle. c) Failure of 2CV-1530-1 (Loop 1 to CCW) will prevent isolation of credited SW Loop 1 from SW Loop 2 resulting in potential pump run-out conditions. d) Loss of power and control to lake sluice gate 2CV-1472-5 if 2P-4B is the operable pump. Spurious operation of this sluice gate could result in a loss of SW loop 2.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required for components associated with SW (vital auxiliaries).
VFDR ID: DD-03  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Spurious operation of 2CV-4920-1 and 2CV-4921-1 preventing suction from the BAMTs. b) Loss of the credited charging pumps 2P-36B and 2P-36C(G). Currently an electrical raceway fire barrier system (wrap) is credited for both 2P-36B and 2P-36C and is identified for potential removal based upon the FRE.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required as 2CV-4920-1 and 2CV-4921-1 are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. b) No further action is required as 2P-36B and 2P-36C are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.
End of Fire Area DD
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-83  Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2055-JJ Lower South Piping Penetration Room 2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A or 2P-36B available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-B from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 19 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.32 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area consists of piping penetration rooms, has no automatic suppression system, and firefighting activities are from controlled manual methods using hose station(s). Flow of water from any zone in this area can be controlled by the fire brigade. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-84  Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Licensing Actions Licensing Action: Appendix R, Exemption 06, FA - EE-L (Originally a portion of FA - EE), Not Meeting III.G Criteria, NRC approval letter 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the door of issue has been modified and determined to be acceptable.
Licensing Action: Appendix R, Exemption 15, FA - EE-L (Originally a portion of FA - EE), Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-105 Penetration Seal Analysis For Penetrations 2084-09-0012 & -0016 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
Engineering Evaluation ID: CALC-89-R-2002-123 Penetration Seal Analysis For Penetration 2109-01-0064 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low and moderate fire durations, available smoke detection system, suppression system on one side and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-85  Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-127 Penetration Seal Analysis For Penetration 2055-04-0059 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on available smoke detection system and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items with equivalent exposed surface area.
 
Engineering Evaluation ID: CALC-89-R-2002-13 Penetration Seal Analysis for Penetration 2083-03-0043 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and detection to protect against the hazards of th e area, as well as limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-23 Penetration Seal Analysis for Penetration 2055-07-0173 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression to protect against the hazards of the area, as wel l as limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-46 Penetration Seal Analysis For Penetrations 2055-02-0079 & -0097 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection on both sides of the barrier, and the response by the fir e brigade to suppress a fire in the incipient stage on either side the differences identified will have no affect on the seal sin ce the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-86 Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-47 Penetration Seal Analysis For Penetrations 2055-05-0021, -0023, -0025, -0028, -0030, & -0043 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The lack of damming material on one side does not degrade the installed configuration. The installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller. A smaller seal is advantageous since it would have less exposure area to the fire and the installed configuration is sealed the full depth of the barrier ~ 18".
Engineering Evaluation ID: CALC-89-R-2002-57 Penetration Seal Analysis for Penetration 2055-07-0160 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  This configuration has differences between the tested detail and the as-b uilt penetration in that the field has a two (2)-inch larger core bore. However, the differences are considered negligible and the seal is acceptable as is for the following reasons: 1) the loading for the near side is less than one (1) hour and the far side is below grade (no loading), and 2) the near side is protected by a ionization detection system and available manual fire suppression equipment.
Engineering Evaluation ID: CALC-89-R-2002-71 Penetration Seal Analysis For Penetration 2055-01-0198 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will h ave no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-73 Penetration Seal Analysis For Penetration 2055-02-0137 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the low fire duration, smoke detection on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side, the differences identified will have no affect on the se al since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-87 Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-99 Penetration Seal Analysis For Penetration 2109-01-0032 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Engineering Evaluation ID: CALC-92-R-2026-01 Fire Barrier 86-10 Evaluation Summary: Purpose:  Evaluate duct shaft located south of column D2 and between columns 3 and 4 west of the Auxiliary Building elevator shaft. Basis for Acceptability:  The non-rated floor of this shaft is acceptable for the hazards based upon being surrounded by NRC and Insurance rated three hour barriers, no penetrating items, limited combustibles, and detection in area EE (2084-DD).
Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
Engineering Evaluation ID: CALC-ANO2-FP-06-00002 Penetration Seal FB-2055-05-0005 86-10 Evaluation Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller. A smaller seal is advantageous since it would have less exposure area to the fire.
 
Engineering Evaluation ID: CALC-ANO2-FP-06-00005 Ventilation Duct Penetrations from F.Z. 2084-DD to F.Z. 2111-T 86-10 Evaluation Summary: Purpose:  Evaluate ventilation duct penetrations to determine if it is acceptable to be utilized in a rated fire boundary. Basis for Acceptability:  The ventilation duct penetrations installed through the floor of Fire Zone 2111-T and ceiling of 2084-DD are adequately designed and supported to ensure a minimum one hour fire rating which is adequate based upon fire hazards in the area.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-88 Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3 -hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual firefighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
Engineering Evaluation ID: CALC-ANO2-FP-09-00025 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area EE Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area EE to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on factors such as combustible loading, smoke detection systems, suppression systems and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00028 Engineering Evaluation of Penetration Seals in Fire Area HH Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetration seals in Fire Area HH to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
 
Engineering Evaluation ID: CALC-ANO2-FP-11-00001 Fire Protection Engineering Evaluation for Unit 2 Fire Area EE Upper and Lower Summary: Purpose:  This evaluation is to evaluate and document the acceptability of the fire barrier between the upper EE and lower EE fire areas to limit fire spread between the two (2) fire zones. Basis for Acceptability:  The combustible loading, smoke detection systems, suppression system in upper EE, and the response by the fire brigade (with firefighting equipment in the area) would prevent a fire from growing past the incipient stage and limit the spread of a fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-89  Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Building penetrations to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual firefighting equipment for those areas in the unit's Auxiliary Buildings.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2055-JJ Lower South Piping Penetration Room No Yes No No No No No Yes No Yes No Yes 2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room No P No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-90  Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-08 Title: ANO-2 Fire Area EE-L Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2CV-1075-1 EFW discharge valve. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area EE-L was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area EE-L has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 and shows no additional DID methods are required beyond those inherent to the fire area No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-91  Fire Area ID: EE-L - South Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: EEL-01  VFDR: Fire damage to cables in the area may impact EFW functions resulting in the following:
a) Loss of control to normally closed 2CV-1075-1 results in loss of EFW flow from 2P-7B to the credited SG B.
b) Loss of power and control to 2CV-1025-1 results in loss of control (isolation) of EFW flow to the non-credited SG A.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) Recovery action associated with 2CV-1075-1.
b) No further action is required for 2CV-1025-1.
End of Fire Area EE-L
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-92  Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2111-T Lower South Electrical Penetration Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A or 2P-36B available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-B from condensate storage with SW as a backup Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 20 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.33
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-93  Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has automatic suppression system. The expected worst case flow of water from is within the capabilities of drains and scupper in the adjoining Fire Area JJ. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criter ia. Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-130 Penetration Seal Analysis for Penetration 2138-01-0030 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Based on the fact that the Thermo Lag is porous and would allow the silicone foam to adhere to it, the addition of the conduit with the insulating material would not have a detrimental effect on the ability of the three (3)-ho ur rated block-out and is thus acceptable.
Engineering Evaluation ID: CALC-ANO2-FP-06-00005 Ventilation Duct Penetrations from F.Z. 2084-DD to F.Z. 2111-T 86-10 Evaluation Summary: Purpose:  Evaluate ventilation duct penetrations to determine if it is acceptable to be utilized in a rated fire boundary. Basis for Acceptability:  The ventilation duct penetrations installed through the floor of Fire Zone 2111-T and ceiling of 2084-DD are adequately designed and supported to ensure a minimum one hour fire rating which is adequate based upon fire hazards in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3 -hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual firefighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-94 Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00025 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area EE Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area EE to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on factors such as combustible loading, smoke detection systems, suppression systems and the response by the fire brigade to suppress the fire in the incipient stage with the manual firefighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-11-00001 Fire Protection Engineering Evaluation for Unit 2 Fire Area EE Upper and Lower Summary: Purpose:  This evaluation is to evaluate and document the acceptability of the fire barrier between the upper EE and lower EE fire areas to limit fire spread between the two (2) fire zones. Basis for Acceptability:  The combustible loading, smoke detection systems, suppression system in upper EE, and the response by the fire brigade (with firefighting equipment in the area) would prevent a fire from growing past the incipient stage and limit the spread of a fire.
Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Buildings penetrations to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual fire fighting equipment for those areas in the unit's Auxiliary Buildings.
Engineering Evaluation ID: ER-ANO-2003-0397-000 Evaluate lack of structural steel 3 hour coating in Fire Zones 112-I and 2111-T. Summary: Purpose:  Evaluate lack of structural steel 3 hour coating in Fire Zones 112-I and 2111-T. Basis for Acceptability:  Lack of structural steel coating is justified based upon full suppression in 2111-T.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-95  Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2111-T Lower South Electrical Penetration Room Yes Yes No No No No Yes Yes No Yes No Yes P - Indicates a partial system is installed.
Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-96  Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-09 Title: ANO-2 Fire Area EE-U Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2CV-1026-2 EFW discharge valve  2CV-1076-2 EFW discharge valve  2EFW-5A EFW manual cross-tie valve  2EFW-5B EFW manual cross-tie valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area EE-U was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area EE-U has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights"
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-97  Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None    VFDRs  VFDR ID: EEU-01  VFDR: Fire damage to cables in the area may impact EFW functions resulting in the following:
a) Loss of control to 2CV-1036-2 (IN 92-18) resulting in potential spurious closure and loss of EFW flow from 2P-7B to the credited SG B.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-98  Fire Area ID: EE-U - Lower South Electrical Penetration Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: EEU-02  VFDR: Fire damage to control cables could result in a loss of the Control Room trip capability of the Reactor Coolant Pumps 2P-32A an d B. Securing the pumps is required to assure normal pressurizer spray is secured and prevent potential RCP seal damage.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Additional circuit analysis on the RCP pump motors concluded that fire damage to the RCP cables would not prevent the Control Room from tripping the pumps for a fire in Fire Area EE-U. Since the Control Room will retain the ability to trip the RCPs aft er fire in Fire Area EE-U, this is not a VFDR, and the associated basic events have been excluded from the model in the Post-Transition Baseline Case.
VFDR ID: EEU-03  VFDR: Loss of control circuit to Pressurizer Heater bank #1resulting in the loss of remote Control Room trip capability needed to sec ure the heaters post fire.
Loss of this function could challenge the Pressure Control Performance Criterion. This condition represents a variance from th e deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
End of Fire Area EE-U
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-99  Fire Area ID: FF - Emergency Feedwater Pump Room (Motor Driven) Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2025-JJ EFW Pump Room (Motor Driven)
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 21 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.13
 
Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station from Fire Area DD. This room is lower than fire area and separated from redundant equipment by a watertight door. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-100 Fire Area ID: FF - Emergency Feedwater Pump Room (Motor Driven) Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3-hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual fire fighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
Engineering Evaluation ID: CALC-ANO2-FP-09-00034 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area FF Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area FF to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetration seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection system (near side only) and the response by the fire brigade to suppress the fire in the incipient stage.
 
Engineering Evaluation ID: CALC-ANOC-FP-07-00003 Water Tight Fire Doors Evaluation Units 1 & 2 86-10 Evaluation Summary: Purpose:  10 CFR 50, Appendix R, requires fire barriers separating redundant trains of safe shutdown equipment to be separated by rated fire barriers that includes all opening, i.e. fire doors, fire dampers, penetration seals, etc. However, th ere are locations in the plant that are susceptible to flooding that are equipped with water tight doors that are not UL Listed or FM Approved to be used in a 3-hour rated barrier. Thus, this evaluation will determine if the water tight fire doors are acceptab le to be used in the 3-hour rated fire barriers at ANO by comparison with UL Listed 3-hour fire doors. The guidance provided in Generic Letter 86-10 will be utilized for the evaluation. Basis for Acceptability:  The water tight doors installed at ANO in 3-hour rated fire barriers have been determined to be acceptable for use based on the hazards in the areas. Although these doors are not 3-hour rated fire doors, they will provide the protection needed in the areas they are used.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-101 Fire Area ID: FF - Emergency Feedwater Pump Room (Motor Driven) Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2025-JJ EFW Pump Room (Motor Driven) No Yes No No No No No Yes No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
 
VFDRs This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area FF
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-102 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 97-R Unit-1 Cable Spreading Room 129-F Control Room 2199-G Control Room 2119-H Printer Room 2136-I Health Physics Room 2137-I Upper South Electrical Penetration Room 2150-C Core Protection Calculator Room (Old CPC Room) 2098-C CPC Room 2098-L Unit 2 Cable Spreading Rooms
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-103 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36B available with gravity feed from BAMTs, normal injection path isolated, and manual alignment to RCS using HPSI header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-B from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV 2A-4 aligned to onsite EDG. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Control Room abandonment. 6. Process Monitoring Instrumentation is available in the Technical Support Center via SPDS to monitor pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 22 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.34
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-104 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. Automatic suppression in this area is limited to the cable spreading room. The flow rate of the spreading room deluge system can result in ponding in other areas, but only in 2093-P can it present a problem. If 2093-P is impacted a single train of on-site power can be impacted but the opposite train is unaffected. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Licensing Action: Appendix R, Exemption 12, FA G, Not Meeting III.G.3 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Licensing Action: Appendix R, Exemption 14, FA  G, Not Meeting III.G.3 Criteria, NRC approval letter 2CNA 108802 dated 10/26/1988. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations Engineering Evaluation ID: CALC-87-E-0024-01 Evaluation of Fire Seal Requirements at Doorway Elev. 386 Feet ANO-1 CR to ANO-2 CR Summary: Purpose:  Evaluate seal requirements at doorway between ANO-1 and ANO-2 Control Rooms. Basis for Acceptability:  The embedded steel plate located at the joint in the doorway between Control Rooms prevents a fire originating at areas outside Fire Area G from spreading into the Control Rooms.
 
Engineering Evaluation ID: CALC-89-R-2002-01 Penetration Seal Analysis for Penetration 2098-07-0094 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area as well as superior penetration seal construction as compared to the detail, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-105 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-02 Penetration Seal Analysis for Penetration 2098-07-0098 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability: Because of availability of fire suppression and detection to protect against the hazards of the area, as well as limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-03 Penetration Seal Analysis for Penetration 2098-07-0098 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-06 Penetration Seal Analysis for Penetration 2098-04-0068 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  The qualification of multiple conduits through a silicone foam blockout is contained in Fire Test Report SWRI 03-6004-006 (CTP-1001A) where test blockout number six demonstrated the ability of multiple conduits to successfully pass the ASTM-E119 fire endurance test. This evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-07 Penetration Seal Analysis for Penetration 2098-04-0066 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  This configuration has differences between the tested detail and the as-b uilt penetration in that the re is one additional cable tray in the as-built. The qualification of multiple cable trays through a silicone foam blockout is contained in the Fire Test Report SWRI 03-6004-006 (CTP-1001A) where test blockout number one qualifies multiple cable trays through a silicone foam blockout. Overall this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-106 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-08 Penetration Seal Analysis for Penetration 2098-07-0096 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-109 Penetration Seal Analysis For Penetration 2153-01-0048 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-119 Penetration Seal Analysis for Penetration 2118-02-0693 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, one side continuously manned, and the fire brigade's response are adequate for the hazards. The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains equivalent areas.
 
Engineering Evaluation ID: CALC-89-R-2002-121 Penetration Seal Analysis For Penetration 2150-01-0020 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains equivalent areas.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-107 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-124 Penetration Seal Analysis For Penetration 2154-01-0039 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on available smoke detection system and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and th e installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-126 Penetration Seal Analysis For Penetration 2139-03-0092 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on available smoke detection system, and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
 
Engineering Evaluation ID: CALC-89-R-2002-130 Penetration Seal Analysis for Penetration 2138-01-0030 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Based on the fact that the Thermo Lag is porous and would allow the silicone foam to adhere to it, the addition of the conduit with the insulating material would not have a detrimental effect on the ability of the three (3)-ho ur rated block-out and is thus acceptable.
Engineering Evaluation ID: CALC-89-R-2002-15 Penetration Seal Analysis for Penetration 2108-07-0076 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and detection to protect against the hazards of th e area, as well as the limited combustible loading, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-108 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-30 Penetration Seal Analysis for Penetration 2128-02-0006 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Considering that the seal 1) meets or exceeds all the specified requirements except for the type/number of penetrating items, 2) the deviating penetrants are small and will transfer a relatively small amount of heat, and 3) the favorable conditions associated with the physical location of the seal, this seal is judged to be acceptable for thi s specific application.
Engineering Evaluation ID: CALC-89-R-2002-32 Penetration Seal Analysis For Penetration Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression systems on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identifi ed will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is equivalent in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-37 Penetration Seal Analysis for Penetration 2118-01-0038, 47, 55 Summary: Purpose:  Evaluate penetration seals to determine if it is acceptable to utilize them in a three (3)-hour rated fire boundary since the seals deviate from the tested configuration. Basis for Acceptability:  The seals have been determined to contain conservative deviations, but are acceptable for the following reasons: 1. Availability of fire suppression and detection to protect against the hazards of the area. 2. Continuous manning of one side of the barrier (Control Room side). 3. The cross sectional area of the conduits in the field configuration is less than the tested detail. 4. The available free area of the field configuration is less than the detail. 5. The fire barrier does not separate redundant safe shutdown equipment. 6. CTP1001A demonstrates that 6" of SF is sufficient to provide a three hour rated fire seal.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-109 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-51 Penetration Seal Analysis For Penetration 2118-02-0677 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller. A smaller seal is advantageous since it would have less exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-62 Penetration Seal Analysis For Penetration 2154-01-0078 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will h ave no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-63 Penetration Seal Analysis For Penetration 2154-01-0081 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will h ave no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-65 Penetration Seal Analysis For Penetration 2098-09-0174 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-110 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-84 Penetration Seal Analysis For Penetrations 2096-03-0095 & -0096 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-88 Penetration Seal Analysis For Penetrations 2118-01-0221 & -0228 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-90 Penetration Seal Analysis For Penetrations 2118-01-0048, -0239, -0259, & -0261 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-91 Penetration Seal Analysis For Penetration 2118-02-0334 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-111 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-94 Penetration Seal Analysis For Penetration 2153-01-0116 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and manual fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3 -hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual firefighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00010 Penetration Seal Analysis for Penetration 2139-01-0041 Summary: Purpose:  Evaluate and document the acceptability of ANO-2 penetrations FB-2139-01-0041 to be used in a 3-hour rated fire area boundary based on approved fire tests. Basis for Acceptability:  The fire duration on the bottom side of the seal is less than the tested configuration (top side has a sprinkler system) thus the seal is adequate for the hazard.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00012 Fire Protection Engineering Evaluation for Ventilation Opening in Fire Area 2136-I Summary: Purpose:  Evaluate the lack of a fire damper in the ventilation openings in Fire Zone 2136-I. Basis for Acceptability:  The opening is considered to be adequate for the hazards in the area since the fire barrier does not separate redundant trains of safe shutdown equipment. Additionally, the near side is protected by a smoke detection system and partial suppression system and the fire brigade would respond in a timely manner to begin firefighting activities before a fire could grow past the incipient stage.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-112 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00017 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area B-2 Summary: Purpose:  This evaluation is to document the acceptability of unit 2 penetration seals in Fire Area B-2 to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  Based on the combustible loading, the smoke detection and the response by the fire brigade to suppress a fire in the incipient stage on either side, the penetration seals are considered to be adequate for the hazards in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00030 Fire Protection Engineering Evaluation for Penetrations Seals in Fire Area TT Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area TT to be used in a 3-hour rated fire area boundary. Basis for Acceptability: The penetration seals are considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, suppression systems (far sides only) and the response by the fire brigade to suppress the fire in the incipient stage.
Engineering Evaluation ID: CALC-ANOC-FP-08-00002 Fire Area G Partial Suppression 86-10 Evaluation Summary: Purpose:  Evaluate lack of complete area wide suppression in Fire Area G. Basis for Acceptability: Based upon detection and fire boundaries it has been determined that the installed fire protection features will promptly detect any fire in its incipient stages and the fire be extinguished to limit any damage to one train of equipment.
Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Buildings penetrations to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual fire fighting equipment for those areas in the unit's Auxiliary Buildings.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-113 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-09-00011 Fire Protection Engineering Evaluation of Units 1 & 2 Aux Bldg Elevator Doors Summary: Purpose:  Evaluate elevator doors as part of the 3-hour fire boundary.
Unit 2:  335' 2040-JJ Fire Area DD 354' 2073-DD Fire Area HH 386' 2136-I Fire Area G 404' 2151-A Fire Area B-2 The elevator doors were previously evaluated in Calculation 85-E-0053-04 however the calculation will be superseded by this evaluation as part of the NFPA 805 transition project.
Basis for Acceptability:  Based on the low and moderate (Fire Zone 2136-I only) combustible loading, the availability of the smoke detection systems (and suppression system in Fire Zone 67-U) and the availability of the fire brigade with manual firefighting equipment, the elevator doors are considered to be adequate for the hazards in the area and acceptable for the 3-hour rated fire barriers.
 
Engineering Evaluation ID: CALC-ANOC-FP-09-00013 Engineering Evaluation for Penetration Seals in Fire Area G Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area G to be used in a 3-hour rated fire area boundary.
Basis for Acceptability:  The penetration seals have been considered to be adequate for the hazards in the area based on the combustible loading, smoke detection system, or line type heat detectors, suppression systems and the response by the fire brigade to suppress the fire in the incipient stage.
 
Engineering Evaluation ID: CALC-ANOC-FP-10-00001 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area G, Part 2 Summary: Purpose:  Evaluate and document the acceptability of penetrations in Fire Area G to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seals are considered to be adequate for the hazards in the area based on the combustible loading, smoke detection system or line type heat detectors, suppression system or the response by the fire brigade to suppress the fire and prevent significant damage (with firefighting equipment in the area).
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-114 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 97-R ANO-1 Cable Spreading Room P Yes No No No No No No Yes Yes Yes Yes 129-F Control Room P Yes No No No No No No Yes Yes Yes Yes 2199-G Control Room No Yes No No No No No Yes No Yes No Yes 2119-H Printer Room No Yes No No No No No Yes No Yes No Yes 2136-I Health Physics Room P Yes No No No No Yes Yes Yes Yes Yes Yes 2137-I Upper South Electrical Penetration Room Yes Yes No No No No Yes Yes Yes Yes Yes Yes 2150-C Core Protection Calculator Room (Old CPC Room) No Yes No No No No No Yes No Yes No Yes 2098-C CPC Room Yes Yes No No No No Yes Yes Yes Yes Yes Yes 2098-L Unit 2 Cable Spreading Rooms Yes Yes No No No No Yes Yes No Yes Yes Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-115 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-10 Title: ANO-2 Fire Area G Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2CV-4816 and 2CV-4817 Letdown isolation valves  2P-32A reactor coolant pump  2P-32B reactor coolant pump  2P-32C reactor coolant pump  2P-32D reactor coolant pump  New AFW pump and suction & discharge valves  2CV-1036-2 EFW discharge valve  2CV-1039-1 EFW discharge valve  2CV-1075-1 EFW discharge valve  2CV-1076-2 EFW discharge valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. (continued)
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-116 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
Summary (continued) The following  modifications are area specific and credited to reduce risk  in this fire area:  2CV-4698-1 circuit protected to prevent spurious  2SV-4669-1 and 2SV-4670-2 circuits modified to prevent spurious  2CV-1036-2 circuit protected to prevent spurious and eliminate IN 92-18 issues  2CV-1039-1 circuit protected to prevent spurious and eliminate IN 92-18 issues  2CV-1075-1 circuit protected to prevent spurious and eliminate IN 92-18 issues  2CV-1076-2 circuit protected to prevent spurious and eliminate IN 92-18 issues IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. EFW discharge valve 2CV-1036-2, 2CV-1039-1, 2CV-1075-1, and 2CV-1076-2 modifications resolve the IN 92-18 issues in this fire area. Additional Fire Area Considerations The available detection systems in Fire Area G were credited in the manual suppression curves in the HGL and MCA analysis to limit the growth of the fire. The wet pipe water suppression system in Fire Zone 2136-I is credited in the HGL and the MCA analysis to limit the fire growth in Fire Area G. The automatic preaction suppression system in Fire Zone 2137-I is credited i n the HGL and MCA analysis to limit the fire growth in Fire Area G. The automatic Halon suppression system in Fire Zone 2098-C (CPC Room) is credited in the HGL and MCA analysis to limit the fire growth in Fire Area G. Fire Area G has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights"
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-117 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods for Echelon 1 (prevention) and Echelon 2 (detection & suppression). Defense in depth actions (Echelon 3) identified to protect selected equipment listed by verifying/performing the following:  2P-89B HPSI pump secured  2P-60B LPSI/SDC pump tripped  2P-35B Containment spray pump tripped  2CV-5630-1 RWT tank outlet valve closed (prevent RWT inventory draining to sump or for HPSI alignment to sump if required) 2CV-5631-2 RWT tank outlet valve closed (prevent RWT inventory draining to sump or for HPSI alignment to sump if required) 2CV-4920-1 BAMT outlet valve opened (alignment to charging pumps)  2CV-4921-1 BAMT outlet valve opened (alignment to charging pumps)CV-1206 RCP seal injection valve closed  2CV-4873-1 VCT outlet valve closed  2T-1 Pressurizer heaters secured  2P-36A/B/C Charging pumps secured and locally operated following valve alignment  2K-4B (EDG #2) locally secured  2A-4 ES bus locally aligned/operated - this includes actions for 2A-408, 2A-409, and 2A410  2B-6 ES load center locally aligned/operated  2D24-2 breaker for 2C108 Power Supply  2D24-4 breaker for DC Control Power for 2A4  2D24-6 breaker for 2E21 Power Supply  2D24-8 breaker for 2C108 Power Supply  2D24-9 for DC Control Power for 2B6  2D24-10 breaker for High Point Vent Panel 2C336-2  2CV-4840-2 Charging header opened  2CV-1504-2 Service water outlet valve to EDG #2 opened  2CV-4950-2 RWT suction for charging pumps opened  2P-4C service water pump started  2CV-0795-2 Suction valve for 2P-7A opened  2VUC-30 SPDS room HVAC transferred to ANO-1 power source if required No procedural changes, recoveries, or modifications are needed for maintenance of DID for this fire area
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-118 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and the industry accepted standards that are specifically documented within the FRE calculation.
Comments: None VFDRs  VFDR ID: G-01  VFDR: Fire damage to control cables in the area may affect pressure control functions. The control circuits affected result in the following:
: a. Loss of control and multiple spurious operations of pressurizer LTOP relief valves may result in an RCS depressurization. The combination that can cause this is 2CV-4740-2 and either 2CV-4741-1 or 2CV-4698-1.
: b. Loss of control to pressurizer heater banks #1 through #6.
Loss of these functions could challenge the Pressure Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805, Section 4.2.4, with the following actions:
: a. Modification is required to prevent spurious opening 2CV-4698-1. No further action is required for either 2CV-4740-2 or 2CV-4741-1.
: b. Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-119 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-02  VFDR: ANO-2 manual actions performed for Alternate Shutdown isolate potential drain, vent, and flow diversion paths that can compromi se the RCS or available makeup inventory. Charging pumps are aligned through the HPSI header by opening manual valve 2CVC-115. The preferred source of borated water for charging is to use the BAMTs followed by alignment to the RWT. Access to align the BAMTs may not be immediately available, as it requires transit of Fire Zone 2136-I in Fire Area G. Should an inadvertent drain-down of RWT inventory to the containment sump occur, then the HPSI pumps are aligned to the sump, and used for inventory make-up. Fire damage to control cables in the area may affect inventory control functions resulting in the following:
: a. Loss of control prevents closing 2CV-4816 and 2CV-4817 (AOVs) to secure RCS letdown. b. Loss of control and spurious opening of either isolation valve 2CV-5649-1 or 2CV-5650-2 could result in RWT drain-down to containment sump. c. Loss of control and the inability to secure RCPs 2P-32A, 2P-32B, 2P-32C, and 2P-32D may result in a seal LOCA. d. Loss of control and spurious start of Containment Spray pumps 2P-35A and B could result in a flow diversion from the RWT. e. Loss of control and ability to secure/run as necessary Charging pumps 2P-36A, 2P-36B, and 2P-36C. 2P-36A is not credited to run as red train power is not restored. f. Loss of control and ability to secure/run as necessary HPSI pumps 2P-89A, 2P-89B, and 2P-89C. 2P-89A is not credited to run in a scenario where RWT inventory drains to the sump as red train power is not restored. g. Loss of control and spurious closure of 2CV-4920-1and 2CV-4921-1prevent charging pumps from being aligned to the BAMTs.h. Loss of control and inability to open 2CV-4950-2 prevent Charging pumps from being aligned to the RWT. i. Loss of control and inability to close Charging flow control valve 2CV-4840-2 may result in inability to secure pressurizer spray.j. Loss of control and inability to open HPSI control valve 2CV-5015-1 (IN 92-18) and 2CV-5055-1 (IN 92-18) prevent Charging pump alignment through the HPSI injection path. k. Loss of control and inability to close RWT outlet valve 2CV-5630-1 or 2CV-5631-2 could result in inability to align HPSI pumps to containment sump. l. Loss of control and inability to open 2CV-5648-2 (IN 92-18) can prevent alignment of HPSI to the sump. m. Loss of control and spurious opening of RCS vents 2SV-4636-1, 2SV-4636-2, 2SV-4668-1, 2SV-4668-2, 2SV-4669-1, and 2SV-4670-2 result in inventory loss in excess of the Charging pumps capacity. n. Loss of control and inability to close 2CV-4941-2 (AOV) for VCT makeup demineralized water may result in RCS dilution. o. If alignment to the sump is required then the following manual valves will need repositioning as necessary:  2BS-26, 2SI-9A, and 2SI-9B. p. If alignment for the Charging pumps is used then manual valves 2CVC-1168 and 2CVC-1169 will need to be opened for local monitoring (2PI-4843) of Charging pump NPSH.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-120 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-02 (continued)
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805, Section 4.2.4, with the following actions:
: a. Recovery actions to locally close 2CV-4816 and 2CV-4817. b. No further action for 2CV-5649-1 or 2CV-5650-2. c. Recovery actions to locally secure RCPs 2P-32A, 2P-32B, 2P-32C, and 2P-32D. d. No further action required for 2P-35A. Defense in depth action for 2P-35B.
: e. 2P-36A, 2P-36B, and 2P-36C are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. Defense in depth action to secure/operated Charging pumps. f. No further action required for 2P-89A and 2P-89C. Defense in depth action for 2P-89B.
: g. 2CV-4920-1 and 2CV-4921-1 are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. Defense in depth action to open these valves. h. 2CV4950-2 is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. Defense in depth action to open this valve. i. 2CV-4840-2 is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. Defense in depth action to open this valve. j. No further action for 2CV-5015-1 and 2CV-5055-1 as HPSI is only credited in the Fire PRA for loss of coolant or feed-and-bleed sequences. k. No further action required for 2CV-5630-1 and 2CV-5630-2. Defense in depth action to close these valves. l. No further action required for 2CV-5658-2.
: m. Modification is required to prevent spurious opening of 2SV-4669-1 and 2SV-4670-2. The ability to maintain 2SV-4669-1 and 2SV-4670-2 closed results in no further actions required for 2SV-4636-1, 2SV-4636-2, 2SV-4668-1, and 2SV-4668-2. n. No further action required for 2CV-4941-2.
: o. No further action required for 2BS-26, 2SI-9A, and 2SI-9B.
: p. 2CVC-1168 and 2CVC-1169 are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-121 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-03  VFDR: ANO-2 provides actions to align flow from the turbine driven EFW pump 2P-7A to SG-B to control decay heat removal. SG-A is isolated. Fire damage to control cables in the area may affect EFW functions resulting in the following:
: a. Loss of control and spurious closure of isolation valve 2CV-1050-2 (IN 92-18) in steam path from SG-B to 2P-7A turbine driver 2K-3 results in failure of EFW pump. b. Loss of control and inability to open EFW pump 2P-7A steam admission valve 2CV-0340-2 (IN 92-18) results in failure of EFW pump. c. Loss of control to EFW panel 2C-143 results in a loss of turbine driven pump 2P-7A. d. Loss of control and spurious closure of 2CV-0795-2 results in loss of stored condensate source for EFW pump 2P-7A. e. Loss of control and inability to open 2CV-0711-2 (IN 92-18) and close drain valve 2SV-0712-2 prevent alignment of EFW pump 2P-7A suction to Service Water (SW) for long term source of feedwater after depletion of stored condensate. f. Loss of control and spurious closure of 2P-7A to SG-B valves 2CV-1039-1 (IN 92-18) and/or 2CV-1076-2 (IN 92-18) results in loss of EFW flow to credited steam generator. g. Loss of control and spurious opening of 2P-7A to SG-A block valve 2CV-1026-2 (IN 92-18) and 2CV-1037-1 (IN 92-18) diverts flow to non-credited steam generator. h. Loss of control and spurious start of non-credited motor driven EFW pump 2P-7B may result in overcooling. i. Loss of control for MSIV valve 2CV-1060-2 (AOV) and its associated bypass 2CV-1090-2 (IN 92-18) result in inability to isolate credited SG-B. j. Loss of control for MSIV valve 2CV-1010-1 (AOV) and its associated bypass 2CV-1040-1 (IN 92-18) result in inability to isolate non-credited SG-A. k. Loss of control capability of atmospheric dump block valve 2CV-1052 (IN 92-18) resulting in inability to isolate credited SG-B. l. Loss of control and spurious opening of SG-B blowdown block valve 2CV-1066-1(AOV) resulting in inability to isolate credited SG-B. m. Loss of control capability of atmospheric dump block valve 2CV-1002 (IN 92-18) resulting in inability to isolate non-credite d SG-A. n. Loss of control and the inability to secure AFW pump 2P-75 and the Condensate pumps 2P-2A, 2P-2B, 2P-2C, and 2P-2D results in an uncontrolled feed-water source to the steam generators.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-122 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-03 (continued)
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
: a. No further actions for 2CV-1050-2. b. No further actions for 2CV-0340-2. c. No further actions for 2C-143 (2P-7A).
: d. Defense in depth action only for 2CV-0795-2.
: e. No further actions for 2CV-0711-2.
: f. Recovery action and modification for 2CV-1039-1 and 2CV-1076-2. g. No further action for 2CV-1026-2 and 2CV-1037-1. h. No further actions for 2P-7B.
: i. No further actions for 2CV-1060-2 and 2CV-1090-2.
: j. No further actions for 2CV-1010-1 and 2CV-1040-1.
: k. No further actions for 2CV-1052.
: l. Recovery action to locally close 2CV-1066-1. m. No further actions for 2CV-1002. n. No further actions for 2P-75 and the Condensate pumps 2P-2A, 2P-2B, 2P-2C, and 2P-2D.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-123 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-04  VFDR: Fire damage to control cables in the area may affect vital auxiliary electrical functions resulting in the following:
: a. Loss of control and spurious closure of 2A-410 and 2A-310 resulting in the 2A-4 bus inadvertently tied to the 2A-3 bus. b. Loss of control and inability to trip supply breaker 2A-409 prevents transfer to EDG 2K-4B. c. Loss of control and the inability to close breaker 2A-408 preventing EDG 2K-4B from energizing engineered safety bus 2A-4. d. Loss of control to 480VAC load center 2B-6 resulting in a loss of power to ESF green train motor control centers. e. Loss of control to EDG 2K-4B, starting air solenoids 2SV-2829-2, 2SV-2830-2,  2SV-2831, and meter and relay cabinet 2E-21 prevents remote start and results in the loss of on-site power to ESF bus 2A-4. f. Loss of control to diesel fuel transfer pump 2P-16B results in loss of long term fuel source for EDG 2K-4B. g. Loss of control to EDG 2K-4B room ventilation fans 2VEF-24C and 2VEF-24D results in loss of onsite power source. h. Spurious start of non-credited EDG 2K-4A and the inability to secure, combined with loss of jacket cooling, can result in lo ss of a redundant on-site power source.
Loss of these functions could challenge the Vital Auxiliaries (Electrical) Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
 
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805, Section 4.2.4, with the following actions:
: a. Defense in depth action only for 2A-410 and no further action for 2A-310. b. Defense in depth action only for 2A-409.
: c. Defense in depth action only for 2A-408.
: d. Defense in depth action only for 2B-6.
: e. Defense in depth action only for 2K-4B.
: f. No further action for 2P-16B. g. No further action for 2VEF-24C and 2VEF-24D. h. No further action for 2K-4A.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-124 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-05  VFDR: Fire damage to control cables in the area may affect Service Water (SW) functions. SW provides cooling to the emergency diesel generator, HPSI pump seal & bearing coolers, and is the assured long-term source of feedwater to the steam generators via its connection to EFW system once depletion of condensate occurs. The control circuits affected result in the following:
: a. Loss of control and spurious closure of sluice gates 2CV-1474-2 (2P-4C) and 2CV-1472-5 (2P-4B) depriving the SW pumps of a suction source. b. Loss of control and spurious closure of either SW discharge valve 2CV-1460 (AOV) or 2CV-1480-2 (IN 82-18) to reservoir (Lake Dardanelle) resulting in a loss of SW flow. c. Loss of control and inability to close one of the two ACW isolation valves 2CV-1425-1(IN 92-18) or 2CV-1427-2 (IN 92-18) results in a diversion of SW. d. Loss of control and spurious closure of SW cross tie valves 2CV-1421-2 (IN 92-18) or 2CV-1422-2 (IN 92-18) if 2P-4C is OOS (Out of Service) with 2P-4B aligned to SW header 2. e. Loss of control and spurious opening of both cross tie valves 2CV-1418-1(IN 92-18) and 2CV-1419-1(IN 92-18) resulting in flow diversion to SW header 1. f. Loss of control and spurious closure of SW header #2 block valve 2CV-1406-2 (in 92-18) deprives ESF loads (HPSI Pump, etc.) of essential cooling water. g. Loss of control and spurious closure of EDG heat exchangers outlet valve 2CV-1504-2 results in loss of heat removal. h. Loss of control and inability to close the SW to fuel pool heat exchanger block valve 2CV-1526-2 (IN 92-18) resulting in SW header 2 and 1 being cross-tied and potential for SW pump run out. i. Loss of control and inability to close the SW to CCW block valve 2CV-1531-2 (IN 92-18) resulting in SW header 2 and 1 being cross-tied and potential for SW pump run out with 1 pump running. j. Loss of control for SW pump 2P-4C an d/or 2P-4B(G) results in loss of SW.
Loss of these functions could challenge the Vital Auxiliaries (SW) Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-125 Fire Area ID: G - Unit 2 Alternate Shutdown Areas Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: G-05 (continued)
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805, Section 4.2.4, with the following actions:
: a. No further actions for 2CV-1474-2 and 2CV-1472-5. b. No further actions for 2CV-1460 and 2CV-1480-2. c. No further actions for 2CV-1425-1 and 2CV-1427-2.
: d. No further actions for 2CV-1421-2 and 2CV-1422-2.
: e. No further actions for 2CV-1418-1 and 2CV-1419-1.
: f. No further action for 2CV-1406-2. g. Defense in depth action only for 2CV-1504-2. h. No further action for 2CV-1526-2.
: i. No further action for 2CV-1531-2.
: j. No further action for 2P-4B(G). Defense in depth action only for 2P-4C.
VFDR ID: G-06  VFDR: Fire damage to control cables can result in the loss of the ANO-2 power source to the SPDS room cooling. Room heat-up could challenge SPDS equipment required to maintain Alternate Shutdown indication in the TSC (Technical Support Center).
: a. Loss of power to SPDS room cooler 2VUC-30 results in room heatup and loss of instrumentation required for alternate shutdown.
Loss of this function could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Condition Report CR-ANO-2-2001-00396 documents that ventilation is assured for this fire. A defense in depth action is taken t o realign the power supply to an available ANO-1 source via the installed transfer switch.
End of Fire Area G
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-126 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2076-HH Electrical Equipment Room 2081-HH Upper North and Lower North Piping Penetration Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-A from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. Variance from the determinist ic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 23 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.14
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-127 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station. Flow of water from any zone in this area can be controlled by the fire brigade. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Licensing Action: Appendix R, Exemption 08, FA  GG, Not Meeting III.G.3 Criteria , NRC approval letter 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-129 Penetration Seal Analysis for Penetration 2048-05-012 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Based on the similarity of tested and installed penetrating items, the availability of fire extinguis hers and detection to protect against the hazards of the area as well as the limited combustible loading, it is considered that this penetration seal offers equivalent protection to the surrounding hazards and therefore the deviation from the detail is negligible and considered acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-31 Penetration Seal Analysis for Penetration 2048-06-0040, 41, 42 Summary: Purpose:  Evaluate penetration seals to determine if it is acceptable to utilize them in a three (3)-hour rated fire boundary since the seals deviate from the tested configuration. Basis for Acceptability:  Considering that the seals 1) meet or exceed all the specified requirements except for the size of penetrating items, 2) the seal depth is over twice that of the specified detail, and 3) the favorable conditions associated wit h the physical location of the seals, these seals are judged to be acceptable for this specific application.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-128 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-34 Penetration Seal Analysis for Penetration 2081-02-0155 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Considering that 1) the seal has inherent features that add to its fire resistivity, 2) the location of each seal is favorable with respect to the hazards in the area, and 3) detection and manual suppression are available to mitigate the consequences of a fire, the modified seal is judged to be acceptable for this specific application.
Engineering Evaluation ID: CALC-89-R-2002-36 Penetration Seal Analysis For Penetration 2091-01-0087 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is an equivalent in heat transmission.
 
Engineering Evaluation ID: CALC-ANO2-FP-06-00006 Piping Chases in Room 2076 Going to Room 2091 86-10 Evaluation Summary: Purpose:  Evaluate two piping chases to determine if they provide an acceptable fire barrier between Fire Areas GG and B-3. Basis for Acceptability:  The installed penetrations are equivalent to the fire tested configurations of electrical conduit penetrations and fire doors. The access panels were verified to meet the structural integrity and heat transfer minimum standards for fire rated doors
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3 -hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual fire fighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-129 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00014 Fire Protection Engineering Evaluation for Unit 2 Fireflex Seals Summary: Purpose:  This evaluation is to evaluate ANO-2 penetration seals utilizing fireflex as a sealing material and document their acceptability to be used in a 3 - hour rated fire area boundary based on the hazards in the area. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection systems, and the fire brigade's response and are adequate for the hazards.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00027 Fire Protection Engineering Report of Penetration Seals in Fire Area B-3 Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetration seals in Fire Area B-3 to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is adequate for the hazards based on the combustible loading, smoke detection systems, suppression system on the top side (Fire Zone 2112-BB) and the response by the fire brigade to suppress the fire in the incipient stage with the manual fire fighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00035 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area GG Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area GG to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual fire fighting equipment in the area.
Engineering Evaluation ID: CALC-ANOC-FP-09-00003 Units 1 & 2 HELB Doors Fire Protection Engineering Evaluation Summary: Purpose:  This evaluation is to document in an engineering report form the fire protection engineering evaluation for the ANO-1 and ANO-2 fire doors that are also classified as HELB doors with modified or missing door latches. Basis for Acceptability:  The room is protected by a smoke detection system that alarms in the Control Room and the prompt response by the fire brigade with access to manual fire fighting equipment would prevent any fire (in the unlikely event one does occur) from building sufficient pressure to open the door.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-130 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Buildings penetrations to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual fire fighting equipment for those areas in the unit's Auxiliary Buildings.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2076-HH Electrical Equipment Room No Yes No No No No No Yes No Yes No Yes 2081-HH Upper North and Lower North Piping Penetration Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-131 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-11 Title: ANO-2 Fire Area GG Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2CV-0789-1 EFW pump condensate suction valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area GG was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area GG has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-132 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: GG-01 VFDR: Fire damage to control cables in the area result in the potential spurious operation of valves 2CV-1480-2 (IN 92-18) (Loop 2 SW Return to Lake Dardanelle), 2CV-1481-1 (IN 92-18) (Loop 1 SW Return to Lake Dardanelle), 2CV-1525-1 (SW to fuel pool HX 2E27A), and 2CV-1530-1 (Loop 1 SW to CCW). This could result in failure of either valve in an undesired position with the inability to reposition. SW provides cooling to the EDGs, shutdown cooling heat exchangers, and LPSI pumps. In addition, the SW system provides a long-term source of feedwater to the SGs via its connection to the EFW system. Off site power is available eliminating the immediate need for SW cooling of the EDG.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required for components associated with SW (vital auxiliaries).
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-133 Fire Area ID: GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: GG-02  VFDR: Fire damage to control cables could potential impact the following:
a) Loss of Control Room capability for credited EFW pump 2P-7A to SG 2E24A. b) Failure to control MSIV 2CV-1010-1 and the spurious operation of associated MSIV bypass 2CV-1040-1 preventing isolation of SG 2E24A c) Spurious operation of main steam inlet valve 2CV-0340-2 to EFW pump 2P-7A turbine driver 2K-3. d) Spurious operation of valve 2CV-1037-1 (IN 92-18) providing EFW from 2P-7A to SG 2E24A isolation. e) Loss of isolation capability to SG 2E24B through atmospheric dump valve 2CV-1052 and MSIV 2CV-1060-2. f) Spurious closure of 2CV-0711-2, SW to pump 2P-7A, upon depletion of condensate.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2P-7A. b) No further action is required for 2CV-1010-1 and 2CV-1040-1. c) No further action is required for 2CV-340-2. d) No further action is required for 2CV-1037-1. e) No further action is required for 2CV-1052 and 2CV-1060-2. f) No further action is required for 2CV-0711-2.
End of Fire Area GG
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-134 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2063-DD Sample Room 2072-R Upper Volume Control Tank Room, Lower Tank & Pump Room 2073-DD Access Room, Pump Room, Tank Room (2B62 & Resin Addition Room) 2096-M Motor Control Center (2B63) 2106-R Degasifier Vacuum Pump Room 2107-N Corridor (North of Stairway 2001)
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-B from condensate storage with SW as a backup Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-135 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 24 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.15 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has only partial automatic suppression system and the majority of firefighting activities are from controlled manual methods using hose station(s). Ponding is minimal due to large area. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Engineering Evaluations Engineering Evaluation ID: CALC-84-D-2002-01 Fire Resistant Hatch Covers Summary: Purpose:  Design fire resistant hatch covers for installation in opening located in rooms 2040 (Elev. 335') and 2069 (Elev. 354
').Basis for Acceptability:  Hatches are designed to withstand the effects of a 3-hour fire from below and a 1-hour fire from abov e which is adequate for the hazard.
Engineering Evaluation ID: CALC-89-R-2002-04 Penetration Seal Analysis for Penetration 2073-05-0140 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Licensing Action: Appendix R, Exemption 09, FA - HH, Not Meeting III.G.3 Crit eria, NRC approval lette r 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-136 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-09 Penetration Seal Analysis for Penetration 2073-05-013 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-10 Penetration Seal Analysis for Penetration 2073-05-0152 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-105 Penetration Seal Analysis For Penetrations 2084-09-0012 & -0016 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
 
Engineering Evaluation ID: CALC-89-R-2002-120 Penetration Seal Analysis For Penetration 2066-03-0018 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-137 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-122 Penetration Seal Analysis for Penetration 2279-03-0028 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Based on the similarity of tested and installed penetrating items, the availability of fire extinguis hers and detection to protect against the hazards of the area as well as the limited combustible loading, it is considered that this penetration seal offers equivalent protection to the surrounding hazards and therefore the deviation from the detail is negligible and considered acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-13 Penetration Seal Analysis for Penetration 2083-03-0043 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and detection to protect against the hazards of th e area, as well as limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-28 Penetration Seal Analysis for Penetration 2074-01-0001 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  The installed configuration significantly differs from the approved detail. However, the differences can be justified by the low combustible loading, the available detection and suppression capabilities and by comparative analysis to other test configurations. Based on the hazards in the area and considering the above factors, the installed configuration i s deemed to be capable of maintaining a 3 hour barrier.
 
Engineering Evaluation ID: CALC-89-R-2002-53 Penetration Seal Analysis For Penetration 2058-01-0009 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection on both sides of the barrier, and the response by the fir e brigade to suppress a fire in the incipient stage on either side the differences identified will have no affect on the seal sin ce the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-138 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-78 Penetration Seal Analysis For Penetrations 2107-02-0028 & 2107-05-0015 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on low and moderate fire durations, the smoke detection and fire suppression system on the top side, and the fire brigade's response and are adequate for the hazards.
Engineering Evaluation ID: CALC-89-R-2002-79 Penetration Seal Analysis For Penetration 2106-05-0078 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on low fir e durations, the smoke detection and fire suppression system on the top side, and the fire brigade's response and are adequate for the hazards.
 
Engineering Evaluation ID: CALC-89-R-2002-84 Penetration Seal Analysis For Penetrations 2096-03-0095 & -0096 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Engineering Evaluation ID: CALC-89-R-2002-98 Penetration Seal Analysis For Penetration 2105-01-0001 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-139 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00010 Penetration Seal Analysis for Penetration 2139-01-0041 Summary: Purpose:  Evaluate and document the acceptability of unit 2 penetrations FB-2139-01-0041 to be used in a 3-hour rated fire area boundary based on approved fire tests. Basis for Acceptability:  The fire duration on the bottom side of the seal is less than the tested configuration (top side has a sprinkler system) thus the seal is adequate for the hazard.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00014 Fire Protection Engineering Evaluation for Unit 2 Fireflex Seals Summary: Purpose:  This evaluation is to evaluate ANO-2 penetration seals utilizing Fireflex as a sealing material and document their acceptability to be used in a 3-hour rated fire area boundary based on the hazards in the area. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection systems, and the fire brigade's response and are adequate for the hazards.
Engineering Evaluation ID: CALC-ANO2-FP-09-00028 Engineering Evaluation of Penetration Seals in Fire Area HH Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetration seals in Fire Area HH to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seal is considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, and the response by the fire brigade to suppress the fire in the incipient stage with the manual fire fighting equipment in the area.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-140 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-10-00001 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area KK Summary: Purpose:  This evaluation is to evaluate and document the acceptability of a unit 2 penetration seals in Fire Area KK to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The combustible loading, smoke detection systems (and flame detection system on the top side), suppression system on the top side and the response by the fire brigade (with fire fighting equipment in the area) would prevent a fire from growing past the incipient stage.
Engineering Evaluation ID: CALC-ANO2-FP-10-00002 Engineering Evaluation for Penetration Seals in Fire Area QQ Summary: Purpose:  This evaluation is to evaluate and document the acceptability of an ANO-2 penetration seals in Fire Area QQ to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The combustible loading, smoke detection systems (and flame detection system on the top side), suppression system on the top side and the response by the fire brigade (with fire fighting equipment in the area) would prevent a fire from growing past the incipient stage.
 
Engineering Evaluation ID: CALC-ANOC-FP-07-00003 Water Tight Fire Doors Evaluation Units 1 & 2 86-10 Evaluation Summary: Purpose:  10 CFR 50, Appendix R, requires fire barriers separating redundant trains of safe shutdown equipment to be separated by rated fire barriers that includes all opening, i.e. fire doors, fire dampers, penetration seals, etc. However, th ere are locations in the plant that are susceptible to flooding that are equipped with water tight doors that are not UL Listed or FM Approved to be used in a 3-hour rated barrier. Thus, this evaluation will determine if the water tight fire doors are acceptab le to be used in the 3-hour rated fire barriers at ANO by comparison with UL Listed 3-hour fire doors. The guidance provided in Generic Letter 86-10 will be utilized for the evaluation. Basis for Acceptability:  The water tight doors installed at ANO in 3-hour rated fire barriers have been determined to be acceptable for use based on the hazards in the areas. Although these doors are not 3-hour rated fire doors, they will provide the protection needed in the areas they are used.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-141 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANOC-FP-09-00011 Fire Protection Engineering Evaluation of Units 1 & 2 Aux Bldg Elevator Doors Summary: Purpose:  Evaluate elevator doors as part of the 3-hour fire boundary.
Unit 2: 335' 2040-JJ Fire Area DD 354' 2073-DD Fire Area HH 386' 2136-I Fire Area G 404' 2151-A Fire Area B-2 The elevator doors were previously evaluated in calculation 85-E-0053-04 however the calculation will be superseded by this evaluation as part of the NFPA 805 transition project. Basis for Acceptability:  Based on the low and moderate (Fire Zone 2136-I only) combustible loading, the availability of the smoke detection systems (and suppression system in Fire Zone 67-U) and the availability of the fire brigade with manual fire fighting equipment, the elevator doors are considered to be adequate for the hazards in the area and acceptable for the 3-hour rated fire barriers.
 
Engineering Evaluation ID: CALC-ANOC-FP-09-00012 Engineering Evaluation for Units 1 & 2 Partial Suppression Systems Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains that are protected by a 1-hour fire wrap. Basis for Acceptability:  The areas of concern (1-hour fire wrap) are adequately covered by the suppression systems and thus a fire would not damage both trains and prevent them from performing their designed functions.
 
Engineering Evaluation ID: CALC-ANOC-FP-10-00001 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area G, Part 2 Summary: Purpose:  Evaluate and document the acceptability of penetrations in Fire Area G to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The seals are considered to be adequate for the hazards in the area based on the combustible loading, smoke detection system or line type heat detectors, suppression system or the response by the fire brigade to suppress the fire and prevent significant damage (with fire fighting equipment in the area).
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-142 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2063-DD Sample Room No No No No No No No No No No No No 2072-R Upper Volume Control Tank Room, Lower Tank & Pump Room No Yes No No No No No Yes No Yes No Yes 2073-DD Access Room, Pump Room, Tank Room (2B62 & Resin Addition Room) P Yes No No No No Yes Yes Yes Yes No Yes 2096-M Motor Control Center (2B63) No Yes No No No No No Yes No Yes No Yes 2106-R Degasifier Vacuum Pump Room No Yes No No No No No Yes No Yes No Yes 2107-N Corridor (North of Stairway 2001) No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-143 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-12 Title: ANO-2 Fire Area HH Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. The following  modifications are area specific and credited to reduce risk  in this fire area:  2CV-1026-2 circuit modified to prevent spurious operation. 2CV-1076-2 circuit modified to prevent spurious operation. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area HH was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Taking credit for suppression in Fire Zone 2073-DD (CALC-PRA-A2-05-010), Fire Area HH has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights"
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-144 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-145 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: HH-01  VFDR: Fire damage to control and power cables in the area may impact EFW functions resulting in the following:
a) Spurious closure of EFW valve 2CV-1036-2 (IN 92-18) which feeds the credited SG B. b) Loss of valve 2CV-1060-2 isolation capabilities to the credited SG B. c) Loss of DC control to valves 2CV-0795-2 and 2CV-0711-2 (IN 92-18) may result in isolation of SW flow to EFW 2P-7B pump from SW Loop 2 for long-term source of feedwater upon depletion of condensate. d) Loss of isolation capability for MSIV 2CV-1010-1, MSIV Bypass 2CV-1090-2 (IN 92-18) and Atmospheric Dump valve 2CV-1001 and associated atmospheric dump isolation valve 2CV-1002 (IN 92-18) to SG A.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-1036-2. b) No further action is required for 2CV-1060-2. c) No further action is required for 2CV-0795-2 and 2CV-0711-2. d) No further action is required for 2CV-1010-1, 2CV-1090-2, 2CV-1001 and 2CV-1002.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-146 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: HH-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Spurious trip of the Charging Pumps 2P-36A from a SIAS signal and loss of Control Room start function. b) Spurious closure of valve 2CV-4840-2 could result in loss of charging flow path requiring manual realignment through manual valve 2CVC-115 and 2CV-5103-1 using the HPSI header. c) Closure of VCT discharge valve 2CV-4873-1 could result in gas binding. Gas binding requires the running charging pump to be secured until BAMT or RWT is aligned for suction source d) Loss of open capability for valves 2CV-4920-1 and 4921-1 resulting in isolation of BAMT supply to charging pumps. e) Loss of DC power to Pressurizer Heater banks #1 and #2 preventing trip capability from the Control Room. f) Failure or spurious operation of 2CV-4950-2 could result in loss of charging pump suction from the RWT after transfer from the BAMTs.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required as 2P-36A is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. b) No further action is required as 2CV-4840-2 is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. c) No further action is required as 2CV-4873-1 is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. d) No further action is required as 2CV-4920-1 and 4921-1 are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. e) Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016. f) No further action is required as 2CV-4950-2 is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-147 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: HH-03 VFDR: Fire damage to cables in the area may impact SW functions. SW is the long-term source of feedwater to the SGs via its connecti on to EFW system once condensate is depleted. Offsite power is avai lable eliminati ng the immedi ate need for SW c ooling of the EDG.
a) Spurious closure of SW to "B" ESF header valve 2CV-1406-2 (IN 92-18) will result in a loss of SW loop 2. b) Spurious closure of 2CV-1480-2 (IN 92-18) will result in loss of the SW return path for loop 2 to Lake Dardanelle. c) Spurious opening of SW to SFP heat exchanger valve 2CV-1526-2 will prevent isolation of credited SW Loop 2 from SW Loop 1 resulting in potential pump run-out conditions. d) Loss of power and control to lake sluice gate 2CV-1472-5 if 2P-4B is the operable pump. Spurious operation of this sluice gate could result in a loss of SW loop 2. e) Loss of power and control to lake sluice gate 2CV-1474-2 if 2P-4C is the operable pump. Spurious operation of this sluice gate could result in a loss of SW loop 2.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-1406-1. b) No further action is required for 2CV-1480-2. c) No further action is required for 2CV-1526-2. d) No further action is required for 2CV-1472-5. e) No further action is required for 2CV-1474-2.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-148 Fire Area ID: HH - Unit 2 General Area 354' Elevation Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: HH-04 (CR-ANO-C-2006-00048-033)
VFDR: Fire damage to cables in the area could result in the loss of Control Room Ventilation.
a) Loss of power and control for Control Room HVAC condensing unit 2VE-1B may result in room heat-up that could challenge equipment and Control Room habitability.
Loss of this function could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: The evaluation performed by CALC-10-E-0010-05 "ANO-2 Auxiliary Building Integrated Room Heat Up Model to Support NFPA-805 Analysis" has demonstrated that Control Room HVAC is not necessary to support post fire shutdown.
End of Fire Area HH
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-149 Fire Area ID: II - North Switchgear Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2101-AA North Switchgear Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36B or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-4 aligned to offsite power. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 25 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.16
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-150 Fire Area ID: II - North Switchgear Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to manual methods. The primary entrance to this area for fire brigade members is from the north door which leads to the large open area of the turbine building. Opening this door minimizes any ponding concerns. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions
 
Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations
 
Engineering Evaluation ID: No engineering evaluations are applicable to this fire area. Summary: N/A  Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2101-AA North Switchgear Room No Yes No No No No No No No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-151 Fire Area ID: II - North Switchgear Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-13 Title: ANO-2 Fire Area II Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area II was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area II has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE. No procedural changes or modifications are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-152 Fire Area ID: II - North Switchgear Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: II-01  VFDR: Fire damage to control cables for 2A-4 to 2A-3 tie breaker could close the breaker and affect alignment of ESF bus 2A-4.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: The safe shutdown equipment details for 2A-410 in Fire Area II indicates that only cable 2C436M can cause spurious closure. Upon further review of the route path for this cable it has been determined that cable 2C436M leaves tray DC267 prior to entering Fi re Area II. Cable 2C436M is therefore free of fire damage in Fire Area II and 2A-410 will not spuriously close. The action associated with 2A-410 to de-energize control power & open the breaker is not necessary and this will be reflected in the next revision of the Safe Shutdown calculation.
VFDR ID: II-02  VFDR: Fire damage to control cables for 2P-32A (breaker 2H-11), 2P-32C (breaker 2H-22) and 2P-32D (breaker 2H-12) can result in a spurious re-start of 2P-32A, 2P-32C and 2P-32D, respectively. Securing the pumps assures normal pressurizer spray is secured a nd prevents potential RCP seal damage that could result in a seal LOCA.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Additional circuit analysis on the RCP pump motors concluded that fire damage to the RCP cables would not prevent the Control Room from tripping the pumps for a fire in Fire Area II. Since the Control Room will retain the ability to trip the RCPs after fire in Fire Area II, this is not a VFDR, and the associated basic events have been excluded from the model in the Post-Transition Baseline Case.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-153 Fire Area ID: II - North Switchgear Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: II-03  VFDR: Fire damage to power and control cables for SW crossover 2P-4B to 2P4A valves 2CV-1418-1 (IN 92-18) and 2CV-1419-1 (IN 92-18) could cause spurious operation and potential damage resulting in diversion of flow from Loop 2. Impact will cause pump run-out of either 2P-4B(G) or 2P-4C depending upon availability. These valves are required for operation of SW.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required for components associated with SW (vital auxiliaries).
VFDR ID: II-04  VFDR: Fire damage to control cables for the Pressurizer Heater Banks 1 and 2 results in the loss of remote Control Room trip capability needed to secure the heaters post fire.
Loss of this function could challenge the Pressure Control Performance Criterion. This condition represents a variance from th e deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
VFDR ID: II-05  VFDR: Fire damage to cables in the area may impact inventory control functions resulting in the following:
a) Loss of power source for valves 2CV-4920-1 and 4921-1 preventing alignment of BAMT supply to charging pumps.
Loss of these functions could challenge the Reactivity Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: No further action is required as 2CV-4920-1 and 4921-1 are components in the CVCS system. CVCS is only credited in the Fire PR A to mitigate non-fire induced ATWS scenarios.
End of Fire Area II Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-154 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2109-U Corridor
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A or 2P-36C available with gravity feed from BAMTs, normal injection path isolated, and manual alignment to RCS using HPSI header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-B from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 aligned to offsite power. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC impacted in this area. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 6. Process Monitoring Instrumentation is available in the Control Room via SPDS to monitor pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 26 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.17
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-155 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area is equipped with an automatic suppression system. Ponding in the fire area is minimal and the presence of drains, a scupper, sill plates, and curb plates prevents migration of fire suppression water to other areas. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Licensing Actions Licensing Action: Appendix R, Exemption 07, FA - JJ, Lack of Barrier/ Separation III.G.2 Criteria, NRC approval letter 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-05 Penetration Seal Analysis for Penetration 2104-08-0133 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-06 Penetration Seal Analysis for Penetration 2098-04-0068 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  The qualification of multiple conduits through a silicone foam blockout is contained in Fire Test Report SwRI 03-6004-006 (CTP-1001A) where test blockout number six demonstrated the ability of multiple conduits to successfully pass the ASTM-E119 fire endurance test. This evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-156 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-07 Penetration Seal Analysis for Penetration 2098-04-0066 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  This configuration has differences between the tested detail and the as-b uilt penetration in that the re is one additional cable tray in the as-built. The qualification of multiple cable trays through a silicone foam blockout is contained in the Fire Test Report SwRI 03-6004-006 (CTP-1001A) where test blockout number one qualifies multiple cable trays through a silicone foam blockout. Overall this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Engineering Evaluation ID: CALC-89-R-2002-106 Penetration Seal Analysis For Penetration 2104-07-0121 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low and moderate fire durations, available smoke detection system, fire suppression on one side and the fire brigade's response are adequate for the hazards. The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains fewer metallic penetrants.
This is advantageous since it would have less heat conductive exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-107 Penetration Seal Analysis for Penetration 2104-02-0091 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of the availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-12 Penetration Seal Analysis for Penetration 2104-02-0101 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-157 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-123 Penetration Seal Analysis For Penetration 2109-01-0064 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low and moderate fire durations, available smoke detection system, suppression system on one side and the fire brigade's response are adequate for the hazards. The differences identified will have no effect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains ewer metallic penetrants. This is advantageous since it would have less heat conductive exposure area to the fire.
Engineering Evaluation ID: CALC-89-R-2002-128 FB-2104-02 GAP EVAL Summary: Purpose:  Evaluate the adequacy of fire barrier FB-2104-02 as a component of a three (3)-hour rated fire boundary. Basis for Acceptability:  This barrier is considered adequate based on the fire duration on both sides, smoke detection systems on both sides (and line type heat detection system and suppression systems on the corridor side) and the response by the fire brigade to suppress a fire in the incipient stage on either side.
 
