Text

Evaluation of Potential Severe Accidents During LOW Power and Shutdown Operations at Surry,Unit 1.Analysis of Core Damage Frequency from Internal Fires During Mid-Loop Operations.Main Report
	ML18151A227
Person / Time
Site:	Surry
Issue date:	07/31/1994
From:	Chu T, Ho V, Hou Y, Lin J, Musicki Z, Nathan Siu, Yang J; BROOKHAVEN NATIONAL LABORATORY, MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE, PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.)
To:	; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
	CON-FIN-L-1922 BNL-NUREG-52399, NUREG-CR-6144, NUREG-CR-6144-V03-P1, NUREG-CR-6144-V3-P1, NUDOCS 9408180144
	Download: ML18151A227 (231)
	v • d • e

{{#Wiki_filter:** NUREG/CR-6144 BNL-NUREG-52399 Vol. 3, Part 1 Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1 .Malysis of Core Damage Frequency from

ernal Fires During Mid-Loop* Operations Main Report Prepared by Z. Musicki, T. L. Chu, V. Ho, Y.-M. Hou, J. Lin, J. Yang, N. Siu Brookhaven National Laboratory Prepared for U.S. Nuclear Regulatory Commission 9408180144 940731 '

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NUREG/CR-6144 BNL-NUREG-52399 Vol. 3, Part 1 Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1 Analysis of Core Damage Frequency from Internal Fires During Mid-Loop Operations ain Report Manuscript Completed: January 1994 Date Published: July 1994. Prepared by

Z. Musicki, T. L. Chu, V. Ho1, Y.-M. Hou1, J. Lin1, J. Yang, N. Siu2 Brookhaven National Laboratory Upton, NY 11973 Prepared for Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC FIN L1922 Inc., 4590 MacArthur Boulevard, Newport Beach, CA 92660-2027 IT, Cambridge, MA, currently at EG&G, Idaho Falls, ID 83415

ABSTRACT Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analysis that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program were to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a Level-3 PRA. A phased approach was used in the Level-1 program. In Phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the Phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed Phase 2 analysis.

. In Phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the Phase 1 study. The objective of the Phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the Level-I study includes plant damage state analysis and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associated, Inc.. Volume 6 documents the accident progression, source terms, and consequence analysis.

In the Phase 2 study, system models applicable for shutdown conditions were developed and supporting thermal hydraulic analyses were performed to determine both the timing of the accidents and success criteria for systems. Initiating events that may occur during mid-loop operations were identified and accident sequence event trees were developed and quantified. In the preliminary quantification of the mid-loop accident sequences, it was found that the decay heat at which the accident initiating event occurs is an important parameter that determines both the success criteria for the mitigating functions and time available for operator actions. In order to better account for the decay heat, a "time window" approach was developed. In this approach, time windows after shutdown were defined based on the success criteria established for the various methods that can be used to mitigate the accident. Within each time window, the decay heat and accident sequence timing are more accurately defined and new event trees developed and quantified accordingly. Statistical analysis of the past outage data was performed to determine the time at which a mid-loop condition is reached, and the duration of the mid-loop operation. Past outage data were used to determine the probability that an accident initiating event occurs in each of the time windows. This probability is used in the quantification of the accident sequences. The major objective of the Surry internal fire analysis was to provide an improved understanding of the risk arising from internal fire-related events. The mean core damage frequency of the Surry plant due to internal fire events during mid-loop operations is 2E-05 per year, and the 5th and 95th percentiles are lE-06 and 8E-05 , per year, respectively. This can be compared with the mean core frequency of lE-05 per year estimated in the NUREG-1150 study for accidents initiated by fires during full power operations. iii

TABLE OF CONTENTS Abstract . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii Table of Contents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix List of Tables .................. : ....... , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x Executive Summary ............................................................... xiii Foreword . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xix Acknowledgements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxi

1. Introduction and Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Scope of the Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 Results of Previous Fire Risk Analyses for Surry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3 Methodology Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.4 Assumptions and Simplifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.5 Internal PRA Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.6 Plant Visits and Interaction with Virginia Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.7 Organization of the Report................................. . . . . . . . . . . . . . . 1-3 1.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

2. Methodology Overview ........................................................ 2-1 2.1 Fire Frequency Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2 Scenario Development ................................. ." ................ 2-1 2.3 Fire Growth Modeling .................................................. 2-1 2.4 Fire Suppression and Damage Fraction Calculation..................... . . . . . . . . 2-1 2.5 Fault Tree and Event Tree Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.6 Model Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3

3. Important Fire Areas Analyzed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 Unit 1 Emergency Switchgear Room (ESGR) ................................. 3-1 3.2 Cable Vault and Tunnel (CVT) ............................................ 3-2 3.3 Containment (CI) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.4 Normal Switchgear Room (NSGR) (Unit 1) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.5 Fire Areas Not Analyzed .............................. *. . . . . . . . . . . . . . . . . . 3-3

4. Fire Frequency Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 Methodology ......... *. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2 Sandia Fire Data Base . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.3 Criteria for Inclusion/Exclusion of Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.4 Exposure Time at Power and Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.5 Pertinent Events from Data Base . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.6 Bayesian Updating Scheme . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.7 Component Based Fire Frequency ......................................... 4-3 4.8 Insights . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5

5. Scenario Development 5-1 V NUREG/CR-6144

Table of Contents (continued) 5.1 Assumptions and Simplifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Initiators Considered . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 53 Important Equipment and Cable Locations .. , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.4 Types of Fire Damage to Components .......................*...........*.. 5-3 5.5 Important Locations and Scenarios Within Fire Zones . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 5.5.1 ESGR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 5.5.2 cvr . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 5.53 CT . . . . . * . * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 5.5.4 NSGR .....*.......................... ~ . . . . . . . . . . . . . . . . . . . . . . 5-8 5.6 Scenario Quantification ............................ ~ . . . . . . . . . . * . . . . . . . . . . 5-8 5.7 References .........*................................................. 5-8

6. Fire Growth Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . 6-1 6.1 COMPBRN Ille Capabilities and Limitations ................................ . 6-1 6.2 Modeling of Fires in Important Fire Areas ..................*................ 6-2 6.2.1 ESGR ...................................................... . 6-2 6.2.2 cvr ....................................................... . 6-3 6.3 Fire Analysis . . . . . . . . . . . * . . . . . . . . . . . . . . . . . . ... *. . . . . . . . . . . . . . . . . . . . . . . .. 6-3 6.3.1 Oil Fire ......................................*., ............. . 6-3 6.3.2 Cable Fires .................................................. . 6-4 6.3.3 Cabinet Fire ...*............................................... 6-5 6.3.4 Trash Fire .*.................................................. 6-5 6.4 Sensitivity Study ........*....................................... *...... . 6-5 6.4.1 Pool Fire Size * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * . . . . . . . . . . . . . . ... . 6-5 6.4.2 Effect of Wall on Plume Entrainment .............................. . 6-6

6.4.3 Forced Ventilation . . . . . . . . . . . . . . . . . . . . . . . ..... *. . . . . . . . . . . . . . . . .. 6-6 6.4.4 Forced Ventilation and Dooiway Opening ..........*................. 6-6 6.5 References ~ . . . . . . . . . . . . . . . . . . . ....... , . . . . . . . . . . . . . * . . . . . . . . . . . . . . .. 6-7

7. Fire Suppression Modeling . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * . . . . . . 7-1 7.1 Transition Diagram of Fire Detection and Suppression ......................... . 7-1 7.2 Calculation of Parameters for fus ..***.***.**...****..*******.*..**.*.... : .* 7-2 7.3 References .*......................................................... 7-2

8. Damage Fraction Calculations .................................................... . 8-1 8.1 ESGR ............................................................. . 8-1 8.2 cvr ............................................................... . 8-1 8.3 CT ................................................................ . 8-1 8.4 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . 8-2

9. Control Room Fire Risk Analysis ................................................. 9-1 9.1 Fire Scenario Identification. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-2 9.1.1 Shutdown Accident Mitigation Equipment and Control Locations . . . . . . . . . . 9-3 9.1.2 Significant Scenario Identifica,tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4 9.2 Fire Scenario Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-9*

9.2.1 Control Room Walkdown ........................................ 9-10 NUREG/CR-6144 vi

Table of Contents (continued) 9.2.2 Industrial Fire Events Specialization ................................. 9-11 9.2.3 Control Room Fire Frequency Assessment ............................ 9-13 9.2.4 Control Room Fire Frequency Apportionment ......................... 9-14 9.2.5 Scenario Frequency Quantification .................................. 9-14 9.3 References ........................................................... 9-17

10. Modification of the Internal PRA Model ........................................... 10-1 10.1 References ........................................................... 10-1

11. Quantification ............................................................... 11-1 11.1 Summary of Procedure ................................................. 11-1 11.2 Quantification of HEP for High-Level Human Action Events .................... 11-1 11.2.1 High Level HEP Quantification ................................... 11-1 11.2.2 Recovery Actions .............................................. 11-1 11.3 Successful System Correction ............ _. ............................... 11-1 11.4 Uncertainty Analysis ................................................... 11-2 11.5 References .......................................................... 11-3

12. Results .................................................................... 12-1 12.1 Point Estimate Results ................................................. 12-1 12.2 Uncertainty Results .................................................... 12-2 12.3 Conclusions ......................................................... 12-3

13. Plant Damage State Analysis .................................................... 13-1 13.1 Definition of Plant Damage State Indicators ............................... 13-1 13.2 PDS Analysis, Rules ................................................. 13-1 13.3 PDS Uncertainty Analysis ............................................. 13-1 APPENDIX A-List of Plants and Pertinent Times APPENDIX B-Pertinent Events from the Fire Data Base Bl Generic Events B2 Surry Events APPENDIX C-Derivationof Phi, The Trash Combustible Fraction APPENDIX D-Cable Tray Trace Information APPENDIX E-Derivation of the Hot Short Spurious Operation Probability APPENDIX F-Fire Suppression Model APPENDIX G-High Level Human Errors Tables APPENDIX I-Event Trees Used in Fire Analysis APPENDIX H-Not Used APPENDIX J-Fault Trees Modified for Fire Events APPENDIX K-Input for Scenario Initiating Event Uncertainty Analysis.

Kl Algebraic Expressions for Fire Scenario Initiating Events K2 Uncertainty Data For Variables in Appendix K.1 K3 Description of Variables APPEND IX L-Fire Scenario Initiating Event Histograms APPEND IX M-Sequence Quantification vii NUREG/CR-6144

Table of Contents (continued) APPENDIX N-Basic Event Reports N.1 Basic Event Uncertainty Report N.2 Basic Event Descriptions APPEND IX 0-The Most Important Cutsets for the Total Core Damage Frequency Due to Fire at Midloop APPENDIX P-Basic Event Importance Reports P.1 Basic Event Importance Sorted by Fussell-Vesely Importance P.2 Basic Event Importance Sorted Alphabetically APPEND IX Q-Rules for Recovery Actions and Successful System Correction NUREG/CR-6144

viii
S.1 LIST OF FIGURES Fire Area Contribution to CDF ...............*................................ xv,*

3.1 Unit 1 Emergency Switchgear Room Fire Doors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-4 3.2 Unit 1 Cable Vault and Tunnel Fire Doors ...................................... 3-5 33 Emergency Switchgear Room Equipment Layout and Cable Tray Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6 3.4 Plan View of the Cable Vault and Tunnel with the Cross Section of the Tunnei Showing Approximate Cable Tray Locations and the MCCs . . . . . . . . . . . . . . . . . . . . 3-7 3.5 Containment Layout at 18'4" Showing Critical Tray Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-8 3.5a Containment Layout at RHR Flats Showing Critical Tray Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-9 3.6 Unit 1 and Unit 2 Normal Switchgear Rooms .................................... 3-10 3.7 Surry Station 4 kV Electrical Distribution System ................................. 3-11 5.1 Emergency Switchgear Room Equipment Layout and Cable Tray Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.2 Plan View of the Cable Vault and Tunnel with the Cross Section of the Tunnei Showing Approximate Cable Tray Locations and the MCCs .................... 5-10 Containment Layout at 18'4" Showing Critical Tray Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5-11 5.3a Containment Layout at RHR Flats Showing Critical Tray Locations ..............*......................... 5-12 7.1 Fire Supression Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 9.1 Surry Unit 1 Main Control Room Equipment Arrangement .......................... 9-19 9.2 F,(r) for Control Room Fires ................................................ 9-20 93 illustration of Five Sev:erity Factor Calculations for Scenario BB-1-1 ................... 9-21 12.1 Fire Area Contribution to CDF .........*................................... ' .. 12-4 12.2 Contn"bution of Windows to Fire CDF ......................... ~ ............... 12-5 12.3 POS Contribution to Fire CDF ....*.......................................... 12-6 12.4 Significant Scenario, POS, Window Combinations Contributors to Fire CDF ............. 12-7 12.5 Scenario Contribution to CDF ................................................ 12-8 ix NUREG/CR-6144

LIST OF TABLES

S.1 Summary of Point Estimate CDFs for Fire Events ................................. xvii S.2 Results of the Fire Uncertainty Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xviii 1.1 Shutdown Fire Frequencies [6] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.2 Results of Coarse Screening Fire Analysis for POS 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.3 Surry Fire Initiating Event Frequencies for Full Power Operation (/yr) . . . . . . . . . . . . . . . . . 1-6 1.4 Surry Fire Area Full Power Core Damage Frequency............................... 1-7 1.5 Summary of Results-Core Damage Frequency by Initiating Event and Plant Operational States . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 3.1 Surry Power Station - Units 1 and 2 Identification of Fire Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3-12 3.2 Combustible Loading Summary+ .............................................. 3-13 3.3A Surry Power Station - Unit 1 Existing Fire Protection Features by Important Fire Area ..................... 3-14 3.3B Miscellaneous Passive Fire Protection Items - Appendix R ........................... 3-15 3.4 Room Volume and Ventilation Rate for Select Rooms at Surry Summary of HVAC Information ........................................ 3-16 3.5 Important Loads 4160 VAC Emergency Bus lH, 11 Cabinets ................................ 3-17 3.6 Important'Loads 480 VAC Emergency Bus lH, 10 Cabinets ................................ 3-18 3.7 Important Loads 480 VAC Emergency Bus lHl, 7 Cabinets ................................ 3-19 3.8 Important Loads 4160 VAC Emergency Bus lJ, 11 Cabinets ................................ 3-20 3.9 Important Loads 480 VAC Emergency Bus lJ, 10 Cabinets ................................. 3-21 3.10 Important Loads 480 VAC Emergency Bus lJl, 7 Cabinets ................................. 3-22 3.11 Important Loads MCC lHl-1, 40 Cabinets ............................................. 3-23 3.12 Important Loads MCC lJl-1, 42 Cabinets .............................................. 3-24 3.13 Important Loads MCC lHl-2 ....................................................... 3-25 3.14 Important Loads MCC lJl-2 ........................................................ 3-26 4-.

4.1 Exposure Time, All U.S. Plants to 12/89 ........................................ 4-6 4.2 Exposure Time, Surry Units .................................................. 4-7 4.3 Import-=(;:::=:~~~*~'.::*i~~'.".. ~ _s_".'~ ~~-~~*-*.~'.". ....... NUREG/CR-6144 - X

4.3A 4.4 Important Cable and Transfont Fire Events, Sandia Fire Data Base Cable Fires ........................................................ 4-10 Pertinent Surry Fire Events .................................................. 4-11 List of Tables (continued) 45 Parameters of Generic Fire Frequency Distribution (/yr) and Its Lognormal Fit from the Bayesian Updating and for Various Fire Categories and Evidence for Surry Events in that Category .. 4-12 4.5A Parameters of Generic Fire Frequency Distn"bution (/yr) and its Lognormal Fit from the Bayesian Updating and for Transients in Various Fire Areas+ and Evidence for Surry Events in that Category ................................ 4-13 4.6 Parameters of Surry-Specific Fire Frequency Distn"bution (/yr) from Bayesian Updating for Various Categories+ ........................... 4-14 4.7 Point Values for Plant-Wide Fire Frequency (/sdy) ................................ 4-15 4.8 Component* Based Frequency Derivation Point Estimates, Fire Frequency (/sdy) ................................... 4-16 4.9 Cable Fire Frequency Derivation per Tray-Foot Frequency (/sdy/ft) ...*...........................* ................... 4-17 4.10 Transient Fire Frequency Derivation per Square Foot .............................. 4-18 4.11 Cable Tray Length in Fire Areas Unit 1 *........................................ _4-19 4.11A Cable Tray Length in Fire Areas Unit 2 *........................................ 4-20 4.12 Oass B Fire Frequency ..................................................... 4-21 Room (Floor Elevation) .................................................... 5-13 Cable Insulation 'Iypes Used at Surry Unit 1 for Select Systems Cable Types+ ...................................................... 5-18 5.3 Fire Scenario Quantification ................................................. 5-19 5.4 Initiating Event Frequencies (/yr.) for the Scenarios vs.Pos-Windows Combinations ........ 5-20 6.1 Physical Property Parameters . . . . . . . . . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.2 Model Parameters . . . . . . . * * * . . * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 6.3 Cable Damage Time for Oil Fires in .ESGR Room J ............................... 6-10 6.4 Probabilistic Distributions Assumed for Uncertainty Analysis ......................... 6-11 6.5 Results of Uncertainty Analysis ............................................... 6-12 6.6 Cable Damage and Ignition Time for Oil Fire in Cable Vault Tunnel (fime in Minutes) ..' .. 6-13 6.7 Cable Damage and Ignition Time for Cable Fires in the Vault/funnel Region (rime in Minutes) *........................................................ 6-15 6.8 Effect of Oil Pool Size on Combustion Characteristics .............................. 6-16 6.9 Effect of 9il Pool Size on Cable Damage Time ................................... 6-23 7.1 Classification of Fire Areas into Suppression Categories of Table 1.8, Appendix F . . . . . . . . . 7-4 7.2 Parameters and for Fire Areas .......................... *...................... 7-4 8.1 Nonsuppression Fraction, f,,., vs. Time in ESGR . . . . . . . . . . .. ...... ... . .. . . . . . . .... 8-3 8.2 Nonsuppression Fraction, f,,., vs. Time in CVT ........ : . . . .. ...... ... . .. . . . . . . .... 8-3 8.3 Non.suppression Fraction, f,,., vs. Time in CT . . . . . . . . . . . . . .. ...... ... . .. . . . . . . .... 8-3 8.4 Damage Fractions for Fire Scenarios . . . . . . . . . . . . . . . . . . . .. . .. . . . . ... . .. . . . . . . .... 8-3 xi NUREG/CR-6144

List of Tables (continued) 9.1 9.2 List of Surry Unit 1 Shutdown PRA Basic Events for Equipment with Controls in Main Control Room . .- ...................................... 9-22 Control Locations for Surry Unit 1 Shutdown Accident Mitigation Equipment ........................................................ 9-25 9.3 Scenario Description and Preliminary Screening Results *............................ 9-28 9.4 Summary of Control Room Walkdown Observations and Frequency Apportionment for Benchboards 1-1 and 1-2, and Vertical Boards-1 and 1-2 ...... 9-31 9.5 Empirical Database Specializ.ation ............................................. 9-32 9.6 Summary of Control Room Age Assessment ..................................... 9-36 9.7 Main Characteristics of the Noninformative Prior ................................. 9-41 9.8 Main Characteristics of the Posterior ........................................... 9-42 9.9 Control Room Cabinet Fire Frequency Apportionment ..*.......................... 9-43 9.10 Fire Frequency Apportionment of Panel Subsections in Benchboard Boards 1-1 and 1-2 ........................................ 9-44 9.11 Fire Frequency Apportionment of Panel Subsections in Vertical Boards 1-1 and 1-2 ............................................ 9-45 9.12 Summary of Scenario Quantification ........................................... 9-46 12.1 Quantification of Sequences Greater than 10-7/yr .................................. 12-9 12.2 Total Core Damage Frequency (/yr) ........................................... 12-11 12.3 Total CDF for Each POS and Scenario ...........*............................ 12-13 12.4 CDF for Window 1 ...................**... -..*.............................. 12-17 12.5 Conditional Core Damage Probability vs. Scenario ......*......................... 12-19 12.6 Conditional Core Damage Probability vs. Scenario and POS .................... ; . . . . 12-21 12.7 Conditional Core Damage Probability for Window 1 .............................. 12-25 12.8 Results of Uncertainty Analysis for Total Core Damage Frequency (1/yr.) .............. 12-27 12.9 Quantiles of Uncertainty Distribution for Total Core Damage Frequency (/yr) ........... 12-28 13.1 Plant Damage State Definitions ...............*...*.......*................... 13-2 13.2 Plant Damage State Point Estimates and Uncertainty Parameters ...................... 13-3 NUREG/CR-6144 xii

EXECUTIVE

SUMMARY

S.1 Background and Objectives This document presents the results of a Level 1 fire risk assessment of the Surry Power Station Unit 1 for accidents initiated during mid-loop operations. It is a part of the program to assess the risk of a Pressurized Water Reactor (PWR) during low power and shutdown operations. This program was initiated in support of the Nuclear Regulatory Commission (NRC) response to the Chernobyl accident, and was later modified by the staffs follow-up actions to the March 20, 1990 Vogtle incident. The work was performed by Brookhaven National Laboratory (BNL) for the U.S. NRC Office of Nuclear Regulatory Research (RES). A phased approach was taken in this program. In Phase 1, a broadly scoped screening analysis, which included internal fires and flooding, was completed in November, 1991. This screening analysis produced a preliminary Level 1 Probabilistic Risk Assessment (PRA) for accidents initiated during low power and shutdown. It also provided insights on potential accident scenarios and potentially vulnerable configurations during low power and shutdown conditions. Phase 2 of the study focused on a detailed analysis of mid-loop operation. Mid-loop operation was selected because many incidents have occurred during mid-loop operations throughout the world, and recent studies, including Phase 1 of this program, found that the core damage frequency during mid-loop operation is comparable to that of power operation. This report documents the results and findings of the Phase 2 fire risk assessment. The work on internal events, internal flooding, seismic analysis, and Level 2/3 analysis are reported in separate volumes. urry Unit 1 was chosen for this study in part because the Surry plant was previously analyzed in the Reactor Safety dy and NUREG-1150 and in part because Virginia Power offered their cooperation. The core damage frequency ring low power and shutdown calculated in this study is compared with the core damage frequency calculated in NUREG-1150 for accidents during full power. The Surry plant contains two Pressurized Water Reactors (PWRs), each rated at 788 megawatts (electrical) capacity, and is located near Surry in Virginia. Grand Gulf, a boiling water reactor, was selected as the plant to be analyzed in a parallel study that was performed by Sandia National Laboratories (SNL). S.2 Methodology In order to take into consideration different shutdown configurations, the same outage types, plant operational states, and time windows as those used in internal event analysis were used in the fire risk assessment. Outages were grouped into four different types: refueling, drained maintenance, non-drained maintenance with use of the residual heat removal (RHR) system, and non-drained maintenance without the use of the RHR system. Due to the continuously changing plant configuration in any outage, plant operational states (POSs) were defined and characterized within each outage type. Each POS represents a unique set of operating conditions (e.g., temperature, pressure, and configuration). For example, in a refueling outage, up to 15 POSs were used They represent the evolution of the plant throughout a refueling from low power down to cold shutdown and refueling, and back up to low power. An extensive effort was made to collect Surry-specific data needed to characterize each POS. This included review of operating and abnormal procedures for shutdown operations, review of shift supervisor's log books, review of monthly operating reports, and performance of supporting thermal hydraulic calculations. In Phase 1 of the fire risk assessment project, a scoping study was performed on mid-loop operations and refueling operations. Critical rooms were identified and modeled by assuming that a fire in such rooms destroys all the equipment contained in the room (including any cables passing through). The scenarios were then used to modify the nt PRA model, and were subsequently quantified. The most important scenarios were: a fire in the emergency itchgear room (ESGR) which causes a station blackout and disables the steam generator (SG) feed and bleed, xiii NUREG/CR-6144

charging pumps and gravity feed; a fire in the control room which causes a station blackout and disables gravity f . Low Head Injection (LHI) MOVs and SG PORVs; a fire in the auxiliary building which causes a long-term los RHR and disables charging pumps; and a fire in the safeguards area which causes a long- term loss of RHR, SG feed and bleed, and gravity feed. In Phase 2, a detailed fire risk assessment was performed for mid-loop POSs. Three mid-loop POSs, in which the reactor coolant system (RCS) level is lowered to the mid-plane of the hot leg, were selected for detailed analysis. Two of them occur in a refueling outage, POSs R6 and RlO, and one in a drained maintenance outage, POS D6. They are characterized by different decay heat levels and different plant configurations, such as the number of RCS loops that are isolated, and whether or not the RCS has a large vent. R6 represents a mid-loop operation that takes place early in a refueling outage. This mid-loop operation allows fast draining of the RCS loops to permit eddy current testing of the steam generator tubes. R10 takes place after refueling operation is completed to allow additional maintenance of equipment in the RCS loops. D6 represents mid-loop operation in which maintenance activities require the plant to go to mid-loop, and is characterized by the highest decay heat level among the three mid-loop POSs. In order to more accurately define the decay heat level when an accident is initiated, a time window approach was developed. A total of four time windows after shutdown were defined, each with its unique set of success criteria . reflecting the decay heat level. For POSs R6 and D6, all four windows were needed. For POS R10, only time windows 3 and 4 are applicable. A statistical analysis on the time to mid-loop and the duration of mid-loop was performed to determine the probability that a given accident occurs in a particular time window, conditional on the accident occurring. In this approach, an event tree was developed for each accident initiating event, POS, and time window. Surry-specific fire frequencies for important plant areas and various types of fire: cable fires, transient fires, equipment fires (e.g., switchgear panels, pumps, etc.) were assessed using the latest available information. The

considered in this study as a localized phenomenon (as it usually is in nuclear power plants). Hence, the analysis pinpoints the precise location of possible sources of fire and vulnerable equipment and/or electrical cables within a plant fire area. For that reason, the cables of the most important systems (for this study) have been traced with a high degree of confidence. Fire growth calculations (with COMPBRN-IIIe) have been performed for the most vulnerable locations uncovered in our analysis. A sophisticated, transition-diagram type of suppression model has been used, which, in conjunction with the fire growth model, gives the damage fraction (i.e., the probability of damage given a fire in that location). The damage fraction was used in the event tree and fault tree models to arrive at the core damage frequency (CDF) from a given fire scenario. The scenario-dependent human error probabilities (HEPs) have been estimated using the same human reliability analysis (HRA) method used in the internal event analysis. A cutset by cutset analysis of possible operator recovery actions was also performed. Uncertainty analysis of the core damage frequency was also performed by propagating the uncertainties associated with the parameters used in the model.

S.3 Results and Insights Table S.1 summarizes the point estimate results of the quantification. Table S.2 summarizes the results of the uncertainty analysis on the total core damage frequency due to fires. No uncertainty analysis was done for individual sequences. Note that the CDF is the frequency at which core damage occurs when the plant is at mid-loop. It accounts for the fact that the plant is at mid-loop only a small fraction of the time. The quantification indicates that certain scenarios in the H and J compartments of the emergency switchgear room, one scenario in the cable vault and tunnel, and one containment scenario dominate the CDF. The most dominant scenarios occur in the cable vault and tunnel (due to proximity of many emergency cables from both divisions in a closed, constrained space) and in the J room of the ESGR, where many emergency cables from both the H and the J divisions come together in proximity (before entering the control room). In the containment, the relatively high CDF is due to a relative! NUREG/CR-6144 xiv

nario frequency combined with non-separation of RHR trains over significant distances. Other scenarios are also portant, due to a moderate damage from the fire combined with a relatively high scenario frequency. Figure S.1 shows the contribution of various fire areas to the core damage frequency. POSs D6 and R6 are much more important than RlO (as RlO occurs in later time windows). D6 is more important than R6 due to constraints imposed by a drained maintenance outage and its tendency to occur in earlier time windows. The earlier time windows are more important than the later ones, with window 4 being relatively unimportant. Windows 1 and 2 are of the highest importance, with window 2 being significantly more important than window 1. While the decay heat is higher and the success criteria are more stringent in window 1, this window doesn't last as long and the outages tend to occur in the later time windows. The most risk significant fire initiator occurs in the cable, vault tunnel area, in window 2 and POS D6, followed by a few scenarios in the J room of the ESGR, in the same window and POS. Before the recovery actions were applied, the core damage frequency due to fire events at mid-loop was 2.7E-05/yr. After application of recovery actions, the core damage frequency is 1.7E-05/yr. Note that this point estimate differs from the total core damage frequency of Table S.1, i.e., 1.SE-05/yr. This is because the total core damage frequency of Table S.1 was calculated simply as the sum of the frequencies of the sequences, while the point estimate of 1.7E-05/yr was calculated using the minimum-cutset-upper-bound method implemented in the IRRAS computer code. The recovery actions do not reduce the core damage frequency by much, because they will only be effective in windows 3 and 4 (in order to satisfy the 24 hour success criterion), whereas windows 1 and 2 are more dominant. ble S.2 summarizes the result of the uncertainty analysis for core damage accidents initiated by fires. The results were obtained by performing uncertainty analysis using 500 Latin Hypercube Sampling (LHS) samples. Also shown in the table is the uncertainty analysis results of the internal event analysis as well as the mean value of the internal fire analysis of NUREG-1150 .

NUREG/CR-6144

ESGR MCR 4.9% NSGR 0.3% ~. CT 3.4% Figure S.1 Fire Area Contribution to CDF

Fire Area Emergency Table S.1 Summary of Point Estimate CDFs for Fire Events (/yr)

R6 4.lE-06 D6 8.2E-06 RlO 2.lE-07 Total 1.3E-05 Switchgear Room Containment 7.0E-08 5.5E-07 5.0E-09 6.3E-07 Cable Vault and 1.3E-06 2.7E-06 7.4E-08 4.0E-06 Tunnel Normal 1.5E-08 3.5E-08 1.4E-09 5.lE-08 Switchgear Room Main Control 7.0E-08 5.3E-07 4.4E-09 6.0E-07 Room Total 5.5E-06 1.2E-05 2.9E-07 1.8E-05

xvii NUREG/CR-6144

Table S.2 Results of the Fire Uncertainty Analysis Fire Analysis- Internal Events Mid-Loop Analysis-Operation Mid-Loop Operation (CDF/yr. while (CDP/yr. while at mid-loop) at mid-loop) Mean* 2.2E-05 4.9E-06 5th Percentile 1.4E-06 4.8E-07 50th Percentile 9.lE-06 2.lE-06 95th Percentile 7.6E-05 1.SE-05 Error Factor 7.2 5.7 Point Estimate 1.SE-05 5.lE-06

NUREG-1150 estimated a mean CDF of l.lE-05 per year for power operation .

NUREG/CR-6144 xviii

FOREWORD (NUREG/CR-6143 and 6144) Low Power and Shutdown Probabilistic Risk Assessment Program Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analysis that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power

and shutdown could be significant contributors to risk.

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects performed by Brookhaven National Laboratory(BNL) and Sandia National Laboratories(SNL), with the seismic analysis performed by Future Resources Associates. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents due to internal events, internal fires, internal floods, and seismic events initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes t of a level-3 PRA. e results of the program are documented in two reports, NUREG/CR-6143 and 6144. The reports are organized as follows: For Grand Gulf: NUREG/CR-6143 - Evaluation of Potential Severe Accidents during Low Power and Shutdown Operations at Grand Gulf, Unit 1 Volume 1: Summary of Results Volume 2: Analysis of Core Damage Frequency from Internal Events for Operational State 5 During a Refueling Outage Part 1: Main Report Part 1A: Sections 1 - 9 Part 1B: Section 10 Part lC: Sections 11 - 14 Part 2: Internal Events Appendices A to H Part 3: Internal Events Appendices I and J Part 4: Internal Events Appendices K to M Volume 3: Analysis of Core Damage Frequency from Internal Fire Events for Plant Operational State 5 During a Refueling Outage Volume 4: Analysis of Core Damage Frequency from Internal Flooding Events for Plant Operational State 5 During a Refueling Outage Volume 5: Analysis of Core Damage Frequency from Seismic Events for Plant Operational State . 5 During a Refueling Outage xix NUREG/CR-6144

Foreword (continued) Volume 6: Evaluation of Severe Accident Risks for Plant Operational State 5 During a Refueling Outage Part 1: Main Report Part 2: Supporting MELCOR Calculations For Surry: NUREG/CR-6144- Evaluation of Potential Severe Accidents during Low Power and Shutdown Operations at Surry Unit-1 Volume 1: Summary of Results Volume 2: Analysis of Core Damage Frequency from Internal Events during Mid-loop Operations Part 1: Main Report Part lA: Chapters 1 - 6 Part lB: Chapters 7 - 12 Part 2: Internal Events Appendices A to D Part 3: Internal Events Appendix E Part 3A: Sections E.1- E.8 Part 3B: Sections E.9 - E.16 Part 4: Internal Events Appendices F to H Part 5: Internal Events Appendix I Volume 3: Analysis of Core Damage Frequency from Internal Fires during Mid-loop Operati Part 1: Main Report Part 2: Appendices Volume 4: Analysis of Core Damage Frequency from Internal Floods during Mid-loop Operations Volume 5: Analysis of Core Damage Frequency from Seismic Events during Mid-loop Operations Volume 6: Evaluation of Severe Accident Risks during Mid-loop Operations Part 1: Main Report Part 2: Appendices NUREG/CR-6144 xx

ACKNOWLEDGEMENTS Cooperation, technical assistance, and plant visit logistics arrangements of personnel at Virginia Power (especially, but not limited to, Candee Lovett, Bob Scanlan, Mike Kacmarcik, Ed Cosby, and Ron Muschingheim) were excellent, tireless, and indispensable to this work. The contribution of Florence O'Brien in completion of this manuscript was essential and appreciated. This work was sponsored by the U.S. Nuclear Regulatory Commission's Office of Research.

xxi NUREG/CR-6144

1.1 Scope of the Study

1. INTRODUCTION AND OVERVIEW Phase 2 is a continuation of Phase lA work of Sony Low Power and Shutdown Study -- fire risk scoping analysis. The analysis presented here is a detailed look at fire risk, including location dependency, plant specific data, fire growth, and fire suppression modeling. This evaluation has been done for the mid-loop plant operational states, i.e., POSs 6 (refueling and drained maintenance outage) and 10 (refueling outage). These are the POSs identified in Phase 1 as having the highest risk potential.

In this report, both the point estimates and the uncertainty of Level 1 risk are presented Fires outside the plant are not considered (e.g., a fire in the transformer yard that causes a loss of offsite power). 1.2 Results of Previous Fire Risk Analyses for Surry In Phase 1A of this study (which was a screening analysis), critical rooms were identified and modeled by assuming that a fire in such rooms destroys all the equipment within (including any cables passing through). The scenarios were then used to modify the plant Previous Risk Analyses (PRA) model, and were subsequently quantified The most important scenarios were: fire in the emergency switchgear room (ESGR) causes station blackout and disables the steam generator (SG) feed and bleed, charging pumps and gravity feed; fire in the control room causes station blackout, disables gravity feed, Low Head Injection (LHI) MOVs and SG PORVs; fire in the auxiliary building causes

     -term loss of residual heat removal (RHR) and disables charging pumps; and fire in the safeguards area causes
     -term loss of RHR, SG feed and bleed, and gravity feed.

Generic room fire frequencies were used and Phase 1 (internal PRA screening analysis) event trees were modified to suit a particular room fire. A simplified suppression model was used Tables 1.1 and 1.2 show the fire frequencies and the associated core damage frequencies for important scenarios in the Phase lA work; Tables 13 and 1.4 show the same information from the 1150 work on fires at full power. Note that all the tables in the chapter show only the background material and results of earlier studies; no results of this study are presented herein. 1.3 Methodology Overview In this work, we use Sony updated fire frequencies for important plant areas and various types of firei cable fires, transient fires, and equipment fires (e.g., switchgear panels, pumps, etc.). The fire is considered in this study as a localized phenomenon (as it usually is in nuclear power plants). Hence, the analysis pinpoints the precise location of possible sources of fire and vulnerable equipment and/or electrical cables within a plant fire area. For that reason, the cables of most important systems (for this study) have been traced with a high degree of confidence. Fire growth calculations (with COMPBRN-IIIE) have been performed for the most vulnerable locations uncovered in our analysis. A more sophisticated, transition-diagram type of suppression model has been used, which, in conjunction with the fire growth model, gives us the damage fraction (i.e., the probability of damage given a frre in that location). This is used in the fire analysis-modified plant modeJ of Phase 21 (fault trees and event trees) to arrive at the core damage frequency from a given fire scenario. An analysis of operator recovery is also performed

1-1 NUREG/CR-6144

Introduction and Overview 1.4 Assumptions and Simplifications Some assumptions and simplifications were used in this study. As in the internal PRA, the 24-hour mission time was kept as the success criterion. Cutset probability cutoff was kept at 10-8, and the fire scenario cutoff (meaning pre sequence quantification)was set at 10-1* Multiple initiators were not considered ( e.g., loss of offsite power subsequent to a fire) for probabilistic reasons. For example, loss of offsite power probability in any 24 hour period is about 7.E-4 for the Surry plant at mid-loop. Multiplying this by the mid-loop time fraction of 0.066 means that the scenario frequency would have to be a few times 10-3 in order for it to survive screening. No such scenarios were found. Barrier failure was not considered as it was deemed probabilistically unimportant. For initiator-causing fires that occur near an important barrier (very few examples were found), the combination of fire frequency, probability of equipment failures causing an initiator, the mid-loop POS fraction, and barrier failure probability (on the order of 10-2-10*3)2 was such that the fire scenario frequency was below the screening value. Also, comparing this probability to the conditional core damage probability for most scenarios (0.1) leads to screening of this consideration. It should be noted that in our visits to the plant (including during shutdown), it was observed that all normally closed fire doors were kept closed. Certain cables were not traced due to lack of resources and/or because the scenarios would be probabilistically unimportant, or would not lead to an initiator, or would be irrelevant post-initiator for the particular initiators possible in the particular fire area (various compressed air systems, diesel generator cables, ventilation system cables, etc.). In some instances assumptions were made for the location of these and other cables (i.e. cables in conduits for w

no precise tracing system is available at Surry).