Engineering Evaluation ID: CALC-89-R-2002-24 Penetration Seal Analysis For Penetration 2108-02-0007 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection systems on both sides of the barrier, (and line type heat detection system and suppression systems on the corridor side) and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is an equivalent in heat transmission.
Engineering Evaluation ID: CALC-89-R-2002-98 Penetration Seal Analysis For Penetration 2105-01-0001 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-158 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-99 Penetration Seal Analysis For Penetration 2109-01-0032 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on the smoke detection and fire suppression systems, and the differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is lower in heat transmission.
 
Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
 
Engineering Evaluation ID: CALC-ANO2-FP-06-00004 Penetration Seal FB-2104-07-0007 86-10 Evaluation Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is smaller. A smaller seal is advantageous since it would have less exposure area to the fire.
Engineering Evaluation ID: CALC-ANO2-FP-09-00009 Unit 2 Structural Steel FP Fire Protection Engineering Evaluation Summary: Purpose:  This fire protection engineering evaluation is to evaluate and document the lack of structural steel fire proofing in rooms noted in this evaluation in a 3 -hour rated fire area boundary. Basis for Acceptability:  The rooms are protected by smoke detection systems that alarm in the Control Room (and suppression systems in the electrical penetration rooms) and the prompt response by the fire brigade with access to manual fire fighting equipment would prevent any fire (in the unlikely event one does occur) from damaging the structural steel.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-159 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00025 Fire Protection Engineering Evaluation of Penetration Seals in Fire Area EE Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area EE to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on factors such as combustible loading, smoke detection systems, suppression systems and the response by the fire brigade to suppress the fire in the incipient stage with the manual fire fighting equipment in the area.
Engineering Evaluation ID: CALC-ANO2-FP-09-00029 Fire Protection Engineering Evaluation for Penetration Seals in U-2 Fire Area JJ Summary: Purpose:  This evaluation is to evaluate and document the acceptability of ANO-2 penetration seals in Fire Area JJ to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the fire duration, smoke detection systems on both sides of the barrier (and line type heat detection system and suppression systems on the near side) and the response by the fire brigade to suppress a fire in the incipient stage on either side.
Engineering Evaluation ID: CALC-ANO2-FP-09-00030 Fire Protection Engineering Evaluation for Penetrations Seals in Fire Area TT Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area TT to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetration seals are considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, suppression systems (far sides only) and the response by the fire brigade to suppress the fire in the incipient stage.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2109-U Corridor P Yes No No No No Yes Yes Yes Yes Yes Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-160 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-14 Title: ANO-2 Fire Area JJ Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions 2A-113 switchgear breaker for offsite power  2A-309 ESF switchgear breaker for offsite power  2CV-1036-2 EFW discharge valve  2CV-1075-1 EFW discharge valve  2CV-5649-1 containment sump isolation valve  2CV-5650-2 containment sump isolation valve  2PIS-0789 EFW condensate suction  2P-7B EFW pump  2P-89 HPSI pump suction Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. The following modifications are area specific and credited to reduce risk in this fire area:  2A3 switchgear DC control power  2A-308 EDG supply breaker  2A-309 Offsite supply breaker  2A-310 crosstie breaker  2B-6 load center  2P-32A, 2P-32B, 2P-32C, and 2P-32D  DC control power  2CV-1036-2 circuit modification to resolve IN 92-18 issue  2CV-1075-1 circuit modification to resolve IN 92-18 issue  2CV-4816 and 2CV4817 Letdown valve control circuit
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-161 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation
 
Risk Summary (continued)
 
Summary (continued) IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. The modifications to 2CV-1036-2 and 2CV-1075-1 will eliminate t he IN 92-18 issue for this fire area. Additional Fire Area Considerations The detection system located in Fire Area JJI was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area JJ has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence. Partial Suppression was credited in this zone for core damage mitigation of specific scenarios (CALC-PRA-A2-05-010).
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. The available suppression will provide additional assurances that any unknown weaknesses or uncertainties will not adversely affect the ability to meet nuclear safety performance criteria. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-162 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: JJ-01  VFDR: Fire damage to control and power cables in the area may impact EFW functions resulting in the following:
a) Loss of valves 2CV-1036-2 and 2CV-1075-1 (IN 92-18) which feed the credited SG B. b) Loss of valve 2CV-1038-2 which may result in flow diversion to SG A without instrumentation and ADV control capability. c) Loss of DC power to the engineered safeguards bus 2A-3 feeding EFW pump 2P-7B. d) Loss of MSIV 2CV-1060-2 isolation capability of the credited SG B. e) Spurious start of the non-credited EFW pump 2P-7A and loss of Control Room trip function. f) Loss of MSIV 2CV-1010-1 isolation capability of the non-credited SG A.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) Recovery action and modification associated with 2CV-1036-2 and 2CV-1075-1 (see Attachment S). b) No further actions are required for 2CV-1038-2. c) Recovery action associated with 2P-7B (recovery will not be required after  DC power and controls are modified). d) No further actions are required for 2CV-1060-2. e) No further actions are required for 2P-7A. f) No further actions are required for 2CV-1010-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-163 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: JJ-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Spurious opening of valve 2CV-5631-2 could result in RWT drain-down to containment sump. b) Spurious trip of the charging pumps 2P-36A and C(R) from a SIAS signal. c) Spurious start of the HPSI pumps 2P-89A, B and C(R) from a SIAS signal. d) Spurious Containment Spray Pumps 2P-35A and B start from a containment spray signal could result in a flow diversion and a loss of DC control power could result in a loss of Control Room trip capability. e) Loss of DC control circuit power to Pressurizer Heater banks #1 through #6 resulting in loss of trip capability from the Control Room. f) Multiple spurious operations of Pressurizer LTOP relief valves 2CV-4731-2 and 2CV-4740-2 may result in an RCS inventory loss path. g) Spurious opening of 2CV-5086-2 and 2CV-5084-1 (SDC RCS isolation) due to impact on power cables for theses high-low pressure interface valves may result in a loss of RCS inventory.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) Failure of 2CV-5631-2 is not a contributor to core damage sequences in the fire PRA and therefore is not risk significant. b) No further action required as 2P-36A and 2P-36C are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. c) Recovery action associated with 2P-89A, B and C(R). d) Failure of 2P-35A or 2P-35B is not a contributor to core damage sequences in the fire PRA and therefore is not risk significant. e) Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016. f) No further actions are required for 2CV-4731-2 and 2CV-4740-2. g) No further actions are required for 2CV-5086-2 and 2CV-5084-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-164 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: JJ-03  VFDR: Fire damage to DC power cables could result in a loss of the Control Room trip capability of the RCPs 2P-32A through D. Securi ng the pumps assures normal pressurizer spray is secured and prevents potential RCP seal damage.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with modification to install a secondary DC power supply directly to the RCP trip breakers at switchg ear 2H-1 and 2H-2 (see Attachment S).
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-165 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: JJ-04  VFDR: Fire damage to cables in the area could result in the following impacts on vital auxiliaries:
a) Loss of DC control power to Switchgear 2A-3, also resulting in loss of SW Pump 2P-4A and 2P-4B(R) required for extended EFW supply upon depletion of CST source. b) Spurious trip of the off-site supply breaker 2A-309 to credited bus 2A-3 c) Spurious start of the non-credited EDG and closure of output breaker 2A-308 connecting to bus 2A-3. d) Normal power supply to the battery charger 2D-32B. Alternate power supply cables are not impacted for a fire in this area but require manual transfer. e) Spurious operation of SW crosstie valves 2CV-1421-2 and 2CV-1422-2 resulting in flow diversion of SW required for extended EFW supply upon depletion of condensate storage tank (CST) source. f) Loss of off-site power to 2A-1 from Startup Transformer #3 due to failure of automatic power transfer from Unit Auxiliary transformer. Manual realignment is required. Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions (see Attachment S):
a) Modification to maintain DC control power to Switchgear 2A-3. b) Recovery action and modification associated with 2A-309. c) Modification associated with 2A-308. d) No further actions are required for 2D-32B. e) No further actions required for 2CV-1421-2 or 2CV-1422-2 (cables exit tray EB208 prior to entering Fire Area JJ). f) Recovery action and modification to maintain DC control power to 2A-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-166 Fire Area ID: JJ - Corridor Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: JJ-05 (CR-ANO-C-2006-00048-033)
VFDR: Fire damage to cables could result in the loss of Control Room Ventilation. Room heat-up could challenge equipment and Control Room habitability.
Loss of this function could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: The evaluation performed by CALC-10-E-0010-05 "ANO-2 Auxiliary Building Integrated Room Heat Up Model to Support NFPA-805 Analysis" has demonstrated that Control Room HVAC is not necessary to support post fire shutdown.
 
VFDR ID: JJ-06 (CR-ANO-C-2006-00048-007 and 050)
VFDR: Fire damage to cables in the area could result in loss of required instrumentation:
a) Battery feed to 2D02 DC Load Center results in loss of all DC power. b) Fault causing a blown fuse feeding 2D32B concurrent with a loss of power supply to 2D32A results in the loss of battery charging capability.
Loss of this function could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further actions required for 2D02. b) No further actions required for 2D32B and 2D32A.
End of Fire Area JJ
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-167 Fire Area ID: K - Tank Vaults Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 16-Y Clean Waste Receiver Tank Room 2020-JJ Boron Holdup Tank Vault Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 27 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.18 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. This fire area is a tank vault with a low combustible loading, has no automatic suppression system, and firefighting activities are limited to manual methods. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-168 Fire Area ID: K - Tank Vaults Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations
 
Engineering Evaluation ID: CALC-89-R-2002-21 Penetration Seal Analysis for Penetration 2017-05-0063 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and detection to protect against the hazards of th e area, as well as the limited combustible loading, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-43 Penetration Seal Analysis for Penetration 2017-05-0077 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Since the referenced test substantiates the ability of a large diameter mechanical penetrant through a sleeve with a zero annulus and sealed with 10" of silicone foam to successfully withstand a 38-minute fire endurance test and the fact that the referenced penetration is located in a fire zone with a fire duration of less than thirty (38) minutes this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-45 Penetration Seal Analysis for Penetration 2016-05-0002 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  The seals utilized at ANO that consist of sleeves with pipes utilizing an annulus that is three (3) inches or less, ten (10) inches of silicone foam with one (1) inch of damming material and located in a barrier with fire duration of less than forty-three (43) minutes are considered acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-169 Fire Area ID: K - Tank Vaults Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 16-Y Clean Waste Receiver Tank Room No No No No No No No No No No No No 2020-JJ Boron Holdup Tank Vault No No No No No No No Yes No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
 
VFDRs This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area K
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-170 Fire Area ID: KK - South EDG Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2093-P South EDG Room 2114-I EDG Air Intake Room 2115-I Boric Acid Makeup Tank Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 28 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.19
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-171 Fire Area ID: KK - South EDG Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. Floor drains in the area are adequate to minimize ponding from suppression activities (automatic and manual) well below the height of the 12" curb plates that prevent a flow of water outside of the fire-affected area. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: Appendix R, Exemption 13, FA - KK (Originally FA - B), Not Meeti ng III.G.2 Criteria, NRC appr oval letter 2CNA108802 dated 10/26/1988. Licensing Basis: This exemption is no longer required because a fire in this zone has been shown to not result in a loss of offsite power.
Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-121 Penetration Seal Analysis For Penetration 2150-01-0020 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based low fire duration, available smoke detection system, and the fire brigade's response are adequate for the hazards. The differences identified will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration contains equivalent areas.
 
Engineering Evaluation ID: CALC-89-R-2002-60 Penetration Seal Analysis for Penetration 2147-01-0017 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and automatic fire detection to protect against th e hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-172 Fire Area ID: KK - South EDG Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-78 Penetration Seal Analysis For Penetrations 2107-02-0028 & 2107-05-0015 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on low and moderate fire durations, the smoke detection and fire suppression system on the top side, and the fire brigade's response and are adequate for the hazards.
Engineering Evaluation ID: CALC-ANO2-FP-10-00001 Fire Protection Engineering Evaluation for Penetration Seals in Fire Area KK Summary: Purpose:  This evaluation is to evaluate and document the acceptability of a unit 2 penetration seals in Fire Area KK to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The combustible loading, smoke detection systems (and flame detection system on the top side), suppression system on the top side and the response by the fire brigade (with firefighting equipment in the area) would prevent a fire from growing past the incipient stage.
 
Engineering Evaluation ID: CALC-ANOC-FP-09-00001 Ventilation Opening in Units 1 & 2 EDG Rooms Summary: Purpose:  This evaluation is to evaluate the lack of fire dampers in the ventilation openings for the EDG Rooms. Basis for Acceptability:  The smoke and flame detection systems and suppression system in the EDG rooms would detect and suppress a fire in the incipient stage and prevent its growth. The smoke detection system in the exhaust fan rooms would also detect a fire in the incipient stage and alert Operations personnel to the fire. Manual firefighting equipment is located in rooms adjacent to the EDG room and exhaust rooms. The configuration of the exhaust fan rooms would also prevent fire to propagate to the other exhaust fan room or EDG rooms.
Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-173 Fire Area ID: KK - South EDG Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2093-P South EDG Room P Yes No No No No Yes Yes Yes Yes No Yes 2114-I EDG Air Intake Room No Yes No No No No No Yes No Yes No Yes 2115-I Boric Acid Makeup Tank Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-174 Fire Area ID: KK - South EDG Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-15 Title: ANO-2 Fire Area KK Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. No recovery actions are credited for this fire area. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area KK was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Taking credit for suppression in Fire Zone 2093-P (CALC-PRA-A2-05-010), Fire Area KK has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-175 Fire Area ID: KK - South EDG Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: KK-01 VFDR: Fire damage associated with an auxiliary relay external to the motor control circuit results in spurious trip of charging pumps 2P-36A, B, and C and prevents operation from the Control Room.
Loss of this function could challenge the Inventory Control Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: No further action is required as 2P-36A; B and C are components in the CVCS system. Failure of the CVCS system is not a contributor to core damage sequences in the fire PRA and therefore is not risk significant.
 
VFDR ID: KK-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Loss of power and control for valves 2CV-4920-1 and 4921-1 preventing alignment of BAMT supply to charging pumps.
Loss of these functions could challenge the Inventory and Reactivity Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: No further action is required as 2CV-4920-1 and 4921-1 are components in the CVCS system. CVCS is only credited in the Fire PR A to mitigate non-fire induced ATWS scenarios.
End of Fire Area KK
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-176 Fire Area ID: L - Diesel Fuel Storage Vault Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description TKVLT Diesel Fuel Storage Vault
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as a backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 29 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.20 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of this fire area. Fire suppression activities would only diesel fuel storage and transfer. Plant equipment in other areas is isolated from effect of fire in this fire area. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-177 Fire Area ID: L - Diesel Fuel Storage Vault Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: No engineering evaluations are applicable to this fire area. Summary: N/A Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET TKVLT Diesel Fuel Storage Vault Yes Yes No No No No No No No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area L Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-178 Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2099-W West D.C. Equipment Room 2103-V West Battery Room Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36B or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A feeding SG-B from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-4 aligned to offsite power. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4C or 2P-4B feeding SW header 2. 5c. Vital Auxiliaries (HVAC) Green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 30 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.21
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-179 Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station from Fire Area JJ. Any water from manual suppression is expected to migrate out the open doorway where the presence of drains and a scupper will be adequate to accommodate the flow. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criter ia.
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-128 FB-2104-02 GAP EVAL Summary: Purpose:  Evaluate the adequacy of fire barrier FB-2104-02 as a component of a three (3)-hour rated fire boundary. Basis for Acceptability:  This barrier is considered adequate based on the fire duration on both sides of the barrier, smoke detection systems on both sides of the barrier (and line type heat detection system and suppression systems on the corridor side) and the response by the fire brigade to suppress a fire in the incipient stage on either side.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2099-W West D.C. Equipment Room No Yes No No No No No Yes No Yes No Yes 2103-V West Battery Room No Yes No No No No No No No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-180 Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-16 Title: ANO-2 Fire Area MM Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions The following equipment is recovered in the post transition baseline case:  2A-113 offsite power breaker  2A-213 offsite power breaker  2CV-1025-1 EFW discharge valve  2CV-1036-2 EFW discharge valve  2CV-1075-1 EFW discharge valve  2CV-1038-2 EFW discharge valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area MM was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area MM has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-181 Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-182 Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: MM-01  VFDR: Fire damage to cables in the area may impact EFW functions resulting in the following:
a) Loss of DC control cables to valves 2CV-0205-2 (IN 92-18) and 2CV-0340-2 results in loss of steam flow to the EFW 2P-7A pump feeding the credited SG B. b) Loss of DC control cables to valve 2-CV-0711-2 (IN 92-18) and 2CV-0795-2 may result in isolation of SW and CST supply to EFW 2P-7A pump. c) Loss of DC control cable to valve 2CV-1076-2 results in isolation of the EFW flow to SG B. d) Loss of power and control cables to valve 2CV-1026-2 results in loss of control of EFW flow to the non-credited SG A. e) Spurious closure of valve 2CV-1039-1 results in isolation of the EFW flow to SG B.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2CV-0205-2 and 2CV-0340-2. b) No further action is required for 2-CV-0711-2 and 2CV-0795-2. c) No further action is required for 2CV-1076-2. d) No further action is required for 2CV-1026-2. e) No further action is required for 2CV-1039-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-183 Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: MM-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Spurious Containment Spray Pumps 2P-35A start from a containment spray signal could result in a flow diversion and a loss of DC control power could result in a loss of Control Room trip capability.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: No further action is required as spurious operation of the containment spray pump is not a contributor to core damage sequences in the fire PRA, and therefore is not risk significant.
 
VFDR ID: MM-03  VFDR Loss of DC control circuit power to Pressurizer Heater banks #1, #3, and #5 resulting in the loss of remote Control Room trip capability needed to secure the heaters post fire.
Loss of this function could challenge the Pressure Control Performance Criterion. This condition represents a variance from th e deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
 
VFDR ID: MM-04  VFDR: Fire damage to DC power cables could result in a loss of the Control Room trip capability of the RCPs 2P-32A and D. Securing t he pumps is required to assure normal pressurizer spray is secured and prevent potential RCP seal damage.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-184 of NFPA 805 Section 4.2.4 with modification of the DC control power circuit to switchgear 2H-1 and 2H-2 (see Attachment S).
Fire Area ID: MM - West Battery and DC Equipment Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: MM-05  VFDR: Fire damage to cables in the area could result in a loss of DC control power to Switchgear 2A-3 resulting in a loss of power to load center 2B-5 and motor control center 2B-51.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with no further action required.
Open items Item ID: MM-HVAC
 
== Description:==
Analysis is required to determine post fire room heat-up in locations requiring operator manual actions given a loss of HVAC. The result of this analysis will be used to evaluate availability of equipment required for post fire shutdown and habitability for feasibility of operator manual actions based on room temperature.
Disposition: The analysis and evaluation for post fire room heat-up is performed in EC-35075. In Fire Area MM, excessive temperatures have been identified in Fire Zone 2099-W following a loss of ventilation. A modification is proposed to modify Fire Door DR-265 to allow normally open positioning of the door with automatic closure features in the event of a fire on either side of the barrier. Th e modification will open Fire Area MM to Fire Area JJ allowing natural circulation to reduce post fire temperatures to an acceptable level (see Attachment S).
Status: Closed  Corrective Action Ref.: CR-ANO-C-2006-0048 CA-55 End of Fire Area MM
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-185 Fire Area ID: NN - Unit 2 Containment Building Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2032-K Containment Building, South Side 2033-K Containment Building, North Side Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using atmospheric dump valves. Dependant upon fire location EFW pumps 2P-7A or 2P-7B feeding available SG-A (North) or SG-B (South) from condensate storage with SW as backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
 
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 31, 32 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachments 8.35, 8.36, 8.39
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-186 Fire Area ID: NN - Unit 2 Containment Building Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. The physical configuration of the containment building prevents water from migrating to other areas. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: Appendix R, Exemption 04, FA - NN, Not Meeting III.G.2 Criteria, NRC approval letter 0C NA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Licensing Action: Appendix R, Exemption 17, FA - NN, RCP Oil Collection, Not Meeting III.O Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. Licensing Basis: This licensing action will be transitioned into the NFPA 805 fire protection program as previously approved and considered compliant under 10 CFR 50.48(c). Refer to LAR Attachment K for detailed discussion of this licensing action.
Licensing Action: Appendix R, Exemption 19, FA - NN, RCP Oil Fill Line, Not Meeting III.O Criteria, NRC approv al letter 2CNA069701 dated 6/14/1997. Licensing Basis: This licensing action will be transitioned into the NFPA 805 fire protection program as previously approved and considered compliant under 10 CFR 50.48(c). Refer to LAR Attachment K for detailed discussion of this licensing action.
Engineering Evaluations Engineering Evaluation ID: CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary: Purpose:  This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Buildings penetrations to be used in a 3 hour rated fire area boundary. Basis for Acceptability:  The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual fire fighting equipment for those areas in the unit's Auxiliary Buildings.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-187 Fire Area ID: NN - Unit 2 Containment Building Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2032-K Containment Building, South Side P P No No No No Yes Yes No Yes No Yes 2033-K Containment Building, North Side P P No No No No Yes Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-188 Fire Area ID: NN - Unit 2 Containment Building Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-17 Title: ANO-2 Fire Area NN Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs. Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area NN was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area NN has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-189 Fire Area ID: NN - Unit 2 Containment Building Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: NN-01  VFDR: Fire damage to cables in the area could result in a potential to spurious open both high-low pressure interface valves 2CV-5086
-2 (IN 92-18) and 2CV-5084-1(IN 92-18) due to fire-induced 3 phase fault to the power cables. Power to both valves is de-energize d during normal high-pressure operations.
 
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: MOVs 2CV-5084-1 and 2CV-5086-2 are high pressure to low pressure interfaces, and their concurrent spurious opening could have severe consequences. However, both valves are normally de-energized, and the power cables for 2CV-5086-2 are divided such that each phase is in a separate cable. Thus, concurrent damage to at least four cables resulting in energizing a total of six cond uctors in the proper phase sequence is required. This is judged to be incredible. Per NUREG/CR-6850, the probability of a three phase AC hot short is 5.00E-08 for thermo-set cables (applicable to ANO-2). Therefore, no further evaluation of this VFDR is necessary in the FPRA. VFDR ID: NN-02  VFDR: Fire damage to control cables for 2P-32A (breaker 2H-11), 2P-32B (breaker 2H-21), 2P-32C (breaker 2H-22) and 2P-32D (breaker 2H-12) can result in a spurious re-start of 2P-32A, 2P-32B, 2P-32C and 2P-32D, respectively. Securing the pumps assures normal pressurizer spray is secured and prevents potential RCP seal damage.
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Additional circuit analysis on the RCP pump motors concluded that fire damage to the RCP cables would not prevent the Control Room from tripping the pumps for a fire in Fire Area NN. Since the Control Room will retain the ability to trip the RCPs after fire in Fire Area NN, this is not a VFDR, and the associated basic events have been excluded from the model in the Post-Transition Baseline Case. End of Fire Area NN
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-190 Fire Area ID: OO - Unit 2 Intake Structure Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description INTAKE Unit 2 Intake Structure
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A and 2P-4C available to feed SW headers 1 and 2. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 33 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachments 8.22
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-191 Fire Area ID: OO - Unit 2 Intake Structure Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
The ANO-2 intake contains all SW pumps, sluice gates, and valves for both trains of equipment. All other equipment and cables utilized to maintain safe and stable are outside of the area of fire suppression activity. The physical configuration of the structure and locations of equipment prevents a credible fire from affecting both trains and therefore suppression activities only directed at fire impacted equipment. The open grating to the SW bays prevents any ponding that could be caused by either automatic or manual suppression methods. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: Appendix R, Exemption 01, FA - OO, Not Meeting III.G.2 Criteria, NRC approval letter 2CNA109902 dated 10/1/1999. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Licensing Action: Appendix R, Exemption 02, FA - OO, Not Meeting III.G.2 Criteria, NRC app roval letter 0CNA038328 dated 3/22/83. Licensing Basis: This exemption is no longer required because the FRE has found that the fire area is compliant with NFPA 805 Section 4.2.4.
 
Engineering Evaluations Engineering Evaluation ID: CALC-A-FP-2005-001 Fire Protection Appendix R Suppression and Detection Partial 86-10 Evaluation Summary: Purpose:  The purpose of this fire protection engineering evaluation is to evaluate and document the partial suppression systems to protect redundant trains of equipment. Basis for Acceptability:  This evaluation has determined that the installed fire protection features will promptly detect any f ire in its incipient stages and the fire extinguished to limit any damage to one train of equipment.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-192 Fire Area ID: OO - Unit 2 Intake Structure Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET Intake Unit 2 Intake Structure P Yes No No No No No No No No No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-193 Fire Area ID: OO - Unit 2 Intake Structure Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-18 Title: ANO-2 Fire Area OO Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2CV-1470-1 sluice gate valve for 2P-4A  2CV-1474-2 sluice gate valve for 2P-4C Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area OO was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area OO has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights"
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-194 Fire Area ID: OO - Unit 2 Intake Structure Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. The available detection will provide additional assurances that any unknown weaknesses or uncertainties will not adversely affect the ability to meet nuclear safety performance criteria. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-195 Fire Area ID: OO - Unit 2 Intake Structure Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs  VFDR ID: OO-01  VFDR: Fire damage to power and control cables in the area could result in the following impacts on vital auxiliaries:
a) Loss of power to all three SW pumps (2P-4A, swing pump 2P-4B(R) & (G), and 2P-4C). SW is the long-term source of feedwater to the SGs via its connection to EFW system once condensate is depleted. Offsite power is available eliminating the immediate need for SW cooling of the EDG. Currently an electrical raceway fire barrier system (wrap) is credited for both 2P-4A and 2P-4C and is identified for potential removal based upon the FRE.
b) Loss of power and control to lake sluice gate 2CV-1470-1 if 2P-4A is the operable pump. Spurious operation of this sluice gate could result in a loss of a SW source to 2P-4A.
c) Loss of power and control to lake sluice gate 2CV-1474-2 if 2P-4C is the operable pump. Spurious operation of this sluice gate could result in a loss of a SW source to 2P-4C.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further actions are required for 2P-4A, 2P-4B or 2P-4C. b) Recovery action associated with 2CV-1470-1. c) Recovery action associated with 2CV-1474-2.
End of Fire Area OO
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-196 Fire Area ID: QQ - North Emergency Diesel Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description 2094-Q North EDG Room 2114-I EDG Air Intake Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2. 5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 34 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.23 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. Floor drains in the area are adequate to minimize ponding from suppression activities (automatic and manual) well below the height of the 12" curb plates that prevent a flow of water outside of the fire-affected area. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-197 Fire Area ID: QQ - North Emergency Diesel Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: Appendix R, Exemption 13A, FA - QQ (Originally FA - B), Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. Licensing Basis: This exemption is no longer required because a fire in this zone has been shown to not result in a loss of offsite power.
Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-79 Penetration Seal Analysis For Penetration 2106-05-0078 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration.
Basis for Acceptability:  The installed configurations are acceptable to be used in a 3-hour rate fire barrier based on low fir e durations, the smoke detection and fire suppression system on the top side, and the fire brigade's response and are adequate for the hazards.
 
Engineering Evaluation ID: CALC-ANO2-FP-10-00002, Engineering Evaluation for Penetration Seals in Fire Area QQ Summary: Purpose:  This evaluation is to evaluate and document the acceptability of a ANO-2 penetration seals in Fire Area QQ to be used in a 3-hour rated fire area boundary.
Basis for Acceptability:  The combustible loading, smoke detection systems (and flame detection system on the top side), suppression system on the top side and the response by the fire brigade (with firefighting equipment in the area) would prevent a fire from growing past the incipient stage.
Engineering Evaluation ID: CALC-ANOC-FP-09-00001 Ventilation Opening in Units 1 & 2 EDG Rooms Summary: Purpose:  This evaluation is to evaluate the lack of fire dampers in the ventilation openings for the EDG Rooms.
 
Basis for Acceptability:  The smoke and flame detection systems and suppression system in the EDG rooms would detect and suppress a fire in the incipient stage and prevent its growth. The smoke detection system in the exhaust fan rooms would also detect a fire in the incipient stage and alert Operations personnel to the fire. Manual firefighting equipment is located in rooms adjacent to the EDG room and exhaust rooms. The configuration of the exhaust fan rooms would also prevent fire to propagate to the other exhaust fan room or EDG rooms.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-198 Fire Area ID: QQ - North Emergency Diesel Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2094-Q North EDG Room Y Yes No No No No Yes Yes No No No No 2114-I EDG Air Intake Room No Yes No No No No No Yes No No No No P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area QQ
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-199 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2097-X East D.C. Equipment Room 2100-Z South Switchgear Room 2102-Y East Battery Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-A from condensate storage with SW as a backup. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 aligned to offsite power. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red train Control Room HVAC. Variance from the deterministic requirements of NFPA 805 exists
 
for this performance goal. A FRE is required. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. 
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-200 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 35 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachments 8.25 Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to manual methods. The presence of curb plates to adjoining areas confines water to the fire affected area. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criteria.
Licensing Actions Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations Engineering Evaluation ID: CALC-89-R-2002-107 Penetration Seal Analysis for Penetration 2104-02-0091 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of the availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-12 Penetration Seal Analysis for Penetration 2104-02-0101 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability: Because of availability of fire suppression and detection to protect against the hazards of the area, as well as the limited combustible loading, and superior penetration seal construction as compared to the tested configuration, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-201 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-ANO2-FP-09-00029 Fire Protection Engineering Evaluation for Penetration Seals in U-2 Fire Area JJ Summary: Purpose:  This evaluation is to evaluate and document the acceptability of unit 1 penetration seals in Fire Area JJ to be used in a 3-hour rated fire area boundary. Basis for Acceptability: The penetrations are considered adequate for the hazards in the area based on the fire duration on both sides of the barrier, smoke detection systems on both sides of the barrier (and line type heat detection system and suppression systems on the near side) and the response by the fire brigade to suppress a fire in the incipient stage on either side. Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2097-X East D.C. Equipment Room No Yes No No No No No Yes No Yes No Yes 2100-Z South Switchgear Room No Yes No No No No No No No Yes No Yes 2102-Y East Battery Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-202 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-19 Title: ANO-2 Fire Area SS Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2CV-1038-2 EFW discharge valve  2CV-1425-1 ACW isolation valve  2CV-1470-1 Sluice gate for 2P-4A  2EFW-802 EFW manual suction valve  2CV-1026-2 EFW discharge valve  2CV-1037-1 EFW discharge valve  2CV-1039-1 EFW discharge valve  2CV-1076-2 EFW discharge valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-203 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
Summary (continued) The following modifications are area specific and credited to reduce risk in this fire area:  2A-4 circuit reroute modification  2A-409 circuit reroute modification  2B-6 circuit reroute modification  2P-32A, 2P-32B, 2P-32C, and 2P-32D DC control power  2A-3 DC control power and circuit reroute modification  2A-308 DC control power and circuit reroute modification  2A-309 DC control power and circuit reroute modification  2A-310 DC control power and circuit reroute modification  2CV-0789-1(2PIS-0789-1) circuit reroute modification  2CV-1040-1AC power source modification  2D-27 circuit reroute modification  2K-4A circuit reroute modification  2P-16A circuit reroute modification  2P-36A circuit reroute modification  2SV-0724-1 circuit reroute modification  2SV-2809-1 circuit reroute modification  2SV-2810-1 circuit reroute modification  2SV-2811 circuit reroute modification IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. Additional Fire Area Considerations The detection system located in Fire Area SS was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area SS has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-204 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None VFDRs  VFDR ID: SS-01  VFDR: Fire damage to control and power cables in the area may impact EFW functions resulting in the following:
a) Spurious start of the non-credited EFW pump 2P-7A and loss of Control Room trip function. b) Spurious closure of valve 2CV-1038-2 which feeds the credited SG A. c) Loss of remote operation from Control Room to credited EFW pump 2P-7B requires local operation from the switchgear.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions: 
 
a) No further actions are required for 2P-7A. b) Recovery action associated with 2CV-1038-2. c) Modifications associated with switchgear (2A-3) to maintain control of 2P-7B (see Attachment S).
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-205 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: SS-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Spurious trip of the charging pumps 2P-36A from a SIAS signal and loss of Control Room start function. b) Spurious containment spray pumps 2P-35A and 2P-35B start from a containment spray signal could result in a flow diversion and a loss of DC control power could result in a loss of Control Room trip capability.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required as 2P-36A is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. b) No further action is required as spurious operation of the containment spray pumps are not a contributor to core damage sequences in the fire PRA, and therefore is not risk significant.
 
VFDR ID: SS-03  VFDR: Fire damage to DC power cables could result in a loss of the Control Room trip capability of the RCPs 2P-32A through 2P-32D. Securing the pumps assures normal pressurizer spray is secured and prevents potential RCP seal damage that could result in a se al loss of coolant accident (LOCA).
Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with modification to install a secondary DC power supply directly to the RCP trip breakers at switchg ear 2H-1 and 2H-2 (see Attachment S).
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-206 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: SS-04  VFDR: Loss of DC control circuit power to Pressurizer Heater banks #1, #2, #4, and #6 resulting in a loss of remote trip from the Con trol Room. Loss of this function could challenge the Pressure Control Performance Criterion. This condition represents a variance from th e deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
 
VFDR ID: SS-05  VFDR: Fire damage to power and control cables in the area could result in the following impacts on vital auxiliaries:
a) Losses of control capability from the Control Room to credited SW pump 2P-4A and swing pump 2P-4B(R). SW is the long-term source of feed-water to the SGs via its connection to EFW system once condensate is depleted. Offsite power is available eliminating the immediate need for SW cooling of the EDG.
b) Loss of valve 2CV-1425-1 prevents isolation of Auxiliary Cooling Water needed to control SW pump run-out if only one pump is available.
c) Loss of power and control to lake sluice gate 2CV-1470-1, if 2P-4A is the operable pump. Spurious operation of this sluice gate could result in a loss of a source of SW to 2P-4A.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805. Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further actions are required for 2P-4A or 2P-4B(R).
b) Recovery action associated with 2CV-1425-1.
c) Recovery action associated with 2CV-1470-1.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-207 Fire Area ID: SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: SS-06  VFDR: Fire damage to power and control cables in the area could result in the following impacts on vital auxiliaries:
a) Loss of DC control power to load centers B-3 and B-4 concurrent with a fault on 2B-32 and 2B-42 load cables can result in loss of Startup Transformer #3 cooling, degrading load capacity of transformer, and possible loss of offsite power to 2A-1 and 2A-2.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Loss of transformer cooling is a long term degradation issue not associated with post-fire shutdown. 
 
VFDR ID: SS-07 (CR-ANO-C-2006-00048-033)
VFDR: Fire damage to cables could result in the loss of Control Room Ventilation. Room heat-up could challenge equipment and Control Room habitability.
Loss of this function could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from t he deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: The evaluation performed by CALC-10-E-0010-05 "ANO-2 Auxiliary Building Integrated Room Heat Up Model to Support NFPA-805 Analysis" has demonstrated that Control Room HVAC is not necessary to support post fire shutdown.
End of Fire Area SS
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-208 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Zone ID Description 2108-S Electrical Equipment (2B9/2B10) Room
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A or 2P-36B available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7B feeding SG-B from condensate storage with SW as a backup Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to offsite power. Variance from the deterministi c requirements of NFPA 805 exists for this performance goal. A FRE is required. 5b. Vital Auxiliaries (SW) SW pump 2P-4A or 2P-4B feeding SW header 1. Variance from the deterministic requirements of NFPA 805 exists for this performance goal. A FRE is required. 5c. Vital Auxiliaries (HVAC) Red train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 36 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.25 Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-209 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Fire Suppression Activities Effect on Nuclear Safety Performance Criteria
 
Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. This fire area has no automatic suppression system and firefighting activities are limited to controlled manual methods using a hose station from Fire Area JJ. Any water from manual suppression is expected to migrate out the open doorway where the presence of drains and a scupper will be adequate to accommodate the flow. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance criter ia.
Licensing Actions
 
Licensing Action: No licensing actions are applicable to this fire area. Licensing Basis:
N/A  Engineering Evaluations
 
Engineering Evaluation ID: CALC-89-R-2002-15 Penetration Seal Analysis for Penetration 2108-07-0076 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to utilize it in a three (3)-hour rated fire boundary sinc e the seal deviates from the tested configuration. Basis for Acceptability:  Because of availability of manual fire suppression and detection to protect against the hazards of th e area, as well as the limited combustible loading, this evaluation has determined this deviation from tested design as having negligible impact and therefore acceptable.
Engineering Evaluation ID: CALC-89-R-2002-24 Penetration Seal Analysis For Penetration 2108-02-0007 Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection systems on both sides of the barrier (and line type heat detection system and suppression systems on the corridor side) and the response by the fire brigade to suppress a fire in the incipient stage on either side. The differences identified will have no affect on the seal since the installed configuration a nd the tested configuration both contain similar penetrating items and the installed configuration is an equivalent in heat transmission.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-210 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID: CALC-89-R-2002-32 Penetration Seal Analysis For Penetration Summary: Purpose:  Evaluate penetration seal to determine if it is acceptable to be utilized in a three (3)-hour rated fire boundary sin ce it deviates from the tested configuration. Basis for Acceptability:  Based on the fire duration, smoke detection and manual suppression systems on both sides of the barrier, and the response by the fire brigade to suppress a fire in the incipient stage on either side the differences identifi ed will have no affect on the seal since the installed configuration and the tested configuration both contain similar penetrating items and the installed configuration is equivalent in heat transmission.
 
Engineering Evaluation ID: CALC-ANO2-FP-09-00030 Fire Protection Engineering Evaluation for Penetrations Seals in Fire Area TT Summary: Purpose:  This evaluation is to evaluate and document the acceptability of penetrations in Fire Area TT to be used in a 3-hour rated fire area boundary. Basis for Acceptability:  The penetration seals are considered to be adequate for the hazards in the area based on the combustible loading, smoke detection systems, suppression systems (far sides only) and the response by the fire brigade to suppress the fire in the incipient stage.
Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2108-S Electrical Equipment (2B9/2B10) Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed. Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-211 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation: CALC-09-E-0008-20 Title: ANO-2 Fire Area TT Risk Evaluation Summary: The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205. Credited Recovery Actions There following equipment is recovered in the post transition baseline case:  2A-309 ESF offsite power breaker  2CV-1036-2 EFW discharge valve  2CV-1075-1 EFW discharge valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:  New AFW source independent of existing EFW/AFW pumps. Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2. The following modifications are area specific and credited to reduce risk in this fire area:  2B-5 circuit reroute modification  2P-7B circuit reroute modification  2CV-0789-1 circuit reroute modification  2CV-1036-2 circuit reroute modification  2CV-1075-1 circuit reroute modification  2CV-4816 and 2CV-4817 Letdown valves circuit reroute modification IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. The circuit reroutes for 2CV-1036-2 and 2CV-1075-1 resolves IN 92-18 issues for this fire area. Additional Fire Area Considerations The detection system located in Fire Area TT was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer. Fire Area TT has been screened for hot gas layer analysis, multi-compartment analysis, and HGL effects on zone of influence.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-212 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)
CDF: Refer to Attachment W "Fire PRA Insights"  LERF: Refer to Attachment W "Fire PRA Insights" DID Maintained: The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area. Safety Margin Maintained: All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.
Comments: None Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-213 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs VFDR ID: TT-01  VFDR: Fire damage to control and power cables in the area may impact EFW functions resulting in the following:
a) Spurious closure of valves 2CV-1036-2 and 2CV-1075-1 (IN 92-18) which feeds the credited SG B. b) Spurious opening of valve 2CV-1052, atmospheric dump to SG B. c) Loss of MSIV 2CV-1060-2 isolation capability to the credited SG B. d) Loss of control capability of the credited EFW pump 2P-7B from the Control Room. e) Spurious closure of 2CV-0789-1 resulting in isolation of CST supply to EFW Pump 2P-7B. f) Spurious closure or failure to open 2CV-0716-1 resulting in loss of SW supply to EFW pump 2P-7B needed for long-term source of feedwater upon depletion of condensate. g) Loss of MSIV 2CV-1010-1 and MSIV bypass 2CV-1040-1 isolation capability to non-credited SG A. h) Loss of EFW pump 2P-7B discharge valve 2CV-1038-2 isolation capabilities to non-credited SG A.
Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance fro m the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) Recovery actions and modification to re-route cables associated with 2CV-1036-2 and 2CV-1075-1 (see Attachment S). b) No further action is required for 2CV-1052. c) No further action is required for 2CV-1060-2. d) Modification to re-route cables associated with 2P-7B (see Attachment S). e) Modification to re-route cables associated with 2CV-0789-1 (see Attachment S). f) No further action is required for 2CV-0716-1. g) No further action is required for 2CV-1010-1 and 2CV-1040-1.
h) No further action is required for 2CV-1038-2.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-214 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: TT-02  VFDR: Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:
a) Loss of valves 2CV-4816 and 2CV-4817 prevent isolation of letdown. b) Spurious opening or failed open of valve 2CV-5630-1 could result in RWT drain-down to containment sump in conjunction with spurious opening of 2CV-5649-1. c) Spurious trip of the charging pumps 2P-36A and 2P-36B or losses of Control Room start function. d) Loss of open capability for valves 2CV-4920-1 and 4921-1 resulting in isolation of BAM supply to charging pumps. e) Spurious closure or loss of open capability for valve 2CV-4950-2 results in isolation of RWT supply to charging pumps.
Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) Modification to re-route cables associated with 2CV-4816 and 2CV-4817 (see Attachment S). b) No further action is required for 2CV-5630-1 and 2CV-5649-1. c) No further action is required for 2P-36A and 2P-36B. d) 2CV-4920-1 and 2CV-4921-1 are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios. e) No further action is required as 2CV-4950-2 is a component in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-215 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: TT-03  VFDR: Fire damage to cables in the area could result in the following impacts on vital auxiliaries:
a) Spurious trip and loss of breaker control for Load Center 2B-5.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with modification to re-route impacted cables associated with load center 2B-5 (see Attachment S).
VFDR ID: TT-04 VFDR: Loss of DC control circuit power to Pressurizer Heater banks #3, #4, #5, and #6 resulting in a loss of remote trip from the Con trol Room. Loss of this function could challenge the Pressure Control Performance Criterion. This condition represents a variance from th e deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-216 Fire Area ID: TT - Electrical Equipment (2B9/2B10) Room Compliance Basis: NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID: TT-05  VFDR: Fire damage to power and control cables in the area could result in the following impacts on vital auxiliaries:
a) Loss of control capability from the Control Room to the credited SW pumps 2P-4A and swing pump 2P-4B(R). SW is the long-term source of feedwater to the SGs via its connection to EFW system once condensate is depleted. Offsite power is available eliminating the immediate need for SW cooling of the EDG.
b) Loss of valve 2CV-1427-2 prevents isolation of Auxiliary Cooling Water needed to control SW pump run-out if only one pump is available.
c) Loss of 2CV-1530-1 (SW Loop 1 to CCW Heat Exchangers) and 2CV-1525-1 (SW Loop 1 to Fuel Pool Heat Exchanger) prevents isolation of credited SW loop to prevent pump run-out.
Loss of these functions could challenge the Vital Auxiliaries Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.
Disposition: This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:
a) No further action is required for 2P-4A and 2P-4B(R). b) No further action is required for 2CV-1427-2. c) No further action is required for 2CV-1530-1 and 2CV-1525-1.
End of Fire Area TT
 
Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-217 Fire Area ID: YD - Yard Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Fire Zone ID Description YARD Yard
 
Performance Goal Method Of Accomplishment Comments 1. Reactivity Control Manual reactor trip from the Control Room. Long term reactivity control by initial inventory addition from BAMTs prior to switching to refueling water tank.
: 2. Inventory Control Letdown isolated and RCPs secured to maintain seal integrity. Charging pump 2P-36A, 2P-36B, or 2P-36C available with gravity feed from BAMTs using normal charging path to RCS and auxiliary pressurizer spray secured.
: 3. Pressure Control RCS vent paths are secured. Pressurizer heaters are de-energized and normal pressurizer spray secured (RCPs turned off). Pressure maintained by inventory addition.
: 4. Decay Heat Removal Main steam isolated, normal feedwater secured, and steam release using MSSVs, if atmospheric dump valves are not immediately available. EFW pump 2P-7A or 2P-7B feeding SG-A and SG-B from condensate storage with SW as backup.
5a. Vital Auxiliaries (Electrical) ESF 4.16KV switchgear 2A-3 and 2A-4 aligned to onsite EDGs. 5b. Vital Auxiliaries (SW) SW pumps 2P-4A, 2P-4B, and 2P-4C available to feed SW headers 1 and 2.
5c. Vital Auxiliaries (HVAC) Red and green train Control Room HVAC. 6. Process Monitoring Instrumentation is available in the Control Room to monitor neutron flux, pressurizer level, RCS pressure, RCS temperature, and credited SG level and pressure. Backup from SPDS is available.
Reference Document CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results, Rev. 4, Attachment 37 CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment, Rev. 9, Attachment 8.26
 
Fire Suppression Activities Effect on Nuclear Safety Performance Criteria Safe and stable conditions can be achieved and maintained utilizing equipment and cables outside of the area of fire suppression activity. Each outdoor deluge system provides local protection, specifically for individual transformers, and sub-grade basins installed to catch and carry away oil and water to a remote separator. Discharge of manual suppression water to adjacent areas is non-consequential as site grading carries any water away from structures and equipment. Fire suppression activities will therefore not adversely affect the plant's ability to achieve the nuclear safety performance c riteria. Fire Area ID: YD - Yard Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition Enclosure 1 to 2CAN121202 Page C-218 Compliance Basis: NFPA 805 Section 4.2.3.2 - Deterministic Approach Licensing Actions Licensing Action: Appendix R, Exemption 16, FA - YD, Not Meeting III.G.2 Criteria, NRC approval letter 2CNA108802 dated 10/26/1988. Licensing Basis: This exemption is no longer required because the postulated loss of tank level indication does not cause a fire related need for plant shutdown, due to the Technical Specification required minimum volume maintained during power operation, and remaining deterministically compliant under NFPA 805 criteria.
Licensing Action: Appendix R, Exemption 18, FA - YD, Emergency Lighting, Not Meeting III.J Criteria, NRC approval letter 2CNA108802 dated 10/26/1988.Licensing Basis: This exemption is no longer required because NFPA 805 does not require 8-hour battery backed emergency lights.
Engineering Evaluations Engineering Evaluation ID: No engineering evaluations are applicable to this fire area. Summary: N/A  Required Fire Protection Systems and Features Required?  Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET YARD Yard P* P* No No No No No No No No No No
* Associated with deluge system for main and unit auxiliary/startup transformers.
P - Indicates a partial system is installed Separation - Required for Chapter 4 Separation Criteria LA- Required for NRC-Approved Licensing Action EEEE- Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.
VFDRs  This fire area is in deterministic compliance and has no VFDRs.
End of Fire Area YD
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-1 D. NEI 04-02 Non-Power Operational Modes Transition (NEI 04-02 Table F-1) 1.3.1 Nuclear Safety Goal The nuclear safety goal is to provide 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.
Implementing Guidance F.1 Review Existing Outage Management Processes
 
Define Higher Risk Evolutions (HREs), if not already defined in plant outage management procedures. The HRE definition should consider the following:
 
Time to boil  Reactor coolant system and fuel pool inventory  Decay heat removal capability Review  OP-1015.048 is the ANO Shutdown Operations Protection Plan (SOPP) and defines HREs as "Activities, plant configurations, or conditions during outages where the plant is more susceptible to an event causing a loss of Key Safety Function."
 
The Shutdown Conditions dealt with by the SOPP are divided into six conditions based on fuel location, RCS and fuel transfer canal inventory, and RCS status of either intact or open. These are numbered from the condition with the lowest relative risk to shutdown operations (Condition 1 - Reactor Vessel Defueled) to the condition with the highest relative risk (Condition 6 - Reduced Inventory, RCS open, Fuel in the Reactor Vessel).
 
Unit Applicability 2 Comments None  Reference Document Document Detail CALC-09-E-0008-02, ANO-2 NFPA 805 Non Power Sections 3.1, 4.2 Operations Assessment, Rev. 0 OP-1015.048, Shutdown Operations Protection Plan, Rev. 9 All
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-2 1.3.1 Nuclear Safety Goal The nuclear safety goal is to provide 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.
Implementing Guidance F.2 Identify Components and Cables
 
The identification of systems and components to be included in this Non-Power Operations (NPO) Review begins with the identification of the Plant Operational States (POSs) that need to be considered.
 
Review As described in NUMARC 91-06, the five Key Safety Functions (KSFs) are:
Decay Heat Removal Capability  Inventory Control  Reactivity Control  Containment Closure  Electrical Power Availability Based on FAQ 07-0040, Revision 4, the POSs considered for equipment and cable selection in the ANO NPO review are:
POS 1  POS 2  POS 3  The evaluation of these POSs resulted in the exclusion of the Containment Closure KSF from further consideration. SFP cooling was also excluded from the DHR (referred to as the SDC
 
system on ANO-2) KSF. The remaining KSFs were evaluated to determine which POS required consideration for selection of equipment and cable necessary to maintain the KSF. The summary of each KSF in relationship to the POS considered in the ANO-2 NFPA 805 NPO Assessment are:
Decay Heat Removal Capability
 
An evaluation of the SDC system during POS 2 (mid-loop) effectively bounds POS 1a (drain-down). A loss of SDC during POS 3 is not an immediate concern due to the large inventory available and long times to boil.
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-3 Inventory Control
 
An evaluation of the Inventory Control KSF (drain paths & makeup) during POS 2 (mid-loop) effectively bounds POS 1a (drain-down) and POS 3 (drain paths).
 
Reactivity Control
 
The inclusion of source range nuclear instrumentation assures reactivity changes are quickly identified and actions can be taken to assure maintenance of this KSF during all POS.
Electric power Availability
 
Offsite power and both trains of onsite emergency power are evaluated to assure this support function is maintained for all POS considered for Decay Heat Removal, Inventory Control, and Reactivity Control KSFs. Electrical power will be limited to those electrical systems needed to directly support equipment required for Decay Heat Removal, Inventory Control, and Reactivity Control.
 
The equipment needed for each KSF was determined by review of applicable P&IDs, single line diagrams, schematics, and procedures to determine the extent needed for NPO. Markups of the P&IDs and single line diagrams are performed and included as an attachment to CALC E-0008-04. Fault tree development, equipment operating states, and circuit analysis needed for NPO are described in detail in CALC E-0008-02, "ANO-2 NFPA 805 Non Power Operations Assessment."
Unit Applicability 2 Comments None Reference Document Document Detail CALC-09-E-0008-02, ANO-2 NFPA 805 Non Power Sections 4.2, 4.3, 4.4, and 4.5 Operations Assessment, Rev. 0
 
CALC-09-E-0008-04, ANO-2 NFPA 805 NPO Fault Tree and All P&ID Attachments, Rev. 0
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-4 1.3.1 Nuclear Safety Goal The nuclear safety goal is to provide 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.
Implementing Guidance F.3 Perform Fire Area Assessments (Identify Pinch Points)
Identify locations where:
Fires may cause damage to the equipment (and cabling) credited above, or  Recovery actions credited for the KSF are performed (for those KSFs that are achieved solely by recovery action, i.e., alignment of gravity feed).
Fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling) thereby eliminating a pinch point.
Review  The pinch point analysis is performed using ARC software. ARC software extracts the necessary data from the Plant Data Management System (PDMS) and maps it to the CAFTA fault tree. Each Fire Area for NPO is evaluated to determine which equipment could be rendered unavailable. Equipment which could spuriously operate or result in the loss of a KSF was given a compliance strategy (typically a manual action) to allow NPO compliance (top gate success). This effectively captures equipment necessary to maintain a KSF in any fire area/zone but could be compromised due to a fire. This provides for each fire area a maximum set of recovery actions that may potentially be required to restore each KSF.
Areas not in deterministic compliance have recovery actions noted and a risk-informed process is used to determine if the defense-in-depth strategy is adequate to maintain the KSFs. This was performed using the following sequence for each impacted area/zone:
Determine the NFF (NPO Compartment F ire Ignition Frequency)  Review area/zone for detection and suppression  Consideration of recovery actions  Circuit failure likelihood No pinch point was excluded in the current NPO analysis but may be considered a viable option for future plant changes.
Unit Applicability 2  Comments None  Reference Document Document Detail CALC-09-E-0008-02, ANO-2 NFPA 805 Non Power Section 6.0 Operations Assessment, Rev. 0
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-5 1.3.1 Nuclear Safety Goal The nuclear safety goal is to provide 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.
Implementing Guidance F.4 Manage Risks Associated with Fire-Induced Vulnerabilities During the Outage
 
During Those NPO Evolutions Where Risk is Relatively Low The following actions are considered to be adequate to address minor losses of system capability or redundancy:
Control of Ignition Sources o Hot Work (cutting, welding and/or grinding) o Temporary Electrical Installations o Electric Portable Space Heaters  Control of Combustibles o Transient Fire Hazards o Modifications o Flammable and Combustible Liquids and Gases  Compensatory Actions for Fire Protection System Impairments o Openings in Fire Barriers o Inoperable Fire Detectors or Detection Systems o Inoperable Fire Suppression Systems  Housekeeping During Those NPO Evolutions that are Defined as HREs Additional fire protection defense-in-depth measures will be taken during HREs by:
Managing risk in fire areas that contain known pinch points  Managing risk in fire areas where pinch points may arise because of equipment taken out of service
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-6 For those areas, consider combinations of the following options to reduce fire risk depending upon the significance of the potential damage:
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 lineup modifications (removing power from equipment once it is placed in its desired position)  Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during increased vulnerability  Use of recovery actions to mitigate potential losses of key safety functions  Identification and monitoring in-situ ignition sources for "fire precursors" (e.g., equipment temperatures)
In addition, for KSF Equipment removed from service during the HREs, the impact should be evaluated based on KSF equipment status and the Fire Area Assessment to develop needed contingency plans/actions.
Review  The normal fire protection programs such as combustible and hot work control are maintained during NPO modes. Operability of detection and suppression systems is maintained.
 
In fire areas/zones where a pinch point is created, a risk-informed evaluation is performed to determine if defense-in-depth strategies are adequate to assure maintenance of each KSF. The type of equipment present and its role in maintenance of KSFs provide locations where no hot work is to be performed during NPO without additional compensatory actions in place, such as securing of equipment in the safe position (i.e. power removed). Identification of modifications is included to reduce risk.
 
Unit Applicability 2  Comments None  Reference Document Document Detail CALC-09-E-0008-02, ANO-2 NFPA 805 Non Power Sections 6.0 and 6.1 Operations Assessment, Rev. 0 CALC-08-E-0016-01, Fire Probabilistic Risk Assessment Plant All Partitioning and Fire Ignition Frequency Development, Rev. 0 EN-DC-127, Control of Hot Work and Ignition Sources, Rev. 11 All EN-DC-161, Control of Combustibles, Rev. 7  All EN-DC-330, Fire Protection Program, Rev. 1 All
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-7 VFDR ID NPO-Procedure
 
Description Operations procedures for NPO are required for transition to NFPA-805 based upon the insights gained from the ANO-2 NPO calculation. This can be accomplished by either incorporation into an existing procedure such as OP-2203.049 "Fires in Areas Affecting Safe Shutdown" or the development of a new procedure (refer to Attachment S of this Enclosure). This task will be completed during NFPA-805 implementation following iss uance of the NRC SER. The NPO procedure will incorporate:
Available equipment by fire affected area  Manual recovery actions  Compensatory Actions Disposition This open item is being tracked to completion by CR-ANO-C-2006-00048 CA 72
 
Status Open  Corrective Action CR-ANO-C-2006-00048 CA 72 Reference  Include in LAR/TR Yes FRE / Change Eval /
Mod Reference
 
Arkansas Nuclear One - Unit 2 Att. D - NEI 04-02 Non-Power Operational Modes Enclosure 1 to 2CAN121202 Page D-8 VFDR ID NPO-RCS-SDC
 
Description ANO-2 has no redundancy with respect to the single RCS drop line with three in-series valves 2CV-5038-1, 2CV-5084-1, and 2CV-5086-2. The risk associated with the RCS drop line valves is low, but the consequences of a spurious failure is high (loss of SDC without recovery) as all three valves are NRC IN 92-18 concerns and can fail in the closed position. Procedural changes will be made to secure valve(s) by removing power and/or modification of the credited valve(s) performed to prevent spurious operation (refer to Attachment S of this Enclosure). The impacted fire areas for each valve are as follows:
2CV-5038-1:  DD, HH, B-2  2CV-5084-1:  JJ, EE-L  2CV-5086-2:  HH Disposition This open item is being tracked to completion by CR-ANO-C-2006-00048 CA 72 One or more valves in this path require either physical or electrical modification to eliminate IN 92-18 issues. Procedural changes can be made to immediately secure (open breaker) any valve that will not be required to remain in service once it is opened to establish flow. The most economical approach would be to procedurally disable two of the three valves in the drop line by opening the breaker after the system is aligned. The single valve selected to remain active will require circuit modification to prevent spurious operation or physical modification/replacement of the valve/operator to prevent damage that could prevent repositioning.
 
Status Open  Corrective Action CR-ANO-C-2006-00048 CA 72 Reference  Include in LAR/TR Yes  FRE / Change Eval /
Mod Reference
 
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-1 E. NEI 04-02 Radioactive Release Transition Radioactive Release Analysis Radiation release to any unrestricted area due to the direct effects of fire suppression activities (but not involving fuel dam age) shall be as low as reasonably achievable and shall not exceed applicable 10 CFR, Part 20, limits.
 
Note that any changes to the pre-fire plans resulting from this analysis, in addition to those changes necessary to meet other transition requirements, are tracked and controlled within the transition project. These changes will be incorporated into the pre-fire plans as part of the NFPA 805 implementation process, following NRC approval, unless the change is necessary to meet pre-805 requirements. While one enhancement was noted during the radioactive release review, no pre-fire plan required changes were
 
identified.
 
A review of ANO-2 and common ANO-1/ANO-2 fire pre-plans (PFP-U1 and PFP-UC) and fire brigade training materials was performed to identify Fire Protection Program (FPP) elements (e.g., systems / components / procedural control actions / flow pa ths, etc.) that are being credited to meet the radioactive release goals, objectives, and performance criteria during all plant oper ating modes, including full power and non-power conditions.
 
Training Review A review of fire brigade training materials to ensure that tr aining materials deal specifically with the containment and monito ring of potentially contaminated fire suppression water was performed.
 
Procedure OP-1063.020, Fire Brigade Training Program, describes the Fire Brigade Training Sequence to assure the capability to fight potential fires is established and maintained. The procedure is applicable to the initial training and retraining of the Fire Brigade personnel at ANO-1 and ANO-2. Fire Protection Engineering has responsibility for reviewing and being knowledgeable of the training requirements of the Fire Brigade and for assessing the effectiveness of the Fire Brigade training.
 
Initial training for Fire Brigade Members and Support Members consists of a scheduled program of instruction as detailed in the Fire Brigade Training Program and Course Summary. The Initial Fire Brigade Training class includes classroom training, hands on training and practice, and fire-fighting scenarios using controll ed fire environments. Fire Brigade Leader Training is provide d to Operations personnel to ensure that personnel meeting the requirements for Fire Brigade Leader are capable of taking charge at the scene of a fire.
 
Continuing Fire Brigade Training is taught on a periodic basis to ensure that the capability to fight fires is maintained. Top ic areas (content) of the Initial Training are repeated every two years in the Continuing Training Program. Fire Brigade members and su pport members also attend Annual Practice Class and drills to maintain needed skills.
 
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-2 Training on radiological release potential is provided in lesson plan ASLP-FP-CAFRS, "Responding to Fires in Controlled Access" in the Fire Brigade training program. This lesson plan addresses radioactive contamination and the need for monitoring and containment. Specifically, the areas of "Flooding Concerns" and "Ventilation Concerns" are addressed. The lesson plan states that "consideration must be given to the path the smoke and gases will take when they are evacuated."  Additionally, "any ventilation path that does not provide for the smoke and gases from the fire to be monitored for radiological contamination should be discussed with the Control Room and Radiation Protection prior to being used."  These principles are further supported and enhanced in Fire Brigade Leader training. Radioactive materials areas outside "Controlled Access" are also addressed in the pre-fire plans.
 
Pre-fire Plan Review As stated previously, a comprehensive review of the ANO-2 and common ANO-1/ANO-2 pre-fire plans was performed. Each pre-fire plan contains information which may be utilized by Fire Brigade or other support personnel in responding to a fire within the f acility. In addition, information has been included which may be pertinent to operations support personnel in performing safe shutdown activities in response to any single fire scenario. The analysis and review of fire-related radioactive release included locat ions that have the potential for contamination such that specific steps are included for containment and monitoring of potentially contam inated fire suppression water. Pre-fire plans contain the following information:
 
Occupancy Manual Suppression Fire Brigade Access Ventilation Plant Personnel Egress Safe Shutdown Impacts Lighting/Communication Guidelines for Fire Attack Hazards Special Precautions/Notes Fixed Fire Systems A Map of the Fire Zone
 
Closure of the Containment Building is controlled per Operations procedure OP-1015.008, Unit 2 SDC Control, and a breach list i s maintained during Modes 5 and 6. Containment openings are internal to the plant with the exception of the Containment Equipmen t Hatch and the Emergency Escape Hatch. Closure of the Equipm ent Hatch for Containment integrity during Modes 5 and 6 is established by a Containment closure plan with a specific closure time identified. While a specific closure time is not specified for the defueled condition, plant procedure OP-2104.033, Containment Atmosphere Control, provides instructions for Operations to contro l ventilation in Containment to maintain negative pressure and thereby prevent effluent flow from the Equipment Hatch or Emergenc y Personnel Hatch. In addition, Radiation Protection monitors the airflow at the Equipment Hatch per procedure EN-RP-131, Air Sampling, and maintains a continuous air monitor.
 
Based on the volume of Containment for collection of smoke and the location of the Equipment Hatch in relation to the top of Containment (~150' below top of dome), the potential for smoke migration to lower levels is generally not considered credible p rior to containment integrity and monitoring actions being taken. Large ignition sources such as RCPs and their associated oil supply were Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-3 considered the largest contributor. Due to lack of large components such as these pumps/motors operating during this plant configuration, no ignition source was identified. Additionally, with the heightened personnel attendance and monitoring of Containment, the potential for fire hazards large enough to present a potential release is unlikely. Administrative controls f or hot work and handling of transient combustibles during outages further enhance the prevention, detection and response elements of defense-in-depth for this area, ensuring the potential for radioactive release is minimized. Therefore, radiation release to a ny unrestricted area due to the direct effects of fire suppression activities (but not involving fuel damage) is expected to be as low as reasonably achievable and not exceed applicable 10 CFR Part 20 limits.
 
With regard to the Auxiliary Building, the Auxiliary Building drains are collected and monitored per operating procedures OP-1104.014, "Dirty Liquid Waste and Drain Processing" for ANO-1 and OP-2104.014, "LRW and BMS Operations" for ANO-2.
 
The Turbine Building is generally open to the outdoors. Potential sources of radioactivity are generally contained within stee l vessels and piping that are not expected to be breached as a result of firefighting activities. Existing smoke vents or roof exhaust f ans are the preferred engineering controls for a fire in the Turbine Building. Turbine Building drains are routed to the turbine build ing sump and retained for monitoring prior to processing and/or release.
 
The pre-fire plans address areas where run-off or ponding of fire suppression water may be an issue.  "Guidelines for Fire Atta ck" and "Special Precautions/Notes" sections of the pre-fire plans contain specific steps to implement based on the potential probl ems for the given fire zone. The pre-fire plans also address post-fire ventilation. For example, if the decision is made to re-en ergize permanent plant ventilation systems, it is important to energize exhaust air first.
 
The Common pre-fire plans specifically address three areas that store low level radioactive waste or materials. These three ar eas are common to both units and are identified as the Old Radwaste Storage Building, the Low Level Radwaste Storage Building, and the RP Storage Building.
 
Support Documentation is provided in Table E-1 below that lists pre-fire plans by fire zone for ANO-2 and also for areas common to ANO-1 and ANO-2. Fire Zones were "screened in" for consideration based on radiation levels greater than "low" or the presence of radiological hazards identified in the pre-fire plan. Areas for controlled storage of radioactive sources, transitional areas, and some isolated areas with low levels of fixed contamination are not listed.
 
Engineering Controls A review of engineering controls to ensure containment of gaseous and liquid effluents (e.g., smoke and fire fighting agents) w as performed for ANO-2 and areas common between ANO-1 and ANO-2. This review included all plant operating modes (including full power and non-power conditions). Where applicable, the specific engineering controls are provided in Table E-1 below.
 
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-4 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 404 2147-A Chemical Storage Area (Boric Acid Makeup Room) Y Y 2VEF14A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 404 2148-A Access Corridor Y Y 2VEF14A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 404 2150-C Core Protection Calculator Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 404 2151-A Fuel Handling Area Y Y 2VEF14A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 404 2152-D Computer Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 404 2153-A Ventilation Equipment Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 404 2154-E Control Element Drive Mechanism Equipment Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 404 2155-A Steam Pipe Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 404 2156-A Containment Purge Air Equipment Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 404 2200-MM Turbine Building 404' N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 386 2114-I Emergency Diesel Generator Air Intake Area N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-5 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 386 2115-I Boric Acid Makeup Tank Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 386 2119-H Printer Room Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 386 2136-I Health Physics Area (CA-2) Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 386 2137-I Upper South Electrical Penetration Room, Hot Instrument Shop, and Decontamination Rooms Y Y 2VEF8A/B 2VEF38A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 386 2183-J Upper North Electrical Penetration Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 386 2199-G Control Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 386 2200-MM Turbine Building 386' N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 386 2243-NN Chemistry Lab and Offices N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2091-BB North Electrical Equipment (2Y22/2Y24) Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2092-PP Chiller Equipment Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2093-P South Diesel Generator Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2094-Q North Diesel Generator Room N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-6 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 372 2096-M Motor Control Center (2B63) Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2097-X East DC Equipment Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2098-C Core Protection Calculator Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2098-L Cable Spreading Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2099-W West DC Equipment (2Y11/2Y13) Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2100-Z South Switchgear (2A4) Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2101-AA North Switchgear (2A3) Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2102-Y East Battery Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2103-V West Battery Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2106-R Vacuum Degasifier Tank Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 372 2107-N Corridor Area (In Front of Degasifier and 2B63 Rooms) Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 372 2108-S Electrical Equipment (2B9/2B10) Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2109-U Diesel Generator Room Access Corridor & Motor Control Center N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-7 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 372 2111-T Lower South Electrical Penetration Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2112-BB Lower North Electrical Penetration Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2200-MM Turbine Building 372' N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 372 2242-OO Heating & Ventilation Mechanical Equip Room and Auxiliary Operator Shack Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2063-DD Sample Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2068-DD Hot Machine Shop Y Y 2VEF14A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 354 2072-R Volume Control Tank Room and Pump Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 354 2073-DD Access Corridor, Pump and Tank Area  Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 354 2076-HH Electrical Equipment (Motor-Generator Sets) Room Y Y None OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 354 2078-QQ Heat Exchanger Equipment Area N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-8 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 354 2081-HH Upper North Piping Penetration Area Y Y 2VEF8A/B 2VEF38A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 354 2084-DD Upper South Piping Penetration Room and Equipment Area Y Y 2VEF8A/B 2VEF38A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 354 2172-ZZ Radiation Protection (RP) Office Area/Break Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2178-AAA Lube Oil Reservoir N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2200-MM Turbine Building 354' N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2229-SS(A) Demineralizer Equipment N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2229-SS(B) Office Area N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2230-RR Boric Acid MU Tank Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2231-TT Plant Heating Boiler Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 354 2261-UU Plant Heating Boiler Day Tank N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 335 2019-JJ Boric Acid Condensate Tank Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-9 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 335 2020-JJ Boron Management System (BMS) Holdup Tank Vault and Pump Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2024-JJ Emergency Feedwater (EFW) Pump 2P7A (Turbine) Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2025-JJ EFW Pump 2P7B (Motor Driven) Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2032-JJ Spent Resin Storage Tank Room Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2040-JJ Access Corridor; Charging Pump; Radwaste and Boron Management System Areas Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2045-XX Turbine Oil Storage Tank Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 335 2055-JJ Lower South Piping Penetration Area Y Y 2VEF8A/B 2VEF38A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-10 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 335 2081-HH Lower North Piping Penetration Area Y Y 2VEF8A/B 2VEF38A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2172-ZZ HP Office Area/Break Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 335 2177-YY Neutralizer Tank Room N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 335 2200-MM Turbine Building 335' N N N/A N/A N/A Not Required.
Pre-fire Plan Elev. 335 2223-KK Pipeway Equip Access Area (Auxiliary Building Extension) Y Y 2VEF52 OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 335 2225-WW Regenerative Waste Pump & Tank Room Y Y 2VEF51A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 317 2006-LL General Access Area Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 317 2007-LL East Pump Area and Gallery Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 317 2010-LL Center Pump Area Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-11 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Elev. 317 2011-LL Tendon Gallery Access Y Y 2VSF23E to 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan Elev. 317 2014-LL West Pump Area Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied. Pre-fire Plan 2032-K Containment Building, South Side Y Y 2VEF15 OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied. Prefire Plan 2033-K Containment Building, North Side Y Y 2VEF15 OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied. Pre-fire Plan 2149-B Stairway No. 2001 Y Y 2VEF8A/B OP-2104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied. Prefire Plan 2158-F Stair No. 2055 N N N/A N/A N/A Not Required.
Pre-fire Plan Diesel Fuel Vault Diesel Fuel Storage Vaults N N N/A N/A N/A Not Required.
Pre-fire Plan Main Transformers ANO-2 Main Transformer Area N N N/A N/A N/A Not Required.
Pre-fire Plan Manholes (East) Yard Manholes 2MHO1E, 2MHO2E and 2MHO3E N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-12 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB FIRE ZONE COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plan Manholes (West) Yard Manholes 2MHO1W, 2MHO2W and 2MHO3W N N N/A N/A N/A Not Required. Pre-fire Plan PASS Bldg PASS Building Y Y 2VEF73  2VEF71 OP-1104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plan ANO-2 Acid &
Caustic ANO-2 Acid, Caustic and Ammonia Tank Building N N N/A N/A N/A Not Required.
Pre-fire Plan ANO-2 Cooling Tower Cooling Tower Pumphouse Control Building N N N/A N/A N/A Not Required. Pre-fire Plan ANO-2 Intake ANO-2 Intake Structure N N N/A N/A N/A Not Required.
NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB PFP COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plans for Support Facilities 1 Administration Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 2 Old Chlorination Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 3 Oily Water Separator Facility Y N N/A N/A N/A Rad. Levels are low.
Pre-fire Plans for Support Facilities 4 Hydrogen and CO 2 Gas Bottle Storage Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 5 Lube Oil Storage Building (Warehouse #14
& #21) N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-13 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB PFP COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plans for Support Facilities 6 Old AP&L Warehouse (Warehouse #1 & #2) Y N N/A N/A N/A Rad. Levels are low.
Pre-fire Plans for Support Facilities 7 Old Radwaste Storage Building (Warehouse #19) Y Y None OP-1104.014 The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plans for Support Facilities 8 Pipe Fitter's Welding Shop / Lunchroom (Warehouse #17 & #20) N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 9 Old NSSS Warehouse (Warehouse #3) N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 10 Low Level Radwaste Storage Building Y Y Exhaust Fan A/B Sump The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
Pre-fire Plans for Support Facilities 11 Network Computer Diesel & Plant Services Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 12 Central Support Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 13 New Maintenance Facility N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 14 Cable Reel Storage Warehouse (Warehouse #6) N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 15 New NSSS Warehouse (Warehouse #5) N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 16 Insulation Warehouse (Warehouse #7)  N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 17 Alternate AC Generator Building N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-14 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB PFP COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plans for Support Facilities 18 Receiving Warehouse (Warehouse #4) N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 19 System Engineering Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 20 Primary Access Point N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 21 Technical Support Building  N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 22 ANO Sally Port N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 22 FZ-3065 ANO Sally Port
* replaces FZ-3046 and FZ-3047 N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 23 Vacuum Degassifier Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 24 Start-Up Boiler Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 25 Generation Support Building  N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 26 Reeves E. Ritchie Training Facility N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 27 Simulator Building (Training Facility)  N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 28 New Fabrication Shop  N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 29 Bulk Storage Warehouse #12 N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 30 Freight Receiving Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 31 Bulk Diesel Fuel Storage Tank, T-25 N N N/A N/A N/A Not Required.
Arkansas Nuclear One - Unit 2 Att. E - NEI 04-02 Radioactive Release Transition Enclosure 1 to 2CAN121202 Page E-15 NEI 04-02 Table E Radioactive Release Transition Radioactive Release Compartment Review ENGINEERING CONTROLS TAB PFP COMPARTMENT* RAD CONCERNS SCREENED IN Smoke Water TRAINING REVIEW RESULTS CONCLUSIONS Pre-fire Plans for Support Facilities 32 Turbine Rotor Maintenance Facility N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 33 Reactor Vessel Head Assembly Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 34 Switchyard Control Building N N N/A N/A N/A Not Required.
Pre-fire Plans for Support Facilities 35 RP Storage Building Y Y None None The NFPA 805 performance requirements for Training are satisfied. The NFPA 805 performance requirements for radiological release are satisfied.
* The "COMPARTMENT" consists of the Pre-Fire Plan book "TAB", the "FIRE ZONE" or "PFP" (Pre-Fire Plan designator), and the room/area description.
 
Arkansas Nuclear One - Unit 2 Att. F - Fire-Induced MSOs Resolution Enclosure 1 to 2CAN121202 Page F-1 F. Fire-Induced Multiple Spurious Operations Resolution The following provides the guidance from FAQ 07-0038, Revision 3, along with the process and results. Step 1 - Identify potential MSOs of concern Information sources that may be used as input include:  Post-fire safe shutdown analysis (NEI 00-01, Revision 1, Chapter 3)  Generic lists of MSOs (e.g., from Owners Groups and/or later versions of NEI 00-01, if endorsed by NRC for use in assessing MSOs)  Self assessment results (e.g., NEI 04-06 assessments performed to addressed RIS 2004-03)  PRA insights (e.g., NEI 00-01, Revision 1, Appendix F)  Operating Experience (e.g., licensee event reports, NRC Inspection Findings, etc.)
Results of Step 1:
 
Information sources that were used as input to the ANO expert panel assessment held September 27th and 28th, 2005, and described in Step 2 include the following.
: 1. United States Nuclear Regulatory Commission Regulatory Issue Summary 2004-03, Risk-Informed Approach for Post-Fire Safe-Shutdown Circuit Inspections , Revision 1, December 29, 2004.
: 2. United States Code of Federal Regulations, Title 10, Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979. 3. United States Nuclear Regulatory Commission Generic Letter 86-10, Implementation of Fire Protection Requirements , April 24, 1986.
: 4. ANO Calculation CALC-85-E-0087-01, SSCA Safe Shutdown Capability Assessment , Revision 6.
: 5. ANO Calculation CALC-99-R-0002-01, Evaluation of High / Low Pressure Interface Valves with Respect to 10 CFR 50 Appendix R , Revision 1.
: 6. ANO Calculation CALC-85-E-0086-02, Manual Action Feasibility Methodology and Common Results , Revision 2.
: 7. Nuclear Energy Institute Technical Report NEI 04-06, Guidance for Self-Assessment of Circuit Failure Issues, Draft Revision L, March 2005.
: 8. United States Nuclear Regulatory Commission Inspection Procedure Attachment 71111.05T, Fire Protection (Triennial), April 21, 2006.
: 9. United States Nuclear Regulatory Commission Auxiliary and Power Conversion Systems Branch (APCSB) Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976, August 23, 1976.
: 10. ANO Calculation CALC-85-E-0087-24, Safe Shutdown Cable Analysis, Revision 0 Arkansas Nuclear One - Unit 2 Att. F - Fire-Induced MSOs Resolution Enclosure 1 to 2CAN121202 Page F-2 These information sources for the expert panel assessment included the post-fire safe shutdown analysis (Item 4), PRA insights (expert panel experience), and operating experience (Items 5 and 6 and expert panel experience). The expert panel meeting was also part of the NEI 04-06 assessment performed to address RIS 2004-03 (see Items 7 and 10). Thus, with the exception of a generic list of MSOs, the information sources recommended by FAQ 07-0038 were utilized for the expert panel assessment.
 
A PWROG generic list of MSOs was not yet available at the time of the expert panel meeting in 2005. However, the list of PWR generic MSOs from Revision 2 of NEI 00-01 was evaluated to ensure that applicable MSOs from this list have been included in the NSCA and FPRA model.
 
Step 2 - Conduct an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. 1 Section F.4.2).
The expert panel should focus on system and component interactions that could impact nuclear safety. This information will be used in later tasks to identify cables and potential locations where vulnerabilities could exist.
 
The documentation of the results of the expert panel should include how the expert panel was conducted including the members of the expert panel, their experience, education, and areas of expertise. The documentation should include the list of MSOs reviewed as well as the source for each MSO. This documentation should provide a list the MSOs that were included in the PRA and a separate list of MSOs that were not kept for further analysis (and the reasons for rejecting these MSOs for further analysis).
Describe the expert panel process (e.g., when it was held, what training was provided to the panel members, what analyses were reviewed to identify MSOs, how was consensus achieved on which MSOs to keep and any dispute resolution process criteria used in decision process, etc.). Note: The physical location of the cables of concern (e.g., fire zone/area routing of the identified MSO cables), if known, may be used at this step in the process to focus the scope of the detailed review in further steps.
Results of Step 2:
On September 27th and 28th, 2005, a panel of plant and industry personnel met at ANO to identify those combinations of spurious actuations which, if they occurred concurrently, could be risk significant.
 
A specific intent of assembling this panel was to ensure potential combinations that may not have been considered previously due to the plant's existing licensing basis would be identified. The safe shutdown analysis addresses spurious actuation on an any-and-all, one-at-a-time basis. Except for high-low pressure interfaces, concurrent spurious actuations had not been previously considered. In particular, the synergistic effects of concurrent failures in different systems may not have been considered.
 
Arkansas Nuclear One - Unit 2 Att. F - Fire-Induced MSOs Resolution Enclosure 1 to 2CAN121202 Page F-3 The panel discussion focused on potential transients that could adversely affect achievement and maintenance of the post-fire safe shutdown functions. Thus, the discussion focused on those fire-induced transients that would require operator action in the first hour after the fire and subsequent reactor trip, and those that could potentially damage equipment that may be required later, such as the credited LPSI pump used for shutdown cooling. The panel also considered whether the synergistic effects of concurrent spurious actuation in different systems serving different safe shutdown functions could adversely affect safe shutdown.
 
The alternate approach described in NEI 04-06 was used to ensure all potentially risk-significant combinations were evaluated for all fire areas. Since all potential spurious actuations of equipment identified as required for safe shutdown were in the process of being evaluated, the initial focus was placed on identifying those spurious actuations that, if they occurred concurrently, could result in an unrecoverable plant condition or lead to unrecoverable equipment damage. The safe shutdown equipment list had been validated and the circuit analysis validation had been completed at the time of the assessment.
 
The panel assembled at ANO focused on identifying those spurious actuations and combinations thereof that could be risk significant. An initial screening was performed by the panel based on the function affected, the potential consequences, and the time available to mitigate the potential transient. In this manner, spurious actuations and combinations of spurious actuations that did not require a mitigating action in the first hour after the reactor trip
 
were identified.
 
The expert panel included members with experience in electrical design engineering, mechanical design engineering, nuclear design engineering, system engineering, fire protection, safe-shutdown analysis, operations, reactor safety analysis, maintenance, probabilistic risk assessment, and accident management. A list of the expert panel members, including their education, experience and area(s) of expertise is contained in on-site documentation.
 
Step 3 - Update the Fire PRA model and NSCA to include the MSOs of concern.
This includes the:  Identification of equipment (NUREG/CR-6850, Task 2)  Identification of cables that, if damaged by fire, could result in the spurious operation (NUREG/CR-6850, Task 3, Task 9)  Identify routing of the cables identified above, including associating that routing with fire areas, fire zones and/or FPRA physical analysis units, as applicable.
 
Include the equipment/cables of concern in the NNSCA. Including the equipment and cable information in the NSCA does not necessarily imply that the interaction is possible since separation/protection may exist throughout the plant fire areas such that the interaction is not possible.
 
Note: Instances may exist where conditions associated with MSOs do not require update of the FPRA and NSCA analysis. For example, FPRA analysis in NUREG/CR-6850, Task 2, Component Selection, may determine that the particular interaction may not lead to core damage, or pre-existing equipment and cable routing information may determine that the particular MSO interaction is not physically possible. In other instances, the Arkansas Nuclear One - Unit 2 Att. F - Fire-Induced MSOs Resolution Enclosure 1 to 2CAN121202 Page F-4 update of the PRA may not be warranted if the contribution is negligible. The rationale for exclusion of identified MSOs from the FPRA and NSCA should be documented and the configuration control mechanisms should be reviewed to provide reasonable confidence that the exclusion basis will remain valid.
Results of Step 3:
The FPRA addresses spurious operations, including MSOs, identified in the post-fire safe shutdown analysis (SSA). These include those that resulted from the expert panel review and from review of the more recent PWR owners group (PWROG) generic list of MSOs (as applicable). The Fire PRA model includes a correlation of safe shutdown components to PRA basic events and a correlation of PRA basic events to safe shutdown components.
The NSCA includes equipment and cables of concern identified during the expert panel review and during review of the more recent PWROG generic list of MSOs (as applicable). As noted in FAQ 07-0038, including the equipment and cable information in the NSCA does not necessarily imply that the interaction is possible since separation or protection may exist throughout the plant fire areas such that the interaction is not possible.
The PWROG generic list of MSOs was not yet available at the time of the expert panel meeting in 2005. However, the list of PWR generic MSOs from Revision 2 of NEI 00-01 was evaluated to ensure that applicable MSOs from this list have been included in the NSCA and FPRA model.
Step 4 - Evaluate for NFPA 805 Compliance The MSO combinations included in the NSCA should be evaluated with respect to compliance with the deterministic requirements of NFPA 805, as discussed in Section 4.2.3 of NFPA 805.
For those situations in which the MSO combination does not meet the deterministic requirements of NFPA 805 (VFDR), the issue with the components and associated cables should be mitigated by other means (e.g., performance-based approach per Section 4.2.4 of NFPA 805, plant modification, etc.)
The performance-based approach may include the use of feasible and reliable recovery actions. The use of recovery actions to demonstrate the availability of a success path for the nuclear safety performance criteria requires that the additional risk presented by the use of these
 
recovery actions be evaluated (NFPA 805 Section 4.2.4).
 
Results of Step 4:
The MSO combinations included in the NSCA were evaluated with respect to compliance with the deterministic requirements of NFPA 805 Section 4.2.3, "Deterministic Approach."  For those situations in which the MSO combination did not meet the deterministic requirements of NFPA 805, the components and associated cables were added to the scope of the FREs performed for the associated fire area. Table B-2 describes the NSCA methods and Table B-3 provides the transition results for each fire area, indicating which areas required performance-
 
based analysis.
The performance-based analyses are described in Section 4.5 and the results are provided in
 
Attachment W.
Arkansas Nuclear One - Unit 2 Att. F - Fire-Induced MSOs Resolution Enclosure 1 to 2CAN121202 Page F-5 Step 5 - Document Results The results of the process should be documented. The results should provide a detailed description of the MSO identification, analysis, disposition, and evaluation results (e.g., references used to identify MSOs; the composition of the expert panel, the expert panel process, and the results of the expert panel process; disposition and evaluation results for each MSO, etc.). High level methodology utilized as part of the transition process should be included in the 10 CFR 50.48(c) License Amendment Request/Transition Report.
 
Results of Step 5:
The list of PWR generic MSOs from Revision 2 of NEI 00-01 was evaluated to ensure that applicable MSOs from this list have been included in the NSCA and FPRA model. This
 
evaluation is documented in CALC-ANO2-FP-09-0016, ANO-2 NFPA-805 Evaluation of Multiple Spurious Operations (MSOs).
Arkansas Nuclear One - Unit 2 Att. G - Recovery Action Transition Enclosure 1 to 2CAN121202 Page G-1 G. Recovery Actions Transition In accordance with the guidance provided in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205, the following methodology was used to determine recovery actions required for compliance (i.e., determining the population of post-transition recovery actions). The methodology consisted of the following steps: Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition. Step 2: Determine the population of recovery actions that are required to resolve 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.
An overview of these steps and the results of their implementation are provided below.
Step 1 - Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s)
The first task in the process of determining the post-transition population of recovery actions is to apply the NFPA 805 definition of recovery action and the RG 1.205 definition of primary control station to determine those activities that are taken at primary control station(s).
 
Results of Step 1:
 
Based on the definition provided in RG 1.205, and the additional guidance provided in FAQ 07-0030, no primary control stations were identified.
Step 2 - Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk or defense-in-depth criteria)
 
On a fire area basis all VFDRs were identified in the NEI 04-02 Table B-3 (See Attachment C). Each VFDR not brought into compliance with the deterministic approach was evaluated using the performance-based approach of NFPA 805, Section 4.2.4. The performance-based evaluations resulted in the need for recovery actions to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth.
Random failures that are not fire induced were addressed in the fire PRA, but are not considered recovery actions requiring approval as allowed by RG 1.205.
 
Results of Step 2:
 
The final set of recovery actions are provided in Table G Recovery Actions.
Arkansas Nuclear One - Unit 2 Att. G - Recovery Action Transition Enclosure 1 to 2CAN121202 Page G-2 Step 3 - Evaluation of the Additional Risk of the Use of Recovery Actions NFPA 805, Section 4.2.3.1, does not allow recovery actions when using the deterministic approach to meet the nuclear safety performance criteria. However, the use of recovery actions is allowed by NFPA 805 using a risk informed, performance-based, approach, provided that the additional risk presented by the recovery actions is evaluated in accordance with NFPA 805, Section 4.2.4.
 
Results of Step 3:
The set of recovery actions that are necessary to demonstrate the availability of a success path for the nuclear safety performance criteria (see Table G-1) were evaluated for additional risk using the process described in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205 and compared against the guidelines of RG 1.174 and RG 1.205. The additional risk is provided in Attachment W.
 
All of the recovery actions were reviewed for adverse impact and dispositioned. None of the recovery actions were found to have an adverse impact on the Fire PRA.
 
Step 4 - Evaluation of the Feasibility of Recovery Actions Recovery actions were evaluated against the feasibility criteria provided in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205.
 
Results of Step 4:
Each of the feasibility criteria in FAQ 07-0030 were assessed for the recovery actions listed in Table G-1. The results of the assessment are included in EC-27716, "ANO2 Fire Area Risk Evaluations for Transition to NFPA-805."
 
Feasibility is based in part on ANO-2 Functional Requirements included in CALC-85-E-0086-02, "Manual Action Feasibility and Common Results."
Implementation items resulting from the feasibility evaluation are included in the corrective action program. These items include:
Development/revision of procedures. Revisions to the Training Program to reflect procedure changes.
These items are included in Attachment S.
Step 5 - Evaluation of the Reliability of Recovery Actions The evaluation of the reliability of recovery actions depends upon its characterization. The reliability of recovery actions that were modeled specifically in the FPRA were addressed using FPRA methods (i.e., HRA).
Arkansas Nuclear One - Unit 2 Att. G - Recovery Action Transition Enclosure 1 to 2CAN121202 Page G-3 The reliability of recovery actions not modeled specifically in the FPRA is bounded by the treatment of additional risk associated with the applicable VFDR. In calculating the additional risk of the VFDR, the compliant case recovers the fire-induced failure(s) as if
 
the variant condition no longer exists. The resulting delta risk between the variant and compliant condition bounds any additional risk for the recovery action even if that recovery action were modeled.
Results of Step 5:
Specific recovery actions were added to the FPRA. For the bounding reliability treatment see results in Attachment W.
 
Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition Enclosure 1 to 2CAN121202 Page G-4 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS B-2 2A-309 2A-3 SUPPLY BREAKER De-energize DC Control Power to 2A-309 at 2A-3, 2A-309 located in Fire Area II, Fire Zone 2101-AA. Then verify open/ manually open 2A-309 in Fire Area II, Fire Zone 2101-AA.
B2-01 RA B-2 2P-32A REACTOR COOLANT PUMP Secure 2P-32A. B2-03 RA B-2 2P-32B REACTOR COOLANT PUMP Secure 2P-32B. B2-03 RA B-2 2P-32C REACTOR COOLANT PUMP Secure 2P-32C B2-03 RA B-2 2P-32D REACTOR COOLANT PUMP Secure 2P-32D. B2-03 RA B-3 2CV-0789-1 EFW PUMP 2P-7B CONDENSATE SUCTION VALVE De-energize 2CV-0789-1 at panel 2B-5, breaker 2B-514, located in Fire Area II, Fire Zone 2101-AA. Verify open/ manually open 2CV-0789-1 in Fire Area FF, Fire Zone 2025-JJ.
B3-01 RA B-4 2P-32A REACTOR COOLANT PUMP Manually trip 2P-32A at breaker 2H-11, local panel 2H-1, in Fire Area B-2, Fire Zone 2200-MM.
B4-01 RA B-4 2P-32B REACTOR COOLANT PUMP Manually trip 2P-32B at breaker 2H-21, local panel 2H-2, in Fire Area B-2, Fire Zone 2200-MM.
B4-01 RA B-4 2P-32C REACTOR COOLANT PUMP Manually trip 2P-32C at breaker 2H-22, local panel 2H-2, in Fire Area B-2, Fire Zone 2200-MM.
B4-01 RA B-4 2P-32D REACTOR COOLANT PUMP Manually trip 2P-32D at breaker 2H-12, local panel 2H-1, in Fire Area B-2, Fire Zone 2200-MM.
B4-01 RA EE-L 2CV-1075-1 2P-7B DISCHARGE TO SG-B Locally open EFW discharge valve following power failure. 2CV-1075-1 located in Fire Area GG, Fire Zone 2081-HH.
EE-L-01 RA EE-U 2CV-1026-2 2CV-1076-2 2P-7A DISCHARGE TO SG-A 2P-7A DISCHARGE TO SG-BLocally open EFW discharge valve following power failure. 2CV-1026-2 located in Fire Area EE-L, Fire Zone 2084-DD. 2CV-1076-2 located in Fire Area GG, Fire Zone 2081-HH.
EE-U-01 RA EE-U 2EFW-5A 2EFW-5B EFW PUMP DISCHARGE CROSSOVER ISOLATION Open Manual Valves 2EFW-5A&B to Divert 2P-7A Flow to B Discharge Line Following Fire Damage to 2P-7A motor operated valves (MOVs). 2EFW-5A located in Fire Area FF, Fire Zone 2025-JJ. 2EFW-5B located in Fire Area CC, Fire Zone 2024-JJ.
EE-U-01 RA Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition Enclosure 1 to 2CAN121202 Page G-5 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS G 2CV-4816 LETDOWN THROTTLE CV Locally vent air from the actuator for 2CV-4816 to fail the valve closed in Fire Area EE, Fire Zone 2084-DD.
G-02 RA G 2CV-4817 LETDOWN THROTTLE CV Locally vent air from the actuator for 2CV-4817 to fail the valve closed in Fire Area EE, Fire Zone 2084-DD.
G-02 RA G 2P-32A REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32A at 2H-11 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/ manually trip 2H-11 in Fire Area B-2, Fire Zone 2200-MM.
G-02 RA G 2P-32B REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32B at 2H-21 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/ manually trip 2H-21 in Fire Area B-2, Fire Zone 2200-MM.
G-02 RA G 2P-32C REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32C at 2H-22 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/ manually trip 2H-22 in Fire Area B-2, Fire Zone 2200-MM.
G-02 RA G 2P-32D REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32D at 2H-12 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/ manually trip 2H-12 in Fire Area B-2, Fire Zone 2200-MM.
G-02 RA G NEW AFW PUMP NEW AUXILIARY FEEDWATER (AFW) PUMP Manually start and align 2P-75B AFW pump locally to establish primary to secondary heat removal.
G-03 G-04 G-05 RA G 2CV-1036-2 2CV-1039-1 2CV-1075-1 2CV-1076-2 2P-7B DISCHARGE TO SG-B 2P-7A DISCHARGE TO SG-B 2P-7B DISCHARGE TO SG-B 2P-7A DISCHARGE TO SG-BLocally open Emergency Feedwater discharge valves following power failure.
G-03 RA G 2P-89B HIGH PRESSURE SAFETY INJECTION (HPSI) PUMP Locally open breaker 2A-406 to prevent start of HPSI pump. G-02 RA*
G 2P-60B LOW PRESSURE SAFETY INJECTION (LPSI) PUMP Locally open breaker 2A-405 to prevent start of LPSI pump. N/A RA*
G 2P-35B CONTAINMENT SPRAY PUMP Locally open breaker 2A-404 to prevent start of Containment Spray Pump.
G-02 RA*
Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition Enclosure 1 to 2CAN121202 Page G-6 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS G 2CV-5630-1 2CV-5631-2 RWT OUTLET VALVES Close both RWT outlet valves locally. G-02 RA*
G 2CV-4920-1 2CV-4921-1 BORIC ACID MAKEUP TANK (BAMT) GRAVITY FEED VALVES Open both BAMT Gravity Feed valves locally. G-02 RA* G 2CV-4873-1 VOLUME CONTROL TANK (VCT) OUTLET VALVE Close VCT outlet valve locally. N/A RA* G 2T-1 PRESSURIZER HEATERS Turn OFF and operate pressurizer heaters as necessary. G-01 RA* G 2P-36A/B/C CHARGING PUMPS Stop and operate Charging pumps locally as needed. G-02 RA*
G 2K-4B EMERGENCY DIESEL GENERATOR #2 (EDG #2) Place EDG #2 in LOCKOUT locally. G-04 RA*
G 2A-4 4160V VITAL POWER SWITCHGEAR De-energize 2A-4 locally to prevent spurious operation. Energize 2A-4 to restore power to vital equipment.
G-04 RA* G 2B-6 480V VITAL POWER LOAD CENTER De-energize 2B-6 locally. G-04 RA*
G 2D-24-2 2D-24-4 2D-24-6 2D-24-8 2D-24-9 2D-24-10 2C108 POWER SUPPLY 2A4 DC CONTROL POWER 2E21 POWER SUPPLY 2C108 POWER SUPPLY 2B6 DC CONTROL POWER HI POINT VENT PNL 2C336-2Open breakers to remove DC power to various equipment.
G-02 G-04 RA* G 2CV-4840-2 CHARGING HEADER ISOLATION Locally verify open Charging header isolation. G-02 RA* G 2CV-1504-2 EDG #2 SERVICE WATER OUTLET Locally verify open EDG #2 Service Water outlet. G-05 RA* G 2CV-4950-2 RWT SUCTION VALVE Verify RWT suction valve open for Charging capability if necessary.
G-02 RA* G 2P-4C SERVICE WATER PUMP Align Loop 2 Service Water header locally. G-05 RA*
Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition Enclosure 1 to 2CAN121202 Page G-7 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS G 2CV-0795-2 2P-7A SUCTION MOV Verify open 2P-7A Condensate suction MOV locally. G-03 RA* G 2VUC-30 SPDS COMPUTER ROOM COOLER Transfer SPDS Room Cooler 2VUC-30 to ANO-1 power source if required.
G-06 RA* GG 2CV-0789-1 EFW PUMP 2P-7B CONDENSATE SUCTION De-energize 2CV-0789-1 at panel 2B-53, breaker 2B-53D2, located in Fire Area B-3, Fire Zone 2091-BB. Verify open/ manually open 2CV-0789-1 in Fire Area FF, Fire Zone 2025-JJ.
GG-02 RA JJ 2A-113 2A-1 SUPPLY BREAKER De-energize DC Control Power to 2A-113 at 2A-1 located in Fire Area B-2, Fire Zone 2200-MM. Verify closed/manually close 2A-113 in Fire Area B-2, Fire Zone 2200-MM.
JJ-04 RA JJ 2A-309 2A-3 SUPPLY BREAKER De-energize DC Control Power to 2A-309 at 2A-3 located in Fire Area II, Fire Zone 2101-AA. Verify closed/manually close 2A-309 in Fire Area II, Fire Zone 2101-AA.
JJ-04 RA JJ 2CV-1036-2 2CV-1075-1 2P-7B DISCHARGE TO SG-BLocally open EFW discharge valve following power failure. 2CV-1075-1 and 2CV-1036-2 lo cated in Fire Area GG, Fire Zone 2081-HH.
JJ-01 RA JJ 2CV-5649-1 CONTAINMENT SUMP SUCTION ISOLATION De-energize 2CV-5649-1 at panel 2B-52, breaker 2B-52G3, located in Fire Area DD, Fire Zone 2040-JJ. Verify open/ manually open 2CV-5649-1 in Fire Area B-6, Fire Zone 2014-LL (action performed in conjunction with 2CV-5650-2).
JJ-02 RA JJ 2CV-5650-2 CONTAINMENT SUMP SUCTION ISOLATION De-energize 2CV-5650-2 at panel 2B-62, breaker 2B-62G3, located in Fire Area HH, Fire Zone 2073-DD. Verify open/ manually open 2CV-5650-2 in Fire Area AA, Fire Zone 2007-LL (action performed in conjunction with 2CV-5649-1).
JJ-02 RA JJ 2PIS-0789 EFW PUMP 2P-7B CONDENSATE SUCTION De-energize and manually open 2CV-0789-1 in Fire Area FF, Fire Zone 2025-JJ, prior to starting an EFW pump.
JJ-01 RA JJ 2P-7B EFW PUMP Manually start 2P-7B at switchgear 2A3, breaker 2A-311 located in Fire Area II, Fire Zone 2101-AA.
JJ-01 RA JJ 2P-89 HPSI PUMP Locally close minimum flow recirculation valve for the HPSI pumps 2CV-5628-2 in Fire Area DD, Fire Zone 2040-JJ.
JJ-02 RA Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition Enclosure 1 to 2CAN121202 Page G-8 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS MM 2A-113 2A-213 4.16KV SWITCHGEAR At the switchgear, align offsite power to bus 2A-1 and 2A-2. MM-01 RA MM 2CV-1025-1 2CV-1036-2 2CV-1038-2 2CV-1075-1 2P-7B DISCHARGE TO SG-A 2P-7B DISCHARGE TO SG-B 2P-7B DISCHARGE TO SG-A 2P-7B DISCHARGE TO SG-BLocally open EFW discharge valves following fire induced control and power failure. 2CV-1025-1 and 2CV-1038-2 located in Fire Area EE-L, Fire Zone 2084-DD. 2CV-1075-1 and 2CV-1036-2 lo cated in Fire Area GG, Fire Zone 2081-HH.
MM-01 RA OO 2CV-1470-1 SERVICE WATER (SW) TO 2P-4A De-energize 2CV-1470-1 at panel 2B-54, breaker 2B-54E4, located in Fire Area II, Fire Zone 2101-AA. Verify open/ manually open 2CV-1470-1 in Fire Area OO, Fire Zone INTAKE.Note: Valve operator is installed external to the intake structure and not in the impacted area.
OO-01 RA OO 2CV-1474-2 SW TO 2P-4C De-energize 2CV-1474-2 at panel 2B-62, breaker 2B-62H3, located in Fire Area HH, Fire Zone 2073-DD. Verify open/ manually open 2CV-1474-2 in Fire Area OO, Fire Zone INTAKE.Note: Valve operator is installed external to the intake structure and not in the impacted area.
OO-01 RA SS 2CV-1038-2 EFW FROM 2P-7B TO SG-A ISOLATION De-energize 2CV-1038-2 at panel 2B-63, breaker 2B-63H3, located in Fire Area HH, Fire Zone 2096-M. Verify open/ manually open 2CV-1038-2 in Fire Area EE-L, Fire Zone 2084-DD.
SS-01 RA SS 2CV-1425-1 AUXILIARY COOLING WATER (ACW) ISOLATION De-energize 2CV-1425-1 at panel 2B-54, breaker 2B-54D5, located in Fire Area II, Fire Zone 2101-AA. Verify closed/ manually close 2CV-1425-1 in Fire Area OO, Fire Zone INTAKE.
SS-05 RA SS 2CV-1470-1 SW TO 2P-4A De-energize 2CV-1470-1 at panel 2B-54, breaker 2B-54E4, located in Fire Area II, Fire Zone 2101-AA. Verify open/ manually open 2CV-1470-1 in Fire Area OO, Fire Zone INTAKE.
SS-05 RA SS 2EFW-802 2P-7A/B SUCTION FROM 2T-41A/B Align EFW/AFW suction to QCST T-41B on low-low level in CST aligned to EFW/AFW (2T-41A or 2T-41B).
SS-01 RA Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition Enclosure 1 to 2CAN121202 Page G-9 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS SS 2CV-1026-2 2CV-1037-1 2CV-1039-1 2CV-1076-2 2P-7A DISCHARGE TO SG-A 2P-7A DISCHARGE TO SG-A 2P-7A DISCHARGE TO SG-B 2P-7A DISCHARGE TO SG-BAlign DC operated valves prior to battery discharge. 2CV-1037-1 and 2CV-1026-2 located in Fire Area EE-L, Fire Zone 2084-DD. 2CV-1039-1 and 2CV-1076-2 located in Fire Area GG, Fire Zone 2081-HH.
N/A RA TT 2A-309 2A-3 SUPPLY BREAKER De-energize DC Control Power to 2A-309 at 2A-3 located in Fire Area II, Fire Zone 2101-AA. Verify closed/manually close 2A-309 in Fire Area II, Fire Zone 2101-AA. TT-01 RA TT 2CV-1036-2 EFW FROM 2P-7B TO SG-B ISOLATION De-energize 2CV-1036-2 at panel 2B-63, breaker 2B-63H1, located in Fire Area HH, Fire Zone 2096-M. Verify open/ manually open 2CV-1036-2 in Fire Area GG, Fire Zone 2081-HH.TT-01 RA TT 2CV-1075-1 EFW FROM 2P-7B TO SG-B FLOW CONTROL VALVE De-energize 2CV-1075-1 at panel 2B-53, breaker 2B-53J2, located in Fire Area B-3, Fire Zone 2091-BB. Verify open/
manually open 2CV-1075-1 in Fire Area GG, Fire Zone 2081-HH.TT-01 RA  RA - Recovery Action RA* - Defense in Depth Measure PCS - Primary Control Station
 
Arkansas Nuclear One - Unit 2 Att. H - NEI 04-02 FAQs Summary Table Enclosure 1 to 2CAN121202 Page H-1 H. NFPA 805 Frequently Asked Question Summary Table This table includes the approved FAQs that have not been incorporated into the current endorsed revision of NEI 04-02 and utilized in this submittal:
Table H NEI 04-02 FAQs Utilized in LAR Submittal No. Rev. Title FAQ Ref. Closure Memo 06-0008 9 NFPA 805 Fire Protection Engineering Evaluations ML090560170 ML073380976 06-0022 3 Acceptable Electrical Cable Construction Tests ML090830220 ML091240278 07-0030 5 Establishing Recovery Actions ML103090602 ML110070485 07-0032 2 Clarification of 10 CFR 50.48(c), 10 CFR 50.48(a) and GDC 3 Clarification ML081300697 ML081400292 07-0035 2 Bus Duct Counting Guidance for High Energy Arcing Faults ML091610189 ML091620572 07-0038 3 Lessons learned on Multiple Spurious Operations ML103090608 ML110140242 07-0039 2 Lessons Learned - NEI B-2 Table ML091420138 ML091320068 07-0040 4 Non-Power Operations Clarification ML082070249 ML082200528 07-0042 0 Fire Propagation from Electrical Cabinets ML080230438 ML091460350 ML092110537 07-0054* 1 Demonstrating Compliance with Chapter 4 of NFPA 805 ML103510379 ML110140183 08-0043 1 Electrical Cabinet Fire Location ML083540152 ML091470266 ML092120448 08-0044 0 Large Oil Fires ML081200099 ML091540179 ML092110516 08-0046 0 Incipient Fire Detection Systems ML081200120 ML093220197 ML093220426 08-0047 1 Spurious Operation Probability ML082770662 ML082950750 08-0048 0 Revised Fire Ignition Frequencies ML081200291 ML092180383 ML092190457 08-0049 0 Cable Fires ML081200309 ML091470242 ML092100274 08-0050 0 Non Suppression Probability ML081200318 ML092510044 ML092190555 08-0052 0 Transient Fire Growth Rate and Control Room Non-Suppression ML081500500 ML091590505 ML092120501 08-0053 0 Kerite-FR Cable Failure Thresholds ML082660021 ML120060267 09-0056 2 Radioactive Release Transition ML102810600 ML102920405 10-0059 5 NFPA 805 Monitoring ML111180481 ML120750108 12-0062 1 UFSAR Content ML121430035 ML121160046
* The FAQ submittal number was 08-0054, but the NRC closure memo for the FAQ was listed as 07-0054. FAQ 07-0054 was used to be consistent with the Closure Memo.
Arkansas Nuclear One - Unit 2 Att. I - Definition of Power Block Enclosure 1 to 2CAN121202 Page I-1 I. Definition of Power Block The methodology of the review process is discussed in Section 4.1.3 of this enclosure. For the purposes of establishing the structures included in the FPP in accordance with 10 CFR 50.48(c) and NFPA 805, plant structures listed in the following table are considered to be part of the
 
power block.
 
Table I Power Block Definition Power Block Structures Fire Area(s) Auxiliary Building Various (Refer to FHA) Containment Building NN Electrical Manholes 2MH01 through 03 (E & W) Emergency Diesel Fuel Oil Storage Tank Vault L Intake Structure OO Radwaste Storage Buildings 1 YD Turbine Building B-2 (Fire Zone 2200-MM) 1 The Radwaste Storage Buildings include Warehouse #19 (Old Radwaste Storage Building), the Radiation Protection (RP) Storage Building (Pole Barn and Mockup Area), and the Low Level Radwaste Building.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-1 J. Fire Modeling Verification and Validation (V&V)
This attachment documents the Verification and Validation (V&V) basis for the ANO-2 Fire Probabilistic Risk Assessment (FPRA) fire modeling applications. Plant specific fire modeling used to support the ANO-2 FPRA consists of the following:
The calculation of the Main Control Room (MCR) operator abandonment times (ANO CALC ANO2-FP-09-0013);  The use of generic fire modeling treatments and its associated supplements, as applicable, to develop Zones of Influence (ZOI) (PRA-A2-05-003);  A detailed assessment of plant specific fire scenarios involving secondary cable tray combustibles (1JMW21022.002-06); and  An assessment of the fire resistance of embedded conduit used as a basis for excluding such conduit from fire zones (EC-494).
Main Control Room Abandonment Report The goal of the MCR abandonment report titled "Evaluation of Unit 2 Control Room Abandonment Times at the Arkansas Nuclear O ne Facility" (ANO CALC ANO2-FP-09-00013) is to compute the time operators would abandon the ANO-2 MCR given a fire in either the ANO-1 or ANO-2 MCR. The abandonment times are assessed for various electronic equipment fires and for ordinary combustible fires as defined by the discretized heat release rate conditional probability distributions presented in NUREG/CR-6850. The abandonment time in the MCR is estimated by calculating the time to reach threshold values for temperature and visibility, as identified by NUREG/CR-6850.
 
The focus of the MCR abandonment evaluation is on the first twenty-five minutes after ignition because the non-suppression probability (NSP) decreases to 0.001 at twenty minutes (NUREG/CR-6850, NUREG/CR-6850, Supplement 1). The abandonment calculations are performed using the zone fire model Consolidated Fire and Smoke Transport (CFAST),
Version 6.0.10 (National Institute of Standards and Technology (NIST) Special Publication (SP) 1026 and NIST SP 1041).
 
The MCR area geometry and fire parameters for t he simulations fall within the model limits listed in NIST SP 1026 and NIST SP 1041. Specifically, the vent area to enclosure volume ratio is less than two and the aspect ratio of the enclosures is less than five (for the true geometry).
The physical input dimensions are adjusted to account for obstructions and boundary heat losses and the resulting model geometry has a length-to-width aspect ratio greater than five for some spaces. However, the input geometry conserves the boundary area, room volume, and enclosure height. Therefore, a corridor flow model is intentionally avoided because the true geometry has an aspect ratio that is within the model limitations.
 
The verification for the CFAST model (Version 6.0.5) is provided in NUREG-1824, Volume 5. Supplemental verification for CFAST, Version 6.0.10, is provided as an appendix to the CALC-ANO2-FP-09-00013 report as well as in NIST SP 1086.
 
The non-dimensional parameters that affect the model results, as documented in NUREG-1824, Volumes 1 and 5, and NUREG-1934, include the model geometry, the global equivalence ratio, the fire Froude Number, and the flame length ratio. Non-dimensional parameters that relate to target exposure conditions (heat flux) and sprinkler actuation (ceiling jet) are not applicable to this calculation because these output parameters are not used.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-2 The non-dimensional geometry parameters (length-to-height and width-to-height, which range from about 1.5 - 5 for the true geometry, depending on whether the fire is in the operator area or the equipment area) fall within the NUREG-1824, Volume 1 validation range (0.6 - 5.7). CFAST, Version 6.0.10, does not use a fire diameter [NIST SP 1026, NIST SP 1041]; therefore, it is possible to specify a fire that falls within the range of fire Froude Numbers considered in the NUREG-1824, Volume 1, validation documentation. Nevertheless, the source fires considered in CALC-ANO2-FP-09-00013 are consistent with those described in NUREG/CR-6850 as well as those used in the NUREG/CR-4527, Volume 2 (control room fire tests), and are considered to be consistent with the NUREG-1824, Volume 1, validation effort.
The global equivalence ratio applicable to the entire ANO-2 MCR domain (equipment area and the operator area) for normal Heating, Ventilation, and Air-Conditioning (HVAC) conditions may be assessed using the ratio of the maximum supported fire size to the fire size postulated.
Based on the fresh air supply flow of 0.62 m 3/s (1,300 cfm), the maximum fire size that could be supported is about 2.2 MW (2,120 Btu/s). The maximum fire size postulated is about 4 MW (3,790 Btu/s) for the workstation fire scenario; thus, the maximum global equivalence ratio is expected to be about 1.8, which exceeds the NUREG-1824, Volume 1, validation range of 0.04 - 0.6. However, the maximum average heat release rate, which better reflects the oxygen consumption that would be expected over the twenty-five minute interval, is about 2.4 MW (2,270 Btu/s). This means the maximum global equivalence ratio is expected to be on the order of 1.1, which still exceeds the NUREG-1824, Volume 1, validation range of 0.04 - 0.6. When the initial oxygen reservoir in the gross ANO-2 MCR volume (820 m 3 (28,960 ft 3)) is considered, it can be shown that this oxygen reservoir is capable of supporting a 1,310 kW (1,240 Btu/s) source fire for twenty-five minutes at a global equivalence ratio of 0.6. This means that the maximum global equivalence ratio decreases to 0.495 and thus falls within the range considered by NUREG-1824, Volume 1. The conditions for the smoke purge mode involve a fresh air supply that is 4.5 times greater than the normal HVAC mode; thus, the maximum equivalence ratio during this HVAC mode is on the order of 0.12.
 
In the case of no forced ventilation, the maximum global equivalence ratio is determined using the initial mass of oxygen available. The bounding case with respect to the NUREG-1824, Volume 1, validation space is when the boundary doors remain closed. The initial oxygen reservoir can support a 1,310 KW (1,240 Btu/s) fire for twenty-five minutes at an equivalence ratio of 0.6. In the case of the large (Bin 15) transient fuel package fire, abandonment is predicted in 2.8 minutes; thus, the equivalence ratio at this time would be 0.067, which falls within the range considered in NUREG-1824, Volume 1. Similarly, in the case of the multiple cable bundle electrical panel fire (Bin 15), abandonment is predicted in 5 minutes. The average fire size at this time is about 300 KW (284 Btu/s), and the global equivalence ratio is about 0.6(300/1310)(5/25) = 0.027, which falls slightly below the minimum value of 0.04 considered in NUREG-1824, Volume 1. This means that there is a sufficient supply of oxygen available for the fire up to the time at which abandonment is predicted. Further, all scenarios begin with an equivalence ratio of 0.0, including those that form the NUREG 1824, Volume 1, validation basis, thus the scenarios evaluated in the ANO-2 MCR are not inconsistent with the validation scenarios considered in NUREG-1824, Volume 1, si mply because the equivalence ratio is low. Given that this is the most adverse electrical panel fire scenario postulated, the global equivalence ratio at the predicted abandonment time is expected to be comparable or lower for the less severe electrical panel fire scenarios and the transient fire scenarios. Consequently, even when the HVAC is inoperative and the boundary doors are closed, the maximum global equivalence ratio within the MCR domain is expected to remain within the NUREG-1824, Volume 1, validation range up until the time at which abandonment is predicted.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-3 Finally, the flame length ratio is normally met, but in the case of the largest fire sizes postulated, the flame height may reach or exceed the ceiling height. Because sprinkler actuation and thermal radiation to targets are not computed with the CFAST model, this parameter is not an applicable metric. Rather, the plume entrainment below the hot gas layer controls the layer descent time and the concentration of soot products in the layer. This aspect of the model is not affected by the flame height to ceiling height ratio. Consequently, the application of CFAST to model fires in the ANO-2 control room falls entirely within the NUREG-1824, Volume 1, validation space.
 
Additional V&V studies are contained in NIST SP 1086 and NRL/MR/6180-04-8746. These studies have a broader parameter validation space than NUREG-1824, Volume 1. NIST SP 1086 is based in part on the methods of American Society for Testing and Materials (ASTM)
E1355. NRL/MR/6180-04-8746 provides a Navy specific V&V study, which includes an assessment of CFAST, Version 3.1.7, predictions in multiple enclosures and multiple elevation configurations. These additional studies extend the range of the validation space to include configurations and conditions applicable to the MCR abandonment sensitivity analysis (Appendix B of CALC-ANO2-FP-09-00013).
 
The MCR abandonment report also provides benchmark and validation simulations for CFAST as applicable to the ANO-2 MCR area. In particular, the control room tests documented in NUREG/CR-4527, Volume 2, are used to provide additional validation basis for control room application of CFAST. Table J-1 provides a summary of the validation and verification basis for CFAST, Version 6.0.10, as applied in the MCR abandonment report.
 
Generic Fire Modeling Treatments The Generic Fire Modeling Treatments (Hughes Associates) document is used to establish zones of influence for specific classes of ignition sources and primarily serves as a screening calculation in the FPRA under NUREG/CR-6850, Sections 8 and 11. The "Generic Fire
 
Modeling Treatments" document has two fundamental uses within the FPRA:
Determine the ZOI inside which a particular ignition source is postulated to damage targets or ignite secondary combustible materials; and Determine the potential of the ignition sources to generate a hot gas layer within an enclosure that can either lead to full room burnout or invalidate the generic treatment ZOIs for a particular class of combustible materials.
 
The ZOI is determined using a collection of empirical and algebraic models and correlations.
The potential for a hot gas layer having a specified temperature to form within an enclosure is determined using the zone model CFAST, Version 6.0.10 (NIST SP 1026, NIST SP 1041).
Verification The calculation development and review process in place at the time the Generic Fire Modeling Treatments document was prepared included contributions from a calculation preparer, a calculation reviewer, and a calculation approver. The responsibilities for each are as follows:
 
The calculation preparer develops and prepares the calculation using appropriate methods.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-4 The calculation reviewer provides a detailed review of the report and supporting calculations, including spreadsheets and fire model input files. The reviewer provides comments to the preparer for resolution.
The calculation approver provides a reasonableness review of the report and approves the document for release.
The calculation preparation occurred over a two year period ending in 2007. The review stage was conducted in 2007 at the completion of the preparation stage. The calculation was approved January 23, 2008. The approved document, the signature page, and an affidavit were transmitted to the NRC Document Control Desk in Washington, D. C. on January 23, 2008.
 
In the case of the empirical equations/correlations that form part of the basis of the Generic Fire Modeling Treatments document, a considerable amount of verification was performed during the preparation stage by the preparer. The empirical equations/correlations were solved with
 
Excel TM spreadsheets using either direct cell solutions (algebraic manipulation) or Visual Basic macros. All direct cell solutions were validated by the preparer through the use of alternate calculation. For simple equations, this entailed matching spreadsheet solution to the solution obtained using a hand calculator. For more complex solutions, the alternate calculation verification entailed either subdividing the problem into many sub-components and matching the solution using a hand calculator, or matching the solution to a verified solution (i.e., the NUREG-1805 solid flame heat flux models). The veri fication of the Visual Basic macros also depended on the type of macro. In situations where the macro is used to perform multiple direct computations, the macro results were verified against the verified spreadsheet solutions that were verified through alternate calculation. In cases where the macro is used to find a root, the root was verified to be a zero by direct substitution into an alternate form of the solved equation.
 
The empirical equations/correlations were further verified by the reviewer using a Design Review method as indicated in the signature sheet. An independent reviewer was provided access to the draft report and all supporting calculation materials in late 2007. The reviewer conducted a detailed review of the implementation of the equations within the spreadsheets and the reporting of the equation result in the draft report. Comments and insights were provided to the preparer over the review period and were addressed to the satisfaction of the reviewer.
Upon the completion of the review, a revised draft was prepared for review by the approver. The approver provided a higher level reasonableness check of the methods, approach, and the results. Comments and insights that were provided by the approver were addressed to the satisfaction of the reviewer and Revision 0 of the report was prepared and approved on
 
January 23, 2008.
 
The verification for the CFAST model (Version 6.0.5) is provided in NUREG-1824, Volume 5. Supplemental verification for CFAST, Version 6.0.10, is provided as an appendix to the CALC-ANO2-FP-09-00013 report as well as in NIST SP 1086.
 
Validation
 
The empirical equations and correlations are drawn from a variety of sources that are documented in various chapters of the Society of Fire Protection Engineers (SFPE)
Handbook of Fire Protection Engineering, peer reviewed journals (e.g., the Fire Safety Journal
), or engineering textbooks. The empirical models primarily fall into three groups:
 
Flame height
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-5 Plume temperatures Heat fluxes (at a target location)
Table J-2 of this attachment identifies the empirical models that are used either directly or indirectly in the Generic Fire Modeling Treatments report. The table also identifies the original correlation source documentation and the correlation range in terms of non-dimensional parameters. The table also provides where applicable supplemental validation work that may have been performed on the correlations and provides limits applied in the Generic Fire Modeling Treatments report as applicable.
 
Except for the cable tray ZOI calculation, the flame height calculation is used only as a means of placing a limit on the applicability of the ZOI tables, which are based on the plume temperature and thermal radiation heat flux. The flame height calculation for axisymmetric source fires is robust and has considerable pedigree. The original documentation and basis of the flame height correlation is Heskestad (1981) as noted in Table J-2 of this attachment. Although there are earlier forms of the flame height equation, Heskestad provides a link between the flame height and plume centerline temperature calculation and identifies the range over which the plume equations are applicable. Because the flame height and plume centerline temperature equations are linked, the plume centerline range cited by Heskestad applies to the flame height calculation as well. The plume centerline temperature equations, and thus the flame height correlation, are applicable over the following range as noted in Table J-2 (Heskestad, 1981; Heskestad, 1984):
 
Where c p is the heat capacity of ambient air (kJ/kg-K [Btu/lb-&deg;R]), T is the ambient temperature (K [&deg;R]), g is the acceleration of gravity (m/s 2 [ft/s 2]), p is the ambient air density (kg/m 3 [lb/ft 3]), Q is the fire heat release rate (KW [Btu/s]), r is the stoichiometric fuel to air mass ratio, D is the fire diameter (m [ft]), and H c is the heat of combustion of the fuel (kJ/kg [Btu/lb]). Application of Equation (J-1) depends on the fuel as well as a non-dimensional form of the fire heat release rate (fire Froude Number). In practice, the heat of combustion to air fuel ratio for most fuels will fall between 2,900 - 3,200 kJ/kg (1,250 - 1,380 Btu/lb), and for typical ambient conditions the 
 
Q 2/5/D ratio for which the plume equations have validation basis is between 7 - 700 KW 2/5/m (2.1 - 208 Btu 2/5/s2/5-ft) [Heskestad, 1984]. For fire sizes on the order of 25 KW (24 Btu/s) or greater, this means that the plume centerline equation is valid for heat release rates of 100 KW/m 2 (8.81 Btu/s-ft
: 2) to well over 3,000 KW/m 2 (264 Btu/s-ft 2). For weaker fires (heat release rates less than 100 KW/m 2 [8.81 Btu/s-ft 2]), the tendency of the model is clearly to over-predict the temperature and flame height; thus for applications outside the range but below the lower limit, the result will be conservative. The concern is, therefore, entirely on the upper range of the empirical model. The tables in the Generic Fire Modeling Treatments document are specifically developed with transient, lubricant spill fires, and electrical panel fires with a heat release rate per unit area within the validation range. When the heat release rate per unit area falls outside the applicable range, the table entry is not provided and it is noted that the source heat release rate per unit area is greater than the applicable range for the correlations.
This applies to the flame height and the plume temperature for axisymmetric source fires.
. . ~-5 < log 10 c p Tg(H c/r)3 Q 2 D 5~< 5.~-5 < log 10 c p Tg(H c/r)3 Q 2 D 5~< 5.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-6 The flame height and plume centerline temperature for line type fires (fires having a large aspect ratio) are applied only to cable tray fires. The correlation used has pedigree and has existed in its general form since at least Yokoi. Most recently, Yuan et al. provided a basis for the empirical constant using experimental data with source fires having a width of 0.015 m - 0.05 m (0.05 - 0.16 ft) and a length of 0.2 - 0.5 m (0.7 - 1.64 ft). When normalized, the applicable height to heat release rate per unit length range Z/Q' for the correlations based on the experiments of Yuan et al. is between 0.002 and 0.6. This range includes the flame height as well as the elevation at which the temperature is between 204 - 329&deg;C (400 - 625&deg;F), the temperature at which cable targets are considered to be damaged under steady state exposure conditions. Yuan et al. also provide a tabular comparison of the empirical constant against seven preceding line fire test series, which include a broader range of physical fire sizes and dimensions. The Yuan et al. constant is greater than the other seven and thus the temperatures and flame heights are more conservatively predicted using the Yuan et al. data. The application of the Yuan et al. correlation in the Generic Fire Modeling Treatments document falls within the normalized applicability range reported by Yuan et al.
 
Four flame heat flux models are used in the Generic Fire Modeling Treatments document as described in Table J-2 of this attachment:  the Point Source Model, the Simple Method of Shokri and Beyler, the Method of Mudan and Croce, and the Detailed Method of Shokri and Beyler. The former two are simple algebraic models using the heat release rate, separation distance, and the fire diameter. The latter two are considered detailed radiant models that account for the emissivity of the fire and the shape of the flame. Due to limitations in the target placement, the (Simple) Method of Shokri and Beyler are shown to be inapplicable for calculating the ZOI dimensions. Similarly, for the fuels considered, it is shown that the Method of Mudan and Croce produces a net heat flux that exceeds the fire size. The ZOIs are, therefore, determined using the Point Source Model and the Detailed Method of Shokri and Beyler. The method that produces the largest ZOI dimension is used for each fuel and fire size bin.
 