It was assumed, for purposes of fire frequency calculations, that other plants are similar to Surry in terms of the number of pieces of equipment of a certain type in certain plant fire areas (e.g., number of switchgear cabinets in the emergency switchgear room). Relay fires were screened probabilistically, because the combination of a relay fire frequency, mid-loop fraction, and the conditional probability of initiator occurrence was below the screening value. The reader should refer to Sections 3.5, 4.3 and 5.1 for an in-depth discussion of assumptions and simplifications and justifications thereof. It should be noted that both power and control cables were traced and the spurious operation following hot shorts was treated. Lack of electrical coordination between the 4160 V emergency buses and the 480 V load centers was also taken into account. 1.5 Internal PRA Results The fire core damage PRA model relies heavily on the internal PRA model1, as the latter is used (with modifications, to account for equipment taken out by the fire) in our calculations. Certain numerical results of the internal PRA model were used in screening out select fire scenarios. Table 1.5 presents a condensed summary of the internal PRA results. NUREG/CR-6144 1-2

Introduction and Overview

Plant VISits and Interaction with Virginia Power It should be noted that the plant was visited on several occasions for better understanding of the systems, layout, fire protection features and for cable tracing. One of these visits was during the refueling outage in early 1992 at Surry 1; access was obtained to various areas of the plant (including the containment and all the other areas considered in this study). In addition, there have been many occasions of written and telephone correspondence when a better understanding of a certain aspect was desired.

1.7 Organization of the Report It_ should be noted that development of the main control room (MCR) fire scenarios and associated frequencies was treated as a separate task, which was carried out at a separate location. The product of this task was input into the plant model and quantified together with the other areas. However, it was necessary that the writeup on all aspects of scenario development and quantification for this area be put into a separate chapter (Chapter 9) and not included with the independent analysis of other areas (Chapters 3-8). After a brief summary of methodology used, in Chapter 2, the important fire areas are identified im Chapter 3. Equipment within and fire protection features for each area are also shown in this chapter. Fire frequency calculation methodology and results are presented in Chapter 4. The scenario development is shown in Chapter 5 for each fire area of interest. Fire growth modeling and results are presented in Chapter 6. Fire suppression is discussed in Chapter 7 and the damage fraction calculations in Chapter 8. The modifications of the internal PRA model are 1

        *bed in Chapter 10 and the* quantification aspects in Chapter 11. The results are presented in Chapter 12.

References

1. T-L. Chu, Z. Musicki, P. Kohut, et al., "Evaluation of Potential Severe Accidents During Low Power and Shutdown.

Operations at Surry Unit-1: Analysis of Core Damage Frequencies from Internal Events During Mip-Loop Operations", NUREG/CR-6144, Volume 2, June 1994.

2. M.P. Bohn, et al., "Analysis of Core Damage Frequency: Surry Power Station, Unit 1 External Events," December 1990, NUREG/CR-4550, Vol. 3, Rev. 1, Part 3.

1-3 NUREG/CR-6144

Introduction and Overview Table 1.1 Shutdown Fire Frequencies (6) from a previous study (not used in this study) Frequency (per shutdown year) Area Type Mean 5th 50th 95th Control Room 6.3x10*3 2.lx104 25x10*3 2.4x10*2 (7.19x 10*7/hr) Cable Spreading Room 5.6x10*3 1.7x104 2.0xl0-3 1.4x10*2 (6.39x10*7/hr) Auxiliary Building 7.2x10*2 6.1x10*3 s.ox10*2 2.ox10*1 (8.22x 10-6/hr) Turbine Building 45x10*2 9.9x104 1.9x10*2 l.Sxto*1 (5.14x 10-6/hr) Containment 4.7x10-2 (5.37x10-6/hr) NUREG/CR-6144 1-4

Introduction and Ovetview Table 1.2 Results of Coarse Screening Fire Analysis for POS 6 (from a previous study - not results of this study) Fire Area Event Tree Sequence CDP* Early or Containment Late (E/L) Oosed or Open (C/0)

1. Cable Vault and RHR 5 L E 0 Tunnel 8 M E 0

3. Emergency Switch SB 5 L E 0 Gear Room 9 H E 0

5. Control Room SB 4 L E 0 7 H E 0

15. Containment RHR 4 L E 0 7 M E 0

17. Auxiliary Building RHR 3 M E 0 5 H E 0

19. Safeguard Area RHR 3 H E 0

31. Turbine Building RHR 5 L E 0 8 M E 0

"' &lq uence CU!' ranKing, J-'low M-meru.um, H-'hl'.h g 1-5 NUREG/CR-6144

Table 1.3 Surry Fire Initiating Event Frequencies for Full Power Operation (/yr), from NUREG/CR-4550 Fire Area Mean 5th 50th 95th Percentile Percentile Percentile Control Room 1.8E-3 1.2E-6 9.6E-4 7.4E-3 Cable Vault/funnel 7.SE-3 3.0E-6 1.8E-3 1.6E-2 Emergency Switchgear Room 8.0E-3 2.0E-5 2.4E-3 1.7E-2 Auxiliary Building 6.6E-2 2.7E-2 5.9E-2 1.6E-1 Pump Room 3.7E-3 (Lognormal EF = 3) I °'

Table 1.4 Surry Fire Area Full Power Co mage Frequency, from NUREG/cr-4550 Fire Area Core Damage Frequency (/yr) Mean Emergency Switchgear Room 6.09E-6 Control Room 1.58E-6 Cable Vault{funnel 1.49E-6 Auxiliary Building 2.18E-6 Charging Pump Service 3.92E-8 Water Pump Room Total 1.13E-5

Table S.1 Summary of Results-Core-Damage Frequency by Initiating Event and Plant Operational States I I Initiating Event I IE Frequency I Core-Damage Frequency (per year)

1. Loss of RHR R6 RlO D6 Total RHR2A-Ovcr Draining 1.6E-02/Demand 1.SE-7 S.3E-8 2.6E-7 4.9E-7 RHR2B-Failure to Maintain Level 1.2E-0S/hr 2.lE-08 2.0E-8 2.9E-8 7.0E-8 RHR3-Non-Recoverable Loss of RHR 4.lE-06/hr 1.SE-7 8.4E-9 3.0E-7 4.6E-7 RHR4-Non-Recoverable Loss of Operating Train of RHR S.3E-06/hr 7.6E-9 l.2E-9 2.3E-8 3.2E-8 RHR5-Recoverable Loss of RHR 2.lE-05/hr 4.0E-8 4.lE-09 9.3E-8 1.4E-7

2. LOOP-Loss of Offsite Power 7.0E-06/hr Ll-Both lH and 1J Energized 6.2E-06/hr 3.3E-7 7.0E-8 7.6E-7 1.2E-6 L2-1H and 2H energized, not lJ 7.4E-07/hr 1.0E-7 1.3E-8 1.?E-7 2.9E-7 L3-1H energized, not lJ, unit 2 blackout 3.SE-08/hr 4.2E-8 1.3E-8 9.9E-8 1.5E-7 Bl-Unit 1 Black Out 2.0E-08/hr 4.SE-8 l.lE-8 1.?E-7 2.3E-7 B2-2 Unit Blackout 3.2E-09/hr 3.SE-8 4.2E-8 1.lE-7 1.9E-7

3. 4KV-Loss of 4kv Bus 2.lE-05/hr 1.4E-7 1.9E-8 2.4E-7 4.0E-7

4. VITAL-Loss of Vital Bus S.6E-06/hr 2.SE-8 S.lE-9 7.3E-8 UE-7

5. AIR-Loss of Outside Instrument Air 2.lE-6/hr 7.9E-10 - 3.2-9 4.0E-9

6. CCW-Loss *of CCW 3.SE-06/hr 6.3E-8 l.lE-10 2.lE-7 2.7E-7

7. SWOR-Loss of Emergency Switchgear Room Cooling 1.SE-08/hr 3.6E-8 1.2E-8 7.4E-8 1.2E-7

8. ESFAS-lnadvertent Safety Feature Actuation 1.lE-04/hr 2.7E-7 2.7E-8 6.SE-7 9.SE-7

9. Dilute-Boron Dilution (estimated CDF) 2.0E-07/hr 6.SE-08 I TOTAL I I I 1.5E-6 I 3.0E-7 I 3.3E-6 I 5.lE-6' I

Not including ~ron dilution

2. METHODOLOGY OVERVIEW A more detailed discussion of certain aspects of the methodology and the results is presented in Chapters 3-10.

2.1 Fire Frequency Calculations Fire frequency calculations were performed for important fire areas and for equipment found in these plant areas. Equipment classes were broken into subclasses, e.g. switchgear panels were divided up into bus panels, monitor control center (MCC) panels, uninterruptable power supply (UPS) panels, etc. Cable fires and transient fires were also considered. Large fires and small fires were calculated separately. Generic and Suny-specific fire events were obtained from the updated Sandia Fire data base1, which considers events through 1989. These events were used to calculate Suny- specific fire frequencies in specific categories. These fire frequencies were then prorated to account for the fraction of fires in that category that occurs at a specific location. 2.2 Scenario Development Fire areas in which important equipment is located were identified. The areas are such that a fire in a certain location within the area might cause an initiating event by disrupting RHR operation. Cable routing information within the area was also included in the analysis, both for power and control cables for important systems. Important systems are the ones which are identified in the internal PRA for providing alternate core cooling paths. Systems and components affected when the fire damages certain equipment and/or cable runs in a fire area location would result in an initiator and possibly degraded configuration of alternate core cooling paths. This information is eventually ded in the event trees and fault trees from the internal PRA. Whether or not a certain cable would be damaged fire in a certain location would be decided by the fire growth calculations. The probability of damage would be ermined by the fire suppression model. 2.3 Fire Growth Modeling Once a certain critical location is identified (e.g., cables for important equipment pass nearby), the fire growth model is employed to determine if a fire in that location would damage the cable (which may be located a certain horiwntal or vertical distance from the fire source). If damage does occur, the time to damage is also calculated. - The tool used for fire frequency calculations is the computer code COMPBRN-IIIE. While this program is an important advance in the fire analysis, the results are only a rough representation of reality, as certain aspects are not modeled well and simplifying assumptions have to be made2.3. For instance, the treatment of fire growth and the hot gas layer is such that ventilating the room will impede target damage (i.e., increase the time to damage). In contrast, the Suny Appendix R analysis4 states that ventilating the room will aid the fire growth as more oxygen is brought in . and the products of combustion are taken out (certain fire suppressing agents will be diluted as well). For this reason (and also to prevent the fire from spreading to adjacent rooms), the ventilation to the affected area is cut off and the dampers and any sliding doors are closed once fire protection actuation occurs. In our analysis, to be conservative in accordance with the code modeling, the ventilation is cut off. 2.4 Fire Suppression and Damage Fraction Calculation After a fire has initiated, and detection occurs, suppression efforts will start to bring it under control. The detection suppression may consist of any combination of manual and automatic systems. These systems will act in such a that a certain probability exists that the fire will be suppressed before damage to a critical cable occurs. (The time 2-1 NUREG/CR-6144

Methodology Overview to damage is input from the fire growth calculations above.) The probability of damage in a certain scenario is t damage fraction, and is input as a basic event representing damage to the particular component in our PRA model, given occurrence of fire in that location. It should be noted that this analysis is a conservative simplification as, in reality, fire growth and suppression occur simultaneously (to a large extent) and there is interaction between them. The suppression model uses transition rates among various states in the suppression path toward the eventual control of the fire5*6*7* The transition rates and a particular combination of suppression steps relate to the probability of suppression in a certain time and in a plant area that contains the specified combination of detection and suppression equipment. A simplified, parametric approach has been developed7 to translate this transition model into a readily usable tool for calculating the non suppression probability (i.e., damage fraction). This suppression analysis is a more realistic approach than t~e one taken in the 1150 study8-9, because it breaks down the suppression process into spetjfic steps, and attempts to model options available at each step. 2.5 Fault Tree and Event Tree Modifications The damage fraction calculated in 2.4 is input into the modified internal PRA model as a basic event representing a possible failure mode in a given component following a fire. The internal PRA model for a given initiator (caused by fire) is modified to account for additional equipment that may be damaged in a fire. Failure modes due to fire (e.g., spurious operation or power loss) are added to the internal failure modes of the component. Certain branches in the event tree will be prohibited, e.g., restoration of the 4 kV bus cannot be accomplished if the bus panel

destroyed in a fire.

The event tree will show core damage sequences given a fire in a certain location and given occurrence of the initiator by the fire. 2.6 Model Quantification The quantification will proceed by multiplying the sequence probability by the location fire frequency and the conditional probability of the initiator to yield the sequence frequency. The latter two frequencies are sometimes combined into the scenario frequency. The sequence frequencies are evaluated by IRRAS 5.0, the computer program used in the internal PRA quantification. The modeling and quantification of human actions (for input into the PRA model} is done in two stages as in the internal PRA10* The high level human error probabilities (HEPs) are evaluated for each scenario (i.e., location-dependent fire event and its associated system impacts). The values are input into the system model, which is evaluated and yields sequence cutsets. The cutset-dependent recovery action HEPs are then applied to the most important sequence cutsets. The quantification is performed with knowledge of IRRAS algorithm limitations at high failure probabilities. Compensatory steps are taken to account for that. NUREG/CR-6144 2-2

Methodology OveIView References

l. Sandia Fire Data Base, updated version of "Nuclear Power Plant Fire Data Base, Based on NUREG/CR-4586,SAND86-0300 and EPRI NP-3179". The latest version contains events through December 1989.

2. N. Siu, "Modeling Issues in Nuclear Plant Fire Risk Analysis", presented at the EPRI Workshop on Fire Protection, Baltimore, MD, February 9-10, 1989.

3. Vincent Ho, Nathan Siu and George Apostolakis, "COMPBRN III - A Fire Haz.ard Model for Risk Analysis".

4. Virginia Power Company, "10CFRSO Appendix R Report, Surry Power Station - Units 1 and 2".

5. Nathan Siu and George Apostolakis, "Modeling 'The Detection And Suppression of Fires in Nuclear Power Plants," from International ANS/ENS Topical Meeting on Probabilistic Safety Methods and Applications, February 24- March 1,1985.

6. N. Siu and George Apostolakis, "A Methodology for Analyzing the Detection and Suppression of Fires in Nuclear Power..!_>lants", in Nuclear Science And Engineering, 94, pp. 213-226 (1986).

I \

                ,1lo i

7. Pickard, Ust8"__,"nd Garrick, Inc. (PLG, Inc.), "Beznau Risk Analysis Plant with NANO", prepared for Nordostschweitzerische Kraftwerke AG, PLG-0511, December 1989.

M.P. Bohn, J A Lambright, "Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150", SANDSS-3102, November 1990.

9. M.P. Bohn, et al., "Analysis of Core Damage Frequency: Surry Power Station, Unit 1 External Events",

NUREG/CR-4550, Vol. 3, Rev. 1, Part 3.

10. T-L Chu, Z. Musicki, P. Kohut, et al.," Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry Unit-1: Analysis of Core Damage Frequency from Internal Events During Mid-Loop Operations", NUREG/CR-6144, Volume 2, June 1994.

2-3 NUREG/CR-6144

3. IMPORTANT FIRE AREAS ANALYZED Table 3.1 shows Surry fire areas.

The fire areas descnbed below are included in our fire risk assessment analysis. A f'rre in these areas could cause an initiator and damage important equipment and systems. The important areas are the emergency switchgear room (ESGR), the cable vault and tunnel (CVI), the containment (CI), the main control room (MCR), and the normal switchgear room (NSGR). Tables 3.2, 33, and 3.4 show the floor areas and the combustible loadings, fire protection features, and the volumes and ventilation rates, respectively, of these areas. Figures 3.1 and 3.2 show fire door locations in the ESGR and the CVf, respectively. Important equipment and cable locations are shown in Figures 33, 3.4, 35, and 3.6 for the ESGR, CVT, CT, and NSGR, respectively. It should be noted that a fire in any area would show up on the fire panel in the main control room. 3.1 Unit 1 Emergency Switchgear Room (ESGR) This room is located at level 9'6" of the Service Building (below the control room and the normal switchgear room). It is split into two compartments, the H (on the left) and the J; however, there are no doors between the two. The Unit 2 ESGR is right next to it and these two ESGRs are separated by a sliding door, which closes upon actuation (manual) of the Halon fire suppression gas system. All other doors are kept closed. The fire protection system consists of smoke detectors, portable fire extinguishers, and a manually actuated Halon s room contains various switchgear panels: the 4 kV 1H and 1J and associated stub buses (which contain RHR and component cooling water (CCW) breakers); the 480 V 1H, 1Hl, 1J and lJl buses (and associated stepdown transformers); (MCCs) lHl-1 and lJl-1; the four UPSs. Important loads on these panels are shown in Tables35-3.12. The pump motors and controls are usually connected to the bus breakers, whereas the valve motors and controls are usually connected to the MCCs. An exception to this is the charging pump cooling and service water pumps which are connected to MCClHl-1 and MCClJl-1. The important valves of concern in this study are connected to the MCCs in the CVT area, not the MCCs here. The auxiliary shutdown panel, used in case of an MCR fire is also located in the ESGR. The J room is more wlnerable than the H room, because the control cables from the H side (as well 3;s the J side) pass into the J room before going up into the control room. The control cables are routed from the respective bus breaker or the MCC cabinet to the lower part of the J room and up into the MCR. The power cables run from the bus/MCC into the CVT (usually) and then into the other areas and to the component in question. Power cables for charging pumps A and C and CC pump B run only a short distance in the CVT before entering the auxiliary building. Similarly, charging pump cooling and service water pump cables run only a short distance in the CVT from the ESGR. For purposes of this analysis, power cables for charging pumps A and C and CC pump B run directly into the auxiliary building. Similarly Charging Pump Cooling (CPq and Charging Pump Service Water (CPSW) pump cables do not enter the CVT. Any cables associated with a valve modeled in this study and found in the ESGR will be control, rather than power cables (because they will be coming down from the CVT MCCs). Most pump cables will be control, too, as the power cables usually just have a short spur from the respective bus before entering other areas (CVT, auxiliary building). 3-1 NUREG/CR-6144

Important Fire Areas 3.2 Cable Vault and Tunnel (CVT) The CVT is located in the auxiliary building, between the containment, the safeguards building, and the ESGR. Its floor elevation is 13'0" and consists of three areas: narrow and long cable vault on the service building (i.e., ESGR) side, a long (40 ft) and narrow tunnel, and the high-ceiling containment penetrations vault. Fire protection consists of smoke and heat detectors and an automatic CO2 system, as well as open head and closed head sprinkler systems. Portable fire extinguishers are provided, too. The door between the tunnel and the containment penetrationsvault automatically closes upon actuation of the Cardox system; all other fire doors are kept closed. The CT vault area contains four MCCs that power and control all valves of interest. The MCCs are arranged in groups of two, so MCC lHl-2 consists of MCC lHl-2-N and MCC lHl-2-S, and likewise for the J side. Tables 3.13 and 3.14 present important loads on the MCCs. There is no physical barrier between the H and the J side, and in some areas there is not much separation between trains of some systems (e.g., the tunnel area, some containment and safeguards penetrations). The tunnel is the most vulnerable area here. 3.3 Containment (CT) For our analysis, the important area is elevation 18'4". Near the RHR flats, the important elevation is -13'0". The containment is a large open area. Portable extinguishers are provided as well as fire detection. (At shutdown, t is also a lot of activity inside.) At power, the standpipe provided is dry; at shutdown it is pressurized. A lot oft is generated inside and the welding frequency is high at shutdown. Any spilled oil would not be able to collect on the grated walkways. For the cables of interest, there is no fire-initiating equipment nearby, except for the RHR pumps at the end of the run for the pump motor cables. The cables of interest are the RHR pump cables, as no other important system's cables are routed through this part of the containment (power cable for RHR MOV-1700 is within 1 ft of the pump cables, but this is not important as the valve is in the correct position). Significantly, the cables for the two RHR pumps are located in the same tray inside the containment. A fire at pump B (or anywhere along the way) would affect both cables. The RHR pumps contain about 6 gal of oil each. 3.4 Normal Switchgear Room (NSGR) (Unit 1) This room is located above the ESGR. The Unit 1 and Unit 2 NSGRs are separated by a wall. The fire protection features include smoke and heat detectors, an automatic CO2 system, and portable extinguishers. This room contains many switchgear panels. The transfer buses D, E, and F, feeding both units' emergency buses, are located here, as are the normal buses lA, lB, and 1C, which can be used to power the emergency buses of both units via backfee~ if reserve station power or transfer buses are not available. (Normal buses 2A, 2B and 2C in Unit 2 NSGR can be used likewise; see Fig. 3.7 depicting the plant 4 kV distribution system.) NUREG/CR-6144 3-2

rtant Fire Areas Analyzed Fire Areas Not Analyzed Safeguards building: This area contains low presssure injection (LPI) and Auxiliary Feed Water (AFW) equipment ( cables were traced inside which showed both B trains in the same tray). However, a fire in this area would not disable the RHR and cause an initiator (unless it were to propagate through the penetrations, which is not an important scenario). Many other areas were screened for the same reason ( e.g., Emergency Diesel Generator (EDG) rooms). Chargingpump service water pump room (CPSWPR) (also known as mechanical equipment room #4 or MER4) and mechanical equipment room #3(MER3): These rooms contain charging pump service water pumps, a necessary support system for the charging pumps. A fire cannot knock out both pumps as they are in separate rooms; if it did, a loss of charging would lead to a delayed recoverable loss of RHR (assuming the operator didn't isolate the letdown). This is a mild initiator (from internal PRA results) and loss of charging in this manner is not considered significant. Any intervening rooms (such as U2 ESGR) between the Ul ESGR and the CPSWPR or the MER3 through which the CPSW cables would pass were not considered for the same reason. Turbine building (fB): This area contains the station service (SS) and the reserve station service (RSS) power cables that feed the plant electrical system and run between the transformers outside the TB and the NSGR. This was screened probabilistically, based on redundancy in the SS and RSS system, the fact that the trays are high above the floor (at least 21 ft), and there are no apparent fire sources in their vicinity. Auxiliary building: This area contains charging system pumps and valves, component cooling pumps ( also CC109A and Ives) and charging pump cooling pumps. The CPC system is not needed for water temperatures in question. The r systems' cables were traced inside the auxiliary building. The CCW system has four pumps, however, only two ps' cables are coming from Unit 1. (These are collocated in the same tray for about 40 ft; charging pump C power cable is in the same tray, also.) The scenarios were screened out due to probabilistic arguments based on the short run of the tray in question, lack of fire sources, redundancy, in systems and equipment locations. 3-3 NUREG/CR-6144

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Important Fire Areas Analyzed Table 3.1 Surry Power Station - Units 1 and 2 Identification of Fire Areas Fire Area Description 1 CVT-1 Unit 1 Cable Vaults and Tunnels 2 CVT-2 Unit 2 Cable Vaults and Tunnels 3 ESGR-1 Unit 1 Emergency Switchgerar and Relay Rooms 4 ESGR-2 Unit 2 Emergency Switchgear and Relay Rooms 5 CR Main Control Room 6 EDG-1 Emergency Diesel Generator Room 1 7 EDG-2 Emergency Diesel Generator Room 2 8 EDG-3 Emergency Diesel Generator Room 3 15 RC-1 Unit 1 Containment 16 RC-2 Unit 2 Containment 17 AB Auxiliary, Fuel, and Decontamination Buildings, and Boron Recovery Room 18AFOPH-1 Fuel Oil Pump House Room 1 18B FOPH-2 Fuel Oil Pump House Room 2 19 SG-1 Unit 1 Safeguards Area 20 SG-2 Unit 2 Safeguards Area 28 INS Intake Structure 31 TB Turbine Building, including Machine Shop Building, Condensate Polishing Building, Office Building, and Service Building (including Cable Spreading Rooms, Normal Switchgear Rooms, MER 1 & 2 Personnel Facility Area, HP Area, Auxiliary Boiler) 45 MER-3 Mechanical Equipment Room #3 54 CPSWPR Charging Pump Service Water Pump Room NUREG/CR-6144 3-12

Description Table 3.2 Combustible g Summary+ Sq. Ft. 1275 BTU 879,880,563 BTU/Sq. Ft. 01 Unit 1 Penetration Cable Vault 830 590,596,300 Unit 1 Service Bldg. Cable Vault 1188 836,000 Unit 1 MCC Rod Drive Room 3293 1,471,312,863 Total 446,800 03 640,140,550 A Unit 1 Emergency Switchgear Room 2670 655,024,400 B Unit 1 Relay Room 1050 1,329,876,795 Total 3720 333,302 05 236,144,000 Control Room Complex 4800 236,144,000 Total 4800 49,197 15 1,596,258,200 Unit 1 Reactor Containment 11200 1,596,258,200 142,523 Total 11200 31 197,649,375 Unit 1 Normal Switchgear Room 3840 51,300 + Appendix R for Surry gives a breakdown of combustibles in each fire area. Cable insulation contributes almost all of the combustible loading in the areas of interest except in the control room where there is a 50% contribution from class A combustibles and the containment with a 20% contribution from lube oil and grease.

Only rooms of interest in a given fire area are shown, e.g. turbine building (fire area 31) has many different rooms.

Table 3.JA Surry Power Station

Unit 1 Existing Flre Protection Features by Important Flre Area Minimum Rating Fire Area Detection Suppression Fire Area Fire Severity1 Boundaries (hrs) 1 Smoke and Heat Total Flood CO2, 3 High (CVf-1) Detection manual sprinkler system with closed head portion and open head portion 2 Smoke Detection Total Flooding 3 High (ESGR-1) Halon 1301 System 5 Smoke Detection None 3 Moderate (CR)

Unit 1 Heat and Smoke Total flood CO2 Normal SWG Room Detection 1 Low: 0-30 min equivalent fire severity (EFS) Moderate:30-90 min EFS High: greater than 90 minutes EFS

portant Fire Areas Analyzed

Table 3.3B Miscellaneous Passive Fire Protection Items - Appendix R

                                                                                                                      *I A. Containment                                                                                                         I

1. Radiant Energy Shield

a. Units 1 and 2, between RHR pump motors, elev.(-) 13 ft - 0 in.

2. Solid Metal Cable Tray Cover

a. Units 1 and 2, on bottom of lowest horizontal cable tray in cable penetration area, and on top of all cable trays in cable penetration area; elev. 18 ft - 4 in between columns 6-9 (Unit 1), and 9-12 (Unit 2).

3. Cable Tray Firestops - various locations shown on plant drawings.

B. Auxiliary Building

1. Cable Tray Firestops

a. On elevation 13 ft O in, in cable tray C2-102 at the locations shown on drawing.

C. Emergency Switchgear Rooms

1. Fire Stops

a. Units 1 and 2 at various locations shown on plant drawings.

3-15 NUREG/CR-6144

Table 3.4 Room Volume and Ventilation Rate for Select Rooms at Surry Summary of HVAC Information Room Description Room Ventilation Air Changes Volume (Ft3) Rate (CFM) Per Hour Unit 1 Cable Vault & Tunnel 45,085 5,000* 6.7* Unit 1 Normal Switchgear Room 73,260 30,000 24.6 Unit 1 Erner. Swgr. & Relay Rooms 64,953 9,400* 8.7* Control Room Complex Unit 1 Control Cabinet 3,510 7,000 120 Control Room Annex 15,551 2,000 7.7 Unit 1 Computer Room 4,200 3,000 42.9 Unit 1 A.C. Room 2,925 10,000 205

I 0\

Has been upgraded to a higher level (with corresponding changes in fire detector locations) since this analysis was performed.

Important Fire Areas Analyzed Table 3.S Important Loads 4160 VAC Emergency Bus lH, 11 Cabinets Location: Emergency Switchgear Room Source: Transfer Bus F, Breaker 15F3 Breaker Load 15H3 Emergency Generator No. 1 15H4 St. General Auxiliary Fd. PP 3A 15H5 Charging PP 1-CH-P-lA 15H6 Charging PP 1-CH-P-lC 15H7 4160V/480V Emergency XFMR lH and XFMR lH-1 ISM8 Normal Feed from Transfer Bus F 4160V lH Stub Bus 15H9 4160V Bus lH, Stub Bus Tie 15HI0 Component Cooling PP 1-CC-0-lA 15Hll Residual Heat Removal PP 1-RH-P-lA

3-17 NUREG/CR-6144

Important Fire Areas Analy7.ed Table 3.6 Important Loads 480 VAC Emergency Bus lH, 10 Cabinets Location: Emergency Switchgear Room Source: 4180 V AC Emergency Bus lH, Breaker 15H7 Breaker Load 14H3 Low Head Safety Inj. PP 1-Sl-P-lA NUREG/CR-6144 3-18

Important Fire Areas Analyzed Table 3.7 Important Loads 480 VAC Emergency Bus 1Bl, 7 Cabinets Location: Emergency Switchgear Room Source: 4160 VAC Emergency Bus lH, Breaker 15H7 Breaker Load 14H13 MCC lHl-2 Supply 14H14 MCC lHl-1 Supply 3-19 NUREG/CR-6144

Important Fire Areas Analyzed Table 3.8 Important Loads 4160 VAC Emergency Bus U, 11 Cabinets Location: Emergency Switchgear Room Source: Transfer Bus D, Breaker 1503 41(,()V Erner. Bus 1J Breaker Load 1512 Charging PP 1-CH-P-10 1513 Emergency Generator No. 3 1514 St Gen. Aux. Fd. PP 3B 1515 Charging PP 1-CH-P-lB 1517 4Hi0V/480V Emergency XFMR 1J and XFMR IJ-1 1518 Normal Feed from Transfer Bus D 41(,()V 1J Stub Bus 1519 4HiOV Bus lJ, Stub* Bus, Tie 15110 Component Cooling PP 1-CC-P-lB 15111 Residual Heat Removal PP 1-RH-P-lB NUREG/CR-6144 3-20

Important Fire Areas Analyzed Table 3.9 Important Loads 480 VAC Emergency Bus lJ, 10 cabinets Location: Emeregency Switchgear Room Source: 41(,() VAC Emergency Bus lJ, Breaker 15J7 Breaker Load 14J3 Low Head Safety lnj. PP 1-Sl-P-lB
3-21 NUREG/CR-6144

Important Fire Areas Analyzed Table 3.10 Important Loads 480 VAC Emergency Bus lJl, 7 Cabinets Location: Emergency Switchgear Room Source: 4160 VAC Emergency Bus lJ, Breaker 15J7 Breaker Load 14Jl2 lJl-lA Supply EDG3 14Jl4 MXX Ul-2 Supply 14Jl6 MCC Ul-1 Supply NUREG/CR-6144 3-22

Important Fire Areas Analyzed Table 3.11 Important Loads MCC lHl-1, 40 Cabinets Location: Emergency Switchgear Room Source: 480 VAC Emergency Bus IHI, Breaker I4HI4 Breaker Load IHI-I-11 MCC lHI-IA Emerg. Gen Room #1 Feeder lHl-1-13 I-SW-P-IOA Chg Pump Service Water Pump IHl-1-23 MOV-I867C Cold Leg Safety Injection IHI-1-31 MOV-SW-102A CC Ht Exchange Inlet lHl-1-33 MOV-1869A Chg Pump SI Stop Valve lHl-1-43 *UPS-lA-I Feeder 1Hl-1-52B *UPS-IA-2 Feeder Bypass lHl-1-53 1-CC-P-2A Chg Pump Cool Water Pump

3-23 NUREG/CR-6144

Table 3.12 Important Loads Important Fire Areas Analyzed

MCC lJl-1, 42 Cabinets Location: Emergency Switchgear Room Source: 480 VAC Emergency Bus lJl, Breaker 14116 Breaker Load lJl-124 MOV-SW-102B CC Ht Exch Inlet lJl-142 1-SW-P-lOB Chg Pump Serv Water Pump lJl-143 1-EE-P-lC Emerg. Gen. Fuel Oil Pump lJl-144 1-CC-P-2B Chg Pump Cooling Water Pump lJl-153 UPS lB-1 Feeder 1Jl-174B UPS lB-2 Fdr. Bypass lJl-183 lJl-191 MOV-18670 Cold Leg Safety Injection MOV-1869B Chg Pump Safety Injection Stop NUREG/CR-6144 3-24
Location: Cable Vault Table 3.13 Important Loads MCC 1Bl-2 Source: 480 VAC Emergency Bus lHl, Breaker 14H13 Breaker Load lHl-2-North, 44 cabinets lHl-211 MOV-FW-260A FW Cross Tie lHl-212 MOV-1860A Lo Hd Safety Injection Suction lHl-213 MOV-LCV-1115B Chg Pump Suction RWSf lHl-223 MOV-LCV-1115C Chg Pump Suction VCT lHl-232 MOV-1862A LHSI Pump Suction RWST lHl-242 MOV-1842 Alt High Hd Cold Leg lHl-281 MOV-1890A Lo Hd SI to A Loop Hot Leg lHl-291 MOV-1890C Lo Hd SI to Cold Leg lHl-2101 MOV-1864A Lo Hd SI Pump Disch
lHl-2 South, 43 Cabinets lHl-212 lHl-213 MOV-FW-151A Aux FW Header MOV-1720A RHR Outlet Isol lHl-223 MOV-FW-151E Aux FW Header lHl-224 (2D)
UPS-lA-2 Feeder lHl-233 MOV-FW-151C Aux FW Header lHl-241 MOV-1270A Chg Pump Suction lHl-251 MOV-1286A Chg Pump Disch lHl-262 MOV-1286C Chg Pump Disch lHl-282 MOV-1267B Chg Pump Suction lHl-283 MOV-1700 RHR Inlet Isol lHl-2103
UPS-lA-1 Feeder Bypass
3-2.'i NUREG/CR-6144

Important Fire Areas Analyzed Location: Cable Vault Table 3.14 Important Loads MCC lJl-2

Source: 480 VAC Emergency Bus lJl, Breaker 14114 Breaker Load lJl-2 West, 37 cabinets lJl-212 MOV-FW-lSlD Aux Fd Water Header lJl-213 MOV-1269A Chg Pump Suction lJl-222 MOV-FW-151B Aux Fd Water Header lJl-223 MOV-1269B Chg Pump Suction lJl-232 MOV-FW-151F Aux Fd Water Header lJl-233 MOV-1270B Chg Pump Suction lJl-243 MOV-1286B Chg Pump Discharge 1J1-244A (4D-1) UPS lB-1 FDR. Bypass lJl-252 MOV-1287B Chg Pump Discharge lJl-271 MOV-1287C Chg Pump Disch lJl-2101 (lOA) UPS lB-2 Feeder MCC lJl-2 East, 40 cabinets lJl-2122 MOV-1862B LHSI Pump Suction RWST lJl-2141 MOV-FW-260B Aux Fd Water Cross Tie lJl-2151 MOV-LCV 1115D RWST Chg Pump Suction lJl-2152 MOV-1860B LHSI Suction lJl-2162 MOV-1864B LHSI Disch lJl-2182 MOV-1890B LHSI B Loop Hot Leg lJl-2101 MOV-1720B RHR Outlet Isol lJl-2102 MOV-1701 RHR Inlet Isol NUREG/CR-6144 3-26

4. FIRE FREQUENCY CALCULATIONS

~ 1 Methodology Sandia Fire Data Base1 was used to compile a list of possible fire events in the categories of interest and plant areas of interest. Certain criteria were applied as to the admissability of events. Generic fire frequency distribution in various categories was obtained by applying a Bayesian updating scheme. The generic results were then updated with Surry-specific data to arrive at the Surry-specific frequency distributions. The mean of a distribution is used as a point estimate of fire frequency in that category. Fire frequencies were computed in three general areas: cable fires (both self-ignited and externally ignited), transient fueled fires (specialized to important plant fire areas), and equipment fires. The latter ones were subdivided into fire size categories (small, medium, large) and specialized to a particular type of equipment (e.g., MCC panel, RHR pump, etc.). 4.2 Sandia Fire Data Base Pertinent fire events in various categories were compiled from the Sandia data base1, both for Surry and for the nuclear industry as a whole. While Phase lA fire analysis used the older version of this data base (covering events reported through,June 1985), this study had access to the just released latest version (including events through December 1989). The data base has a summary and a narrative part. The summary shows the plant and date of the event, the operating mode of the plant, the fire source and affected equipment, dollar loss and impact on plant operations, and time and method of suppression. The narrative goes into more detail on the circumstances of the fire but is usually only a few sentences

g. Both parts have to be used in order to classify the fire for the purposes of this study. The events can be also uped by plant and this feature was used to compile Surry-specific events. In some fraction of cases, certain important information is missing from the data fields or the narrative and assumptions have to be made.