The Point Source Model and the Method of Shokri and Beyler have been shown in the NUREG-1824, Volume 3, verification and vali dation study to provide reasonably accurate predictions when the target separation to fire diameter R/D f ratio is between 2.2 and 5.7 (NUREG-1824, Volume 1). Furthermore, the fire size ranges considered in the Generic Fire Modeling Treatments report are between about 25 - 12,000 KW (24 - 11,400 Btu/s) and the heat release rates per unit area range between about 100 - 3,000 KW/m 2 (8.1 - 264 Btu/s-ft
: 2) for all fuels and fire size bins.
 
Using this information, the following table may be assembled for the applicable target heat flux range, based on the NUREG-1824, Volume 1, validation range:
 
Fire Size KW (Btu/s) Heat Release Rate Per Unit Area, KW/m 2 (Btu/s-ft
: 2) Fire Diameter, m (ft) Point Source Model Heat Flux Range, KW/m 2 (Btu/s-ft
: 2) Shokri and Beyler Heat Flux Range, KW/m 2 (Btu/s-ft
: 2) 25 (24) 100 (8.8) 0.56 (1.9) 0.07 - 0.45 (0.006 - 0.04) 0.36 - 3.8 (0.03 - 0.4) 25 (24) 3,000 (264) 0.1 (0.3) 2 - 13.6 (0.2 - 1.2) 2.84 - 10 (0.3 - 0.9) 12,000 (11,400) 100 (8.8) 12.4 (41) 0.07 - 0.45 (0.006 - 0.04) 0.55 - 5 (0.05 - 0.4) 12,000 (11,400) 3,000 (264) 2.3 (7.4) 2 - 13.6 (0.2 - 1.2) 0.45 - 4.6 (0.04 - 0.4)
.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-7 The threshold heat fluxes that define the steady state ZOI dimensions range from 5.7 - 11.4 KW/m 2 (0.5 - 1 Btu/s-ft 2). Transient ZOI dimensions, addressed in the Supplemental Generic Fire Modeling Treatments: Transient Fuel Package Ignition Source Characteristics report may approach 16 - 18 KW/m 2 (1.4 - 1.6 Btu/s-ft 2). Clearly, the steady state ZOI dimensions based on critical heat fluxes of 5.7 - 11.4 KW/m 2 (0.5 - 1 Btu/s-ft
: 2) overlay with the range of valid predicted heat fluxes identified in NUREG-1824, Volume 1. Fuels that identify the most conservative value over a range of heat release rates per unit area (transient and electrical panels) will thus include at least one point within the validation range (i.e., 5.7 KW/m 2 [0.5 Btu/s-ft 2]). Since the algorithm searches for the most adverse value, the result will be at least as conservative as the value obtained within the model validation range.
There are combinations of fuels and source strength ranges that do not produce heat fluxes that fall within the validation range. This is especially true for the higher target heat flux values
 
(11.4 KW/m 2 [1 Btu/s-ft 2] and higher) combined with the lower transient fuel package heat release per unit area range (200 - 1,000 KW/m 2 [17.6 - 88.1 Btu/s-ft 2]). This is addressed through an extended validation range of the heat flux models provided by the SFPE (1999). As noted in Table J-2 of this attachment, the SFPE assessed the predictive capabilities of the Point Source Model and the Detailed Method of Shokri and Beyler against available pool fire data. The pool diameters ranged from 1 - 80 m (3.3 - 262 ft). The conclusion was that the Point Source Model was conservative, but not necessarily bounding, when the predicted heat flux is less than 5 KW/m 2 (0.44 Btu/s-ft
: 2) and the empirical constant (radiant fraction) is 0.21. The method is bounding when a safety factor of two is applied to the predicted heat flux. The application in the Generic Fire Modeling Treatments document uses an empirical constant (radiant fraction) of 0.35, indicating the application is essentially bounding. Similarly, it was concluded that the Method of Shokri and Beyler is conservative when the predicted heat flux is
 
greater than 5 KW/m 2 (0.44 Btu/s-ft
: 2) and the method is bounding when a safety factor of two is applied to the predicted heat flux. The implementation in the Generic Fire Modeling Treatments document is conservative, though not bounding. Although the SFPE considered fire diameters greater than about 1 m (3.3 ft), smaller diameter pool fires are not optically thick and have a lower emissive power (SFPE Handbook of Fire Protection Engineering, Section 3-1). Thus, the use of the methods for smaller fires is conservative though outside the SFPE validation range.
The use of the heat flux models largely falls within the NUREG-1824, Volume 1, validation parameter space range; however, there are cases where this is not so. For larger diameter fires, the SFPE provides comprehensive validation against full scale test data of the methods applied. The application in the Generic Fire Modeling Treatments report and the applicable supplements necessarily fall within the validation range or are more conservative because the solution algorithm identifies the most adverse solution among the methods. Smaller fires may fall outside the validation range of both studies, but such fires have a lower emissive power and are conservatively treated using the methods designed for high emissive power source fires.
A number of other empirical models that appear in the generic fire modeling treatments are applied within the stated range of the models or the data for which the models were developed.
For example, the cable heat release rate per unit area model is based on cables that have a small scale heat release rate that ranges between 100 - 1,000 KW/m 2 (8.8 - 88.1 Btu/s-ft 2). The solution tables are provided for this range. The unconfined spill fire model (heat release rate reduction factor) is based on observations of pool fires having a diameter between 1 - 10 m (3.3 - 33 ft). The diameter range for which ZOI data is provided is 0.7 - 5 m (2.2 - 17 ft). The lower range value is less of a concern due the reduction in the optical thickness of the fire when the diameter falls below 1 m (3.3 ft). The upper range is maintained in the ZOI solutions. The offset distance for flame extensions outside a burning panel have an upper observational limit of about 1,000 KW (950 Btu/s), though it is applied in a normalized form (extension to panel height ratio). The ratio is applied as determined from the test data.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-8 The CFAST applications in the Generic Fire Modeling Treatments report consist of simple geometries with a single natural vent path connected to an ambient boundary condition. The simulations are used to determine the time after the start of the fire that the hot gas layer temperature reaches a predetermined critical temperature. No consideration for the hot gas layer depth is made; if the hot gas layer temperature reaches the critical temperature at any time, then this time is the sole output parameter used in the Generic Fire Modeling Treatments report. The enclosure geometry is specified as a function of the volume in such a way as to minimize the heat losses to the boundary. Three vent configurations are evaluated for each volume-room geometry-vent fraction; the most adverse result among the three vent configurations is used.
The room geometry and fire parameters for the Generic Fire Modeling Treatments simulations fall within the model limits listed in NIST SP 1026 and NIST SP 1041. Specifically, the vent area to enclosure volume ratio is less than two and the aspect ratios of the enclosures are less than five. The non-dimensional parameters that affect the model results as documented in NUREG-1824, Volumes 1 and 5, and NUREG-1934 include the model geometry, the global equivalence ratio, the fire Froude Number, and the flame length rati
: o. The non-dimensional parameters that relate to target exposure conditions (heat flux) and sprinkler actuation (ceiling jet) are not applicable to this calculation because these output parameters are not used. The non-dimensional geometry parameters (length-to-height and width-to-height, which range from 3.3 - 4.3) fall within the NUREG-1824, Volume 1, validation range (0.6 - 5.7). As previously noted, CFAST does not use a fire diameter; therefore, it is possible to specify a fire that falls within the range of fire Froude numbers considered in the NUREG-1824, Volumes 1 and 5, validation documentation. 
 
The source fires considered are consistent with those described in NUREG/CR-6850 and thus those that are the subject of the NUREG-1824, Volume 1, validation effort. The global equivalence ratio does exceed the ratio validated in NUREG-1824, Volume 1, in some cases by a significant margin. Large fires in very small volumes with low ventilation could effectively result in equivalence ratios that even exceed the maximum values observed in fully developed fires (3 - 5) [SFPE Handbook of Fire Protection Engineering, Sections 2-5 and 3-4, 2008].
However, the limiting oxygen index used in the model is zero, which forces the combustion process to use all available oxygen within the enclosure and the heat release rate to decrease to a value set by the natural ventilation oxygen inflow. The maximum temperature over the course of the fire occurs at some time prior to the oxygen being consumed in the enclosure, thus the global equivalence ratio for the data reported is based on a condition where it is less than unity and within the validation basis of NUREG-1824, Volume 1. Further, for a given volume and fire size, an optimum ventilation condition will occur over the vent range considered. Because of potential variations in a ventilation condition, the FPRA uses the most adverse time over the reported range and effectively performs an optimization on this parameter.
Finally, the flame length ratio is not always met, especially for large fires postulated in small enclosures. Because sprinkler actuation and thermal radiation to targets are not computed within the CFAST model, this parameter is not an applicable metric. Rather, the plume entrainment below the hot gas layer controls the layer decent time and the concentration of soot products in the layer. This aspect of the model is not affected by the flame height to ceiling height ratio. Consequently, the application of CFAST in the Generic Fire Modeling Treatments document falls within the NUREG-1824, Volume 1, validation parameter space.
Additional V&V studies, which are useful for extending the range of applicability of the model, are contained in NIST SP 1086 and NRL/MR/6180-04-8746. These studies have a broader parameter validation space than NUREG-1824, Volume 1. NIST SP 1086 is based in part on Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-9 the methods of ASTM E1355. NRL/MR/6180-04-8746 provides a Navy specific V&V study, which includes an assessment of CFAST, Version 3.1.7, predictions in multiple enclosures and multiple elevation configurations. These additional studies extend the range of the validation parameter space to include configurations and conditions presented in Appendix B of the Generic Fire Modeling Treatments report.
Appendix B of the Generic Fire Modeling Treatments report provides an in depth analysis of the parameters used as input and Table B-2 indicates the basis for the input parameter selection.
The parameters are either selected as absolutely bounding over the credible range or establish an application limit (e.g., elevated temperature environment and boundary thermal properties).
A summary of the validation basis for both the CFAST and the empirical models is provided in Tables J-1 and J-2 of this attachment. Based on the information in the tables and the preceding discussion, it is shown that that the empirical fire model applications in the Generic Fire Modeling Treatments document either fall within the original correlation bounds or they are outside the bounds, but used in a way that is demonstrably conservative. Likewise, CFAST is used within the model limitations described in the User's Guide (NIST SP 1041) and the Technical Reference Guide (NIST SP 1026). The results as reported in the Generic Fire Modeling Treatments document are based on conditions that meet the NUREG-1824, Volumes 1 and 5, validation space, although there are input specifications that fall outside this range. The use of the Generic Fire Modeling Treatments document in the FPRA performs an optimization over the ventilation fraction and necessarily is based on a condition that falls within the NUREG-1824, Volumes 1 and 5, validation space for the global equivalence ratio. Given these considerations, it is concluded that the CFAST application in the Generic Fire Modeling Treatments document has a validation and verification basis that meets the requirements of NFPA 805, Section 2.4.1.2.3.
 
Generic Fire Modeling Treatments Supplements There are five supplements to the Generic Fire Modeling Treatments document, two of which are used by the ANO-2 FPRA [PRA-A2-05-003; PRA-ES-05-004]:
Supplement 2: "Supplemental Generic Fire Modeling Treatments: Hot Gas Layer Tables;" and Supplement 3: "Supplemental Generic Fire Modeling Treatments: Transient Fuel Package Ignition Source Characteristics."
Supplement 1 - "Supplemental Generic Fire Modeling Treatments: Closed Electrical Panels," Supplement 4 - "Supplemental Generic Fire Modeling Treatments: Transient Target Response to Transient Ignition Source Fire Exposures," and Supplement 5 - "Supplemental Generic Fire Modeling Treatments: Solid State Control Component ZOI and Hot Gas Layer Tables"
[1JMW21020.000-02] are not used in the ANO-2 FPRA [PRA-A2-05-003; PRA-ES-05-004].
Supplement 2 Supplement 2, "Supplemental Generic Fire Modeling Treatments: Hot Gas Layer Tables," provides hot gas layer tables for additional critical temperatures and ignition source heat release rates, including some ignition source-secondary fuel package combinations. In addition, the
 
ZOI dimensions for sensitive component targets are provided. The hot gas layer tables and ZOI dimensions are calculated using the same calculation procedures as were used for the original Generic Fire Modeling Treatments report, but with different input parameters. These procedures were verified and approved as previously described. The validation basis for the ZOI dimensions is identical to that of the original Generic Fire Modeling Treatments report.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-10 Hot gas layers are provided for a single generic secondary combustible configuration:  two horizontal cable trays located 0.3 m (1 ft) above an electrical panel ignition source that may be located in the open, near a wall, or in a corner. The cable trays are simultaneously ignited five minutes after ignition at a single point and the fire is allowed to propagate laterally in each direction. The horizontal flame propagation rate used in the analysis for thermoset cables is as recommended in NUREG/CR-6850 (0.3 mm/s [0.1 in/s]). This propagation rate has been
 
shown to be broadly applicable to the cable class in NUREG/CR-7010, Volume 1. The assumed heat release rate per unit area for the cables is constant and assumed to be
 
225 KW/m 2 (19.8 Btu/s-ft 2), which is slightly less than the value recommended in NUREG/CR-7010, Volume 1, for thermoplastic cables (250 KW/m 2 [22 Btu/s-ft 2]). The applicable value for thermoset cables is 150 KW/m 2 (13.2 Btu/s-ft
: 2) per NUREG-7010, Volume 1. Wall and corner configurations are addressed using the 'Image' or 'Mirror' Method in which the source heat release rate and area are doubled for a wall configuration and quadrupled for a corner configuration. The enclosure boundary surface area and ventilation are also doubled and quadrupled for wall and corner configurations, respectively. This treatment takes advantage of the proportionality of the entrainment to the fire perimeter and the constant plume angle (Beyler, 1986; SFPE Handbook of Fire Protection Engineering, Section 2-1, 2008; Thomas et al.; NIST-GCR-90-580)and results in more adverse conditions when the entrainment/fire perimeter ratio is reduced.
The secondary combustible hot gas layer tables are primarily used as a tool for addressing scenarios with the potential to involve one or two cable trays as secondary combustibles. Scenarios with more adverse cable tray arrangements are addressed in the plant specific detailed fire modeling report (1JMW21022.002-06).
 
Supplement 3 The focus of Supplement 3 to the Generic Fire Modeling Treatments report, "Supplemental Generic Fire Modeling Treatments: Transient Fuel Package Ignition Source Characteristics"
[1JMW21020.000-01], is to provide an analysis of, and basis for, the transient ignition source heat release rate per unit area, the fire duration, and flame height. The analysis uses the original transient fire test data referenced in NUREG/CR-6850 to estimate the transient ignition source characteristics of interest in order to provide a narrower range of input parameters for the ZOI calculations addressed in the G eneric Fire Modeling Treatments report and Supplement 2 to the Generic Fire Modeling Treatments report [1JMW21020.000-01].
Supplement 3 is primarily an analysis of test data; however, several revised ZOI tables using the results of the analysis are provided. The ZOI tables were determined using the same processes and fire models used to generate the original ZOI tables in the Generic Fire Modeling Treatments report. The validation and verification developed for the Generic Fire Modeling Treatments report for the model is thus applicable to this supplement.
Detailed Fire Modeling Calculations
 
The detailed fire modeling calculations as documented in 1JMW21022.002-06 assess the potential for hot gas layers to exceed certain critical temperature thresholds when secondary combustibles are involved. The calculation provides detailed calculations for approximately forty specific ignition source - cable tray confi gurations at ANO-2, primarily those that involve more than two cable trays.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-11 The calculation uses three different fire models:
FLASH-CAT, as incorporated in plant specific Excel TM spreadsheets (1JMW21022.002-06; NUREG/CR-7010, Volume 1);
CFAST, Version 6.1.1 (NIST SP 1026; NIST SP 1041); and Plant specific plume temperature calculation conducted using an Excel TM spreadsheet.
The FLASH-CAT model (NUREG/CR-7010, Volume 1), which essentially involves a group of recommended heat release rate and flame spread parameters for cables in cable trays, is used in the 1JMW21022.002-06 to generate the heat release rate contribution from secondary combustibles. Finally, CFAST is used to generate hot gas layer tables for specific plant spaces and source fire configurations and to predict the Optical Density (OD), a parameter that is proportional to the soot concentration (SFPE Handbook of Fire Protection Engineering, Section 2-13) in Fire Zone 2098-C. The OD concentration in Fire Zone 2098-C is used to quantify when the ionization detection system would actuate, which in turn triggers a Halon suppression system. The CFAST results for the hot gas layer temperature are evaluated over a range of natural ventilation conditions (0.001 - 10 percent of the boundary). The large natural ventilation range considered in the analysis readily encompasses the ability of a forced ventilation system to provide oxygen while conservatively ignoring the mixing or diluting aspects of such systems. In other words, a forced ventilation system is not postulated to provide more oxygen than is already assumed over the range of natural ventilation conditions and the system would tend to improve the result when dilution of the hot gas layer is considered.
 
The purpose of the plant specific plume temperature calculation is to determine the severity factor (via the heat release rate necessary to cause secondary combustibles to ignite) and to determine the time at which a fire could damage a cable tray target in Fire Zone 2098-C. This information is compared against the OD concentration, which is used to determine the time at which the Halon suppression system actuates to show that the suppression system will actuate before target damage occurs. The plume calculation uses the Heskestad Plume [Heskestad, 1981; Heskestad, 1984; SFPE Handbook of Fire Protection Engineering, Section 2-1, 2008] and is solved using an Excel TM spreadsheet.
FLASH-CAT The FLASH-CAT application in 1JMW21022.002-06 is used to generate the temporal heat release rate for specific cable tray arrangements. The input parameters that are used are those recommended in NUREG/CR-7010, Volume 1, and the initial conditions (initial area and ignition criteria) are those recommended in NUREG/CR-6850. The calculation itself is performed using an Excel TM spreadsheet.
The verification basis for the FLASH-CAT model as incorporated in the Excel TM spreadsheet involves numerical comparisons against resu lts presented in NUREG/CR-7010, Volume 1. These comparisons are provided with the detailed fire modeling report (1JMW21022.002-06) and serve as the verification that the m odel is correctly implemented as an Excel TM spreadsheet. The validation for the FLASH-CAT model is provided in NUREG/CR-7010, Volume 1, using about thirty different cable samples. The samples include cables having the same or similar materials as the predominant cable types used at ANO-2 (e.g., chlorosulfonated polyethylene per ER-ANO-2003-0450-000) such that the results and conclusions are applicable.
An added measure of conservatism is provided in the FLASH-CAT analysis by assuming thermoplastic cable flame spread and propagation properties (1JMW21022.002-06).
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-12 There is no validation range per se specified for the FLASH-CAT model (NUREG/CR-7010, Volume 1). Rather, it may be inferred that if the configuration is similar (i.e., horizontal cable tray stacks) and the cable composition is similar, the results are applicable and NUREG/CR-7010, Volume 1, serves as the validation basis. The FLASH-CAT applications described in 1JMW21022.002-06 involve horizontal cable tray stacks with some vertical or vertically sloped segments involving materials that are among those tested. The horizontal
 
segments conform to the NUREG/CR-7010, Volume 1, test configuration, but the vertical segments do not. However, the vertical segments are conservatively assumed to propagate at a faster rate as recommended in NUREG/CR-6850. Therefore, the FLASH-CAT application has a validation and verification basis that meets the requirements of NFPA 805, Section 2.4.1.2.3.
CFAST  CFAST, Version 6.1.1 (NIST SP 1026, NIST SP 1041), is used to generate hot gas layer tables that provide the time various temperature thresholds are reached in the specific spaces using the FLASH-CAT temporal heat release rates and to predict the OD within Fire Zone 2098-C.
The CFAST hot gas layer temperature applications are identical to the approach adopted in the Generic Fire Modeling Treatments document with the following exceptions:
 
The height of the specific space is used in lieu of a generic room shape.
The room volume is used rather than the generic room volumes. The length and the width of the space are determined by minimizing the surface area given a height and volume. The fire heights are set based on the particular fuel packages examined. Thus, panel fires are modeled at the top of the panel and transient fuel packages are modeled 0.6 m (2 ft) above the floor.
 
The CFAST analysis assesses the time the hot gas layer temperature reaches threshold values over a range of ventilation conditions (0.001 - 10 percent of the boundary area). The ventilation condition that results in the most adverse time for a given scenario is used in the FPRA (PRA-A2-05-003).
 
The verification for the CFAST model (Version 6.0.5) is provided in NUREG-1824, Volume 5.
Supplemental verification for CFAST, Version 6.1.1, is provided as an appendix to the 1JMW21022.002-06 report as well as in NIST SP 1086.
 
The validation for CFAST described for the original Generic Fire Modeling Treatments document applies, except as follows:
 
The equivalence ratio for some ventilation cases will fall outside the NUREG-1824, Volume 1, validation parameter space. However, at least one ventilation condition will be within this range, and the results are thus no less conservative than a case that falls within the NUREG-1824, Volume 1, validation parameter space. In general, the most adverse results will be predicted when the equivalence ratio is near unity (optimum burning conditions). Validation work has been performed for CFAST at these equivalence ratios (e.g., NRL/MR/6180-04-8746, NIST SP 1086) and applies to the
 
ANO-2 calculation.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-13 The calculation of the OD in Fire Zone 2098-C uses the specific room geometry and various electrical panel heat release rate fire size bins. The application falls within the NUREG-1824, Volume 1, non-dimensional validation parameter space for all parameters except for the equivalence ratio and the fire Froude Number. In low heat release rate bin fire scenarios, the equivalence ratio can be lower than the mini mum value of 0.04 considered in NUREG-1824, Volume 1. However, as was the case for the MCR analysis, this means that there is a sufficient supply of oxygen available for the fire up to the time at which the smoke detectors are predicted to actuate. Further, all scenarios begin with an equivalence ratio of 0.0, including those that form the NUREG 1824, Volume 1, validation basis, so that the scenarios evaluated in Fire Zone 2098-C are not inconsistent with the validation scenarios considered in NUREG-1824, Volume 1, simply because the equivalence ratio is low. In the case of the fire Froude Number, fire sizes are considered over a range of values that include fires within the NUREG-1824, Volume 1, validation range. The analysis conclusions are based on the most adverse conditions predicted. As such, the use of fire Froude Numbers that fall outside the validation is reasonable for this application given the results are either based on values that fall within the validation range or are more conservative. This is described in the appendix section of
 
1JMW21022.002-06.
 
Based on these considerations, it is concluded that the V&V basis for the CFAST application analysis meets the NFPA 805, Section 2.4.1.2.3, requirements.
Excel TM Plume Temperature Calculation The plume temperature at a fixed elevation above an ignition source is computed using the Heskestad Plume model as programmed into an Excel TM spreadsheet. The Heskestad Plume model is included in the FDT S calculation suite (NUREG-1805), and is one of the calculations/functions NUREG-1824, Volume 3, assessed as part of its V&V study.
Consequently, NUREG-1824, Volume 3, serves as the validation basis for the Excel TM plume temperature calculation. The applicability of the NUREG 1824, Volume 1, validation studies to the fire scenarios evaluated in 1JMW21022.002-06 is summarized in the appendix section of 1JMW21022.002-06. Specifically, the fire Froude Number Q
* and the target height to dimensionless fire diameter ratio H/D
* are the only parameters that are applicable to a plume temperature calculation. It is shown in the appendix section of 1JMW21022.002-06 that the values for these parameters either fall within the NUREG-1824, Volume 1, validation range or produce conservative results.
 
The verification of the spreadsheet was conducted through a comparison of the Excel TM spreadsheet predictions with values obtained by a previously verified method (NUREG-1805, NUREG-1824, Volume 1, NUREG-1824, Volume 3). This verification is documented in the appendix section of 1JMW21022.002-06.
 
Based on these considerations, it is concluded that the V&V basis for the Excel TM spreadsheet computation meets the NFPA 805, Section 2.4.1.2.3, requirements.
 
Embedded Conduit Fire Resistance
 
The fire resistance for conduit embedded in concrete boundaries is determined in the calculation titled "Thermal Analysis of Concrete Embedded Conduit" (ANOC-FP-07-00001). This calculation serves as part of the basis for excluding impact to cables in embedded conduit (EC-494) and is therefore implicitly credited in the FPRA.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-14 The fire resistance of concrete embedded conduit is calculated using the finite difference conduction heat transfer model HEATING, Version 7.3 (Technical Report PSR-199) for various types and sized conduits and conduit embed depths. HEATING, Version 7.3, was developed at the Oak Ridge National Laboratories as a general purpose finite difference heat transfer model for use in the commercial and government nuclear industries. There are a number of validation and verification reports and benchmark solution cases for general applications of the HEATING model (e.g., Technical Report K/CSD/INF-89/4, Technical Report K/CSD/TM-61, Technical Report ORNL/NUREG/CSD-2/V2/R3). A verifica tion and validation study for fire related applications is documented in NRL/MR/6180-04-8746. The model verification summarized in NRL/MR/6180-04-8746 is based on the methodology developed by Wickstrm ([Wickstrm, 1999; Wickstrm et al., 1999; Plsson et al., 2000) for which the solutions of eight fire exposure configurations of increasing complexity are provided. The simplest cases have exact analytic solutions whereas the more complex cases involve a comparison against a baseline heat transfer solution generated by the conduction finite element model TASEF (Sterner et al). The model validation documented by NRL/MR/6180-04-8746 is based on eight test cases of increasing complexity for which measured data is available, one of which includes steel embedded in concrete. These sixteen V&V cases are consistent with the ASTM E1355 procedure for providing fire model V&V and thus meet the NFPA 805, Section 2.4.1.2.3, requirement for using a fire model that has undergone a V&V process and is applied within its limitations.
 
The embedded conduit calculation ANOC-FP-07-00001 provides a detailed description of the Wickstrm et al. verification case involving a convection fire exposure to a two-dimensional concrete slab as documented in NRL/MR/6180-04-8746, a similar configuration to the concrete embedded conduit evaluated at ANO-2. This verification case has an analytic solution and serves as the model benchmark for the calculation and serves as a demonstration that the application is within the model limitations. A parameter sensitivity analysis is provided in embedded concrete calculation, including material property uncertainty, boundary condition uncertainty, and mesh dependencies. Table J-1 provides a summary of the validation and verification basis for HEATING, Version 7.3, as applied in embedded conduit calculation.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-15  Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation Application V & V Basis Discussion MCR Abandonment Calculation of operator abandonment times in the MCR. NUREG 1824, Volumes 1 & 5 NIST SP 1026 NIST SP 1041 NIST SP 1086 NUREG/CR 4527, Volume 2 NRL/MR/6180-04-8746 The abandonment time in the MCR is determined by computing the time for the visibility and temperature to reach thresholds as specified in NUREG/CR-6850. CFAST, Version 6.0.5, has been validated for certain configurations in terms of predicting the temperature increase in an enclosure in accordance with NUREG-1824, Volume 5. In addition, NUREG/CR-4527, Volume 2, provides full scale test data of electrical panel fires in control room like structures. These tests are modeled using the CFAST, Version 6.0.10, and the results are documented in report entitled "Evaluation of Unit 2 Control Room Abandonment Times at the Arkansas Nuclear One Facility."  CFAST, Version 6.0.10, is found to provide a reasonable and conservative estimate of both the hot gas layer temperature and visibility as a function of time given the input fire size for a control room like enclosure. This information is documented in Appendix D of the Report entitled "Evaluation of Unit 2 Control Room Abandonment Times at the Arkansas Nuclear One Facility" (CALC-ANO2-FP-09-00013). The MCR abandonment application falls within the non-dimensional parameter space for the NUREG-1824, Volumes 1 and 5, V&V report as estimated using the methods described in NUREG-1934. The application also falls within the model limits as specified in NIST SP 1026 and NIST SP 1041. Additional V&V documentation is provided in NIST SP 1086 and NRL/MR/6180-04-8746 that expand the validation parameter space from that included in NUREG-1824, Volumes 1 and 5, including multiple compartment applications.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-16 Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation Application V & V Basis Discussion Generic Fire Modeling Treatments Definition of zones of influence about specific classes of ignition sources. Scenario screening for the multi-compartment analysis. NUREG 1824, Volumes 1, 3, 5 NIST SP 1026 NIST SP 1041 NIST SP 1086 Table J-2 Table J-2 provides a summary of the validation basis for the empirical models used in the Generic Fire Modeling Treatments. The Generic Fire Modeling Treatments report uses CFAST, Version 6.0.10, in a simple geometry that minimizes the boundary heat losses given a volume. For the volume postulated, the configuration produces the most adverse result regardless of the actual dimensions used. The application falls within the model limits as specified in NIST SP 1026 and NIST SP 1041. Except for the global equivalence ratio, the non-dimensional parameters fall within the V&V space of NUREG-1824, Volumes 1 and 5. Although equivalence ratios are considered over a much larger range than addressed by the NUREG-1824, Volume 1, validation tests, the results are based on a single time point based on an equivalence ratio that is close to unity or lower and thus may fall directly within the NUREG-1824, Volume 1, validation parameter space. Additional validation results that consider the higher predictive capability under higher equivalence ratios are provided in NIST SP 1086. Supplemental Generic Fire Model Treatments: Hot Gas Layer Tables, (Supplement 2) Definition of zones of influence about specific classes of ignition sources for use in the FPRA. Scenario screening for the multi-compartment analysis. NUREG 1824, Volume 5 NIST SP 1026 NIST SP 1041 NIST SP 1086 NUREG/CR-4527, Volume 2 NRL/MR/6180-04-8746 NUREG/CR-6850 NUREG/CR-7010, Volume 1 Table J-2 The same methods developed in Generic Fire Modeling Treatments document are used to generate additional hot gas layer tables and ZOI definitions. The treatment of secondary applies to a two tray configuration located above an electrical panel ignition source and used propagation rates recommended by NUREG/CR-6850 and validated for a wide range of cable compositions in NUREG/CR-7010, Volume 1.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-17 Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation Application V & V Basis Discussion Supplemental Generic Fire Model Treatments: Transient Ignition Source Strength (Supplement 3) Characterization of the heat release rate per unit area, fire duration, and flame height for transient ignition sources. Provides revised ZOI tables for transient fuel packages based on the analysis of the transient fire test data. NUREG 1824, Volumes 1, 3, 5 NIST SP 1026 NIST SP 1041 NIST SP 1086 Table J-2 The supplement provides an analysis of the transient fuel package fire tests in order to better characterize the heat release rate per unit area, the fire duration, and the flame height. These parameters are used in the development of the ZOI in the original Generic Fire Modeling Treatments document and, prior to the development of Supplement 3, were conservatively bounded. Supplement 3 provides the basis for a narrower parameter value range as determined from the actual fire test reports on which the NUREG/CR-6850 conditional probability distribution was established. Revised ZOI tables are developed for transient ignition source fuel packages using the results of the fire test data analysis. The ZOIs are computed using the same processes as the original Generic Fire Modeling Treatments report and the V&V basis is therefore the same. Detailed Fire Scenario Calculations Calculation of the time the hot gas layer reaches critical temperature thresholds for scenarios involving secondary combustibles (multiple cable trays). Calculation of the actuation time of a Halon suppression system. NUREG 1824, Volumes 1, 3, 5 NUREG/CR-7010, Volume 1 NIST SP 1026 NIST SP 1041 NIST SP 1086 NRL/MR/6180-04-8746 Detailed evaluations are provided for specific ignition source-secondary combustible configurations involving multiple cable trays. In addition, a simple analysis of the detector actuation time (and the Halon suppression actuation time) versus the predicted damage time for a cable tray target is conducted in Fire Zone 2098-C. Three fire modeling tools are used in this assessment:  FLASH-CAT (NUREG/CR-7010, Volume 1), CFAST, Version 6.1.1 (NIST SP 1041), and a plant specific Excel TM spreadsheet plume temperature computation (1JMW21022.002-06). The FLASH-CAT model is used to compute the temporal heat release rate profiles for specific cable tray arrangements where secondary combustibles are included. CFAST, Version 6.1.1, is used to compute the time the hot gas layer temperature reaches various threshold values given the ignition source and secondary combustible heat release rates. CFAST, Version 6.1.1, is also used to calculate the OD in Fire Zone 2098-C as a function of time for various electrical panel fire scenarios. The plant specific Excel TM spreadsheet is used to assess the fire size or damage time at targets above an ignition source using the Heskestad Plume model (Heskestad, 1981; Heskestad, 1984; SFPE Handbook of Fire Protection Engineering, Section 2-1, 2008). (continued)
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-18 Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation Application V & V Basis Discussion Detailed Fire Scenario Calculations (continued)
NUREG-1824, Volume 5, provides the verification and validation basis for CFAST. Supplemental verification is provided in 1JMW21022.002-06 for the specific CFAST version used. Verification and validation for the FLASH-CAT model is provided in 1JMW21022.002-06 via comparisons with NUREG/CR-7010, Volume 1, results. Verification and validation of the plant specific Excel TM spreadsheet plume temperature computation is provided in NUREG-1824, Volume 3, and in the appendix section of 1JMW21022.002-06. CFAST, Version 6.1.1, is used within the NUREG-1824, Volume 1, parameter space for at least one ventilation condition per hot gas layer fire scenario; the most adverse ventilation condition is selected for each scenario; thus, the results are at least as conservative as a case that falls within the NUREG 1824, Volume 1, validation space. CFAST, Version 6.1.1, is also used within the NUREG-1824, Volume 1, validation space for the OD calculation in Fire Zone 2098-C, except for low fire Froude Numbers for some heat release rate bins. The FLASH-CAT model uses the recommended input parameters of NUREG/CR-7010, Volume 1, and is used to calculate the heat release rate in horizontal cable trays containing cables similar to those tested. Therefore, the application falls within the validated range for FLASH-CAT. The plant specific Excel TM spreadsheet plume temperature computation is used within the NUREG-1824, Volume 1, validation space or is shown to be conservative for the application in the appendix section of 1JMW21022.002-06. Thermal Analysis of Concrete Embedded Conduit Basis for excluding concrete embedded raceways from adjacent fire zones per EC-464. NRL/MR/6180-04-8746 Technical Report K/CSD/INF-89/4 Technical Report K/CSD/TM-61 Technical Report ORNL/NUREG/CSD-2/V2/R3 The fire resistance of conduit embedded in concrete is calculated using the finite difference model HEATING, Version 7.3, for various concre te cover thicknesses, conduit diameters, and conduit types. The base finite difference model is a one-dimensional geometry, but it includes material properties that vary with temperature and boundary conditions that vary with time. Several two-dimensional geometries are evaluated and compared with the one-dimensional counterparts and it is shown that the one-dimensional model is universally conservative in this application. A sensitivity analysis is provided that demonstrates the results are not dependent on material property or boundary condition uncertainty, unless there is sustained flame impingent.                          (continued)
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-19 Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation Application V & V Basis Discussion Thermal Analysis of Concrete Embedded Conduit (continued)  NRL/MR/6180-04-8746 provides a verification and validation assessment of HEATING as applied to fire exposure configurations us ing the method recommended by Wickstrm (Wickstrm, 1999; Wickstrm et al., 1999; Plsson et al., 2000). Other validation and verification studies on older revisions are documented in Technical Report K/CSD/INF-89/4, Technical Report K/CSD/TM-61, and Technical Report ORNL/NUREG/CSD-2/V2/R3. A validation case involving a two dimensional slab exposed to a convection boundary condition as a model application benchmark is provided in the embedded conduit calculation.
 
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-20  Table J-2 V & V Basis for Fire Models / Model Correlations Used:  Generic Treatment Correlations Correlation Location in Hughes Associates Original Reference Application Original Correlation Range Subsequent Validation and Verification Limits in Hughes Associates Flame Height Page 18 Heskestad
[1981] Heskestad
[1984] Provides a limit on the use of the ZOI. In practice, wood and hydrocarbon fuels, momentum or buoyancy dominated, with diameters between 0.05 - 10 m (0.16 - 33 ft). Directly NUREG 1824, Volume 3] Indirectly NUREG 1824, Volume 5  (Correlation used in CFAST) 4H C D 2 < 3000  Point Source Model Page 19 Modak Lateral extent of ZOI - comparison to other methods. Isotropic flame radiation. Compared with data for 0.37 m (1.2 ft) diameter PMMA pool fire and a target located at a R o/R ratio of 10. NUREG 1824, Volume 3 SFPE (1999) Predicted heat flux at target is less than 5 KW/m 2 (0.44 Btu/s-ft
: 2) per SFPE (1999). Method of Shokri and Beyler Page 19 Shokri et al. Lateral extent of ZOI - comparison to other methods. Pool aspect ratio less than 2.5. Hydrocarbon fuel in pools with a diameter between 1 - 30 m (3.3 - 98 ft). Vertical target, ground level. SFPE (1999) NUREG 1824, Volume 3 Ground based vertical target. Method of Mudan (and Croce) Page 20 Mudan Lateral extent of ZOI - comparison to other methods. Round pools; Hydrocarbon fuel in pools with a diameter between 0.5 - 80 m (1.64 - 262 ft). SFPE (1999) Total energy emitted by thermal radiation less than total heat released.
Method of Shokri and Beyler Page 20 Shokri et al. Lateral extent of ZOI. Round pools; Hydrocarbon fuel in pools with a diameter between 1 - 50 m (3.3 - 164 ft). SFPE (1999) NUREG 1824, Volume 3 Predicted heat flux at target is greater than 5 KW/m 2 (0.44 Btu/s-ft
: 2) per SFPE. Shown to produce most conservative heat flux over range of scenarios considered among all methods considered.
~-5 < log 10 c p Tg(Hc/r)3 Q 2 D 5~< 5.~-5 < log 10 c p Tg(Hc/r)3 Q 2 D 5~< 5.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-21 Table J-2 V & V Basis for Fire Models / Model Correlations Used:  Generic Treatment Correlations Correlation Location in Hughes Associates Original Reference Application Original Correlation Range Subsequent Validation and Verification Limits in Hughes Associates Plume heat fluxes Page 22 Wakamutsu et al. Vertical extent of ZOI. Fires with an aspect ratio of about 1 and having a plan area less than 1 m 2 (0.09 ft 2). Wakamatsu et al. (larger fires) SFPE Handbook of Fire Protection Engineering, Section 2-14, 2008 Area source fires with aspect ratio ~ 1. Used with plume centerline temperature correlation; most severe of the two is used as basis for the ZOI dimension. This is not a constraint in the fire model analysis for the cases evaluated.
Plume centerline temperature Page 23 Yokoi  Beyler Vertical extent of ZOI. Alcohol lamp assumed to effectively be a fire with a diameter ~0.1 m (0.33 ft). NUREG 1824, Volume 3 SFPE Handbook of Fire Protection Engineering, Section 2-1, 2008 Area source fires with aspect ratio ~ 1. Used with plume flux correlation; most severe of the two is used as basis for the ZOI dimension. Hydrocarbon spill fire size Page 51 SFPE Handbook, Section 2-15 (2002) Determine heat release rate for unconfined hydrocarbon spill fires. Hydrocarbon spill fires on concrete surfaces ranging from ~1 to ~10 m (3.3 - 33 ft) in diameter. None. Based on limited number of observations. None. Transition from unconfined spill fire to deep pool burning assumed to be abrupt. Flame extension Page 100 SFPE Handbook of Fire Protection Engineering, Section 2-14 (2002) Determine the fire offset for open panel fires. Corner fires ranging from ~10 to ~1,000 KW (9.5 - 948 Btu/s). Fires included gas burners and hydrocarbon pans. None. Based on limited number of observations. None. Offset is assumed equal to the depth of the ceiling jet from the experiments. Line source flame height Page 101 Delichatsios Determine the vertical extent of the ZOI. Theoretical development. SFPE Handbook of Fire Protection Engineering, Section 2-14 (2008) None. Transition to area source assumed for aspect plan ratios less than four. Maximum of area and line source predictions used in this region.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-22 Table J-2 V & V Basis for Fire Models / Model Correlations Used:  Generic Treatment Correlations Correlation Location in Hughes Associates Original Reference Application Original Correlation Range Subsequent Validation and Verification Limits in Hughes Associates Corner flame height Page 108 SFPE Handbook of Fire Protection Engineering, Section 2-14 (2002) Determine the vertical extent of the ZOI. Corner fires ranging from ~10 to ~1,000 KW (9.5 - 948 Btu/s). Fires included gas burners and hydrocarbon pans. None. Correlation form is consistent with other methods; comparison to the dataset from SFPE Handbook, Section 2-14 (2002) and SFPE Handbook, Section 2-14 (2008) provide the basis. None. Air mass flow through opening Page 140 Kawagoe Compare mechanical ventilation and natural ventilation. Small scale,  scale, and full scale single rooms with concrete and steel boundaries. Vent sizes and thus opening factor varied. Wood crib fuels. Drysdale SFPE (2004) None. SFPE (2002) spaces with a wide range of opening factors. Line fire flame height Page 210 Yuan et al. Provides a limit on the use of the ZOI; Extent of ZOI for cable tray fires. 0.002 < Z/' < 0.6 In practice, from the base to several times the flame height from 0.15 - 0.5 m (0.5 - 1.64 ft) wide gas burners. None. Correlation form is consistent with other methods; comparison to dataset from Yuan et al. provides basis. None. Cable heat release rate per unit area Page 210 NBSIR 85-3196Provides assurance that the method used is bounding. Cables with heat release rates per unit area ranging from about 100 - 1000 KW/m 2 (8.8 - 88 Btu/s-ft 2). None. Correlation predicts a lower heat release rate than assumed in the Treatments and is based on test data.
Line fire plume centerline temperature Page 212 Yuan et al. Provides a limit on the use of the ZOI; Extent of ZOI for cable tray fires. 0.002 < Z/' < 0.6 In practice, from the base to several times the flame height from 0.15 - 0.5 m (0.15 - 1.64 ft) wide gas burners. None. Correlation form is consistent with other methods; comparison to dataset from Yuan et al. provides basis. None.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-23 Table J-2 V & V Basis for Fire Models / Model Correlations Used:  Generic Treatment Correlations Correlation Location in Hughes Associates Original Reference Application Original Correlation Range Subsequent Validation and Verification Limits in Hughes Associates Ventilation limited fire size Page 283 Babrauskas Assessing the significance of vent position on the hot gas layer temperature. Ventilation factors between 0.06 - 7.51. Fire sizes between 11 - 2,800 KW (10 - 2,654 Btu/s) Wood, plastic, and natural gas fuels. SFPE (2004) None. Provides depth in the analysis of the selected vent positions. The global equivalence ratio provides an alternate measure of the applicability of the analysis and for reported output is within the validation range of CFAST.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-24 References
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: 45. SFPE Handbook of Fire Protection Engineering, Section 2-13, "Smoke Production and Properties," Mulholland, G. W., The SFPE Handbook of Fire Protection Engineering , 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008. 46. SFPE Handbook of Fire Protection Engineering, Section 2-14, "Heat Fluxes from Fires to Surfaces," Lattimer, B. Y., The SFPE Handbook of Fire Protection Engineering , 3 rd Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2002.
: 47. SFPE Handbook of Fire Protection Engineering, Section 2-14, "Heat Fluxes from Fires to Surfaces," Lattimer, B. Y., The SFPE Handbook of Fire Protection Engineering , 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 48. SFPE Handbook of Fire Protection Engineering, Section 2-15, "Liquid Fuel Fires," Gottuk, D. and White, D., The SFPE Handbook of Fire Protection Engineering , 3 rd Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2002.
: 49. SFPE Handbook of Fire Protection Engineering, Section 3-1, "Heat Release Rates," Babrauskas, V., The SFPE Handbook of Fire Protection Engineering , 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 50. SFPE Handbook of Fire Protection Engineering, Section 3-4, "Generation of Heat and Chemical Compounds in Fires," Tewarson, A., The SFPE Handbook of Fire Protection Engineering , 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 51. Shokri, M and Beyler, C. L., "Radiation from Large Pool Fires," SFPE Journal of Fire Protection Engineering, Vol. 1, No. 4, pp. 141-150, 1989.
: 52. Sterner, E., and Wickstrm, U., "TASEF - Temperature Analysis of Structures Exposed to Fire," SP Report 1990:05, SP Swedish National Testing and Research Institute, Bors, Sweden, 1990.
: 53. Technical Report K/CSD/INF-89/4, "HEATCHEK: A Computer Program to Automate Verification of New Versions of HEATING," Chu, W., Union Carbide Corp., Nuclear Div., Gaseous Diffusion Plant, Oak Ridge, Tennessee, 1989.
: 54. Technical Report K/CSD/TM-61, "HEATING6 Verification," Bryan, C. B., Childs, K. W., and Giles, G. E., Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1986.
Arkansas Nuclear One - Unit 2 Att. J - Fire Modeling V&V Enclosure 1 to 2CAN121202 Page J-28
: 55. Technical Report ORNL/NUREG/CSD-2/V2/R3, "HEATING: A Computer Program for Multidimensional Heat Transfer Analysis (Version 6.1), Sect. F10 of SCALE: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation," Childs, K. W., Giles, G. E., Bryan, C. B., and Cobb, C. K., Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1990.
: 56. Technical Report PSR-199, "HEATING 7: Multidimensional, Finite-Difference Heat Conduction Analysis Code System," Childs, K. W., Oak Ridge National Laboratory (ORNL),
Oak Ridge, TN, 1998.
: 57. Thomas, P. H., Hinkley, P. L., Theobald, C.R., and Sims, D. L., "Fire Technical Paper No. 7," H. M. Stationery Office, Joint Fire Research Organization, London, 1963.
: 58. Wakamatsu, T., Hasemi, Y., Kagiya, K., and Kamikawa, D., "Heating Mechanism of Unprotected Steel Beam Installed Beneath Ceiling and Exposed to a Localized Fire: Verification Using the Real-scale Experiment and Effects of the Smoke Layer,"
Proceedings of the Seventh International Symposium on Fire Safety Science, International Association for Fire Safety Science, London, UK, 2003.
: 59. Wickstrm, U., "An Evaluation Scheme of Computer Codes for Calculating Temperature in Fire Exposed Structures," in S. Grayson (Ed.), Interflam '99 (pp. 1033-1044), London: Interscience Communications Ltd., 1999.
: 60. Wickstrm, U., and Plsson, J., "A Scheme for the Verification of Computer Codes for Calculating Temperature in Fire Exposed Structures," SP Report 1999:36, Bors, Sweden: SP Swedish National Testing and Research Institute, 1999.
: 61. Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current," Report Number 34, Building Research Institute, Tokyo, Japan, 1960.
: 62. Yuan, L. and Cox, F., "An Experimental Study of Some Line Fires," Fire Safety Journal , 27, 1996. 
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-1 K. Existing Licensing Action Transition Licensing Action Appendix R Exemption 01, FA - OO, Not Meeting III.G.2 Criteria
 
Basis Date: 10/01/1999 Transitioned
? No Basis: Exemption request per 2CAN109703 / 2CAN029905 provides the justification for not providing suppression/detection for the below 354' elevation of the Intake Structure per Section III.G.2.c, which was approved by the NRC in 2CNA109902 based on:
Low combustible loading in the area  Fire detection is provided in areas above this elevation  Normally floor of bays are water covered at this elevation  Redundant components protected with 1-hour fire rated barriers The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on fixed fire suppression and smoke detection in the area.
Unit Fire Area Name Description ANO-2 OO Unit 2 Intake Structure Fire Zone Name Description INTAKE Intake Structure (Unit 2)
Reference Document
 
2CNA109902, Issuance of Exemption from the Requirements of 10 CFR 50, Appendix R, Section III.G.2.c for ANO, Unit No. 2 (TAC No. MA1545), 10/1/1999, Attachment, SER
 
Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-2 Licensing Action
 
Appendix R Exemption 02, FA - OO, Not Meeting III.G.2 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN078202 and 0CAN118210 provides, for elevations 354' and 366', the justification for lack of meeting separation with detection/suppression per III.G.2, which was approved by the NRC in 0CNA038328 based on:
Manual fire suppression capability is available for this area  Low combustible loading in the area  Fire detection is provided in this area (elevations 354' and 366')  Fire suppression is provided on the 366' elevation  Partial radiant energy shield at 366' elevation The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.2 separation with fixed fire suppression and smoke detection.
Unit Fire Area Name Description ANO-2 OO Unit 2 Intake Structure Fire Zone Name Description INTAKE Intake Structure (Unit 2)
 
Reference Document 0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, ANO-2 SER, Section 2.0
 
Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-3 Licensing Action
 
Appendix R Exemption 03, FA - CC, Not Meeting 3-hour Rated Barrier, III.G.2 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN078202 / 0CAN088508 provides the following justification for the lack of 3-hour rated barrier for redundant EFW components which was approved by the NRC in 0CNA038328 based on:
Fire area boundary door (water tight door) is adequate for the hazard on both sides of the fire area boundary.
An EEEE (ANOC-FP-07-00003) encompasses the watertight doors. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since fire area boundary requirements are addressed by an
 
EEEE. Unit Fire Area Name Description ANO-2 CC Emergency Feedwater Pump Room (Turbine Driven)
Fire Zone Name Description 2024-JJ Emergency Feedwater Pump Room (Turbine Driven)
Reference Document
 
0CNA038328, 10 CFR 50. Appendix R Exemptions Approved, 3/22/1983, Attachment, ANO-2
 
SER, Section 5.0
 
Evaluation
 
This Fire Area was not transitioned due to an EEEE for this configuration; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-4 Licensing Action
 
Appendix R Exemption 04, FA - NN, Not Meeting III.G.2 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN078202 provides the justification for lack of 20' separation with no intervening combustibles to meet the requirements of Appendix R, Section III.G.2, which was approved by the NRC in 0CNA038328
 
based on:
Firestops along selected cable trays  Partial radiant energy shield  Separation is ~17 feet versus 20 feet  Limited access to prevent transient combustible concern The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.2 separation with no intervening combustibles or fire hazards.
Unit Fire Area Name Description ANO-2 NN Unit 2 Containment Building Fire Zone Name Description 2032-K Containment Building South Side 2033-K Containment Building North Side
 
Reference Document 0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, ANO-2
 
SER, Section 6.0
 
Evaluation This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-5 Licensing Action
 
Appendix R Exemption 05, FA - DD, Not Meeting III.G.2 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN078202 provides the justification for the lack of full coverage with an automatic suppression system and full height 1-hour rated fire barriers between redundant charging pumps and associated cables which
 
were approved by the NRC in 0CNA038328.
Manual fire suppression capability is available for this area  Low combustible loading in the area  Fire detection is provided in this area  Partial height walls  Floor drains to limit flammable lube oil spill size  Firestops along selected cable trays and one hour fire barrier on selected cables  The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.2 requirements for automatic suppression system and full height 1-hour rated fire barriers between redundant charging pumps and associated
 
cables. Unit Fire Area Name Description ANO-2 DD Unit 2 General Area 335' Elevation Fire Zone Name Description 2040-JJ Corridor Reference Document
 
0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, ANO-2
 
SER, Section 7.0
 
Evaluation This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-6 Licensing Action
 
Appendix R Exemption 06, FA - EE-L (Originally a portion of FA - EE), Not Meeting III.G Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN078202 and updated by 0CAN118210 provides the following justification for a fire area boundary not meeting III.G criteria for 3-hour fire rating of a door, which was approved by the NRC in 0CNA038328.
The NRC approval was for a rated fire door to have a low pressure blow out capability to meet room pressurization limitations that should not impact the fire door rating from reasonable fires within this area.
An EEEE (ANOC-FP-09-00003) encompasses this door which is now constructed with an adequate latch throw to maintain the fire door rating and is properly evaluated for room pressurization. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since fire area boundary requirements are addressed by an EEEE.
Unit Fire Area Name Description ANO-2 EE-L South Piping Penetration Rooms Fire Zone Name Description 2055-JJ Lower South Piping Penetration Room
 
Reference Document 0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, ANO-2 SER, Section 8.0
 
Evaluation
 
This Fire Area was not transitioned due to an EEEE showing compliance; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-7 Licensing Action
 
Appendix R Exemption 07, FA - JJ, Lack of Barrier/Separation III.G.2 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN078202 provides the following justification for the lack of 20' separation of panels free of intervening combustibles, which was approved by the NRC in 0CNA038328.
Conduits associated with one panel protected with one-hour rated fire barrier and existing early warning detection system and automatic water deluge system with flame retardant coating provided on intervening cable trays.
The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.2 separation with no intervening combustibles or fire hazards.
Unit Fire Area Name Description ANO-2 JJ Corridor Fire Zone Name Description 2109-U Corridor
 
Reference Document 0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, ANO-2
 
SER, Section 9.0
 
Evaluation This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-8 Licensing Action
 
Appendix R Exemption 08, FA - GG, Not Meeting III.G.3 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN118210 provides the justification for lack of suppression and detection required by Appendix R, Section III.G.3, in Fire Area GG (Fire Zone 2081-HH), which was approved by the NRC in 0CNA038328.
The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.3, separation with no intervening combustibles or fire hazards.
Unit Fire Area Name Description ANO-2 GG Unit 2 North Electrical and Piping Penetration Area Fire Zone Name Description 2081-HH Upper North and Lower North Piping Penetration Room Reference Document
 
0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, Exemption IV, SER, Section 10.0 Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-9 Licensing Action
 
Appendix R Exemption 09, FA - HH, Not Meeting III.G.3 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN118210 provides the justification for lack of suppression required by Appendix R, Section III.G.3 in Fire Area HH (Fire Zones 2096-M, 2106-R, and 2107-N), which was approved by the NRC in 0CNA038328.
The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.3, separation with no intervening combustibles or fire hazards.
Unit Fire Area Name Description ANO-2 HH Unit 2 General Area 354' Elevation Fire Zone Name Description 2096-M Motor Control Center (2B63) 2106-R Degasifier Vacuum Pump Room 2107-N Corridor (North of Stairway 2001)
 
Reference Document 0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, Exemption IV, SER, Section 10.0
 
Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-10 Licensing Action
 
Appendix R Exemption 10, FA - B-3, Not Meeting III.G.3 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN118210 provides the justification for lack of suppression required by Appendix R, Section III.G.3 in Fire Area B-3 (Fire
 
Zone 2091-BB), which was approved by the NRC in 0CNA038328.
The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.3, separation with no intervening combustibles or fire hazards.
Unit Fire Area Name Description ANO-2 B-3 North Penetration Areas Fire Zone Name Description 2091-BB North Electrical Equipment Room Reference Document
 
0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, Exemption IV, SER, Section 10.0
 
Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-11 Licensing Action
 
Appendix R Exemption 11, FA - B-2, Not Meeting III.G.3 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN118210 provides the justification for lack of suppression required by Appendix R, Section III.G.3 in Fire Area B-2 (Fire
 
Zone 2223-KK), which was approved by the NRC in 0CNA038328.
The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on Appendix R, Section III.G.3, separation with no intervening combustibles or fire hazards.
Unit Fire Area Name Description ANO-2 B-2 Unit 2 General Plant Multiple Elevations Fire Zone Name Description 2223-KK Pipeway Equipment Access R oom (Aux. Bldg. Extension)
Reference Document
 
0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, Exemption IV, SER, Section 10.0
 
Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-12 Licensing Action
 
Appendix R Exemption 12, FA - G, Not Meeting III.G.3 Criteria Basis Date: 03/22/1983
 
Transitioned
? No  Basis: Exemption request per 0CAN118210 provides the justification for lack of full suppression required by Appendix R, Section III.G.3 in Fire Area G (Fire Zones 2150-C and 2136-I), which was approved by the NRC in 0CNA038328.
Manual fire suppression capability is available for this area  Low combustible loading in the area  Fire detection is provided in this area  The diesel fuel pumps can be cross-connected if necessary The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on fixed fire suppression in the area.
Unit Fire Area Name Description ANO-2 G Unit 2 Alternate Shutdown Areas Fire Zone Name Description 2136-I Health Physics Corridor 2150-C Old CPC Room
 
Reference Document 0CNA038328, 10 CFR 50, Appendix R Exemptions Approved, 3/22/1983, Attachment, Exemption IV, SER, Section 10.0
 
Evaluation This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-13 Licensing Action
 
Appendix R Exemption 13, FA - KK (Originally FA - B), Not Meeting III.G.2 Criteria Basis Date: 10/26/1988
 
Transitioned
? No  Basis: Exemption request per 0CAN088508 provides the justification for inability to provide at least 20' separation with area suppression/detection for the redundant diesel generator exhaust fans (Fire Zone 2114-I) as required by Appendix R, Section III.G.2, which was approved by the NRC in 2CNA108802.
Low combustible loading in the area  No intervening combustibles  3-hour rated fire door between redundant trains  Open roof that prevents smoke/heat accumulation Per engineering request ER-ANO-2002-0745-001, a fire in this fire zone is no longer analyzed to cause a loss of offsite power (i.e., the diesel generators are not needed for a fire in this area). Therefore, this exemption is not transitioned.
Unit Fire Area Name Description ANO-2 KK Unit 2 South Emergency Diesel Generator and Boric Acid Makeup Tank Fire Zone Name Description 2114-I EDG Air Intake Room
 
Reference Document
 
2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, SER, Section 2.0
 
Evaluation
 
This Fire Area was transitioned using updated analyses on where a loss of offsite power can occur. This exemption is no longer required.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-14 Licensing Action
 
Appendix R Exemption 13A, FA - QQ (Originally FA - B), Not Meeting III.G.2 Criteria Basis Date: 10/26/1988
 
Transitioned
? No  Basis: Exemption request per 0CAN088508 provides the following justification for inability to provide at least 20' separation with area suppression/detection for the redundant diesel generator exhaust fans (Fire Zone 2114-I) as required by Appendix R, Section III.G.2, which was approved by the NRC in 2CNA108802.
Low combustible loading in the area  3-hour rated fire door between redundant trains  Open roof that prevents smoke/heat accumulation  No intervening combustibles Per engineering request ER-ANO-2002-0745-001, a fire in this fire zone is no longer analyzed to cause a loss of offsite power (i.e., the diesel generators are not needed for a fire in this area). Therefore, this exemption is not transitioned.
Unit Fire Area Name Description ANO-2 QQ North Emergency Diesel Fire Zone Name Description 2114-I EDG Air Intake Room
 
Reference Document
 
2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, SER, Section 2.0 Evaluation
 
This Fire Area was transitioned using updated analyses on where a loss of offsite power can occur. This exemption is no longer required.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-15 Licensing Action
 
Appendix R Exemption 14, FA - G, Not Meeting III.G.3 Criteria Basis Date: 10/26/1988
 
Transitioned
? No  Basis: Exemption request per 0CAN088508 provides the justification for lack of suppression required by Appendix R, Section III.G.3 in Fire Area G (Fire
 
Zone 2199-G), which was approved by the NRC in 2CNA108802.
Manual fire suppression capability is available for this area  Moderate combustible loading in the area  Fire detection is provided in this area  Continuous manning in the area (Control Room Operators)
The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on fixed fire suppression in the area.
Unit Fire Area Name Description ANO-2 G Unit 2 Alternate Shutdown Areas Fire Zone Name Description 2199-G Unit 2 Control Room
 
Reference Document
 
2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, SER, Section 3.0
 
Evaluation
 
This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-16 Licensing Action
 
Appendix R Exemption 15, FA - EE-L (Originally a portion of FA - EE), Not Meeting III.G.2 Criteria Basis Date: 10/26/1988
 
Transitioned
? No  Basis: Exemption request per 0CAN088404 provides the justification for the lack of full coverage with an automatic suppression system, which was approved by the
 
NRC in 2CNA108802.
Manual fire suppression capability is available for this area  Moderate combustible loading in the area  Fire detection is provided in this area  Cable trays are not intervening combustibles between safe shutdown valves  Local manual operation of valves The NFPA 805 transition compliance strategy is a performance based approach that is in accordance with Section 4.2.4. This exemption is no longer required and will not be transitioned to the NFPA 805 licensing basis since the compliance strategy of Section 4.2.4 is not based on fixed fire suppression in
 
the area.
Unit Fire Area Name Description ANO-2 EE-L South Piping Penetration Rooms Fire Zone Name Description 2055-JJ Lower South Piping Penetration Room 2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room
 
Reference Document 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, SER, Section 4.0
 
Evaluation This Fire Area was transitioned using the performance based approach; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-17 Licensing Action
 
Appendix R Exemption 16, FA - YD, Not Meeting III.G.2 Criteria Basis Date: 10/26/1988
 
Transitioned
? No  Basis: Exemption request per 0CAN108710 provides the justification for lack of adequate separation of QCST level transmitters/cables in the yard area (FA -
 
YD), which was approved by the NRC in 2CNA108802.
Manual fire suppression capability is available for this area  Physical configuration of the area  Smoke and hot gas dissipation in open air The lack of adequate separation of QCST level transmitters/cables in the area is no longer a concern under NFPA 805 criteria due to the Technical Specification required large volume of water that must be maintained during power operation. This area remains deterministically compliant as no fire related plant shutdown is required with the postulated loss of this
 
instrumentation.
Unit Fire Area Name Description ANO-2 YD Miscellaneous Yard Locations Fire Zone Name Description YARD Miscellaneous Yard Locations
 
Reference Document 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, SER, Section 5.0
 
Evaluation This Fire Area was found to be deterministically compliant; therefore, this exemption is no
 
longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-18 Licensing Action
 
Appendix R Exemption 17, FA - NN, RCP Oil Collection, Not Meeting III.O Criteria Basis Date: 10/26/1988
 
Transitioned
? Yes  Basis: Exemption request per 0CAN088404 provides the following justification for inability to contain entire oil supply of all RCPs in lube oil collection system and meet the SSE requirements from Appendix R, Section III.O, which was approved by the NRC in 2CNA108802.
The RCP Oil Collection Systems at ANO-1 and 2 each contain two tanks. These tanks are each designed to hold the contents of one reactor coolant pump's lube oil inventory with margin. Oil leakage from the remaining pump in each RCS loop will be drained into the appropriate tank, until the tank capacity is reached, and then to an open curbing where it can be
 
safely contained. The ANO-2 RCP oil collection system, like Unit One's, was not designed to withstand an SSE. The Reactor Coolant Pump motor lube oil systems are integral with the pump motors. These motors, which are not seismically qualified, i.e., which are not required to function after a SSE, are seismically supported. The RCPs, RCP motors, and the integral lube oil systems contained within those pump motors are all designed, engineered, and installed such that a reasonable assurance of withstanding a Safe Shutdown Earthquake has been provided. Because the lubrication oil systems for the reactor coolant pumps are qualified to remain functional during and after a SSE, oil leaks should not result from such an event. Further, because the installed oil collection system itself is designed so that the dropping of its components during an SSE should not cause loss of operability of safety-related equipment nor cause a fire, the systems for each plant would not degrade safety features within containment.
Unit Fire Area Name Description ANO-2 NN Unit 2 Containment Building Fire Zone Name Description 2032-K Containment Building South Side 2033-K Containment Building North Side
 
Reference Document
 
0CAN088404, Results of Analysis Against NRC Cl arification/Interpretation of Appendix R to 10 CFR 50, 8/15/1984, Attachment, Section IV.B
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-19 Evaluation
 
0CAN088404, Attachment, Section IV.B concludes, "Because the lubrication oil systems for the reactor coolant pumps are qualified to remain functional during and after a SSE, oil leaks should not result from such an event. Further, because the installed oil collection system itself is designed so that the dropping of its com ponents during an SSE should not cause loss of operability of safety-related equipment nor cause a fire, the systems for each plant would not degrade safety features within containment."
Reference Document
 
2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, Section III & SER, Section 6.0
 
Evaluation
 
The SER (2CNA108802) in Section 6.3 notes, "On the basis that the lube oil system at ANO-2 is capable of withstanding the SSE without rupture and that the existing oil collection system will channel random leaks to a vented and closed container, the existing design conforms with the above staff guidance." 
 