4.3 Criteria for Inclusion/Exclusion of Events Certain criteria were applied. Events that can happen only during power operation ( e.g., Reactor Coolant Pumps (RCP) electrical fires, or oil coming in contact with hot piping) were thrown out. By the same token, certain events can only happen at shutdown, and not in power operation. This is sometimes a gray area, and a judgement call occasionally had to be made. Such events were noted and a separate time period for frequency calculations was used for them. Construction events were also excluded. The time period for inclusion of events was from initial criticality of the plant to its decommissioning or December 1989, whichever was sooner. Data from dissimilar components of the same class was not used For instance, fires in RCPs or Main Feedwater (MFW) pumps were not judged relevant for estimating the frequency of fires in RHR pumps. (CCW pumps were judged to be similar, however.) Also disregarded were events in irrelevant fire areas, components or fuels (e.g., transformer yard fires, turbine generator fires, hydrogen fires). Diesel generator fires are irrelevant because such an event would not cause an initiator (meaning loss of RHR), unless the plant was already running on the EDG due to a prior loss of offsite power. 4-1 NUREG/CR-6144

Fire Frequency Calculations In the latter case, any diesel generator fires would be folded into the category "failure of the EDG to run for dura of mission time" which would be included in the analysis of loss of offsite power in the internal PRA. Relevant events from all types of nuclear power plants were counted. Fires that were too small or weak to cause damage were disregarded, e.g., smoldering fires, self-extinguishing fires, events where smoke but no flame was observed (or a burning smell but no open flame), or events where deenergizing equipment extinguished the fire. For surviving fires, classification into small, large and medium fires was performed depending on several factors: dollar amount of loss, if the unit had to be shut down, the extent of the damage, time to suppression and what type of extinguishing equipment was used. For example, if the fire was successfully suppressed in a short time by portable extinguishers only, it was judged to be a small fire. If the whole MCC panel or a transformer was destroyed, it was judged to be a "large fire". 4.4 Exposure Time at Power and Shutdown In order to calculate frequency of fires, the exposure time of U.S. nuclear power plants and Surry has to be known. This time is calculated, for a particular plant, as the difference between the date of decommissioning or December 1989 (whichever was sooner) and the date of initial criticality of the plant. The table in Appendix A was used for that purpose. It shows the pertinent dates for each plant along with its exposure time. Most of the data comes from the data base (or was calculated). Some missing dates were taken from the latest world list of nuclear power plants in Nuclear News2* The resulting exposure time for all U.S. plants and for Surry are shown in Tables 4.1 and 4.2, respectively. category "shutdown time only" is also shown. This is computed for all U.S. plants by assuming an average down of 35%. For Surry, data was taken from the Phase lA report3 on the frequency and duration of various types o shutdown configurations to calculate an average downtime fraction. 4.5 Pertinent Events from Data Base Tables 4.3 and 4.4 list a summary of pertinent events in various categories for generic data and Surry-specific data. 4.6 Bayesian Updating Scheme For calculating the fire frequency distnbution, a Bayesian updating program was employed with a noninformative prior (1/lambda) and the Poisson likelihood function. The number of events and number of plant years for each category was input. The resulting frequency distribution is shown in Table 4.5, along with the parameters of the lognormal distribution that would fit it and the number of observed events at Surry in that category. The lognormal distribution was input as a prior, along with Surry evidence and plant years in that category in a second Bayesian updating scheme. The resulting Surry-specific fire frequency distribution is shown in Table 4.6 and the point estimates used in Table 4.7. The frequencies for (operation+ shutdown) and for (shutdown only) events are simply added to arrive at a frequency per shutdown year. Frequencies for cable self-ignited and externally ignited fires (which are approximately equal) are added together. NUREG/CR6144 4-2

Fire Frequency Calculations Component Based Fire Frequency The fire frequency mean value from Table 4.6 was converted to component based frequencies by dividing by the number of components of that type (e.g., MCC panels, pumps etc.). It is assumed that all plants have the same number of components of the pertinent type per plant. The number of components and the corresponding component based fire frequency are shown in Table 4.8. The corresponding numbers for the cable and transient fires are shown in Tables 4.9 and 4.10. Note that in the equipment category, an event that never happened at a nuclear power plant was added: large switchgear equipment fire. This is postulated to be a large and smoky fire that disables not only the initiating component but also all the other switchgear in the room (medium voltage switchgear equipment is susceptible to arcing damage in a smoky environment). According to the Sandia cabinet fire test results4, most cabinet fires evolve large amounts of smoke. The cable fire frequency is adjusted for the difference between Surry and an average plant in the number of cable tray feet. The average number of tray-feet per plant5, 85,000 tray-feet, was obtained from a sample of 32 plants started up since May, 1974. The Surry units have an average of about 25,000 cable tray feet, as shown in the breakdown in Table 4.11 (compiled from Appendix R data6). The Bayesian updating evidence for Surry was adjusted for the lower exposure in terms of tray-feet-years. The modified numbers are shown in Table 4.9 as well as the frequency per tray-foot-year. Note that the above analysis is somewhat nonconservative, as it assumes equal likelihood of fires in control cables vs. power cables; this may be true for externally ignited fires but not for the self-ignited ones. However, this nonconservatism will be offset by applying the total mix of control and power cable tray-footage to the tray-foot frequency number to arrive at the frequency of the scenario. e transient fire frequency in Table 4.10 is calculated per square foot of the appropriate plant area. The cvr, ESGR NSGR fall into the "switchgear room" type of areas and these were lumped together to arrive at this frequency for Surry. We can distinguish between A and B types of combustibles. A is the regular trash type, whereas B is the liquid fuel (e.g., oil). It can be assumed that an A type fire will not lead to damage to nearby components. The frequency of B fires is then: lambdatr, areaB = lambdatr, area

phi where phi is the fraction of transient fires involving B type fuels. The calculation of phi is shown in Appendix C (taken from Ref. 5) and the results for the plant fire areas of interest are given in Table 4.12. The ratio of large oil fires to small oil fires will be 30:70, i.e., the same as in the 1150 analysis7*8; the same characteristics of the fuel source will be assumed (See Chapter 6.) The frequency of fire for a particular scenario will be calculated by multiplying the values in Table 4.12 by the area of influence of the fire along the cable run.

The calculated frequencies in this chapter are per shutdown year, and in the scenario calculations are to be multiplied by the average fraction of time the plant spends in the mid-loop POS of interest (i.e., by 0.017 for R6, 0.015 for RlO, and 0.035 for D6). 4-3 NUREG/CR-6144

Fire Frequency Calculations 4.8 Insights No prevalence of fires at shutdown in the data base was noticed, as compared to power operation fires (after the construction events are taken out). It is true that greater potential exists for fires in certain categories (e.g., transient or welding igniting cables or other equipment fires). It is also true that possibility of some types of fires is reduced (e.g. deenergized equipment, oil dripping on hot piping). A fire at shutdown is liable to be detected much sooner and extinguished in its early phases, because of increased floor traffic. (Credit is taken for this by disallowing events that were discovered in the smoking stage (without flames) or early enough such that deenergizing equipment extinguished the fire.) Increased vigilance by licensees may play a part in this also. At Surry, a fire watch is in place at welding operations; fire doors are kept closed. One can compare the generated fire frequencies with several other sources, namely the 1150 full power fire analysis7, the Phase lA study3 or the recently released EPRI FIVE report9

The first two studies give fire frequency per plant fire area, whereas the last reference also shows equipment based fire frequencies. The ESGR fire frequency in this Phase 2 study (when all the contributions from equipment, transients, and cables are added up) is about the same as that reported in the 1150; however, the CVT frequency is about 4 times lower. This may be due to disallowing construction events and instances where the initial fire intensity was too small. (One can argue there is less justification for the latter at power, as the fire may go undiscovered longer). The Phase lA study containment fire frequency is about twice the one derived from data developed here; again, this may be due to the same reasons as above (the Phase lA study was a scoping study and thus more conseivative). The FIVE report equipment fire frequencies are also higher than the ones here; they seem to have counted all the events, even the ones where the fire severity was too small or no actual fire resulted (the methodology presented therein is supposed to be used for screening purposes).

NUREG/CR6144 4-4

Fire Frequency Calculations

zt.9 References

1. Sandia Fire Data Base, updated version of "Nuclear Power Plant Fire Data Base, Based on NUREG/CR-4586, SAND86-0300 and EPRI NP-3179". The latest version contains events through December 1989.

2. World List of Nuclear Power Plants", Nuclear News, August 1992.

3. T-L. Chu, et al., "PWR Low Power And Shutdown Accident Frequencies Program, Phase lA - Coarse Screening Analysis", November 1991.

4. J.M. Chave7., S.P. Nowlen, "An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets, Part II: Room Effects Tests", NUREG/CR-4527 Vol. 2, Sandia National Laboratories, November 1988.

5. Memorandum from N. Siu to T-L. Chu and Z. Musicki, titled "Cable and Transient Fire Frequencies", dated July 31, 1992.

6. Virginia Power Company, "10CFR50 Appendix R Report, Surry Power Station - Units 1 and 2".

7. M.P. Bohn, J.A. Lambright, "Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150", SAND88-3102, November 1990.

~ . P . Bohn, et al., "Analy~is of Core Damage Frequency: Surry Power Station, Unit 1 External Events", NUREG/CR-4550, Vol. 3, Rev. 1, Part 3.

9. Professional Loss Control, Inc., "Fire Induced Vulnerability Evaluation Methodology (FIVE) Plant Screening Guide", December 1991.

4-5 NUREG/CR-6144

Fire Frequency Calculations Table 4.1 Exposure Time, All U.S. Plants to 12/89 Mode of Operation Exposure Time (y) Power + Shutdown 1396 Shutdown Only 490 NUREG/CR6144 4-6

Fire Frequency Calculations Table 4.2 Exposure Time, Surry Units Unit Exposure Time (y) Surry 1 17.4 Surry 2 16.8 Surry Total 34.2 Surry Shutdown 13.7

4-7 NUREG/CR-6144

Table 4.3 Important Fire Events Initiated in Pertinent Equipment rrom Sandia Fire Data Base Type or Equipment MCC Small Fires Medium Fires Large Fires INO Plant Years INO Plant Years INO Plant Years 121 Trojan 14.0 14 Oyster Creek 20.6 138 Millstone 2 14.2 185 Unknown BWR 12.6 254 Unknown PWR 11.1 268 Trojan 14.0 404 Quad Cities 2 17.7 415 Fermi2 4.5 Bus 186 Unknown 11.5 437 Oconee 1 16.7 344 Unknown 11.5 235 Unknown PWR 11.1 246 Unknown BWR 12.6 256' Unknown PWR 11.1 324 Humboldt Bay 13.4 342 Yankee Rowe 29.3 385 Yankee Rowe 29.3 401 Kewaunee 15.8 429 Palo Verde 1 4.6 Small Transformer Medium Large

Table 4.3 Important Fire Events Initiated In Pertinent Equipment from Sandia Fire Data Base (continued) 59 Turkey Point 3 17.2 396 Cook 1 14.9 56 Vermont Yankee 17.8 193 Fort St. Vrain 15.6 444 LaSalle 2 5.8 Pump 269 Palisades 13.6 358 Palisades 18.6 392' Peach Bottom 3 15.3 UPS 157' Oyster Creek 20.6 Relay 159 Duane Arnold 15.8 156 Peach Bottom 3 15.3 166 Peach Bottom 2 16.2 169 Peach Bottom 3 15.3 368 TMI-1 15.5 388' Brunswick 1 13.2 419' TMl-1 15.5 5

The fire event can only occur at shutdown

Fire Frequency Calculations Table__4.3A Important Cable and Transient Fire Events, Sandia Fire Data Base Cable Fires Self-Ignited: INO Plant (Area) Years 7 San Onofre 1 225 8 San Onofre 1 225 51 Quad Cities 2 17.7 240 Browns Ferry 1 16.3 Externally Ignited: 100 Browns Ferry 1 (CSR) 16.3 264 Unknown PWR 11.1 326 Oconee 3 152 Transient Fires INO Plant (Area) Years 57' Point Beach 1 (TB) 19.1 80 Robinson 2 (CI) 19.2 87 Unknown PWR (TB) 11.1 105 Unknown PWR (?) 11.1 128 Browns Ferry 1 (CI) 16.3 137 Fort Calhoun (CI) 16.3 158 Brunswick 1 (TB) 13.2 270 Pilgrim (RB) 175 2755 Ginna (RB) 20.1 288 Browns Ferry 1 (?) 16.3 309 San Onofre 1 (SWGR) 22.5 431 5 Wolf Creek (CB) 4.6 4385 Dresden 3 (CI) 18.9 s = This fire event could only occur while the plant is shutdown. NUREG/CR-6144 4-10

Fire Frequency Calculations

INO 735 Table 4.4 Pertinent Sun-y Fire Events Equipment/Type of Fire RHR Pump (Small)

Surry Unit (Fire Area) 1 (CT) 2385 Transient 2 (CT) s = This fire event can only occur in shutdown 4-11 NUREG/CR-6144

Fire Frequency Calculations Table 4.5 Parameters of Generic Fire Frequency Distribution (/yr) and Its Lognormal Fit from the Bayesian Updating and for Various Fire Categories and Evidence for Suny Events in that Category Type of Fire 5% 50% Mean 95% p. (1 e MCCSmall 2.35-3 4.8-3 5.0-3 8.62-3 -5.4 0.395 0 MCCLarge 3.83-5 5.12-4 7.17-4 227-3 -8.13 124 0 Bus Small 2.83-3 55-3 5.7-3 9.6-3 -5.26 0.371 0 Bus Large 2.6-4 12-3 1.4-3 3.7-3 -6.93 0.807 0 Bus Small, Shutdown Only 1.08-4 15-3 2.0-3 65-3 -7.08 124 0 Transformer, Small 5.98-4 1.9-3 2.15-3 459-3 -6.40 0.619 0 Transformer, Large 2.6-4 12-3 1.4-3 3.7-3 -6.93 0.807 0 Pump, Small 2.6-4 12-3 1.4-3 3.7-3 -6.93 0.807 0 Pump, Small Shutdown Only 1.1-4 15-3 2.0-3 65-3 -7.08 124 1 j UPS Shutdown Only

1.1-4 15-3 2.0-3 65-3 -7.08 124 0 Relay, Small 9.8-4 2.6-3 2.9-3 5.7-3 -6.05 0535 0 Relay, Medium 2.6-4 1.2-3 1.4-3 3.7-3 -6.93 0.807 0 Relay, Small Shutdown Only 73-4 3.5-3 4.1-3 1.0-2 -5.91 0.796 0 Cable, Self-Ignited 9.8-4 2.6-3 2.9-3 5.7-3 -6.05 0535 0 Cable Externally Ignited 6.0-4 1.9-3 22-3 4.6-3 -6.40 0.619 .o Transient 33-3 6.2-3 6.4-3 1.0-2 -5.16 0337 0 Transient, Shutdown Only 2.7-3 7.4-3 7.8-3 15-2 -5.06 0521 1 NUREG/CR-6144 4-12

- Fire Frequency Calculations Table__4.SA Parameters of Generic Fire Frequency Distribution (/yr) and Its Lognormal Fit from the Bayesian Updating and for Transients in Various Fire Areas+ and Evidence for Sun-y Events in that Category Number Type of Fire Fire of 5% 50% Mean 95% µ (] e Area Events Transient CT/RB 5 1.4-3 3.4-3 3.6-3 6.7-3 -5.79 0.472 0 TB 2 2.6-4 12-3 1.4-3 3.7-3 -6.93 0.807 0 SWGR 2 2.6-4 12-3 1.4-3 3.7-3 -6.93 0.807 0 Transient, Shutdown Only CT/RB 2 73-4 3.5-3 4.1-3 1.0-2 -5.91 0.796 1 TB 2 73-4 3.5-3 4.1-3 1.0-2 -5.91 0.796 0 SWGR 1 1.1-4 1.5-3 2.0-3 6.5-3 -7.08 1.24 0

  + CT/RB: containment/reactor building TB:      turbine building SWGR:      aggregate of switchgear rooms.

4-13 NUREG/CR-6144

Fire Frequency Calculations Table 4.6 Parameters of Surry-Specific Fire Frequency Distribution (/yr) from Bayesian Updating for Various Categories+ Type of Fire 5% 50% Mean 95% (fire area) MCCSmall 1.85-3 5.34-3 6.56-3 1.54-2 MCCLarge 4.0-5 2.9-4 6.2-4 22-3 Bus Small 1.5-4 1.4-3 3.4-3 12-2 Bus Large 8.8-5 6.2-4 1.3-3 4.4-3 Bus Small, Shutdown Only 1.1-4 8.2-4 1.8-3 6.3-3 Transformer Small 5.5-4 2.1-3 2.9-3 8.1-3 Transformer Large 2.9-4 1.4-3 2.2-3 6.6-3 Pump Small 2.4-5 2.9-4 9.4-4 3.6-3 Pump Small, Shutdown Only 1.9-3 6.1-3 7.9-3 2.0-2

UPS, Shutdown Only 1.13-4 8.3-4 1.7-3 6.1-3 Relay Small 1.1-3 3.1-3 9.3-3 3.9-3 Relay Medium 3.9-5 2.9-4 62-4 22-3 Relay Small, Shutdown Only 1.6-3 5.0-3 6.4-3 1.6-2 Large, Large 4.9-7 2.9-6 5.2-6 1.7-5 Switchgear Cable, Self-Ignited 1.8-4 1.4-3 2.9-3 1.0-2 Transients (CT/RB) 1.5-4 1.2-3 2.4-3 8.7-3 Transients (TB) 8.4-5 6.1-4 1.3-3 4.5-3 Transients (SWGR) 92-5 6.4-4 1.3-3 4.4-3 Transients, Shutdown Only (CT/RB) 2.6-3 9.8-3 1.4-2 3.7-2 Transients, Shutdown Only (TB) 1.1-4 8.3-4 1.7-3 62-3 Transients, Shutdown Only (SWGR) 1.9-4 1.6-3 3.8-3 1.4-2

  + NUlc:     All bre frequencies are per p,Iant year, not p er p anel year or cabinet year, similarly for pumps.

NUREG/CR-6144 4-14

Fire Frequency Calculations Table 4.7 Point Values for Plant-Wide Fire Frequency (/sdy) Category Frequency (/sdy) MCCSmall 6.6-3 MCCLarge 62-4 Bus Small 52-3 Bus Large 13-3 Transformer Small 2.9-3 Transformer Large 22-3 Pump Small 8.8-3 UPS 1.7-3 Relay Small 1.0-2 Relay Medium 62-4 Large, Large Switchgear 52-6 Cable 2.9-3 Transient, CT 1.6-2 Transient, TB 3.0-3 Transient, SWGR+ 52-3 + Aggregate of rooms containing important switchgear equipment. 4-15 NUREG/CR-6144

Fire Frequency Calculations Tabl!l 4.8 Component Based Frequency Derivation Point Estimates, Fire Frequency (/sdy) Equipment Frequency/ # Pieces* Frequency/ Plant Unit Eq't) MCC Small Fire 6.6-3 34 1.9-4 MCC Large Fire 6.2-4 34 1.8-5 Bus Small Fire 5.2-3 20 2.6-4 Bus Large Fire 13-3 20 6.5-5 1 Transformer Small Fire 2.9-3 19 15-4 Transformer Large Fire 2.2-3 19 1.2-4 Pump Small Fire 8.8-3 4+ 2.2-3 UPS 1.7-3 4 4.2-4 Relay Small Fire 1.0-2 102 1.0-3 102 Relay Medium Fire 6.2-4 6.2-5

Panels, not individual cabinets for switchgear equipment; there are about 40 cabinets/MCC, 7 cabinets/bus panel

+ 2RHR + 2CCW/Unit . 1 one per 480 V bus, two per UPS 2 estimate NUREG/CR-6144 4-16

Fire Frequency Calculations Table 4.9 Cable Fire Frequency Derivation per Tray-Foot Frequency (/sdy/ft) Cable, total Average number of Cable frequency fire cable-tray-feet/plant (/sdy/tray-ft) frequency(/sdy)

          . 2.9-3                85,000                  3.4-8 4-17                                     NUREG/CR-6144

Fire Frequency Calculations Table 4.10 Transient Fire Frequency Derivation per Square Foot Fire Area Area-wide frequency Floor area (Ft2) Fire frequency (/sdy/ft2) CT 1.6-2 11,200 1.4-6 SWGR+ 5.2-3 11,123 4.7-7

+ ESGR and CVf will be in this category.

NUREG/CR-6144 4-18

Fire Frequency Calculations Table 4.11 Cable Tray Length in Fire Areas Unit 1 Fire Area Name Cable Tray- Feet 01A Penetration Cable Vault 2649 om Sexvice Building Cable Vault 1632 03A Emergency Switchgear Room 1838 03B Relay Room 1697 08C Diesel Generator #3 45 15 Containment 5082 17B Auxiliary Building, 13'2" 400 17C Auxiliary Building, 27'6" 907 17D Auxiliary Building, 45'10" 135

17E 17F 28A 31A Fuel Building, 6'10" Fuel Building, 47'4" Intake Structure Turbine Building, 9'6" 120 470 239 3339 31B Turbine Building 27'0" 4790 31C Turbine Building, 58'6" 204 31H Unit 1 Normal Switchgear Room 1125 31J Mechanical Equipment Room #1 234 31K Mechanical Equipment Room #2 141 31L Administrative Offices 82 31R Condensate Polishing Building 596 31S Unit 1 Cable Spreading Room 1761 45 Mechanical Equipment Room #3 93 I Total I Unit 1 I 27,579 I

4-19 NUREG/CR-6144

Fire Frequency Calculations Table 4.11A Cable Tray Length in Fire Areas Unit .2 Fire Area Name Cable Tray- Feet 02A Penetration Cable Vault 2448 02B Service Building Cable Vault 2444 04A Unit 2 Emergency Switchgear Room 2422 04B Relay Room 1503 16 Containment 5120 31D Turbine Building, 9'6" 2861 31E Turbine Building, 27'0" 3172 31F Turbine Building, 58'6" 75 311 Unit 2 Normal Switchgear Room 550 31T Cable Spreading Room 1976 Total Unit2 22,571 Total Unit 1 + Unit 2 = 50,150 Tray Feet Average per Unit = 25,075 Tray Feet NUREG/CR-6144 4-20

Table 4.12 Class B Fire Frequency Fire Frequency Calculations 8

                                                                                           -Large 2                                                2                          2 Fire Area Air (/sdy/ft  )                 ti>               A (/sdy/ft   )            Air (/sdy/ft )

CT 1.4-6 038 53-7 1.6-7 SWGR 4.7-7 029 1.4-7 42-8 4-21 NUREG/CR-6144

5. SCENARIO DEVELOPMENT

. t h i s chapter, the analysis of scenarios is presented, along with quantification. Cable routing information in relation to important equipment in significant fire areas will be presented, along with impact of fire at various locations within a fire area. 5.1 Assumptions and Simplifications Some cables are routed in conduit; others are in covered trays; the rest are in uncovered trays. For purposes of COMPBRN-IIIE, the uncovered trays are relatively easy to model. The covered trays and conduit cannot be realistically modeled (by COMPBRN). As there is some question as to how effective these barriers are for preventing cable damage, we assumed that they (the barriers) do not exist. Therefore, the results generated will be somewhat conservative (in addition to some conservative aspects of COMPBRN modeling). For information, most trays in the ESGR are uncovered. Most trays in the CVf are covered. In the tunnel part, which is a critical location, the trays are stacked 8 deep on each side. The bottom 6 trays on the H side and 5 trays on the J side are covered. These trays carry mostly control cables for the valves (going from the vault MCCs down into the ESGR). The upper 2 trays on the H side and 3 trays on the J side are uncovered. The upper two trays on either side carry power cables for the pumps. In the containment (CI), most trays are covered; however, the covering is mostly on the top (possibly to prevent water collection especially after a spray actuation). The cable in conduit cannot be traced from the engineering drawings used for this purpose (It would be very difficult to trace conduits, as apparently greater latitude was allowed in their placement). An assumption is made in this case at places the conduit in the worst location (within reason-see below); there are not many cases like this. Conduit ds to be used in the containment for the final spur between the cable tray and the component in question. Of the mportant systems considered and in the significant areas considered, conduit is used for LPI A pump power cable in the ESGR and the CVf (and on in the safeguards area); charging pump service water pump lOA power and control cables in the ESGR; AFW A pump in the ESGR and the CVf (and on in the safeguards area); and charging pump A in the auxiliary building. The conservative assumption would be to place the conduit next to another train, and this was done when it seemed reasonable based 9n known information. For instance, charging pump A conduit was placed next to charging pump C cable, for a certain distance, in the auxiliary building (which is not an important area). However, in some other cases, it was believed reasonable that an 'A' train conduit would be placed in the H part of a fire area, not the J part (if the two areas are distinctly separate, e.g., in the tunnel). This assumption will not necessarily have a large impact on the core damage frequency, based on the scenarios considered. Modeling of switchgear panel fires by COMPBRN showed that no damage would result to the cables nearby as long as the flames did not penetrate the panel casing. However, according to the Sandia cabinet fire test results1, in a small fraction of fires, the flames were observed to shoot 50 cm horizontally from the cabinet. (The tests also show that the fire doesn't propagate between double walled cabinets of the type used at Surry.) COMPBRN results show that cables immersed in flames will suffer damage. (This seems to be almost a necessary and sufficient condition for cable damage). The panels are 7.5 to 8.5 ft high. Based on COMPBRN results, all the cables in the ESGR within the 50 cm horizontal distance of any panel are within reach of a small fire placed near the top of the cabinet. Therefore, the assumption is that a large panel fire (one that penetrates the casing) will damage all cables within the 50 cm horizontal distance from the panel. If a critical cable tray is not close to the panel, it is usually very far away such that propagation of fire along cable trays (from the burning panel) would lead to very long damage times and small damage .ctions, hence, this kind of propagation can be screened out., Another assumption is that such a fire will destroy the 5-1 NUREG/CR-6144

Scenario Development whole panel (i.e., all cabinets within the panel). A small fire will destroy only the cabinet of origination and the - of the panel and the cables nearby will be intact. (As for the CVT panels, i.e., the MCCs in the penetrations v a ' P the cable trays for systems not directly associated with these panels are horizontally outside the zone of influence of even a large panel fire). A conservative assumption was made in the tunnel area where all combinations of cable fires were rolled into the worst case. The worst case is when the fire is initiated in the lowest tray; this causes damage to all the trays on both sides of the tunnel. However, a fire in an upper tray will not likely damage the trays below, but in order to limit the number of scenarios it was assumed that all initiation locations were equivalent. It should be noted that this is a relatively good assumption, as the upper two trays on either side carry pump power cables, while the lower trays carry valve control cables. Loss of pumps is irrecoverable, whereas most valves can be manually operated. One explanation will help the reader better understand certain scenarios in the ESGR and certain aspects of tracing cables there. The important systems cables are pretty much grouped together (horizontally) within certain passageways (see Figure 5.1). The important trays within these passageways are stacked vertically, but in such a way that they are, for all systems, within reach of a large fire on the floor ( except for some stretches of cable for occasional systems). Of course, as explained above, they are also within reach (vertically) of large panel fires within the critical horizontal distance. Some important systems' cables ( e.g., both RHR pumps power) are also within reach of even a small oil fire on the floor. Since most ESGR trays are in this elevation band, conclusions can be reached for systems with unknown cable routing. Certain systems were not included in the analysis and/or cable tracing:

1) EDG cables were not traced because this information is irrelevant; the EDGs would be used only in case of of power. If the.loss of power was caused by a large fire in the ESGR, such a fire would destroy the originat bus panel, therefore recovery via EDG would be impossible. A small fire (i.e., in the supply breaker that would cause a demand for the EDG) would not damage any nearby cables.

2) Circulating Water (CW) condenser inlet/outlet valve cables were not traced because this information would be important only for a (LOOP) initiator. Such an initiator can only be caused by a large fire in the NSGR. This event will not damage (CW MOV) cables, which pass through the ESGR, 40 ft below.

3) Service air cable routing was not considered because this system is powered from the NSGR 480 V buses 1C2 (in Unit 1 NSGR) and 2A2 (in Unit 2 NSGR). (The outage compressors are also powered from the NSGR and there is a diesel compressor.) The backup instrument compressors are powered from the MCCs in the CVT.

A single fire would not disable all these locations. Disregarded due to probabilistic concerns, system impact of likely initiators and system configuration. The likely fire-induced initiators in the NSGR (LOOP, loss of a bus) will already include (irrecoverable) loss of part of the system in their impact; the fact that a cable may have been damaged is immaterial.

4) The ESGR cooling system was not considered due to redundancy in the system.

5) Process instrumentation cables were not traced; however, as per discussion of ESGR cable layout above, one has a pretty good idea of scenario impact on this function. As all the trays in the tunnel scenario (the only one in the CVT) are damaged, the impact on instrumentation is known there also.

NUREG/CR-6144 5-2 **

Scenario Development . 6 ) SG PORV control cables were not traced, also, the precise routing of supply to the semi-vital bus (controlling the PORVs) from the ESGR MCCs to the MCR is not known. In the earlier phase of this project the understanding was that the SG safeties are always open in mid-loop, so there was no need to trace the PORV cables. While this proved incorrect, the same comment as in 5) applies to this system. It was assumed that a fire in either H or J room and along the probable routing path will disable this system.

7) Relay fires were screened based on probability and system impact.

8) CC valve 109A and B cables were not traced as the information was not available at the time of analysis. It is expected that impact will be minimal.

9) Only one train of Unit 1 CCW, charging pump cooling and charging pump service water is usually operating in mid-loop. The assumption is that it is train A Likewise, it is assumed that train A of RHR and charging is the operating train at the time of fire.

10) When the Reactor Water Storage Tank (RWST) suction MOVs to the charging system are disabled (due to damage to control or power cables to both MOVs), and if the control cables to charging pump A are damaged, this represents an irrecoverable failure of charging pump A Other assumptions and simplifications are described in Chapter 2.

5.2 Initiators Considered e initiators considered are RHR3, RHR4, RHR5 (irrecoverable loss of RHR, irrecoverable loss of operating RHR train, recoverable RHR loss), Ll (type of loss of offsite power), Bl (Unit 1 blackout), 4 kV (loss of lH bus) and VITAL (loss of vital bus). Others were screened out due to probabilistic concerns, or redundancy in widely separated systems or based on importance of initiator from the internal PRA. 5.3 Important Equipment and Cable Locations Figures 5.1, 5.2, and 5.3 show the juxtaposition of important cables and equipment in the ESGR, CVT, and CT, respectively (See also the discussion in Chapter 2). Table 5.1 presents summarized cable routing information for important components, while Table 5.2 presents cable insulation information. Appendix D contains figures showing cable tray layout in the areas considered. It should be noted that the floor elevation in the ESGR is 9'6". COMPBRN predicts damage to cable trays up to elevation of 22'0" from a large fire and 20'6" from a small fire (and all trays above the initiating tray in a cable fire). The floor elevation of the CVT is 13'0". COMPBRN predicts damage to all cables in the tunnel from a large oil fire anywhere on the tunnel floor or a cable fire in the lowest tray. Important CT floor elevations are 18'4" and -13'0" (RHR flats). COMPBRN results for the ESGR also hold true here (with respect to the vulnerable tray elevation off the floor). 5.4 Types of Fire Damage to Components A fire can damage control and/or power cables for valves and pumps. It can disable instrumentation. Certain electrical failures can propagate between panels.

5-3 NUREG/CR-6144

Scenario Development Spurious operation of components can happen due to hot shorts, when the fire occurs in the control room, at MCC, or bus cabinet, or along the routing of the control cable. Isolation switches are provided to protect certain components from this phenomenon for fires in select locations. For instance, hot short switches are in the ESGR for protection against fires in the MCR. Disabling a power cable for an MOV will not cause failure if the valve is in the correct position; e.g., MOV-1700 (RHR RCS isolation valve) is open while the RHR is running, so cutting off power to this valve will not disable the RHR system. However, if the valve is not in the correct position for the intended operation of the system (e.g., charging suction from the RWS1), damage to the power cable will disable the pump. This is recoverable, as the operator can manually operate the valve. Damaging a valve control cable will disable it if it is not in the correct position. As before, this is recoverable, after power is disconnected from the motor. If the control cable is for a valve that is in the correct position, there will likely be no change in the system status. However, in a certain fraction of cases, spurious operation of the valve due to hot shorts will result. This will disable the valve, recoverably as above. For a pump, damage to the power cable will disable it irrecoverably. (Repair takes too long). Damage to the control cable will disable it irrecoverably if the pump is idle. (It cannot be started). If the pump is running, it will likely continue to run. However, in a certain fraction of cases, spurious operation (i.e., tripping) of the pump due to shorts results, and the pump is lost permanently. The layout of control circuitry was considered in order to evaluate the probability of hot shorts causing spurious operation given damage to the control cable. The analysis is given in Appendix E. While the control circuitry for different valves may have slightly different arrangements, the result is that the conditional probability of hot short induced spurious operation is on the order of 0.1. The same is assumed to hold true for pumps. This is based on a combinatorial analysis alone and does not include other effects, e.g., the probability that the contact will be made or that it will be "good". The number derived will be used for spurious operation of all components. In case of a loss of instrumentation cables, redundant process monitoring is provided at the remote monitoring panels outside the control room. These have cable routing independent of the ESGR and CVf. At Surry, fire- induced electrical faults can propagate in certain situations. Appendix R analysis identifies lack of electrical coordination between the 480 V panels lH (lJ) or lHl (lJl) and the corresponding 4.16 kV bus lH (lJ) feeding them2

This means that a fire at one of the 480 V buses which causes (or is caused by) an electrical short will cause the 4 kV bus supply breaker to open before the fault can be isolated at the 480 V level. Therefore, a loss of 4 kV bus will result.

S.S Important Locations and Scenarios Within Fire Zones 5.5.1 ESGR NUREG/CR-6144 5-4

Scenario Development

     . Large fire at 4 kV lH bus panel. The fire destroys the whole panel and nearby cables. Initiator: 4 kV (loss of lH 4 kV emergency bus). No recovery is possible.

System impact: irrecoverable loss of all lH loads, including RHR pump lA and charging service water pump lOA. Possible spurious operation of MOV-1700 due to control cable damage would disable both RHR trains. SG PORVs are also assumed to be lost. Recoverable loss of LPI MOV 1890C (cold leg injection path). Loss of AFW pump 3A and valves 151A, C and E (control cable damage. Loss of recirculation spray train A, both inside and outside). E2. Large fire at the 480 V panel lHl and 4 kV lH stub bus. The fire destroys the whole panel and nearby cables. Initiator: 4 kV (loss of lH 4 kV bus). This is caused by lack of coordination between the 480V lH and 4 kV lH buses; this allows an electrical fault to propagate to the 4kV bus, where the supply breaker will open to protect the upstream components. Unlike El, this loss of 4 kV bus can be recovered. System impact: irrecoverable loss of RHR train A, CC pump A, LPI pump A, charging pump lA and 1C, and charging pump service water pump A Possible spurious operation of MOV 1700 and 1720A. SG PORVs and LPI MOV 1890C are disabled. AFW valves 151A, C, and E are disabled Recirculation spray train A (both inside and outside systems) are disabled. E3. Large fire at the 480 V panel lH. The fire destroys the whole panel and nearby cables. Initiator: same as in E2, due to lack of electrical coordination between 480V bus lH and 4 kV bus lH. System impact: same as in E2. E4. Large fire at 480 V lJl panel and the 4 kV 1J stub bus. The fire destroys the whole panel and nearby cables. Initiator: RHR3 (irrecoverable loss of RHR). This is caused by loss of RHR B train in combination with other RHR "lures (spurious operation of pump A, or MOV 1700, or MOV 1701, or loss of HCV-1758 instrumentation followed failure of operators to control flow such that pump cavitation is prevented). stem impact: recoverable loss of 1J 4 kV bus due to the coordination problem. Permanent loss of CC pump B, LPI pump B, charging pump service water pump B, LPI pump A, charging pump B and C, and charging pump A (due to loss of RWST suction and damage to its control cable). Possible spurious operation of CC pump A Possible spurious operation of the pressurizer block valves. Possible spurious operation of LPI RWST suction valves. SG PORVs disabled. Loss of process instrumentation forces the operators to rely on remote monitoring panels. All Unit 1 AFW is lost (pumps 3A and 3B and valves 151A, B, C, D, E, Fare all disabled due to control cable damage). ES. Large fire at 480 V 1J panel. The fire destroys the whole panel and nearby cables. Initiator: RHR3 (irrecoverable loss of RHR). Caused by the same type of failures as in E5. System impact: recoverable loss of the 4 kV 1J bus due to coordination problems. All methods of forced primary feed and bleed from Unit 1 are lost permanently. SG PORVs disabled. Possible spurious operation of pressurizer block valves. Loss of process instrumentation forces the operators to rely on remote monitoring panels. All Unit 1 AFW is lost as in E4. E6. Large fire at MCC lJl-1. The fire destroys the whole panel and nearby cables. Initiator: RHR3 (irrecoverable loss of RHR). This is caused by loss of train B (due to control cable damage) in combination of various failures causing the loss of train A These failures are: spurious closure of valve MOV-1700, spurious stopping of pump lA, or operator failure due to loss of valve HCV-1758. System impact: all important controls for the H division are lost. This means loss of charging pump C (possible spurious operation of pump A), LPI pump A, AFW pump A, and loss of H division valves for these systems. In addition, both charging pump service water pumps will be lost, as well as CPC pump 2B. This will indirectly cause

of all charging. In addition, recirculation spray train A will be lost, and the pressurizer PORVs will he lost.

5-5 NUREG/CR-6144

Scenario Development E7. Large fire at UPS lB-1. The fire destroys the whole panel and nearby cables.

Initiator: VB (loss of vital bus). This is caused by similar failures as in E6; in addition, spurious closure of MOV-1701, or simultaneous spurious closure of MOV-1720A and 1720B will also contribute to the initiator.

System impact: AFW pump 3A is lost (due to control cable damage) as areAFW valves 151B, D and F (for the same reason). Control is lost to the pressurizer PORVs, and the block valves may actuate spuriously. Recirculation sprays are lost. Charging pump cooling pumps 2A and 2B are lost (control cable damage) as is charging pump service water pump 10B. Charging MOVs 1115C and E are disabled. E8. Very large fire at any of the panels causes loss of all switchgear equipment in both Hand J compartment of the ESGR, due to interaction of smoke with switchgear. Initiator: Unit 1 station blackout, without possibility of recovery, as 4 kV breakers are damaged, and neither the diesel generator nor the offsite power can be connected to either 4 kV bus. System impact: All methods of cooling are permanently lost, except for Unit 2 charging cross-connect. Gravity feed can be recovered, but no recovery is possible post gravity feed. Possible loss of all process instrumentation forces the operators to use remote monitoring panels. E9. Transient or cable fire in the H room, along the critical path. Initiator: RHR4 (irrecoverable loss of RHR train A) due to spurious operation of pump A (control cable damaged), or RHR5 (recoverable loss of RHR) due to spurious operation of MOV 1700 (control cable damaged). System impact: LPI pump A and charging pump C disabled (due to damage to control cables). SG PO RVs disabled. LPI cold leg injection MOV-1890C disabled. Charging pump service water pump 10A disabled (power cable damage). Charging pump RWST suction MOV 1115B disabled. Possible spurious operation of charging A and/or CCW A pump. Possible spurious operation of charging pump suction and/or discharge MOVs (this disables charging tra* irrecoverably). AFW pump 3A and MOVs 151A, C, and E are lost due to control cable damage. Train A of i and outside recirculation sprays is lost. ElO. Transient or cable fire in the J room along the critical path. Initiator: RHR3 (irrecoverable loss of both trains of RHR due to spurious operation of MOV 1700, or MOV 1701, bo~h with loss of control cables for RHR pump A and B). System impact: SG PORVs disabled; both LPI pumps disabled; charging pumps Band C disabled; CCW pump B disabled; charging RWST suction MOVs disabled. Charging pump A disabled (control cable damaged and RWST suction disabled). Possible spurious operation of CCW pump A Possible spurious operation of pressurizer PORV block valves. Unit 1 AFW is lost due to control cable damage for pumps 3A and 3B and all six 151 MOVs. Ell. Small fire at 4 kV lH stub bus panel; a fire at any of the three cabinets (one of which is supply to RHR pump A) will lead to irrecoverable loss of one train of RHR. Initiator: RHR4 System impact: irrecpverable loss of CCW pump A E12. UPS fire (large fire at UPS-lA-1 or UPS-lA-2, both located in the H room, or a small fire at UPS-lB-1 located in the J room). Initiator: VB (loss of vital bus) No other impacts El3A Small fire in a 480 V cabinet in the H room. Initiator: 4 kV (loss o{ lH 4 kV bus). Recovery by diesel is possible. The bus is lost due to lack of electrical coordination: NUREG/CR-6144 5-6

Scenario Development System impact: Loss of all lH loads. LPI pump A and inside and outside recirculation sprays train A are lost permanently. E13B. Small fire in a 4kV cabinet in the H room. Initiator: 4 kV (loss of lH 4 kV bus). Recovery is not possible due to electrical fault. System impact: Loss of all lH loads. 5.2.2 CVT CVl. A cable fire or a large transient fire in the tunnel/service building vault area destroys all the cables there (on , both sides, i.e. both the H and J trains). Initiator: RHR3 (irrecoverable RHR loss due to damage to pumps A and B power cables). System impact: both trains of LPI disabled. Charging pump B disabled. CCW pump A disabled. SG PORVs disabled. Process instrumentation disabled, causing operators to rely on remote monitoring panels outside the control room (in the cable spreading room). MO Vs for charging alignment to RWST disabled. Possible spurious operation of charging suction/discharge valves for all pumps. Possible spurious operation of pressurizer PORV block valves 1535 and 1536. All Unit 1 AFW is lost (pumps 3A and 3B have damage to the power cables, while valves MOV 151A, B, C, D, E, F are disabled due to control cable damage). Inside and outside recirculation sprays are lost. eenedout due to low frequency: Large fire at MCC-lHl-2 that damages nearby cable trays and destroys the MCC ntrolling A train valves). The trays are well outside the range (horizontally) of the casing-penetrating large fires served in the Sandia cabinet fire tests. Screened out due to low frequency: a cable fire in the safeguards penetration area knocks out RHR pump lA and both LPI pumps. Similar reasoning applies to a fire in the 1 containment penetrations area. '1 5.5.3 CT Cfl. A cable or a transient fire along the cable tray holding power cables for both RHR pumps. Initiator: RHR3 (irrecoverable loss of RHR). System impact: no other impacts, except for loss of power to RHR/RCS isolation MOVs. Cf2. A fire at one of the two RHR pumps. Such an event will likely just disable the pump, as the tray CA (carrying the final cable run to the pumps) is too high above the pumps. It is not known whether the cable from the tray to the pump is in a conduit; if it isn't, a fire can propagate along the cable to the cable tray, and possibly disable the other pump also. A fire at pump A (assumed to be the running pump), were it to reach the tray, would have to travel several feet along the tray in order to reach pump B cable. It should be noted that there is a radiant energy shield between the two pumps. A fire at pump B would much more easily disable pump B, assuming it reaches tray CA, as both pumps' cables are routed together in the vicinity of pump B. While a pump B fire would have a more serious impact, it would also have a lower frequency, as the pump is idle (There have been pump fires set by welding sparks, however). Conservatively, it is assumed that a fire at either pump (at the same RHR pump fire frequency) will disable both RHR trains.