NFPA 805 Chapter 3 Reference
 
3.3.12  Reactor Coolant Pumps
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-20 Licensing Action
 
Appendix R Exemption 18, FA - YD, Emergency Lighting, Not Meeting III.J Criteria Basis Date: 10/26/1988
 
Transitioned
? No  Basis: Exemption request per 0CAN088508 provides the following justification for lack of emergency lighting in the yard area (FA - YD), which was approved by the
 
NRC in 2CNA108802.
The diesel-backed security lighting provides adequate lighting and operations personnel are equipped with flashlights.
This exemption is no longer required because NFPA 805 does not require 8-hour battery backed emergency lights.
Unit Fire Area Name Description ANO-2 YD Miscellaneous Yard Locations Fire Zone Name Description YARD Miscellaneous Yard Locations
 
Reference Document 2CNA108802, Evaluation of Exemptions from the Technical Requirements of Appendix R, 10/26/1988, Attachment, Section III
 
Evaluation This Fire Area was found to be deterministically compliant; therefore, this exemption is no longer required under the new licensing basis.
 
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-21 Licensing Action
 
Appendix R Exemption 19, FA - NN, RCP Oil Fill Line, Not Meeting III.O Criteria Basis Date: 06/14/1997
 
Transitioned
? Yes  Basis: Exemption request per 2CAN129612 provides the following justification for inability to contain remote oil addition line leakage in the RCP lube oil collection system as required by Appendix R, Section III.O, which was approved by the
 
NRC in 2CNA069701.
In order to minimize the potential for an oil fire due to a leak from the lines of the remote addition system that do not have an oil collection system, the following compensatory actions will be taken each time oil is added:  Limit initial oil addition to two gallons. Verify that the two gallons of oil has reached the reservoir of the correct reactor coolant pump motor (i.e., verify that the oil has not leaked from the oil fill line). Add the remaining oil only after confirmation that the initial two gallons has reached the appropriate oil reservoir. Limit the total oil added to less than the amount calculated to result in an indicated reservoir level of 95 percent. Verify the oil addition funnel is empty prior to closing the lube oil manifold ball valve after oil has been added. Remove any oil in the drip pan under the lube oil manifold prior to exiting the containment building. Inspect for evidence of smoke following the oil addition. If smoke is detected, a fire brigade will be dispatched to the area. If at any point during oil addition the licensee determines that the oil is not reaching the desired location, it will terminate the activity and initiate a condition report to assess the situation.
Unit Fire Area Name Description ANO-2 NN Unit 2 Containment Building Fire Zone Name Description 2032-K Containment Building South Side 2033-K Containment Building North Side Reference Document
 
2CAN129612, 10 CFR 50, Appendix R Exemption Request - Reactor Coolant Pump Lube Oil System, 12/23/1996, Attachment Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-22 Evaluation "Remote Oil Addition System Description As a result of oil consumption during power operations, the need arises to periodically add oil to the reactor coolant pump (RCP) motor lube oil reservoirs. Prior to 1991, this was accomplished by transporting oil into the containment building cavities (inside the D-ring) and adding the oil using the oil fill connection on the motor. This resulted in both ALARA (an estimated 750-1000 mR per entry) and personnel safety concerns (e.g., heat stress and climbing ladders carrying containers of oil).
 
In an effort to minimize radiation exposure to maintenance personnel and resolve the personnel safety concerns, a temporary modification was initiated and installed on 2P32A reactor coolant pump. This modification installed a funnel at the 426' elevation outside the D-ring and routed a hose to the oil fill connection of the motor upper reservoir. Subsequent temporary modifications were added on 2P32B and 2P32D as the need arose. Oil addition efforts following the installation of these temporary modifications resulted in greatly reduced radiation dose (reduced
 
to 50-75 mR per entry).
 
A permanent remote oil addition system was installed in April 1994. This modification installed a gravity feed system to both RCP motors in each D-ring. The system currently consists of a ten gallon capacity funnel with spill protection, a lubrication oil (LO) addition valve manifold and isolation ball valves (with an oil collection pan underneath), 3/4" stainless steel (SS) tubing, and a SS flexible connection to the RCP motor reservoir fill connections.
 
From the outlet of the ball valves on the LO addition manifold, lines are routed into the D-ring through the feedwater pipe penetration and to their respective oil fill connection. All lines are seismically mounted to remove any II/I concerns. A 22" SS flexible hose is used between the routed tubing and the vendor supplied oil fill connection to minimize the effect of any translated vibration and thermal movement. Neither the 3/4" SS tubing nor the SS flexible hose are protected by an oil collection system. The connections to the motors are protected by installed oil collection system pans. The maximum oil level in the reservoirs is lower than the oil addition connection.
 
Tubing runs inside the D-ring are connected with compression fittings. Fittings of this type are highly reliable, especially when used in low pressure applications, and not typically subject to leakage. The minimum slope of the installed tubing is 1/4" per 12" run. This ensures the system drains following each use and remains a dry system. Following initial installation of the remote oil addition system, a system flush and a leak test were conducted to ensure the system functioned as designed.
 
Since the installation of the remote oil addition system, oil has been added to the RCP motor upper oil reservoirs on several occasions. During each oil addition, the appropriate increase in reservoir has been observed on the plant monitoring system (PMS) computer. In each case, the resultant indicated change in level verified the oil reached the reservoir. This system has been demonstrated leak tight and it is not likely that significant leakage would develop.
 
No indication of smoke, from either odor or visual observation, has been noted during containment entries to add oil. Additionally, smoke detectors located in the vicinity of each RCP motor have not indicated the presence of smoke after remote oil addition evolutions.
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-23 Consequences of Remote Oil Addition System Leakage
 
The routing of the oil addition tubing, as described above, is from the inside containment wall on the 426' elevation, under the grating, through the feedwater piping penetration, and through the cavities to the motors. The oil addition tubing is routed such that no leakage from the system could reach any fibrous blanket insulation located in the containment building. All insulation that leakage from the system could come into contact with is a SS reflective type. The expected maximum temperature on this type of insulation is less than 200 &deg;F under worst case design conditions and therefore, would not be an ignition source.
Should the system leak from the fittings around the valve manifold, the leakage would be retained in the drip tray located under the manifold. Personnel involved in the filling evolution would then remove the oil before it could become a fire hazard.
 
The ANO-2 Fire Hazards Analysis for the containment building considers the entire building as a single fire area. However, the containment building was divided into two zones for the purposes of evaluating the effects of fires. The zones were divided on the basis of clear space without intervening combustibles and provision of fire stops where cables provide a pathway between zones. The north half of the containment building (Zone 2033-K) contains cabling and equipment associated with safe shutdown that is separated by over 30 feet from cabling and equipment in the south half of the containment building (Zone 2032-K). The effects of postulated oil loss outside the D-ring during oil addition evolutions were specifically evaluated for each zone.
 
Due to the separation of redundant components, the oil addition lines for the "C" and "D" pumps (fire zone 2033-K) pose no threat to the safe shutdown capability.
 
The oil addition lines for the "A" and "B" pumps are located in fire zone 2032-K. This zone contains redundant channels of instrumentation for safe shutdown. Therefore, the effects of a potential fire from a spill occurring during oil addition were analyzed (even though spill prevention is provided at the funnel and manifold). The oil would tend to migrate to the containment floor where it would flow to the containment sump. Any oil leaked inside the D-ring would flow to the equipment drains under the RCPs. The temperature of the equipment in the area is expected to remain significantly below the autoignition temperature (700 &deg;F for R&O 68 oil and 635 &deg;F for R&O 32 oil). A pre-action suppression system and ionization smoke detection with control room alarms protect the cable trays in the containment building cable spreading areas. Two ionization type smoke detectors are installed in each RCP area inside the D-ring. The detection system provides early warning of possible fire conditions and the suppression system is designed to control the spread of fire, if it did occur.
 
The redundant safe shutdown components found in the containment building consist of the reactor coolant system (RCS) high point vents, the low temperature overpressurization (LTOP) valves, the shutdown cooling suction valves, and the following instrumentation: steam generator (SG) level, SG pressure, pressurizer level and pressure, neutron flux monitoring, and RCS temperature.
 
The SG level and pressure instruments and their associated cables are located on opposite sides of the reactor vessel and are of sufficient distance apart to preclude a RCP lube oil fire from affecting redundant channels of instrumentation. A similar condition is applicable to the RCS temperature instrumentation and the neutron flux
 
monitors.
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-24 In the unlikely event an oil spill did occur from an oil addition operation, it would be administratively limited to two (2) gallons. A fire resulting from a spill of two gallons of oil would have to travel 20 feet to reach the nearest safe shutdown component of interest (pressurizer pressure wide range transmitter). Due to the lack of oil addition system pressure, the size of the containment building (over 10,500 sq ft), the limited size of the postulated fire (two gallons of oil), the radiant fire scenario (target only sees 40% of the heat generated), and the fact that there is not a direct line of sight from the source to the target, it is not credible that damage will occur to redundant channels of safe shutdown instrumentation outside the D-ring.
At least two RCS high point vent valves would have to spuriously operate to create an uncontrolled vent path. Although this event is highly improbable, it can be mitigated by isolating letdown and then utilizing a charging pump which is capable of providing make-up in excess of the vent path losses.
Inside the containment building, the control cabling for the LTOP valves cannot produce a spurious operation. The power cables are deenergized, with the breakers outside containment. Since the valves are normally closed (i.e., their safe shutdown position), a fire cannot cause the valves to fail in the open position.
The shutdown cooling valves are not required to be opened until cold shutdown conditions. During normal operations, the breakers for these valves are locked open; thus preventing any spurious operations. The valves may be manually operated when required.
Other safe shutdown components are backed up by their redundant component(s) located outside the containment building. The necessity of the pressurizer heaters was analyzed and determined not to be necessary for safe shutdown.
RCS pressure boundary components consist of passive mechanical components such as heat exchanges, piping, tanks, manual valves, and check valves. These components are not fire sensitive and will not be damaged by credible plant fires based on the defense-in-depth fire protection philosophy at ANO.
Due to system design (slanting lines that do not retain oil) and the process for adding oil, the likelihood of an oil spill in excess of 2 gallons is very remote. Further, should a leak occur during filling, the amount of the spill would be much less than the contents from one RCP motor, which has been analyzed from a safe shutdown perspective and found to be acceptable.
Therefore, in the unlikely event of a fire involving the maximum postulated lube oil leak from one RCP, the ability to achieve and maintain safe shutdown is assured. Additionally, fire suppression equipment is readily available for response to a fire.
 
Compensatory Measures
 
In order to minimize the potential for an oil fire due to a leak from the lines of the remote addition system that do not have an oil collection system, the following actions will be taken each time oil is added through this system:
Initial oil addition will be limited to two gallons. The PMS will be utilized to verify that the two gallons has reached the reservoir of the correct RCP motor (the oil level can be determined to within 1/2 gallon).
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-25 Only after confirmation that the initial quantity reached the appropriate reservoir, will any remaining oil be added. The total oil volume to be added will be limited to less than the amount calculated to result in an indicated reservoir level of 95%. The oil addition funnel will be verified empty prior to closing the LO manifold ball valve after oil has been added. Should any oil collect in the drip pan under the LO manifold, it will be removed prior to exiting the containment building.
If at any point during a remote addition operation, it is determined that added oil is not reaching the desired location, the activity will be terminated and a condition report initiated to assess the situation. Personnel responsible for adding oil to the system also inspect for evidence of smoke following the oil addition. If smoke is detected, a fire brigade will be dispatched to the area. Fire suppression equipment is readily available for use in responding to such an event.
Conclusions The remote oil addition system is infrequently used during power operation. The system is dry when not in use. When utilized, the unpressurized system is limited to two gallons of oil for initial addition. Only after the oil is verified to have reached the appropriate RCP lube oil reservoir (confirmation of leakage integrity of the system), may additional oil be added.
Oil that may leak from this system could fall onto metal reflective insulation protecting RCS piping. The maximum temperature of the surface of this insulation is well below the autoignition temperature of the oil. Therefore, no credible ignition source is present.
 
Should a fire occur, no redundant trains of safe shutdown equipment will be affected due to the limited amount of oil available (two gallons) and the configuration of safe shutdown equipment.
Based on the above, the intent of 10CFR50 Appendix R (the ability to achieve and maintain safe shutdown of the plant in the event of a single fire) is accomplished without having a full oil collection system on the RCP remote oil addition system."  Reference Document 2CNA069701, Exemptions to Certain Requirements of 10 CFR 50, Appendix R, Section III.O, 6/14/1997, Attachment, ANO-2 SER, Section 3.0, 4.0, and 5.0
 
Evaluation "3.0 DISCUSSION The purpose behind the reactor lube oil collection system is to prevent a major lube oil fire from occurring inside of the reactor containment as a result of a lube oil leak from the RCPs.
Periodically, as a result of oil consumption during power operations, the licensee needs to add oil to the RCP motor lube oil reservoirs. Prior to 1991, the licensee accomplished oil addition by going into the containment building cavities (inside the D-rings) and adding the oil using the fill connection on the motor. This resulted in both ALARA (750-1000 mR per entry) and personnel safety concerns (e.g., heat stress and climbing ladders while carrying containers of oil).
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-26 In 1991, the licensee installed a temporary modification to resolve the ALARA and safety concerns. This modification installed a funnel outside of the D-ring an routed a hose to the oil fill connection of the motor upper reservoir. Oil addition efforts following the installation of the temporary modification resulted in reduced radiation dose (reduced to 50-75 mR per entry).
 
Later, in April 1994, the licensee installed a permanent remote oil addition system. This modification installed a gravity feed oil fill system to both RCP motors in each D-ring. The system consists of a ten gallon capacity funnel with spill protection, a lubrication oil addition valve manifold and isolation ball valves (with an oil collection pan underneath), 3/4 inch diameter stainless steel tubing with a stainless steel flexible connection to each coolant pump motor reservoir fill connection. All lines are seismically mounted. Neither the stainless steel tubing nor the flexible hose are protected by an oil collection system. The connections to the motors are protected by installed oil collection system pans. The maximum level in the reservoirs is lower than the oil addition connections. The minimum slope of the installed tubing is 1/4 inch per 12 inches of tubing run. This ensures that the system drains following each use and remains a dry system. This tubing is not pressurized. Oil is present in the tubing only during fill evolutions. By procedure, the licensee limits the initial oil addition to two gallons, until verification that the oil reservoir on the appropriate pump motor has shown a corresponding increase. This limits the amount of oil which could leak and cause a fire to two gallons.
The oil addition tubing is also routed such that no leakage from the system could reach any fibrous insulation located in the containment building. Should the system leak from the fittings around the valve manifold, the leakage would be collected by the drip tray located under the manifold. Personnel involved in the filling evolution would then remove the oil before it could become a fire hazard.
Any oil leakage outside of the reactor cavities would migrate to the containment sump and present no fire hazard to safe shutdown equipment. Any oil leakage inside of a reactor cavity would flow to the equipment drains under the reactor coolant pumps. The temperature in the area is expected to remain below the autoignition temperature of the lube oil.
A pre-action sprinkler system with an ionization-type smoke detection system with control room alarms protect the cable trays in the containment building cable spreading areas. In addition, two ionization-type smoke detectors are installed in the area of the reactor coolant pumps inside of each cavity. The detection system provides early warning of possible fire conditions and the suppression system is designed to control the spread of fire, if one were to occur.
In order to minimize the potential for an oil fire due to a leak from the lines of the remote addition system that do not have an oil collection system, the licensee will take the following compensatory actions each time oil is added.
 
Limit initial oil addition to two gallons. Verify that the two gallons of oil has reached the reservoir of the correct reactor coolant pump motor (i.e., verify that the oil has not leaked from the oil fill line). Add the remaining oil only after confirmation that the initial two gallons has reached the appropriate oil reservoir. Limit the total oil added to less than the amount calculated to result in an indicated reservoir level of 95 percent.
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-27 Verify the oil addition funnel is empty prior to closing the lube oil manifold ball valve after oil has been added. Remove any oil in the drip pan under the lube oil manifold prior to exiting the containment building. Inspect for evidence of smoke following the oil addition. If smoke is detected, a fire brigade will be dispatched to the area.
If at any point during oil addition the licensee determines that the oil is not reaching the desired location, it will terminate the activity and initiate a condition report to assess the situation.
10 CFR 50.48 requires nuclear power plants licensed prior to January 1, 1979, to implement 10 CFR Part 50, Appendix R, Section III.O. Section III.O requires that the licensee have a collection system "capable of collecting lube oil from all pressurized and unpressurized leakage sites in the reactor coolant pump lube oil systems."  It also specifies that "leakage points to be protected shall include lift pump and piping, overflow lines, lube oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and lube oil reservoirs where such features exist on the reactor coolant pumps."  The underlying purpose of the rule is to ensure that leaking oil will not lead to a fire which could damage safe shutdown systems during normal or design basis accident conditions.
 
4.0  EVALUATION The staff was concerned that damage to the stainless steel tubing could result in a lube oil leak into the containment during the addition of lube oil, that the leaked lube oil would ignite, and that the resulting fire could affect the ability to achieve and maintain post-fire safe shutdown conditions.
Due to compensatory actions taken by the licensee during oil addition, the maximum potential oil leak is two gallons. The maximum oil leak is limited to two gallons primarily because the fill operation will be performed in batches. No more than two gallons of oil will be in the oil fill lines at any one time. The ANO-2 fire hazards analysis for the containment building considers the entire building as a single fire area. The containment building is divided into two fire zones.
The zones were divided on the basis of clear space without intervening combustibles and provision of fire stops where cables provide a pathway between zones.
Due to separation of redundant components, a leak from the oil addition lines for the "C" and "D" reactor coolant pumps pose no significant threat to the safe shutdown capability. The fire zone containing the "A" and "B" reactor coolant pumps contains redundant channels of instrumentation for safe shutdown. Outside of the reactor cavities, oil from a leak would migrate to the containment floor and then to the containment sump. In the event a fire were to occur, it would be limited to two gallons of oil. The oil would have to spread twenty feet to reach the nearest safe shutdown component (pressurizer pressure wide range pressure transmitter). Due to the lack of oil addition system pressurization, the size of the containment (10,500 sq. ft.), the limited fire size caused by two gallons of lube oil, and no direct line of sight between the transmitters and postulated fire location, there is reasonable assurance that damage would not occur to safe shutdown equipment.
Fire detection and manual fire suppression equipment is available in the vicinity of the lube oil fill lines. In the event of a fire, it is expected that the detector will alarm and the fire brigade will respond to extinguish the fire in its incipient stages. This provides further assurance that a worst-case postulated fire would not damage safe shutdown equipment.
Arkansas Nuclear One - Unit 2 Att. K - Existing Licensing Action Transition Enclosure 1 to 2CAN121202 Page K-28
 
==5.0  CONCLUSION==
 
Therefore, contingent on the use of the compensatory measures that are itemized in the licensee's exemption request dated December 23, 1996, the staff concluded that the design of the oil filling system and the level of protection provided during oil fill operations provide reasonable assurance that a lube oil fire will not occur. The staff also concluded that in the event of a worst-case postulated fire, due to not having a lube oil collection system for the reactor coolant pump lube oil fill lines, it would be of limited magnitude and extent. In addition, such a fire would not cause significant damage in the containment building and would not prevent the operators from achieving and maintaining safe shutdown conditions. The staff concluded, therefore, that special circumstances are present in that an oil collection system for the reactor coolant pump lube oil fill lines is not necessary to achieve the underlying purpose of the rule, and that an exemption as described herein is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security. The licensee's request for exemption should, therefore, be granted."  NFPA 805 Chapter 3 Reference 3.3.12  Reactor Coolant Pumps
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-1 L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))
In accordance with 10 CFR 50.48(c)(2)(vii) Performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard.
 
In accordance with NFPA 805, Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied.
 
In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).
Entergy requests formal approval of performance based exceptions to requirements in Chapter 3 of NFPA 805 as follows.
NFPA 805 Section 3.3.3
 
NFPA 805 Section 3.3.3 states:
"Interior Finishes. Interior wall or ceiling finish classification shall be in accordance with NFPA 101, Life Safety Code
, requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requirements for Class I interior floor finishes."
ANO utilizes an epoxy floor coating system that does not meet the exact requirements of NFPA 805, Section 3.3.3.
 
NFPA 101 requirements for interior floor finishes state that the floor finish shall be characterized by a critical radiant flux not less than 0.45 W / cm
: 2. In addition, the NRC issued Information Notice (IN) 2007-26 to address the combustibility of epoxy floor coatings at commercial nuclear
 
power plants. Per IN 2007-26, the NRC defined a non-combustible material as:
: a. A material which in the form in which it is used and under the conditions anticipated, will not ignite, burn, support combustion, or release flammable vapors when subjected to fire or heat; and b. Material having a structural base of noncombustible material, as defined in a., above, with a surfacing not over 1/8-inch thick that has a flame spread rating not higher than 50 when measured using the test protocol of American Society for Testing and Materials (ASTM) E 84, "Standard Test Method for Surface Burning Characteristics of Building Materials.
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-2 NFPA 805 has re-defined the IN 2007-26 definition of non-combustible material to limited combustible material:
 
"Material that, in the form in which it is used, has a potential heat value not exceeding 3500 Btu/lb (8141 kJ/kg) and either has a structural base of noncombustible material with a surfacing not exceeding a thickness of 1/8 in. (3.2 mm) that has a flame spread rating not greater than 50, or has another material having neither a flame spread rating greater than 25 nor evidence of continued progressive combustion, even on surfaces exposed by cutting through the material on any plane."  NFPA 805 defines non-combustible material as:
 
"Material that, in the form in which it is used and under the conditions anticipated, will not ignite, burn, support combustion, or release flammable vapors when subjected to fire or heat."  A previous ANO evaluation of the acceptability of the epoxy floor coatings was performed in response to NRC published IN 2007-26, August 13, 2007, regarding combustibility of epoxy floor coatings at commercial nuclear facilities. The results are documented in CR-ANO-C-2008-01315 (excerpts below).
 
ANO evaluated coating samples taken from areas containing safety related equipment to determine the contribution that epoxy floor coating may have to combustible loads in safety related areas of the plant.
 
"The energy required to support combustion of the floor coatings used at ANO will not be produced by an incipient stage compartment fire. Direct impingement of flame or heat onto the coatings will not cause propagation of flame beyond the influence zone of the heat source, as exhibited with incidental hot work contact with floor surfaces. Manual or automatic suppression will provide protective cover to preclude floor coating involvement in fire severity should a fire proceed past the incipient stage."
"The epoxy floor coatings currently applied at the ANO site could potentially be considered a slight contributor, typically of less than 3 minutes, to fire severity, only if the compartment progressed to flashover conditions and automatic or manual suppression is never attempted. Considering the epoxy floor materials used and the conditions anticipated, it can be reasonably concluded that the epoxy floor coatings of the type utilized at ANO do not present a primary fire hazard, will not propagate fire from one fire area to another, or
 
exacerbate the severity of a compartment fire."
Basis for Request:
The coatings permitted at ANO, with the exception of Duochem 9400, are either NFPA Class A qualified or ASTM E84 tested with a flame spread index less than 50. All epoxy floor coatings have been determined by ANO evaluation to have a negligible contribution with regard to combustible loading. In addition, the epoxy coating is on the floor. The ASTM E-84 test is conducted with the material on the ceiling of a tunnel. This configuration would allow the flame to directly impinge on the ceiling surface, enhancing flame spread. With the material on the floor, the heat flux to the surface is much less than would be expected in the ceiling configuration since the convective flame is directing the heat away from the surface. This would mean that the overall flame spread would be expected to be much less, even with a slightly greater thickness.
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-3 Acceptance Criteria Evaluation:
Nuclear Safety and Radiological Release Performance Criteria:
The use of epoxy floor coating does not affect nuclear safety as it in general meets the definition of a limited combustible material with isolated thickness excesses. The floor coating materials were evaluated to have a negligible effect on combustibility. Application of epoxy floor coatings is controlled via ANO procedures to ensure that the amount of material does not add appreciable amounts of combustible material to the plant. Therefore there is no impact on the nuclear safety performance criteria.
 
The use of epoxy floor coatings has no impact on the radiological release performance criteria.
The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the floor coating materials. The floor coatings do not change the radiological release evaluation performed that potentially contaminated water is contained and smoke monitored. Floor coatings do not add additional radiological materials to the area or challenge systems boundaries that contain such.
 
Safety Margin and Defense-in-Depth:
The use of epoxy floor coating does not affect safety margin as it in general meets the definition of a limited combustible material with isolated thickness excesses. The floor coating materials were evaluated to have a negligible effect on combustibility. Application of epoxy floor coatings is controlled via ANO procedures. These precautions and limitations on the use of these materials have been defined by the limitations of the analytical methods used in the development of the fire probabilistic risk assessment (PRA). Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.
The three echelons of defense-in-depth are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The use of epoxy floor coatings does not affect echelons 1, 2 and 3. The use of epoxy floor coatings does not directly result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.
 
==
Conclusion:==
 
NRC approval is requested for the use of epoxy floor coatings as a performance-based method that provides an equivalent level of fire protection to NFPA 805, Section 3.3.3.
 
ANO-2 determined that the performance based approach satisfies the following criteria:
Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release  Defense in Depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)  Safety Margin
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-4 NFPA 805 Section 3.3.5.1 NFPA 805 Section 3.3.5.1 states:
"Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays with solid metal top and bottom covers."
ANO has wiring above suspended ceilings that may not comply with the requirements of this code section.
 
Suspended ceilings and their supports are non-combustible and combustibles in concealed spaces are minimal.
 
The majority of ANO areas currently with suspended ceilings inside the NFPA 805 defined power block are office areas in the Turbine and Auxiliary buildings. These areas are not risk significant with the exception of the Control Rooms. The Control Rooms previously identified cabling above the suspended ceiling. The quantity of video/communication/data cabling above the suspended ceilings in the Control Rooms is very low and results in limited combustible loading. In addition, the existing fire detection capability and/or the Control Room Operators who are continuously present in the area would identify the presence of smoke.
 
These areas are assumed to have wiring above the suspended ceilings including that needed for power, control, and video/communication/data. Power and control cables at ANO are IEEE-383-1974 or equivalent. FAQ 06-0022 identified acceptable electrical cable construction tests for all areas of the power block and the majority of plant cables meet those requirements.
 
Video/communication/data cables that have been field routed above suspended ceilings are low voltage. Existing video/communication/data cabling may not be plenum rated, but is not generally susceptible to shorts that would result in a fire, and meets plant specific requirements documented in plant procedure OP-6030.109, "Installation of Electrical Cable & Wire."
Basis for Request:
The basis for the approval request of this deviation is:
 
The NFPA 805 requirement is excessive in that plenum rating should not be applied to wiring above suspended ceilings that are not used as a plenum and have stagnant air versus flowing air. Only a limited amount of the cable installed above the suspended ceilings in these areas is not rated for plenum use, IEEE-383-1974 equivalent, or routed in conduit. The cable is low voltage (less than 480V) and, therefore, less susceptible to self-ignition and electrical shorts that could result in a fire in the enclosed space. There are no additional ignition sources in the areas above the suspended ceilings. For the cables that do not meet the NFPA 805, Section 3.3.5.1 criteria, the majority meet one of the cable qualifications listed within FAQ 06-0022. Plant procedure OP-6030.109, "Installation of Electrical Cable & Wire," contains adequate guidance to ensure suitable cable qualification criteria was provided and is maintained.
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-5 Acceptance Criteria Evaluation:
Nuclear Safety and Radiological Release Performance Criteria:
The location of wiring above suspended ceilings does not affect nuclear safety. Power and control cables comply with IEEE-383 or equivalent. Other wiring, while it may not be in armored cable, in metallic conduit, or plenum rated, is low voltage cable not susceptible to shorts that would result in a fire. Therefore, there is no impact on the nuclear safety performance criteria.
The location of cables above suspended ceilings has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the type of cables or locations of suspended ceilings. The location of cables does not change the radiological release evaluation performed that potentially contaminated water is contained and smoke monitored. The cables do not add additional radiological materials to the area or challenge systems boundaries that contain such.
 
Safety Margin and Defense-in-Depth:
Power and control cables comply with IEEE-383-1974 or equivalent. The use of these materials has been defined by the limitations of the analytical methods used in the development of the FPRA. Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.
The three echelons of defense-in-depth are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The prior introduction of non-listed video/communication/data cables routed above suspended ceilings does not impact fire protection defense-in-depth. Echelon 1 is maintained by the current cable installation procedures documenting the requirements of OP-6030.109. The introduction of cables above suspended ceilings does not affect Echelons 2 and 3. The video/communication/data cables routed above suspended ceilings do not result in compromising automatic fire suppression functions, manual fire suppression functions, fire protection for systems and structures, or post-fire safe shutdown capability.
 
==
Conclusion:==
 
NRC approval is requested to permit the pres ence of cable located above the suspended ceilings located in the power block which do not meet the requirements of NFPA 805, Section 3.3.5.1. The cabling is not enclosed in metal conduit, is not armored, is not enclosed in metal cable trays, and is not plenum rated cable. Adequate controls for such cabling are provided using existing plant procedure OP-6030.109, "Installation of Electrical Cable & Wire."
ANO-2 determined that the performance based approach satisfies the following criteria:
Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release  Defense in Depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)  Safety Margin
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-6 NFPA 805 Section 3.3.5.2 NFPA 805 Section 3.3.5.2 states:
"Only metal tray and metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components."
Installation of raceway systems is addressed in approved procedures. Cable tray and conduit material is primarily of substantial metal construction. However, use of Schedule 40 PVC is allowed by procedure for underground and embedded applications.
 
Site procedures state:
 
"Underground concrete encased conduits, sizes three (3) inches and larger shall be of an approved heavywall ABS or Schedule 40 PVC. Slab and wall embedded conduits three (3) inches and larger shall be Schedule 40 PVC or rigid steel as required by the Plant Data Management System. All conduits two (2) inches and smaller and all exposed conduits shall be rigid steel."
Basis for Request:
The basis for the approval request of this deviation is:
 
Access points to embedded conduit are required to be rigid steel. The nonmetallic conduit is used only in concrete embedded applications, thus providing physical protection and separation for the conduit. The plastic conduit, while a combustible material, is not subject to flame/heat impingement from an external source which would result in structural failure, contribution to fire load, and/or damage to the circuits contained within where the conduit is embedded in concrete. Failure of circuits within the conduit resulting in a fire would not result in damage to external targets. The NEC allows use of Rigid Nonmetallic Conduit for underground and embedded applications.
Acceptance Criteria Evaluation:
 
Nuclear Safety and Radiological Release Performance Criteria:
 
The use of nonmetallic conduit for raceways embedded in concrete is allowed by NFPA 70, National Electric Code (NEC), and provides adequate physical and electrical protection for cables. The use of plastic conduit in embedded locations does not affect nuclear safety as the material in which conduits are run within an embedded location are not subject to the failure mechanisms potentially resultant in circuit damage or resultant damage to external targets.
Therefore there is no impact on the nuclear safety performance criteria.
 
The use of plastic conduit in embedded installations has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-7 not dependent on the type of conduit material. The conduit material does not change the radiological release evaluation performed that potentially contaminated water is contained and smoke monitored. The conduits do not add additional radiological materials to the area or challenge systems boundaries that contain plastic conduits.
 
Safety Margin and Defense-in-Depth:
 
The plastic conduit material is embedded in a non-combustible configuration. The use of these materials has been defined by the limitations of the analytical methods used in the development of the FPRA. Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.
 
The three echelons of defense-in-depth are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The use of plastic conduit in embedded installations does not impact fire protection defense-in-depth. The plastic conduit in embedded installations does not affect echelons 1, 2 and 3. The plastic conduits do not directly result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire safe shutdown capability.
 
==
Conclusion:==
 
NRC approval is requested for the use of nonmetallic conduit for raceways embedded in concrete.
 
ANO-2 determined that the performance based approach satisfies the following criteria:
 
Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release  Defense in Depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)  Safety Margin NFPA 805 Section 3.3.12 (1)
NFPA 805 Section 3.3.12(1) states:
 
"The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oil from all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil system."
The ANO oil collection system is designed and was reviewed in accordance with 10 CFR 50 Appendix R, Section III.O to collect leakage from pressurized and non-pressurized leakage sites in the RCP oil system. This may not include collection of oil mist as a result of pump/motor operation. Oil misting is not leakage due to equipment failure, but an inherent occurrence in the operation of large rotating equipment. It is normal for large motors to lose some oil through seals and the oil to potentially become 'atomized' in the ventilation system. This atomized oil Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-8 mist can then collect on surfaces in the vicinity of the RCP as the pump design is not completely sealed to permit airflow for cooling. The oil mist resulting from normal operation will not adversely impact the ability of a plant to achieve and maintain safe shutdown even if ignition occurred. Redundant RCPs are available to achieve and maintain safe shutdown and detection is provided in the immediate vicinity of the pumps in order to detect a fire should one occur.
 
In addition, NRC GL 86-10, Implementation of Fire Protection Requirements, dated April 24, 1986; Question 6.2 (shown below) discussed oil dripping. The response concluded that there was no concern with oil consumption (which is an oil misting phenomena), but that an oil fire started from a pressurized leakage point and/or spilled leakage should be addressed.
Question 6.2 It would appear that a literal reading of Section III.0 regarding the oil collection system for the reactor coolant pump could be met by a combination of seismically designed splash shields and a sump with sufficient capacity to contain the entire lube oil system inventory. If the reactor coolant pump is seismically designed and the nearby piping hot surfaces are protected by seismically designed splash shields such that any spilled lube oil would contact only cold surfaces, does this design concept conform to the requirements of the rule?
Response  If the reactor coolant pump, including the oil system, is seismically designed and the nearby hot surfaces of piping are protected by seismically designed splash shields such that any spilled lube oil would contact only cold surfaces, and it could be demonstrated by engineering analysis that sump and splash shields would be capable of preventing a fire during normal and design basis accident conditions, the safety objective of Section III.0 would be achieved. Such a design concept would have to be evaluated under the exemption process. The justification for the exemption should provide reasonable assurance that oil from all potential pressurized and unpressurized leakage points would be safely collected and drained to the sump. The sump should be shown capable of safely containing all of the anticipated oil leakage. The analysis should verify that there are no electric sources of ignition.
The requirement for holding the inventory of the RCP lubricating oil system is not met without tank overflow, but an exemption for this configuration which keeps the overflow oil away from potential ignition sources was previously approved by the NRC for ANO-2.
 
Correspondence 2CNA108802 includes the following discussion:
 
6.1 Exemptions Requested The licensee requested approval of exemptions from the technical requirements of Section III.O of Appendix R to 10 CFR Part 50 to the extent that it requires the reactor coolant pump (RCP) oil collection system to be sized to hold the contents of the entire lube oil system for all pumps and to be designed to withstand a safe shutdown earthquake (SSE).
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-9
 
===6.2 Discussion===
The RCP Oil Collection Systems at each unit contains two tanks. These tanks are each designed to hold the contents of one reactor coolant pump's lube oil inventory with margin. Oil leakage from the remaining pump in each RCS loop will be drained into the appropriate tank, until the tank capacity is reached, and then go to an open curbing where it can be safely contained. The system is located above the floor of the Containment Building. Safe shutdown circuitry is routed approximately forty feet above the elevation outside the primary shield walls containing the reactor, RCPs, and other primary system components. The shield wall separates the heavy concentrations of safe shutdown circuitry in the electrical penetration areas from the RCPs and the Oil Collection System itself. Additionally, that circuitry is protected by localized automatic fire suppression and detection capability. The Reactor Coolant Pump motor lube oil systems are integral with the pump motors. The licensee stated in the August 15, 1984 submittal, that the lube oil systems are qualified to remain functional during and after an SSE. 6.3 Evaluation The technical requirements of Section III.O of Appendix R have not been met because the oil collection system for the RCPs has not been sized to hold the oil from all of the pumps and is not seismically designed.
Generic Letter 86-10 states:
  "Where the RCP lube oil system is capable of withstanding the safe shutdown earthquake (SSE), the analysis should assume that only random oil leaks from the joints could occur during the lifetime of the plant. The oil collection system, therefore, should be designed to safely channel the quantity of oil from one pump to a vented closed container. Under this set of circumstances, the oil collection system would not have to be seismically designed."  On the basis that the lube oil system at ANO-2 is capable of withstanding the SSE without rupture and that the existing oil collection system will channel random leaks to a vented and closed container, the existing design conforms with the above staff guidance.
 
===6.4 Conclusion===
Based on the above evaluation, the licensee's alternative design of the oil collection system provides an equivalent level of safety to that achieved by compliance with Section III.O of Appendix R. Therefore, the licensee's request for exemption should be approved. An external RCP oil fill system without oil spill collection on the supply lines is addressed in an exemption that was approved by the NRC for ANO-2 due to administrative controls during the fill process.
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-10 Correspondence 2CNA069701 includes the following discussion:
5.0 - CONCLUSION Therefore, contingent on the use of the compensatory measures that are itemized in the licensee's exemption request dated December 23, 1996, the staff concluded that the design of the oil filling system and the level of protection provided during oil fill operations provide reasonable assurance that a lube oil fire will not occur. The staff also concluded that in the event of a worst-case postulated fire, due to not having a lube oil collection system for the reactor coolant pump lube oil fill lines, it would be of limited magnitude and extent. In addition, such a fire would not cause significant damage in the containment building and would not prevent the operators from achieving and maintaining safe shutdown conditions. The staff concluded, therefore, that special circumstances are present in that an oil collection system for the reactor coolant pump lube oil fill lines is not necessary to achieve the underlying purpose of the rule, and that an exemption as described herein is authorized by law, will not present or (sic) undue risk to public health and safety, and is consistent with the common defense and security.
The licensee's request for exemption should, therefore, be granted.
 
Historically, there have been no fires attributed to oil misting based on normal operation in the industry. Fires have occurred due to oil leakage from equipment failure such as cracked welds on piping or inadequate collection pan design. ANO does not have a history of significant oil loss from the RCPs as a result of oil misting or oil leakage that is not contained by the properly designed and installed oil leakage collection system.
 
Basis for Request:
The basis for the approval request of this deviation is:
 
The oil collection system is designed to collect leakage from pressurized and nonpressurized leakage sites in the RCP oil system. Oil misted from normal operation is not leakage; it is normal motor oil consumption. Oil misted from normal operation does not significantly reduce the oil inventory. The oil historically released as misting does not account for an appreciable heat release rate or accumulation near potential ignition sources or non-insulated reactor coolant piping. RCPs are not required to achieve or maintain fire safe shutdown.
Acceptance Criteria Evaluation:
 
Nuclear Safety and Radiological Release Performance Criteria:
 
The oil mist resulting from normal operation will not adversely impact nuclear safety. There are redundant RCPs available as necessary. In addition, the RCPs are not required to achieve and maintain fire safe shutdown. Therefore there is no impact on the nuclear safety performance criteria.
 
The potential for oil mist from the RCPs has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials. The entire Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-11 Containment Building in which the RCPs are located is an environmentally sealed radiological area during power operations. The oil mist does not add additional radiological materials to the area or challenge systems boundaries that contain such.
Safety Margin and Defense-in-Depth:
 
The oil mist resultant from normal operation will not adversely impact the ability of a plant to achieve and maintain fire safe shutdown even if ignition occurred. There are redundant RCPs, however the RCPs are not required to achieve and maintain fire safe shutdown. The use of this equipment has been defined by the limitations of the analytical methods used in the development of the FPRA. Therefore, the inherent safety margin and conservatisms in these
 
methods remain unchanged.
 
The three echelons of defense-in-depth are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The potential for oil mist from RCPs does not impact fire protection defense-in-depth. Echelon 1 is maintained by the oil collection system and RCP design. The introduction of a small amount of oil misting does not affect echelons 2 and 3. The potential for oil mist from the RCPs does not result in compromising automatic fire suppression functions, manual fire suppression functions, fire protection for systems and structures, or post-fire safe shutdown capability.
 
== Conclusion:==
 
NRC approval is requested for the potential of oil misting from the RCPs due to normal motor consumption not captured by the oil collection system designed for pressurized and non-pressurized leakage and spillage.
ANO-2 determined that the performance based approach satisfies the following criteria:
 
Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release  Defense in Depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)  Safety Margin NFPA 805 Section 3.5.3, NFPA 20 (1969) Sections 457, 511c, 626a, 626d.e2 and 626d.e5 NFPA 805 Section 3.5.3 states:
 
"Fire pumps, designed and installed in accordance with NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of the largest pump or pump power source."
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-12 NFPA 20 (1969) Section 457 states:
Conformance: Motors furnished for centrifugal fire pump use shall be guaranteed to conform to these specifications.
NFPA 20 (1969) Section 511c states:
 
511c. All controllers shall be specifically approved for fire pump service.
 
ANO does not meet NFPA 20 Sections 457 or 511c that require the electric fire pump motor and electric fire pump controller to be Underwriters Laboratories, Inc. (UL) Listed/Approved for fire
 
pump service.
 
At the time of purchase October 30, 1969, in accordance with the Purchase Order Specification, the electric drive motor for the electric fire pump was not available as an UL listed motor for fire pump service; therefore, the motor could not be purchased as UL listed due to the larger service requirements associated with the fire pump. The Purchase Order Specification states that "UL will allow us to use a 350 HP motor for this rating."  The 3 phase / 60 cycles / 4160 VAC /
1770 RPM rated at 400 HP motor was ordered to meet the fire pump size requirements; nevertheless, the fire pump motor in use is not UL listed for fire pump service. A similar issue existed for the fire pump controller. However, the fire pump controller was evaluated to meet the design data requirements needed for the size and type for the electrically driven fire pump and drive motor.
 
NFPA 20 (1969) Section 626a states:
626a.General. The battery shall have sufficient capacity, at 40 &deg;F., to maintain the engine manufacturers recommended cranking speed during the following 6-minute cycle (15 seconds crank and 15 seconds rest in 12 consecutive cycles). The fire pump manufacturer shall provide a certification that the battery which was furnished complies with this requirement.
NFPA 20 (1969) 626d.e2 and 626d.e5 states:
 
Battery Chargers. The rectifier shall be of the semiconductor type, and the charger for the lead acid battery shall be capable of delivering a current within the range of 50 to 100 percent of the 20-hour discharge rate of the battery. ANO does not meet NFPA 20 Sections 626a, 626d.e2 and 626d.e5 for the Cummins Diesel Engine controller, since vendor documents do not identify a certification for the batteries and do not identify the discharge rate of the lead acid batteries. The vendor manual does identify the battery charger rectifier as being of a semiconductor type (silicone diode rectifier).
 
The vendor manual and design drawing identify the Cummins Diesel Fire Pump Engine with two lead acid battery banks D08 and D09. The battery charging rectifier function is to automatically adjust its output to the battery's requirement and to the demands of the indicating lamps that draw small amounts of current during when in stand-by. Vendor documents do not identify the battery discharge rate for the lead acid batteries.
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-13 The vendor manual for the diesel engine fire pump controller states that this equipment is UL Listed and Factory Mutual Research Corporation (FM) Approved for fire service. The vendor's diesel engine fire pump controller is manufactured, inspected, and tested to obtain UL listing and FM approvals for fire pump service. The fire pump controller sub-components (battery charger, relays, and etc.) were certified by the vendor for fire pump service. In addition, a review of historical fire pump testing found no issues identified by maintenance or during the diesel fire pump test, with battery problems related to battery discharge that would impact engine start.
 
Basis for Request:
The basis for approval request of this deviation is:
 
The electrical fire pump configuration required the larger size 4160 VAC fire pump motor and the 4160 VAC fire pump controller, which was not UL Listed/Approved for fire pump service in 1969. In addition, historical evidence and procedural testing requirements have shown that the 4160 VAC electric motor, electric fire pump, and electric fire pump controller configuration used at ANO, while not in explicit agreement with the code requirement for a UL Listing, meets the intent of electrically driven fire pump design size, type, and function. The electric driven fire pump and electr ic pump controller was manufactured in accordance with the National Electrical Code (NEC). The electrical fire pump configuration meets the demands for the fire protection water supply system at ANO. In review of ANO documents, no issues were identified in association with past diesel fire pump tests, specifically with battery problems related to the rectifiers or battery discharge that would prevent the engine from starting. The vendor manual for the diesel engine fire pump controller states that this equipment is UL Listed and FM Approved for fire service. The diesel fire pump meets the demands for the fire protection water supply system. Acceptance Criteria Evaluation:
 
Nuclear Safety and Radiological Release Performance Criteria
:
The 4160 VAC fire pump motor and the 4160 VAC electrical fire pump controller were not UL Listed/Approved for fire pump service at the time of purchase in 1969 due to UL not having the high voltage 4160 VAC electric fire pump motor and controller rated for fire service in 1969.
The vendor manual for the diesel engine fire pump controller states that this equipment is UL Listed and FM Approved for fire service. The vendor's diesel engine fire pump controller is manufactured, inspected and tested to obtain UL listing and FM approvals for fire pump service.
The fire pump controller sub-components (battery charger, relays, and etc.) were certified by the vendor for fire pump service. In addition, a review of historical fire pump testing found no issues identified by maintenance or during the diesel fire pump test with battery problems related to battery discharge that would impact engine start.
 
The deviations described above have no impact on the nuclear safety performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-14 A radiological release review was performed as, documented in Table G-1, based on the manual fire suppression activities in areas containing, or potentially containing, radioactive materials and is not impacted by the motor driven fire pump and fire pump controller purchased as not UL listed/approved for fire pump service in 1969. Therefore, this deviation has no impact on radiological controlled areas (RCAs) or the radiological release performance criteria.
 
Safety Margin and Defense-in-Depth
:  The fire protection water supply system has redundant capacity to supply the demands of the system. Therefore, the safety margin inherent in the analysis for the fire event has been
 
preserved.
The three echelons of defense-in-depth are to 1) prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control, and extinguish fires that do occur, thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The pumps (electric fire pump or diesel fire pump), at 100 percent flow rate and pressure, have the excess capacity to supply the demands of the fire protection system in addition to the greatest hose reel demand and, therefore, do not affect echelons 1, 2 and 3.
Conclusion
:  NRC approval is requested for the aforementioned deviation to allow the use of 4160 VAC electric fire pump motor and 4160 VAC electric fire pump controller, since the motor and electric fire pump controller were ordered to meet the appropriate electric fire pump size, type, and configuration requirements. However, at the time of purchase in 1969, the fire pump motor and controller were not UL listed for fire pump service.
NRC approval is also requested for the aforem entioned deviation related to the diesel engine fire pump vendor documents, which do not list or identify a certification for the batteries and the battery charger discharge rate of the lead acid batteries. The vendor manual for the diesel engine fire pump controller states that this equipment is UL Listed and FM Approved for fire service. The vendor's diesel engine fire pump controller is manufactured, inspected and tested to obtain UL listing and FM approvals for fire pump service. The fire pump controller sub-components (battery charger, relays, and etc.) were certified by the vendor for fire pump service. In addition, a review of historical fire pump testing found no issues identified by maintenance or during the diesel fire pump test with battery problems related to battery discharge that would impact engine start.
Historical evidence and procedural testing requirements have shown that the existing configuration meets the intent of electrically driven fire pump and the diesel fire pump design function for ANO. ANO-2 has determined that the performance-based approach satisfies the following criteria:
Satisfies the performance, goals performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release  Defense in Depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)  Safety Margin
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-15 NFPA 805 Section 3.5.16 NFPA 805 Section 3.5.16 states:
"The fire protection water supply system shall be dedicated for fire protection use only.
Exception No. 1:  Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.
Exception No. 2:  Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section."
NFPA 24 (1995) Section 5-7 states:
 
"Domestic Service Use Prohibited. The use of hydrants and hose for purposes other than fire-related services shall be prohibited."
The ANO fire protection water supply system is used for installation of a temporary fire pump to allow both units to supply a protracted and continual supply of cooling water during unit outages
 
when the Auxiliary Cooling Water (ACW) system is removed from service. Past practices of allowing use of the Fire Water system for non-fire water demands during outages have been authorized by engineering and incorporated into Operations procedures.
 
Significant margin exists in the fire protection water supply system above that required for suppression system demands. EC-27142 provides an evaluation that addresses the use of fire water for non-fire issues based on the results of a hydraulic model.
Basis for Request:
The basis for the approval request of this deviation is:
 
The fire protection water supply system has excess capacity. The use of the fire protection water supply system is procedurally controlled.
Acceptance Criteria Evaluation:
 
Nuclear Safety and Radiological Release Performance Criteria
:
The use of the fire protection water supply system for temporary cooling is evaluated as a temporary modification and controlled by approved procedures. The fire protection water supply system has excess capacity to supply the demands of the system to the greatest hose reel demand as evaluated by EC-27142 using a hydraulic model. Administrative controls consisting of procedural direction or continuously stationed individual ensure that a hose station or hydrant is secured or otherwise made available in the event of a fire. Therefore, use of the fire protection water supply system for temporary cooling has no impact on the nuclear safety
 
performance criteria.
 
Arkansas Nuclear One - Unit 2 Att. L - NFPA 805 Chapter 3 Requirements for Approval Enclosure 1 to 2CAN121202 Page L-16 The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the alternate use of the fire water supply system. Therefore, the use of the fire protection water supply system for non-fire protection uses, including the use of hydrants and hoses for purposes other than fire, has no impact on radiological controlled areas or the radiological release
 
performance criteria.
 
Safety Margin and Defense-in-Depth:
The fire protection water supply system has excess capacity to supply the demands of the system to the greatest hose reel demand. Therefore, the safety margin inherent in the analysis for the fire event has been preserved.
 
The three echelons of defense-in-depth are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The use of the fire protection water supply system for non-fire protection uses, including the use of hydrants and hoses for purposes other than fire, does not impact fire protection defense-in-depth. Administrative controls consisting of procedural direction or continuously stationed individual ensure that a hose station or hydrant is secured or otherwise made available in the event of a fire. The pumps have the excess capacity to supply the demands of the fire protection system in addition to the greatest hose reel demand and do not affect echelons 1, 2 and 3.
Conclusion
:
NRC approval is requested for the use of the AN O fire protection water supply system for purposes other than fire protection water supply.
ANO-2 determined that the performance based approach satisfies the following criteria:
 
Satisfies the performance, goals performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release  Defense in Depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)  Safety Margin
 
Arkansas Nuclear One - Unit 2 Att. M - License Condition Changes Enclosure 1 to 2CAN121202 Page M-1 M. License Condition Changes The current ANO-2 fire protection license conditions 2.C.(3)(b) and 2.C.(3)(e) are being replaced consistent with the standard license condition in Regulatory Position 3.1 of RG 1.205. Note that the latter license condition was simply a table of historical modifications which have been addressed as discussed in Section 2.1 of the Transition Report (TR). The revised license condition will replace 2.C.(3)(b). License condition 2.C.(3)(e) will be shown as "Deleted per Amendment xxx, date."  The amendment number and date may be inserted by the NRC upon issuance of the revised Operating License (OL), consistent with the format used for previously deleted sections listed on this page of the OL.
 
Information currently on the affected pages is moved to the following pages as required to accommodate the new license condition. Such changes are administrative in nature and have no impact on the technical content of any license condition.
In support of this change, Entergy has developed an ANO-2 specific FPRA as described in Sections 4.5 and 4.7.3, and Attachments U, V, and W of the TR. A mark-up of the proposed changes to the OL is provided in Enclosure 2 of this letter. A clean, revised copy of the pages is provided in Enclosure 3 of this letter.
License condition 2.C.(10), involving mitigation strategies associated with large fires and explosions (security related), will be maintained as is.
 
A review was conducted of the ANO-2 OL NPF-6 by the site Licensing staff and one or more NFPA 805 Transition Team members. The review was performed by performing electronic searches of the docketed correspondence files by using the Entergy Licensing Research System (Autonomy). The system contains site licensing documents, including documents pertaining to the operating license, the Technical Specifications, the fire protection program, the
 
SAR, correspondence sent to the NRC, and correspondence received from the NRC. The
 
correspondence sent to the NRC includes any outstanding license amendment request submittals.
 
No other license conditions were identified as needing to be revised or superseded.
 
Arkansas Nuclear One - Unit 2 Att. N - Technical Specification Changes Enclosure 1 to 2CAN121202 Page N-1 N. Technical Specification Changes Technical Specification (TS) 6.4.1.c will be deleted.
6.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:
: c. Fire Protection Program implementation Deleted  Entergy determined that this change to the TS is adequate for adoption of the new fire protection licensing basis since the requirement for establishing, implementing, and maintaining fire protection procedures is contained in the regulation (10 CFR 50.48(a) and 50.48(c)
NFPA 805 Chapter 3).
 
A mark-up of the proposed change to the TS is provided in Enclosure 2 of this letter. A clean, revised copy of the page is provided in Enclosure 3 of this letter.
 
A review was conducted of the ANO-2 TSs and TS Bases by Entergy Licensing staff and one or
 
more NFPA 805 Transition Team members. The review was performed by reading the TSs and performing electronic searches. Outstanding LARs that have been submitted to the NRC were also reviewed for potential impact on the TSs.
 
No other TSs or TS Bases were identified as needing to be revised or superseded.
 
Arkansas Nuclear One - Unit 2 Att. O - Orders and Exemption Enclosure 1 to 2CAN121202 Page O-1 O. Orders and Exemptions Exemptions
 
Attachment K includes a detailed listing of exemptions granted against 10 CFR 50, Appendix R. Only two exemptions are being retained:
: 1) Appendix R Exemption 17, FA - NN, RCP Oil Collection, Not Meeting III.O Criteria
: 2) Appendix R Exemption 19, FA - NN, RCP Oil Fill Line, Not Meeting III.O Criteria
 
Attachment K provides the basis for rescinding or retaining exemptions, as applicable.
 
Orders No Orders need to be superseded or revised.
 
ANO-2 implemented the following process for making this determination:
A review of the ANO-2 docketed correspondence was performed by the site Licensing staff and the NFPA 805 Transition Team. The review was performed by performing electronic searches of the docketed correspondence files by using the Entergy Licensing Research System (Autonomy). The system contains site licensing documents, including documents pertaining to the operating license, the Technical Specifications, the fire protection program, the SAR, correspondence sent to the NRC, and correspondence received from the NRC. The correspondence sent to the NRC includes any outstanding license amendment request submittals.
A specific review was performed of the license amendment that incorporated the mitigation strategies required by Section B.5.b of Commission Order EA-02-026 (TAC No. MD4495) to ensure that any changes being made to ensure compliance with 10 CFR 50.48(c) do not invalidate existing commitments applicable to the plant. The review of this order demonstrated that changes to the fire protection program will not affect measures required by B.5.b.
 
Arkansas Nuclear One - Unit 2 Att. P - RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4) Enclosure 1 to 2CAN121202 Page P-1 P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)
No risk-informed or performance-based alternatives to compliance with NFPA 805 (per 10 CFR 50.48(c)(4)) were utilized by ANO-2.
 
Arkansas Nuclear One - Unit 2 Att. Q - No Significant Hazards Evaluation Enclosure 1 to 2CAN121202 Page Q-1 Q. No Significant Hazards Evaluations Pursuant to 10 CFR 50.91, Entergy Operations, Inc. (Entergy) has made the determination that this amendment request involves a "No Signific ant Hazards Consideration" by applying the standards established by the NRC regulations in 10 CFR 50.92.
 
To the extent that these conclusions apply to compliance with the requirements in NFPA 805, these conclusions are based on the following NRC statements in the Statements of Consideration accompanying the adoption of alternative fire protection requirements based on NFPA 805.
 
Criterion 1: The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated Operation of Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with the proposed amendment does not result in a significant increase in the probability or consequences of accidents previously evaluated. The proposed amendment does not affect accident initiators or precursors as described in the ANO-2 Safety Analysis Report (SAR), nor does it adversely alter design assumptions, conditions, or configurations of the facility, and it does not adversely impact the ability of structures, systems, or components (SSCs) to perform their intended function to mitigate the consequences of accidents described and evaluated in the SAR. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions as required by the accident analysis. The SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions.
 
The purpose of this amendment is to permit ANO-2 to adopt a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection requirements that are an acceptable alternative to the 10 CFR Part 50, Appendix R, fire protection features (69 FR 33536; June 16, 2004).
 
The purpose of the fire protection program is to provide assurance, through defense-in-depth, that the NRC's fire protection objectives are satisfied. These objectives are: (1) preventing fires from starting; (2) rapidly detecting and controlling fires and promptly extinguishing those fires that do occur, thereby limiting fire damage; (3) providing an adequate level of fire protection for SSCs important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed; and (4) ensuring that fires will not significantly increase the risk of radioactive releases to the environment. In addition, fire protection systems must be designed such that their failure or inadvertent operation does not adversely impact the ability of the SSCs important
 
to safety to perform their safety-related functions.
 
Arkansas Nuclear One - Unit 2 Att. Q - No Significant Hazards Evaluation Enclosure 1 to 2CAN121202 Page Q-2 NFPA 805, taken as a whole, provides an acceptable alternative for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 10 CFR Part 50, meets the underlying intent of the NRC's existing fire protection regulations and guidance, and achieves defense-in-depth along with the goals, performance objectives, and performance criteria specified in NFPA 805, Chapter 1. In addition, if there are any increases in core damage frequency (CDF) or risk as a result of the transition to NFPA 805, the increase will be small, bounded by the delta risk requirements of NFPA 805, and consistent with the intent of the Commission's Safety Goal Policy.
Engineering analyses, which may include engineering evaluations, probabilistic risk assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met. The SAR documents the analyses of design basis accidents (DBAs) at ANO-2. All accident analysis acceptance criteria will continue to be met with the proposed amendment.
The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions for the radiological consequence evaluations in the ANO-2 SAR. In addition, the applicable radiological dose acceptance criteria will continue to be met.
Based on the above, the implementation of this amendment to transition the Fire Protection Plan (FPP) at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a significant increase in the probability of any accident previously evaluated. In addition, all equipment required to mitigate an accident remains capable of performing the assumed function. Therefore, the consequences of any accident previously evaluated are not significantly increased with the implementation of this amendment.
Criterion 2: The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from Any Accident Previously Evaluated Operation of ANO-2 in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Previously analyzed accidents with potential offsite dose consequences were included in the evaluation of the transition to NFPA 805. The proposed amendment does not impact these accident analyses. The proposed change does not alter the requirements or functions for systems required during accident conditions as assumed in the licensing basis analyses and/or DBA radiological consequences evaluations.
Implementation of the new risk-informed, performance-based fire protection licensing basis, which complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in RG 1.205, will not result in new or different kinds of accidents. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). No new modes of operation are introduced by the proposed amendment, nor will it create any failure mode not bounded by previously evaluated accidents. Further, the impacts of the proposed change are not directly assumed in any safety analysis to initiate an accident sequence.
Arkansas Nuclear One - Unit 2 Att. Q - No Significant Hazards Evaluation Enclosure 1 to 2CAN121202 Page Q-3 The requirements in NFPA 805 address only fire protection and the impacts of fire effects on the plant have been evaluated. The proposed fire protection program changes do not involve new failure mechanisms or malfunctions that could initiate a new or different kind of accident beyond those already analyzed in the SAR. Based on this, as well as the discussion above, the implementation of this amendment to transition the FPP at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not create the possibility of a new or different kind of accident from any accident previously evaluated.
 
Criterion 3: The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety  Operation of ANO-2 in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The transition to a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed in the SAR to mitigate accidents. The proposed change does not adv ersely impact systems that respond to safely shut down the plant and maintain the plant in a safe shutdown condition.
In addition, the proposed amendment will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without implementation of appropriate compensatory measures.
The risk evaluations for plant changes, in part as they relate to the potential for reducing a safety margin, were measured quantitatively for acceptability using the delta risk (i.e., CDF and LERF) criteria from Section 5.3.5, "Acceptance Criteria," of NEI 04-02, as well as the guidance contained in RG 1.205. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based methods of NFPA 805 do not result in a significant reduction in the margin of safety. As such, the proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits.
Based on the above, the implementation of this amendment to transition the FPP at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c), will not significantly reduce a margin of safety.
 
NFPA 805 continues to protect public health and safety and the common defense and security because the overall approach of NFPA 805 is consistent with the key principles for evaluating risk-informed licensing basis changes, as described in RG 1.174, is consistent with the defense-in-depth philosophy, and maintains sufficient safety margins. Based on the above discussion, the three standards of 10 CFR 50.92(c) are sa tisfied. Therefore, the amendment request to transition the FPP at ANO-2 to one based on NFPA 805, in accordance with 10 CFR 50.48(c),
involves no significant hazards consideration.
 