  "tiator: RHR3.

5-7 NUREG/CR-6144

Scenario Development System impact: none. The same HEPs as for the internal PRA will be used, hence the same conditional core damage probability will result. 5.5.4 NSGR Nl. Small fire at 4 kV transfer bus F, supply breaker for the 1H bus. Initiator: 4 kV (loss of 4 kv 1H bus) System impact: This will cause loss of 1H AND 2J (Unit 2) buses (under normal configuration). Recovery of 1H possible from EDG. Recovery from reserve station transformers is prohibited cir from service transformers - same as E13. N2. Large fire at 4 kV transfer bus F (destroys the whole panel). Initiator: 4 kV (loss of 4 kV 1H bus) System impact: this event causes loss of transfer bus F and normal bus 1A, so recovery via backfeed is prohibited, as well as from reserve station transformers. The loss caused would be to (Unit 1) 1H bus and (Unit 2) 2J bus. Recovery of 1H bus possible via EDG#l. N3. Very large smoky fire in the NSGR disables all switchgear equipment there. Initiator: Loss of offsite power (irrecoverable). System impact: recovery of offsite power prohibited. However, recovery via diesel generator is allowed. 5.6 Scenario Quantification Table 5.3 shows the results of fire scenario quantification (per plant year). These numbers include the mid-loop POS fraction. The values include the nonsuppression fraction (presented in Chapter 8). The scenario quantification (i.e., initiating event frequency) for each window and POS combination is given in Table 5.4. (These values also include the nonsuppression fraction.) Table 5.4 also includes quantification of scenarios in the main control room, developed in Chapter 9. These are scenarios Ml-M13 and MA-MH Note that development and quantification of fire scenarios in the control room is presented in Chapter 9. Appendix K shows the algebraic expressions used in calculating the fire scenario initiating event frequencies and uncertainty distributions. Appendix L shows the fire scenario initiating event histograms (i.e. uncertainty distributions) 5.7 References

1. J.M. Chavez, "Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets, Part 1: Cabinet Effects Tests," NUREG/CR-4527-Pl, Sandia National Labs, Nov. '86.

2. BNL Memorandum, dated July 9, 1991, from R.G. Fitzpatrick to T-L. Chu and N. Siu, titled "Review of Surry Appendix R Report (Chapter 9)."

NUREG/CR-6144 5-8

          ~
                                                                             \
            ~                                              /                                       Critical Tray iV.         Location UOORS
                       >      A"

u,..ilT I CADLE VAlA. T/TUH:L 1-J FIRE AllfA I

J 6 5
-.cfl...a .. naa CPI. IUCI di . . . . . . *.1.- ll"*r I ff CwtCIMII**
(
--11 .***.
PLAN EL. IS' *C"

Figure !.3 Containment Layout at 18'4" Showing Critical Tray Locations 5-11
ua.ss LllDO'l"I 1---"=-+.:...-ff--.--.;....+-:r~ ;!t.!J!~rL
DETACHED PLAN EL -13'-0
Figure 5.3a Containment Layout at RHR Flats Showing Critical Tray Locations 5-12 *
Table 5.1 Room (Floor Elevation)
Scenario Development System Component ESGR (9'6") 1 CVT (13'0") 1 CT (18'4", 13'0") 1 (train) tray height tray height tray height RHR HCV-1758 A17 18'7" Al 17'9" Al 25'8" (instrument) 17'7" A5 17'4" CND2 to- 11'6" 18'7" RH-P-lA (H) BlO 20'6" B2 22'4" B 28'8" (power) Bl 23'4" 15'0" 31'2" 1'0" 7'9" RH-P-lB(J) BlO 19'7" B3 23'4" B 28'8" (power) 20'6" 32'10" 15'0" l'O" 9'9" RH-P-lA (H) la161 22'7" Control la16 21'7" 23'0" RH-P-lB (J) Cll 19'7" Control MOV-1700- (H) C14 23'4" C 27'8" Power 22'6" 15'0" (also MOV 1720A 23'0" O'O" power) C7 16'4" 8'9" 25'4" CND 8'9"
-1'6" MOV-1700 (H) ClO 23'6" C14 23'4" Control 22'0" 22'6" la16 21'7" C7 16'4" 23'0" MOV-1720A (H) ClO 23'6" C14 23'4" Control 22'0" 22'6" la16 21'7" Cl 21'4" 23'0" MOV-1701 (J) C15 23'4" C6 63'6" (also MOV-1720B) 21'4" CNDto 13'0" (power) C6 15'4" 25'4"
5-13 NUREG/CR-6144
Scenario Development Table S.l Room (Door Elevation) (continued) System Component ESGR (9'6") 1 CVT (13'0") 1 CT (18'4", 13'0") 1 (train) tray height tray height tray height
.RHR MOV-1701 (J) ClO 23'6" C15 23'4" (also MOV 1720B) 20'7" 21'4" (control) 24'0" C6 15'4" 20'0" 16'4" C8 20'9" PPRS MOV-1535 (H) C14 22'6" C 27'8" Power 23'4" CNDto 61'3" C7 16'4" 25'4" MOV-1535+ (H) ClO 23'6" C14 23'2" Control 20'T' 23'0" la162 18'0" C7 16'4" la161 22'T' MOV-1536 (J) C15 23'4" C2 30'8" Power 21'4" 36'3" C6 15'4" 43'0" 25'4" CNDto 61'3" MOV-1536+ (J) ClO 23'6" C15 21'4" Control 20'T' 23'4" 24'0" C6 16'4" 20'0" C8 20'9" HPIA MOV-1115D (J) & C15 23'4" MOV-1115E 21'4" Power MOV-1115D (J) ClO 23'6" C15 21'4" Control la15 22'0" 23'4" C6 16'4" C8 20'9" MOV-1115C (H) & C14 22'6" MOV-1115B 23'4" Power 21'4" 23'0" MOV-1115C (H) & ClO 23'6" C14 23'0" MOV-1115B a16 23'0" C7 16'4" Control 22'T' 23'0" NUREG/CR-6144 5-14
Table 5.1 Room (Floor Elewtion) (continued)
Scenario Development System Component ESGR (9'6") 1 cvr (13'0") 1 CT (18'4", 13'0)1 (train) tray height tray height tray height HPI MOV-1115E (J) ClO 23'6" C15 21'4" Control 20'7 23'4" 24'0" C6 16'4" 20'0" C8 20'9" MOV-1269A (J) C15 23'4" Power++ 21'4" MOV-1269A (J) ClO 23'6" C15 21'4" Control++ la15 22'0" 23'4" C6 16'4" C4 20'4" C8 20'9" MOV-1267A (H) C14 22'6" Power* 23'4" 21'4" 23'0" MOV-1267A (H) ClO 23'6" C14 23'0" Control* la16 03'0" C7 16'4" 22'7 23'0" CH-P-lA (H) ClO 23'6" Control & lal6 23'0" CH-P-lC (H) CH-P-lB (J) BS 22'4" Power CH-P-lB (J) ClO 23'6" Power la15 22'0" HPI 1-SW-p-lOa (h) CND Cooling' Power, control l~SW-P-lOb (J) la15 25'3" Power 22'7 ClOl 19'6" 22'9" 1-SW-P-lOb (J) la15 22'7 Control 21'7 22'0" 5-15 NUREG/CR-6144
Scenario Development Table 5.1 Itoom (Floor Elewtion) (continued) System Component ESGR (9'6")1 cvr (13'0")1 CT (18'4", 13'0")1 (train) tray height tray height tray height LPI0 MOV-1862A (H) & C14 23'4" MOV-1860A Power MOV-1862A (H) & la16 23'0" C14 23'0" MOV-1860A C7 16'4" Control C20 23'6" 20'7" MOV-1860B (J) & C15 23'4" MOV-1862B 21'4" Power C2 20'4" Cl 31'2" MOV-1860B (J) & ClO 20'1" C15 23'4" MOX-1862B 20'0" 21'4" Control C6 16'4" C8 16'4" 20'9" SI-P-lB (J) BS 22'4" Power 32'0" SI-P-lB (J) Cll 19'7' Control SI-P-lA (H) CND CND Power SI-P-lA (H) la16 23'0" Control ccwd 1-CC-P-lB BlO 20'6" Power 1-CC-P-lB Cll 19'7' Control 1-CC-P-lA BlO 20'6" B2 22'4" Power BS 22'4" 1-CC-P-lA la16 20'7' 21'7' 23'0" AFW ID MOV MS-102 C14 23'4" Control C3 20'4" NUREG/CR-6144 5-16
Scenario Development Table 5.1 Room (Floor Elewtion) (continued) System Component ESGR (9'6") 1 cvr (13'0") 1 CT (18'4", 13'0") 1 (train) tray height tray height tray height AFW MS-102BPWR A 172 16'7" C8 20'9" cs 16'4" C7 16'4" C2 21'4" C2 20'4" C2 31'3" C12 MDP3BPWR B2 22'4" MDP3APWR CND CND MDP 3B Control ClO 20'7" ClO 24'0" Cll 19'7 MDP3A Control A16 20'7 21'0" 23'0" MOV 151B, D,F C15 23'4" Power C6 15'4" C6 63'6" C6 25'4" CND MOV 151B, D,F ClO 23'6" C6 15'4" Control 24'0" C8 20'9" 11:./. 20'0" short Section MOV 151 A,C,E C14 23'4" C 27'8" Power C7 16'4" CND C7 25'4" MOV 151 A,C,E ClO 23'6" C7 16'4" Control A16 22'7" CND C7 25'4" + Note: Control cables for 'H' train components first run through the 'H' room in the ESGR, then through the 'J' room, whereas the 'J' train control cables are wholly contained in the 'J' room. Hence, tray ClO for MOV 1535 is in the 'H' room, whereas tray ClO for MOV 1536 is in the 'J' room. a HPI routing in the auxiliary building not shown here ++ All other suction/discharge MOVs for CH-P-lB have the same cable routing.
All other suction/discharge MOVs for CH-P-lA & CH-P-lC have the same cable routing.
5-17 NUREG/CR-6144
Scenario Development Table 5.1 Room (Floor Elevation) (continued) b Charging pump component cooling not needed for RWST water injection; cable routing through Unit 2 ESGR and beyond not shown here. c Cable routing through the safeguards building not shown here. d This table does not show routing into the auxiliary building. 1 Floor elevations of mterest in the fire area. 2 CND is a conduit, indeterminate routing. NUREG/CR~6144 5-18
Scenario Development Table S.2 Cable Insulation Types Used at Surry Unit 1 for Select Systems Cable Types+
System Power Control Insulation (Mills)* Total Cable Insulation Total Thickness (Mills)* Cable (in) Thickness (in) AFW 45pei-45np .58 45pei-80np .90 llOpei-Alintlk 1.39 45pei-60np .75 ' 110pei-60np 1.39 ccw llOpei-Alintlk 1.48 45pei-80np .90 RHR 25pei-45np .375 llOpei-Alintlk 1.39 45pei-60np .75 45pei-45np .58 45pei-80np .90 PPRS 45pei-45np .58 47 thw 45pei-80np .90 Charging 45pei-45np .58 45pei-80np .90 llOpei-Alintlk 1.48 45pei-60np .69 47thw LPI 80pei-Alintlk 2.72 45pei-60np .69 80pei-65np 2.70 45pei-60np .75
+ Explanation of symbols:
pei polyethylene crosslinked np - neoprene thw - thermal wrap Alintlk - Aluminum intlk. armor
the number preceding the symbol is the thickness of that layer in mills (innermost layer first).
5-19 NUREG/CR-6144
Scenario Development Table 5.3 Fire Scenario Quantification Scenario Initiator Frequency+ (/yr) El 4kV 2.8-6 E2 4kV 7.8-6* E3 4kV 7.8-6* E4 RHR3 6.0-6* E5 RHR3 6.0-6* E6 RHR3 5.3-7 E7 VB 8.0-6 E8 Bl 3.4-7 E9 RHR4 2.8-7** ElO RHR3 7.8-7** Ell RHR4 7.4-6 E12 VITAL 8.8-5*,a E13A 4kV 5.5-5 E13B 4kV 1.7-5 Nl 4kV 1.7-5 N2 4kV 8.4-6 N3 L1 3.4-7 CVl RHR3 8.1-6b en RHR3 1.2-5c Cf2 RHR3 2.9-4
+ Multiply by 0.26 for R6, 0.23 for RlO, 0.51 for D6.
Stepdown transformer treated as a contributor
** Approximate length of cable -40 ft. in either H or J room; 4 trays stacked in the H room, 11 trays in the J room. In either room the trays with RHR Cables are out of reach of a small oil fire. Zone of influence of large oil fire 3 ft on either side of the tray. Trash fires treated as oil fires. Probability of spurious operation is 0.1.
a Two transformers per UPS, treated as separate contributors. b Trash fires treated as oil fires. Small fire scenario much milder and of much smaller frequency, so it was neglected. NUREG/CR..:6144 5-20
Scenario Development c Trash fires treated as oil fires. No credit for grating surface. Both small and large fires can reach a 51 ft section of cable tray. Fires cannot develop on one side of the tray, as the tray is next to the crane wall. 5-21 NUREG/CR-6144
Table !.4 Initiating Event Frequencies (/yr) for the Scenarios vs. Pos*Wlndow Combinations Windows-POS El E2 E3 E4 E5 E6 E7 EB E9 ElO Scenario : W1D6 1.7-7 4.8-7 4.8-7 3.7-7 3.7-7 3.3-8 4.8-7 2.1-8 1.7-8 4.6-8 W1R6 1.2-8 3.3-8 3.3-8 2.5-8 2.5-8 2.2-9 3.3-8 1.4-9 1.2-9 2.9-9 W2D6
6.4-7 1.8-6 1.8-6 1.4-6 1.4-6 1.2-7 1.8-6 7.9-8 6.3-8 1.7-7 W2D6 3.7-7 1.1-6 1.1-6 8.1-7 8.1-7 7.1-8 1.1-6 4.6-8 3.7-8 1.0-7 W3D6 5.5-7 1.5-6 1.5-6 1.2-6 1.2-6 1.0-7 1.6-6 6.8-8 5.4-8 1.5-7 W3R10 1.0-8 2.9-8 2.9-8 2.2-8 2.2-8 1.9.9 2.9-8 1.3-9 1.0-9 2.7-9 W3R6 2.8-7 7.9-7 7.9-7 6.1-7 6.1-7 5.4-8 7.9-7 3.5-8 2.8.8 7.6-8 Ul W4D6 1.1-7 2.9-7 2.9-7 2.3-7 2.3-7 2.0-8 3.0-7 1.3-8 1.0-8 2.8-8
~ W4R10 6.3-7 1.8-6 1.8-6 1:4-6 1.4.6 1.2-7 1.8-6 7.7-8 6.2-8 1.7-7 W4R6 2.3-8 6.6-8 6.6-8 5.1-8 5.1-8 4.5.9 6.6-8 2.9-9 2.3-9 6.3-9
Table !.4 (Continued) Initiating Event Frequencies (/yr) for the Scenarios '18. Pos-Wlndow Combinations Windows-POS Ell E12 E13A E13B Nl N2 N3 en Cf2 CVl Scenario W1D6 4.5-7 7.2-6 3.4-6 1.1-6 1.1-6 5.2-7 2.1-8 7.2-7 1.8-5 5.0-7 W1R6 3.1-8 4.9-7 2.3-7 7.2-8 7.2-8 3.5-8 1.4-9 4.9-8 1.2-6 3.4-8 W2D6 1.7-6 2.7-5 1.3-5 3.9-6 3.9-6 1.9-6 7.9-8 2.7-6 6.7-5 1.9-6 W2R6 9.9-7 1.6-5 7.3-6 2.3-6 2.3-6 1.1-6 4.6-8 1.6-6 3.9-5 1.1-6 W3D6 1.5.-6 2.3-5 1.1-5 3.4-6 3.4-6 1.7-6 6.8-8 2.3-6 5.7-5 1.6-6 VI W3R10 2.7-8 4.3-7 2.0-7 6.3-8 6.3-8 3.1-8 1.3-9 4.3-8 1.1-6 3.0-8 ~ W3R6 7.5-7 1.2-5 5.5-6 1.7-6 1.7-6 8.5-7 3.5-8 1.2-6 3.0-5 8.2-7 W4P6 2.8-7 4.4-6 2.1-6 6.5-7 6.5-7 3.2-7 1.3-8 4.4-7 1.1-5 3.1-7 W4R10 1.7-6 2.7-5 1.2-5 3.8-6 3.9-6 1.9-6 7.7-8 2.7-6 6.6-5 1.8-6 W4R6 6.2-8 9.9-7 4.6-7 1.4-7 1.4-7 7.1-8 2.9-7 9.9-8 2.5-6 6.8-8
Table 5.4 (Continued) Initiating Event Frequencies (/yr) for the Scenarios vs. Pos-Wlndow Combinations Windows-POS Ml M3 M4 MS M13 MA MC MG MH Scenario W1D6 8.3-8 6.1-8 1.0-8 7.0-9 3.0-9 1.1-6 2.4-8 4.4-7 5.6-8 W1R6 5.7-9 4.2-9 6.8-10 4.8-10 2.1-10 7.5-8 1.7-9 3.0-8 3.8-9 W2D6 3.1-7 2.3-7 3.7-8 2.6-8 1.1-8 4.1-6 9.1-8 1.6.6 2.0-7 W2R6 1.8-7 1.3-7 2.2-8 1.5-8 6.6-9 2.4-6 5.3-8 9.5-7 1.2-7 W3D6 2.7-7 2.0-7 3.2-8 2.3-8 9.8-9 3.5-6 7.8-8 1.4-6 1.8-7 VI ~ W3R10 5.0-9 3.6-9 6.0-10 4.2-10 1.8-10 6.5-8 1.5-9 2.6-8 3.3-9 W3R6 1.4-7 1.0-7 1.7-8 1.2-8 5.0-9 1.8-6 4.0-8 7.2-7 9.2-8 W4D6 5.1-8 3.7-8 6.2-9 4.3-9 1.9-9 6.7-7 1.5-8 2.7-7 3.4-8 W4R10 3.1-7 2.2-7 3.7-8 2.6-8 1.1-8 4.0-6 8.9-8 1.6-6 2.1-7 W4R6 1.1-8 8.3-9 1.4-9 9.6-10 4.2-10 1.5-7 3.3-9 6.0-8 7.6-9
6. FIRE GROWTH ANALYSIS 6.1 COMPBRN-Ille Capabilities and Limitations The COMPBRN-IIIe fire growth code1 was used to calculate fire growth and cable damage. COMPBRN-IIIe is an updated version of COMPBRN-II12 and is developed as a menu-driven interactive computer program.
The physical models used in COMPBRN-IIIe are the same as that in COMPBRN-III, except for a few modifications related to radiation heat transfer. In addition to the point estimate analysis contained in COMPBRN, the updated version of Ille has an uncertainty analysis module and can assist the users in performing a probabilistic fire risk analysis directly. COMPBRN uses a quasi-static approach to simulate the process of fire growth in an enclosure. The enclosure is divided into two distinct homogeneous, stably stratified regions. The hot gases accumulating under the ceiling due to fire plume entrainment and buoyancy are defined as the upper layer. Elements in this layer are heated by convective heat flux from the hot gases. The lower region is assumed to be thermally inert and contains relatively quiescent cool air, which remains at ambient conditions at all times. The burning rate of a fuel element is used to determine the heat output of the element, which is transferred to other objects via radiation. An element is considered damaged or ignited if its surface temperature exceeds the user-specified damage or ignition temperatures. The physical models of COMPBRN have been reviewed extensively.3*4 In general, the COMPBRN-IIIe is geared toward modeling relatively small fires in large enclosures. The hot gas layer is assumed to be formed within the first time step of simulation. Thus, the code's predictions are expected to be most reasonable for fire scenarios involving large fuel loads during their pre-flashover burning period. The uncertainties of COMPBRN are associated with the physical property data for combustibles and model parameters. The property data are needed to define the physical behavior of the fuel, and the model parameters are needed to represent the uncertainties of the simplified physical models. A complete list of the physical properties required as input to the code is given in Table 6.1. Among the 14 property parameters, seven parameters are significant in the damage time assessment because they dominate the analysis results. The seven parameters are density, specific heat, thermal conductivity, heat value, damage temperature, specific burning rate constant, and reflectivity. The values used for cable and oil for the present analysis are included in Table 6.1. These values are selected based on referencesS- 9 and recommendations from COMPBRN-IIIe. They represent a reasonable estimate of ~bles with polyethylene/neoprene type insulation and engine oil encountered in nuclear power plants. The seven model parameters which are related to the physical modeling of doorway, heat transfer, enclosure wall, and flame/plume entrainment are listed in Table 6.2, together with the values suggested by COMPBRN-IIIe. These suggested values are used in the present analysis. Many of the suggested values were determined by comparisons with experiments such as SNL/UL Cable Tray Fire Experiments. 2 It should be noted that many of these experiments were performed under limited test conditions. The validity of these suggested values for nuclear power plants under accident conditions has not been fully demonstrated. Probabilistic distributions can be specified for all the 14 property parameters and 7 model parameters to perform the uncertainty analysis. The COMPBRN-IIIe allows many distribution types: point value, uniform, normal, lognormal, shifted lognormal, negatively skewed lognormal, exponential, and negatively skewed exponential. Some of the distribution types require three values (point, low, and high) as input to run the
uncertainty analysis.

Thus, considerable experience and knowledge is required to perform a mea.ningful 6-1 NUREG/CR-6144
Fire Growth Analysis uncertainty analysis. Without this experience and knowledge, the uncertainty analysis itself could involve a large degree of uncertainty. In addition to the simplified physical models used in the code, the code has several limitations to simulate fire accident in nuclear power plants. The following are some examples:
1. The code does not perform well when the fire source is too close to the ceiling, within the hot gas layer, or when a target is directly on top of a flame.

2. Only one doorway can be modeled and window openings can not. The location of a doorway cannot be specified.

3. The ventilation ports can only be located at the ceiling or on the floor. The fraction of flow entering or leaving each port is specified by users.

4. All rooms are treated as rectangular in shape. All objects, such as cable trays, must be oriented parallel to one of the axes.

5. Vertical or slanted burning objects cannot be modeled with accuracy.

6. The interference effect of several layers of objects, such as cable trays piled together with a small distance between them, cannot be modeled properly.
6.2 Modeling of Fires in Important Fire Areas The important areas where critical cables are located and where potential fire could occur have been identified in Section 3 of this report. Since all these areas are much larger than the postulated fire size, it is expected that the COMPBRN results will not be affected by the selection of any specific room. Therefore, only the ESGR and CVT are selected as representatives for the present fire growth analysis. The results of this analysis would be applicable to other rooms. 6.2.1 ESGR The Surry emergency switch gear room contains two sub-rooms, i.e., Rooms Hand J. The height of each room is 15 ft and 6 in. The floor areas are about 1570 and 1480 ft2 for Rooms H and J, respectively. In Room J, there are seven critical cable trays with elevations at 9, 10, 10.5, 11, 12.5, 13, and 13.5 ft above the floor. In Room H, there are 4 critical cable trays at elevations of 9, 11, 13, and 14 ft above the floor. Since both rooms are similar in size and Room J has more cable trays at various elevations, it is selected for the COMPBRN analysis. Room J also has many noncritical cable trays at elevations varying from 7 ft to 14 ft above the floor. In modeling fires in Room J of the ESGR, all the seven critical cable trays and one noncritical cable tray at 7 ft are included to simulate the responses of cables at different elevations. Under normal conditions, the forced ventilation system *provides a flow of about 9400 cfm for the ESGR. However, during a fire accident, it is assumed that the ventilation is turned off and is therefore not modeled in the base case of the analysis. The effect of ventilation is, however, included in the sensitive study presented in Section 6.4. The doorway between Rooms J and H is modeled in the analysis. NUREG/CR-6144 . 6-2
Fire Growth Analysis
6.2.2 CVT There are 16 cable trays located along the cable tunnel as shown in Figure 5-2. There are four critical cable trays (Bl, B2, AS, and C7) and 4 noncritical cable trays (Cl, C3, Al, and A3) on the left side of the tunnel.
Similarly, there are four critical cable trays (B3, B5, C4, and C6) and four noncritical cable trays (C2, A2, A4, and A6) on the right side of the tunnel. The elevations of these trays vary from 1.2 ft to 8.4 ft and are modeled in the COMPBRN analysis. The width, height, and length of the cable tunnel are about 10 ft., 9 ft 4 in and 40 ft, respectively. In modeling the size of the room, the service building vault area is included so that the total floor area is about 830 ft2. A door (2 ft 8 in by 6 ft 8 in) at the end of the tunnel is assumed open during the entire transient. Doors at the vault area are assumed closed. A forced ventilation with a flow rate at 6-7 room/hr is also included in the analysis. 6.3 Fire Analysis 6.3.1 Oil Fire (a) ESGR Room J Oil is one of the major transient fuel sources commonly found during maintenance. Only fires from a pool of oil spilled into an open surface is analyzed. The spray fire which is out of the scope of the COMPBRN code is not considered. The size of the oil pool in nuclear power plants is generally within one to 10 gal. Thus, three oil pool sizes were postulated in the analysis as shown below: Pool Gallon Mass Diameter Depth Size kg m cm Small 1 3.4 0.61 1.3 Medium 5 16.9 0.76 4.2 Large 10 33.8 0.91 5.8 The size, i.e., surface area of the pool, determines the oil burning rate and burning time for a given amount of oil mass. Using the input data given in Table 6.1, the fire would last approximately three, 10 and 14 minutes for the small, medium, and large fires, respectively. In performing the COMPBRN analysis, the horizontal distance between the oil pool and cable tray was varied in order to estimate the shortest distance at which cables would not be damaged by fires from the oil pool. All oil fires are assumed to be located at the floor level. The results of the predicted cable damage time for fires in ESGR Room J are summarized in Table 6.3. It is seen that
the distances beyond which cable would not be damaged are about 2, 3 and 5 ft for the small, medium, and large fires, respectively. When the fire source is within the critical distance, most of the cables at various elevations are damaged in a few minutes. However, no cable ignition is predicted for most of the cases except when a large fire source is located directly under or within 2 ft of the cable trays. For these two cases, cable ignition is predicted within minutes following the cable damage. There could be a large degree of uncertainty for the small fire case. For this case, 6-3 NUREG/CR-6144
Fire Growth Analysis the quantity of oil is only 3.4 kg and is exhausted in about three minutes. During this short burning time, many cable trays at higher elevations are at temperatures slightly less than the user-specified damage temperature. Hence the code predicts no damage for these cables. This prediction of no damage could be alternated by a minor change of the input parameters. A limited scope of uncertainty analysis was performed for the small and large fires in Room J. Probabilistic distributions were specified only for a few property parameters as shown in Table 6.4. The predicted results are given in Table 6.5. The results are expected to depend strongly on the assumed distributions in Table 6.4. (b) CVT The three oil pools used for fire analysis for the ESGR were applied to the cable vault tunnel. Two fire locations were assumed. In the first case, the fire source is about 3 ft from the left wall, which is about 6 in to the cable trays on the left side (Bl, B2, Cl, C3, Al, A3, A5, and C7) and 4.5 ft from the cable trays on the right side (B3, BS, C2, C4, A2, A4, A6, and C6). In the second case, the fire is located at about 5 ft from the left wall (i.e., middle of the tunnel). In all cases, the fire is at the floor level. The COMPBRN-predicted results are summarized in Table 6.6. It is seen that for the case of fire sources located at 3 ft from the left wall, cable trays on the left side of the tunnel are damaged in about two to three minutes due to the adjacency to the fire source. COMPBRN also predicts a rapid increase in temperatures beyond the user-specified ignition temperature. The ignition of cable trays at the left side of the tunnel causes a large fire which propagates to the right side and causes damag and ignition of these cables in about two min. Again, the predictions for the small fire case may involve a large degree of uncertainty because of the short burning time. For the case in which the fire source is located at the middle of the tunnel, the damage time of cables on both sides is delayed by a few minutes. No cable ignition is predicted due to the relatively large distance between the cable trays and fire source. 6.3.2 Cable Fires Cable fires are simulated in the CVf region. Fire is initially started at cable tray C7 which is located on the left side of the tunnel and is about 1.2 ft above the floor. The fire source covers a section of 2.5 ft of cables in C7. The COMPBRN-predicted results are given in Table 6.7. It appears that fire initiated at C7 immediately causes damage and ignition of cable trays located directly above the fire source. Within a few minutes the fire propagates to the right side of the tunnel and causes damage and ignition of cables on that side. In about 10 min, all trays in the tunnel region are ignited and burning. In this scenario it is assumed that none of the cable trays has metal cover. It is recognized that some of the cable trays in the tunnel region are covered by a metal sheet. The presence of a metal cover would serve as a radiation shield and therefore delay the potential damage time. The effect of a metal cover was not modeled in the present study. An attempt was made to simulate cable fires initiated at a higher elevation. A section of 2.5 ft of cables in tray C3 was selected as the fire source. The C3 tray is located at about 5.2 ft above the floor. For this case, COMPBRN predicted the damage and ignition of cable trays located directly above the fire source at about 2 minutes. The cable trays are Cl, B2, and Bl. No damage and ignition of other trays is predicted for a tim NUREG/CR-6144
6-4
Fire Growth Analysis period of 15 minutes. Since the fire source is located in the upper section of the room, it appears that the COMPBRN code has difficulty in properly simulating the hot gas layer and ventilation flow through the doorway. Interpretation of these predicted results must be cautious. 6.3.3 Cabinet Fire The COMPBRN code was used to analyze the potential cabinet fire in the ESGR. The purpose of the analysis is to simulate the impact of fires in the MCCs and cabinet on critical cables located nearby. Since COMPBRN can not model the cabinet as an enclosure with venting holes on the sides or top of the cabinet, the scenario is simply simulated as a cable fire located at an elevation equivalent to the upper section of the cabinet. A steel barrier is put directly above the fire source to simulate the top cover of the steel cabinet. The cabinet is located below the critical cables at elevations of 10 and 11 ft, and is near the critical cables at 9, 11.5, and 13.5 ft elevation. In the first case, we assumed that the fire is enclosed within the cabinet and the flame cannot be seen by these critical cables above or near the cabinet. This is done by using the noncommunication option provided by the code. The results show that the cabinet fire does not damage any cables near the cabinet for a time period of 30 min. This case is apparently not conservative. Therefore, the non-communication option of the code was removed for the second case of our analysis. The removal of the noncommunication option implies that the flame could reach out and be seen by all critical cables nearby. The COMPBRN results for this case show that the cables at the 10 and 11 ft. elevation directly above the cabinet are damaged at 2 and 5 min, respectively. Within a period of 30 minutes of cabinet fire, neither damage nor ignition to another cable is predicted by the code .
6.3.4 Trash Fire Trash fire is another transient fire source encountered during maintenance. Since trash fire consists of the burning of a solid fuel package, such as paper and cotton rags, there is a great deal of uncertainty in the anticipated behavior of a particular fuel package. Many experimental works have observed poor repeatability in many fuel packages because of the large variations of fuel configuration, packing density, and effective surface area. The existing fire model, such as the one used in COMPBRN, cannot accurately and reliably simulate the trash fire due to the lack of fundamental fuel properties to characterize the fire growth rate and heat release. Therefore, trash fire analysis*is not performed in the present study. The impact of trash fire on cable damage probably could be estimated by considering an equivalent oil fire which has about the same heat content.
6.4 Sensitivity Study 6.4.1 Pool Fire Size In Section 6.3.1, the diameter of the fire pool is assumed to be 2 ft and 3 ft for the small and large fires, respectively. In sensitivity studies, the pool size is extended to include 1 and 3 ft for the small fire, and 2 and 4 ft for the large fire. The surface of the pool has a direct effect on the fire burning rate and burning time based on the COMPBRN model as shown in Table 6.8. In the analysis, the fire source is assumed to be 1 ft from the cable trays for the small fire, and 3 ft for the large fire. The predicted cable damage time is summarized in Table 6.9. It appears that reducing the fire size reduces the potential for cable damage. When the fire size is increased, the cable* damage time is hastened. COMPBRN also predicted cable ignition when 6-5 NUREG/CR-6144
Fire Growth Analysis the fire size is increased to 3 ft for the small fire and 4 ft for the large fire. Cable ignition is not predicted
for these base cases.
6.4.2 Effect of Wall on Plume Entrainment The presence of a wall could affect the air entrainment when the fire source is near the wall or at a comer. This wall effect is considered in COMPBRN. The user's manual1 suggests that the plume entrainment coefficient be changed from 2.0 to 1.5 for pool fire next to a wall, and to 1.25 for a fire at a comer. The potential for a fire occurring near a wall is simulated by considering the fire pool located next to the back wall of Room J in the ESGR. It is assumed that the pool fire is near the three cable trays at elevations of 7, 9, and 10 ft above the floor. The horizontal distances from the cable trays are 1 ft for the small fire and 3 ft for the large fire. The COMPBRN results show that the presence of the wall has no apparent effect on the large fire and the predicted damage times remain the same as for the base case. However, the wall does have a stronger effect on the small fire. The sensitivity study shows that only the cable tray at 7 ft is damaged at 2 min and no damage is predicted for the other two trays at higher elevations. In the base case, the two cable trays at elevations of 9 and 10 ft are also damaged at 3 min just before the oil is exhausted. 6.4.3 Forced Ventilation The fire scenario discussed in Section 6.3 assumes that forced ventilation in the ESGR is turned off but that doors remain open during a fire accident. In the sensitivity study, the potential of failure to tum off the ventilation system is examined. It is assumed that the fresh air enters Room J from ports on the floor and discharges at the ceiling. A flow rate of 4700 cfm is used in the analysis. For the case in which a small fire is located 1 ft from the cable trays, COMPBRN predictions show no change in the cable damage time. Fo the case in which a large fire is located 3 ft from the cable trays, COMPBRN predicts a delay of cable damage time by about 1 min. 6.4.4 Forced Ventilation and Doorway Opening The fire scenario considered for the cable vault/tunnel region assumes that the forced ventilation is maintained and the door at the end of the tunnel is open throughout the fire accident. Since the ventilation flow tends to delay the cable damage time, the effect of no ventilation is considered for the sensitivity study. It is also considered that the doorway is closed as the forced ventilation is turned off. For oil fires, the COMPBRN predictions show that the potential for cable damage appears to be higher when the ventilation is turned off and doors are closed. For the case in which a small fire is initiated at 3 ft from the left wall of the tunnel, cable trays A5 and C7 are predicted to be damaged at about 3 min. No damage is predicted for these two trays in the base case. For the case of a large fire initiated in the middle of the tunnel, cable ignition is predicted for all cables at about 5 to 6 min. The base case predicts no cable ignition. The effects of ventilation and doorway are also examined for cable fires initiated at cable tray C7, which is located near the left side of the tunnel. For this case, there is no significant change in cable damage and ignition times. The times are alternated by only 1 or 2 min for some cable trays. NUREG/CR-6144 - 6-6
Fire Growth Analysis
The above results may depend on the location of ports where air intake and discharge takes place. In the analysis, it is assumed that fresh air enters from ports on the floor and discharges from the ceiling. No change of the ports' location was included in the sensitivity study.
6.5 References
1. V. Ho., S. Chien and G. Apostolakis, "COMPBRN Ille: An Interactive Computer Code for Fire Risk Analysis," EPRI NP-7282, Electric Power Research Institute, May 1991.

2. V. Ho, N. Siu, G. Apostolakis and G. Flanagan, "COMPBRN III - A Computer Code for Modeling Compartment Fires," NUREG/CR-4566, Oak Ridge National Laboratory, July 1986.

3. C. Ruger, J. L. Boccio and M.A. Azarm, "Evaluation of Current Methodology Employed in Probabilistic Risk Assessment of Fire Events at Nuclear Power Plants," NUREG/CR-4229, Brookhaven National Laboratory, May 1985.

4. J. A. Lambright, S. P. Nowlen, V. P. Nicholette and M. P. Bohn, "Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues," NUREG/CR-5088, Sandia National Laboratories, December 1988.

5. M. P. Bohn, et al., "Analysis of Core Damage Frequency: Surry Power Station, Unit 1 External Events,"
NUREG/CR-4550, Vol. 3, Rev. 2, Part 3, Sandia National Laboratories, December 1990.
6. A. Tewarson, J. L. Lee and R. F. Pion, "Categorization of Cable Flammability, Part 1: Laboratory Evaluation of Cable Flammability Parameters," NP-1200, Part 1, Electric Power Research Institute, October 1979.

7. S. P. Nowlen, "An Investigation of the Effects of Thermal Aging on the Fire Damageability of Electric Cables," NUREG/CR-5546, Sandia National Laboratories, May 1991.

8. C. J. Hilado, "Flammability Handbook for Plastics," 3rd Edition, Technomic Publishing Co., 1982.

9. M. Brandyberry and G. Apostolakis, "Response Surface Approximation of a Fire Risk Analysis Computer Code," Reliability Engineering and System Safety, Vol. 29, 1990, p. 153.
6-7 NUREG/CR-6144
Fire Growth Analysis Table 6.1 Physical Property Parameters Property Values Used for Present Analysis Parameters Cable Oil Density, Kg/m3 1710 900 Specific Heat, J/Kg/K 1040 2100 Thermal Conductivity, w/m/k 0.092 0.145 Heat Value, MJ/Kg 20.6 46.7 Pilot Ignition Temperature, K 773 400 Spontaneous Ignition Temperature, K 776 486 Damage Temperature, K 623 - Ventilation Control Burning Rate Constant 0.11 0.11 Specific Burning Rate Constant, Kg/m2/J 0.0043 0.061 Surface Control Burning Rate Constant, Kg/J 0.18 X 10"6 0.2 X 10*6 Combustion Efficiency 0.7 0.9 I Fraction of Flame Heat Released as Radiation 0.4 0.45 Absorption Coefficient for Flame Gases, 1/m 1.4 1.4 Reflectivity 0.2 0.35 NUREG/CR-6144
6-8
Fire Growth Analysis Table 6.2 Model Parameters Model Parameters Suggested I I Value I Heat Transfer Coefficient for Heat Transfer in a Flame, w/m2/K 22 Convective Heat Transfer Coefficient Outside of Hot Gas Layer, w/m2/K 10 Coefficient of Inflow Air through Doorway 0.6 Coefficient of Discharge for Doorway 0.7 Absorption Coefficient of Hot Gas 1.3 Heat Transfer Coefficient for Ceiling and for Objects in the Hot Gas Layer, w/m2/k 10 Buoyant Plume entrainment Coefficient* 2.0
The buoyant plume entrainment coefficient
= 2.0 for pool fire unaffected by enclosure = 1.5 for pool fire next to a wall = 1.25 for pool fire at .a corner 6-9 NUREG/CR-6144
Fire Growth Analysis Table 6.3 Cable Damage Time for Oil Fires in ESGR Room J (Time in Minutes) Fire Source Distance from Cable, Ft Cable Small Fire Medium Fire Large Fire Elevation Ft 1 2 2 3 0 I 2 3 4 4.5 5 7 2 - 2 - 1 2 3 5 10 - 9 3 - 3 - 1 2 3 7 - - 10 3 - 5 - 1 2 4 - - - 11 3 - 8 - 1 2 5 - - - 11.5 - - - - 1 2 5 - - - 12.5 - - - - - 4 9 - - - 13.5 - - - - 3 4 - - - - Note: 1. Cable at 7 ft is non-critical
2. - No damage predicted

3. Cable ignition is predicted for large fires at O and 2 ft from the cable tray.
NUREG/CR-6144 - 6-10
Fire Growth Analysis Table 6.4 Probabilistic Distributions Assumed for Uncertainty Analysis Distribution Point Low High Value Value Value Cables Heat Value J/Kg Normal 2.06(7) 1.85(7) 2.31(7) Pilot Ignition Temperature, K Neg-log 773 750 780 Spontaneous Ignition Temperature, K Neg-log 773 750 780 Damage Temperature, K Normal 623 598 643 Specific Burning Rate Constant, Kg/m 2/s Neg-log 4.3(-3) 1(-4) 7.5(-3) Reflectivity Normal 0.2 0.1 0.3 Oil ,,, . . " Specific Burning Rate Constant, Kg/m 2/s Normal 6(-2) 5.1(-2) 7.1(-2) Combustion Efficiency Uniform 0.9 0.85 0.95
Note: Point values are assumed for all other property parameters

6-11 NUREG/CR-6144
Fire Growth Analysis Table 6.5 Results of Uncertainty Analysis Cable Elevation, Ft I I 9 I 11.5 I 13.5 I I Small Fire, 1 ft Away from Cables I Probability of No Damage before 10 min 0.22 0.95 0.96 Probability of Damage within 10 min 0.78 0.05 0.04 I Large Fire, 4 ft away from Cables I Probability of No Damage before 30 min 0.22 0.63 0.86 Probability of Damage within 30 min 0.78 0.37 0.14 NUREG/CR-6144 . 6-12 **
Fire Growth Analysis Table 6.6 Cable Damage and Ignition Time for Oil Fire in Cable Vault Tunnel (Time in Minutes) Small Fire Medium Fire Large Fire Cable No. Damage Ignition Damage Ignition Damage Ignition I Fire starts at 3 ft from the left wall I Bl - - 3 5 2 3 B2 - - 2 5 2 3 Cl - - 2 5 2 3 C3 - - 2 5 2 3 Al 3 - 2 5 2 3 A3 3 - 2 5 2 3 A5 - - 3 7 2 3 C7 - - 3 7 2 4 B3 - - 6 7 4 5 BS - - 6 7 4 5 C2 - - 6 7 4 5 C4 - - 6 7 4 5 A2 - - 6 7 4 5 A4 - - 6 7 4 5 A6 - - 7 8 5 5 C6 - - 7 8 5 6 I Fire starts at 5 ft from the left wall I Bl - - - - 3 - B2 - - 7 - 3 - Cl - - 5 - 3 - C3 - - 5 - 3 - Al - - 5 - 3 - A3 - - 6 - 3 -
6-13 NUREG/CR-6144
Fire Growth Analysis Cable Table 6.6 Cable Damage and Ignition Time for Oil Fire in Cable Vault Tunnel Small Fire (Time in Minutes) (continued) Medium Fire Large Fire No. Damage Ignition Damage Ignition Damage Ignition A5 - - - - 3 - C7 - - - - 6 - B3 - - - - 3 - BS - - 7 - 3 - C2 - - 5 - 3 - C4 - - 5 - 3 - A2 - - 5 - 3 - A4 - - 6 - 3 - A6 - - - - 3 - C6 - - - - 6 - NUREG/CR-6144
6-14
Fire Growth Analysis
I Table 6.7 Cable Damage and Ignition Time for Cable Fires in the Vault/Tunnel Region Tray No.
I Damage I (Time in Minutes) Ignition I Tray No. I Damage I Ignition I Bl 5 5 B3 8 8 B2 5 5 BS 8 8 Cl 5 5 C2 8 8 C3 2 3 C4 8 8 Al 2 3 A2 8 8 A3 2 3 A4 8 8 A5 2 3 A6 9 9 C7 0 0 C6 9 10 6-15 NUREG/CR-6144
Fire Growth Analysis I Small Fire (1 Gal) Pool Size, ft Table 6.8 Effect of Oil Pool Size on Combustion Characteristics 1 2 3 I Area, m2 0.073 0.29 0.6567 Depth, m 0.05 0.013 0.0058 Burning Rate, Kg/s 0.0044 0.0175 0.0394 Burning Time, s 771 193 86 I Large Fire (10 Gal) I Pool Size, ft 2 3 4 Area,m2 0.29 0.657 1.167 Depth, m 0.13 0.058 0.0325 Burning Rate, Kg/s 0.0175 0.0394 0.071 Burning Time, s 1930 858 476 Note: Burning rate and time are based on 100% efficiency. NUREG/CR-6144
6-16
Fire Growth Analysis
I Table 6.9 Effect of Oil Pool Size on Cable Damage Time I Pool Size, Ft.
I I Small Fire (1 Gal) I Cable Elevation, Ft 1 2 3* 7 - 2 1 9 - 3 1 10 - 3 1 11 - 3 1 11.5 - - 1 13.5 - - 3 I Large Fire (10 Gal) I Cable Elevation, Ft 2 3 4* 7 - 3 2 9 - 3 2 10 - 4 2 11 - 6 2 11.5 - 5 2 13.5 - - 3
Cable ignition predicted

6-17 NUREG/CR-6144
7. FIRE SUPPRESSION MODELING
The suppression model is Mferent than the 1150 model where a mean suppression time was used1*'. The conditional probability of nonsuppression in the 1150 model is given by:
(7.1) where fns is the nonsuppression fraction, td is the predicted time to damage (corresponding to 7G below, in section* 7 .1) and t. is the mean suppression time, equaling 42 minutes according to Reference 2. The suppression model used in this study is somewhat more sophisticated. The nonsuppression fraction will depend on the type of suppression and detection equipment available in a particular fire area. The mean time to suppression will have a finer granularity, too (and will be only implicitly present in the model); it is broken down into stages in the suppression process, with each stage having a success/failure decision box attached to it, and the stages connected by branches representing transition rates. The nonsuppression fraction will have a different distribution (supported by numerical calculations) than the one in equation (7.1). The parameters of this model will be calculated from field data (nuclear and non-nuclear) and specialized to each type of fire area and fire severity. 7 .1 Transition Diagram of Fire Detection and Suppression Successful suppression will prevent damage to a component if the fire started nearby. The conditional probability of damage to a component given a fire event is then equal to the probability that suppression will not extinguish the fire before damage. (In case of electrical cables, damage is defined as reaching a certain critical temperature of the insulation). A more detailed explanation of the suppression model is given in Appendix F, taken from reference 3. The nonsuppression probability is the fraction of cases where the time to detect and control the fire (hazard time, t8 ) is greater than the time needed to damage the component (growth time, ~). i.e.: (7.2) As explained in Appendix F, to can be replaced by its mean value, 7G, and equation (7.2) becomes: (7.3) (7.4) where FI t8 is the cumulative probability function for t8 . The suppression analysis in this study is based on the transition diagram model of suppression described elsewhere3 *4 *5
Figure 7 .1 shows the suppression model. The suppression process in a particular case progresses between the states in Figure 7.1, starting with detection, according to the conditional probabilities of the branches
(<h_'s and 1'1K's) and the transition rates (A;/s) along the branches (i.e., between the states). K are the state designators, whereas ij are the branches' identifiers (as per Figure 7 .1) .
7-1 NUREG/CR-6144
Fire Suppression Modeling 7 .2 Calculation of Parameters for fns The parameters of the model (q,'s, 1/t's and >..'s) for a particular type of fire area are calculated, in Appendix F, from available data on suppression in nuclear power plants, and elsewhere. The types of fire areas (from suppression standpoint) are distinguished based on availability of detectors, special systems (e.g., CO2 , Halon, sprinklers), and manual or automatic actuation of such systems. The types of fire areas encountered are given in Table 1.8, Appendix F. For the areas of interest here, the ESGR is in category IV (has detectors but the Halon system is manually actuated), the CVT (particularly the tunnel area) is category V (detectors with automatic actuation of Cardox, plus a manually actuated sprinkler system), whereas the CT is in category II (detectors, no special systems except for standpipe). (The MCR suppression is described in Chapter 9). This is shown in Table 7 .1. Once the parameters of the transition model are known (shown in Tables 1.4 and 1.5 of Appendix F), it is possible to calculate the nonsuppression probability for a given growth time and for a particular distribution of time along the particular path taken in a particular case. This involves integration over all possible paths and all possible time allotments of time to various stages along the suppression path. A simplification of the above procedure is developed in Appendix F, where it is shown that the distribut1on in tH is approximately Weibull: (7.5) where the parameters ('Ye, ~c) can be determined from the mean and variance of tH using the following equations: (7.6) (7.7) The mean and variance of hazard time and for a particular class of rooms, E(tH IC) and Var(tH IC), are calculated in Appendix F and shown in Table 1.9 there. The "high severity" data in the "hazard time mean" and "hazard time variance mean" columns are used in this study. The parameters 'Ye and fo are calculated using these data in equations 7 .6 and 7. 7 and the calculated values are shown in Table 7 .2 for various room classes of interest. These values will be applied in Chapter 8, using equations (7.5) and (7.4), to calculate fns for cases of interest. 7 .3 References
1. M.P. Bohn et al., "Analysis of Core Damage Frequency: Surry Power Station, Unit 1 External Events",
NUREG/CR-4550 Vol. 3, Rev. 1, Part 3, December 1990.
2. J.A. Lambright et al., "Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues", NUREG/CR-5088, January 1989.

3. Pickard, Lowe and Garrick, Inc. (PLG, Inc.), "Bemau Risk Analysis Plant with NANO", prepared for Nordostschweitzerische Kraftwerke AG, PLG-0511, December 1989.
NUREG/CR-6144 - 7-2
Fire Suppression Modeling Nathan Siu and George Apostolakis, "Modeling The Detection And Suppression of Fires in Nuclear Power Plants", from International ANS/ENS Topical Meeting on Probabilistic Safety Methods and Applications, February 24-March 1, 1985.
5. N. Siu and George Apostolakis, "A Methodology for Analyzing the Detection and Suppression of Fires in Nuclear Power Plants", in Nuclear Science And Engineering 94, pp. 213-226 (1986).
7-3 NUREG/CR-6144
,- ~,
In1t1nt1neoua Tran1ition1 Delayed Tr1n1ition1 DI Detected ln1t1nt1neou1ly? A Aut11111tic Detactora Available? SH Initial Manuel Supprenaion Succeaaful7 (by on*1ite pereonnel) SA Pixed Suppre1aion Sy1tema Available and Successfully Actuated? SS Pixed Supp~eaalon Systema Successful? Figure 7.1 Fire Suppression Model
Fire Suppression Modeling
Table 7.1 Classification of Fire Areas into Suppression Categories of Table 1.8, Appendix F FIRE AREA ESGR CVT CT Suppression Category IV V II Table 7.2 Parameters ~c and "(C for Fire Areas FIRE AREA ESGR CVT CT
~c 17.5 21.7 18.6 "(C .5 .575 .55 NUREG/CR-6144 7-5
8. DAMAGE FRACTION CALCULATIONS The damage fractions are calculated from equations (7.4) and (7.5), using data in Table 7.2. The resulting damage fractions (i.e., fns in equation (7.4)) are given in Tables 8.1, 8.2, and 8.3 for the ESGR, CVI', and CT, respectively, and for the damage times of interest.
8.1 ESGR In the ESGR, the damage times of interest are 3 min for a small fire (used for switchgear panel fires that reach out through the casing) and 5 min, 2 min and 1 min for large fires within 3, 2, and Oft (center to edge) of the cable tray, respectively. This is based on Table 6-3 showing the tray elevation-horizontal displacement-damage time relationship and Table 5.1 showing critical cable tray elevations. (The elevations in Table 6-3 are relative to the 9'6" floor elevation, whereas elevations in Table 5.1 are absolute.) The RHR cables (whose damage is needed in order to have an initiator) are generally in higher trays. Therefore, the damage fraction (from Table 8.1) is .66 for a switchgear panel fire; .59 for a large oil fire centered in the swath between 2.5 and 3 ft from the tray; .71 for a large oil fire centered in the swath between 1.5 ft and 2.5 ft from the tray; .94 for a large oil fire centered within 2.5 ft on either side from the tray centerline. (The tray is 2 ft wide.) The aggregate damage fraction is then 0.84 for a large oil fire, 0.66 for a small oil fire, and 0.66 for a switchgear fire. For cable fires in the ESGR, growth time results from cable fires in the CVI' (see below) were conservatively used. For a short stack of trays (4-5 trays above the initiating tray), the damage will propagate within 2 min to all the trays (damage will be immediate in the initiating tray). For a tall stack (10-11 trays above the initiating tray), half the trays will sustain damage within 2 min, the other half within 5 minutes (again, damage is immediate in: the initiating tray). Therefore, keeping in mind that the RHR-cable-carrying trays are near the top in the ESGR, and that the H room contains a 4-tray stack, whereas the J room has an 11-tray stack, the aggregate damage fraction for cable fires (arising from all possible combinations) in the H room is .78 and in the J room it is .68. 8.2 CVT In the CVf, the Chapter 6 results for a sensitivity analysis of the growth time, subsection 6.4.4, are used instead of the base case. This is due to the base case (for CVI) keeping the doors open and ventilation on (an optimistic case for COMPBRN modeling). When the ventilation is turned off and the doors are closed, a large oil fire on the tunnel centerline will cause damage to all the trays within 3 min; a cable fire starting in a bottom tray will propagate to the top tray within 5 min, and to the other side of the tunnel within 7 min. Then, using Table 8.2, the damage fractions are .73 for a large oil fire and .65 for a cable fire. 8.3 CT In the CT, the connecting cable tray between the two RHR pumps is too high to be directly damaged by a pump fire; however, propagation along the vertical pump cable (assuming it is not in a conduit) to the tray is probable; such an event cannot be modeled by COMPBRN. The damage fraction will be conservatively assumed to be 1, as not enough is known about the propagation time of a fire along a vertical cable. For the transient fire in scenario CTI, the 45' section of tray reachable by an oil fire (large or small) is at a relative elevation of 10'4". The results of Table 6-3 are used, as they apply for a large room. The results will 8-1 NUREG/CR-6144
Damage Fraction Calculations be modified somewhat, as the ventilation option will make no difference due to the size of the containment. According to Subsection 6.4.3, damage time due to a large fire at a distance of 3 ft should be increased by 1 min, while other results are unchanged. Using these results, and applying the same reasoning as in section 8.1 (keeping in mind that the trays are anchored to the crane wall), results in a damage fraction of .84 for a large fire and .66 for a small fire, from Table 8.3. The aggregate damage fraction for transient fires is 0.71. For the cable fires in scenario CTl, a 50.75' section contains 3 trays (below or at the RHR cables-carrying tray), a 31.2' section contains 2 such trays and a 19.7' section contains one tray. Applying the same logic as in 8.1 and Table 8.3, the 3-tray section will have a damage fraction of 0.81, the two-tray section will have a damage fraction of 0.86, _and the one-tray section will have a damage fraction of 1. The aggregate damage fraction for cable fires is 0.86. 8.4 Results Table 8.4 shows the results of damage fraction calculations for each room as described above. (As there are no cables involved, NSGR fire scenarios have a damage fraction of 1.) NUREG/CR-6144. 8-2
Damage Fraction Calculations Table 8.1 Nonsuppression Fraction, fns, vs. Time in ESGR Time (Min) 1 2 3 4 5 7 10 0.94 .71 .66 .62 .59 .53 .47 Table 8.2 NonSuppression Fraction, fns vs. Time in CVT 1 2 3 4 5 7 10 II Tlllle :ID)
.95 .78 .73 .68 .65 .59 .53 I I Table 8.3 NonSuppression Fraction, fns, vs. Time in CT I
Time (Min) 1 2 3 4 5 7 10 fns .95 .75 .69 .65 .62 .56 .49 II Table 8.4 Damage Fractions for Fire Scenarios Room Scenarios Damage Fractions ESGR El-E7 0.66 E9 0.84 0.66 0.78 1 ElO 0.84 0.66 0.68 1 E8 1 Ell-El3B 1 CVT CVl 0.73 0.692 CT CTl 0.71 0.862 CTI 1 NSGR Nl-N3 1 1 First number is for large oil fires, second number is for small oil fires, and third number is for cable fires 2 First number is for large oil fires, second number is for cable fires 8-3 NUREG/CR-6144
~- 9. CONTROL ROOM FIRE RISK ANALYSIS The main control room (MCR) at Surry serves as the nerve center of controls for all plant operations. The MCR houses a multitude of the instrumentation and controls for various plant systems. Due to the close spatial arrangement of the controls (e.g., control switches, control wiring, circuit boards, controllers) in the MCR, occurrence of fires can simultaneously damage control capability of multiple pieces of equipment Thus, fires are considered to be a potential risk contributor. The purpose of this section is to assess the risk of fires to the MCR at Surry during mid-loop. The scope of the analysis presented in this section is limited to the portion of the MCR that is dedicated to Unit 1 operations. Fire scenarios that can cause core damage are identified. These scenarios are then screened qualitatively. The occurrence frequency of the scenarios that survive the preliminary screening are then assessed The overall process in the control room frre risk analysis is inherently counteractive and must be balanced in a meaningful practical fire risk analysis. Ideally, all potential hazardous scenarios that can cause any conceivable damage would be identified to ensure that all locations and all possible hazards within the control room are fully examined On the other hand, to most efficiently use available resources and to maintain a proper balance throughout the risk assessment process, only the relatively risk significant scenarios are analyzed in detail. This top-down approach to risk assessment minimizes the effort in quantifying the risk associated with unimportant scenarios. Therefore, the scenarios developed for the MCR should be as comprehensive as possible, but they remain at a manageable number for the subsequent frequency, core impact, and recovery action analyses. A fire scenario in this section is defined at the location level. A scenario describes all possible fires that occur at or near a particular location and damage the components within that location. Each scenario has a unique combination of the fire sources, fire initiators, location and magnitude of the fires, their occurrence frequencies, and the damage (or impact) caused by the fires to the plant. Fire scenarios developed for the MCR are generally grouped into several impact categories (i.e., grouping the scenarios with the same damage impact). One of these impact categories, consisting of a large number of scenarios, is identified as "no impact on the shutdown risk". For example, fire-induced control failure of the main feedwater is important to the plant power operation but is not a risk contributor during mid-loop operation at plant shutdown. To streamline the effort in the identification of risk significant fire scenarios, the analysis approach is to focus on the fire scenarios that can potentially impact or degrade systems that are required to be operating during mid-loop operation at shutdown such as the residual heat removal system, component cooling system, or charging system. Therefore, this study begins with identifying the control locations for the equipment that is involved in the decay heat removal operation or supports the operation of RHR. In addition, control locations for the equipment that can be used for the mitigation of a loss of RHR event are also identified. This analysis conservatively assumes that fires occurring around the control locations would cause an adverse impact on the modeled function. For the postulated impact, all possible fires that could cause the same damage are considered in the occurrence frequency estimation of the scenario. The fire scenarios are then incorporated into the overall plant risk model as the fire-induced initiating events to determine their risk contribution (i.e., in terms of core damage frequency). In summary, the control room fire analysis consists of the following major steps:
1. Generate a list of the equipment considered in providing the shutdown cooling or the mitigation of shutdown accidents.

2. Locate the controls for the above equipment.
9-1 NUREG/CR-6144
Control Room Fire Analysis
3. Postulate fire scenarios based on impact on the shutdown cooling, the RCS inventory control functions, and the mitigation of loss of RHR events.

4. Screen the scenarios qualitatively according to effective fire sizes.

5. Estimate the scenario occurrence frequency.
The first three steps in this analysis are described in Section 9.1, fire scenario development. Section 9.2 discusses the fire scenario quantification process. The references cited in this section are listed in Section 9.3. 9.1 Fire Scenario Identification The Surry MCR is a common area serving both Units 1 and 2. It is located on elevation 27' -0 of the Service Building. The main control boards for Unit 1 are completely independent of those for Unit 2. The control room is manned 24 hours continuously; and manual fire suppression is available with the fire extinguishers inside the control room and a hose station located in the turbine building. The potential fire sources in the control room are expected to be either the transient combustibles or the electrical equipment such as control cables. The fire induced by the transient combustibles that would damage the control equipment or cause a control room evacuation is unlikely because of the continuous presence of the operators and the easy access to the frre extinguishers. Furthermore, based on past industrial experience (Ref. 9-1), there has been no evidence of transient combustible induced fires in the control room area. Therefore, this analysis concentrates its effort in the cabinet cables-induced fires, and transient combustible induced fires are not considered. In addition to the control features provided in the main control room, alternate control capabilities are available outside the control room such as the auxiliary shutdown panel (Ref. 9-2) and switchgear room. T a b l . 5-2a of Reference 9-2 lists the plant equipment that can be controlled from the auxiliary shutdown panel in the event of evacuation from the main control room due to fires. In this analysis, credits can be taken for this type of equipment in the evaluation of control room fires that would impact the corresponding controls in the MCR. A simplified layout of the control room is provided in Figure 9-1. As shown in the figure, the control room contains:
1. Benchboard sections 1-1 and 1-2

2. Vertical board sections 1-1, 1-2, 1-3, 1-4, 1-5, 1-6, and 1-7

3. Semi-vital bus and the associated transformer

4. Vital bus cabinets 1-11 and 1-IV and terminal box for vital bus train B

5. 125V DC cabinets 1-1 and 1-2

6. 120V AC cabinet

7. Miscellaneous panels such as PAMC panel and junction boxes The Benchboards anµ Vertical Boards (Ref. 9-3) contain controls for the reactor coolant system, engineered safeguards systems (i.e., High-Head Safety Injection system, Low-Head Safety Injection system, Containme NUREG/CR-6144 . 9-2
Control Room Fire Risk Analysis Spray system, and Recirculation Spray system), auxiliary systems (including Chemical and Volume Control System, Residual Heat Removal system, Component Cooling Water system, Charging Pump Cooling Water system) and secondary systems (e.g., Main Steam system, Steam Generator Feedwater system, Auxiliary Feedwater system, and Condensate system), emergency power system (Diesel Generators 1 and 3), and the power generation systems. The semi-vital bus and its associated transformer (480V/120V) are located behind the Vertical Board section 1-1. This bus supplies power for the operation of the steam generator power-operated relief valves. Vital buses II and IV and their associated terminal box are located on the west wall of the control room. These buses provide uninterruptible power supply to vital instruments and equipment. The 125V DC and 120V AC cabinets distnbute the required control power for solenoid-operated valves and control relays, and the power supply for indicating lights, monitors, recorders, and annunciators in the main control room. The control devices ( e.g., control switches) for the equipment in various systems are mounted on the front panel of the control boards or cabinets. In this analysis, the control associated components (such as control cables, fuses, contacts, etc.) are generally assumed to be installed inside a panel or board in the proximity of their associated control device. Fire incidents occurring in the aforementioned boards, cabinets, or panels could lead to different impacts on plant safety depending on their location of occurrence and severity. For example, a small fire occurring on the left side of Benchboard section 1-1 could damage the controls of the shutdown accident mitigation equipment such as the low head safety injection pumps. Another fire with similar severity occurring at a place where the controls for the containment spray system are located might contribute much less to the shutdown risk because the containment integrity may not be maintained during plant shutdown. Thus a significant scenario induced by a small fire is possible only if it occurs near the controls for the equipment of concern. On the other hand, a fire with greater severity could cover a large area, and thereby cause a more severe impact but with less frequency. From the past nuclear industrial operating experience (Ref. 9-1), small fires generally cause minor impact but they may occur more frequently. Large fires (with higher severity) might occur at a lower occurrence frequency but result in a more severe impact due to a larger damage zone. To analyze the control room fires that would contribute significantly to the risk during shutdown, only those could disrupt the operations of the RHR or the RCS inventory control are considered. These fires may or may not cause impacts on the shutdown accident mitigation equipment. Therefore, the controls for the shutdown accident mitigation equipment need to be located first and then fires with possible combinations of the initiating location and severity that could cover the controls for the equipment of interests are considered. In the following sections, the process of fire scenario identification is described. 9.1.1 Shutdown Accident Mitigation Equipment and Control Locations As mentioned before, fire scenarios being analyzed are those fires that could possibly cause damages to the controls of the shutdown accident mitigation equipment. To postulate a significant fire scenario, information about the control location for the shutdown accident mitigation equipment is required. The shutdown accident mitigation equipment are taken from the basic events modeled in each of the fault trees developed for the
event tree functionai top events. These fault tree models include 9-3 NUREG/CR-6144
Control Room Fire Analysis
1. RCS makeup at mid-loop,

2. Restoration of RHR,

3. Steam generator reflux cooling,

4. RCS feed and bleed, and

5. RCS gravity feed.
The basic events list is provided in Table 9-1. Also provided in the table is the basic event description. Because the description provides sufficient information to determine its associated equipment in most cases, the conversion of the basic events into equipment is not performed. Instead, only annotations in parentheses are added to those basic events that require more information for identifying the associated equipment. Also noted in this table, for the equipment with different failure modes modeled ( e.g., component cooling water pump 1B fails to start on demand, designated as CCW-MDP-FS-CCPlB, and fails to run, CCW-MDP-FR-CCPlB), more than one basic event may be listed. A screening is first performed to exclude those basic events that are not of interest, such as human errors, common cause faults, maintenance and testing events, and equipment that is not equipped with a control feature in the main control room. This equipment includes check valve, manual valve, heat exchanger, strainer, tank, pipe segment, and air dryer. The remaining basic events are thus used as the source of information for generating the list of potential shutdown accident mitigation equipment that can be controlled from the main control room. The outline drawings (Ref. 9-3) for the control boards and panels are used to locate the controls for each piece of equipment listed in Table 9-1. The outline drawings provide information about the physical arrangement of the instrumentation and control (such as indicators, annunciators, controllers, recorders. control switches, etc.) on the front panels. The control location of the shutdown accident mitigatio equipment identified in this review process is presented in Table 9-2. The location information in Table 9-2 is listed in terms of the control board and the subsection on the control board. For example, the controls for charging pumps lA, lB, and lC are all located on the BenchBoard 1-1, subsection 3W, thus the location is identified as BB-1-1 (3W). Based on the control locations, the fire-induced impact on the controls of the shutdown accident mitigation equipment is postulated for the significant fire scenarios. 9.1.2 Significant Scenario Identification To ensure plant safety during mid-loop operation at shutdown, the residual heat generated in the fuel must be removed and water inventory in the reactor coolant system is required to be controlled at a level where the mid-loop operation can be successfully performed. The RHR system is used to provide the residual heat removal function. The reactor coolant inventory control is maintained by the charging and letdown operations. If a fire event does not cause any perturbation to these two functions, the fire is not considered to be a significant risk contributor. For this reason, the significant fire scenarios in this study are defined as those that might potentially damage the controls for shutdown accident mitigation equipment and interrupt the heat removal or the inventory control function. Therefore, the critical control locations are identified from Table 9-2:
1. Subsections 3AE, 3V, 3W, and 3Z in Benchboard 1-1.

2. Subsections 3A, 3B, 3F, 3G, and 3H in Benchboard 1-2.

3. Subsections 3BC, 3BD, and 3B in Vertical board 1-1.
NUREG/CR-6144 . 9-4
Control Room Fire Risk Analysis
4. Subsection 3AQ in Vertical Board 1-2.
5. Diesel Generator 1 control panel in Vertical Board 1-6.

6. Diesel Generator 3 control panel in Vertical board 1-7.

7. 120V AC vital bus 1-II.
The failure modes resulting from control damage due to fires are also of interest in the scenario identification process. From the industry experience, the most common failures of the control equipment are open circuits or shorting to ground in the control circuitry. Hot shorts are addressed in several control room fire scenarios in which an undesired operation of the plant equipment is initiated. For example, fire-induced open circuits or shorting to ground in the control circuitry for the operating RHR pump would not have any adverse impact on the continued RHR operation, but a hot short in the circuitry could generate a spurious signal that causes the pump to trip. In this case, only fires with hot shorts are considered since the open circuits or grounding do not have any significant impact. For the postulated scenarios, control failure with open circuit or shorting to ground is generally assumed unless another failure mode is specified. Fire scenarios are thus postulated by considering the fire-induced control damage in the critical control locations listed above. To provide insights for the scenario postulation, each of the fire scenarios identified is discussed as follows: BB-1-1: Benchboard 1-1 subsection 3Z contains the controls for the RHR pumps lA and lB, the RHR heat exchanger flow control valve HCVl 758, the heat exchanger bypass valve HCV1605, the hot leg suction isolation valves MOVl 700 and MOVl 701, the cold leg injection isolation valves MOVl 720A and MOVl 720B, and the RHR letdown flow control valve HCV1142. The RHR system provides shutdown cooling when the RCS is depressurized to below 450 psig and cooled to below 350 F. Loss of RHR decay heat removal function induced by fires is the key concern of this scenario. This scenario addresses several ways of failures that might lead to a loss of RHR, and they are
1. failures of both RHR pumps to operate

2. failure of MOV1700 or MOV1701 to remain open

3. failures of MOV1720A and MOV1720B to remain open

4. failure of FCV1605 to control heat exchanger bypass flow

5. failure of HCV1758 to control heat exchanger flow

6. failure of HCV1142 to control letdown flow Most of the above failures require at least one hot short in the control circuitry.
9-5 NUREG/CR-6144
Control Room Fire Analysis BB-1-2: Benchboard 1-1 subsections 3W and 3V contain the controls for charging pumps lA, lB, a n d . lC, charging pump cooling pumps 2A and 2B, charging pump service water pumps lOA and 10B, RWST supply isolation valves M0V1l15B and MOV1115D, and VCT supply isolation valves MOV1115C and MOV1115E. Operation of the charging pumps requires pump seal cooling from the charging pump cooling system and lube oil cooling from the charging pump service water system. Loss of charging flow and High Hat Safety Injection (HHS!) are considered in this fire scenario which can be induced by
1. failure of the three charging pumps

2. failure of the charging pump cooling pumps

3. failure of the charging pump service water pumps

4. failure of charging pump suction supplies The above failures require at least one hot short in the control circuitry.
BB-1-3: Benchboard 1-2 subsections 3A and 3B contain the controls for feed breakers 15Fl and 15Dl, which provide power from the reserve station service transformers C and A to 4 kV buses F and D, respectively. Buses F and D are the normal power supply to 4 kV buses lH and 1J which provide the motive power to the charging pumps and the RHR and CCW pumps (via stub buses). Loss of normal power supply to 4 kV buses lH and lJ is considered in this fire scenario. The above failures require at least one hot short in the control circuitry. BB-1-4: This scenario addresses fires with a medium severity such that the damage zone covers subsections 3AE and 3Z on Benchboard 1-1. Subsection 3AE contains the controls for th motor-operated isolation valves (1867C and 1867D) at the charging pump discharge to the RCS cold leg, and the low head pressure injection (LHSI) pumps lA and lB and their associated valves. Fire-induced closure of these two valves would cause loss of HHS! through the RCS cold legs. The controls in subsection 3Z are as descnbed in scenario BB-1-1. The impact of this scenario is considered as loss of RHR, HHS! through cold legs, and LHSI. The above failures require at least one hot shorts in the control circuitry. BB-1-5: This scenario addresses fires with a large severity such that the damage zone covers subsections 3AE, 3Z, 3W, and 3V. The controls of interests in these subsections are as descnbed in scenarios BB-1-1, BB-1-2, and BB-1-4. The fire-induced impact is considered as the combined impact of these three scenarios (i.e., loss of RHR, charging flow, HHS!, and LHSI.) The above failures require at least one hot short in the control circuitry. BB-1-6: This scenario addresses fires with a large severity such that the damage zone covers subsections 3AE, 3Z, 3W, 3V, 3H, 3G, and 3F. Subsections 3H, 3G, and 3F contain the controls for the steam generator PORVs, main steam trip valves, and steam dump valves. The steam generator PORVs or the steam dump valves are required to provide the steam relief path in the event of a loss of RHR. Fire-induced failure of the controls is assumed to cause a loss of SG steam relief path. The controls of interests in subsections 3AE, 3Z, 3W, and 3V are as described in scenarios BB-1-1, BB-1-2, and BB-1-4. The fire-induced impact is considered as loss of RHR, charging flow, HHS!, LHSI, and SG steam relief path. The above failures require at least one hot short in the control circuitry. NUREG/CR-6144 . 9-6
Control Room Fire Risk Analysis BB-1-7: This scenario addresses fires with a large severity such that the damage zone covers the entire benchboards 1-1 and 1-2. The controls of interests in these boards are as described in scenarios BB-1-1, BB-1-2, BB-1-4, and BB-1-6. The impact of this scenario is considered as the combined impact of the four scenarios (i.e., loss of RHR, charging flow, HHSI, LHSI, steam generator steam relief path, and normal power supply to 4 kV buses lH and lJ). The above failures require at least one hot short in the control circuitry. BB-1-8: This scenario addresses fires with a medium severity such that the damage zone covers subsections 3Z, 3W, and 3V on benchboard 1-1. The controls located in these subsections are as described in scenarios BB-1-1 and BB-1-2. The fire-induced impact is considered as the combined impact of these two scenarios (i.e., loss of RHR, charging flow, and HHSI). The above failures require at least one hot short in the control circuitry. BB-1-9: This scenario addresses fires with a large severity such that the damage zone covers subsections 3Z, 3W, 3V, 3H, 3G, and 3F. The controls located in these subsections are as described in scenarios BB-1-1, BB-1-2, and BB-1-6. The fire-induced impact is considered as loss of RHR, charging flow, HHSI, and steam generator steam relief path. The above failures require at least one hot short in the control circuitry. BB-1-10: This scenario addresses fires with a large severity such that the damage zone covers subsections 3Z, 3W, 3V, 3H, 3G, 3F, 3B, and 3A in Benchboards 1-1 and 1-2. The controls of interest in these subsections are as descnbed in scenarios BB-1-1, BB-1-2, BB-1-3, and BB-1-6. The impact of this scenario is considered as loss of RHR, charging flow, HHSI, steam generator steam relief path, and normal power supply to 4 kV buses 1H and lJ. The above failures require at least one hot short in the control circuitry. '\,* BB-1-11: This scenario addresses fires with a medium severity such that the damage zone covers subsections 3W, 3V, 3H, 3G, and 3F in Benchboards 1-1 and 1-2. The controls of interest in these subsections are as described in scenarios BB-1-2 and BB-1-6. The impact of this scenario is considered as loss of charging flow, HHSI, and steam generator steam relief path. The above failures require at least one hot short in the control circuitry. BB-1-12: This scenario addresses fires with a large severity such that the damage zone covers subsections 3W, 3V, 3H, 3G, 3F, 3B, and 3A in Benchboards 1-1 and 1-2. The controls of interest in these subsections are as descnbed in scenarios BB-1-2, BB-1-3, and BB-1-6. The impact of this scenario is considered as loss of charging flow, HHSI, steam generator steam relief path, and normal power supply to 4 kV buses 1H and lJ. The above failures require at least one hot short in the control circuitry. BB-1-13: This scenario addresses fires with a large severity such that the damage zone covers subsections 3H, 3G, 3F, 3B, and 3A in Benchboard 1-2. The controls of interest in these subsections are as descnbed in scenarios BB-1-3 and BB-1-6. The impact of this scenario is considered as loss of steam generator steam relief path and normal power supply to 4 kV buses lH and lJ. The above failures require at least one hot short in the control circuitry. 9-7 NUREG/CR-6144
Control Room Fire Analysis VB-1-1: Vertical Board 1-1 subsection 3BH contains the controls for CCW return headers A and B trip valves TV109A and TV109B. The CCW system provides cooling water to the RHR heat exchangers for decay heat removal and the RHR pump seal coolers. Fire-induced closure of both valves would block the CCW return flow which, in turn, fails the CCW system. The failures descnbed in this scenario does not require a hot short. VB-1-2: Vertical Board 1-1 subsections 3BD and 3BC contain the controls for the charging pump suction and discharge isolation valves (MOV1267A and MOV1286A for pump lA, MOV1269A and MOV1286B for pump lB, and MOV1270A and MOV1286C for pump lC). Failure of the suction isolation valves or the pump discharge isolation valves to remain open would cause loss of charging flow and HHSI which is considered in this fire scenario. The above failures require at least one hot short in the control circuitry. VB-1-3: Vertical Board 1-2 subsection 3AQ contains the control for CCW pumps lA and lB. RHR shutdown cooling requires heat removal by the CCW system. Loss of the CCW system is the key concern of this fire scenario. The above failures require at least one hot short in the control circuitry. VB-1-4: This scenario addresses fires with a medium severity such that the damage zone covers subsections 3BH, 3BD, and 3BC on the Vertical Board 1-1. The controls located in these subsections are as descnbed in scenarios VB-1-1 and VB-1-2. The fire-induced impact is thus the combined impact of these two scenarios (i.e., loss of CCW, charging flow, and HHSI). The above failures require at least one hot short in the control circuitry. VB-1-5: This scenario addresses fires with a medium severity such that the damage zone covers subsections 3AQ, 3BD, and 3BC in Vertical Boards 1-1 and 1-2. The controls of concern located in these subsections are as described in scenarios VB-1-2 and VB-1-3. The fire-induced impact is thus the combined impact of these two scenarios (i.e., loss of CCW, charging flow, and HHSI). The above failures requii:e at least one hot short in the control circuitry. VB-1-6: This scenario addresses fires with a large severity such that the damage zone covers subsections 3AQ, 3BD, 3BC, and 3BH in Vertical Boards 1-1 and 1-2. The controls of concern located in these subsections are as described in scenarios VB-1-1, VB-1-2 and VB-1-3. The fire-induced impact is thus the combined impact of these three scenarios (i.e., loss of CCW, charging flow, and HHSI). The above failures require at least one hot short in the control circuitry. VB-6-DGl: The diesel generator No. 1 control panel in Vertical Board 1-6 contains the controls for diesel generator No. 1, the feed breakers (ACB 15H8 from the normal supply 4 kV bus F and ACB 15H3 from the emergency supply) to the 4 kV bus lH, and the tie breaker 15Hl (from 4 kV bus lJ). Fire-induced failure of the controls could cause a loss of 4 kV bus lH which, in turn, fails thepower'supply to one train of the charging, RHR, and CCW pumps. The above failures require at least one hot short in the control circuitry. VB-7-DG3: The emergency diesel generator 3 control panel in Vertical Board 1-7 contains the controls for diesel generator 3, which is shared with Unit 2, and the feed breakers (ACB 15J8 from the normal supply and ACB 15J3 from the emergency supply). Failure of these controls induced by fires may lead to a loss of 4 kV bus 1J which would cause the loss of power supply to on NUREG/CR-6144 - 9-8
Control Room Fire Risk Analysis train of the charging, RHR, and CCW pumps. However, this train is assumed to be the standby train. VBC-1-II-1: The 120V AC vital bus 1-II supplies control power for the CCW return header B trip valve TV109B. Failure of this vital AC bus causes closure of the valve which can lead to a loss of CCW cooling to the RHR heat exchanger in service. The failures described in this scenario does not require hot short. Fire scenario involving individual failure of the controls in subsection 3AE in benchboard 1-1 or subsections 3H, 3G, and 3F in Benchboard 1-2 is not included in the above list because the impact does not cause any disturbance to the RHR or charging systems. However, the control damages are evaluated with some other control failures that would cause a loss of RHR (e.g., scenario BB-1-4). Fires in the 125V DC cabinets 1-1 and 1-2 and the 120V AC cabinet 1-1 located inside Vertical Board 1-2 are not evaluated in this analysis. This is based on a review of the loads that are supplied by the cabinets (load list in Ref. 9-3). This leads to a conclusion that no significant impact due to fires can be resulted. The identified MCR fire scenarios are summarized in Table 9-3. A preliminary screening is performed to exclude those fire scenarios that are judged to be highly unlikely or insignificant. The screening is based on the following observations:
1. The voltage of the electrical equipment in the MCR is relatively low (e.g., 120 V, 24 V, etc.), thus the
likelihood of a severe fire that could cause a large damage zone in a short period of time is deemed to be extremely low.
2. Fires inside the MCR cabinets are expected to generate smoke and odors that would catch the attention of the operators immediately which would prevent the fires from further progression to cause severe damage.