Arkansas Nuclear One - Unit 2 Att. R -Environmental Considerations Enclosure 1 to 2CAN121202 Page R-1 R. Environmental Considerations Evaluation Entergy Operations, Inc. (Entergy) has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Entergy has determined that the proposed amendment meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50.
The purpose of the proposed amendment is to permit Arkansas Nuclear One, Unit 2, (ANO-2) to adopt a new fire protection licensing basis that complies with the requirements of 10 CFR 50.48(a) and (c) and the guidance in Regulatory Guide 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and appropriate performance criteria for licensees to identify fire protection requirements that are an acceptable alternative to the 10 CFR 50, Appendix R, fire protection features (69 FR 33536, June 16, 2004).
Accordingly, Entergy evaluated the proposed change against the categorical exclusion requirements of 10 CFR 51.22(c)(9), which state that in order for a license amendment to be excluded from the need for an environmental review, it must meet the following criteria:
(i) The amendment involves no significant hazards consideration; (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and (iii) There is no significant increase in individual or cumulative occupational radiation exposure.
As stated in Attachment Q, the proposed amendment does not involve a significant hazards consideration.
Compliance with NFPA 805 explicitly requires the attainment of performance criteria, objectives, and goals for radioactive releases to the environment. The radioactive release goals provide reasonable assurance that a fire will not result in a radiological release that affects the public, plant personnel, or the environment. The NFPA 805 transition has been evaluated based on fire suppression activities, but not involving fuel damage, and does not create any new source
 
terms. Therefore, the proposed amendment will not change the types or amounts of any effluents that may be released offsite.
Furthermore, the proposed change will not significantly alter the types or increase the amount of individual or cumulative occupational radiation exposures based on the results of the evaluation performed regarding fire fighting activities. In addition, the modifications being implemented as a part of the transition to NFPA 805 at ANO-2 will reduce the need for recovery actions within the plant, which may function to lower overall operator occupational exposures in many scenarios.
Therefore, Entergy has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. Entergy has also determined that the amendment involves no significant hazards consideration. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-1 S. Plant Modifications and Items to be Completed During Implementation Table S-1, Plant Modifications, provided below includes a description of the modifications along with the following information
:  A problem statement,  Risk ranking of the modification,  An indication if the modification is currently included in the FPRA,  Compensatory measure in place, and  A risk-informed characterization of the modification and compensatory measure.
The following ranking legend should be used when reviewing the table:  High = Modification which would have an impact on FPRA and affect multiple Fire Areas. Med = Modification which would have an impact on FPRA and affect individual Fire Areas, or include IN 92-18 modifications. Low = Modification which would have no or insignificant impact on risk.
Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-1 Med (PRA) 2 In Fire Area HH, a separation issue was identified on the EFW valves 2CV-1026-2 and 2CV-1076-2. During a fire induced circuit failure the feedwater valves may be impacted by a fire in Fire Zone 2096-M. LAR Source:
Attachment C ANO plans to relocate interposing relays and affected cables associated with 2CV-1026-2 and 2CV-1076-2 from Fire Area HH, Fire Zone 2096-M, to the adjacent room in Fire Area G, Fire Zone 2098-C. Circuits for 2CV-1026-2 and 2CV-1076-2 are currently routed through Fire Area G and no new impacts will be generated by this modification. Yes Yes This modification is specifically credited from a PRA perspective. Modification reduces the risk in Fire Area HH of a fire induced circuit failure for EFW valves 2CV-1026-2 and 2CV-1076-2 in Fire Zone 2096-M. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-2 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-2 High (PRA) 2 In Fire Area JJ, a separation issue was identified that impacts the DC power cables control wiring on both trains. If a fire event occurred, this could result in the loss of equipment that would otherwise be available. Additional considerations are potential spurious operations at switchgear 2A-3 that may result in a loss of power to the safety bus. LAR Source:
Attachment C ANO plans to modify the circuits as described to eliminate impacts in Fire Area JJ associated with these components. 2A-3, 2A-308, 2A-309, and 2A-310 - The red train 125V DC panel 2D-23 that supplies control power for 2A-3 and 2B-5 is planned for relocation to Fire Area MM from Fire Area JJ. Control power cables are planned to be rerouted using embedded conduits from Fire Area MM to Fire Area II to avoid Fire Areas JJ and SS. This allows post-fire control of 2A-3 bus from the control room. 2CV-1036-2
- Auxiliary relays 2CR1036A, B, C, and D are currently installed in MCC 2B-61 and are planned to be relocated to MCC 2B-63. This would eliminate cables that are routed through Fire Area JJ associated with this valve. This eliminates a loss of 2CV-1036-2 due to a fire in Fire Area JJ. (continued) Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas. The modification limits the risk of a potential spurious operation and a loss of DC power to safety bus for switchgear 2A-3 due to a fire induced circuit failure. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-3 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-2    2CV-1075 The reroute of DC control power to bus 2A-3 and load-center 2B-5 listed above assures MCC 2B-53 remains available to power this valve. Control cables from 2C-17 to MCC 2B-53 are planned to be rerouted using an embedded conduit between Fire Area G and II to avoid Fire Areas JJ, SS, and TT. New dedicated fuses are planned to be installed in 2C-17 for 2CV-1075-1 control relays so that failure of cables in scheme 2S113 will not impact 2CV-1075-1.
2B Cables are planned to be rerouted to control room panel 2C33-2 from 2B-6 using an embedded conduit between Fire Zone 2100-Z to the cable
 
spreading room Fire Area G. This eliminates an impact in Fire Area JJ. 2CV-4816 & 2CV-4817
- A reroute of cable 2I016N is planned by using embedded conduit C4080 that is located between Fire Area G (cable spreading room) to Fire Area EE-L. Cable 2I016N is also planned to be separately fused in panel C-09 to prevent failure due to a loss of cable 2I016P. This eliminates circuit impacts in Fire Areas TT, JJ, and EE-U.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-4 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-3 High (PRA) 2 In Fire Area MM, fire induced circuit failure could impact DC power cables feeding circuit breakers at switchgear 2A-1, 2A-2, 2H-1, and 2H-2. The failure of 2A-1 and 2A-2 could prevent alignment to an offsite power source. The failure of 2H-1 and 2H-2 could prevent tripping the reactor coolant pumps (RCPs) from the control room. LAR Source:
Attachment C ANO plans to install backup DC control power to switchgear 2A-1, 2A-2, 2H-1 and 2H-2 with automatic transfer capability in the event the normal DC control power source is lost. The new backup DC power source will be located completely within Fire Area B-2 in proximity to the switchgear either on elevation 372' or below at elevation 354'. This eliminates impacts to switchgear DC control power due to a fire in any other ANO-2 fire area and allows tripping of the RCPs in those areas.
Inclusive in this modification will be changes to the control power circuits for switchgear 2H-1 and 2H-2 to allow tripping the RCPs in a scenario where a fire originates internally to a switchgear cubicle. This design will prevent fire damage to a load cubicle from disabling the ability to trip the line breakers and remove power to the RCPs. The opposite scenario where fire damages the line breakers would not prevent the RCP load breakers from being tripped. This modification will require the line and load breakers to be separately fused and fed as described: (continued) Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas. Modification to install an alternate DC power source reduces the risk of a fire induced circuit failure to the DC power cables feeding RCP circuit breakers 2H-1 and 2H-2 which could prevent tripping the RCPs from the control room. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-5 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-3    2H Internal DC control wiring jumpers will be removed to isolate the line and load cubicles. The DC control power for line breakers 2H-13, 2H-14, and 2H-15 will be isolated from the DC control power for the load breakers 2H-10, 2H-11, and 2H-12.
2H Internal DC control wiring jumpers will be removed to isolate the line and load cubicles. The DC control power for line breakers 2H-23, 2H-24, and 2H-25 will be isolated from the DC control power for the load breakers 2H-20, 2H-21, and 2H-22.
 
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-6 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-4 High (PRA) 2 In Fire Area TT, a separation issue was identified that impacts the power cables for EFW, chemical and volume control system (CVCS), and service water (SW) components. LAR Source:
Attachment C ANO plans to modify the circuits as described to eliminate impacts in Fire Area TT associated  with these components. 2CV-1036 Auxiliary relays 2CR1036A, B, C, and D are currently installed in MCC 2B-61 and are planned to be relocated to MCC 2B-63. This would also eliminate cables that are routed through Fire Area TT associated with this valve. This eliminates a loss of 2CV-1036-2 due to a fire in Fire Area TT. 2CV-1075 Cables for this valve between panels 2C-39 to 2C-17 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable
 
spreading room. Control cables from 2C-17 to MCC 2B-53 are planned to be rerouted using an embedded conduit between Fire Area G and II to avoid Fire Areas JJ, SS, and TT. New dedicated fuses are planned for installation in 2C-17 for 2CV-1075-1 control relays so that failure of cables in scheme 2S113 will not impact 2CV-1075-1. (continued) Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas. The modification reduces the risk of a fire induced circuit failure for EFW/CVCS/SW components and power cables (2B-5, 2CV-0789-1, 2CV-1036-2, 2CV-1075-1, 2CV-4816, 2CV-4817, and 2P-7B) in Fire Area TT. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-7 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-4    2P-7B - Cables for this pump between panels 2C-39 to 2C-17 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room. New conduits are also planned to be installed. 2CV-0789 Cables for this valve between panels 2C-39 to 2C-17 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room. Control cables from 2C-17 to MCC 2B-53 are planned to be rerouted using an embedded conduit between Fire Area G and II to avoid Fire Areas JJ and SS. 2CV-4816 & 2CV-4817
- A reroute of cable 2I016N is planned by using embedded conduit C4080 that goes between Fire Area G (cable spreading room) to Fire Area EE-L. Cable 2I016N is also planned to be separately fused in panel C-09 to prevent failure due to a loss of cable 2I016P of cable. This eliminates circuit impacts in Fire Areas TT, JJ, and EE-U. (continued) 2B Cables for this load center between panels 2C-39 to 2C-33-1 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-8 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-4    2B Cables for this load center between panels 2C-39 to 2C-33-1 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room.
S1-5 High (PRA) 2 In Fire Area SS, a fire induced circuit failure could impact the DC power on both trains resulting in the loss of the various components. LAR Source:
Attachment C ANO plans to modify the circuits as described to eliminate impacts in Fire Area SS associated with these components. 2A-3 and 2A-310
- The red train 125V DC panel 2D-23 that supplies control power for 2A-3 and 2B-5 is planned to be relocated from Fire Area JJ to Fire Area MM. Control power cables are planned to be rerouted using embedded conduits from Fire Area MM to Fire Area II to avoid Fire Areas JJ and SS. This allows post-fire control of 2A-3 bus from the control room. 2A-4, 2A-409, & 2B-6
- Cables are planned to be rerouted to control room panel 2C33-2 from 2A-4 and 2B-6 using an embedded conduit between Fire Zone 2100-Z to the cable spreading room Fire Area G. This eliminates an impact in Fire Zone 2097-X and Fire Area JJ. (continued) Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas. The modification reduces the risk of a fire induced circuit failure that could result in the loss of DC power for both trains. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-9 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-5    The 125V DC control power from 2D-24 to 2A-4 is planned to be rerouted using a new conduit to avoid an impact against cables G2D2404A and B in Fire Zone 2097-X. 2CV-0789-1 & 2PIS-0789-1
- The power cable for 2PIS-0789-1 (for 2CV-0789-1) is planned to be re-routed using an embedded conduit from Fire Area G to Fire Area II to avoid Fire Area SS. 2CV-1040 This valve is not directly impacted but is failed due to a loss of AC. The red train 125V DC panel 2D-23 that supplies control power for 2A-3 and 2B-5 is planned to be relocated from Fire Area JJ to Fire Area MM. Control power cables are planned to be routed using embedded conduits from Fire Area MM to Fire Area II to avoid Fire Areas JJ and SS. This assures 2CV-1040-1 will have a source of power and eliminates an impact in Fire Area SS. (continued)
 
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-10 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-5    2A-308, 2A-309, 2D-27, 2K-4A, 2P-16A, 2P-36A, 2SV-0724-1, 2SV-2809-1, 2SV-2810-1, and 2SV-2811 - The cables associated with these components are planned to be re-routed to avoid Fire Area SS by using embedded conduits and as required the installation of a new raceway in Fire Area B-2 directly under Fire Area SS on elevation 372'. The new raceway in Fire Area B-2 is planned to be installed above the vertical zone of influence for any postulated fire source. This eliminates impacts for 2A-308, 2A-309, 2D-27, 2K-4A, 2P-16A, 2P-36A, 2SV-0724-1, 2SV-2809-1, 2SV-2810-1, and 2SV-2811 in Fire Area SS. S1-6 Med (92-18) 2 Motor Operated Valves (MOVs) will be modified to meet requirements per IN 92-18. The NPO assessment determined that any one of the RCS drop line valves can fail in a closed and unrecoverable position resulting in a loss of SDC. LAR Source: Attachment D NPO-RCS-SDC ANO plans to modify the control circuit for 2CV-5038-1 to prevent spurious closure. This is planned to be similar to the inhibit circuit modification on CV-1275 for ANO-1. Procedural controls to secure power by opening breakers are planned to be implemented for 2CV-5084-1 and 2CV-5086-2. No Yes The NPO modification reduces the risk of fire induced MOV circuit failures (hot shorts, open circuits and short to ground). This MOV modification can prevent a non-recoverable position failure resulting in the loss of shutdown cooling. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-11 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-7 Med (PRA) 2 In Fire Area G, MOVs will be modified to meet requirements per IN 92-18. Four EFW discharge valves can fail in an unrecoverable position. LAR Source:
Attachment C ANO plans to modify the control circuit for MOVs 2CV-1075-1, 2CV-1076-2, 2CV-1036-2, and 2CV-1039-1 to prevent fire induced spurious operation from the main control room, Fire Area G. This will be accomplished by separating the cable conductors, inclusive of internal panel wiring, that can cause spurious valve operation and protecting them with grounded metallic raceway and the use of grounded metallic barriers. This will prevent contact with potentially energized conductors from both intracable and intercable hot shorts. MOV 2CV-1075-1 control cables R2B53J2C and R2B53J2N that enter panel 2C-17 or 2C-39 from floor penetrations have been identified as the cables of concern applicable to this modification. MOV 2CV-1076-2 control cables G2D26C1D, G2D26C1E, and G2D26C1L that enter panel 2C-16 or 2C-40 from floor penetrations have been identified as the cables of concern applicable to this modification. (continued) Yes Yes This modification is specifically credited from a PRA perspective. The modification reduces the risk of fire induced MOV circuit failures (hot shorts, open circuits and short to ground). This MOV modification can prevent a non-recoverable position failure. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-12 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-7    MOV 2CV-1036-2 control cable G2B63H1E that enters panel 2C-40 from a floor penetration has been identified as the cable of concern applicable to this modification. MOV 2CV-1039-1 control cable R2D27B2E that enters panel 2C-39 from a floor penetration has been identified as the cable of concern applicable to this modification. S1-8 High (PRA) 2 In Fire Areas B-3 and G, spurious opening of MOV 2CV-4698-1 pressurizer low temperature - overpressure (LTOP) relief can result from a fire in motor control center (MCC) 2D-27. LAR Source:
Attachment C ANO plans to modify the control circuit for 2CV-4698-1 to prevent fire induced spurious opening in Fire Areas B-3 and G. This will be accomplished by separating the cable conductors, inclusive of internal panel wiring, that can cause spurious opening and protecting the conductors with a grounded metallic raceway and the use of grounded metallic barriers. This will prevent contact with potentially energized conductors from both intracable and intercable hot shorts. Control cable R2D27A3J that enters MCC 2D-27 in Fire Area B-3 and the other end of cable that enters cabinet 2C-09 in Fire Area G from the floor penetrations have been identified as the cable of concern applicable to this modification. Yes Yes This modification is specifically credited from a PRA perspective. The modification in Fire Areas B-3 and G to install flexible metallic conduit protects the valve control cable in MCC 2D-27 and cabinet 2C-09 which reduces the risk of fire induced circuit failures (such as spurious opening). This modification can prevent a non-recoverable position failure. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-13 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-9 High (PRA) 2 RCS High Point Vent Valves 2SV-4670-2 and 2SV-4669-1: Solenoid valves are provided in the RCS system for a means of venting. These "energize to open" solenoids provide a vent path fr om the top of the pressurizer. The final two, in parallel, solenoids determine the vent path, either to the containment atmosphere (2SV-4670-2) or the quench tank (2SV-4669-1). In Fire Area G, PRA determined that solenoid valves 2SV-4670-2 and 2SV-4669-1 control circuits shall be protected with metallic sleeves and/or barriers to prevent a spuriously opening of 2SV-4670-2 or 2SV-4669-1 due to a fire induced hot short circuit failure resulting in an uncontrolled RCS vent path release. LAR Source:
Attachment C ANO plans to modify the RCS vent solenoid valve control circuits with the installation of metallic sleeves and/or barriers as described to eliminate impacts in Fire Area G associated with the following components: 2SV-4670 Control circuit in cabinet 2C-336-2 modification is planned to prevent conductor (wire P3) of cable G2SI122E from contacting energized conductors by installing a grounded metallic sleeve and/or barriers up to load side of hand-switch 2HS-4670-2. 2SV-4669 Control circuit in cabinet 2C-336-1 modification is planned to prevent conductor (wire P3) of cable R2SI121E from contacting energized conductors by installing a grounded metallic sleeve and/or barriers up to load side of hand-switch 2HS-4669-1.
Yes Yes This modification is specifically credited from a PRA perspective. The modification in Fire Area G to install grounded metallic sleeves and/or barriers protects control circuit cable G2SI122E in cabinet 2C-336-2 and control circuit cable R2SI121E in cabinet 2C-336-1 to reduce the risk of fire induced circuit failures (such as spurious opening). In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-14 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-10 Med (PRA) 2 In Fire Area B-4 an incipient fire detection system is not installed in control element drive mechanism (CEDM) room panels 2C-70, 2C-71, 2C-72, 2C-73, 2C-75, 2C-80, and 2C-409. However an early warning fire detection system in accordance with NFPA 72, Fire Alarm Detection Code, is required by the PRA in accordance with FRE CALC-09-E-0008-05. LAR Source:
Attachment C ANO plans to provide a modification in the CEDM room in Fire Area B-4 to install incipient detection in cabinets 2C-70, 2C-71, 2C-72, 2C-73, 2C-75, 2C-80, and 2C-409. Fire detection signal cable is planned to be routed from each air sampling detector to the control room fire panel 2C-343-3. Yes Yes This modification is specifically credited from a PRA perspective. The early warning fire detection system modification in Fire Area B-4 reduces the risk of a fire induced circuit and equipment failures that could result in the loss of CEDM room panels 2C-70, 2C-71, 2C-72, 2C-73, 2C-75, 2C-80, and 2C-409. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
S1-11 High (PRA) 2 At ANO the availability of feedwater to ANO-2 SGs was identified as an issue by PRA. Also identified by PRA was ANO's inability to perform high risk and time sensitive actions, such as control of auxiliary feedwater (AFW), outside of the ANO-2 Control Room. LAR Source: Attachment C (listed globally in multiple Fire Areas for new AFW pump in Risk Summaries) ANO plans to install a new AFW pump capable of feeding either of the ANO-2 SGs. The AFW would be designed to meet or exceed the flow requirements of ANO-2 Emergency Feedwater (EFW) Pump 2P-7B (380 gpm @
1100 psig). The new pump, controls and motor operated valves would be designed to be installed in a manner that protects the assumptions in the PRA. The preferred source of suction for the new pump is planned to be from an available source (i.e. Condensate Storage Tank). (continued)
Yes Yes The AFW modification is specifically credited from a PRA perspective to provide a reliable additional source of feedwater. The local control panel modification is specifically credited from a PRA perspective to provide an alternate means to perform required actions outside the ANO-2 Control Room. This modification reduces the risk of not being able to perform necessary operator actions to shutdown the plant, if either Control Room can't be manned. (continued)
 
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-15 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-11    The discharge piping is planned to be routed through the  Turbine Building to ANO-2 Auxiliary Building Rooms 2081 and 2084 for the tie-ins to the EFW System piping. The AFW tie-ins are planned to discharge into the ANO-2 EFW downstream of all EFW injection valves to ensure a single area fire does not disable AFW. The AFW pump would be designed to have the capability to be operated from the ANO-2 Control Room and also locally. The design will ensure electrical isolation from Control Room functions to prevent a fire in the ANO-2 Control Room from affecting local control of AFW components. The AFW pump and associated motor operated valves would be designed to be powered by diverse non-safety related power sources to prevent a single failure from disabling equipment operation. The AFW pump would be designed to include controls and monitoring instrumentation to ensure proper water flow to the SGs. The local controls and monitoring instrumentation are planned to be located and powered with redundant power supplies in a manner that protects the assumptions in the PRA. Also, the local control panel modification reduces the risk of availability issue with of feedwater supply to the ANO-2 SGs. Manual actions are credited in fire areas that contain redundant safe shutdown equipment. These actions have been demonstrated feasible and are therefore considered adequate compensatory measures until compliance can be achieved by transitioning to a 10CFR50.48(c) licensing basis.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-16 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-12 Med (PRA) 2 In Fire Area B-3, excessive temperatures have been identified in Fire Zone 2091-BB following a loss of ventilation. LAR Source:
Attachment C ANO plans to modify the control wiring for fans 2VEF-63 and 2VEF-64 to isolate the control room and allow the local controls to override a "stop" signal generated from within Fire Area G, either from handswitch positioning or fire-induced circuit damage. This eliminates fire impacts in Fire Area G and assures either 2VEF-63 or 2VEF-64 will remain available except for a fire in Fire Area B-3, Fire Zone 2091-BB. Yes No This modification supports a basic assumption from a PRA perspective.
S1-13 Med (PRA) 2 In Fire Area MM, excessive temperatures have been identified in Fire Zone 2099-W following a loss of ventilation. LAR Source:
Attachment C ANO plans to provide a modification to fire door DR 265 to allow normally open positioning with automatic closure features in the event of a fire. This allows natural circulation to prevent long term room overheating impact on equipment located in Fire Zone 2099-W, West DC Equipment Room, by allowing an opening to Fire Zone 2109-U, Corridor, in Fire Area JJ. Yes No This modification supports a basic assumption from a PRA perspective.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-17 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-14 Low (Code) C With regard to NFPA 50A, Gaseous Hydrogen Systems, code non-compliance issues were identified in the Hydrogen Gas Bottle Storage Room related to inadequate vent piping and room ventilation. The hydrogen storage room light switch was identified as not meeting Article 501 for Class I, Division II locations of the National Electric Code (NEC). LAR Source: Attachment A, Section 3.3.7.1 ANO plans to provide a modification to move the hydrogen bottles and manifold from the Hydrogen Gas Bottle Storage Room to a concrete slab located outside this room and open to atmosphere. This addresses hydrogen ventilation concerns and eliminates the need for electrical upgrades. No No The subject hydrogen gas system bottle storage area is not credited by the PRA. This modification will be completed to meet NFPA 805 code requirements.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-18 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-15 Med (PRA) C NFPA 805 non-compliance issues were encountered when smaller fire areas were defined such that multiple walls, dampers, penetration seals, and doors were credited and used in the PRA model as rated fire barriers in the NRC regulatory basis for NFPA 805. Multiple walls and doors barriers will require upgrading to comply with NFPA 805. LAR Source: Attachment A, Section 3.11.2 ANO plans to provide an adequate-for-the-hazard evaluation and if necessary a modification to upgrade fire barrier walls, dampers, penetration seals, and doors to rated barriers for those barriers credited for deterministic compliance and subsequently credited in the Fire PRA analysis. These barriers have been previously identified as NRC regulatory basis to ensure compliance with NFPA 805 and have compensatory measures established. The barriers to be addressed as identified by EC-1956 are 2005-2, 2005-3, 2067-4, 2082-3, 2091-1, 2091-2, 2091-3, 2091-4, 2107-4, 2110-2, 2110-4, 2110-7, 2112-2, 2112-8, 2112-10, 2133-5, 2133-6, 2147-8, 2148-4, 2148-5, 2149-5, 2152-2, 2154-2, 2154-3, 2154-5, 2158-10, 2224-2, 2224-3, 2228-10, 2239-4, 2239-5, 2256-4, 2256-5, 2256-6, 2256-8, 2134-1, and 2155-1. Yes Yes This modification will be completed to meet NFPA 805 code requirements. In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-19 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-16 Low (Code) C NFPA 10 non-compliance issues (such as incorrect number of fire extinguishers for travel distance, incorrect type and size for the hazard area) were identified with ANO portable fire extinguishers. LAR Source:
Attachment A, Section 3.7 ANO plans to provide a modification to resolve the NFPA 10 code deficiencies identified in CALC-ANOC-FP-09-00009. In general, this modification would involve portable fire extinguisher physical relocation, substitution of existing extinguishers, and documentation updates to reflect these plant changes. The results will ensure the proper number of fire extinguishers to meet travel distance requirements in coverage areas, adequately sized fire extinguishers, and the correct type of extinguisher that is rated for the fire hazard in each area. No No The subject fire extinguishers are not credited in the FPRA. This modification will be completed to meet NFPA 805 code requirements.
 
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-20 Table S-2 items provided below are those items (procedure changes, process updates, and training to affected plant personnel) t hat will be completed prior to the implementation of new NFPA 805 fire protection program.
Table S-2 Implementation Items Item Unit Description LAR Section / Source S2-1 C Develop a monitoring program required by NFPA 805 that will include a process to monitor and trend the fire protection program based on specific goals established to measure effectiveness. LAR Section 4.6 and Attachment A (NEI-04-02 B-1 Table) Section 3.2.3 (3) S2-2 2 Revise or develop fire protection flushing activity to perform fixed water spray system flushing and drainage of underground lead-in connections in accordance with NFPA 15, 1977 Edition Code. Attachment A (NEI-04-02 B-1 Table) Section 3.9.1 (2) S2-3 2 Perform an evaluation for NFPA 14, 1983 Edition Code non-compliance for standpipe manual hose station 2HR-75 horizontal water header to determine if additional hangers are required, since hose station water header wall hangers were found not secured. Attachment A (NEI-04-02 B-1 Table) Section 3.6.1 S2-4 C Revise fire protection administrative procedure EN-DC-161, Control of Combustible, to include the following:  In accordance with NFPA 30, applicable NFPA Standards are considered to be equivalent to those NFPA Standards identified in the current license basis (CLB) for procedures and systems in the fire protection program that are transitioning to NFPA 805. Terminology for zero transient combustibles and changes needed to support FPRA assumptions. Attachment A (NEI-04-02 B-1 Table) Sections 3.3.1.2 (5) S2-5 2 Revise existing procedure(s) or develop a new procedure(s) for NPO required to transition to NFPA 805 based upon insights gained from ANO-2 NPO calculation. Attachment D (NEI-04-02 Non-Power Operational Modes) VFDR NPO-ProcedureS2-6 2 Revise OMA procedures/documents to include feasibility criteria in FAQ 07-0030 for the recovery actions listed in Table G-1 of Attachment G, Recovery Action Transition. Attachment G (NEI-04-02 OMA) Step 4 S2-7 C Develop or revise technical documents and procedures that relate to new FP design and licensing basis (e.g., ANO Fire Protection Program, OP-1003.014, Technical Requirements Manual, Design Basis Document, Pre-Fire Plans, Maintenance and Surveillance Procedures, Configuration Control Program, Training and Qualification Guidelines, etc.) as required for implementation of NFPA 805. LAR Sections 4.7.1, 4.7.2, and 4.7.3 S2-8 2 Revise technical documents for NFPA 13 for acceptance of the partial area sprinkler system in Fire Area G for NFPA 805. The existing partial area sprinkler system for Fire Area G has been previously approved by the NRC in an exemption under Appendix R. As transitioned to NFPA 805, the previously approved exemption is being withdrawn, so documentation updates will be needed to remove the reference to an exemption and provide an independent basis. Attachment A (NEI-04-02 B-1 Table) Section 3.9.1 (1)
Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed Enclosure 1 to 2CAN121202 Page S-21 Table S-2 Implementation Items Item Unit Description LAR Section / Source S2-9 C Develop or create a PRA review plan of action to revise the PRA model for each modification or implementation item completed that is credited either directly or indirectly by PRA. The PRA review plan will ensure the as-built change-in-risk from each modification or implementation item does not exceed the PRA model change-in-risk estimates reported in the LAR. LAR Section 4.8.2
 
Arkansas Nuclear One - Unit 2 Att. T - Clarification of Prior NRC Approvals Enclosure 1 to 2CAN121202 Page T-1 T. Clarification of Prior NRC Approvals Introduction For ANO-2, there are no elements of the Fire Protection Current Licensing Basis (CLB) for
 
which specific NRC clarification is required.
 
Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-1 U. Internal Events PRA Quality In accordance with RG 1.205 position 4.3:
"The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard "supporting requirements" important to the NFPA 805 risk assessments, the
 
NRC position is that Capability Category II is generally acceptable. Licensees should justify use of Capability Category I for specific supporting requirements in their NFPA 805 risk assessments, if they contend that it is adequate for the application. Licensees should also evaluate whether portions of the PRA need to meet Capability Category III, as described in the PRA Standard."
The ANO-2 PRA analysis as it relates to Table U-1 is discussed in Section 4.5.1.1 of this enclosure. The following table provi des specific information related to each Category II requirement.
 
Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IE-C3 CALCULATE initiating event frequencies on a reactor year basis
[Note (3)]. INCLUDE in the initiating event analysis plant
 
availability, such that the frequencies are weighted by the fraction of time the plant is at power. Issue:  ANO-2 explicitly calculated the total reactor critical years as total reactor critical hours divided by 8766 hours per year and used this to calculate the initiating event frequencies (IEFs). However, there is no evidence that ANO-2 adjusted these IEFs to reflect average plant availability. This is, in essence, equivalent to assuming that the plant operates at full power all year. ANO-2 needs to adjust their initiating event frequencies to account for average plant availability. Partial Adjust the affected Initiating events to accurately reflect the annual frequency based upon plant availability.
ANO-2 Initiating Events Analysis  Work Package, PRA-A2-01-003S06, Rev 1, May
 
2008 Technical The initiating event frequencies are set to True or False in the quantification
 
of the FPRA.
This issue does not affect
 
the FPRA results Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-2 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IE-C10 COMPARE results and EXPLAIN differences in the initiating event analysis with generic data sources to provide a reasonableness check of the results. There is no evidence in the Initiating Event Notebook that this has been done. Table 7 could be compared to similar plants and also NUREG/
CR-5750 and NUREG/CR-6928 to assess the final values. The ISLOCA IE frequency needs to be reviewed, compared and understood. It is very small. Partial Enhance the documentation to include a table of initiating event comparisons to generic data to meet the expectations required by this
 
SR. ANO-2 Initiating
 
Events Analysis  Work Package, PRA-A2-01-003S06, Rev 1, May 2008 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-3 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IE-C12 In the ISLOCA frequency analysis, INCLUDE the following features of plant and procedures that influence the ISLOCA frequency: (a) configuration of potential pathways including numbers and types of valves and their relevant failure modes and the existence, size, and positioning of relief valves (b) provision of protective interlocks (c) relevant surveillance test procedures (d) the capability of secondary system piping (e) isolation capabilities given high flow/ differential pressure conditions that might exist following breach of the secondary system. Some of the components in the ISLOCA fault tree model appear to have incorrect mission times.
The LPSI MOV, e.g., 2CV5017 rupture, has a mission time of 36 hours. However, the mission time should probably be 8760 hours because the MOV rupture is not likely to be annunciated in the control room as assumed. Therefore, it could potentially be in an undetected failed state for an extended period. The same comment may apply to the second check valve. It could be potentially in an undetected failed state for an extended period. Met Based on the input from this assessment, the database and logic for
 
the ISLOCA events were reviewed and revised to reflect the 18-month interval as recommended. 
 
This issue has been addressed and
 
resolved ISLOCA  Initiating
 
Event Frequency, PRA-A2-01-003S08, Rev. 2, Dec 2008 Technical No impact. Issue has been resolved.
Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-4 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IE-D1 DOCUMENT the initiating event analysis in a manner that facilitates PRA applications, upgrades, and peer review. Some of the documentation is not adequate to meet this requirement. The following items should be addressed: (1) What is the basis for the 80/20 split between reactor trip and turbine trip events?  (Assumption 8, Section 2.2 and Section 5.1 page 19). (2) Appendix C contains calculations for loss of feedwater/condensate (T2). Page 48 contains the Bayesian update for this event. Recommend explaining how these are used in the PRA model and which is used in the base model. The value in Table 7 appears to be from the Bayesian update which is inconsistent with the discussion in Section 5.3.1. (3) Loss of Lake Dardanelle IE - need to document the change from 2E-04/yr to 1E-05/yr - it should be explained how the IE frequency for this event was reduced to 1E-05/yr. (4) As stated in Section 4.7, the list of initiating events was reviewed to ensure it was complete and accurate. Need some documentation of the extent of the review, feedback, and how comments were addressed. Partial Enhance the documentation to address the issues identified during the review of this SR. ANO-2 Initiating
 
Events Analysis  Work Package, PRA-A2-01-003S06, Rev 1, May 2008 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-5 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-A4 For each modeled initiating event, using the success criteria defined for each key safety function (in accordance with SR SC-A4), IDENTIFY the necessary operator actions to achieve the defined success criteria.  [See Notes (1) and (2).] The operator actions are not specifically identified even though some are mentioned. Partial Enhance documentation to specifically discuss the results of procedural review that did not identify any new HRAs for Accident Sequences that are needed to protect the safety functions ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation No impact is expected for documentation issues AS-A5 DEFINE the accident sequence model in a manner that is consistent with the plant-specific: system design, EOPs, abnormal procedures, and plant transient response. There is no reference to the System design, EOPs, or abnormal procedures in the accident sequence notebook. It would be helpful if the EOP or abnormal procedure used for each accident sequence was noted. Partial The event trees and top logic are consistent with Emergency Operating Procedures (EOP) / Abnormal Operating Procedures (AOPs). Documentation
 
will be enhanced for clearer links to EOP/AOPs.
ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan
 
2006 Documentation No impact is expected for documentation
 
issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-6 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-A10 In constructing the accident sequence models, INCLUDE, for each modeled initiating event, sufficient detail that differences in requirements on systems and operator responses are captured. Where diverse systems and/or operator actions provide a similar function, if choosing one over another changes the requirements for operator intervention or the need for other systems, MODEL each separately.
The operation actions are not specified in either the accident sequence detailed description or the event tree. An example would be a detailed discussion of the once through cooling and the operator actions required.
Partial Operator actions are included in the accident sequence logic.
Documentation of operator actions needs to be enhanced.
ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-7 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-B1 For each modeled initiating event, IDENTIFY mitigating systems impacted by the occurrence of the initiator and the extent of the impact. INCLUDE the impact of initiating events on mitigating systems in the accident progression either in the accident sequence models or in the system models. The special initiators do not address the impact of these initiators on the mitigating systems. Partial ANO-2 uses a linked fault tree approach where the initiating events are placed in the tree with the appropriate system model.
A table will be added to the accident sequence report to show how the IE fails both the front
 
line and support systems. ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-8 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-B2 IDENTIFY the dependence of modeled mitigating systems on the success or failure of preceding systems, functions, and human actions. INCLUDE the impact on accident progression, either in the accident sequence models or in the system models. For
 
example, (a) turbine-driven system dependency
 
on SORV, depressurization, and containment heat removal (suppression pool cooling) (b) low-pressure system injection success dependent on need for RPV depressurization The dependencies are not addressed in the Accident Sequence notebook. This is especially true of operator actions and how the failure of an operator action would affect subsequent operator actions. Partial ANO-2 uses a linked fault tree approach where the dependencies of the operator actions are placed in the tree with the appropriate system model. Dependencies between operator actions are addressed
 
within a rule file that is applied post-quantification.
ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-9 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-B3 For each accident sequence, IDENTIFY the phenomenological conditions created by the accident progression. Phenomenological impacts include generation of harsh environments affecting temperature, pressure, debris, water levels, humidity, etc. that could impact the success of the system or function under consideration [e.g., loss of pump net positive suction head (NPSH), clogging of flow paths]. INCLUDE the impact of the accident progression phenomena, either in the accident sequence models or in the system models. No assumption or statement is made that plant equipment will perform in the environment for which it was designed. There also was not evidence that equipment not specified in the SAR for accident mitigation but still credited in the PRA was reviewed for environmental affects. This could be an assumption that the equipment meets the environmental qualification.
Equipment that is not environmentally qualified need to be analyzed on the impact they have on the applicable accident sequence. Partial Phenomeno-logical effects are already considered in the model.
Accident sequence notebooks and system model notebooks should be revised to identify those environmental effects of the initiating event and the impact on mitigation systems. ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation No impact is expected for documentation issues.
Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-10 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-B6 MODEL time-phased dependencies (i.e.,
those that change as the accident progresses, due to such factors as depletion of resources, recovery of resources, and changes in loads) in the accident sequences. Examples are: (a) For SBO/LOOP sequences, key time- phased events, such as  (1) AC power recovery (2) DC battery adequacy (time-dependent discharge) (3) Environmental conditions (e.g., room cooling) for
 
operating equipment and the control room (continued) This SR was not met due to any discussion of the following:
changes in environmental conditions, shifting of the CSTs, and operator actions. Questions were raised in that ANO-2 uses 40 minutes as the time that the RCP can run without CCW cooling. The industry practice (WCAP 16175) uses 20 minutes.
This difference should be analyzed and resolved. Discuss the following: changes in environmental conditions, shifting of the CSTs, and operator actions.
Partial Accident sequence notebooks and system model notebooks should discuss changes in environmental conditions due to changing factors associated with support systems. The PRA model will be revised to use 20 minutes as available time to trip the RCPs. ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation/Technical No impact is expected for documentation issues. Additionally, the FPRA uses 20 minutes for available time in tripping the RCPs vs. the 40 minutes discussed in
 
the internal events peer review finding.
Therefore, this technical issue has been resolved for the NFPA-805 submittal.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-B6 (AS-B6 continued) (b) For ATWS/failure to scram events (for BWRs), key time-dependent actions such as (1) SLCS initiation (2) RPV level control (3) ADS inhibit (c) Other events that may be subject to explicit time-dependent characterization include (1) CRD as an adequate RPV injection source (2) Long-term make-up to RWST      AS-C2 DOCUMENT the processes used to develop accident sequences and treat dependencies in accident sequences, including the inputs, methods, and results. For example, this documentation typically includes: (continued) This SR was not met. The documentation does not show that the items in this SR have been address. The operator actions have not been addressed in the context of this
 
SR. Partial Accident sequence notebooks
 
Initiating event notebooks and system model notebooks should document the dependencies noted in this
 
SR. ANO-2 Accident Sequence Analysis Work Package, PRA-A2-01-003S01, Rev. 0, Jan 2006 Documentation No impact is expected for documentation
 
issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-12 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-C2 (AS-C2 continued) (a) the linkage between the modeled initiating event in the Initiating Event Analysis section and the accident sequence model; (b) the success criteria established for each modeled initiating event including the bases for the criteria (i.e., the system capacities required to mitigate the accident and the necessary components required to achieve these capacities); (c) a description of the accident progression for each sequence or group of similar sequences (i.e., descriptions of the sequence timing, applicable procedural guidance, expected environmental or (continued)
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 AS-C2 (AS-C2 continued) (c) continued phenomenological impacts, dependencies between systems and operator actions, end states, and other pertinent information required to fully establish the sequence of events); (d) the operator actions reflected in the event trees, and the sequence specific
 
timing and dependencies that are traceable to the
 
HRA for these actions; (e) the interface of the accident sequence models with plant damage states; (f) [when sequences are modeled using a single top event fault tree] the manner in which the requirements for accident sequence analysis have been satisfied.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 SY-A4 Perform plant walkdowns and interviews with system engineers and plant operators to confirm that the systems analysis correctly reflects the as-built, as-operated plant. ANO-2 has a System Notebook Database as part of their overall PSA documentation system. The System Notebook entries for EFW, SW and the AC Power System were reviewed against the list of information provided in the SR. There was no specific entry in the reviewed notebooks that address performing plant walkdowns and interviews with system engineers and plant operators to confirm that the systems analysis correctly reflects the as-built, as operated plant. Walkdowns will need to be performed in order to support NFPA 805 FPRA and Flooding
 
initiator. This SR can be accomplishment during this process. Partial Walkdowns have been performed of the ANO-2 Systems and documented in the Internal Floods analysis. Documentation of walkdowns and insights gained from
 
these walkdowns are to be documented in future updates to the system analysis work packages.
ANO-2 PRA Model System Analysis Work Package, PRA-A2-01-003S11, Rev. 1, May 2008 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-15 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 SY-A8 ESTABLISH the boundaries of the components required for system operation. MATCH the definitions used to establish the component failure data. For example, a control circuit for a pump does not need to be included as a separate basic event (or events) in the system model if the pump failure data used in quantifying the system model include control circuit failures. MODEL as separate basic events of the model, those subcomponents (e.g.,
a valve limit switch that is associated with a permissive signal for another component) that are shared by another component or affect another component, in order to account for the dependent failure mechanism. Boundaries are established for components required for system operation. One super component (EDG) that was to be addressed to close a self assessment observation was not completed. Finding - The EDG air start system is included in the component boundary of the EDG for failure rate and common cause, but is still modeled in the fault tree with non-zero
 
probabilities. Met The EDG air start system is considered part of the EDG component boundary as discussed in NUREG/ CR-6928 and Echelon Calculation PRA-ES-01-003. Therefore, the diesel air start basic events have been set to
 
a probability of 0.0 in the ANO-2 database. The
 
events are, however, retained in the
 
model for Equipment Out Of Service (EOOS) purposes.
ANO-2 PRA Model System Analysis Work Package, PRA-A2-01-003S11, Rev. 1, May 2008 Technical This issue has been addressed and resolved.
Therefore, there is no impact on the NFPA 805 submittal Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-16 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 SY-B8 IDENTIFY spatial and environmental hazards
 
that may impact multiple systems or redundant components in the same system, and ACCOUNT for them in the system fault tree or the accident sequence evaluation.
Example: Use results of plant walkdowns as a source of information regarding spatial/environmental hazards, for resolution of spatial/ environmental issues, or evaluation of the impacts of such hazards. A review of three system notebooks and supporting documents (STMs) did not identify an organized assessment of spatial and environmental hazards that may impact multiple systems or redundant components in the same system. Only for the service water system was a common hazard noted. It was related to blocking of the intake structure in an assumption on page 8 for the traveling screens due to icing and shad runs. Partial Internal Events System Analysis should discuss use of spatial and environmental hazards that may impact multiple systems or redundant components in the same system. ANO-2 PRA Model System Analysis Work Package, PRA-A2-01-003S11, Rev. 1, May 2008 Documentation No impact is expected for documentation issues. It should be noted that the internal flooding analysis has evaluated spatial and environmental hazards.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 HR-C2 INCLUDE those modes of unavailability
 
that, following completion of each unscreened activity, result from failure to restore: (a) equipment to the desired standby or operational status (b) initiation signal or set point for equipment start-up or realignment (c) automatic realignment or
 
power ADD failure modes identified during the collection of plant-specific or applicable generic operating experience that leave equipment unavailable for response in accident sequences. ANO-2 does include the pre-initiator events as appropriate with respect to this SR. There is no direct evidence that ANO-2 evaluated plant-specific or generic operating experience to check for other pre-initiators. Documentation of a review of plant specific information or LERs from other similar plants and incorporating this information into the HRA assessment is required to receive a CC-II/III rating.
Met  CC-I Provide documentation
 
of review of Licensee Event Reports (LERs) and plant specific information relating to potential pre-initiator events ANO-2 Human Reliability Analysis /
Rule Recovery Calculation, PRA-A2-01-003S03, Rev. 2, Dec
 
2008 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-18 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 HR-D3 For each detailed human error
 
probability assessment, INCLUDE in the evaluation process the following plant-specific relevant information: (a) the quality of written procedures (for performing tasks) and administrative controls (for independent review) (b) the quality of the human-machine interface, including both the equipment configuration, and instrumentation and control layout. Not assessed consistent with CC II since the evaluation does not provide an assessment of the quality of the procedures or the quality of the human-machine interface.
Met  CC-I The procedures were reviewed during the HRA evaluation. No issues were noted during development.
Provide documentation in future revision relating to the Quality of procedures and Equipment configuration.
ANO-2 Human Reliability Analysis /
Rule Recovery Calculation, PRA-A2-01-003S03, Rev. 2, Dec
 
2008 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-19 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 HR-D6 PROVIDE an assessment of the uncertainty in the HEPs in a manner consistent with the quantification approach. USE mean values when providing point estimates of HEPs. ANO-2 uses the HRA Toolbox for quantifying their pre-initiator HEPs. For the pre-initiator HEPs, ANO-2 basically uses the ASEP approach and treats the ASEP Basic HEPs as means with the associated error factors. However, as defined on page xv of NUREG/CR-4772, the ASEP BHEP values are medians for a log-normal distribution. Thus, the treatment of the BHEP values for the pre-initiators is mathematically incorrect. Met The HRA Toolbox Excel Spreadsheets have been revised to
 
covert the median values to mean values. This includes changes to the median values for the "General Types of Execution Failures and Associated Probabilities" ANO-2 Human Reliability Analysis /
Rule Recovery Calculation, PRA-A2-01-003S03, Rev. 2, Dec
 
2008 Technical This issue has been addressed and resolved.
Therefore, there is no impact on the NFPA 805 submittal.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 HR-G6 CHECK the consistency of the post-initiator HEP quantifications. REVIEW the HFEs and their final HEPs relative to each other to check their reasonableness given the scenario context, plant history, procedures, operational practices, and experience. There is no evidence that this consistency check has been done. If done this should be documented, if not done this should be completed. This can be addressed by adding an explicit process for reviewing the HEPs for internal consistency with respect to scenario, context, procedures and timing.
Specifically this can evaluate the HEPs with respect to certain expected patterns such as increasing HEPs with decreasing time available, increasing HEPs with increasing stress levels, and increasing HEPs with increasing complexity of the procedures for accomplishing the desired successful outcome. A statement that such an evaluation was performed and, where there were deviations from the expected patterns, and either provide a basis for the deviation or what was done to correct it. Partial The methods applied in the assessment of human reliability for the
 
PSA of ANO-2 are intended to support both a reasonable and defensible assessment of risk and the ability to perform meaningful applications of
 
the PSA models. A consistency check has been done that ensures that the HRAs are reasonable for the scenarios considered. This review will be documented in future updates. ANO-2 Human Reliability Analysis /
Rule Recovery Calculation, PRA-A2-01-003S03, Rev. 2, Dec
 
2008 Documentation No impact is expected for documentation issues Arkansas Nuclear One - Unit 2 Att. U - Internal Events PRA Quality Enclosure 1 to 2CAN121202 Page U-21 Table U-1 Internal Events PRA Peer Review - Findings and Observations (F&Os)
SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 HR-G9 Characterize the uncertainty in the estimates of the HEPs in a manner consistent with the quantification approach, and PROVIDE mean values for use in the quantification of the PRA results. NUREG-1278 contains median values that do not appear to be converted to means before being used in the ANO-2 PRA. For example, spread sheet used for HRA at ANO-2. Met The HRA Toolbox Excel Spreadsheets have been revised to
 
covert the median values to mean values. This includes changes to the median values for the "General Types of Execution Failures and Associated Probabilities" ANO-2 Human Reliability Analysis /
Rule Recovery Calculation, PRA-A2-01-003S03, Rev. 2, Dec
 
2008 Technical This issue has been addressed and resolved.
Therefore, there is no impact on the NFPA 805 submittal.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 DA-A1a ESTABLISH definitions of SSC boundaries, failure modes, and success criteria in a manner consistent with corresponding basic event definitions in Systems Analysis (SY-A5, SY-A7, SY-A8, SY-A10 through SY-A13 and SY-B4) for failure rates and common cause failure parameters, and
 
ESTABLISH boundaries of unavailability events in a manner consistent with corresponding definitions in Systems Analysis (SY-A18). Boundary developed for EDG starting air was outlined in PRA-ES-01-003 included the start air system inside the component boundary. The CAFTA model had the starting air modeled with BEs set greater than zero, effectively placing the starting air outside the component boundary. Met The EDG air start system is considered part of the EDG component boundary as discussed in NUREG/ CR-6928 and Echelon Calculation PRA-ES-01-003. Therefore, the diesel air start basic events have been set to
 
a probability of 0.0 in the ANO-2 database. The
 
events are, however, retained in the model for EOOS purposes.
ANO-2 PRA Model System Analysis Work Package, PRA-A2-01-003S11, Rev. 1, May 2008 Technical This issue has been addressed and resolved.
Therefore, there is no impact on the NFPA 805 submittal.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 DA-C10When using surveillance test data, REVIEW the test procedure to determine whether a test should be credited for each possible failure mode. COUNT only completed tests or unplanned operational demands as success for component operation. If the component failure mode is decomposed into subelements (or causes) that are fully tested, then USE tests that exercise specific
 
subelements in their evaluation. Thus, one subelement sometimes has many more successes than another.  [Example: a diesel generator is tested more frequently than the load sequencer. IF the sequencer were to be included in the diesel generator boundary, the number of valid test would be significantly decreased.] CAT I given based on information listed in Procedure PRA-A2-01-003S05 which does not address decomposing the component failure mode into sub-elements (or causes) that are fully tested, then using tests that exercise specific sub-elements in their evaluation.
Met  CC-I Update procedure
 
CE-P-05.07 with process details that
 
ensure the requirements described in CAT II/III are met. ANO-2 PRA Model System Analysis Work Package, PRA-A2-01-003S11, Rev. 1, May 2008 Technical This is a refinement of
 
the data analysis via specific reviews of the testing procedures. However, since most components are assessed appropriately, the overall impact is expected to be
 
insignificant.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 DA-C12EVALUATE the duration of the actual
 
time that the equipment was unavailable for each contributing activity. Since maintenance outages are a function of the plant status. INCLUDE only outages occurring during plant at power. Special attention should be paid to the case of a multi-plant site with shared systems, when the Technical Specifications (TS) requirements can be different depending on the status of both
 
plants. Accurate modeling generally leads to a particular
 
allocation of outage data among basic events to take this mode dependence into account. In the case that reliable estimates or the start and finish times are not
 
available, INTERVIEW the knowledgeable (continued) Procedure PRA-A2-01-003S05 addresses evaluating maintenance outage as a function of plant status. CAT I given since there is no evidence
 
of INTERVIEW the plant maintenance and operations staff to generate estimates of ranges in the unavailable time per maintenance act for components, trains, or systems for which the unavailabilities are significant basis events. As a suggestion, needs to include interviews and shared equipment between ANO-1 and ANO-2 (i.e., air compressors) in procedure PRA--A2-01-003S05.
Met  CC-I Update procedure
 
CE-P-05.07 with process to perform interviews with plant maintenance and operations staff to generate estimates of ranges in the unavailable time per maintenance act for components, trains, or systems for which the unavailabilities are significant basis events. Document personnel interviews.
ANO-2 PRA Model System Analysis Work Package, PRA-A2-01-003S11, Rev. 1, May 2008 Documentation Operators and System Engineers were involved in the collection of plant data. Therefore, no changes in the data are expected. No impact is expected for documentation issues.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 DA-C12(DA-C12 continued) plant personnel (e.g., engineering, plant operations, etc.) to generate estimates of ranges in the unavailable time per maintenance act for components, trains, or systems for which the unavailabilities are significant basic events.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-C2c For each flood area not screened out using the requirements under other Internal Flood Supporting requirements (e.g. IF-B1b and IF-C5), IDENTIFY the SSCs located in each defined flood area and along flood propagation paths that are modeled in the internal events PRA model as being required to respond to an initiating event or whose failure would challenge normal plant operation, and are susceptible to flood. For each identified
 
SSC, IDENTIFY, for the purpose of determining its susceptibly per IF-C3, its spatial location in the area and any flooding mitigative features (e.g.,
shielding, flood, or spray capability ratings). Equipment height off floor appears to be not recorded for most of the equipment on the walk down sheets. For example, flood area TB-15-250 walkdown sheet on page 460 only 3 of 26 items listed include a height. Met The walkdown sheets have been updated to document the spatial
 
information relating to the components within the flood zones. These spatial impacts have been included in the
 
flooding walkdown reports and documented for the ANO-2 analysis.
ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Documentation This issue has been addressed and resolved. Also, internal
 
flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-C3 For the SSCs identified in IF-C2c, IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms. INCLUDE failure by submergence and spray in the identification process.
EITHER: (a) ASSESS qualitatively the impact of flood-induced mechanisms that are not formally addressed (e.g.,
using the mechanisms listed under Capability Category III of this requirement), by using conservative assumptions; OR (b) NOTE that these mechanisms are not included in the scope of the
 
evaluation. The walkdown sheets identify the components located inside the flood area. The SR requires that components in a flood area be identified and include whether the component is susceptible to failure by submergence or spray. The walkdown sheets are formatted to allow recording whether or not the component is vulnerable to spray. Only several walkdown sheets have the column filled out for vulnerable to spray. It was not documented whether blanks indicate not susceptible to spray or whether a blank indicates that it was not evaluated. Partial The walkdown sheets for the ANO-2 analysis have been updated to be
 
more comprehensive in the identification of those components vulnerable to spray. While most forms are complete in depicting the components susceptibility to spray, some entries remain blank and are in need of updating.
ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Documentation No impact is expected for documentation issues. Also, internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-C7 SCREEN OUT flood sources if it can be shown that (a) the flood source is insufficient (e.g., through spray, immersion, or other applicable mechanism) to cause failure of equipment identified in IF-C2c; OR (b) the area flooding mitigation systems (e.g., drains or sump pumps) are capable of preventing unacceptable flood levels and nature of the flood does not cause failure of equipment identified in IF-C2c (e.g., through spray, immersion, or other applicable failure mechanism); OR (c) the flood only affects the system that is the flood source, and the systems analysis addresses this per SY-A13 and SY-A14 and need not be treated as a separate internal flood event Not Reviewed Not applicable At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team.
ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-D5a GATHER plant-specific information on plant design, operating practices, and conditions that may impact flood likelihood (i.e., material condition of fluid systems, experience with water hammer, and maintenance-induced floods). In determining the flood-initiating event frequencies for flood scenario groups, USE a combination of (a) generic and plant-specific operating experience (b) pipe, component, and tank rupture failure rates from generic data sources and plant-specific experience; and  (c) engineering judgment for consideration of the plant-specific information collected Waterford basically calculates the initiating event frequency for each evaluated flood scenario using generic pipe break frequency data and plant-specific pipe lengths. Not Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed.
The Peer Review Team did perform a
 
review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for ANO-2. The ANO-2 analysis has since been completed, however, a review of plant specific failures is required to address the issue identified in this SR.
ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-D6 INCLUDE consideration of human-induced floods during maintenance through application of generic data. Operator error contributions to flooding are discussed at a very high level. However, basically the only floods considered were catastrophic failures. The flood scenario frequencies were quantified using generic pipe rupture data and plant-specific pipe length. The generic flood frequency sources do not include floods cause by human actions during maintenance. While the operator induced floods may be less severe than the catastrophic pipe failure floods, the frequencies will be higher so should be considered explicitly. Not Met Enhance the internal flood analysis by assessing human induced
 
floods in relation to the application of the generic data. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical While the need to assess human induced floods in relation to the application of the generic data is important for ensuring that
 
the flood frequency is inclusive, Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-D7 SCREEN OUT flood scenario groups if (a) the quantitative screening criteria in IE-C4, as applied to the flood scenario groups, are met, OR (b) the internal flood-initiating event affects only components in a single system, AND it can be shown that the product of the frequency of the flood, and the
 
probability of SSC failure given the flood is two orders of magnitude lower than the product of the non-flooding frequency for the corresponding initiating events in the PRA, and the random (non-flood-induced) failure probability of the same SSCs that are assumed failed by the flood. If the flood impacts multiple systems, DO NOT screen on this basis. Not Reviewed Not applicable At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for ANO-2.
However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E1 For each flood scenario, REVIEW the accident sequences for the associated plant initiating event group to confirm applicability of the accident sequence model. If appropriate accident sequences do not exist, MODIFY sequences as necessary to account for any unique flood-induced scenarios and/or phenomena in accordance with the applicable requirements described in paragraph 4.5.2. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E3 MODIFY the systems analysis results obtained by following the applicable requirements described in paragraph 4.5.2 to include flood-induced failures identified by IF-C3. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E3a SCREEN OUT a flood area if the product of the sum of the frequencies of the flood scenarios for the area, and the bounding conditional core damage probability (CCDP) is less than 109/reactor yr. The bounding CCDP is the highest of the CCDP values for the flood scenarios in an area. Not Reviewed Not applicable At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E4 If additional analysis of SSC data is required to support quantification of flood scenarios, PERFORM the analysis in accordance with the applicable requirements described 4.5.6 Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E5 If additional human failure events are required to support quantification of flood scenarios, PERFORM any human reliability analysis in accordance with the applicable requirements described in Tables 4.5.5-2(e) through 4.5.5-2(h). Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E5a For all human failure events in the internal flood scenarios, INCLUDE the following scenario
 
specific impacts on PSFs for control room and ex-control room actions as appropriate to the HRA methodology being used: (a) additional workload and stress (above that for similar sequences not caused by internal floods) (b) cue availability (c) effect of flood on mitigation, required response, timing, and recovery activities (e.g., accessibility restrictions, possibility of physical harm) (d) flooding-specific job aids and training (e.g., procedures, training exercises) Not Reviewed Not applicable At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E6 PERFORM internal flood sequence quantification in accordance with the applicable requirements described in paragraph 4.5.8. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E6a INCLUDE, in the quantification, the combined effects of failures caused by flooding and those coincident with the flooding due to independent causes including equipment failures, unavailability due to maintenance, and other credible causes. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E6b INCLUDE, in the quantification, both the direct effects of the flood (e.g., loss of cooling from a service water train due to an associated pipe rupture) and indirect effects such as submergence, jet impingement, and pipe whip, as applicable. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E7 For each flood scenario, REVIEW the LERF analysis to confirm applicability of the LERF sequences.
If appropriate LERF sequences do not exist, MODIFY the LERF analysis as necessary to account for any unique flood-induced scenarios or phenomena in accordance with the applicable requirements described in paragraph 4.5.9. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 IF-E8 CONDUCT walkdown(s) to verify the accuracy of information obtained from plant information sources and to obtain or verify inputs to (a) engineering analyses (b) human reliability analyses (c) spray or other applicable impact assessments (d) screening decisions Note: A walkdown may be done in conjunction with the requirements of IF-A4, IF-B3a, and IF-C9. Not Reviewed Met At the time of the Peer Review, the ANO-2 Internal Flood Analysis was not completed. The Peer Review Team did perform a review of the methodology to be used for the internal flood analysis using Waterford as a template since Entergy used the same methodology for
 
ANO-2. However, for some SRs, the peer review team could not perform a review without a complete ANO-2 analysis. The ANO-2 Internal Flood analysis has since been completed in accordance with the methodology reviewed by the peer review team. ANO-2 Internal Flooding Analysis, PRA-A2-01-004, Rev. 0, Feb 2010 Technical Internal flooding modeling issues do not impact fire risk.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 QU-D3 COMPARE results to those from similar plants and IDENTIFY causes for significant differences. For example:  Why is LOCA a large contributor for one plant and not another? ANO-2 used the MSPI Cross comparison report. The MSPI cross comparison compared component importances of similar plants for the five systems in the MSPI program. These are EDGs, EFW, HPSI, and Cooling Water. However, the MSPI comparison is based on component and system importance. The MSPI report does not allow comparison by Initiating Event and Sequence Type, which is the intention of the requirement, as indicated by the example.
Met  CC-I Perform additional comparison of results to similar plants (consider Initiating Events and Sequence types). This can be done via a review of the summary report from similar plants.
ANO-2 PSA Level-1 Model 4p02 Summary Report, PRA-A2 003, Rev. 2,  Dec 2008 Documentation Primarily considered a documentation issue and is not expected to change ANO-2 results. QU-F4 Document the key assumptions and key sources of uncertainty, such as; possible
 
optimistic or conservative success criteria, suitability of
 
the reliability data, possible modeling uncertainties (modeling limitations due to the method selected), degree of completeness in selection of initiating events, possible spatial dependencies, etc. Selection process for determining important assumptions and sources of uncertainty was not delineated.
A list was provided. Hard to determine if important items were documented. Partial ANO-2 has performed a sensitivity and uncertainty analysis on the current model. The ANO-2
 
sensitivity analyses compare favorably with the issues identified by EPRI, in its latest report on sources of uncertainty.
ANO-2 PSA Level-1 Model 4p02 Summary Report, PRA-A2 003, Rev. 2, Dec 2008 Documentation No impact is expected for documentation issues.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 LE-C2b REVIEW significant accident progression sequences resulting in a large early release to determine if repair of equipment can be credited. JUSTIFY credit given for repair (i.e., ensure that plant conditions do not preclude repair and actuarial data exists from which to estimate the repair failure probability [see SY-A22, DA-C14, and DA-
 
D8]). AC power recovery based on generic data applicable to the plant is acceptable. Repair was not addressed Met  CC-I Evaluate the potential to repair equipment in order to credit for mitigation.
ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Technical The current analysis for
 
the LERF results is assumed conservative with respect to potential for restoration of mitigation equipment.
No material impact is expected for this issue.
LE-C8a JUSTIFY any credit given for equipment survivability or human actions under adverse environments. For ANO-2, the systems with components inside containment and thus subject to the severe environment are the containment spray and fan coolers. ANO-2 does not credit sprays and fan coolers for averting LERF. The credited human actions are performed outside containment.
Met  CC-I Expand the documentation of PRA model results to address all required items ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Documentation No impact is expected for documentation issues.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 LE-C9a JUSTIFY any credit given for equipment survivability or human actions that could be impacted by containment failure. A review of Figures 4.1-1 and 4.1-2 of PRA-A2-01-003S12 indicates that ANO-2 does not appear to model equipment operation post-containment
 
failure. Met  CC-I Expand the documentation of PRA model results to address all required items.
ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Documentation No impact is expected for documentation issues. LE-C9b REVIEW significant accident progression sequences resulting in a large early release to determine if engineering analyses can support continued equipment operation or operator actions after containment failure that could reduce LERF. USE conservative or a combination of conservative and realistic treatment for non-significant accident progression sequences. A review of Figures 4.1-1 and 4.1-2 of PRA-A2-01-003S12 indicates that ANO-2 does not appear to model equipment operation post-containment failure. Met  CC-I Expand the documentation of PRA model results to
 
address all required items.
ANO-2 Large Early Release Frequency (LERF) Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Documentation No impact is expected for documentation issues.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 LE-C10 PERFORM a containment bypass analysis in a realistic manner. JUSTIFY any credit taken for
 
scrubbing (i.e., provide an engineering basis for the decontamination factor used). ANO-2 does not take credit for scrubbing for containment bypass events.
Met  CC-I Expand the documentation of PRA model results to address all required items.
ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Technical The current analysis for
 
the LERF results is assumed conservative with respect to potential for scrubbing for bypass sequences.
No material impact is expected for this issue. LE-D1b EVALUATE the impact of accident progression conditions on containment seals, penetrations, hatches, drywell heads (BWRs), and vent pipe bellows and INCLUDE these impacts as potential containment challenges, as required.
If generic analyses are used in support of the assessment, JUSTIFY applicability to the plant being evaluated.
There is no evidence of an evaluation of the impact of the accident progression conditions on containment seals, penetrations, etc. The model this is based on is related to NUREG/CR-6595 so consistency with NUREG/CR-6595 meets CC-I, but there is no discussion of the accident progression conditions on these elements.
Met  CC-I Expand the documentation of PRA model results to address all required items.
ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Documentation The LERF analysis meets
 
the requirements of NUREG/
CR-6595. No material impact is expected for this issue.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 LE-D3 PERFORM a realistic interfacing system failure probability analysis for the significant accident progression sequences resulting in a large early release.
USE a conservative
 
or a combination of conservative and realistic evaluation of interfacing system failure probability for non-significant accident progression sequences resulting in a large early release. INCLUDE behavior of piping relief valves, pump seals, and heat exchangers at applicable temperature and pressure conditions.
The ANO-2 ISLOCA analysis meets CC I which includes a plant specific assessment. A realistic analysis was not performed that examined the failure of piping segments, pump seals, etc. after the last isolation valve. Note that with an ISLOCA frequency of approximately E-12 these are not significant accident progression sequences so the current treatment is acceptable, but this frequency is questionable (see F&O IE-C12-01). Met  CC-I Expand the documentation of PRA model results to address all required items.
ISLOCA  Initiating
 
Event Frequency, PRA-A2-01-003S08, Rev. 2, Dec. 2008 Technical The current analysis for
 
the LERF results is assumed conservative with respect to potential for scrubbing for bypass sequences.
No material impact is expected for this issue.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 LE-D6 PERFORM containment isolation analysis in a realistic manner for the significant accident progression sequences resulting in a large early release. USE conservative or a combination of conservative or realistic treatment for the non-significant accident progression sequences resulting in a large early
 
release. INCLUDE consideration of both the failure of containment isolation systems to perform properly and the status of safety systems that do not have automatic isolation provisions. Containment isolation is addressed by top event (question) 3. This is based on a calculation that is noted not to have been maintained up to date. Since it has not been maintained up-to-date, there is no confidence that the analysis represents a realistic assessment; therefore, this does not meet CC II.
Met  CC-I Revise the identified supporting document and incorporate any
 
new information into the LERF model. ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Technical Although it is possible that some plant changes may have occurred
 
that are associated with containment isolation, this
 
issue is not expected to have a significant affect on the results. The current analysis results are considered sufficient for use in the analysis for NFPA-805 submittal.
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SR Category II Requirements ANO-2 Assessment Comments Met for ANO-2? Long-Term Resolution ANO-2 Ref. Technical or Documentation?
Expected Impact on NFPA 805 LE-E4 QUANTIFY LERF consistent with the applicable requirements of Tables 4.5.8-2(a),
4.5.8-2(b), and 4.5.8-2(c). Although the majority of the SR requirements in these three high level requirements are met, there is no indication that dependencies between multiple HFEs have been addressed.
Met  CC-I Perform dependency analysis between HRA combinations resulting from the Level 1 and Level 2 models ANO-2 Large Early Release Frequency (LERF)
Model, PRA-A2-01-003S12, Rev. 2, Feb 2010 Technical The Level 2 PRA only adds
 
a small number of human actions
 
that are primarily associated with containment isolation. Any new HRA combinations
 
that may have dependencies not accounted for are not expected to have a significant impact on LERF Frequency.
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-1 V. Fire PRA Quality V.1 ANO-2 Fire PRA Quality Review In accordance with RG 1.205 position 4.3:
 
"The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard "supporting requirements" important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable. Licensees should justify use of Capability Category I for specific supporting requirements in their NFPA 805 risk assessments, if they contend that it is adequate for the application. Licensees should also evaluate whether portions of the PRA need to meet Capability Category III, as described in the PRA Standard."
The ANO-2 Fire PRA (FPRA) has undergone a RG 1.200, Revision 2, Peer Review against the ASME PRA Supporting Requirements (SRs) by a team of knowledgeable industry (vendor and utility) personnel. The review was conducted by the Westinghouse Owners Group in June 2009 under LTR-RAM-II-09-046, "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 Arkansas Nuclear One, Unit 2 Fire Probabilistic Risk Assessment."  The conclusion of the review was that the ANO-2 FPRA methodologies being used were appropriate and sufficient to satisfy the ASME/ANS PRA Standard RA-Sa-2009. The review team also noted that the staff
 
appeared to be applying the NUREG/CR-6850 methodologies correctly.
The ANO-2 FPRA has also been subject to two additional focus scope peer reviews. During the evolution of the NFPA 805 project, some changes to the fire scenario methodologies were applied to both refine the model and results, and to more fully comply with approved methods.
The first of the focused reviews primarily evaluated changes associated with fire scenario selection (FSS) technical elements. Two ignition frequency (IGN) elements were also studied. This focused-scope peer review was conducted by the Westinghouse Owners Group in
 
October 2011 under LTR-RAM-II-11-064, "Focused Scope Fire PRA Peer Review for Arkansas Nuclear One Unit 2."
 