3. Fires are assumed not to spread and damage any components outside the cabinet. This assumption is based upon the cabinet configurations within the Surry MCR and the cabinet fire test results in Reference 9-4.
Thus, only fires that occur within a cabinet are considered. Furthermore, of the 451 actual fire events in the industrial fire event data base (Reference 9-1), there is no evidence of cabinet-to-cabinet fire propagation, and for the actual fire events, none of the cabinet fires that has a fire damage zone with a radius greater than 5 feet. For these reasons, a severe fire that would cover several control subsections (or clusters) is judged to be unlikely. To accommodate the uncertainty about this issue, it is conservatively assumed any fires with a fire radius more than 8 feet are excluded from this analysis. Another qualitative evaluation is based on the scenario impact which screened out fire scenario VB-7-DG3 due to no initiating event involved. The results of this preliminary screening are summarized in the third column of Table 9-3. 9 .2 Fire Scenario Quantification The control room fire scenarios are defined at the location level, i.e., each scenario includes fires that occur near or around a particular subsection in a control board and damage the components in that subsection. Once accounted by the fires, all components are assumed to fail in the subsection being considered. Therefore, the occurrence frequency for each scenario is the expected frequency of all possible fires that occur 9-9 NUREG/CR-6144
Control Room Fire Analysis within or around the components prescribed by the scenario and damage those components. Thus, the origin of fires may not be limited to the location where the components prescribed by a particular scenario are, it may also include other locations where fires, with significant severity, can propagate from their origins to damage the components. Moreover, only fires that occur within a cabinet are considered in fire propagation analysis and these fires are assumed not to spread and damage any components 8 ft beyond the fire origin or outside the cabinet. This assumption is based upon the cabinet configurations within the Surry MCR and the cabinet fire tests results in Reference 9-4. Furthermore, of the 451 actual fire events in the industrial fire events data base (Reference 9-1), there is no evidence of cabinet-to-cabinet fire propagation. Therefore, the occurrence frequency of the scenarios is limited to fires initiated within a control room cabinet, and fires occur within one section of a cabinet and propagate to another section in the cabinet. The objective of the scenario frequency assessment is to quantify the Surry-plant-specific MCR fire occurrence frequency for each scenario with the following considerations:
1. The scenario frequency must consistently account for actual empirical industry data and any Surry-plant-specific experience in control room related fires.

2. The scenario frequency must provide a conservative upper bound for the actual frequency of more detailed event scenarios that may eventually be developed for the location. In other words, the total scenario frequency* may be consistently subdivided to more realistically represent any specific event scenario in the location, if it is necessary to develop more detailed models for the location.
The MCR fire scenario occurrence frequency assessment is performed according to the following steps:
1. Collect plant information.

2. Specialize empirical industrial fire events data base.

3. Apply Bayesian update to obtain Surry-plant-specific MCR fire frequency.

4. Apportion the MCR fire frequency to individual cabinets.

5. Assess occurrence frequency for each fire scenario.
Descriptions of the above steps and their results are presented in the following sections. 9.2.1 Control Room Walkdown The Surry MCR was visited by a fire risk analysis team member at the beginning of the analysis. The purposes of the visit included:
1. to familiarize himself with the configuration of the MCR,

2. to examine plant-specific features that might affect the basis for and assumptions of the frequency assessment,

3. to evaluate the traffic level and transient fuel loadings inside the MCR, NUREG/CR-6144 9-10
Control Room Fire Risk Analysis
4. to identify combustible fuels outside the MCR and assess the possibility of fire propagation from these fuels into the MCR, and

5. to observe the back side of each control panel inside the MCR and assess a relative weight ratio of the amount of combustible fuels within each section of the panels observed.
The control room visit resulted in the following observations:
1. Relay cabinets were not found in the MCR, and there was no specific plant features that deviate the Surry MCR from a typical nuclear power plant control room.

2. Although the traffic levels and transient combustible loadings at the MCR during shutdown were considerably higher than during normal operation, all activities within the MCR were closely supervised by the operators. Any fire precursors would be suppressed before a more severe situation could develop. Therefore, the increase in traffic is not expected to significantly increase the frequency of transient fires in the MCR during shutdown.

3. There was no significant amount of transient and in-situ combustibles outside the MCR that have the severity of damaging components within the MCR, if ignited, and affect the operability of the MCR.
4. The first three columns of Table 9-4 summarize the relative weight ratio of combustibles (mainly cables and circuit boards) as observed from the back of the panels during the plarit visit. (Note that the rest of Table 9-4 contains materials to be presented in the following sections.)
9.2.2 Industrial Fire Events Specialization One important step in the fire occurrence frequency assessment for events that lack sufficient plant specific experience is to apply relevant empirical industrial experience of similar situations. Since there have not been any MCR fires at Surry during the past years of its operation, the revised Sandia generic fire event data base (Ref. 9-1) is used as a basis for the Bayesian update process. The data base contains 451 actual fire events that occurred in U.S. nuclear power plants. Only a portion of these events is relevant to the Surry control room operations; therefore, the data base has to be reviewed thoroughly to select only the relevant events to develop a "specialized" generic data base for the Surry MCR. The data base should account for design features of the Surry MCR and characteristics of the hazard sources that can be found in the MCR during shutdown. In order to be consistent with the fire frequency assessment performed for the rest of the plant areas at Surry, a "one-stage" Bayesian approach is taken and the specialized data base is developed accordingly. Two pieces of information are required for the formation of the specialized data base: (1) the number of relevant control room fire incidents, and (2) the number of years that the nuclear industry has accumulated for the control room experience. Fire events that occurred in control rooms were first selected from the empirical database provided in Reference 9-1 and their event summaries were reviewed to determine their applicability to the analysis. Fire 9-11 NUREG/CR-6144
Control Room Fire Analysis events that are classified to occur in control buildings in Reference 9-1 were also screened for their applicability to the Surry control room. This was to ensure the completeness of the fire event data search and to avoid any possible misclassification of fire location in Reference 9-1 since control room is a compartment in a control building. Several control room related fire events and all control building related fire events were screened out and were not included in the specialized database. This may be because of one the following reasons:
1. the event denotes a fire precursor condition only and no adjacent components were affected,

2. the events involves only component failures and such failures have been accounted for in the component failure database,

3. the event denotes a fire occur in a control building but outside of a control room,

4. the event relates to components or operations that are not present in the Surry MCR,

5. the event resulted from a sabotage or arson attempt,

6. the event descnbes a fire scenario due to violation of procedures, or

7. the event describes a fire scenario that occurred during construction and would not occur in the Surry MCR during shutdown.
After the empirical database was screened, seven control room events were included in the specialized database. It is noted that although some events occurred during normal power operation, it is judged that such events are also likely to occur during shutdown and, therefore, are included in the database. Table 9-5 summarized the events selected for the specialized database. In order to account for the number of years of control room experience that the nuclear industry has accumulated, the age of all U.S. commercial nuclear power plants are accumulated. The age is defined as the time between first criticality and the end of december 1989 (or date of decommissioning). Unlike other plant areas where plant operation during shutdown and at power may be different (in terms of combustible loadings), the control room at Surry is considered to have the same characteristics with respect to cabinet fires during both at power and shutdown. Therefore, there is no discount factor for the control room age even only the plant period during shutdown is the focus of this analysis. The fraction of time that the plant in shutdown mode will be considered in the subsequent fault tree and event tree analyses. It is noted that some of the power plants have more than one reactor unit but are equipped with only one control room, the amount of equipment included in such multiple unit control rooms is significantly higher than those with a single reactor unit. For instance, Surry has 2 reactor units but with only one control room. On the other hand, Beaver Valley has 2 control rooms for its 2 reactor units and Plant Trojan has one control room for its single unit reactor. Therefore, the age of control room experience must be carefully adjusted for on a site-to-site basis for the purpose of control room age estimation. Since most control room fires are NUREG/CR-6144 9-12
Control Room Fire Risk Analysis assumed to be originated within control cabinets and control rooms that handle multiple reactor units must have dedicated control cabinets for each unit, those control rooms that handle multiple units are counted as two individual control rooms in terms of control room experience accumulation.
For example, Surry 1 achieved criticality on 7/ln2 which gives 1751 years - time elapsed between 7/ln2 to 12/31/89 - for its control room age. Surry 2 achieved criticality on 3nn3, even though it shares the control room with Unit 1, Surry 2 has its dedicated control cabinets. The age of the Surry 2 control room is then equal to the time elapsed between March 3, 1973 and December 12, 1989, which is 16.83 years. Thus, as a site, Surry accumulated a total of 34.34 years of control room experience. 1bis approach is more realistic than the overly conservative approach which only counts the control room age of Surry 1 which would be 17.51 years, even though its control room has twice the amount of equipment than a typical one-unit-one-control-room plant. Table 9-6 summarizes the control room age estimation. . The data included in the table is provided by Reference 9-1. Tables 9-5 and 9-6 form the specialized industrial fire event data base for the Surry control room. There have been 7 relevant control room fire events within 1341.87 adjusted years of control room operation experience. It is also noted that control room experience at Surry 1 is not counted toward the total generic control room age. This data base provides the basis for the industry experience input to the hazard frequency analysis. 9.2.3 Control Room Fire Frequency Assessment Following the similar approach used in the fire occurrence frequency assessment for other plant areas, a one-stage Bayesian update is used to assess the posterior control room fire frequency. From the specialized data base, there have been seven events in 1341.87 years of control room operation experience. A noninformative lognormal prior is assumed. The industrial experience is entered as the median of the lognormal prior into the RISKMAN Data Analysis Module. The range factor of this generic prior is assumed to be seven. Table 9-7 displays the main characteristics of such prior. The characteristic values of the prior frequency are Mean = 1.05E-02 per year 5th percentile = 7.03E-04 per year 50th percentile = _5.09E-03 per year 95th percentile = 3.55E-02 per year The Surry-plant-specific experience on control room fires is zero fires in 17.51 years for the Unit 1 portion of the shared control room. Such plant specific information is used to update the prior and form the posterior distribution as shown in Table 9-8. The characteristic values of the posterior frequency for the Surry control room are Mean = 7.66E-03 per year 5th percentile = 6.18E-04 per year 50th percentile = 3.92E-03 per year 95th percentile = 2.00E-02 per year 9-13 NUREG/CR-6144
Control Room Fire Analysis 9.2.4 Control Room Fire Frequency Apportionment The fire frequency obtained in the previous section denotes the frequency of fires, at any size, occur within anywhere in the control room area that houses Unit 1 equipment. To assess the occurrence frequency of fires in different locations within the MCR, the control room fire frequency must be apportioned consistently to different areas. Because the control room is 24-hour manned, any transient combustible-initiated fire precursors or exposed cable fire precursors will be detected and extinguished immediately; thus, no damage to the vital equipment will be expected Furthermore, control room cabinet fire test results (Ref. 9-4) and past industrial experience (Ref. 9-1) indicated that (1) fires occurring in a cabinet are very unlikely to propagate and damage components in other cabinets, (2) fires have only occurred in electrical cabinets, and (3) transient combustible-initiated fires have never occurred in the control room area. Therefore, the fire frequency is conservatively apportioned only to the control cabinets in the MCR. Assuming that the density of combustible loadings (cables, circuit boards) within each cabinet is approximately the same and the equipment within a control board is uniformly distributed and is located in the front and back of the panels only, and the occurrence frequency of fire is directly proportional to the amount of combustibles (cables and circuit boards) inside each cabinets, the control room frequency is first apportioned into the different cabinets based on the panel surface areas. Miscellaneous panels and junction boxes that are not being considered in the scenarios are conservatively assumed to have 10% of the total MCR cabine surface area. Fire occurrence frequency of individual cabinets is then equal to the product of its normalized surface area ratio to the MCR total fire frequency (7.66E-3/yr). Table 9-9 summarizes the apportionment of the control room fire frequency into different cabinets based on the panel surface area ratios. Because of the additional information regarding the weight ratio (see the first three columns in Table 9-4) of the combustible loadings within Benchboards 1-1 and 1-2, and Vertical Boards 1-1 and 1-2, the frequencies of these boards are combined and then apportioned based on the combustible loading distributions observed during the control room visit. Table 9-4 presents the additional frequency apportionment of the individual sections within Benchboards 1-1 and 1-2, and Vertical Boards 1-1 and 1-2. Since the weight ratios obtained from the walkdown were taken for group subsections of the boards, the fire frequency for each group of panel subsections in Table 9-4 is further apportioned to each individual panel subsection. Following the assumption that the density of combustible loadings within each subsection in a panel section is approximately the same, the subsection frequency can then be obtained using normalized subsection length and the total fire frequency of the panel since all subsections have the same height. Tables 9-10 and 9-11 present the results of the fire frequency apportionment for each panel subsection in Benchboards 1-1 and 1-2, and Vertical Boards 1-1 and 1-2. 9.2.5 Scenario Frequency Quantification The frequencies presented in Tables 9-9 through 9-11 are the fire occurrence frequencies at each panel subsection. These fire occurring frequencies are then used to assess the occurrence frequency for the scenarios postulated in the previous section with the consideration of the severity of fires occurred at or near the panel subsection of the scenario interested. NUREG/CR-6144 9-14
Control Room Fire Risk Analysis The occurrence frequency of a fire scenario in this analysis is defined as the sum of the frequencies of all possible fires that occur in the subsection prescribed by the scenario, and fires that occur in other subsections in the same panel with significant severity that can damage the components in the subsection prescribed by the scenario. Thus, the occurrence frequency of scenario i in control panel board j, >-;. ;, can be expressed as: where A-MCR = the annual frequency of fire of any severity in the control room
= 7.66E-3 per year (see Table 9-8) = the conditional probability of fire of any severity occurring in control board j given a control room fire has occurred, i.e., the normalized ratio of each cabinet presented in Table 9-9 *.
F 0 s(i, j) = the conditional frequency, given a fire has occurred in control board j, of h:aving an impact on the area of interest of scenario i.
The conditional frequency F 05(i, j) can be expressed as F Gs(i, J) = L fs( i, j, k) !g(i, j, k) k where "k" represents a unique set of subsections in control board j for scenario i. The panel subsections included in "k" are those that if a frre occurs in those subsections, it will be severe enough to damage only the panel subsection prescnbed by scenario i but not so severe as to affect the adjacent panel subsections whose damage is modeled in other scenarios. The two factors fg and ~ ar_e generally known as the geometry factor and the severity factor of the subsections included in "k." The geometry factor and severity factor are interrelated (Ref. 9-5). If the scenario involves the whole control panel, the combined geometry and severity factor, FGs* is 1.0. The geometry factor is used to model t_he fraction of fires that can be initiated within a specific area, say area m, within the control panel j given that a fire has occurred within j. Thus, the product >..McRF/g(i, j, k) denotes the fire occurrence frequency of panel subsection k in control board j in the MCR, and k is a subsection that can damage the panel subsection prescnbed by scenario i. The geometry factor in this analysis is obtained by assessing the normalized length ratio of a subsection to the whole panel. The severity factor is used to model the fraction of fires at different levels of severity that can occur at subsection k, given a fire occurs at panel subsection k. Severity factors of different fire sizes are obtained from analyzing the fire sizes and damage levels of past actual control cabinet fire incidents reported in Reference 9-6. The severity factor is the probability of fires that can grow and propagate to certain fire sizes, 9-15 NUREG/CR-6144
Control Room Fire Analysis which is usually measured in terms of fire radius, without being mitigated. Figure 9-2 illustrates the severity factor function developed for control cabinet fires. Since the MCR at Surry is 24-hour manned, any fires should be detected and suppressed immediately, thus the severity function shown in Figure 9-2 is considered to be overly conservative when applied to the Surry MCR. For instance, it is very unlikely that a fire would be allowed to grow and propagate to a distance of more than 5 ft without being interfered with and extinguished. Therefore, the severity factor presented in Figure 9-2 is adjusted by conservatively assuming that the probability of fires with fire radius of over 8 ft is essentially zero. The following example illustrates the application of the geometry factor and severity factor in the assessment of scenario occurrence frequency: Scenario BB-1-1-1 postulates failure of RHR pumps A and B control in subsection 3Z of bBenchboard 1-1. Any fires within the benchboard with an appropriate severity that can damage components in subsection 3Z are included. It is conservatively assumed that all fires occurring in a cabinet will spread radially along the length of the cabinet with equal distance, and if a fire can propagate to subsection 3Z, regardless of the actual magnitude of the fire, subsection 3Z is declared to be damaged. To facilitate the analysis, it is further assumed that the center of a fire occurs in a panel subsection at the midpoint of the panel subsection. The set of fire origin subsections that can damage 3Z is a unique set. If a fire can damage not only 3Z but also subsections 3AE or 3W(3V, such subsection fire will not be included in the unique set for 3Z but will be included in scenarios BB-1-4 or BB-1-8. For example, in order for a fire originating at subsection 3AD to damage subsection 3Z, the fire size must have a fire radius of at least 3.8 ft (midpoint to midpoint distance between panel subsections 3Z and 3AD measured in Ref. 9-3.) With such fire sizes, subsection 3AE will also be damaged. Therefore, fire originating at 3AD, regardless of fire sizes, cannot be included in the unique set of fire origins for subsection 3Z in scenario BB-1-1. For subsection 3AB, the minimum fire size for a fire originating at 3AB in order to damage 3Z will be fires with radius of at least 2.1 ft from midpoint of subsection 3AB. However, if the fires originating at 3AB are allowed to grow and propagate radially, when the fire size reaches a radius of 3.0 ft or more, subsection 3AE will also be damaged. Thus, the effective fire radius for a fire originating in subsection 3AB to damage subsection 3Z is between 2.1 ft and 3.0 ft Therefore, the conditional probability of fires with a radius of 2.1 ft to 3.0 ft., given the fires occurring in subsection 3AB, must be used. Severity factor ~(2.1 ft) denotes the probability of fires with a radius of larger than 2.1 ft before being suppressed eventually, severity ~(3.0 ft) denotes the probability of fires with a radius of larger than 3.0 ft before being suppressed eventually. Similarly, the unique set of subsection fires that can damage subsection 3Z in scenario BB-1-1 includes the following fires: *
1. Fires initiating in subsection 3Z with an effective fire radius of 0.0 ft to 4.4 ft,

2. Fires initiating in subsection 3AB with an effective fire radius of 2.1 ft to 3.0 ft,

3. Fires initiating in subsection 3AA with an effective fire radius of 0.9 ft to 4.1 ft, and

4. Fires Initiating in subsection 3Y with an effective fire radius of 1.2 ft to 3.3 ft NUREG/CR-6144 9-16
Control Room Fire Risk Analysis The severity factors for each possible subsection fire in this unique set (k = 3Z, 3AB, 3AA, 3Y) are then assessed. Using Figure 9-3 and the notation in the above expressions, the severity factor for fires originating in subsection 3AB to damage 3Z in scenario BB-1-1 is:
~(i=BB-1-1, j=BB 1-1, k=3AB) = ~(2.1 ft.) - ~(3.0 ft.) = 0.23 - 0.08 = 0.153.
Table 9-12 summarizes the fire origin identification, effective fire radius estimation, and severity factor assessment for all scenarios retained from the preliminary screening. It is noted that the effective fire radius for the fires being considered cannot be greater than 8.0 ft. Once the unique set of subsection fires is identified for scenario BB-1-1, the geometry factor for each subsection (3Z, 3AB, 3AA, and 3Y) included in the unique set is required. The geometry factor for subsection 3Z is the fraction of fires that can occur at 3Z, given a fire has occurred in benchboard 1-1. This geometry factor is equal to the normalized ratio presented in Table 9-10, i.e., the geometry factor fg(i=BB-1-1, j=BB 1-1, k=3AB) is 0.613. The product of the total MCR fire frequency and the geometry factor of 3AB gives the frequency of fires occurring at subsection 3AB. The product of the geometry factor and severity factor for scenario BB-1-1, fgfs(i = BB-1-1, j =BB 1-1, k=3AB) denotes the frequency of fires occurring at subsection 3AB and damaging subsection 3Z only before being suppressed The total of such frequency for all panel subsections included in the unique set k (3Z, 3AB, 3AA, and 3Y) is the frequency of subsection 3Z being damaged in a configuration denoted by scenario BB-1-1. Since at least one hot short in the circuitry within subsection 3Z is required for the damage of the control of RHR pump A or other RHR system valves, the probability of hot short must be accounted for in scenario BB-1-1. In some scenarios, more than one hot short may be required to disable the control; it is conservatively assumed that only one hot short is needed to upper bound the hot short failure mode required for the fire scenario to occur. Using the hot short probability assessed in Appendix E, a failure mode factor of 0.1 is multiplied to the combined severity and geometry factor/fire frequency of scenario BB-1-1 to obtain the scenario occurrence frequency (2.04E-5 per year). For scenarios that do not require a hot short to disable the control descnbed in the scenarios, the failure mode factor is essentially equal to one. Similarly, the scenario occurrence frequency for other scenarios is assessed, and the summary of the scenario occurrence frequency assessment is presented in Table 9-12. It is noted that these scenario occurrence frequencies are on a yearly basis, the fraction of such* frequencies applied to mid-loop are considered in the subsequent fault tree/event tree analysis. 9 .3 References
1. "Nuclear Power Plant Fire Data Base," based on NUREG/CR-4586, SAND86-0300, and EPRI NP-3179, updated, June 1992.

2. 10 CPR 50 Appendix R Report, Surry Power Plant - Units 1 and 2, Revision 7, October 1990 9-17 NUREG/CR-6144
Control Room Fire Analysis
3. Outline Drawing 11448-RE-25A, Benchboard Section 1-1, Rev 15 Outline Drawing 11448-RE-25B, Benchboard Section 1-2, Rev 11 Outline Drawing 11448-RE-25C, Main Control Board Vertical Section 1-1, Rev 16 Outline Drawing 11448-RE-25D, Main Control Board Vertical Section 1-2, Rev 14 Outline Drawing 11448-RE-25E, Benchboard Section 1-3, Rev 3 Outline Drawing 11448-RE-25F, Benchboard Section 1-4, Rev 3 Outline Drawing 11448-RE-25G, Benchboard Section 1-5, Rev 1 Outline Drawing 11448-RE-25H, Benchboard Section 1-6, Rev 8 Outline Drawing 11448-RE-25J, Benchboard Section 1-7, Rev 2 Loading Table 11448-FE-llAD, Bus Distnbution Panel AC 1-1, Rev 3 Loading Table 11448-FE-llAE, Bus Distribution Panels DC 1-1 & DC 1-2, Rev 4

4. J.M. Chavez, "An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets, Part I - Cabinet Effects Tests," NUREG/CR-4527, Sandia National Laboratories, April 1987.

5. Pickard, Lowe and Garrick, Inc., "Seismic and Fire Risk Analysis for A Typical Japanese 4-Loop PWR Plant," PLG-0642, July 1988