The second additional peer review focused solely on FSS technical elements. This review was performedin November 2012 by Kazarians & Associates (Mardy Kazarians) and is documented in 5384.R02.121122.DRAFT B, "Focused Scope Peer Review ANO-2 Fire PRA, FSS-A, C, D, E and H."  None of the other FPRA SRs involved changes in the Probabilistic Safety Assessment (PSA) Methodology as defined in ASME/ANS RA-Sa-2009.
 
The summary of the original peer review findings exhibited the following statistics for the evaluation of elements to the combined PRA Standard. For the ANO-2 FPRA, 81.3% of the SRs were assessed at Capability Category II or higher, including 7.6% of the SRs being assessed at Capability Category III. The ANO-2 FPRA had an additional 3.5% of the applicable SRs assessed at the Capability Category I (CC-1) level. The FPRA was found to not
 
meet 15.3% of the applicable SRs.
The Westinghouse Peer Group concluded that the ANO-2 FPRA is consistent with the ASME/ANS PRA Standard and supports risk-informed applications.
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-2 This attachment provides a detailed assessment of each of the findings identified by the Peer Review team. Table V-1 lists all findings of the original peer review and provides the ANO-2 disposition of each finding. Table V-2 lists each focus scope review finding and provides the ANO-2 disposition of each finding. Table V-3 lists those SRs that were assessed as CC-I and provides the disposition for CC-I acceptability in the NFPA application.
 
The Peer Review Reports (LTR-RAM-II-09-046, 5384.R02.121129.DRAFT B, and LTR-RAM-I-11-064) will be available for NRC review via the CERTREC portal.
 
V.2 Alternate Analysis Methods to NUREG/CR-6850 During the development of the ANO-2 FPRA, Entergy applied several unapproved analysis methods (UAM) that are not explicitly endorsed in the NUREG/CR-6850 guidance. Entergy believed that the approach taken with the UAMs provided a more realistic representation in the ANO-2 FPRA model. However, since these methods differ from that provided in NUREG/CR-6850, Entergy has performed a sensitivity analysis for each method to determine the impact against the accepted methods of NUREG/CR-6850. The ANO-2 sensitivity analysis is
 
documented in PRA-A2-05-010, Revision 0, "ANO-2 Baseline Sensitivity Analysis."  Additional details of the UAMs examined in the sensitivity analysis are provided below.
The ANO-2 FPRA model of record submitted with this LAR contains the UAMs. The completed sensitivity analysis described here has two purposes. First, the output of the analysis demonstrates that the unapproved methods and deviations from NUREG/CR-6850 applied in the baseline PRA are bounded by a conservative analysis. The evaluation shows that the current ANO-2 FPRA meets the acceptance criteria of RG 1.205. Secondly the sensitivity demonstrates that the ANO-2 FPRA meets the RG 1.205 requirements with UAMs removed.
 
The conclusion that the ANO-2 model meets the RG 1.205 requirements with UAMs removed is a key conclusion. In an effort to more fully comply with the guidance of NUREG/CR-6850, ANO-2 plans to revise the model of record to remove the UAMs (and match the sensitivity). This revised model is planned following the LAR submittal during the NFPA 805 implementation phase. The goal to revise the model to comply with the guidance is reflected in the listed sensitivity analysis. The analysis demonstrates that the current model of record and the proposed future model of record both meet the acceptance criteria of RG 1.205.
 
The following discussion provides details associated with the sensitivity. The analysis shows the impact of the following model aspects:  Update to EPRI developed ignition frequencies  Removal of electrical panel factor credit  Application of a floor value for NSP for long term fires V.2.1 Fire Ignition Frequencies An update to the ANO fire ignition frequencies from the NUREG/CR-6850 frequencies to the FAQ-08-0048 recommended EPRI ignition frequencies is documented in the sensitivity analysis.
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-3 ANO-2 Application The current model of record utilized the original NUREG/CR-6850 bin frequencies. This ANO fire ignition frequency analysis updated the NUREG data based on a review of the plant specific fire events from the period of January 1, 2001, through December 31, 2006.
 
ANO-2 Sensitivity Analysis The sensitivity analysis shows the impact of applying FAQ-08-0048 methods utilizing the updated EPRI frequencies. The ignition frequency update applies to compartment frequencies, individual fixed ignition source frequencies, and transient frequencies. This included a revision of the ANO-2 Control Room fire ignition frequencies to the FAQ-08-0048 values. Besides updating to the EPRI values, the sensitivity also includes a Bayesian update that extends the fire ignition frequency calculation (with plant specific data) through October, 2012. An analysis was also performed, as required by FAQ 08-0048, to evaluate the analysis results associated with the sensitivity evaluation to address the impact of the use of the original NUREG/CR-6850 ignition frequencies in this sensitivity evaluation.
 
V.2.2 Adjustment Factor for Electrical Cabinet Ignition Frequency The current methodology for defining a split fraction for panel fires which allows for evaluation of separate scenarios based on spread of the fire outside of the source panel was evaluated by the EPRI Alternate Methods review group. The method approved by the review group was used in the baseline analysis provided in the initial LAR submittal. However, the NRC member of this review group dissented from the review group disposition of the methodology. Therefore, the removal of the panel factors is included in the sensitivity evaluation.
ANO-2 Application The electrical panel severity factors developed by ERIN (ERIN factors) evaluate actual fire events in determining the fraction of fires impacting the source panel only, versus those impacting targets outside of the source panel. Since the plant fire data may include some credit for manual suppression, the use of the manual non-suppression factor using the distribution in FAQ 08-0050 may result in some double counting of the effectiveness of manual suppression.
This UAM affects electrical panel calculations in the hot gas layer (HGL) and multi-compartment
 
analysis (MCA).
ANO-2 Sensitivity Analysis The sensitivity documents the changes made to the scenarios to remove credit for the electrical panel factors and incorporate additional Fire Modeling results, as needed, in order to calculate new frequencies for the HGL/MCA results.
 
With adjustments to the ignition frequencies and panel factors, additional FPRA analyses were updated as part of the sensitivity evaluation. The delta risk calculations were revised with the updated inputs. The HGL and MCA were also updated. These changes to the delta risk calculations and HGL/MCA required some additional changes to the model used in the sensitivity (not currently in the model of record). These changes include a handful of changes to credited detection/suppression features, two new proposed modifications (included in Attachment S), and some more detailed HGL/MCA scenarios due to variations in screening
 
values.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-4 The new proposed modifications listed in the previous paragraph are planned modifications listed in Attachment S of this LAR. These include:
S1 A modification of the RCS head vent valves has been proposed in Fire Area G to preclude spurious operation in Fire Zone 2199-G.
S1 Incipient detection for Panel 2C75 in 2154-E, Fire Area B-4, was credited to reduce the probability of a hot gas layer.
V.2.3 Non Suppression Probability for Long Term Fires The application of the NUREG/CR-6850 guidance for the treatment of non-suppression probabilities recommends a minimum default value of 1E-3 (NUREG/CR-6850, Appendix P, Table P-3). NFPA 805, FAQ 08-0050, reviewed the non-suppression probabilities, but maintained the minimum default value of 1E-3.
 
ANO-2 Application
 
A non-suppression probability of 0 was applied for Hot Gas Layer (HGL) analysis of the impact on the fire zone of influence and Multi-Compartment Analysis (MCA) for manual suppression times greater than 60 minutes. This probability was based on the formula used to calculate the probability of non-suppression in FAQ 08-0050. The formula for non-suppression probability is
 
P(t)NS = exp[-(t
* C s)], where  is the corresponding mean suppression rate (1/time) and C s is a scenario-specific adjustment factor which is generally close to 1.
ANO-2 Sensitivity Analysis The sensitivity analysis was performed by changing the non-suppression probability for all fire scenarios with a fire response time greater than or equal to 25 minutes from 0 to 1E-3 in the HGL/MCA evaluation. The majority of the fire scenarios still screened with a non-suppression probability of 1E-3. For those scenarios that did not screen initially, a more detailed evaluation was performed (all screened deterministically with more detailed fire modeling).
 
V.3 Baseline Sensitivity Conclusions
 
The baseline sensitivity analysis examines an ANO-2 FPRA model that more fully complies with the approved methods (UAMs removed). The results show that removal of the unapproved methods does not shift model results beyond the program limits. The revised total CDF and LERF values remain less than 1E-04 (CDF) and 1E-05 (LERF) for a Region II submittal (RG 1.174). The delta CDF/LERF also meet the acceptance criteria of RG 1.205 of less than 1E-05 (CDF) and 1E-06 (LERF).
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-5  Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition CF-A1-01 Incorrect probabilities used in FRANC AlteredEvents Table Closed The FRANC AlteredEvents Table, Appendix F in the ANO-2 Scenario Report (ERIN doc. # 0247-06-0006.05 Rev. 0) appears to have incorrect probabilities in the following instances: 1) Events related to the letdown flow control valves 2CV4816 & 2CV4817 apparently used the wrong probability from NUREG/CR-6850 Table 10-2 for fire zones 2098-C & 2108-S. In these fire zones, database xx2.mdb shows that the applicable cables (2I016Q and 2I016N/P, respectively) are routed through trays instead of conduits as apparently assumed in Appendix F. NUREG/CR-6850 Table 10-2 allows probabilities in the range of 0.02 to 0.1 for inter-cable shorts in trays rather than the 0.01 cited in Appendix F which would be acceptable if the cables were in conduits. 2) Appendix F states the 0.0006 probability used for event YMP2P35BFF in fire zones 2100-Z and 2109-U-C comes from the HFE AHF2CSAS1P. In the HRA toolbox spreadsheet hfe_cp New Fire PRA HFEs (4-23-09).xlsm provided shows that the HEP for this HFE is actually 3.2E-4.3) Appendix F states the 0.22 probability used for event EB12A3XXXF in fire zone 2100-Z comes from the HFE EHF2A309XP. In the HRA toolbox spreadsheet hfe_cp New Fire PRA HFEs (4-23-09).xlsm provided shows that the HEP for this HFE is actually 2.2E-2. Use of incorrect probabilities in the FRANC AlteredEvents Table would introduce errors in results obtained using FRANC. Correct errors in probabilities used or document justification for values currently found in the FRANC AlteredEvents Table. The altered events table was revised to correct, and provide a summary of the basis for, the assigned probabilities. 1) Spurious actuation of these valves requires two, concurrent, proper polarity shorts to another 4-20mA cable. Table 10-2 of NUREG/CR-6850 provides a probability range of 0.02 to 0.1 for an inter-cable short from one multi-conductor cable to another in a tray for a DC circuit. Assuming an average probability of 0.06 for each short, the overall probability is 0.06
* 0.06 or 0.0036. This discussion is documented in Section 12 of the Fire Scenarios Report (PRA-A2-05-003). 2) YMP2P35BF was originally included in the model as an MSO for depletion of the RWT. However, it has since been removed due to the length of time required to deplete the RWT, the operator cues for rapid RWT depletion, reactor building (RB) Spray pump operation, and the alignment to the sump, if required. Therefore, this event is removed from the Altered Events list. 3) EHF2A309XP has been added to the fault tree. Therefore, the altered event for EB12A3XXXF has been removed.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-6 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition CF-B1-01 Other Affected SR FSS-A3, HRA-B1 Problems with AlteredEvents Table & Appendix F in the ANO-2 Scenario Report Closed The FRANC AlteredEvents Table, Appendix F, in the ANO2 Scenario Report (ERIN doc. # 0247-06-0006.05 Rev. 0) makes recurring reference to ANO Engineering Change EC13540, "ANO2 CABLE ROUTING EXCLUSIONS TO SUPPORT FIRE PRA FOR NFPA-805" to document that cables associated with a given event do not go through the cited fire zone. A review of EC13540 shows that it does identify the component and associated cables, but documents different fire zones than those listed in Appendix F. For example, events QAV200798C and QSV200798C for valve 2CV-0798-1 are assigned a probability of zero because the cables for that valve do not pass through fire zones 2007-LL and 2024-JJ. EC13540 shows that cables for valve 2CV 0798-1 do not pass through fire zone 2101-AA without any mention of fire zones 2007-LL and 2024-JJ. A review of PDMS data shown in database file xx2.mdb confirm that 2CV-0798-1 cables do not pass through fire zones 2007-LL and 2024-JJ, but EC13540 can not be used as a reference to support that conclusion. This is applicable to almost all of the events set to zero with reference to EC13540. In three instances, an event was incorrectly set to zero based on EC13540 when cables associated with the component actually do pass through the cited fire zone based on PDMS information found in database xx2.mdb: 1) Event QSV200798R for valve 2SV-0798-1 has an associated cable R2D2301A identified in EC13540 passing through fire zone 2099-W in tray EC122. 2) Event PMV210401R for valve 2CV-1040-1 has an associated cable R2B53B1E identified in EC13540 passing through fire zone 2200-MM in tray EC114. 3) Event ECB2A409XD for breaker 2A409 has associated cables 2A409A thru H and J thru M identified in EC13540 passing through fire zone 2200-MM in trays DA020, 030 thru 035, 037 & 039. (continued)The altered events table in Attachment F of the Fire Scenarios Report (PRA-A2-05-003) has been revised so that the reference to EC13540, "ANO-2 Cable Routing Exclusions to Support Fire PRA for NFPA-805," has been removed for the items mentioned in the finding. Where EC 13540 is referenced, the events can be tied to the EC directly.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-7 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition CF-B1-01 (continued)  Event ECB2A409XR for breaker 2A409 has a probability of 0.1 assigned for fire zone 2200-MM-B while two other events for the same breaker in that fire zone are assigned probabilities of 0. Note that EC13540 does not address this fire zone at all. Inadequate justification provided for setting numerous event probabilities to zero and in three instances event probabilities inappropriately set equal to zero. Use of incorrect probabilities in the FRANC AlteredEvents Table would introduce errors in results obtained using FRANC. Revise EC13540 to address the fire zones listed in the FRANC AlteredEvents Table and fix probabilities where re-analysis of cable routing shows that the cables actually do pass through the cited fire zones.
CS-A7-01 Other Affected SR ES-B1, FQ-D1 LERF/ Containment Bypass Closed Although cables are considered for one containment bypass event (MOV 2CV-3200-2 TRANSFERS OPEN), there are no other containment isolation failures considered in the LERF model. The containment purge system is not considered (it is screened from the internal events model) and other containment bypass paths may exist, but are not identified. There is currently an open internal event PRA F&O (LE-D6-01) on the containment isolation system, expressing no confidence that the analysis represents a realistic assessment. Cable and circuit failure modes affecting containment bypass are not completely considered or dispositioned in the Fire PRA plant response model for LERF. Review potential containment bypass pathways beyond those in the current internal events PRA, including containment purge, that could lead to a LERF. A review of the containment piping and venting paths were reviewed for impact on large early release. No LERF paths were identified for the internal events. Only two potential LERF paths were identified due to multiple spurious operations of the isolation valves:  the containment purge supply and return lines (penetrations 2V-1 and 2V-2). However, each of these lines includes two AOVs and an MOV for isolation. The valves are in series with different power supplies (i.e, red and green powered), have independent circuits, and power supplies that are separated by fire zone with the exception of the control room. The controls within the control room divided between two cabinets (2C16 and 2C17). These cabinets have sufficient separation and barriers to prevent a fire from spreading from one cabinet to another.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-8 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition CS-A8-01 Other Affected SR FSS-C5 Component and Cable Selection Report Thermoplastic Cable Closed ER-ANO-2003-0450-000 Rev 0 is credited in the self-review as determining that only 1% of cables at ANO-2 are thermoplastic and therefore of minimal significance. This ANO-generated document is not sufficiently complete to justify this conclusion. This engineering reply document states that 380 thermoplastic cables were identified in PDMS, but listed only 10 specific cables as being in "safety significant fire zones". A brief review of safe shutdown components conducted with ANO engineering personnel identified a safe shutdown neutron monitoring system as having thermoplastic cables which is not mentioned in the engineering reply document cited above. A complete plant-wide review of thermoplastic cable used for safe-shutdown components should be performed, documented and referenced in the Fire PRA Component and Cable Selection Report. SR requirement to include treatment of thermoplastic cable failure was not satisfied. CALC-ANOC-FP-09-00019, Rev. 0, EC-6964, "Safe Shutdown Cable Jacket Insulation Types at ANO," evaluates the cables at both units and concludes that of the over 4600 cables reviewed, less that 0.3% have thermoplastic insulation. This calculation also confirms that thermoplastic cables are not used in power supply circuits. Thus, thermoplastic insulation at ANO is of minimal significance. CALC-ANOC-FP-09-0019 is referenced in the Fire PRA Scenarios Report. The resolution of ANO-2 CS-A8-01 was reviewed and accepted by the ANO-1 Peer Review Team. CS-A9-01 Other Affected CS-A5 DC Circuits Proper Polarity Hot Shorts on Ungrounded DC Circuits Closed The ANO Fire PRA guidance was contradictory with respect to cable selection with respect to proper polarity hot shorts on ungrounded dc circuits, and it was unclear whether these hot shorts were appropriately considered. "Safe Shutdown Cable Analysis," Calculation Number 85-E-0087-24, Revision 1 was designated as the reference for the scope. Section 4.2.5 states: "For ungrounded DC circuits, two hot shorts of the proper polarity (without grounding) causing spurious operation is not considered credible except for high-low pressure interface components."  However, the ANO-2 self-assessment for CS-A9 states: "Postulated proper polarity hot shorts on ungrounded DC circuits (e.g., RCS head vents) were considered as documented in the PDMS database."  Further, in CALC-85-E0087-24, Safe Shutdown Cable Analysis, conflicting criteria was identified in Section 4.2 of the calculation. Criteria 4.2.2 states that all DC grounded and ungrounded circuits must consider any and all shorts, hot shorts, shorts-to ground and open circuits. Criteria 4.2.3 states that all ungrounded circuits (both AC and DC) will be analyzed as if the circuit is grounded to account for the possibility of experiencing a ground fault. (continued)
The F&O indicates that Criteria 4.2.2, 4.2.3, and 4.2.5 in CALC-85-E-0087-24 are contradictory and that this guidance for cable analysis should be changed. However, the criteria in CALC-85-E-0087-24 come from NUREG-1778 and Generic Letter 86-10, and provide a consistent approach to circuit analysis. This approach keeps the analyst from making false assumptions based upon a "single failure" safety analysis instead of more appropriately considering multiple concurrent failures expected as the result of a fire. The criteria in question are concerned with DC circuit analysis and ungrounded AC circuits which are similar to ungrounded DC in how they respond to circuit failures. The 125 VDC power and control circuits at ANO are ungrounded and typically only shields in low voltage DC instrumentation loops are grounded. (continued)
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-9 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition CS-A9-01 (continued)  However, criteria 4.2.5 then states that for ungrounded DC circuits, two hot shorts of proper polarity (without grounding) are not considered credible except for high-low pressure interface components. Criteria 4.2.5 conflicts with 4.2.2 and 4.2.3 for several reasons:  (a) 4.2.2 criterion requires that any and all hot shorts be considered, and (b) 4.2.3 criterion requires that ungrounded dc circuits be grounded, and it would only take one hot short of proper polarity from the same DC source to spuriously operate the high/low pressure interface. These criteria need to be aligned with each other. Additional contradictory evidence is found in the "NFPA-805 Fire PRA Modeling of Multiple Spurious Operations (MSO)" document CALC-ANO2-FP-09-00016, Rev. 0. For example, the disposition of PWR MSO 10 states that inter-cable hot shorts are not postulated due to the use of thermoset cables. The guidance and discussion of proper polarity hot shorts on ungrounded dc circuits is contradictory and the correct disposition of this requirement could not be determined. Clarify the position on proper polarity hot shorts in ungrounded dc circuits and document that the ANO-2 Fire PRA includes consideration of proper polarity hot shorts on ungrounded dc circuits, besides high-low interface components. Not grounding DC systems allows a single fault on either the positive or negative side of the circuit to occur without jeopardizing the function of the DC system. The criteria used for ungrounded systems provide a methodical approach as discussed below. Criterion 4.2.2 of CALC-85-E-0087-24 is a definition of the scope of DC circuits considered and prevents the exclusion of any circuit associated with safe shutdown equipment without engineering review/analysis. Criterion 4.2.3 of CALC-85-E-0087-24 is a simplified methodology to DC circuit analysis. Since safe shutdown deals with multiple failures it is possible to have failure in a separate circuit create the initial ground fault. Requiring that all ungrounded circuits (both AC and DC) be analyzed as if the circuit is grounded simplifies the analysis to determine circuit failure. A single short/fault in the circuit being reviewed can result in failure since a ground is already assumed to be established. This criterion requires all circuits, grounded and ungrounded, to be evaluated in the same manner. Criterion 4.2.5 of CALC-85-E-0087-24 will allow an exclusion of a specific circumstance where a cable in the subject circuit faults to a second cable (two hot shorts of proper polarity). This unique intercable short requires that it be of the appropriate DC voltage, proper polarity (positive to positive/negative to negative), occurs exclusive of any fire induced grounds, and not be a high-low pressure boundary. This criterion is the DC equivalent of Criterion 4.2.4 which is for 3-phase AC. (continued)
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-10 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition CS-A9-01 (continued)
To summarize, Criterion 4.2.2 provides a baseline, Criterion 4.2.3 a simplified methodology, and Criterion 4.2.5 is an exclusion for a specific set of circumstances. Therefore, the criteria are consistent with each other and no revision to CALC-85-E-0087-24 is necessary. Also, the F&O indicates that additional contradictory evidence is found in the "NFPA-805 Fire PRA Modeling of Multiple Spurious Operations (MSO)" document CALC-ANO2-FP-09-00016, Rev. 0. For example, the disposition of PWR MSO 10 states that inter-cable hot shorts are not postulated due to the use of thermoset cables. This inaccurate wording has been removed from CALC-ANO2-FP-09-00016. Inter-cable hot shorts are postulated in the fire PRA model and NSCA as described in the MSO report. CS-C4-01 Cable Selection/
Overcurrent Protection Closed Electrical distribution system over-current coordination and protection analysis was performed using the methodology of ANO Upper Level Document "Electrical Protection/ Coordination", ULD-0-TOP-12 Revision 3, 11/25/2002. This upper level document references the ANO2 plant specific analysis found in engineering calculation 84-E-0103-01, Revision 8, "General Criteria for Safety Bus", 01/21/2000. No references or discussion of this review is provided in the fire PRA documentation. Provide in the appropriate portion(s) of the fire PRA documentation references to the review of electrical distribution system over-current coordination and protection analysis. References to the review of electrical distribution system over-current coordination and protection analysis are required in Fire PRA documentation to satisfy this SR. Provide in the appropriate portion(s) of the Fire PRA documentation references to the review of electrical distribution system over-current coordination and protection analysis. Section 4.4 of the Component and Cable Selection Report, provides documentation that all circuits and electrical distribution buses credited in the fire PRA have been analyzed for proper over-current coordination and protection. A description of the processes used is included via reference to ULD-0-TOP-12, Upper Level Document ANO-1 and ANO-2 Electrical Protection/ Coordination, Rev. 3 Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-11 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition ES-C1-01 HRA Instrumentation Closed Task 12, Post Fire HRA was reviewed for detailed analysis performed for credited operator actions. For many of the recovery actions control room instrumentation and annunciators were credited as cues. In many cases, these instruments and annunciators (along with supporting cables and power supplies) were not added to the equipment list and validated for function. In some cases, the cues stated may not be necessary since it is expected that the operator takes the manual actions regardless of cues stated. In other cases, the operators would not know the actions are required unless the cues were available. Example of instrumentation needed:  RHGISORCPBO (Operators Fail to Isolate RCP Bleed-off) credits RCP bleed off temperature from annunciator 2K11 for knowledge that spurious operation has occurred. Example of instrumentation (annunciators) credited, though may not be needed include EHF2A309XP (Operator fails to locally close 2A-309 to restore off-site power to 2A-3). This action credits annunciators for low bus voltage and EDG (Emergency Diesel Generator) trouble. However, it appears that the step to verify 2A1 energized from Off Site power would be followed if verification could not take place due to loss of instrumentation. For each HFE, check credited instrumentation and add as appropriate to the equipment list. Ensure satisfaction of ES-B4 such that supporting power supplies are included in equipment/cable selection. At the time of the peer review, most of the Post Fire HRAs included information related to the instrumentation for operator cues in the Fire Scenarios report. Since the peer review, the HRA information was moved to a separate report (PRA-A2-05-007, Rev. 0). The HRA Notebook, Section 4.2 describes the correlation of instrumentation for HFEs to Appendix R instruments that have been confirmed to be available following a fire. The existing HFEs were evaluated to determine if non-Appendix R instrumentation is required. The additional detail on the HRA instrumentation had a minor impact on HRA probabilities and HRA credit in the Fire Scenarios. The addition of detail for the HRAs represents an enhancement of the justification for HRA credit in the fire scenarios and does not impact the Fire PRA methodology. PRA-A2-05-002, ANO2 Fire PRA New Human Failure Events, was revised to indicate the appropriate operator cues. Event RHGISORCPBO was renamed to RHF2BLOFFP to reflect the ANO basic event naming convention, however, this event is not currently credited in the Fire PRA model. EHF2A309XP has sufficient cues to ensure operator actions are taken; additional instrumentation is not necessary. ES-C1-02 HRA Instrumentation Closed Appendix R instrumentation for Steam Generator pressure and level were credited for manual operation of AFW based upon adequate instrumentation demonstrated by the Appendix R analysis. The Appendix R analysis only validated that a minimum of one train was available. Since the PRA model may credit other trains, validation that adequate instrumentation for that train is available to support manual operations was not performed. Evaluate instrumentation availability either within the PRM or through separate evaluation and document. As stated in the F&O, one train of level instrumentation will be available to the operators for manual operation of the AFW. Also, additional level and pressure indication is to be installed at a proposed AFW control and instrumentation panel outside the control room. The culmination of these instruments will ensure sufficient cues for all fire scenarios. In addition, the operator cues (continued)
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-12 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition ES-C1-02 (continued)    for manual start of the AFW pump is based on both SG instrumentation and failure of the MFW and EFW pumps to operate. Since the SG instrumentation is not the only cue used for this operator action and no fire scenario damages all SG instrumentation, the HEP for manual start of AFW is not impacted by a failure of one train of Appendix R instrumentation.
ES-D1-01 Other Affected SR HRA-B3 HRA Instrumentation Closed ANO Fire Scenarios Report, 0247-06-0006.05, Appendix E has numerous documentation issues: 1) Some Basic Events (e.g. AHF2CSAS1P) do not have entries under "Operator Cues and Components Credited with Providing Operator Cues for Pose Fire Safe Shutdown" column, even though the "Required Instrumentation Available" column is marked Yes. 2) Column titles contain references to footnotes, but the footnotes are not in the document 3) The instrumentation identified as required for operator actions are listed in noun format, not instrumentation identifiers (e.g. "Steam generators pressure indications") 4) No explanation is given regarding whether or not all instrumentation listed in column for operator cues & components 5) No link to documentation is provided to basis of conclusion in "Required Instrumentation Available" column Since there is information missing from Appendix E that is necessary for determining whether or not SR ES-C1 is met, this is a documentation finding. It also has an impact on the determination of SR HRA-B3. Revise Appendix E of the report to include missing information and to clearly explain the basis of conclusion of availability of a component; this could be achieved by adding a column to reference the Safe Shutdown Capability Assessment or Safe Shutdown Equipment List - whichever document is more relevant. At the time of the peer review, most of the Post Fire HRAs included information related to the instrumentation for operator cues in the Fire Scenarios report. Since the peer review, the HRA information was moved to a separate report (PRA-A2-05-007, Rev. 0). 1) The column "Operator Cues and Components Credited with Providing Operator Cues for Pose Fire Safe Shutdown" has been populated for all credited HFEs. 2) The footnotes associated with this Appendix have been moved to the Assumptions section of the HRA report. These assumptions address the justification of Fire HRA multipliers and the definitions of accessible and simple actions to account for fire impacts on HRAs. 3) The instruments for each operator action used in the Fire PRA are provided using the component identifiers in the Cue Instruments / Components column. 4) The Cue Instruments / Components column has been added to the table in Appendix A to provide the instruments
 
used. (continued)
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-13 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition ES-D1-01 (continued)    5) The availability of at least one train of Appendix R instrumentation has been verified through the deterministic safe shutdown analysis. The availability of non-Appendix R instruments is documented based on discussions with Operations.
ES-D1-02 Other Affected SR ES-C2, HRA-A3 HRA Instrumentation Closed Attachment C of the Fire Scenario Report 0247-06-0006.05 is the Fire PRA Simulator Review, which assesses the instrumentation needs of the operators during / following a fire event. The review was conducted by a simulator operator and a Senior Reactor Operator and is formatted such that all of the appropriate questions are addressed to ensure that the SR is met. However, the basis for the conclusions is not included in the report. The condition for meeting the Category II requirement is answered by the questions, "Do the operators only rely on the specifically identified instruments?" and "Which tiles cause the operator to take and immediate action?"  The answers are that they do not rely on only one indicator and that there are no immediate actions taken due to annunciation. There is no other supporting documentation to demonstrate that these responses are correct. The ANO team confirmed that the analysis had verified the responses, but there is no documentation of the basis of the conclusions; therefore, it is very difficult to recreate the same conclusions. Since critical documentation is missing, this is considered to be a finding as opposed to a suggestion. Revise Attachment C of the report to include either a discussion regarding the basis for the answers or at a minimum to reference the applicable document. Also, clarify the relevancy of the unanswered questions or provide a response. The HRA Notebook, PRAA205007 was revised to include the information discussed in this F&O.
Appendix A includes all of the instruments that would provide indications for the operator actions. In addition, OP-2203.049 verifies the Appendix R instruments that are available for the worst case fire in each fire area. The operators are trained to rely on these instruments for cues to perform the operator actions. COPD-001,  "Operations Expectations and Standards" Section 5.4.2 provides the following direction; "Numerous events within the industry have occurred because the Operator performing the reactivity change only focused on one parameter to determine core response, such as, only monitoring RCS average temperature following a reactivity addition and ignoring other critical parameters. This practice must be avoided. The individual as well as the Operations team must use diverse and multiple indications to monitor changes to the core as a result of reactivity manipulations." Therefore, a basis is available to support the conclusions of the simulator review.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-14 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition FSS-A4-01 Targets in ZOI - missing a specific conduit Closed Plant drawing E-2872 was reviewed to identify some sample targets (EC3021, EC3022, B4124, B4125, and J4519) within the zone of influence of 2D-35 (Scenario 2109-U-B). All the targets were listed in Attachment A, except B4124. A review and plant walk-down should be performed to determine if B4124 should be added to the list of targets in Attachment A of ERIN Report 0247-06-0006.05. Because this is a conduit, the location of conduits may not necessarily match the location in the drawing. If this conduit is located in the zone of influence of 2D-35, it will need to be listed as a target. Verify PDMS information is consistent with walk-down results and drawings. Conduit B4124 was determined to be within the zone of influence of 2D-35. Therefore, conduit B4124 was added to the list of targets for scenario 2109-U-B in Attachment A of the Scenario Report. This finding is associated with a specific scenario with a missing target. A review of drawings identified targets that were difficult to locate physically in the field. This review was performed for all scenarios in congested fire zones. The walkdown target list includes notes that identify some of the targets that were identified from the drawing review versus the original. FSS-A5-01 Ignition Frequency Closed The ignition frequency used in scenario 2111-T-B is for the electrical panels and not a battery charger. The CDF quantification for this scenario will need to be revised to use the correct ignition frequency. A related F&O FSS-D10-01 identified that a battery charger was missed from the ignition source walk-down sheets and the CDF quantification results were not correct. The CDF quantification for 2111-T-B has been revised to use the appropriate ignition frequency. See -PRA-A2-05-003 - the ANO-2 Fire Scenarios Report for details. The ignition source data sheet and walkdown sheet for compartment 2111-T was revised in CALC-08-E-0016-01 to include battery charger 2D-33. FSS-B1-02 Control room abandonment Closed The glass partition between the two control room fire compartments (129-F and 2199-G) is not a fire rated barrier. The calculation of the control room abandonment CDF for Unit 2 only considers the fire initiating frequency for Unit 2, and does not consider a fire in the Unit 1 control room. The fire ignition frequency for the Unit 2 control room abandonment is under estimated. Include the fire ignition frequency for a Unit 1 control room fire in the Unit 2 control room abandonment CDF. The Evaluation of Unit 2 Control Room Abandonment Times at the Arkansas Nuclear One Facility [ANO2-FP-09-00013] defines the conditions that are assumed to cause control room abandonment and reliance upon alternate shutdown actions. Abandonment times are assessed for ANO-2 for fire scenarios originating in the ANO-1 control room and the sensitivity study also assumes failure of the glass barrier separating the ANO-1 and ANO-2 control rooms.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-15 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition FSS-B2-01 Shutdown following MCR abandonment Closed The control room abandonment (CRA) scenario for Unit 2 (2199-G-B) assumes a CCDP of 0.1 for an alternate shutdown scenario. The basis for this CCDP in the Fire Scenarios Report states that this is "a conservative CCDP, probability of failure of shutdown from outside the control room, given that a deterministic analysis has been performed and feasibility of required actions have been assessed to ensure feasibility of this shutdown scenario for a worst case fire". This is not a sufficient basis to validate a CCDP value of 0.1. A determination of a bounding risk for the MCR abandonment scenario is not demonstrated with the assumed CCDP. A detailed Control Room Abandonment analysis [CALC-09-E-0008-10] was performed to calculate the CCDP for MCR Abandonment. FSS-C5-01 Thermoplastic Cables Closed Per ER-ANO-2003-0450-000, Revision 0, approximately 1% of the cables in the plant are thermoplastic. These cables are primarily COAX used to support instrumentation. The subject ER is limited in scope and does not identify Fire PRA cables they may be thermoplastic and require a lower damage temperature. Significance of effects of different types of cables could be enough to impact results. Determine which (if any) fire PRA credited cables are thermoplastic and use appropriate damage threshold in fire damage scenarios. CALC-ANOC-FP-09-00019, Rev. 0, EC-6964, "Safe Shutdown Cable Jacket Insulation Types at ANO," evaluates the cables at both units and concludes that of the over 4600 cables reviewed, less that 0.3% have thermoplastic insulation. This calculation also confirms that thermoplastic cables are not used in power supply circuits. Thus, thermoplastic insulation at ANO is of minimal significance. CALC-ANOC-FP-09-0019 is referenced in the FPRA Scenarios Report. FSS-C8-01 Fire wraps -Fire-Resistance Rating Closed ANO has not confirmed that the wrap will not be subjected to direct flame impingement or validation that the wrap is qualified for fire impingement was not performed. Lack of basis for wrap validation is significant information; therefore, F&O is finding rather than suggestion. Review credited raceway wraps and either validate that wrap is qualified for fire impingement or validate that wrap is not subjected to fire impingement. The FPRA analysis was revised such that it does not take credit for the 1-hour (Hemyc) fire wrap on service water pump cables in areas OO and B-6 or on charging pump cables in area DD. As shown in the Scenarios Report, in base scenarios for these areas, the wrapped cables are assumed to fail. For specific fire scenarios in these areas, the wrapped cables are assumed to fail if they are within the zone of influence of the fixed or transient ignition source. Thus, the FPRA results for these areas were calculated assuming that the 1-hour fire wrap does nothing to mitigate the extent of fire damage.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-16 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition FSS-D3-01 Fire growth and propagation within cable trays Closed Implementation of the generic fire modeling methods did not consider the effect of fire growth and propagation within cable trays. A review of fire modeling applications within fire compartment 2109-U shows that the HGL determination was based upon bounding estimates for cabinet heat release rates and did not validate that the value would be bounding with the inclusion of fire spread to the cable trays. This is a finding because the analysis may not be bounding if the heat release rate contribution of the cable trays would be involved in a cabinet fire scenario. Consider the effect of fire growth and propagation within cable trays to validate that the value would be bounding with the inclusion of fire spread to the cable trays. The generic fire modeling methods were updated to perform a case to account for hot gas layer effects on cable trays located one foot above the electrical panel. See Section 2.2 of the Multi-Compartment Analysis (PRA-ES-05-004). Therefore, the fire growth and propagation within the cable trays are addressed in the fire scenarios associated with bounding cabinet heat release rates such as with fire compartment 2109-U. FSS-D7-01 Site-specific suppression system unavailability Closed Site-specific suppression system unavailability values were not evaluated. Surveillance requirements are incorporated into site procedures, but specific unavailability values in the cable spreading room are not tracked for Fire PRA purposes. The SR is not met per NUREG/CR-6850 guidance because the methodology does not include maintenance contributions to unavailability, credit for manual actuation of the system, dependent failures, and plant specific data. F&O is a finding since the analysis did not include attributes necessary to meet the NUREG/CR-6850 guidance for uncertainty evaluations. Incorporate the required plant specific data regarding unavailability per NUREG/CR-6850 guidance. The ANO-2 Fire Scenarios Report (PRA-A2-05-003) discusses the review of the fire suppression systems credited in the Fire PRA. Explicit credit for suppression and detection systems is taken for the cable spreading room fire scenario only. The smoke detection system is credited for early detection of a fire in the cable spreading room. A review of impairments associated with the cable spreading room detection system indicated that only individual detectors were out of service for a limited period of time during the period from 2007 through 2009. Therefore, the unavailability of these systems is very low and is considered to be enveloped by the system unreliability data taken from NUREG/CR-6850.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-17 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition FSS-D10-01 Other Affected SR FSS-H10 Verification of ignition sources via walkdowns Closed Compartment 2111-T was one of those areas that was not walked down by ERIN, but was walked down by the FPRA Peer Review. A battery charger 2D33 is located in compartment 2111-T and is not listed in the Walk-down sheet or in the Ignition source data. Further review of the fire scenarios listed in ERIN Report 0247-06-0006.05 showed a fire scenario that takes into account battery charger 2D33 as the ignition source (FRANC scenario 2111-T-B in Attachment A). It is recommended that the fire areas that were not walked down (listed in Table 2-1 of CALC-08-E-0016-01) have the ignition sources re-verified by a walkdown or compared against the list of fire scenarios. Because a fire scenario for the battery charger 2D33 was evaluated using the wrong ignition source frequency (see F&O FSS-A5-01), the battery charger needs to be added to the plant walk-down and the data sheets to ensure the correct ignition frequency is used in the CDF quantification. Because the missing data was due to not performing the plant walk-down, it is recommended that the list of compartments in Table 2-1 of CALC-08-E-0016-01 be re-considered for walk-downs to ensure that all the applicable ignition sources are adequately evaluated as fire scenarios. The ignition source data sheet and walkdown sheet for compartment 2111-T were revised in CALC-08-E-0016-01 to include the battery charger. The walkdown sheets in CALC-08-E-0016-01 were developed during initial ignition frequency walkdowns. Subsequently, additional walkdowns were performed for fire scenario development. During the later walkdowns, access was obtained to some zones previously inaccessible. The notes from the subsequent walkdowns were reviewed to determine if components in addition to those in CALC-08-E-0016-01 were noted. A review of the walkdown sheets resulted in some additional components being added based upon the walkdowns. In addition, other changes were identified and incorporated addressing new information relating to the components identified.
FSS-E2-01 Other Affected SR UNC-A2 Justification for assumption of no target damage from ventilated cabinet fire within 30 minutes Closed Section 7.1.2.1 of the ANO Fire Scenarios Report 0247-06-0006.05 evaluates the severity factor for ventilated cabinets located in fire zones equipped with automatic detection. The evaluation assumes that no target damage will result if the fire is suppressed within 30 minutes. This is an application of expert judgment in lieu of plant-specific data or generic estimates. The justification of the basis for this application has not been adequately developed or supported. Justification has not been provided for the basis of the specific expert judgment discussed above and therefore may not be considered to be valid. Perform a more extensive study to justify the use of the 30-minute non-damage time based on fire brigade response. The non-suppression probabilities for electrical cabinet fires were changed based upon a methodology that has been submitted to the EPRI Fire PRA Methods Panel. These values are based on Panel voltage ratings and do not include any assumptions for suppression times.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-18 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition FSS-G2-01 Multi-Compartment Analysis Fire barrier failure probability Closed Multi-Compartment Screening Analysis methodology includes a manual suppression probability based upon the rating of the fire barriers and that they would withstand a fire for a minimum of 90 minutes. A barrier failure probability is used based upon the bounding assumption that each barrier includes a normally closed door (7.4E-3). In the screening method, the value of manual suppression is multiplied by a barrier random failure probability. Considering that the manual suppression probability is based upon 90 minutes, these two values should not be ANDed. Given that the method is intended to be bounding for purposes of screening, this method may artificially reduce the CDF by up to 3 orders of magnitude. This F&O was written on the analysis when credit for manual non-suppression probability was based on the rating of the barrier. This approach was corrected in the latest revision of the MCA/HGL analysis. The current analysis credits a manual non-suppression probability based on the time required to generate a hot gas layer. The original approach would be susceptible to a dependency between the non-suppression probability and the barrier failure probability. In the current approach the non-suppression probability is not related to the barrier rating and therefore has no significant dependency with respect to the barrier failure probability.
FSS-H7-01 Other Affected SR FSS-E2 Fire brigade response time Closed Section 7.1.2.1 - Severity Factor for Ventilated / Open Cabinets from the Fire Scenarios Report describes values credited in fire brigade suppression prior to damage external to the effected cabinet. The assumptions related to this method have not been justified. The method assumes that fire detection will occur 30 minutes prior to external damage to cabinets. This assu mption has not been justified. NUREG/CR-6850 indicates a bounding assumption that in-cabinet detectors may provide an additional 5 minutes of time for detection. Therefore, without justification 30 minutes should not be used. In addition, the 30 minutes does not account for plant specific response times to the alarm. This value would also need to be considered. The lower branch does provide a 15 minute delay to fire brigade response. The 15 minutes included in NUREG/CR-6850 is loosely based upon control room indication of fire due to failed equipment that results from a developed fire. Justification for credit of this value would need to be provided on how operations would know to send fire brigade based upon a fire limited to a single cabinet with no fire alarm. The concern identified in this F&O no longer applies. The non-suppression probabilities for electrical cabinet fires were changed based upon a methodology that has been submitted to the EPRI Fire PRA Methods Panel. These values are based on Panel voltage ratings and do not include any assumptions for suppression times. Brigade response is only credited for transient fires in the cable spreading room. This is described in Section 9.0 of the Fire Scenarios Report (PRA-A2-05-003).
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-19 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition FSS-H7-02 Fire brigade capability to alter hot gas layer development Closed Section 9.1 of the Fire Scenarios Report contains the method used to evaluate overall compartment damage through hot gas layer development. This section of the report includes credit for fire brigade capability to alter hot gas layer development in 20 minutes which in essence results in a 1.0 probability that if a HGL (625 deg F) does not develop within 20 minutes it will not develop. This is an input/ assumption that is not clearly identified in the report. In addition, the basis for this value is not provided. The model employed does not discuss impacts that would support this time including fire brigade response time, fire characterization, and probabi lities for failure. Brigade response is only credited for transient fires in the cable spreading room. This is described in Section 9.0 of the Fire Scenarios Report (PRA-A2-05-003). This evaluation assumes that manual suppression would have to occur within 5 minutes and before damage to any cable trays. In this case, a hot gas layer would not have time to develop. HRA-A2-01 Other Affected SR HRA-B3 New fire specific HRA Procedures Open Identification of new operator actions for each fire scenario did not include fire area specific EOPs (Emergency Operating Procedures). This can result in the failure to identify operator actions which may de-energize equipment, lead to actions different from those assumed in the Fire PRA, and/or more direct symptom-based procedural actions. For example, for a fire in Fire Area B-3, the operator is specifically directed (by FIRES IN AREAS AFFECTING SAFE SHUTDOWN, 2203.049, Rev. 7) to verify the RCP breakers are OPEN with DC Control Power removed. This is different from the HRA write-up which assumes that 10 minutes will be used to attempt to trip the RCPs in the CR and that "when they [the operators] tried to trip the RCPs from the control room, they would not be successful and would decide to open the associated breakers."  This would be evaluated differently if procedurally guided. It was noted by the Fire PRA staff that the fire-related procedures were planned to be changed to make them more symptom-based and remove many of the steps currently in the procedures. The SR specifically requires, for each fire scenario, identification of any new fire-specific safe shutdown actions called out in the plant fire response procedures. The Fire PRA staff noted that the fire-related procedures were expected to be changed and did not want to add actions to the PRA that would have to be removed before applying the fire PRA. (continued) The new HRA events identified for the Fire PRA are documented in calculation PRA-A2-05-002. This calculation evaluates the probabilities for these new operator actions and any new combinations between these actions and the other HRA events. Each of the HFEs that were assessed for the transition to NFPA-805 and contain a reference to the guiding procedures within the HRA worksheets. Revisions to OP-2203.049 will be required prior to transition to NFPA-805 that will provide additional guidance to the operators for fire in each area of the plant. The transition to NFPA-805 will result in removal of many operator actions currently considered necessary under Appendix R. Therefore, many actions deemed necessary in the current version of OP-2203.049 are not required post transition to 805.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-20 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition HRA-A2-01 (continued)  SR HRA-A2 was assessed as "Not Met" for the reasons discussed above. To "Meet" this SR, ANO-2 should update the fire-related procedures as the fire PRA staff indicated to be planned. Following this update to the fire-related procedures, perform and document a systematic review of the plant fire response procedures and identify fire-specific safe shutdown actions that may be taken by the operator for inclusion in the PRA. It is recommended that all HFEs included in the model be tied directly to either the fire-related procedures or the normal EOPs.
HRA-A4-01 HRA documentation of operations or training personnel to confirm HRA modeling is consistent with plant operation Closed Attachment C of the Scenarios Report (0247-06-0006.05 Revision 0) provides the results of talk- and walk-throughs with operations personnel regarding instrumentation and indication issues during fire scenarios. However, there is no evidence of reviews or talk-throughs with operations or training personnel to confirm that the interpretation is consistent with plant operational and training practices during fire scenarios. SR HRA-A4 requires at least a review of the interpretations of procedures associated with actions in the PRA with operations or training personnel to meet Capability Category I. SR HRA-A4 was assessed as Capability Category I for the reasons discussed above. To achieve Capability Category II/III, ANO-2 needs to perform and document the review or talk-through with operations or training personnel to confirm the interpretation of procedures is consistent with both operational and training practices. This should be performed after the fire-related procedures are revised.  (See F&O HRA-A2-01) Although this SR only meets CC-I, the issue identified does not impact the FPRA results and is ultimately considered a documentation issue associated with ensuring that the information provided by the Shift Managers involved with the interviews are familiar with the Ops and training requirements and expectations. ANO has a high degree of confidence that the Shift Managers are familiar with these expectations since it is one of the responsibilities of the Shift Managers. In addition to the general talk-throughs of fire scenarios and simulator observations in Attachment C of the Fire HRA calculation (PRA-A2-05-007), ANO-2 has documented operator interviews of operator actions identified for the Fire PRA in calculation PRA-A2-05-002 and has identified operator cues and instruments required for each operator action.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-21 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition HRA-D2-01 Other Affected SR PRM-B9 HRA / availability of action during fire Closed The ECCS vent path is modeled as a spurious open vent path (2CV-4740-2 and 2CV-4698-1) equivalent to a medium LOCA. The model logic (Gate FIRE003) includes an operator recovery action (RHFPZRVNTISO) from the control room to isolate the vent path by closing valve 2CV-4740-2. However, 2CV-4740-2 is one of the spuriously opened valves, and may not be remotely operable. The ECCS vent path may not be remotely isolable given a spurious opening due to a fire. The recovery action assigned in the model may not be effective. A detailed circuit analysis of 2CV-4740-2 should be performed. Operator recovery action RHFPZRVNTISO, to close valve 2CV-4740-2, has been deleted from the FPRA model. Additionally, all valves requiring manual operation post fire have been evaluated for 92-18 concerns. If a valve has been identified to require manual manipulation after spurious operation and is indeed a 92-18 concern, the valve has been identified for modification within the FREs. IGN-A5-01 Ignition frequencies correlated to availability factor Closed Table 3-2: some IEFs are not corrected to a "per reactor-year" basis. Some ignition frequencies are based on all modes of operation (e.g., batteries) and are correctly updated with calendar years as described in Section 3.1 of Entergy Calculation CALC-08-E-0016-01 Revision 0. However for those ignition frequencies, plant availability must be factored into the initiating event frequency or it will be conservative by approximately 10%. IGN-A5 requires generic fire ignition frequencies or plant-specific fire frequency updates on a reactor-year basis. This is not done for the following ignition frequency bins: 1, 4, 8-10, 12-19, 23, 26, and 30. Multiply the frequencies of all-mode Ignition Frequencies by the availability factor (critical years/calendar years). IGN-A5 is "Met" because the plant-specific fire frequency updates were revised to reflect a reactor-year basis. The plant availability was used in determining the frequencies by the fraction of time the plant was at-power. CALC-08-E-0016-01 Table 3-2 has been changed to show that all bins were updated on a reactor-year basis.
IGN-B4-01 Other Affected SR IGN-A6 Ignition Frequency/ Bin/frequency error Closed It appears that the 5th and 95th frequencies in Table 3-2 of Entergy Calculation CALC-08-E-0016-01 Revision 0, does not match up with the 5th and 95th frequencies from Table C-3 of NUREG/CR-6850. Additionally, Bin 16 has been subdivided to include isophase bus ducts bins (16c and 16d) based on FAQ 0035, but this is not reflected in Table 3-2. (continued) Table 3-2 of CALC-08-E-0016-01 was revised to reflect the data from NUREG/CR-6850 Table C-3, as well as data from FAQ 07-0035 for bins 16c and 16d and data from FAQ 06-0017 for bins 16a and 16b. Per e-mail from the peer reviewer for this SR (L. Shanley) to J. Renner, dated 9/2/09, 1:36 pm; the portion of this F&O related to the uncertainty information was in error. No action is required.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-22 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition IGN-B4-01 (continued)
Table 3-2 of Entergy Calculation CALC-08-E-0016-01 Revision 0 does not accurately reflect the data used to develop ignition frequencies and compartment initiating event frequencies. Additionally, a spot check of the uncertainty information associated with the generic frequencies provided by the PRA team (FIF Bart Input Template (Updated 11-14-2008).xls) found one error (Bin 1 Range Factor is listed as 11.0 and it should be 10.0); other data appears to be correct. Correct Table 3-2 to reflect the correct information used for the Bayesian Update. Review all data used for the Bayesian update and correct any additional errors (beyond Bin 1), if found. It is also recommended that the document specify the prior distribution parameters used in the BART calculations (i.e., mean and range factor) to allow for reproduction of the update results. Provide details on isophase bus duct frequencies (bins 16c and 16d) from FAQ 0035.
PRM-B2-01 Internal Events PRA Open Items Closed The majority of the Internal Event deficiencies are open. The items have not been dispositioned such that their impact on the Fire PRA could be determined. A review of the items indicate one item in particular (timing for securing RCPs following a loss of CCW) that could have impact on the Fire PRA results. ANO-2 needs to review the disposition of open items from their level 1 internal initiator PRA to ensure that their disposition remains valid in view of the unique aspects of fires. The ANO-2 Internal Events Peer Review (LTR-RAM-II-08-020 id entified 26 findings against the Internal Events supporting requirements. These findings are discussed in Attachment U of this LAR submittal. A qualitative review indicated that most of these findings would have no or minor impact on fire risk. ANO-2 has created a new Fire HFE associated with tripping the RCP outside the control room. The system time window for this HFE uses a shorter available time than that assumed in the internal events analysis. This new available time is based upon WCAP16175-P-A "Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants." (continued)
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-23 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition PRM-B2-01 (continued)
The system time window for the HFE associated with tripping the RCP from within the control room (developed as part of the internal events PRA) will be revised in the next revision to the internal events PRA. The difference in failure probability using the shorter available time for this HFE does not change significantly due to the contribution of the Caused Based Method to the resulting failure probability.
PRM-B9-01 Other Affected SR HRA-D2 Plant Response Model Added Components for HRA actions Closed Components credited to support fire recovery actions have not been reviewed per the requirements of the applicable internal events standard. For example, valves added to support recovery actions that isolate spurious operation flow diversions do not always included credited power supply dependencies and random failure modes for the equipment. Review added components credited in HRA recovery and add appropriate basic events and links within the PRM. Document as required. All new fire recovery actions in the ANO-2 FPRA model have been reviewed per the requirements of the applicable internal events standard and appropriate basic events and links were added to the plant response model to ensure that dependencies are modeled. Therefore, plant response model (PRM)-B9 is "Met." PRM-C1-01 Plant Response Model Documentation of PRM development Closed The Supporting Requirement requires the Fire plant response model be documented consistent with HLR-IE-D, HLR-AS-C, HLR-SC-C, HLR-SY-C, and HLR-DA-E and their SRs in Section 2. No specific documentation could be found that explains the development of the Plant Response Model consistent with the SR requirements above. The supporting requirement to document the PRM (PRM-C1) is not met. The work supporting the PRM (including modifications to the internal events PRA) needs to be tied together in a summary document. Prepare a calculation report that documents the PRM in accordance with SR PRM-C1. This is a documentation issue only. The following discussion was added to Section 1 of the Component and Cable Selection reports linking the requirements of PRM-C1 to the internal events model. "The FPRA model uses the ANO Internal Events model as its basis. The methodology used to update the ANO Internal Events model to generate the FPRA model uses the same methodology used in the development and documentation of the Internal Events model. The FPRA documentation may differ from that used for the internal events model due to FPRA specific requirements, however, the basic intent of the internal events documentation requirements are met. (continued)
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-24 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status 1 Finding/Observation Disposition PRM-C1-01 (continued)
The FPRA therefore meets the same model development and documentation requirements applicable to the internal events model. Therefore, requirements of the ANSI/ASME standard applicable to the PRA model specified in PRM-C1 are met via the internal event model meeting the associated model requirements." Additional details on development of the PRM from the Internal Events PRA model can be found in Section 4 of the ANO-2 Component and Cable Selection Report. UNC-A1-01 Other Affected SR FQ-F1, UNC-A2 Uncertainty  of CDF and LERF results Closed The impact of parameter uncertainties were not estimated or propagated as required in SR QU-E1 and for LERF (LE-F3). Additionally, the uncertainties are not characterized as required by SR QU-E4. Appendix D lists and discusses sources of uncertainty; however the characterization of these uncertainties is not detailed. Uncertainty must be evaluated to include "ESTIMATE the uncertainty interval of the CDF results" per QU-E1, Capability Category I or II (different details, but both require uncertainty interval to be estimated). This is also required for LERF (LE-F3). Propagate or estimate the impact of parameter uncertainties on CDF and LERF. Alternatively, a defined basis can be developed to support the non-applicability of this SR. Appendix D of the Fire PRA Summary Report (PRA-A2-05-004) addresses sensitivity to the sources of uncertainty for the Fire PRA tasks. Additionally, PRA-A2-05-006, "ANO-2 Fire PRA Uncertainty/Sensitivity Analysis" has been developed to calculate the uncertainty associated with the CDF and LERF values of the Fire PRA. The information provided in these two documents satisfies the requirements identified in this F&O. MU-B4-01 Software upgrades Closed There is no reference to a peer review for upgrades. Did not find a section which addressed upgrades (not updates) to the PRA specifically involving changes to key PRA software. EN-DC-151 (PSA Maintenance and Update) procedure was revised to require an industry peer review if a PSA upgrade per ASME/ANS RA-Sa-2009 has occurred.
1 The Status of Closed indicates that the Peer Review Finding has been addressed in the NFPA-805 evaluation and incorporated into the Fire PRA model and documentation.
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-25  Table V-2 Focused Scope Fire PRA Peer Review - Findings and Observations SR Topic Status Finding/Observation Disposition FSS-D8-01 Fire Detection and Suppression System Open This SR requires an assessment of fire detection and suppression system effectiveness in the context of analyzed fire scenarios. The ANO-2 fire PRA credits the smoke-actuated dry pipe sprinkler system in the cable spreading room for preventing target damage from a transient fire. That is, the fire PRA only models target damage if the system fails. ANO-2 did not perform (or at least did not document) a technical basis that the suppression system is capable of preventing target damage. With the current model assuming a 317 kW fire, a relatively short fire growth phase, and the relative proximity of overhead cable trays to the postulated fire, the suppression system may not be capable of suppression prior to target damage. Suggested Resolution - Provide a technical basis that the cable spreading room suppression system can effectively extinguish a transient fire prior to target damage. As an alternate resolution, consider making this area a "zero-transient" area per EN-DC-161 such that a 69 kW transient (versus the current 317 kW) can be modeled per the ANO-2 transient methodology, in addition to being able to credit prompt suppression if the procedure revision requires a continuous fire watch when transient combustibles are present. The technical basis that the suppression system can extinguish the 69 kW fire prior to target damage should also be provided per this Supporting Requirement. The lower heat release rate has been justified as applicable in the cable spreading rooms and will be supported by a revision to EN-DC-161, which will restrict the combustibles allowed within the affected rooms. A Work tracking item (LO-WTANO-2010-00222 CA-00011) has been issued to ensure that this zone is included as a no transient zone. This action requires an update to the procedure for control of combustibles [EN-DC-161] to ensure "zero-transient" zones in the ANO-1 and ANO-2 Fire Scenarios Reports are designated and maintained. As part of the transition to an NFPA-805 licensing basis, the ANO-1 and ANO-2 Fire Scenarios Reports specify that certain fire zones must be maintained as "zero-transient" zones. Credit for manual suppression of a 69 kW transient fire has been credited for this fire zone. No specific credit for the automatic suppression systems has been taken. FSS-H1-01 Transient Fire size documentation Open The ANO-2 fire PRA documentation is generally adequate and this SR is met with an F&O. Procedure EN-DC-161 "Control of Combustibles" Rev 5 was reviewed against Table 8-1 of the Fire Scenarios Report. This review identified that four fire zones (2111-T, 2096-M, 2112-BB, and 2182-J) allowed transient combustibles (in some cases up to 300 pounds) with a roving fire watch. However, the fire PRA model postulates 69 kW transient fires in these areas consistent with its methodology for "zero-transient" areas. (continued) A corrective action has been written to revise EN-DC-161 to address this discrepancy. Work tracking item #LO-WTANO-2010-00222 CA-00011 has been issued to ensure that this zone is included as a no transient zone. This action requires an update to the procedure for control of combustibles [EN-DC-161] to ensure "zero-transient" zones in the ANO-1 and ANO-2 Fire Scenarios Reports are designated and maintained. As part of the transition to an NFPA-805 licensing basis, the ANO-1 and ANO-2 Fire Scenarios Reports specify that certain fire zones must be maintained as "zero-transient" zones.                  (continued)
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-26 Table V-2 Focused Scope Fire PRA Peer Review - Findings and Observations SR Topic Status Finding/Observation Disposition FSS-H1-01 (continued)
After some discussion, it was identified that ANO-2 plans to revise EN-DC-161 to require continuous fire watch for any transient storage in these areas. The Fire Scenarios Report should be revised to reference an ANO-2 action item or CR tracking implementation of this procedure revision. On a related note, ANO-2 may consider crediting prompt suppression for these areas. Suggested Resolution - Revise the Fire Scenarios Report to reference an ANO-2 action item or CR tracking implementation of revision to ENDC-161. While the action item has been written and incorporated into the Paperless Condition Reporting System (PCRS), it was not deemed prudent to incorporate a temporary work tracking item into the Fire Scenario Report. IGN-A7-01 Fire Ignition Frequency Apportioning Closed This SR relates to the fire ignition frequency apportioning methodology. ANO-2 meets this SR for transient ignition sources based on preponderance of evidence; however, one deficiency was noted with the turbine building transient fire frequency. In the turbine building, ANO-2 postulated 12 transient fire scenarios that could affect PRA targets. The area factor for each source was calculated as 100 sq. ft. divided by the total turbine building floor area. The total area factor for all turbine building transients was therefore 1,200 sq. ft. divided by the total turbine bui lding floor area. Based on inspection of plant drawings, the total floor area containing PRA targets of concern is greater than 1,200 sq. ft., and the ANO-2 model is therefore not accounting for some fraction of the turbine building transient fire frequency. The underlying cause of this error was using a nominal 100 sq. ft. for each transient and not postulating enough scenarios to cover the risk-relevant floor area. Alternatively, the area factor for each transient could be inflated to ensure the entire risk-relevant floor area is considered. Suggested Resolution
- Adjust the turbine building transient area factors such that the total floor area containing risk-relevant targets is considered. The floor areas assumed for the transient fires were adjusted based upon walkdowns and drawings to better reflect the area of the scenarios being developed. This approach increased the size of the floor area used in calculating the ignition frequency for each of the scenarios considered. This new information is documented in the Scenario Report Attachment H and Attachment D.
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-27  Table V-3 Fire PRA- Category I Summary SR Topic Status PP-B2 Credit for non-rated fire barriers PP-B3 Credit for spatial separation as a partitioning feature The ANO Plant Partitioning and Fire Ignition Frequency Development report uses the plant areas that are identified in the plant FPP. The ANO-2 Peer Review team noted two instances where this approach allows credit for non-rated fire barriers:  1) the glass partition between the ANO-1 and ANO-2 Control Rooms and 2) the Turbine deck, which does not have a rated barrier between the units. The MCR Abandonment analysis provides an evaluation of fires that would fail the partition and require both Control Rooms to be abandoned. The MCA evaluated fires on the turbine deck to determine if a fire in this area would spread to the other unit. The MCA did not identify any fires on the turbine deck that would spread to the other unit. Detailed evaluation of the impact to fires on the barriers was completed. The detailed evaluation of the non-rated barriers ensures the Category I limitation does not impact the results or conclusions of the model. PP-B5 Credit for active fire barriers The ANO Plant Partitioning was performed without taking credit for active fire barriers. Active barriers are used to separate divisional rooms. Not crediting the active barriers for partitioning only shifts the burden for fire risk evaluation to the Fire Scenario development process. Therefore, this assumption has a minor impact on overall fire risk. Active fire barriers were appropriately credited/modeled in scenario development. While this approach only meets the Category I requirement for portioning (PP), it does not significantly impact the overall model results or conclusions. CS-B1 Analyze Electrical Buses for Overcurrent Coordination Section 4.4 of the Component and Cable Selection Report (PRA-A2-05-005) documents the Electrical Coordination/Protection for ANO-2. This document refers to Upper Level Document ULD-0-TOP-12 for FPRA components in the SSEL. Components that are not in the SSEL are evaluated in Table 4-3 of the calculation. The difference between Category I and Category II is the use of an existing document (Cat I) instead of completing a new analysis for all modeled buses (Cat II). The Category I classification only partially applies. An existing analysis was used for FPRA components in the SSEL. Evaluation of non-SSEL buses meets the Category II requirement. Though the current method only partially meets the Category II requirement, it was judged to be acceptable for the NFPA 805 application. IGN-A10 Provide uncertainty intervals for fire ignition frequencies Section 6.2 of the ANO-2 FPRA Uncertainty/Sensitivity Analysis (PRA-A2-05-006) discusses the uncertainty intervals used for the fire ignition frequencies for propagating uncertainty. The uncertainty from ignition frequency development is primarily from the NUREG guidance. ANO strictly followed NUREG/CR-6850 guidance in developing ignition frequencies using numbers included in the document.
 
Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality Enclosure 1 to 2CAN121202 Page V-28 References
: 1. LTR-RAM-II-09-046, "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 Arkansas Nuclear One, Unit 2 Fire Probabilistic Risk Assessment," September 2009.
: 2. LTR-RAM-I-11-064, "Focused Scope Fire PRA Peer Review for Arkansas Nuclear One Unit 2," December 2011.
: 3. Echelon Calculation PRA-A2-05-003, "Fire Scenarios Report NUREG/CR-6850 Tasks 8 and 11," January 2012.
: 4. Entergy Engineering Change EC13540, "ANO2 Cable Routing Exclusions to Support Fire PRA for NFPA-805."
: 5. Echelon Calculation PRA-A2-05-005, "ANO-2 Fire Probabilistic Risk Assessment Component and Cable Selection Report,"
: 6. ANO Calculation CALC-ANOC-FP-09-00019, Rev. 0, "Safe Shutdown Cable Jacket Insulation Types at ANO," July 2009.
: 7. ANO Calculation CALC-85-E-0087-24, Rev. 1, "Safe Shutdown Cable Analysis," April 2006.
: 8. Echelon Calculation PRA-A2-05-002, "ANO-2 Fire PRA New Human Failure Events."
: 9. Echelon Calculation PRA-A2-05-007, Rev. 2, "ANO-2 Fire Probabilistic Risk Assessment Human Reliability Analysis (HRA) Notebook," November 2011.
: 10. ANO Calculation CALC-08-E-0016-01, "Fire Probabilistic Risk Assessment Plant Partitioning and Fire Ignition Frequency Development," March 2011.
: 11. ANO Calculation ANO2-FP-09-00013, Rev. 0, "Evaluation of Unit 2 Control Room Abandonment Times at ANO Facility."
: 12. ANO Calculation CALC-09-E-0008-010, "ANO-2 Fire Area G Risk Evaluations," November 2011.
: 13. Echelon Calculation PRA-ES-05-004, "Multi-Compartment / Hot Gas Layer Analysis."
: 14. Echelon Calculation PR-A2-05-004, Rev. 0, "Fire Probabilistic Risk Assessment Summary Report, NUREG/CR-6850 Task 16," January 2012.
: 15. Echelon Calculation PR-A2-05-006, Rev. 0, "ANO-2 Fire PRA Uncertainty / Sensitivity Analysis," January 2012.
: 16. Nuclear Management Manual Procedure EN-DC-151, Rev. 2, "PSA Maintenance and Update," January 2011.
: 17. Nuclear Management Manual Procedure EN-DC-161, Rev. 7, "Control of Combustibles," November 2012.
: 18. ASME/ANS RASa-2009 - ASME and ANS combined PRA Standard "Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment."
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-1 W. Fire PRA Insights W.1 Fire PRA Overall Risk Insights Risk insights were documented as part of the development of the FPRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for FPRA development and is useful in identifying the areas of the plant where fire risk is greatest. The risk insights generated were useful in identifying areas where specific contributors might be mitigated via modification. A detailed description of significant risk sequences associated with the fire initiating events that contribute above 1% of the calculated fire risk for the plant was prepared for the purposes of gaining these insights and an understanding of the risk significance of MSO combinations.
These insights are provided in Table W-1.
 
Fire Scenario Selection
 
Fire scenarios were selected based on the definition of 'significant accident sequence' from
 
RG 1.200, Revision 2:
 
Significant accident sequence:  A significant sequence is one of the set of sequences, defined at the functional or systemic level that, when ranked, compose 95% of the CDF or the LERF/LRF, or that individually contribute more than ~1% to the CDF or LERF/LRF.
There are 80 fire scenarios comprising 90% of the cumulative fire CDF and 107 fire scenarios comprising 95% of the cumulative fire CDF. Of these, 32 scenarios contribute 1% or greater on an individual basis. These 32 scenarios (all scenarios contributing 1% or greater on an individual basis) are presented in Table W-1. There is a strong correlation between CDF and LERF. Twenty-six of the top LERF scenarios are also included in the top 32 CDF scenarios presented in Table W-1, and all but two of the LERF scenarios contributing 1% or greater on an individual basis. The two remaining LERF scenarios that account for 1% or more of the total have also been added to Table W-1.
 
Modifications
 
Several modifications were identified in the FREs that contributed to reducing CDF and LERF to within the acceptable criteria. The risk benefits of these proposed modifications are reflected in the CDF and LERF risk values presented in Table W-2.
 
See Attachment S for a complete list of all modifications including additional details of each.
 
Recovery Actions Recovery actions were reviewed for adverse impact on the FPRA. Each human action credited in the FPRA model was evaluated in the ANO-2 Fire PRA Human Reliability Analysis Notebook (PRA-A2-05-007, Revision 0). None of the modeled actions was found to have an adverse
 
impact on the FPRA. Recovery actions were not cr edited given a fire in the room in which the action occurs, or through which the operators must pass to perform the action. Also, for the main control room, recovery actions were not credited given a fire in a panel needed to complete the action.
 
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-2 Safe Shutdown Analysis actions were also reviewed in developing the VFDRs to be used for assessing actions adverse to risk.
 
The risk associated with each VFDR was evaluated in the FRE process. The potential risk of each recovery action is bounded by the CDF and LERF provided in Table W-2. Also, the additional risk of recovery actions for an area was conservatively determined and is provided in Table W-2.
 
See Attachment G for the recovery actions credited in each area.
W.2 Risk Change Due to NFPA 805 Transition In accordance with the guidance in Regulatory Position 2.2.4.2 of RG 1.205, Revision 1:
 
"The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions).
The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk-neutral or represent a risk decrease."
W.2.1 Methods Used to Determine Changes in Risk Variances from Deterministic Requirements (VFDRs)
For a fire area that is not deterministically compliant under NFPA 805, Section 4.2.3.2, deterministic compliance strategies were identified as compensatory measures for the variances from the deterministic requirements (VF DRs). These strategies include NRC-granted exemptions, evaluations that determine specific equipment is free of fire damage (see discussion of embedded conduit below), application of electrical raceway fire barrier systems (wrap), or manual manipulations of controls and equipment in the control room and power block. The VFDRs are lack-of-separation issues and span the NFPA 805 nuclear safety performance criteria: reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring. The compensatory measures already in place for the VFDRs mitigate the risk of the associated failures, thereby ensuring that the risk from fire damage prior to transition to NFPA 805, although not quantified, is acceptable.
For the transition to NFPA 805, the risk from fire damage is quantified. Each fire area that is not deterministically compliant under NFPA 805, Section 4.2.3.2, is evaluated in the risk-informed, performance-based approach under NFPA 805, Section 4.2.4.2, by the Fire Probabilistic Risk Assessment (FPRA). The FPRA provides a current state-of-the-art fire PRA analysis performed using accepted methods.
 
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-3 The aggregate change in risk associated with the VF DRs in a given fire area is evaluated by comparing all fire scenarios in the given fire area against two distinct plant models within the
 
FPRA:  the compliant plant model, and  the post-transition plant model Compliant Case Analysis for a Fire Area The compliant case for an area represents the existing as-built, as-operated plant if all of the VFDRs in the area were eliminated; in other words, if the area was deterministically compliant.
Thus, the compliant case for each area was analyzed as follows.
 
The FPRA scenarios for the area were reviewed to determine which VFDR related components are modeled and for those modeled, which VFDR related components are failed by each specific fire scenario. VFDR related components not modeled within the FPRA (e.g., those involving HVAC systems not required to meet PRA success criteria) were determined to have no impact on the safety functions modeled in the FPRA and thus, no contribution to core damage frequency (CDF) or large early release frequency (LERF).
For each scenario, the specific VFDR related components, which if protected would eliminate the VFDR, were set to their random failure probability instead of to "failed by the fire."  Setting these components to their random failure probability provides an estimate of the fire risk if individual modifications were made to protect or reroute the components, thereby eliminating the VFDRs. The other components in the FPRA model that are impacted by the fire scenario are set to "failed by the fire."
 
Recovery actions (outside control room manual actions to mitigate the direct failure of VFDRs listed in Attachment G) were not credited in the compliant case. Non-recovery actions (manual actions to mitigate non-VFDR failures) were credited in the compliant case. This ensures that the compliant case represents the as-built, as-operated plant except for the eliminated VFDRs in the area, allowing for direct comparison with the post-transition plant model, which credits
 
recovery actions.
 
As a rule, proposed modifications (listed in Attachment S) were not credited in the compliant case. This ensures that the compliant case represents the as-built, as-operated plant, except for the eliminated VFDRs in the area, allowing for direct comparison with the post-transition plant model, which credits the modifications. One noted exception is Modification S1-3, "Backup DC control power to switchgear 2A-1, 2A-2, 2H-1 and 2H-2," which is conservatively credited in both the compliant plant model and the post transition plant model.
 
Post Transition Case Analysis for a Fire Area
 
The post transition case for a fire area represents the plant if the recoveries listed in Attachment G and the modifications listed in Attachment S are used to protect the plant from core damage, mitigating the risk imposed by the VFDRs.
For each scenario, the specific VFDR related components and other components in the FPRA model that are impacted by the fire scenario were set to "failed by the fire."  Some examples of the methods used in the post transition case are provided below.
 
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-4 VFDR components with 1-hour fire wrap; specifically, Service Water pump cables in Fire Areas OO and B-6 and charging pump cables in Fire Area DD were set to "failed by the fire". Since these wrapped components were set to "failed by the fire" in the post transition model, but not in the compliant model, they are included in the  risk calculations. Also, since the resulting risk meets the acceptance criteria, the fire wrap does not need to be maintained. See VFDRs B6-01, DD-03 and OO-01 in Attachment C. VFDR related components with NRC-granted exemptions were set to "failed by the fire."  Since these components were set to "failed by the fire" in the post transition model and the resulting risk meets the acceptance criteria, the NRC-granted exemptions for these components do not need to be maintained in the post transition licensing basis. See Licensing Actions sections of Attachment C for details. VFDR related components in embedded conduits were not set to "failed by the fire" since they were determined to be protected from fire damage as documented in Attachment J. Thus, they have no impact on the  risk calculations.
 
Recovery actions (outside control room manual actions to mitigate the direct failure of VFDRs listed in Attachment G) were credited in the post transition case. Non-recovery actions (manual actions to mitigate non-VFDR failures) were also credited in the post transition case.
 
The proposed modifications (listed in Attachment S) were credited in the post transition case. This ensures that the post transition case represents the plant following transition to NFPA 805 and allows comparison, with the compliant case which does not credit the modifications. The exception is Modification S1-3, "Backup DC control power to switchgear 2A-1, 2A-2, 2H-1 and 2H-2," which is conservatively credited in both the compliant plant model and the post transition plant model.
 
FPRA model changes to incorporate the proposed modifications (listed in Attachment S) and recovery actions (listed in Attachment G) were made using accepted methods. FPRA peer reviews were performed to assess the adequacy of the FPRA model and the results of the peer reviews are described in Attachment V.
To confirm the availability of operator cues for the recovery actions, the actions were correlated to fire safe shutdown analysis instrumentation. Since one train of fire safe shutdown analysis instrumentation is demonstrated to be available via the conservative deterministic post-fire analysis, these cues will remain available post-fire. Current fire procedures provide guidance to the operators for use of the fire safe shutdown analysis instrumentation as cues for evaluation of the need to perform actions. Confirmation of the availability of operator cues is documented in the ANO-2 Fire PRA Human Reliability Analysis calculations (PRA-A2-05-007 and PRA-A2 002).
Change in Risk for a Fire Area
 
Each scenario for a fire area is evaluated as described above, with the following exceptions. If a scenario does not impact VFDR related components, no analysis was performed since the post transition case is equivalent to the compliant case, and the CDF and LERF for that scenario are zero.
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-5 If a scenario contains a VFDR and is low risk (< 1E-08 CDF, < 1E-09 LERF), the scenario CDF and LERF are conservatively approximated by assuming the CDF/LERF for the compliant case is equal to zero. Therefore, the delta risk is bounded by assuming the delta risk is equal to the total risk of the post transition case results for
 
that scenario.
The change in risk for a fire area is computed as the sum of the post-transition plant results
 
minus the sum of the compliant plant results.
 
ANO-2 Control Room Analysis ANO-2 does not have a Primary Control Station outside the Control Room. Therefore, if Control Room abandonment is required, the current Alternate Shutdown Procedure for Control Room abandonment requires Operations personnel to control the plant via local control of components. Plant monitoring is performed from the Technical Support Center using the SPDS.
Post transition to NFPA-805, the plant response to a fire in Fire Area G will differ from the current response. The primary differences are due to the proposed modifications and the insights gained from the fire scenarios analyzed as part of the transition to NFPA-805.
Incorporating the proposed modifications into the Fire PRA analysis allows for analyzing the risk impact of maintaining RCS integrity and providing primary-to-secondary heat removal.
The ANO-2 Control Room is one of several fire zones included in Fire Area G. In order to support the transition to NFPA-805, the ANO-2 Control Room abandonment scenario, the non-abandonment scenarios, and additional scenarios associated with other fire zones within Fire Area G have been analyzed to determine the delta risk for the compliant cases vs. post transition bases. The information provided in the previous sections describing the process used for delta risk determination also applies to the methodology used for Fire Area G. The VFDRs for Fire Area G are provided in Attachment C.
 
In order to calculate the compliant cases for Fire Area G, the VFDRs associated with this fire area were used to determine the compliant cases by identifying the pertinent VFDRs that would be affected in each of the scenarios (i.e. both abandonment and non-abandonment). This process allowed for a compliant case to be developed and analyzed for each of the Fire Area G
 
scenarios.
 
The Post Transition case was analyzed by failing the components affected by the fire and using the Fire PRA model with modifications and recoveries as necessary to determine the risk.
In addition to the proposed modifications and recoveries identified as part of the risk analysis, additional Defense in Depth actions have been identified to enhance plant control and reduce the likelihood that additional equipment is damaged due to spurious operation.
Additional Risk of Recovery Actions
 
In the fire area risk evaluations, credit was taken for plant modifications in addition to the
 
credited recovery actions to meet the acceptance criteria. These proposed modifications were developed and scoped to reduce risk. One modification in particular, the newly proposed Auxiliary Feedwater (AFW) pump, is a significant modification and has been developed to be implemented with more reliable and redundant power supplies. Additionally, this proposed Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-6 modification will have fewer dependencies than the currently configured Emergency Feedwater (EFW) pumps. The design of the AFW pump will be developed to be more reliable than the existing configuration for the EFW systems when mitigating a post-fire transient. The risk reduction, after implementing the AFW pump in the PRA, results in a significant decrease in plant risk. This large decrease in risk enveloped the positive risk increase of the recovery actions and resulted in an overall negative delta CDF and LERF.
 
Regulation Guide 1.205, Section 2.2.4.1 (Revision 1), requires that the risk increase associated
 
with the recovery actions be reported to the NRC as part of the license amendment request (LAR). Therefore, the risk increase from crediting the identified recovery actions, which is enveloped in the overall negative delta CDF/LERF of the fire risk evaluations, was determined as described below.
 
The additional risk of recovery actions for an area was conservatively determined by calculating the difference in CDF and LERF using the post transition case model (with modifications incorporated into the model) in accordance with the guidance in FAQ 07-0030 "Establishing Recovery Actions."
The difference between the two cases of the same scenario (one recovery set to its nominal HEP values, and the other set to zero failure probability) provides an estimated, yet bounding, evaluation of the change in risk of the recovery actions in the area. By starting with the post-transition model, the modifications are credited in both cases of the risk of recovery analysis. This method of analysis removes the modification offset reported in the fire risk evaluation CDF and LERF in Table W-2.
 
W.2.2 Risk Acceptance Criteria
 
The change in CDF and LERF for each fire area is provided in Table W-2.
 
Total Change in CDF and LERF
 
The total change in CDF for this application is calculated to be -4.40E-04/yr (the sum of the calculated delta risk from Table W-2) and the total change in LERF is calculated to be -1.52E-05/yr. These values include credited recovery actions and plant modifications (documented in Attachments G and S, respectively). These changes in the plant CDF and LERF meet the RG 1.174 criteria as the total change in risk associated with the transition to NFPA 805 results are well within the acceptance criteria and the total plant fire risk is below 1E-04/yr for CDF and 1E-05/yr for LERF.
Site Risk from Internal Events
 
Although RG 1.174 does not require calculation of total CDF and LERF, if the increases are below the delta CDF and delta LERF of 1E-06/yr and 1E-07/yr respectively, it does recommend that if there is an indication that the CDF is 'considerably higher' than 1E- 04/yr or if LERF is 'considerably higher' than 1E-05/yr, then the focus should be on finding ways to decrease CDF
 
or LERF.
 
The total CDF including Fire and Internal events has a value of 6.7E-05/yr (Internal Events CDF (9.5E-7/year) + Internal Floods 8.0E-07/yr) + Fire CDF (6.5E-05)), and the total LERF has a value of 1.6E-06/yr (Internal Events LERF (1.1E-07/year) + Internal Floods (5.6E-08/yr) + Fire LERF (1.4E-06/yr)). Both values are below the RG 1.174 criteria of 1E-04/yr (CDF) and 1E-05/yr (LERF).
 
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-7 The aforementioned total CDF and LERF values do not include contribution from external events. Therefore, the contribution to risk from external events is captured below.
Site Risk from External Events
 
Seismic - The Operating Basis Earthquake for ANO is 0.1g and the Design Basis Earthquake for ANO is 0.2g. As part of the IPEEE submittal, ANO-2 performed a Seismic Margin Analysis (SMA). The results of the walkdowns that were performed as part of the SMA verified that the equipment, tanks, distribution systems, structures, and relays are able to withstand the 0.3g Review Level Earthquake at the plant and still provide for its safe shutdown. Based on an updated seismic hazard curve provided by EPRI, the likelihood of a seismic event exceeding 0.3 g peak ground acceleration is 9.28E-06/yr. Given the low seismic frequency with no seismic design outliers, the seismic CDF is estimated to be well below 1E-5/yr and LERF is estimated to be well below 1E-6/yr.
 
Flooding and other External Events - High winds, floods, or off-site industry facility accidents do not contribute significantly to ANO-2 site risk. For the external events the CDF is also estimated to be less than 1E-6/yr. This is consistent with the discussions of the events in Sections 2.3 through 2.11 of NUREG-1407.
 
A bounding estimate of the overall CDF risk due to external events (including seismic, external flooding, and off-site industry facility accidents) is estimated to be less than 1E-5/yr. A total bounding estimate for LERF external events is assumed to be 0.1 of the total CDF, which is less than 1E-6/yr.
 
Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-8  Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2098-C/A CPC Room Severe Fire 5.25% 5.25% The scenario represents hot gas layer fire in the Core Protection Calculator (CPC) room impacting all targets in the room. EFW pumps and AFW pump 2P-75 are unavailable due to fire damage. Top cutsets are associated with random failures of the new AFW pump operation resulting in loss of secondary heat removal as well as failure to trip RCPs in the control room resulting in an RCP seal LOCA. 9.26E-03 1.81E-03 3.40E-06 9.84E-082099-W/A West DC Equipment Room - Severe Fire 4.97% 10.23% This scenario represents a hot gas layer fire scenario in the West DC Equipment room impacting all targets in the room. EFW pumps unavailable. Top cutsets include random failures associated with failure to trip RCPs in the control room resulting in an RCP seal LOCA, as well as loss of AFW pump 2P-75 and new AFW pump operation, resulting in loss of secondary heat removal. 3.88E-03 1.88E-03 3.22E-06 7.71E-082108-S/A Electrical Equipment Room 4.42% 14.64% This case represents hot gas layer fire scenarios in the Electrical Equipment Room impacting all targets in the room. Top cutsets include random failures associated with failure to trip RCPs in the control room resulting in an RCP seal LOCA. 1.95E-03 1.46E-03 2.86E-06 4.47E-082199-G/A Control Room-Abandonment 4.08% 18.72% This case involves control room abandonment. Top cutsets associated are with failure of RCP trip at the switchgear as well as failure of the new AFW pump. 6.97E-02 3.79E-05 2.64E-06 3.52E-08 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-9 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2111-T/A Lower South Electrical Penetration Room 3.14% 21.86% This case represents hot gas layer fire scenarios in the Lower South Electrical Equipment Room impacting all targets in the room. Dominant risk contributor is dependent on in-CR (control room) action to trip RCPs, since component cooling water (CCW) is assumed failed in the model. High pressure injection is unavailable. 3.73E-03 5.45E-04 2.03E-06 4.81E-082098-L/A Cable Spreading Room 2.92% 24.78% This case represents a transient fire which is not suppressed via manual suppression and results in damage to all targets within the cable spreading room. Top risk contributors include spurious hot short of LTOP ECCS valves, 2CV-4698-1 and 2CV-4740-2, random failure of the new AFW pump, and failure to trip the RCPs at the switchgear. 4.75E-01 3.98E-04 1.89E-06 2.25E-082032-K/D Containment Building South Side - Fixed and Transient Fire 2.80% 27.57% This scenario is associated with ignition frequency for non-RCP scenarios, outside of secondary shield wall impacting all targets outside of secondary shield wall. Top cutsets include random failures associated with failure to trip RCPs in the control room resulting in an RCP seal LOCA and failure to initiate a safety injection (SI) signal after failure of associated instrumentation or failure of containment sump valves, preventing recirculation in support of mitigation of the seal LOCA. 5.82E-04 3.11E-03 1.81E-06 3.68E-08 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-10 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2199-G-AJ/A Control Room - Fire in panels 2C33-1 and 2C33-2 2.78% 30.35% This case represents a fire impacting 2C33-1 and 2C33-2 control room panels. Fire damage renders EFW pumps and AFW pump 2P-75 unavailable. Top cutsets associated with random failures include loss of the new AFW pump operation resulting in loss of secondary heat removal as well as failure to trip RCPs in the control room, resulting in an RCP seal LOCA. 9.00E-03 2.00E-04 1.80E-06 5.31E-082100-Z/A South Switchgear Room 2.73% 33.09% This is the base scenario representing a hot gas layer fire in the switchgear room impacting all targets in the room. The top cutsets associated with random failures include failure to trip RCPs in the control room resulting in an RCP seal LOCA as well as failure of new AFW pump and operator actions associated with alignment of EFW discharge valve alignment, resulting in loss of secondary heat removal. HPSI mitigation is unavailable in top cutsets due random failure of service water valves. 1.37E-03 2.92E-03 1.77E-06 4.06E-082073-DD-F/A Access Area, Pump Room, Tank Room - fire in local MCC 2B-62 2.66% 35.74% This case represents a fire at MCC 2B-62. Fire damage causes loss of EFW and AFW. Top cutsets associated with random failures include failure to trip RCPs in the control room. No HPSI available to mitigate RCS loss
. 1.56E-03 1.10E-03 1.72E-06 1.87E-082014-LL/A West HPSI, LPSI & Containment Spray Pump 2.35% 38.09% This case represents the base scenario fire, impacting all components/cables in the zone. Top cutsets associated with random failures associated with failure to trip RCPs in the control room resulting in an RCP seal LOCA. No HPSI available to mitigate RCS loss
. 1.56E-03 9.71E-04 1.52E-06 1.65E-08 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-11 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2007-LL/A East HPSI, LPSI and Containment Spray Pump Area And Gallery 2.35% 40.44% This case represents the base scenario fire for 2007-LL, impacting all components/ cables in the zone. Top cutsets associated with random failures associated with failure to trip RCPs in the control room resulting in an RCP seal LOCA. 1.58E-03 9.63E-04 1.52E-06 1.67E-082101-AA/A North Switchgear Room (2A3 room) 2.19% 42.63% This case represents a hot gas layer fire scenario in the switchgear room impacting all targets in the room. EFW pumps and AFW pump 2P-75 are unavailable due to loss of ability to open the isolation valves. Top cutsets associated with random failures include the new AFW pump operation resulting in loss of secondary heat removal as well as failure to trip RCPs in the control room) resulting in an RCP seal LOCA. 1.14E-03 2.79E-03 1.42E-06 4.57E-082040-JJ-Y/A Tank Rooms, Pump Rooms &
Corridors - Fire at MCC 2B-52 2.18% 44.81% The case represents a fire at MCC 2B-52. Top cutsets for random failures are associated with failure to trip RCPs in the control room resulting in an RCP seal LOCA. 1.57E-03 9.00E-04 1.41E-06 1.54E-082101-AA-B-NS/A North Switchgear Room (2A3 room) - Fire at Switchgear 2A3 1.95% 46.76% The case represents a fire at the 2A3 switchgear. EFW pumps and AFW pump 2P75 are unavailable due to loss of ability to open the isolation valves. Top cutsets for random failures include new AFW pump operation and failure to align EFW discharge valves resulting in loss of secondary heat removal. 1.14E-03 1.10E-03 1.26E-06 4.05E-082033-K/A Containment Building North Side 1.85% 48.61% The case involves a Containment building north side fire impacting all components in the fire zone. Top cutsets include operator failure to reset spurious main steam isolation signal (MSIS) along with random failure of new AFW pump operation. 3.27E-04 3.66E-03 1.20E-06 4.00E-08 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-12 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2199-G-O/A Control Room - Fire in panel 2C17 1.81% 50.42% Top cutsets associated with failure of RCP trip in the control room, as well as failure to prevent SG overfill and failure of 2P-75 and new AFW pump. HPI mitigation capability is unavailable. 3.90E-03 3.00E-04 1.17E-06 2.84E-082199-G-B/A Control Room - Fire in panels 2C04 (including 2C01, 2C02, 2C03) and adjacent panel 2C100 1.61% 52.02% Control Room's, Main Control Board fire. Top risk contributing cutsets require ex-CR action to trip RCPs, since CCW is assumed failed. High pressure injection is unavailable. 7.02E-02 1.79E-03 1.04E-06 1.41E-08 2200-MM-C/A Turbine Building - Fire at Switchgear 2A-1 1.50% 53.53% This case represents a fire at 4160V switchgear 2A-1. Top cutsets for this case are associated with random failure of EDG No. 1 and the new AFW pump. 7.01E-04 1.39E-03 9.72E-07 2.14E-082068-DD/A Hot Machine Shop 1.42% 54.94% This case represents a hot machine shop base scenario impacting all cables in the fire zone. Top cutsets include failure of RCP trip in the control room and failure to align new AFW pump. 1.59E-03 5.77E-04 9.18E-07 1.03E-082199-G-AI/A Control Room -Fire in panel 2C33 1.39% 56.33% This case represents a control room fire in panel 2C33. Fire damage leaves EFW pumps and AFW pump 2P-75 unavailable. Top cutsets are associated with failure of new AFW pump and failure to trip the RCPs in the control room. 9.00E-03 1.00E-04 9.00E-07 2.65E-082109-U-H/A Corridor and Motor Control Center - Fire at MCC 2B-51 1.33% 57.66% This case represents a fire at MCC 2B-51. Top cutsets are associated with failure of the new AFW pump components and failure to align EFW discharge valves. 6.61E-04 1.30E-03 8.60E-07 2.84E-08 2200-MM-D/A Turbine Building - Fire at Switchgear 2A2 1.32% 58.99% This case represents a fire at switchgear 2A-2. Top cutsets associated with random failure of EDG No. 1, the new AFW pump, and failure to trip the RCPs at the switchgear. 6.18E-04 1.39E-03 8.57E-07 1.96E-08 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-13 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2100-Z-B-NS/A South Switchgear Room - Fire at Switchgear 2A4 1.26% 60.24% This case represents a fire at the 2A-4 switchgear. Top cutsets associated with random failures associated with the new AFW pump operation and failure to align EFW discharge valves resulting in loss of secondary heat removal. 6.78E-04 1.20E-03 8.13E-07 2.65E-082136-I-B/A Health Physics and Corridor - Fire at panel 2C330 1.18% 61.42% This case represents a panel fire in 2C330. Scenario risk is dominated by failure of in-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 3.82E-03 2.00E-04 7.64E-07 1.82E-082150-C/A CPC Room 1.15% 62.57% The scenario represents hot gas layer fire in the CPC room impacting all targets in the room. Top risk contributing cutsets requires in-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 3.76E-03 9.19E-04 7.46E-07 1.79E-082154-E-TN10/A CEDM Equipment Room - Transient Fire 1.10% 63.68% Transient fire in 2154-E. Scenario risk is dominated by failure of ex-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 6.30E-02 1.36E-04 7.15E-07 7.89E-092154-E-TN8/A CEDM Equipment Room - Transient Fire 1.10% 64.78% Transient fire in 2154-E. Scenario risk is dominated by failure of ex-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 6.30E-02 1.36E-04 7.15E-07 7.89E-092154-E-TN9/A CEDM Equipment Room - Transient Fire 1.10% 65.89% Transient fire in 2154-E. Scenario risk is dominated by failure of ex-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 6.30E-02 1.36E-04 7.15E-07 7.89E-09 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-14 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)
Contribution Scenario Description
% of Total Cumulative Risk Insights CCDP IF CDF LERF 2154-E-TN11/A CEDM Equipment Room - Transient Fire 1.10% 66.99% Transient fire in 2154-E. Scenario risk is dominated by failure of ex-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 6.29E-02 1.36E-04 7.14E-07 7.86E-09 2200-MM/A Unit 2 Turbine Building 1.04% 68.03% This case represents a transient fire base scenario with non-RCP and 4 kV target impacts. Top cutsets associated with random failures associated with failure to trip RCPs from control room along with failure to align and initiate once through cooling or random failure of an EDG. 5.49E-05 1.27E-02 6.74E-07 1.79E-082072-R/A Upper Volume Control Tank Room and Lower Tank And Pump Room 1.04% 69.07% This case represents a base scenario (Appendix R type fire) in Fire Zone 2072-R. Top risk contributing cutsets requires in-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model. High pressure injection is unavailable. 1.56E-03 4.29E-04 6.71E-07 7.30E-09(1)2137-I-B/A Upper South Electrical Penetration Room and Hot Instrument Shop - Fire at 2B-61 0.97% 70.04% This case represents a fire at MCC 2B-61. Top risk contributing cutsets requires in-CR action to trip RCPs, since CCW is assumed failed in the ANO-2 Model, establish once through cooling, and random failure of AFW pump. 4.86E-04 1.30E-03 6.31E-07 2.04E-08(1)2076-HH/A Electrical Equipment Room 0.79% 70.04% This case represents a base case scenario in the Electrical Equipment Room impacting all targets in the zone. EFW Train B and AFW pump available. EFW Train A lost due to fire damage to the steam supply valves. Seal LOCA mitigation is available. 1.25E-04 4.07E-03 5.10E-07 1.60E-08(1) - Sequences included due to LERF contribution. These sequences contribute less than 1% of CDF, but account for 1% or more of the LERF total.
CCDP - Conditional Core Damage Probability IF - Ignition Frequency Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-15  Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF) 2MH01E concrete manhole east 4.2.4.2 4.89E-09 4.93E-11 yes no 4.89E-09 4.93E-11 n/a 2MH02E concrete manhole east 4.2.4.2 5.09E-09 5.13E-11 yes no 5.09E-09 5.13E-11 n/a 2MH03E concrete manhole east 4.2.4.2 9.99E-08 2.65E-09 yes no -5.66E-06 -1.89E-07 n/a 2MH01W concrete manhole west 4.2.3.2 8.10E-09 1.67E-10 no n/a n/a n/a n/a 2MH02W concrete manhole west 4.2.3.2 8.10E-09 1.67E-10 no n/a n/a n/a n/a 2MH03W concrete manhole west 4.2.3.2 9.23E-09 1.90E-10 no n/a n/a n/a n/a AA Fire Zone 2007-LL ("B" HPSI, LPSI, and Containment Spray Pump room and gallery) 4.2.4.2 1.52E-06 1.67E-8 yes no -9.70E-07 -3.22E-08 n/a AAC Fire Zones SBOD and 2MH12 (alternate AC diesel and nearby manhole) 4.2.3.2 5.11E-08 1.06E-9 no n/a n/a n/a n/a Admin administration building 4.2.3.2 n/a n/a no n/a n/a n/a n/a B-2 miscellaneous turbine building fire compartments 4.2.4.2 5.41E-06 1.23E-07 yes yes -1.19E-04 -3.98E-06 2.08E-06/ 2.79E-08 B-3 Fire Zones 2091-BB, 2112-BB and 2183-J (electrical penetration rooms) 4.2.4.2 4.13E-07 1.32E-08 yes yes -1.96E-06 -6.49E-08 8.49E-08/ 2.79E-09 B-4 Fire Zone 2154-E (CEDM equipment room) 4.2.4.2 3.26E-06 3.61E-08 yes yes 2.60E-07 -5.17E-08 3.22E-06/ 3.51E-08 B-5 Fire Zones 2149-B and 2158-F (stairwells 2001 and 2055) 4.2.3.2 4.06E-09 1.21E-10 no n/a n/a n/a n/a Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-16 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF)
B-6 Fire Zones 2006-LL, 2010-LL, 2011-LL, and 2014-LL (general access, C HPSI pump room, tendon gallery access, and A HPSI, LPSI and Containment Spray Pump room)  4.2.4.2 1.56E-06 1.70E-08 yes no 3.00E-08 3.00E-10 n/a CC Fire Zone 2024-JJ (turbine-driven emergency feedwater pump room) 4.2.3.2 1.49E-09 3.85E-11 no n/a n/a n/a n/a DD Fire Zones 2019-JJ, 2032-JJ, 2040-JJ, and 2068-DD (boric acid condensate tank room, spent resin storage tank room, corridor, and hot machine shop) 4.2.4.2 2.45E-06 2.76E-08 yes no -2.85E-06 -9.64E-08 n/a EE-L Fire Zones 2055-JJ and 2084-DD (piping penetration rooms) 4.2.4.2 3.82E-07 7.53E-09 yes yes -7.88E-07 -2.64E-08 8.99E-08/ 3.03E-09 EE-U Fire Zone 2111-T (lower south electrical penetration room) 4.2.4.2 2.03E-06 4.81E-08 yes yes -5.77E-06 -1.94E-07 9.51E-07/ 3.17E-08 FF Fire Zone 2025-JJ (motor- driven emergency feedwater pump room) 4.2.3.2 1.15E-08 3.66E-10 no n/a n/a n/a n/a G Fire Zones 2199-G, 2119-H, 2136-I, 2137-I, 2150-C, 2098-C, 2098-L, 129-F, and 97-R (control room and other alternate shutdown areas) 4.2.4.2 2.37E-05 5.28E-07 yes yes -7.36E-05 -2.65E-06 9.00E-06/ 1.61E-07 GG Fire Zones 2076-HH and 2081-HH (electrical equipment room and upper north and lower north piping penetration room) 4.2.4.2 1.04E-06 2.19E-08 yes yes -1.12E-05 -3.82E-07 4.41E-07/ 1.46E-08 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-17 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF)
HH Fire Zones 2063-DD, 2072-R, 2073-DD, 2096-M, 2106-R, and 2107-N (sample room, VCT room, 2B-62 room, 2B-63 room, degasifier vacuum pump room, and corridor) 4.2.4.2 3.11E-06 4.65E-08 yes no -5.80E-07 -2.18E-08 n/a II Fire Zone 2101-AA (north switchgear 2A-3 room) 4.2.4.2 2.90E-06 9.31E-08 yes no -1.33E-04 -4.52E-06 n/a JJ Fire Zone 2109-U (corridor) 4.2.4.2 2.70E-06 7.97E-08 yes yes -3.78E-06 -1.21E-07 2.37E-07/ 6.78E-09 K Fire Zones 16-Y and 2020-JJ (clean waste receiver tank room and boron holdup tank vault) 4.2.3.2 6.46E-10 1.44E-11 no n/a n/a n/a n/a KK Fire Zones 2093-P, 2114-I, and 2115-I (south EDG room, EDG air intake room, and boric acid makeup tank room) 4.2.4.2 1.72E-07 5.22E-09 yes no  n/a L Fire Zone TKVLT (diesel fuel storage vault) 4.2.3.2 1.86E-08 4.32E-10 no n/a n/a n/a n/a MM Fire Zones 2099-W and 2103-V (west DC equipment room and west battery room) 4.2.4.2 3.28E-06 7.85E-08 yes yes -2.06E-05 -7.94E-07 8.61E-07/ 2.88E-08 NN Fire Zones 2032-K and 2033-K (containment building south side and containment building north side) 4.2.4.2 3.04E-06 7.76E-08 yes no  n/a OO Intake Structure 4.2.4.2 1.62E-07 1.84E-09 yes yes 1.62E-07 1.84E-09 8.42E-08/ 9.87E-10 Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights Enclosure 1 to 2CAN121202 Page W-18 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF)
QQ Fire Zones 2094-Q and 2114-I (north EDG room and EDG air intake room) 4.2.3.2 3.98E-07 1.26E-08 no n/a n/a n/a n/a SS Fire Zones 2097-X, 2100-Z and 2102-Y (east DC equipment room, south switchgear room and east battery room) 4.2.4.2 2.81E-06 7.43E-08 yes yes -3.75E-05 -1.28E-06 2.66E-06/ 7.04E-08 TT Fire Zone 2108-S (electrical equipment room) 4.2.4.2 2.86E-06 4.47E-08 yes yes -2.37E-05 -7.95E-07 5.07E-07/ 1.76E-08 YD YARD 4.2.3.2 6.74E-07 1.28E-08 no n/a n/a n/a n/a Various Unit 1 - specific fire areas 1 4.2.3.2 6.06E-07 1.87E-8 no n/a n/a n/a n/a TOTAL  6.47E-05 1.39E-06
  -4.40E-04 -1.52E-05 2.02E-05/ 4.01E-07  Indicative of an immeasurable change in risk from the impact of the VFDR on Fire PRA model.
1 ANO-1-specific fire areas were conservatively assessed to contribute to ANO-2 CDF/LERF. Fires in these areas typically do not impact circuits for ANO-2 components and are not expected to cause, or require, an ANO-2 plant trip.  (Fires in the ANO-1 contr ol room and cable spreading room are not included in this value since they are included in the Fire Area G results.)
 
References
: 1. Echelon Calculation PRA-A2-05-002, Rev. 0, "ANO-2 Fire PRA New Human Failure Events," October 2011. 2. Echelon Calculation PRA-A2-05-007, Rev. 0, "ANO-2 Fire Probabilistic Risk Assessment Human Reliability Analysis (HRA)
Notebook," February 2012.
 
Entergy Operations, Inc. Arkansas Nuclear One - Unit 2 Enclosure 2 to 2CAN121202 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition
 
Proposed Operating License and Technical Specification Changes (mark-up)
 
December 17, 2012
 
4 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 (b) Fire Protection EOI shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated December 17, 2012, (and supplements dated ___________) and as approved in the safety evaluation report dated ___________. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, as described in Amendment 9A to the Safety Analysis Report and as approved in the Safety Evaluation dated March 31, 1992, subject to the following provision:
T the licensee 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.satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that may be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and sha ll be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
: 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
: 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10
-7/year (yr) for CDF and less than 1x10
-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
 
4 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 (c) Less Than Four Reactor Coolant Pump Operation EOI shall not operate the reactor in operational Modes 1 and 2 with fewer than four reactor coolant pumps in operation, except as allowed by Special Test Exception 3.10.3 of the facility Technical Specifications.
2.C.(3)(d) Deleted per Amendment 24, 6/19/81.
2.C.(3)(e) Arkansas Power & Light (AP&L) 1 shall complete the following modifications by the indicated dates in accordance with the staff's findings as set forth in the fire protection evaluation report, NUREG
-0223 "Fire Protection Safety Evaluation Report."Deleted per Amendment [TBD], [date]. Implementation Dates for Proposed Modifications Applicable Section of NUREG-0223      Date      3.1  Portable Radio Communication Equipment March 31, 1979 3.2  Separation of Power Cables in Manholes
* 3.3  Protection from Water Spray
* 3.4  Protection of Redundant Cables in the MCC Room (2096
-M)  December 30, 1978 3.5  Protection of Redundant Cables in the Hallway - Elevation 372 (2109
-U)  *, **      3.6  Protection of Redundant Cables in the Cable Spreading Room (2098
-L)
* 3.7  Protection of Redundant Cables in the Switchgear Room (2100
-Z)
* 1 AP&L is the predecessor to Entergy Arkansas, Inc.
Moved to Page 6 5
Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 Applicable Section of NUREG-0223      Date      3.8  Protection of Redundant Cables in the Electrical Equipment Room (2091
-BB)  September 30, 1978 3.9  Protection of Redundant Cables in the Lowe r South Piping Penetration Room (2111-T)    September 30, 1978 3.10  Protection of Safe Shutdown Cables in the Upper South Piping Penetration Room (2084-DD)    September 30, 1978 3.11  Protection of Redundant Reactor Protection System Cables (2136-I)  *, **      3.12  Fire Dampers September 30, 1978 3.13  Portable Extinguisher for the Control Room (2199-J)  November 15, 1978 3.14  Smoke Detectors
  *, **      3.15  Manual Hose Stations (2055
-JJ, 2084-DD, Containment, Elev. 317' of Auxiliary Building)
  *, **      3.16  Portable Smoke Exhaust Equipment December 1, 1978 3.17  Emergency Lighting December 1, 1978 3.18  Reactor Coolant Pump Oil Collection System
* 3.19  Control of Fire Doors March 31, 1979      3.20  Administrative Control Changes December 1, 1978 (Numbers in parentheses refer to fire zone designations in the AP&L fire hazards analysis.)
* Prior to startup following the first regularly scheduled refueling outage.
  ** Technical Specifications covering these items should be proposed not later than 90 days prior to implementation.
 
5 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 Other Changes that may be Made Without Prior NRC Approval
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program  Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard."  Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:  Fire Alarm and Detection Systems (Section 3.8);
Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
Gaseous Fire Suppression Systems (Section 3.10); and,  Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated ___________ to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
 
6 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 Transition License Conditions
: 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2. The licensee shall implement the modifications described in the December 17, 2012, submittal of the ANO-2 NFPA 805 Transition Report, Table S-1, "Plant Modifications Committed," to complete the transition to full compliance with 10 CFR 50.48(c) prior to startup from the second ANO-2 refueling outage following SER issuance.
: 3. The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
  (c) Less Than Four Reactor Coolant Pump Operation EOI shall not operate the reactor in operational Modes 1 and 2 with fewer than four reactor coolant pumps in operation, except as allowed by Special Test Exception 3.10.3 of the facility Technical Specifications.
2.C.(3)(d) Deleted per Amendment 24, 6/19/81.
2.C.(3)(e) Arkansas Power & Light (AP&L) 1 shall complete the following modifications by the indicated dates in accordance with the staff's findings as set forth in the fire protection evaluation report, NUREG
-0223 "Fire Protection Safety Evaluation Report."Deleted per Amendment [TBD], [date]. 2.C.(3)(f) Deleted per Amendment 24, 6/19/81.
2.C.(3)(g) Deleted per Amendment 93, 4/25/89.
2.C.(3)(h) Deleted per Amendment 29, (3/4/82) and its correction letter, (3/15/82).
(i) Containment Radiation Monitor AP&L shall, prior to July 31, 1980 submit for Commission review and approval documentation which establishes the adequacy of the qualifications of the containment radiation monitors located inside the containment and shall complete the installation and testing of these instruments to demonstrate that they meet the operability requirements of Technical Specification No. 3.3.3.6.
2.C.(3)(j) Deleted per Amendment 7, 12/1/78.
2.C.(3)(k) Deleted per Amendment 12, 6/12/79 and Amendment 31, 5/12/82.
From Page 4 6
Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 2.C.(3)(l) Deleted per Amendment 24, 6/19/81.
2.C.(3)(m) Deleted per Amendment 12, 6/12/79.
2.C.(3)(n) Deleted per Amendment 7, 12/1/78.
2.C.(3)(o) Deleted per Amendment 7, 12/1/78.
2.C.(3)(p) Deleted per Amendment 255, 9/28/04.
2.C.(4) (Number has never been used.)
2.C.(5) Deleted per Amendment 255, 9/28/04.
2.C.(6) Deleted per Amendment 255, 9/28/04.
2.C.(7) Deleted per Amendment 78, 7/22/86.
(8) Antitrust Conditions EOI shall not market or broker power or energy from Arkansas Nuclear One, Unit 2. Entergy Arkansas, Inc. is responsible and accountable for the actions of its agents to the extent said agent's actions affect the marketing or brokering of power or energy from ANO, Unit 2.
(9) Rod Average Fuel Burnup Entergy Operations is authorized to operate the facility with an individual rod average fuel burnup (burnup averaged over the length of a fuel rod) not to exceed 60 megawatt-days/kilogram of uranium.
 
Move to Page 7 7
Renewed License No. NPF-6 Amendment No. 288 , 2.C.(3)(o) Deleted per Amendment 7, 12/1/78.
2.C.(3)(p) Deleted per Amendment 255, 9/28/04.
2.C.(4) (Number has never been used.)
2.C.(5) Deleted per Amendment 255, 9/28/04.
2.C.(6) Deleted per Amendment 255, 9/28/04.
2.C.(7) Deleted per Amendment 78, 7/22/86.
(8) Antitrust Conditions EOI shall not market or broker power or energy from Arkansas Nuclear One, Unit 2. Entergy Arkansas, Inc. is responsible and accountable for the actions of its agents to the extent said agent's actions affect the marketing or brokering of power or energy from ANO, Unit 2.
(9) Rod Average Fuel Burnup Entergy Operations is authorized to operate the facility with an individual rod average fuel burnup (burnup averaged over the length of a fuel rod) not to exceed 60 megawatt-days/kilogram of uranium.
(10) Mitigation Strategies The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:
(i) Fire fighting response strategy with the following elements:
: 1. Pre-defined coordinated fire response strategy and guidance 2. Assessment of mutual aid fire fighting assets 3. Designated staging areas for equipment and materials 4. Command and control
: 5. Training of response personnel (ii) Operations to mitigate fuel damage considering the following:
: 1. Protection and use of personnel assets 2. Communications 3. Minimizing fire spread
: 4. Procedures for implementing integrated fire response strategy
: 5. Identification of readily-available pre-staged equipment
: 6. Training on integrated fire response strategy
: 7. Spent fuel pool mitigation measures (iii) Actions to minimize release to include consideration of:
: 1. Water spray scrubbing 2. Dose to onsite responders From Page 6 7
Renewed License No. NPF-6 Amendment No. 288 ,  (11) Upon implementation of Amendment 288 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 4.7.6.1.2.d, in accordance with Specifications 6.5.12.c.(i),
6.5.12.c.(ii), and 6.5.12.d, shall be considered met. Following implementation:
  (i) The first performance of SR 4.7.6.1.2.d, in accordance with Specification 6.5.12.c.(i), shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
(ii) The first performance of the periodic assessment of CRE habitability, Specification 6.5.12.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
(iii) The first performance of the periodic measurement of CRE pressure, Specification 6.5.12.d, shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
 
Moved to Page 8 8
Renewed License No. NPF-6 Amendment No. 288
,294 ,295 , (11) Upon implementation of Amendment 288 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 4.7.6.1.2.d, in accordance with Specifications 6.5.12.c.(i),
6.5.12.c.(ii), and 6.5.12.d, shall be considered met. Following implementation:
  (i) The first performance of SR 4.7.6.1.2.d, in accordance with Specification 6.5.12.c.(i), shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
(ii) The first performance of the periodic assessment of CRE habitability, Specification 6.5.12.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
(iii) The first performance of the periodic measurement of CRE pressure, Specification 6.5.12.d, shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
D. Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security, Safeguards Contingency and Training & Qualification Plan," as submitted on May 4, 2006.
EOI shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the
 
authority of 10 CFR 50.90 and 10 CFR 50.54(p). The EOI CSP was approved by License Amendment No. 294 as supplemented by a change approved by License Amendment No. 295 E. This renewed license is subject to the following additional condition for the protection of the environment:
Before engaging in additional construction or operational activities which may result in an environmental impact that was not evaluated by the Commission, EOI will prepare and record an environmental evaluation for such activity. When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater than that evaluated, in the Final Environmental Statement (NUREG-0254) or any addendum thereto, and other NRC environmental impact assessments, EOI shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation.
From Page 7 8
Renewed License No. NPF-6 Amendment No. 288
,294 ,295 , F. Updated Final Safety Analysis Report Supplement The Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license. Until that update is complete, ANO-2 may make changes to the programs and activities described in the supplement without prior Commission approval, provided that ANO-2 evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements of that section.
The ANO-2 Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. ANO-2 shall complete these activities no later than July 17, 2018, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
Move to Page 9 9
Renewed License No. NPF-6 Amendment No. 288
,294 ,295 , F. Updated Final Safety Analysis Report Supplement The Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license. Until that update is complete, ANO-2 may make changes to the programs and activities described in the supplement without prior Commission approval, provided that ANO-2 evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements of that section.
The ANO-2 Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. ANO-2 shall complete these activities no later than July 17, 2018, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
G. Reactor Vessel Material Surveillance Capsules All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.
: 4. This renewed license is effective as of the date of issuance and shall expire at midnight, July 17, 2038.
FOR THE NUCLEAR REGULATORY COMMISSION
 
Original signed by J. E. Dyer
 
J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:
: 1. Appendix A - Technical Specifications
: 2. Preoperational Tests, Startup Tests and other items which must be completed by the indicated Operational Mode Date of Issuance: June 30, 2005
 
From Page 8 ARKANSAS - UNIT 2 6-3 Amendment No. 255 , ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATIONS
 
6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI ANS 3.1-1978 for comparable positions, except for the designated radiation protection manager, who shall meet or exceed the minimum qualifications of
 
Regulatory Guide 1.8, September 1975.
 
===6.4 PROCEDURES===
6.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:
: a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;
: b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Section 7.1 of
 
Generic Letter 82-33;
: c. Fire Protection Program implementation; Deleted  d. All programs specified in Specification 6.5; and
: e. Modification of core protection calculator (CPC) addressable constants. These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with the CPCs during reactor operation.
Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P, which has been determined to be applicable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.
 
Entergy Operations, Inc. Arkansas Nuclear One - Unit 2 Enclosure 3 to 2CAN121202
 
Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition
 
Revised Operating License and Technical Specification Pages
 
December 17, 2012
 
4 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 (b) Fire Protection EOI shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated December 17, 2012, (and supplements dated_________________ and
_________________) and as approved in the safety evaluation report dated _________________. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that may be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and sha ll be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
: 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
: 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10
-7/year (yr) for CDF and less than 1x10
-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
 
5 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 Other Changes that may be Made Without Prior NRC Approval
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program    Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard."  Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:  Fire Alarm and Detection Systems (Section 3.8);
Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
Gaseous Fire Suppression Systems (Section 3.10); and,  Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated _________________ to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
 
6 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 Transition License Conditions
: 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2. The licensee shall implement the modifications described in the December 17, 2012, submittal of the ANO-2 NFPA 805 Transition Report, Table S-1, "Plant Modifications Committed," to complete the transition to full compliance with 10 CFR 50.48(c) prior to startup from the second ANO-2 refueling outage following SER issuance.
: 3. The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
  (c) Less Than Four Reactor Coolant Pump Operation EOI shall not operate the reactor in operational Modes 1 and 2 with fewer than four reactor coolant pumps in operation, except as allowed by Special Test Exception 3.10.3 of the facility Technical Specifications.
2.C.(3)(d) Deleted per Amendment 24, 6/19/81.
2.C.(3)(e) Deleted per Amendment [TBD], [date].
2.C.(3)(f) Deleted per Amendment 24, 6/19/81.
2.C.(3)(g) Deleted per Amendment 93, 4/25/89.
2.C.(3)(h) Deleted per Amendment 29, (3/4/82) and its correction letter, (3/15/82).
(i) Containment Radiation Monitor AP&L shall, prior to July 31, 1980 submit for Commission review and approval documentation which establishes the adequacy of the qualifications of the containment radiation monitors located inside the containment and shall complete the installation and testing of these instruments to demonstrate that they meet the operability requirements of Technical Specification No. 3.3.3.6.
2.C.(3)(j) Deleted per Amendment 7, 12/1/78.
2.C.(3)(k) Deleted per Amendment 12, 6/12/79 and Amendment 31, 5/12/82.
2.C.(3)(l) Deleted per Amendment 24, 6/19/81.
2.C.(3)(m) Deleted per Amendment 12, 6/12/79.
2.C.(3)(n) Deleted per Amendment 7, 12/1/78.
7 Renewed License No. NPF-6 Amendment No.
Revised by letter dated July 18, 2007 2.C.(3)(o) Deleted per Amendment 7, 12/1/78.
2.C.(3)(p) Deleted per Amendment 255, 9/28/04.
2.C.(4) (Number has never been used.)
2.C.(5) Deleted per Amendment 255, 9/28/04.
2.C.(6) Deleted per Amendment 255, 9/28/04.
2.C.(7) Deleted per Amendment 78, 7/22/86.
(8) Antitrust Conditions EOI shall not market or broker power or energy from Arkansas Nuclear One, Unit 2. Entergy Arkansas, Inc. is responsible and accountable for the actions of its agents to the extent said agent's actions affect the marketing or brokering of power or energy from ANO, Unit 2.
(9) Rod Average Fuel Burnup Entergy Operations is authorized to operate the facility with an individual rod average fuel burnup (burnup averaged over the length of a fuel rod) not to exceed 60 megawatt-days/kilogram of uranium.
(10) Mitigation Strategies The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:
(i) Fire fighting response strategy with the following elements:
: 1. Pre-defined coordinated fire response strategy and guidance 2. Assessment of mutual aid fire fighting assets 3. Designated staging areas for equipment and materials
: 4. Command and control
: 5. Training of response personnel (ii) Operations to mitigate fuel damage considering the following:
: 1. Protection and use of personnel assets 2. Communications 3. Minimizing fire spread 4. Procedures for implementing integrated fire response strategy 5. Identification of readily-available pre-staged equipment
: 6. Training on integrated fire response strategy
: 7. Spent fuel pool mitigation measures (iii) Actions to minimize release to include consideration of:
: 1. Water spray scrubbing 2. Dose to onsite responders 8
Renewed License No. NPF-6 Amendment No. 288
,294 ,295 , (11) Upon implementation of Amendment 288 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 4.7.6.1.2.d, in accordance with Specifications 6.5.12.c.(i),
6.5.12.c.(ii), and 6.5.12.d, shall be considered met. Following implementation:
  (i) The first performance of SR 4.7.6.1.2.d, in accordance with Specification 6.5.12.c.(i), shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
(ii) The first performance of the periodic assessment of CRE habitability, Specification 6.5.12.c.(ii), shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
(iii) The first performance of the periodic measurement of CRE pressure, Specification 6.5.12.d, shall be within 15 months of the approval of TSTF-448. SR 4.0.2 will not be applicable to this first performance.
D. Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security, Safeguards Contingency and Training & Qualification Plan," as submitted on May 4, 2006.
EOI shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the
 
authority of 10 CFR 50.90 and 10 CFR 50.54(p). The EOI CSP was approved by License Amendment No. 294 as supplemented by a change approved by License Amendment No. 295.
E. This renewed license is subject to the following additional condition for the protection of the environment:
Before engaging in additional construction or operational activities which may result in an environmental impact that was not evaluated by the Commission, EOI will prepare and record an environmental evaluation for such activity. When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater than that evaluated, in the Final Environmental Statement (NUREG-0254) or any addendum thereto, and other NRC environmental impact assessments, EOI shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation.
 
9 Renewed License No. NPF-6 Amendment No. 288
,294 , F. Updated Final Safety Analysis Report Supplement The Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license. Until that update is complete, ANO-2 may make changes to the programs and activities described in the supplement without prior Commission approval, provided that ANO-2 evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements of that section.
The ANO-2 Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. ANO-2 shall complete these activities no later than July 17, 2018, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
G. Reactor Vessel Material Surveillance Capsules All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.
: 4. This renewed license is effective as of the date of issuance and shall expire at midnight, July 17, 2038.
FOR THE NUCLEAR REGULATORY COMMISSION
 
Original signed by J. E. Dyer
 
J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:
: 1. Appendix A - Technical Specifications
: 2. Preoperational Tests, Startup Tests and other items which must be completed by the indicated Operational Mode Date of Issuance: June 30, 2005
 
Entergy Operations, Inc. Arkansas Nuclear One - Unit 2 Enclosure 4 to 2CAN121202
 
Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition
 
List of Regulatory Commitments
 
December 17, 2012 to 2CAN121202
 
Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
 
TYPE (Check one) COMMITMENT ONE-TIME ACTION CONTINUING COMPLIANCE SCHEDULED COMPLETION DATE Entergy will complete implementation of the modifications identified in Table S-1
 
of Attachment S  Prior to startup from the second ANO-2 refueling outage following SER issuance Entergy will complete implementation of procedure changes, process updates, and training of affected plant personnel identified in Table S-2 of Attachment S  Within six months following SER issuance}}

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