6. PLG, Inc., "Database for Probabilistic Risk Assessment of Light Water Nuclear Power Plants - Fire Data," Volume, PLG-0500, Revision 0, October 1990.
NUREG/CR-6144 9-18
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u ('fl N I C ('fl "C ~ !:~ -, ('fl Figure 9.3 Illustration of Five Severity Factor Calculations for Scenario BB-1-1 9-21
Control Room Fire ACP-CRB-C0-14H13 14H13 ACP-CRB-C0-14H14 14H14 ACP-CRB-C0-14H15 14H15 ACP-CRB-C0-14J1 14J1 ACP-CRB-C0-14J11 14J11 ACP-CRB-C0-14J14 14J14 ACP-CRB-C0-14J16 14J16 ACP-CRB-C0-15H7 15H7 ACP-CRB-C0-15H8 15H8 ACP-CRB-C0-15H9 15H9 ACP-CRB-C0-15J7 15J7 ACP-CRB-C0-15J8 15J8 ACP-CRB-C0-15J9 15J9 ACP-CRB-C0-1I35 1135 CCW-AOV*OC-109B CCW-MDP*FR-CCP1A CCP1A CPC*MDP-FR*CC2B CC2B CPC*MDP*FR*SW10A SW10A CPC*MDP*FR*SW10B SW10B CPC*MDP*FS-CC2B CC2B CPC*MDP*FS*SW10B SW10B DCP*BDC-ST-BUS1A BUS1A Analysis Table 9.1 List of Surry Unit 1 Shutdown PRA Basic Events for Equipment with Controls in Main Control Room Basic Event COIII). Basic Event Designator ID Description ACP-BAC-ST-V1III V1III VITAL BUS 1III BUSWORK FAILURE ACP-BAC-ST-VB1I VB1 I VITAL BUS 11 BUSWORK FAILURE ACP-BAC-ST-VB1II VB1II VITAL BUS 111 BUSWORK FAILURE ACP-CRB-C0-14H1 14H1 AC CIRCUIT BREAKER 14H1 TRANSFERS OPEN (480V BUS 1H FEED BKR) AC CIRCUIT BREAKER 14H13 TRANSFERS OPEN (480V MCC 1H1*2 FEED BKR) AC CIRCUIT BREAKER 14H14 TRANSFERS OPEN (480V MCC 1H1*1 FEED BKR) AC CIRCUIT BREAKER 14H15 TRANSFERS OPEN (480V BUS 1H1 FEED BKR) AC CIRCUIT BREAKER 14J1 TRANSFERS OPEN (480V BUS 1J FEED BKR) AC CIRCUIT BREAKER 14J11 TRANSFERS OPEN (480V BUS 1J1 FEED BKR) AC CIRCUIT BREAKER 14J14 TRANSFERS OPEN (480V BUS 1J1-2 FEED BKR) AC CIRCUIT BREAKER 14J16 TRANSFERS OPEN (480V BUS 1J1-1 FEED BKR) AC CIRCUIT BREAKER 15H7 TRANSFERS OPEN (4KV BUS 1H TO 4KV/480V XFMR BKR) AC CIRCUIT BREAKER 15H8 TRANSFERS OPEN (BUS 1H FEED BKR, FROM BUS F) AC CIRCUIT BREAKER 15H9 TRANSFERS OPEN (4KV BUS 1H TO STUB BUS BKR) AC CIRCUIT BREAKER 15J7 TRANSFERS OPEN C4KV BUS 1J TO 4KV/480V XFMR BKR) AC CIRCUIT BREAKER 15J8 TRANSFERS OPEN (BUS 1J FEED BKR, FROM BUS D) AC CIRCUIT BREAKER 15J9 TRANSFERS (4KV BUS 1J TO STUB BUS BKR) VITAL BUS 1I AC CKT BREAKER 35 TRANSFERS OPEN ACP-CRB-C0-1II 1II AC CIRCUIT BREAKER TO 1II TRANSFERS OPEN ACP-CRB-CO-FE9AE FE9AE AC CIRCUIT BREAKER FE9AE TRANSFERS OPEN ACP*CRB*CO*FE9AF FE9AF AC CIRCUIT BREAKER FE9AF TRANSFERS OPEN ACP*CRB*CO*FE9AJ FE9AJ AC CIRCUIT BREAKER FE9AJ TRANSFERS OPEN ACP*CRB*CO*FE9AK FE9AK AC CIRCUIT BREAKER FE9AK TRANSFERS OPEN ACP*CRB*CO-FE9BE FE9BE AC CIRCUIT BREAKER FE9BE TRANSFERS OPEN ACP*CRB-CO-FE9BJ FE9BJ AC CIRCUIT BREAKER FE9BJ TRANSFERS OPEN ACP*CRB*CO*FE9BK FE9BK AC CIRCUIT BREAKER FE9BK TRANSFERS OPEN ACP*CRB-CO-III35 III35 VITAL BUS 1111 AC CKT BRKR 35 TRANSFERS OPEN CCW*AOV*OC*109A CC*TV*109A TRANSFERS CLOSED (CCW RETURN HEADER A TCV) CC-TV-1098 TRANSFERS CLOSED (CCW RETURN HEADER B TCV) 1*CC*P*1A FAILS TO RUN CCW*MDP*FR*CCP1B CCP1B MOP CC*P1B FAILS TO RUN CCW*MDP*FS*CCP1B CCP1B MOP CC*P1B FAILS TO START ON DEMAND CPC*AOV*FT*108B 1088 AOV TV*CC*108B FAILS TO OPEN CPC*AOV*FT*108C 108C AOV TV*CC*108C FAILS TO OPEN CPC*ICC*FA*SWPBS SWPBS NO ACTUATION SIGNAL TO START SW PUMP 108 CPC*MDP*FR*CC2A CC2A MOP CC2A FAILS TO RUN AS LONG AS CHRGNG PUMPS (HPI PUMP COOLING WATER PUMP A) MOP CC2B FAILS TO RUN FOR 24 HOURS (HPI PUMP COOLING WATER PUMP B) MOP SW10A FAILS TO RUN AS LONG AS PUMPS CHPI PUMP COOLING SERVICE WATER PUMP A) MOP SW10B FAILS TO RUN FOR 24 HRS (HPI PUMP COOLING SERVICE WATER PUMP B) MOP CC2B FAILS TO START (HPI PUMP COOLING WATER PUMP B) MOP SW10B FAILS TO START (HPI PUMP COOLING SERVICE WATER PUMP B) 125V DC BUS 1A BUSWORK FAILURE DCP*BDC-ST*BUS1B BUS1B 125V DC BUS 18 BUSWORK FAILURE DCP-CRB*C0-19 19 DC CIRCUIT BREAKER 19 TRANSFERS OPEN DCP*CRB*C0-20 20 DC CIRCUIT BREAKER 20 TRANSFERS OPEN DCP-CRB*C0-23 23 DC CIRCUIT BREAKER 23 TRANSFERS OPEN DCP*CRB-C0-24 24 DC CIRCUIT BREAKER 24 TRANSFERS OPEN HPI*MDP-FR-1A6HR 1A6HR CHARGING PUMP CH1A FAILS TO RUN FOR 6 HOURS HPI*MDP-FR*1B6HR 1B6HR CHARGING PUMP CH1B FAILS TO RUN FOR 6 HOURS HPI-MDP*FR-1C12H 1C12H CHARGING PUMP CH1C FAILS TO RUN FOR 12 HOURS HPI*MDP-FS*CH1B CH1B CHARGING PUMP CH1B FAILS TO START ON DMD HPI*MDP*FS*CH1C CH1C CHARGING PUMP CH1C FAILS TO START ON DMD HPI *MOD-FT*101B MOTOR OPERATED DAMPER MOD-1018 FAILS TO OPEN NUREG/CR-6144 9-22
Control Room Fire Analysis Table 9.1 List of Sun-y Unit 1 Shutdwn PRA Basic Events for Equipment with Controls in Main Control Room (continued) Basic Event Coq,. Basic Event Designator ID Description HPl*MOD*FT-101C MOTOR OPERATED DAMPER MOD-101C FAILS TO OPEN HPI-MOV-FT-1115B 1115B HPI MOV 1115B FAILS TO OPEN ON DEMAND (RWST SUPPLY TO HPI PUMP ISOL VLV) HPI-MOV-FT-1115C 1115C HPI MOV 1115C FAILS TO CLOSE (VCT SUPPLY TO HPI PUMP ISOL VLV) HPI-MOV-FT-1115D 1115D HPI MOV 1115D FAILS TO OPEN ON DEMAND (RWST SUPPLY TO HPI PUMP ISOL VLV) HPI-MOV-FT-1115E 1115E HPI MOV 1115E FAILS TO CLOSE (VCT SUPPLY TO HPI PUMP ISOL VLV) HPI-MOV-FT-1867C 1867c HPI MOV 1867C FAILS TO OPEN ON DEMAND (HPI PUMP DISCH TO COLD LEGS ISOL VLV) HPI-MOV-FT-18670 18670 HPI MOV 18670 FAILS TO OPEN ON DEMAND (HPI PUMP DISCH TO COLD LEGS ISOL VLV) HPI-MOV-PG-1269A 1269A MOTOR OPERATED VALVE 1269A PLUGGED (HPI PUMP B SUCTION ISOL VLV) HPI-MOV-PG-1270A 1270A MOTOR OPERATED VALVE 1270A PLUGGED (HPI PUMP C SUCTION ISOL VLV) HPI-MOV-PG-1286B 1286B MOTOR OPERATED VALVE 1286B PLUGGED (HPI PUMP B DISCH ISOL VLV) HPI-MOV-PG-1286C 1286C MOTOR OPERATED VALVE 1286C PLUGGED (HPI PUMP C DISCH !SOL VLV) IAS-AOV-FT-TV126 IA*TV-126 FAILS TO OPEN IAS-AOV-PG-TV125 IA*TV-125 PLUGGED IAS-CPS-FR-IAC*1 TURBINE BUILDING IA COMPRESSOR IA*C-1 FAILS TO RUN IAS-CPS-FS-IAC-1 TURBINE BUILDING IA COMPRESSOR IA*C-1 FAILS TO START LPI-MDP-FR-1A1HR 1A1HR LPI MDP SI1A FAILS TO RUN FOR 1 HOUR LPI-MDP-FR-1B1HR 1B1HR LP! MDP SI1B FAILS TO RUN FOR 1 HOUR LPI-MDP-FS-SI 1A SI1A LP! MDP SI1A FAILS TO START ON DEMAND LPI-MDP-FS-SI1B SI1B LPI MDP SI1B FAILS TO START ON DEMAND LPI-MOV-CC-1890C LPI MOV* 1890C FAILS TO OPEN (LPI PUMP DISCH TO COLD LEGS ISOL VLV) LPI-MOV-PG-1862A 1862A LPI MOTOR OPER VALVE 1862A PLUGGED (LPI PUMP A SUCTION ISOL VLV) LPI-MOV-PG-1862B 1862B LPI MOTOR OPER VALVE 1862B PLUGGED (LPI PUMP B SUCTION ISOL VLV) LPI-MOV-PG-1864A 1864A LPI MOTOR OPERATED VALVE 1864A PLUGGED (LPI PUMP A DISCH TO COLD LEGS ISOL VLV) LPI-MOV-PG-1864B 1864B LPI MOTOR OPERATED VALVE 1864B PLUGGED (LPI PUMP B DISCH TO COLD LEGS ISOL VLV) MSS-AOV-FC-101A 101A SG A PORV BLOCKED MSS-AOV-FC-101B 101B SG B PORV BLOCKED MSS-AOV-FC-101C 101C SG C PORV BLOCKED MSS-AOV-FT-101A 101A SG A PORV FAILS TO OPEN MSS-AOV-FT-101B 101B SG B PORV FAILS TO OPEN MSS-AOV-FT-101C 101C SG C PORV FAILS TO OPEN MSS-NRV-FT-101A NRV FAILS TO OPEN (MAIN STEAM NON-RETURN VLV A) MSS-NRV-FT-101B NRV FAILS TO OPEN (MAIN STEAM NON-RETURN VLV B) MSS-NRV-FT-101C NRV FAILS TO OPEN (MAIN STEAM NON-RETURN VLV C) MSS-STV-FT-101A MSTV FAILS TO OPEN (MAIN STEAM TRIP VLV A) MSS-STV-FT-101B MSTV FAILS TO OPEN (MAIN STEAM TRIP VLV B) MSS-STV-FT-101C MSTV FAILS TO OPEN (MAIN STEAM TRIP VLV C) OEP-BAC-ST°FDRD FDRD RESERVE STATION SERVICE FEEDER D BUSWORK FAllS CBKR 15D1, RSS XFMR TO BUS D) OEP-BAC-ST-FDRF FDRF RESERVE STATION SERVICE FEEDER F BUSWORK FAILS (BKR 15F1, RSS XFMR TO BUS F) OEP-CRB-FT-15H3 15H3 DIESEL GEN #1 CKT BRKR 15H3 FAILS TO CLOSE (BUS 1H FEED BKR, FROM DG 1) OEP-CRB-FT-15J3 15J3 DIESEL GEN #3 CKT BRKR 15J3 FAILS TO CLOSE (BUS 1J FEED BKR, FROM DG 3) OEP-DGN-FC-DG3U2 DG3U2 DIESEL GEN #3 UNAVAIL, ALIGNED TO UNIT 2 OEP-DGN-FR-DG01 DG01 DG 1 FAILS TO RUN 1 HR OEP-DGN- FR-DG03 DG03 DG 3 FAILS TO RUN 1 HR OEP-DGN-FS-DG01 DG01 DIESEL GENERATOR #1 FAILS TO START OEP-DGN- FS-DG03 DG03 DIESEL GENERATOR #3 FAILS TO START PCS-AOV-FT-BYPA BYPA A BYPASS VALVE FAIL TO OPEN TCV 107/BA PCS-AOV-FT-BYPB BYPB B BYPASS VALVE FAIL TO OPEN TCV 107/BB PCS*AOV*FT*MSTVA MSTVA A BYPASS VALVE TCV 105/6A FAILS TO OPEN PCS-AOV*FT*MSTVB MSTVB B BYPASS VALVE TCV 105/6B FAILS TO OPEN PCS-AOV*PG*BYPA A BYPASS TCV 107/BA PLUGGED PCS*AOV*PG*BYPB B BYPASS TCV 107/BB PLUGGED PCS*AOV*PG*MSTVA MSTVA A BYPASS VALVE TCV 105/6A PLUGGED 9-23 NUREG/CR-6144
Control Room Fire Analysis Table 9.1 List of SuITY Unit 1 Shutdown PRA Basic Events for Equipment with Controls in Main Control Room (continued) Basic Event Coq>. Basic Event Designator ID Description PCS*AOV*PG*MSTVB MSTVB B BYPASS VALVE TCV 105/66 PLUGGED RHR*AOV*C0-1758 AOV 1758 OPENS SPURIOUSLY (RHR HX FCV) RHR*AOV*OC-1758 1758 HCV-1758 FAILS SHUT (RHR HX FCV) RHR*AOV-00*1605 1605 FCV-1605 TRANSFERS FULLY OPEN AND REMAINS OPEN (RHR HX BYPASS FCV) RHR*MDP*FR*A24HR A24HR RHR MDP 1A FAILS TO RUN 24 HOURS RHR*MDP*FR*B24HR B24HR RHR MDP 1B FAILS TO RUN FOR 24 HOURS RHR*MDP*FS*RHR1A RHR1A RHR MDP 1A FAILS TO START ON DEMAND RHR*MDP*FS*RHR1B RHR1B RHR MDP 1B FAILS TO START ON DEMAND SAS*CPS*FR*1SAC1 SERVICE AIR COMPRESSOR 1-SA-C-1 FAILS TO RUN SAS*CPS*FR-2SAC1 SERVICE AIR COMPRESSOR 2-SA-C-1 FAILS TO RUN SEMI-VIT-BUS FAILURE OF SEMI-VITAL BUS SP*SI*CLSI-HE SPURIOUS SI/CLSI SIGNAL SW-AOV-FT-SW263 1-SW-263 FAILS CLOSED ON LOSS OF OFFSITE POWER NUREG/CR-6144 9-24
Control Room Fire Analysis Table 9.2 Control Locatiom for Surry Unit 1 Shutdown Aa:ident Mitigation F.quipment Location of Basic Event Ccap. Basic Event Control Designator ID Description BB 1*1 (3AE) HPI*MOV*FT*1867C 1867C HPI MOV 1867C FAILS TO OPEN ON DEMAND (HPI PUMP DISCH TO COLD LEGS ISOL VLV) BB 1*1 (3AE) HPl*MOV*FT*1867D 1867D HPI MOV 1867D FAILS TO OPEN ON DEMAND (HPI PUMP DISCH TO COLD LEGS ISOL VLV) BB 1*1 (3AE) LPl*MDP*FR*1A1HR 1A1HR LPI MDP Sl1A FAILS TO RUN FOR 1 HOUR BB 1*1 (3AE) LPl*MDP*FR*1B1HR 1B1HR LPI MDP Sl1B FAILS TO RUN FOR 1 HOUR BB 1*1 (3AE) LPl*MDP*FS*Sl1A Sl1A LPI MDP Sl1A FAILS TO START ON DEMAND BB 1*1 (3AE) LPl*MDP*FS*Sl1B Sl1B LPI HOP Sl1B FAILS TO START ON DEMAND BB 1*1 (3AE) LPl*MOV*CC*1890C LPI MOV* 1890C FAILS TO OPEN (LPI PUMP DISCH TO COLD LEGS ISOL VLV) BB 1*1 (3AE) LPI *MOV*PG* 1862A 1862A LPI MOTOR OPER VALVE 1862A PLUGGED (LPI PUMP A SUCTION ISOL VLV) BB 1*1 (3AE) LPl*MOV*PG*1862B 18628 LPI MOTOR OPER VALVE 1862B PLUGGED (LPI PUMP B SUCTION ISOL VLV) BB 1*1 (3AE) LPI-MOV*PG*1864A 1864A LPI MOTOR OPERATED VALVE 1864A PLUGGED (LPI PUMP A DISCH TO COLD LEGS ISOL VLV) BB 1-1 (3AE) LPI-MOV-PG-1864B 1864B LPI MOTOR OPERATED VALVE 1864B PLUGGED (LPI PUMP B DISCH TO COLO LEGS ISOL VLV) BB 1-1 (3AE) SP-Sl*CLSI-HE SPURIOUS SI/CLSI SIGNAL BB 1-1 (3V) CPC-ICC*FA-SWPBS SWPBS NO ACTUATION SIGNAL TO START SW PUMP 108 BB 1-1 (3V) CPC-MDP-FR*CC2B CC2B MDP CC2B FAILS TO RUN FOR 24 HCXJRS (HPI PUMP COOLING WATER PUMP B) BB 1-1 (3V) CPC-MDP-FR*SW10B SW10B MOP SW10B FAILS TO RUN FOR 24 HRS (HPI PUMP COOLING SERVICE WATER PUMP B) BB 1-1 (3V) CPC-MDP-FS*CC2B CC2B MOP CC2B FAILS TO START (HPI PUMP COOLING WATER PUMP B) BB 1-1 (3V) CPC-MDP-FS*SW10B SW10B MOP SW10B FAILS TO START (HPI PUMP COOLING SERVICE WATER PUMP B) BB 1-1 (3W) CPC-MDP-FR-CC2A CC2A MOP CC2A FAILS TO RUN AS LONG AS CHRGNG PUMPS (HPI PUMP COOLING WATER PUMP A) BB 1-1 (3W) CPC-ll>P-FR*SW10A SW10A MOP SW10A FAILS TO RUN AS LONG AS PUMPS (HPI PUMP COOLING SERVICE WATER PUMP A) BB 1-1 (3W) HPl*MDP-FR*1A6HR 1A6HR CHARGING PUMP CH1A FAILS TO RUN FOR 6 HOURS BB 1*1 (3W) HPl*MDP-FR*1B6HR 1B6HR CHARGING PUMP CH1B FAILS TO RUN FOR 6 HOURS BB 1-1 (3W) HPl-ll>P-FR-1C12H 1C12H CHARGING PUMP CH1C FAILS TO RUN FOR 12 HOURS BB 1-1 (3W) HPI-MDP-FS-CH1B CH1B CHARGING PUMP CH1B FAILS TO START ON DMD BB 1-1 (3W) HPI-MDP-FS*CH1C CH1C CHARGING PUMP CH1C FAILS TO START ON DMD BB 1-1 (3W) HPl*MOV-FT*1115B 11158 HPI MOY 1115B FAILS TO OPEN ON DEMAND (RWST SUPPLY TO HPI PUMP ISOL VLV) BB 1*1 (3W) HPI-MOV*FT-1115C 1115C HPI MOY 1115C FAILS TO CLOSE (VCT SUPPLY TO HPI PUMP ISOL VLV) BB 1-1 (3W) HPI-MOV*FT-1115D 1115D HPI MOV 1115D FAILS TO OPEN ON DEMAND (RWST SUPPLY TO HPI PUMP ISOL VLV) BB 1-1 (3W) HPI-MOV*FT*1115E 1115E HPI MOV 1115E FAILS TO CLOSE (VCT SUPPLY TO HPI PUMP ISOL VLV) BB 1-1 (3W) SP*Sl*CLSI*HE SPURIOUS SI/CLSI SIGNAL BB 1-1 (3Z) RHR-AOV*C0*1758 ADV 1758 OPENS SPURIOUSLY (RHR HX FCV) BB 1-1 (3Z) RHR-AOV*OC-1758 1758 HCV*1758 FAILS SHUT (RHR HX FCV) BB 1-1 C3Z) RHR-AOV*00*1605 1605 FCV-1605 TRANSFERS FULLY OPEN ANO REMAINS OPEN (RHR HX BYPASS FCV) BB 1-1 C3Z) RHR-MDP-FR*A24HR A24HR RHR MOP 1A FAILS TO RUN 24 HOURS BB 1-1 C3Z) RHR*MDP*FR*B24HR B24HR RHR MOP 18 FAILS TO RUN FOR 24 HOURS BB 1-1 (3Z) RHR-MDP*FS*RHR1A RHR1A RHR MDP 1A FAILS TO START ON DEMAND BB 1*1 (3Z) RHR-MDP*FS*RHR1B RHR1B RHR MDP 18 FAILS TO START ON DEMAND BB 1-2 (3A) OEP-BAC-ST-FDRF FDRF RESERVE STATION SERVICE FEEDER F BUSWORK FAILS (BKR 15F1, RSS XFMR TO BUS F) BB 1-2 (38) OEP*BAC*ST*FDRD FDRD RESERVE STATION SERVICE FEEDER D BUSWORK FAILS (BKR 15D1, RSS XFMR TO BUS D) BB 1-2 (3F) MSS*AOV-FC*101B 1018 SG B PORV BLOCKED BB 1*2 (3F) MSS*AOV*FC*101C 101C SG C PORV BLOCKED BB 1*2 (3F) MSS-AOV-FT*101B 1018 SG B PORV FAILS TO OPEN BB 1*2 (3F) MSS-AOV*FT*101C 101C SG C PORV FAILS TO OPEN
BB 1*2 (3G) MSS-AOV*FC*101A 101A SG A PORV BLOCKED 9-25 NUREG/CR-6144
Control Room Fire Analysis Table 9.2 Control Locations for Surry Unit 1 Shutdown Accident Mitigation Equipment (continued) Location of Basic Event c~. Basic Event Control Designator ID Description BB 1-2 (3G) MSS*AOV*FT*101A 101A SG A PORV FAILS TO OPEN BB 1*2 (3G) MSS*STV*FT*101B MSTV FAILS TO OPEN (MAIN STEAM TRIP VLV B) BB 1-2 (3G) MSS*STV*FT*101C MSTV FAILS TO OPEN (MAIN STEAM TRIP VLV C) BB 1-2 (3H) MSS*STV*FT*101A MSTV FAILS TO OPEN (MAIN STEAM TRIP VLV A) BB 1*2 (3H) PCS*AOV*FT*BYPA BYPA A BYPASS VALVE FAIL TO OPEN TCV 107/SA BB 1*2 (3H) PCS*AOV*FT*BYPB BYPB B BYPASS VALVE FAIL TO OPEN TCV 107/SB BB 1*2 (3H) PCS*AOV*FT*MSTVA MSTVA A BYPASS VALVE TCV 105/6A FAILS TO OPEN BB 1-2 (3H) PCS*AOV*FT*MSTVB MSTVB B BYPASS VALVE TCV 105/6B FAILS TO OPEN BB 1*2 (3H) PCS*AOV*PG*BYPA A BYPASS TCV 107/SA PLUGGED BB 1*2 C3H) PCS*AOV*PG*BYPB B BYPASS TCV 107/SB PLUGGED BB 1-2 (3H) PCS*AOV*PG*MSTVA MSTVA A BYPASS VALVE TCV 105/6A PLUGGED BB 1*2 (3H) PCS*AOV*PG*MSTVB MSTVB B BYPASS VALVE TCV 105/6B PLUGGED VB 1*1 (3BC) HPI*MOV*PG*1270A 1270A MOTOR OPERATED VALVE 1270A PLUGGED (HPI PUMP C SUCTION ISOL VLV) VB 1-1 (3BC) HPI*MOV*PG*1286C 1286C MOTOR OPERATED VALVE 1286C PLUGGED CHPI PUMP C DISCH ISOL VLV) VB 1*1 (3BD) HPI*MOV*PG*1269A 1269A MOTOR OPERATED VALVE 1269A PLUGGED (HPI PUMP B SUCTION ISOL VLV) VB 1*1 (3BD) HPI*MOV*PG*1286B 1286B MOTOR OPERATED VALVE 1286B PLUGGED CHPI PUMP B DISCH ISOL VLV) VB 1*1 C3BH) CCW*AOV*OC*109A CC*TV*109A TRANSFERS CLOSED (CCW RETURN HEADER A TCV) VB 1*1 (3BH) CCW*AOV~OC*109B CC*TV*109B TRANSFERS CLOSED (CCW RETURN HEADER B TCV) VB 1*1 (BACK) SEMI*VIT*BUS FAILURE OF SEMI-VITAL BUS VB 1*2 (3AQ) CCW*MDP*FR*CCP1A CCP1A 1*CC*P*1A FAILS TO RUN VB 1*2 (3AQ) CCW*MDP*FR*CCP1B CCP1B MDP CC*P1B FAILS TO RUN VB 1*2 C3AQ) CCW*MDP*FS*CCP1B CCP1B MDP CC*P1B FAILS TO START ON DEMAND VB 1*2 (3AS) MSS*NRV*FT*101A NRV FAILS TO OPEN (MAIN STEAM NON-RETURN VLV A) VB 1*2 (3AS) MSS*NRV*FT*101B NRV FAILS TO OPEN (MAIN STEAM NON-RETURN VLV B) VB 1*2 (3AS) MSS*NRV*FT*101C NRV FAILS TO OPEN (MAIN STEAM NON-RETURN VLV C) VB 1-6 ACP*CRB*C0*15H8 15H8 AC CIRCUIT BREAKER 15H8 TRANSFERS OPEN (BUS 1H FEED BKR, FROM BUS F) VB 1*6 OEP*CRB*FT*15H3 15H3 DIESEL GEN #1 CKT BRKR 15H3 FAILS TO CLOSE (BUS 1H FEED BKR, FROM DG 1) VB 1-6 OEP*DGN*FR*DG01 DG01 DG 1 FAILS TO RUN 1 HR VB 1*6 OEP*DGN*FS*DG01 DG01 DIESEL GENERATOR #1 FAILS TO START VB 1*7 ACP*CRB*C0*15J8 15J8 AC CIRCUIT BREAKER 15J8 TRANSFERS OPEN (BUS 1J FEED BKR, FROM BUS D) VB 1*7 OEP*CRB*FT*15J3 15J3 DIESEL GEN #3 CKT BRKR 15J3 FAILS TO CLOSE (BUS 1J FEED BKR, FROM DG 3) VB 1*7 OEP*DGN*FC*DG3U2 DG3U2 DIESEL GEN #3 UNAVAIL, ALIGNED TO UNIT 2 VB 1*7 OEP*DGN-FR*DG03 DG03 DG 3 FAILS TO RUN 1 HR VB 1-7 OEP*DGN* FS*DG03 DG03 DIESEL GENERATOR #3 FAILS TO START VB Cab 1-II ACP*BAC*ST-VB1II VB1II VITAL BUS 1II BUSWORK FAILURE VB Cab 1-II ACP*CRB*C0-1 II 1II AC CIRCUIT BREAKER TO 111 TRANSFERS OPEN Not in MCR ACP*BAC*ST*V1III V1III VITAL BUS 1III BUSWORK FAILURE NOt in MCR ACP*BAC*ST*VB1I VB1I VITAL BUS 1I BUSWORK FAILURE Not in MCR ACP*CRB*C0*1I35 1135 VITAL BUS 1I AC CKT BREAKER 35 TRANSFERS OPEN Not in MCR ACP*CRB*CO*III35 I II35 VITAL BUS 1III AC CKT BRKR 35 TRANSFERS OPEN NUREG/CR-6144 9-26
Control Room Fire Analysis Table 9.2 Control Locations for Surry Unit 1 Shutdown Accident Mitigation Equipment (continued) Location of Basic Event Comp. Basic Event Control Designator ID Description Not in MCR ACP-CRB-C0-14H1 14H1 AC CIRCUIT BREAKER 14H1 TRANSFERS OPEN (480V BUS 1H FEED BKR) Not in MCR ACP-CRB-C0-14H13 14H13 AC CIRCUIT BREAKER 14H13 TRANSFERS OPEN (480V MCC 1H1-2 FEED BKR) Not in MCR ACP-CRB-C0-14H14 14H14 AC CIRCUIT BREAKER 14H14 TRANSFERS OPEN (480V MCC 1H1-1 FEED BKR) Not in MCR ACP-CRB-C0-14H15 14H15 AC CIRCUIT BREAKER 14H15 TRANSFERS OPEN (480V BUS 1H1 FEED BKR) Not in MCR ACP-CRB-C0-14J1 14J1 AC CIRCUIT BREAKER 14J1 TRANSFERS OPEN (480V BUS 1J FEED BKR) Not in MCR ACP-CRB-C0-14J11 14J11 AC CIRCUIT BREAKER 14J11 TRANSFERS OPEN (480V BUS 1J1 FEED BKR) Not in MCR ACP-CRB-C0-14J14 14J14 AC CIRCUIT BREAKER 14J14 TRANSFERS OPEN (480V BUS 1J1-2 FEED BKR) Not in MCR ACP-CRB-C0-14J16 14J16 AC CIRCUIT BREAKER 14J16 TRANSFERS OPEN (480V BUS 1J1-1 FEED BKR) Not in MCR ACP-CRB-C0-15H7 15H7 AC CIRCUIT BREAKER 15H7 TRANSFERS OPEN C4KV BUS 1H TO 4KV/480V XFMR BKR) Not in MCR ACP-CRB-C0-15H9 15H9 AC CIRCUIT BREAKER 15H9 TRANSFERS OPEN (4KV BUS 1H TO STUB BUS BKR) Not in MCR ACP-CRB-C0-15J7 15J7 AC CIRCUIT BREAKER 15J7 TRANSFERS OPEN (4KV BUS 1J TO 4KV/480V XFMR BKR) Not in MCR ACP-CRB-C0-15J9 1SJ9 AC CIRCUIT BREAKER 15J9 TRANSFERS (4KV BUS 1J TO STUB BUS BKR) Not in MCR ACP-CRB-CO-FE9AE FE9AE AC CIRCUIT BREAKER FE9AE TRANSFERS OPEN Not in MCR ACP-CRB-CO-FE9AF FE9AF AC CIRCUIT BREAKER FE9AF TRANSFERS OPEN Not in MCR ACP-CRB-CO-FE9AJ FE9AJ AC CIRCUIT BREAKER FE9AJ TRANSFERS OPEN Not in MCR ACP-CRB-CO-FE9AK FE9AK AC CIRCUIT BREAKER FE9AK TRANSFERS OPEN Not in MCR ACP-CRB-CO-FE9BE FE9BE AC CIRCUIT BREAKER FE9BE TRANSFERS OPEN in MCR ACP-CRB-CO-FE9BJ FE9BJ AC CIRCUIT BREAKER FE9BJ TRANSFERS OPEN in MCR ACP-CRB-CO-FE9BK FE9BK AC CIRCUIT BREAKER FE9BK TRANSFERS OPEN in MCR CPC-AOV-FT-108B 108B AOV TV-CC-108B FAILS TO OPEN Not, in MCR CPC-AOV-FT-108C 108C AOV TV-CC-108C FAILS TO OPEN Not in MCR DCP-BDC-ST-BUS1A BUS1A 125V DC BUS 1A BUSWORK FAILURE Not in MCR DCP-BDC-ST-BUS1B BUS1B 125V DC BUS 18 BUSWORK FAILURE Not in MCR DCP-CRB-C0-19 19 DC.CIRCUIT BREAKER 19 TRANSFERS OPEN Not in MCR DCP-CRB-C0-20 20 DC CIRCUIT BREAKER 20 TRANSFERS OPEN Not in MCR DCP-CRB-C0-23 23 DC CIRCUIT BREAKER 23 TRANSFERS OPEN Not in MCR DCP-CRB-C0-24 24 DC CIRCUIT BREAKER 24 TRANSFERS OPEN Not in MCR HPI-MOD-FT-101B MOTOR OPERATED DAMPER MOD-101B FAILS TO OPEN Not in MCR HPI-MOD-FT-101C MOTOR OPERATED DAMPER MOD-101C FAILS TO OPEN Not in MCR IAS-AOV-FT-TV126 IA-TV-126 FAILS TO OPEN Not in MCR IAS-AOV-PG-TV125 IA-TV-125 PLUGGED Not in MCR IAS-CPS-FR-IAC-1 TURBINE BUILDING IA COMPRESSOR IA-C-1 FAILS TO RUN Not in MCR IAS-CPS-FS-IAC-1 TURBINE BUILDING IA COMPRESSOR IA-C-1 FAILS TO START Not in MCR SAS-CPS-FR-1SAC1 SERVICE AIR COMPRESSOR 1-SA-C-1 FAILS TO RUN Not in MCR SAS-CPS-FR-2SAC1 SERVICE AIR COMPRESSOR 2-SA-C-1 FAILS TO RUN Not in MCR SW-AOV-FT-SW263 1-SW-263 FAILS CLOSED ON LOSS OF OFFSITE POWER
9-27 NUREG/CR-6144
Control Room Fire Analysis Table 9.3 Scenario Description and Preliminary Screening Results Scenario Scenario Impact Retained for Hot Short Designator Frequency Required? Assessment? BB-1-1 Small fires in Benchboard 1-1, around subsection 3Z, fail Yes Yes RHR pumps A and B BB-1-2 Small fires in Benchboard 1-1 or 1-2, around subsections Yes Yes 3W and 3V, fail HHSI pumps A, B, and C BB-1-3 Small fires in Benchboard 1-2, around subsections 3B and Yes Yes 3A, fail feed breakers from RSS transformers to 4 kV buses F and D BB-1-4 Medium fires in Benchboard 1-1, covering subsections Yes Yes 3AE and 3Z, fail (a) RHR pumps A and B, (b) HHSI pump discharge to the cold leg isolation valves, (c) LHSI pumps A and B, ( d) recirculation sprays BB-1-5 Large fires in Benchboard 1-1, covering subsections 3AE, 3Z, 3W, and 3V, fail (a) RHR pumps A and B, (b) HHSI pumps A, B, and C, (c) LHSI pumps A and B, ( d) Yes Yes i. recirculation sprays BB-1-6 ~rge fires in Benchboard 1-1, covering subsections 3AE, No Yes 3Z, 3W, 3V, 3H, 3G, and 3F, fail (a) RHR pumps A and B, (b) HHSI pumps A, B, and C, (c) LHSI pumps A and B, ( d) steam generator steam relief path, (e) recirculation sprays BB-1-7 Large fires in Benchboard 1-1 or 1-2, covering subsections No Yes 3AE, 3Z, 3W, 3V, 3H, 3G, 3F, 3B, and 3A, fail (a) RHR pumps A and B, (b) HHSI pumps A, B, and C, (c) LHSI pumps A and B, ( d) SG steam relief path, ( e) feed breakers from RSS transformers to 4 kV buses F and D,
* (f) AFW system, (g) recirculation sprays BB-1-8 Medium fires in Benchboard 1-1, covering subsections 3Z, Yes Yes 3W, and 3V, fail (a) RHR pumps A and B and (b) HHSI pumps A, B, and C BB-1-9 Large fires in Benchboard 1-1 or 1-2, covering subsections Yes Yes 3Z, 3W, 3V, 3H, 3G, and 3F, fail (a) RHR pumps A and B, (b) HHSI pumps A, B, and C, and (c) steam generator steam relief path NUREG/CR-6144 9-28
Control Room Fire Analysis Table 9.3 Scenario Description and Preliminary Screening Results (continued)
Scenario Scenario Impact Retained for Hot Short Designator Frequency Required? Assessment? BB-1-10 Large fires in Benchboard 1-2, covering subsections 32, No Yes 3W, 3V, 3H, 3G, 3F, 3B, and 3A, fail (a) RHR pumps A and B, (b) HHSI pumps A, B, and C, (c) steam generator steam relief path, (d) feed breakers from RSS transformers to 4 kV buses F and D, (e) AFW system BB-1-11 Medium fires in Benchboard 1-1 or 1-2, covering Yes Yes subsections 3W, 3V, 3H, 3G, and 3F, fail (a) HHSI pumps A, B, and C and (b) steam generator steam relief path BB-1-12 Large fires in Benchboard 1-2, covering subsections 3W, Yes Yes 3V, 3H, 3G, 3F, 3B, and 3A, fail (a) HHSI pumps A, B, and C, (b) steam generator steam relief path, (c) feed breakers from RSS transformers to 4 kV buses F and D, (d) AFW system BB-1-13 Large fires in Benchboard 1-2, covering subsections 3H, Yes Yes 3G, 3F, 3B, and 3A, fail (a) steam generator steam relief path, (b) feed breakers from RSS transformers to 4 kV buses F and D, (c) AFW system VB-1-1 Small fires in Vertical Board 1-1, around subsection 3BH, Yes No fail CCW return headers A and B TCVs VB-1-2 Small fires in Vertical Board 1-1 or 1-2, around Yes Yes subsections 3BD and 3BC, fail suction and discharge isolation valves of HHSI pumps A, B, and C VB-1-3 Small fires in Vertical Board 1-2, around subsection 3AQ, Yes Yes fail CCW pumps lA and 1B VB-1-4 Medium fires in Vertical Board 1-1, covering subsections Yes Yes 3BH, 3BD, and 3BC, fail (a) CCW return headers A and B TCVs and *(b) suction and discharge isolation valves of HHSI pumps A, B and C VB-1-5 Medium fires in Vertical Board 1-2, covering subsections No Yes 3AQ, 3BC, and 3BD, fail (a) suction and discharge isolation valves of HHSI pumps A, B, and C and (b) CCW pumps lA and 1B 9-29 NUREG/CR-6144
Control Room Fire Analysis Table 9.3 Scenario Description and Preliminary Screening Results (continued) Scenario Scenario Impact Retained for Hot Short Designator Frequency Required? Assessment? VB-1-6 Large fires in Vertical Board 1-1 or 1-2, covering No Yes subsections 3AQ, 3BC, 3BD, and 3BH,, fail (a) suction and discharge isolation valves of HHSI pumps A, B, and C (b) CCW pumps lA and lB, and (c) CCW return headers A and B TCVs VB-6-DGl Fires in Vertical Board 1-6 fail diesel generator 1 and Yes Yes normal power supply to 4 kV bus lH VB-7-DG3 Fires in Vertical Board 1-7 fail diesel generator 3 and No No normal supply to 4 kV bus 1J VBC-1-11-1 Fires in 120 V AC vital bus 1-11 cabinet fail the cabinet Yes No Note: The terms "small", "medium", and "large" fires in the scenario impact description are used only for distinguishing purposes and are not intended to quantatively define the fire magnitudes.
NUREG/CR-6144 9-30

Control Room Fire Analysis Table 9.4 Summary of Control Room Walkdown Observations and Frequency Apportionment for Benchboards 1-1 and 1-2, and Vertical Boards 1-1 and 1-2 Location Panel Subsection1 Weight Normalized Fire Frequency2 Ratio Weight Ratio (/yr)
Benchboard 1-1 3V, 3W and 3X 3.0 0.07 2.98E-04 Benchboard 1-1 3Y, 3Z and 3AA 3.0 0.07 2.98E-04 Benchboard 1-1 3AB and 3AC 3.0 0.07 2.98E-04 Benchboard 1-1 3AD 1.5 0.03 1.49E-04 Benchboard 1-1 3AE 1.0 0.02 1.0lE-04 Benchboard 1-2 3A and 3B 1.5 0.03 1.49E-04 Benchboard 1-2 3C, 3D and 3E 1.5 0.03 1.49E-04 Benchboard 1-2 3F, 3G and 3H 1.8 0.04 1.80E-04 Benchboard 1-2 3J and 3K 1.5 0.03 1.49E-04 Vertical Board 1-1 3BC through 3BF 6.0 0.14 5.96E-04 Vertical Board 1-1 3BG, 3BH and 3BJ 6.0 0.14 5.96E-04 Vertical Board 1-2 3AQ, 3AR and 3AS 7.2 0.16 7.14E-04 Vertical Board 1-2 3AT, 3AU and 3AV 7.2 0.16 7.14E-04 Sum 44.2 1.00 4.39E-03 Note: 1 Fire frequency for individual subsection within each row is assumed to be linearly proportioned 2 Total fire frequency for Benchboards 1-1 and 1-2, and Vertical Boards 1-1 and 1-2 is 7.96E-4 + 3.58E-3 = 4.38E-3/yr. (See Table 9.9.)
9-31 NUREG/CR-6144
Table 9.S Empirical Data Base Specialization Event Plant Event Date Location Operation Mode Initiated General Description Comment No. Unit Component 188 Unknown 7/4nB Control Power Operation Diode Fire involved in the RPS located in the Not included. Building control building. Operator noticed smoke Component failure only. coming from the RPSCIP panel. Component not in MCR. 225 Three Mile Island 2 7/1'1/79 Control Room Cold Shutdown Resistor A small fire occurred in the radiation Included in the Surry monitoring readout panel for the Allll. Bldg. MCR fire data base. waste gas system monitor. The cause was suspected to be an overheated resistor. 263 Unknown 11/21/80 Control Construction While putting up new wall outside the Not included. Building control room, sparks from metal plate Construction event welding ignited Dashing and nailer board. occurred outside control room. 271 Watts Bar 1 3n/81 Control Construction Computer Fire in a digital computer system in Not included. Unit never Building communications room or the facilities reached critical. control building. Component not in MCR. 318 Salem l 11/9/82 Control Room Power Operation Relay Operator observed a fire in fire detection Included in the Surry instrumentation panel. The fire was resulted MCR fire data base. from the failure of the panel alarm buzzer relay. 323 McGuire 2 2/19/83 Control Hot Shutdown Fan Household fan caught on fire. Approx. 2 ft Not included. Fan is not Building cable tray damaged. used in Surry MCR. 331 Surry l 9/9/83 Control Power Operation ~otor With unit l at power, while attempting to Not included. Building (100%) energize *A" control room air conditioning Component not in MCR. chiller, the local control panel caught fire. Event occurred in control Service water from a blown condenser zinc building outside control plug shorted the motor starter contact at room. thq local chiller control panel. Power Operation
' Relay i
Co,1 failure led to burning of relay due to Not included. 368 Three Mile Island 1 9/27/82 Control ' Building either end-of-life or undefined mechanical Component failure only.
' ' : faijure.
Table 9.5 Empirical Data Base Specialization (continued)
Event Plant Event Date Location Operation Mode Initiated General Description Comment No. Unit Component 372 Hatch 1 3/30/83 Control Room Power Operation Relay Smoke observed from a scram discharge Included in Surry MCR volume high level RPS relay in the control fire data base. room. The fire was extinguished by portable extinguisher. 374 Three Mile Island 1 4/'J/83 Control Power Operation Relay Coil failure led to burning of relay due to Not included. Building either end-of-life or undefined mechanical Component failure only. failure. 37.S Three Mile Island 1 4/19/83 Control Relay Coil failure led to burning of relay due to Not included. Building either end-of-life or undefined mechanical Component failure only.
.. failure .
379 Hatch I 8/'}/83 Control Room Power Operation Relay CR personnel noted a "foul smell" in the Nol induded. Fire CR. A relay was found to have overheated precursor only. and created the offgasing. 380 Three Mile Island I 8/27/83 Control Cold Shutdown Relay Coil failure led to burning of relay due to Not included. Building either end-of-life or undefined mechanical Component failure only. failure. 382 San Onofre 3 10/S/83 Control Room lnslrumentalion Operators noted a "smoky odor" being Not induded. Fire emitted from the Qualified Safety Precursor only. Parameter Display System Train B panel. Problem was traced to a failed circuit card. No open naming was reported and apparently no fire alarm was initiated . 391 Crystal River 3 10/26/8.S Control Room Power Operation Bus : While operating al 95% power, an Not induded. Fire undervol1age condition on the radiation precursor. Event monitoring panel caused overheating of occurred outside control relays and the introduction of smoke into room during at-power the control room. No open name was condition. reported and the relay smoking was a result rather than the cause of the adverse occurrences.
Table 9.5 Empirical Data Base Specialization (continued) Event Plant Event Date I.Aicatlon Opemtion Mode Initiated General Description Comment No. Unit Component 400 McGuire 1 1n./81 Control Room Hot Shutdown Breaker While performing control rod drop timing Not included. Switchgear test, personnel detected smoke in the area fire. of the reactor trip switchgear. 410 Calvert Cliffs 2 3/1/89 Control Room Power Operation Instrumentation While attempting to trip a solenoid Included in Surry MCR operated amiliary feedwater pump fire data base. trip/throllle valve during maintenance, the switch was found to be on fire. The fire was elllinguished using a portable Halon elllinguisher. 412 Waterford 3 7/14/85 Control Room Power Operation Resistor While at 100% power, an overheated Included in Surry MCR resistor caused a trip. The overheating fire data base. circuit board cause a fire in the control panel and was quickly elllinguished. 416 Hatch 1 3/12/83 Control Room Power Operation Relay Personnel smelled smoke around a panel in Included in Surry MCR the control room as a result of a burning fire data base. low reactor water level RPS relay. 419 Three Mile Island 9n/8s Control Hot Shutdown Instrumentation A fire resulted from transfer of power Not included. Nol a Building supplies without clamping and control room fire. Event synchronization. The fire was small and not compatible with focalized to several CRD cabinets in the Surry MCR relay room. configuration. 431 Wolf Creek 10/14/87 Control Refsecueling A maintenance worker was electrocuted No included. Event not Building Outage while cleaning potential transformer compatible to Surry cabinets. A part of the worker's clothing MCR. caught fire. 40 Pilgrim 9/13/83 Control Room Power Operation Relay Failure of a recirculation pump low level Not included. Fire water trip relay resulted in smoke emitting precursor, component from the burned coil coating. failure.
Table 9.5 Empirical Data Base Specialization (continued)
Event Plant Event Date Location Operation Mode Initiated General Description Comment No. Unit Component 442 Shoreham 4f7/8S Control Room Shutdown Fan Failure or cooling for a circuit inverter Included in Surry MCR causing the inverter cabinet to overheat. lire data base. Unit This resulted in the overheating of a achieved criticality and resistor and a capacitor in the invertor produced power, but circuitry. closed down before it could begin commercial operation. 446 Susquehanna 2 7/16/86 Control Room Power Operation Instrumentation Control room personnel observed smoke Not included. Fire coming from a "C" Transversing Incore precursor only. Probe machine.
Control Room Fire Analysis Table 9.6 Summary of Control Room Age Assessment
Site-Plant Number of Date Achieved Reference Control Control Room Criticality Date1 Room Age Arkansas Nuclear One Unit 1 1 8t6n4 12/31/89 15.41 Arkansas Nuclear One Unit 2 shared 1215n8 12/31/89 11.08 Beaver Valley 1 1 5110n6 12/31/89 13.65 Beaver Valley 2 1 8/1/87 12/31/89 2.42 Big Rock Point 1 1 9/27/62 12/31/89 27.28 Braidwood 1 1 5/15/87 12/31/89 2.63 Braidwood 2 shared 3/15/88 12/31/89 1.8 Brown's Ferry 1 1 8t11n3 3/9/85 11.57 Brown's Ferry 2 shared 7/20(74 3/9/85 10.64 Brown's Ferry 3 1 8/8(76 3/9/85 8.59 Brunswick 1 1 10/8(76 12/31/89 13.24 Brunswick 2 shared 3/20n5 12/31/89 14.79 Byron 1 1 2/2/85 12/31/89 4.91 Byron 2 shared 1/1/87 12/31/89 3.0 Callaway 1 1 10/2/84 12/31/89 5.25 Calvert Cliffs 1 1 10nn4 12/31/89 15.24 Calvert Cliffs 2 shared 11/30/76 12/31/89 13.09 Catawba 1 1 1/8/85 12/31/89 4.98 Catawba2 shared 5/15/86 12/31/89 3.63 Clinton 1 1 2/15/87 12/31/89 2.88 Cook 1 1 1/18(75 12/31/89 14.96 Cook2 1 3tl0n8 12/31/89 11.82 Cooper Station 1 2121n4 12/31/89 15.87 Crystal River 3 1 1114n1 12/31/89 12.97 Davis Besse 1 1 8/12(77 12/31/89 12.39 NUREG/CR-6144 9-36
Control Room Fire Analysis Table 9.6 Summary of Control Room Age Assessment Site-Plant Number of Date Achieved Reference Control Control Room Criticality Date 1 Room Age Diablo Canyon 1 1 4/29/84 12/31/89 5.68 Diablo Canyon 2 shared 8/15/85 12/31/89 4.38 Dresden 1 1 10/15/59 10/31/78 19.06 Dresden 2 1 1nno 12/31/89 19.99 Dresden 3 shared 1131n1 12/31/89 18.93 Duane Arnold 1 3/23n4 12/31/89 15.79 Farley 1 1 8J9n1 12/31/89 12.4 Farley 2 shared 5/8/81 12/31/89 8.65 Fermi2 1 6/15/85 12/31/89 4.55 itzpatrick 1 11/17/74 12/31/89 15.13 ort Calhoun 1 1 8J6n3 12/31/89 16.41 Ft. St. Vrain 1 1131n4 8/15/89 15.55 Ginna 1 1 11/8/69 12/31/89 20.16 Grand Gulf 1 1 8/18/82 12/31/89 7.38 Haddam Neck 1 7/24/67 12/31/89 22.45 Harris 1 1/15/87 12/31/89 2.96 Hatch 1 1 w12n4 12/31/89 15.31 Hatch 2 shared 114n8 12/31/89 11.5 Hope Creek 1 1 6/16/86 12/31/89 3.55 Humboldt Bay 1 2/16/63 112n6 13.38 Indian Point 1 1 4J6n6 12/31/89 13.75 Indian Point 2 1 8/2/62 10/31/74 12.25 Indian Point 3 1 5122n3 12/31/89 16.62 Kewaunee 1 3nn4 12/31/89 15.83 a Crosse 1 7/11/67 4/15/87 19.78 9-37 NUREG/CR-6144
Control Room Fire Analysis Table 9.6 Summary of Control Room Age Assessment (continued) Site-Plant Number of Date Achieved Reference Control Control Room Criticality Date 1 Room Age La Salle 1 1 6/21/82 12/31/89 7.53 La Salle 2 shared 3/10/84 12/31/89 5.81 Limerick 1 1 12/22/84 12/31/89 5.03 Limerick2 shared 8/15/89 12/31/89 0.38 Maine Yankee 1 10/23/72 12/31/89 17.2 McGuire 1 1 8/8/81 12/31/89 8.4 McGuire2 shared 5/8/83 12/31/89 6.65 Millstone 1 1 10/26/70 12/31/89 19.19 Millstone 2 1 10/17/75 12/31/89 14.22 Millstone 3 1 1/15/86 12/31/89 3.96 Monticello 1 12/10/70 12/31/89 19.07 Nine Mile Point 1 1 9/5/69 12/31/89 20.33 Nine Mile Point 2 1 5/15/87 12/31/89 2.63 North Anna 1 1 4/5/78 12/31/89 11.75 North Anna 2 1 6/12/80 12/31/89 9.56 Oconee 1 1 4/19/73 12/31/89 16.71 Oconee 2 shared 11/11/73 12/31/89 16.15 Oconee3 1 9/5/74 12/31/89 15.33 Oyster Creek 1 1 5/3/69 12/31/89 20.68 Palisades 1 5/24/71 12/31/89 18.62 Palo Verde 1 1 5/25/85 12/31/89 4.61 Palo Verde 2 1 4/15/86 12/31/89 3.72 Palo Verde 3 1 10/15/87 12/31/89 2.21 Peach Bottom 1 1 6/15/67 11/1/74 7.39 Peach Bottom 2 1 9/16/73 12/31/89 16.3 NUREG/CR-6144 9-38
Control Room Fire Analysis Table 9.6 Summary of Control Room Age Assessment Site-Plant Number of Date Achieved Reference Control Control Room Criticality Date 1 Room Age Peach Bottom 3 shared 8nn4 11nn4 0.24 Perry 1 6/15/86 12/31/89 3.55 Pilgrim 1 1 6/16n2 12/31/89 17.55 Point Beach 1 1 1112no 12/31/89 19.18 Point Beach 2 shared 513on2 12/31/89 17.6 Prairie Island 1 1 1211n3 12/31/89 16.09 Prairie Island 2 shared 12/17/74 12/31/89 15.05 Quad Cities 1 1 10/18/71 12/31/89 18.22 Quad Cities 2 shared 4!26n2 12/31/89 17.69 ancho Seco 1 9;16n4 6/15/89 14.76 iver Bend 1 1 10/15/85 12/31/89 4.21 Robinson 2 1 912ono 12/31/89 19.29 Salem 1 1 12/11/76 12/31/89 13.06 Salem 2 shared 8/8/80 12/31/89 9.4 San Onofre 1 1 6/14/67 12/31/89 22.56 San Onofre 2 1 7/26/82 12/31/89 7.44 San Onofre 3 shared 8/29/83 12/31/89 6.35 Seabrook 1 6/15/89 12/31/89 0.55 Sequoyah 1 1 7/5/80 3/9/85 4.68 Sequoyah 2 shared 11/5/81 3/9/85 3.34 Shippingport 1 12/23/57 10/1/82 24.79 Shoreham 1 2/15/85 5/15/89 4.25 South Texas Project 1 1 3/15/88 12/31/89 1.8 South Texas Project 2 shared 3/15/89 12/31/89 0.8
t. Lucie 1 1 4122n6 12/31/89 13.7 9-39 NUREG/CR-6144
Control Room Fire Analysis Table 9.6 Summary of Control Room Age Assessment (continued) Site-Plant Number of Date Achieved Reference Control Control Room Criticality Date 1 Room Age St. Lucie 2 1 6/2/83 12/31/89 6.59 Summer 1 1 10/22/82 12/31/89 7.2 Surry 1 1 711n2 12/31/89 17.51 Surry 2 shared 3nn3 12/31/89 16.83 Susquehanna 1 1 9/10/82 12/31/89 7.31 Susquehanna 2 shared 5/8/84 12/31/89 5.65 Three Mile Island 1 1 6!5n4 12/31/89 15.58 Three Mile Island 2 1 3!27n8 3!28n9 1.0 Trojan 1 12/15/75 12/31/89 14.05 Turkey Point 3 1 10/20/72 12/31/89 17.21 Turkey Point 4 Vermont Yankee 1 shared 1 6111n3 3!24n2 12/31/89 12/31/89 16.57 17.78 1* Vogtle 1 1 3/15/87 12/31/89 2.8 Vogtle 2 shared 3/15/89 12/31/89 0.8 Waterford 3 1 3/4/85 12/31/89 4.83 WNP2 1 1/19/84 12/31/89 5.95 Wolf Creek 1 1 5/22/85 12/31/89 4.61 Yankee Rowe 1 8/19/60 12/31/89 29.39 Zion 1 1 6119n3 12/31/89 16.55 Zion 2 shared 12/24/73 12/31/89 16.03 Total 1,359.35 Industry generic control room experience excluding Surry-1 is 1341.84 year Note: 1. Reference date is either 12/31/89 or date of decommissioning. NUREG/CR-6144 9-40
Control Room Fire Analysis Table 9.7 Main Characteristics of the Noninformative Prior D A T A M A N 092492 GENERIC DISTRIBUTION UPDATE REPORT MCRPRI MCR GENERIC NONINFORMATIVE PRIOR FIRE FREQUENCY PRIOR - Median S.2200E-03 Range Factor 7.0000E+OO MAIN CHARACTERISTICS OF THE DISTRIBUTION MEAN 1.0509E-02 VARIANCE 2.9012E-04 5 TH PERCENTILE 7.0270E-04 50TH PERCENTILE S.0853E-03 95TH PERCENTILE 3.5460E-02 DISCRETE PROBABILITY DISTRIBUTION VALUE PROBABILITY CUMULATIVE 1.7943E-04 S.OOOOE-03 5.0000E-03 2.9267E-04 5.0000E-03 1.0000E-02 5.0SOOE-04 3.0000E-02 4.0000E-02 6.8010E-04 5.0000E-03 4.SOOOE-02 D A T A M A N 092492 GENERIC DISTRIBUTION UPDATE REPORT 9.2873E-04 5.SOOOE-02 1.0000E-01 1.5336E-03 1.0000E-01 2.0000E-01 2.3562E-03 1.0000E-01 3.0000E-01 3.3186E-03 1.0000E-01 4.0000E-01 4.3754E-03 8.0000E-02 4.SOOOE-01 4.9931E-03 1.0000E-02 4.9000E-01 S.5468E-03 6.0000E-02 S.SOOOE-01 6.5361E-03 S.OOOOE-02 6.0000E-01 8.2815E-03 1.0000E-01 7.0000E-01 1.1700E-02 1.0000E-01 8.0000E-01 1.8165E-02 1.0000E-01 9.0000E-01
2. 7776E-02 4.0000E-02 9.4000E-01 3.3688E-02 5.0000E-03 9.4SOOE-01 4.9640E-02 4.SOOOE-02 9.9000E-01 9.3766E-02 S.OOOOE-03 9.9SOOE-01

1. 7200E-01 S.OOOOE-03 1.0000E+OO
9-41 NUREG/CR-6144
Control Room Fire Analysis Table 9.8 Main Characteristics of the Posterior D A T A M A N 092492 GENERIC DISTRIBUTION UPDATE REPORT MCRPOS MCR POSTERIOR FIRE FREQUENCY POSTERIOR - Number of Events 0 Number of Time Units 1.7510E+Ol MAIN CHARACTERISTICS OF THE DISTRIBUTION MEAN 7.6561E-03 VARIANCE 9.6108E-05 5 TH PERCENTILE 6.1778E-04 50TH PERCENTILE 3.9181E-03 95TH PERCENTILE 1.9964E-02 DISCRETE PROBABILITY DISTRIBUTION VALUE PROBABILITY CUMULATIVE l.7943E-04 5.8167E-03 5.8167E-03 2.9267E-04 S.8052E-03 l.1622E-02 S.OSOOE-04 3.4700E-02 4.6322E-02 6.SOlOE-04 S.7660E-03 S.2088E-02 9.2873E-04 6.3150E-02
1.1524E-01 l.5336E-03 l.1361E-01 2.2885E-01 2.3562E-03 1.1198E-01 3.4083E-01 3.3186E-03 l.lOllE-01 4.5094E-01 4.3754E-03 8.6475E-02 5.3742E-01 4.9931E-03 l.0693E-02 5.4811E-01 5.5468E-03 6.3540E-02 6.1165E-01 6.5361E-03 S.2040E-02 6.6369E-01 8.2815E-03 1. 0095E-01 7.6464E-Ol l.1700E-02 9.5082E-02 8.5972E-01 l.8165E-02 8.4906E-02 9.4463E-01
2. 7776E-02 2.8702E-02 9.7333E-01 3.3688E-02 3.2349E-03 9.7656E-01 4.9640E-02 2.2019E-02 9.9858E-01 9.3766E-02 l.1298E-03 9.9971E-01 l.7200E-01 2.8711E-04 1.0000E+OO NUREG/CR-6144 9-42
Control Room Fire Analysis Table 9.9 Control Room Cabinet Fire Frequency Apportionment Control Cabinet Approx Panel Normalized Cabinet Fire Area (ft2) Ratio Frequency (/yr)4 Benchboard 1-1 and 1-2 (2x12'x4') 96 0.1 7.96E-04 Vertical Board 1-1 and 1-2 (2x2x13.5'x8') 1 432 0.47 3.58E-03 Vertical Board 1-3 (6.4'x8') 51.2 0.06 4.21E-04 Vertical Board 1-4 (6'x5') 30 0.03 2.53E-04 Vertical Board 1-5 (ll.3'x4.5') 50.85 0.06 4.21E-04 Vertical Board 1-6 (11.3'x5.3') 59.89 0.07 4.98E-04 Vertical Board 1-7 (6.7'x5.3') 35.51 0.04 2.91E-04 Semi-Vital bus (2.5'x5') 2 12.5 0.01 l.07E-04 Vital Bus cabinet 1-II (2.5':x5')2 12.5 0.01 1.07E-04 Vital Bus cabinet 1-IV (2.5':x5') 2 12.5 0.01 l.07E-04 125V DC cabinet 1-1 (2.5'x5') 2 12.5 0.01 1.07E-04 125V DC cabinet 1-2 (2.5'x5') 2 12.5 0.01 l.07E-04 120V AC cabinet (2.5'x5') 2 12.5 0.01 1.07E-04 Other miscellaneous panels and junction 92.27 0.10 7.66E-04 boxes (10% of the total panel area of all MCR cabinets) 3 Sum 922.72 1.00 7.67E-03 Note:
1. Both of the front and back panels of the Vertical boards were observed to have instrumentations and related cables.

2. The height of these cabinets is conservatively assumed to be 5 ft.

3. 10% of the total fire frequency is conservatively apportioned to the miscellaneous areas, total surface areas of all control panels is 922.72 sq ft.

4. Individual panel fire frequency is equal to the product of the normalized panel surface area ratio and the total MCR fire frequency (7.656E-3 per year) .
9-43 NUREG/CR-6144
Control Room Fire Analysis Table 9.10 Fire Frequency Apportionment of Panel Subsections in Benchboard Boards 1-1 and 1-2 Benchboard Approx. Panel Normalized Fire Frequency Subsection Length (ft) 1 Ratio2 (/yr)3 3W/3V 2.35 0.606 1.80E-04 3X 1.53 0.394 l.18E-04 3Y 1.27 0.407 l.21E-04 3Z 0.93 0.298 8.88E-05 3AA 0.92 0.295 8.79E-05 3AB 1.19 0.613 l.83E-04 3AC 0.75 0387 l.15E-04 3AD 0.92 1.000 l.49E-04 3AE 1.57 1.000 l.OlE-04 3A/3B 2.23 1.000 l.49E-04 3C 1.00 0.398 5.94E-05 3D 0.88 0.351 5.22E-05 3E 0.63 0.251 3.74E-OS 3F/3G/3H 2.53 1.000 l.80E-04 3J 1.47 0.531 7.91E-OS 3K 1.30 0.469 6.99E-OS Sum 21.47 l.77E-03 Note:
1. Approximate panel subsection length is measured from Drawings 11448-RE-25A through J.

2. The normalized ratio, in this case, is the geometry factor for the panel subsection.

3. Panel subsection fire frequency is the product of the geometry factor and the grouped panel section fire frequency in each row of Table 9-9.
NUREG/CR-6144 9-44
Control Room Fire Analysis Table 9.11 Fire Frequency Apportionment of Panel Subsections in Vertical Boards 1-1 and 1-2 Vertical Board Approx. Panel Normalized Fire Frequency Subsection Length (ft)1 Ratio2 (/yr)3 3BC/3BD 1.80 0.323 1.93£-04 3BE 1.70 0.305 1.82£-04 3BF 2.07 0.372 2.21£-04 3BG 2.67 0.444 2.65£..:04 3BH 1.87 0.311 1.85£-04 3BJ 1.47 0.245 1.46£-04 3AS 1.57 0.261 1.87£-04 3AR 2.37 0.394 2.82£-04 3AQ 2.07 0.344 2.46£-04 3AT 3.58 0.523 3.73£-04 3AU 1.27 0.185 1.32£-04 3AV 2.00 0.292 2.08£-04 Sum 2.62£-03 Note:
1. Approximate panel subsection length is measured from Drawings 11448-RE-25A through J.

2. The normalized ratio, in this case, is the geometry factor for the panel subsection.

3. Panel subsection fire frequency is the product of the geometry factor and the grouped panel section fire frequency in each row of Table 9-9.
9-45 NUREG/CR-6144
Table 9.12 Summary of Scenario Quantification Scenario Target Relevant Geometry Factor Required Fire Severity Combined Severity and Failure Scenario Subsection Fire and Fire Frequency Radius (ft) Factor Geometry Factor/Fire Mode Freq~ency Origin (/yr) Frequency (/yr) Factor (/yr) BB-1-1 3Z 3Z 8.88E-05 0.0 < R < 4.4 0.962 8.54E-05 3AB l.83E-04 2.1 < R < 3.0 0.153 2.SOE-05 3AA 8.79E-05 0.9 < R < 4.1 0.517 4.54E-05 3Y 1.21E-04 1.2 < R < 3.3 0.377 4.56E-05 Total: 2.04E-04 0.10 2.04E-05 BB-1-2 3W/3V 3W/3V 1.SOE-04 0.0 < R < 4.4 0.962 1.73E-04 3X 1.18E-04 2.0 < R < 2.5 0.130 1.53E-05 3K 6.99E-05 1.8 < R < 3.3 0.217 1.52E-05 Total: 2.04E-04 0.10 2.04E-05 BB-1-3 3A/3B 3A/3B 1.49E-04 0.0 < R < 5.6 0.972 1.45E-04 3C 5.94E-05 2.3 < R < 3.3 0.075 4.46E-06 Total: 1.49E-04 0.10 1.49E-05 BB-1-4 3AE, 3Z 3AE l.OlE-04 5.0 < R < 8.0 0.009 9.09E-07 3AD 1.49E-04 3.8 < R < 8.0 0.035 5.22E-06 3AC 1.15E-04 3.0 < R < 7.5 0.058 6.67E-06 3AB l.83E-04 3.0 < R < 6.5 0.054 9.96E-06 3AA 8.79E-05 4.1 < R < 5.4 0.021 1.86E-06 Total: 2.46E-05 0.10 2.46E-06
Table 9.12 Summary of Scenario Quantification (continued)
Scenario Target Relevant Geometry Factor Required Fire Severity Combined Severity and Failure Scenario Subsection Fire and Fire Frequency Radius (ft) Factor Geometry Factor/Fire Mode Freq~ency Origin (/yr) Frequency (/yr) Factor (/yr) BB-1-5 3AE, 3Z, 3AC 1.15E-04 7.5 < R < 8.0 0.001 1.15E-07 , 3W/3V 3AB 1.83E-04 6.5 < R < 8.0 0.005 8.42E-07 3AA 8.79E-05 5.4 < R < 8.0 0.008 6.86E-07 3Z 8.88E-05 5.0 < R < 8.0 0.009 7.99E-07 3Y 1.21E-04 6.1 < R < 8.0 0.006 6.79E-07 Total: 3.12E-06 0.10 3.12E-07 BB-1-8 3Z, 3W/3V 3Z 8.88E-05 4.4 < R < 5.0 0.020 1.78E-06 3Y 1.21E-04 3.3 < R < 6.1 0.043 5.25E-06 3X 1.18E-04 2.5 < R < 6.1 0.073 8.66E-06 3W/3V 1.80E-04 4.4 < R < 5.2 0.009 1.56E-06 Total: 1.73E-05 0.10 l.73E-06 BB-1-9 3Z, 3W/3V, 3W/3V l.BOE-04 5.2 < R < 8.0 0.008 1.49E-06 3H/3G/3F 3K 6.99E-05 6.2 < R < 8.0 0.005 3.77E-07 Total: 1.87E-06 0.10 1.87E-07 BB-1-11 3W/3V, 3K 6.99E-05 3.8 < R < 6.2 0.029 2.00E-06 3H/3G/3F 3J 7.91E-05 3.2 < R < 7.5 0.052 4.07E-06 3H/3G/3F 1.80E-04 5.2 < R < 5.6 0.001 1.98E-07 Total:' 6.27E-06 0.10 6.27E-07
Table 9.12 Summary or Scenario Quantification (continued) Scenario Target Relevant Geometry Factor Required Fire Severity Combined Severity and Failure Scenario Subsection Fire and Fire Frequency Radius (ft) Factor Geometry Factor/Fire Mode Freq~ency Origin (/yr) Frequency (/yr) Factor (/yr) 3A/38, 3E 3.74E-05 6.8<R<8.0 0.004 1.35E-07 3H/3G/3F 3D 5.22E-05 7.8 < R < 8.0 0.001 3.13E-08 Total: l.46E-06 0.10 1.46E-07 88-1-13 3H/3G/3F, 3E 3.74E-05 4.1 < R < 6.8 0.023 8.75E-07 3A/38 3D 5.22E-05 33 < R < 7.8 0.050 2.63E-06 3C 5.94E-05 3.3 < R < 8.0 0.049 2.91E-06 3A/3B 1.49E-04 0.0 < R < 8.0 0.007 1.06E-06 Total: 7.47E-06 0.10 7.47E-07 V8-l-l 38H 3BG 2.65E-04 2.3 < R < 6.6 0.137 3.62E-05 3BH 1.85E-04 0.0 < R < 8.0 0.979 1.81E-04 3BJ l.46E-04 1.7 < R < 8.0 0.301 4.39E-05 Total: 2.69E-04 1.00 2.69E-04 V8-l-2 3BC/3BD 3BF 2.21E-04 4.2 < R < 4.6 0.011 2.43E-06 3BE l.82E-04 2.3 < R < 6.5 0.120 2.17E-05 3BC/BD 1.93E-04 0.0 < R < 8.0 0.979 l.89E-04 3AT 3.73E-04 3.3 < R < 8.0 0.049 1.83E-05 3AU 1.32E-04 5.7 < R < 7.6 0.006 7.39E-07 Total: 2.32E-04 0.10 2.32E-05 e
Table 9.12 Summary or Scenario Quantification (continued) Scenario Target Relevant Geometry Factor Required Fire Severity Combined Severity and Failure Scenario Subsection Fire and Fire Frequency Radius (ft) Factor Geometry Factor/Fire Mode Freql\ency Origin (/yr) Frequency (/yr) Factor (/yr) 3AR 2.82E-04 2.2 < R < 8.0 0.150 4.23E-05 3AS 1.87E-04 4.2 < R < 8.0 0.024 4.49E-06 3AV 2.08E-04 5.9 < R < 7.4 0.004 9.15E-07 Total: 5.98E-05 0.10 5.98E-06 VB-1-4 3BH, 3BG 2.65E-04 6.6 < R < 8.0 0.004 1.14E-06 3BC/3BD 3BF 2.21E-04 4.6 < R < 8.0 0.011 2.43E-06 3BE 1.82E-04 6.5 < R < 8.0 0.005 8.19E-07 Total: 4.39E-06 0.10 4.39E-07 VB-6-DGl DG1 RMS-100 8.07E-05 R > 4.4 0.038 3.07E-06 IS 5.38E-05 R > 2.9 0.084 4.52E-06 GTG 6.41E-05 R > 1.5 0.358 2.29E-05 DGl 7.48E-05 R > 0.0 1.000 7.48E-05 sws 7.48E-05 R > 1.7 0.322 2.41E-05 LFHT 7.48E-05 R > 3.3 0.070 5.24E-06 VENT 7.48E-05 R > 5.0 0.030 2.25E-06 Total: 1.37E-04 0.10 l.37E-05 VBC-1-11-1 VBC-1-11 VBC-1-11 1.07E-04 R > 0.0 1.000 1.07E-04 1.00 1.07E-04
10. MODIFICATION OF THE INTERNAL PRA MODEL The internal PRA IRRAS model1 was used as a starting point in the analysis of the impact of failures caused by fire events.
The model was then modified by adding fire-induced failures to the existing component failures in the fault trees. For instance, for an LPI pump, fire-caused failures due to damage to the control cables or damage to the power cables were added to the failure due to internal causes wherever this event appeared in the fault trees. For a running pump, or a valve, failure modes due to spurious operation were also added. Such failures are caused by fire-induced hot shorts. There were no modifications of the event trees. The event trees used are shown in Appendix I, and the modified fault trees are shown in Appendix J. 10.1
References:
1. T-L. Chu, Z. Musicki, P. Kohut, et al., "Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry Unit-1: Analysis of Core Damage Frequency from Internal Events During Mid-Loop Operations", NUREG/CR-6144, Volume 2, June 1994.
10-1 NUREG/CR-6144
11. QUANTIFICATION 11.1 Summary of Procedure For each fire scenario in Section 5.5, the appropriate initiating event was identified, and the corresponding event tree sequences were used in quantification (e.g., loss of 4 kV bus, station blackout, etc. sequences). The initiating event frequency was set equal to the scenario frequency from Table 5.4 or, for items from Table 9.12 multiplied by the mid-loop fraction (0.066), the probability of the particular POS and the particular window.
Appropriate fire-induced failures were flagged (according to descriptions in Section 5.5), and the scenario- dependent, high-level human error probabilities were set to their appropriate values (see Section 11.2 and Appendix G). The core damage sequences were quantified with a cutset cutoff of 1.E-10. The resulting values were modified for successful systems, where appropriate (because the cutset evaluation algorithm assumes a probability of 1 for success of such systems). Finally, recovery actions were applied to qualifying cutsets greater than 1.E-81)'1' to account for operator actions not credited in the model (e.g., opening of the LPI discharge valves 1890C or 1890B, which are normally closed in mid-loop), or to refine conservatisms. The recovery actions considered were opening of valves 1890B or C when failure of forced cooling is indicated in the cutset and opening of Unit 2 charging cross-tie, under the same circumstances. covery of forced cooling is not included if the cutset indicates that the failure of forced cooling is caused by operator
r. No recovery is applied in cutsets showing more than one human error.
Recovery actions rules are shown in Appendix Q. 11.2 Quantification of HEP for High-Level Human Action Events 11.2.1 High Level HEP Quantification The approach used for the evaluation of operator actions in response to the fire initiating events is the same as that used for the internal event analysis1* The dynamic operator actions analyzed in this study are listed in Table G-1 of Appendix G. All of the human actions with unique numerical values are quantified in Table G-2 of Appendix G, following the descriptions of Table G-1. There are other actions defined in Table G-1 that have the same numerical values as those tabulated in Table G-2 and are also used in the event sequence quantification. 11.2.2 Recovery Actions Recovery actions quantifications are shown in Tables G-3 and G-4. In these tables, only a sample of scenarios is given where these recovery actions apply. In the analysis, these recovery actions were applied to all the applicable scenarios. 11.3 Successful System Correction Another type of correction (or recovery action) occurs in certain sequences with success of the F event (forced feed bleed), and failure of the C event (recirculation), see Appendix I. Some such sequences have a high value for the vent ( > 0.1 ), such that a correction is needed (computer code IRRAS used for modeling always quantifies 11-1 NUREG/CR-6144
Quantification successful systems as having a probability of 1). The complement value (1-F) of the fault tree quantification of F was used to multiply the affected sequence cutsets to arrive at the correct sequence quantification. Likewise, uncertainty analysis of the affected F trees was used to arrive at an uncertainty histogram for this recovery factor. 11.4 Uncertainty Analysis An in-house VAX resident computer code was used for generation of uncertainty distribution for total CDF. Fire scenarios are usually composed of several components ( e.g., there could be a cable fire and a large and a small transient fire, or a bus panel fire may affect more than one component of the RHR system); each component will have several (on the order of 10) multiplicative terms, most of which will have an uncertainty distribution. This results in complicated algebraic expressions with many shared elements for evaluation of initiating events (see Appendix K). A separate computer program was created on the VAX to generate histograms of uncertainty distributions for the initiating events for each scenario (each scenario has 10 POS-window combinations). This is accomplished by combining LHS samples of uncertainty distributions of constituent events in the algebraic expressions for the initiating events. The initiating event histograms that resulted from this procedure are shown in Appendix L. The samples of initiating event distnbutions were input into the VAX uncertainty code, along with the cutset Boolean expression and basic event uncertainty data. It should be noted that a uniform distribution was used for such parameters as the large oil fraction (from 0. to 0.6) and damage fractions for various scenario (the upper bound of 1.0 was used). The latter is subject to many unknown uncertainties, starting with input parameters to the COMPBRN code and including (but not limited to) COMPB modeling itself. For most other variables, a lognormal distribution was assumed. For fire frequencies in vari equipment, and from cable and transient fires, the distributions presented in Chapter 4 were used. For the HEPs that appeared in the cutsets (including the recovery actions), a lognormal distribution was assumed with the following error factors: HEPs of < 1.E-2 were given an error factor of 10; between 1.E-2 and 0.1 resulted in an error factor of 5; from 0.1 to 0.3 were given an error factor of 2; HEPs from 0.3 to 0.5 assumed an error factor of 1.5, and above 0.5, an error factor of 1.0 was assumed. For recovery actions resulting from treatment of successful systems in the event trees, the resulting histograms were generated in accordance with the procedure described in Subsection 11.3. NUREG/CR-6144 11-2
Quantification 11.5 References
1. T-L Chu, Z. Musicki, P. Kohut, et al., "Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry Unit-1: Analysis of Core Damage Frequency from Internal Events During Mid-Loop Operations", NUREG/CR-6144, Volume 2, June 1994.
11-3 NUREG/CR-6144
12. RESULTS 12.1 Point Estimate Results The unrecovered total core damage frequency (CDF)at Surry Unit 1 due to fire events is 2.72-5/yr .. The recovered total core damage frequency is 1.68-5/yr. As can be seen, the recovery actions bring about less than a factor of 2 reduction in the total CDF. This is due to the fact that in windows 1 and 2, the recovery actions considered ( e.g,. Unit 2 charging cross-tie) will only extend the time to core damage, as recirculation is needed within the 24- hour mission time frame. Damage done by fire is considered irrecoverable in this time frame. Windows 1 and 2 are significant, on the other hand, because the decay heat level is the highest, and the success criteria are the most stringent.
Table 12.1 shows the scenario, window, and POS-dependent core damage frequency from sequences exceeding 1.0E-7/yr after recovery. Sequences were quantified down to the 1.E-10/yr level, but only sequences greater than 1.E-8 were considered for recovery. (These added up to 96% of the total CDF for the unrecovered sequences and 88% of the total CDF for the recovered sequences). This table shows the interplay between decay heat, success criteria, and unavailabilities of equipment due to shutdown activities and those caused by a fire scenario. In Table 12.1, the first column denotes the combination of fire scenario, window, and POS of the significant sequences. The POS is the last two or three characters, denoted by D6 (drained maintenance), R6 (first refueling mid-loop) or RlO (second refueling mid-loop). The window designation is in the middle of the string, denoted by WN, where N refers to the window number (1 through 4). Finally, the string is started by the. scenario designator. A detailed description of scenarios appears in Chapters 5 and 9. r the purposes of this chapter, the E designators are the scenarios in the emergency switchgear room, CVl is the cable vault and tunnel scenario, CT are containment scenarios, N are the normal switchgear room scenarios and, M are the main control room scenarios. MN, where N is a number, refers to the benchboard scenario ( designated BB N in Chapter 9) of the same number. MA, where A is an alphabet character, refers to the vertical board scenario (designated VB-X-M in Chapter 9). For the scenarios of interest (others were screened out), MA refers to scenario VB-1-1, MC refers to VB-1-3, MG refers to VB-6-DGl, and MH is scenario VBC-1-11-1. The second column is the sequence number in the appropriate event tree and the last column is the sequence CDF. The 32 sequences in Table 12.1 add up to about 87% of the total CDF ( altogether there were about 600 sequences above the cutoff of 1.E-10 /yr). As can be seen from Table 12.1, POS RlO (second refueling outage) is not represented among the significant sequences, i.e., it is not very risk significant. This is due to the fact that this POS can only occur in Windows 3 and 4, when the decay heat is lower. Also it can be seen that D6 is more heavily represented (21 out of 32 sequences), i.e., it is more risk significant than R6. D6 tends to occur earlier in the outage, i.e., its distnbution is more heavily weighted toward the earlier windows than R6. An option that is available in R6, namely gravity feed, is not available in D6 due to nonremoval of safety valves. Furthermore, in window 1, where D6 occurs much more frequently than R6, another option, reflux cooling is also not available. This is due to the success criteria for this alternative and the fact that one steam generator is assumed to be quickly isolated in this type of outage. Again, Window 1 has the highest decay heat. It can also be seen in Table 12.1 that earlier windows are more heavily represented than later ones. Window 4, which has the lowest decay heat level, is not represented at all. In this window, gravity feed alone can provide sufficient ling to last the 24-hour mission time. Window 3 is represented in those scenarios (E4, ES and CVl) where there eavy damage from the fire, which causes many options to be lost. In the three scenarios cited, all methods of 12-1 NUREG/CR-6144
Results *. forced feed and bleed from Unit 1 are lost, as is Unit 1 AFW and the primary means of relief from the steam generator secondary for reflux cooling (i.e., the steam generator PORVs). Window 2 is more of a contributor than Window 1, because it lasts much longer and a higher proportion of outages fall into this window. As far as the representation of scenarios is concerned, the table shows what can be expected: the scenarios with moderate to heavy damage to the plant, due to multiple fire-induced equipment failures. Most of the important scenarios are in the ESGR (emergency switchgear room), and the cable vault and tunnel, where cables for many important emergency equipment converge. Note that there are no scenarios from the main control room (MCR) or the normal switchgear room (NSGR) in Table 12.1, because the fire scenarios in these areas fail to inflict enough, damage to enough equipment (or the scenario frequency may be low). Some insights about possible plant vulnerabilities may be gleaned from this table. Scenarios E4 and E5 occur in locations where many cables from both the H and the J division (the two emergency divisions at Suny) come together. Similarly, scenario CVl occurs in a tunnel, where many emergency cables from both divisions are located. The H and J sides of the tunnel are about 10 ft apart (wall to wall, only 5 ft between cable trays). A transient or a cable fire will quickly spread to both sides of the tunnel. Another insight comes from consideration of containment scenarios, where it can be seen that the two RHR divisions are not separated. Tables 12.2 through 12.7 show a more detailed picture of the results after recovery actions. Table 12.2 presents total CDP for each scenario; Table 12.3 shows the same information, but broken down into the three POSs; and T a b l e . has the CDF for Window 1, when the decay heat level is the highest. Tables 12.5-12.7 show the conditional damage probability (i.e., given a fire scenario), arranged in the same format as the previous three tables. These tables bear out the above discussion. In Table 12.2, scenarios with the total CDP greater than 1.-6/yr are E3, E4, E7, E13B, and CVl (E2 and E3 are close). These are scenarios with moderate to heavy damage and/or relatively high frequency. In Table 12.3, the scenario-POS combinations that are the highest in frequency are E4D6, E5D6, E7D6, and E7R6, i.e., the high damage scenarios in combination with the risk significant POSs. In Table 12.5, the scenarios with the highest conditional core damage probability (CCDP) are, again, those where multiple equipment damage (preferably from both divisions) is expected. These are the scenarios E4, ES, E6, E7, E8, ElO, and CVl (all of these have a CCDP on the order of 0.5), and, to a lesser extent, M4 and M13. In Table 12.6, one can see that POSs D6 and R6 have a relatively high CCDP compared to POS RlO. In Table 12.7, for Window 1, it can be seen that the following scenarios have a conditional core damage probability of 1 or close to 1: E4, ES, E6, E7, CVl, ES, M4, MG, and ElO. Figures 12.1 through 12.5 pictorially represent some of the discussion points above. Sequence quantification report appears in Appendix M; basic event descriptions and the uncertainty report are given in Appendix N; the. total CDP cutsets are shown in Appendix o;* importance measures are printed out in Appendix P. 12.2 Uncertainty Results The uncertainty analysis was run, as explained in Chapter 11, in a pretty straightforward manner, once all the uncertainty distributions for basic events and fire scenarios were generated. The results of LHS analysis of the to recovered core damage fr~quency, from a sample size of 500 observations, are presented in Tables 12.8 .and 12,, NUREG/CR-6144 12-2
Results As can be seen, the mean value is 2.19-5/yr (vs. the point estimate of 1.68-5/yr), the median is 9.14-6/yr, while the 5th and the 95th percentile are 1.44-6/yr and 7.55-5/yr, respectively. If a lognormal distribution were fitted through the 5th and the 95th percentile, the resulting error factor would be 7.2.4. 12.3 Conclusions Risk significant scenarios are found mainly in the emergency switchgear room (ESGR), the cable vault and tunnel (CVI). In the ESGR, several important scenarios (which are also the most risk significant ESGR scenarios) occur in locations where many cables for the H and the J emergency divisions come together in a close proximity. In the CVT, the tunnel part is a constrained space, where damage would quickly propagate to both divisions (serving many different emergency equipment). In the CT, the risk significance stems from the relatively high fire frequency and non-separation of the two RHR divisions. POSs D6 and R6 are much more risk significant than RlO, with POS D6 more significant than R6. Windows 1, 2, and 3 are much more important than Window 4, and Windows 1 and 2 are more i:rnportant than Window 3. Window 2 is the most risk significant window. The most significant scenario-window-POS combination, according to Fussel-Vessely importance measure, is CV1-W2-D6 (importance of over 10%), followed by E7W2D6, E4W2D6, E5W2D6 and followed by the R6 POS in the same scenario and window combinations.
man error events are not prominent contributors individually in terms of Fussell-Vesely importance range (F-V importance of a few percent). These are the basic events that start with 'A-' (operator errors) and certain events starting with 'R-' (recovery actions, ending with '2CH' or '1869B'). The most important HEPs are recovery via charging cross-tie in window 3 and scenario CVl and operator failure to establish RHR cooling in. window 2 of scenario 2 of scenario E13B.
Part of the reason for lack of prominence of individual HEPs is that there are many HEPS, each applicable in a small fraction of sequences; another reason is in the values assigned to the HEPs; the third reason is that in many important scenarios hardware failures dominate due to heavy damage by fire. Of hardware failures the most important ones are the fire induced spurious operation of certain valves (RHR isolation valve MOVl 700 and pressurizer PORV block valves MOV1535 and 1536), inverter failures, CCW failures and plugging of the sump. Maintenance of the AFW pump 3B is very important.
12-3 NUREG/CR-6144
FIRE AREA CONTRIBUTION TO CDF ESGR .. 69~5% CVT 21.9'.%
t;QNTRIBUTIO"'°F WINDOWS*
TO FIRE CDF W1 W3.
W2 7.5% 76.4%
Sequences > 1 . E- 7
POS CONTRIBUTION TO FIRE CDF D6
. 66.*2%
R10 3.2% R6 30.6%
INIFICANT'SCENARIO, POS, WINDOW COMBINATIONS CONTRIBUTORS TO FIRE CDF w2'o*a 37.9% W1D6 12.9% W2R6 Other 20.8% 22.4% W3D6 W3R6 4.1:% .1.9%
% OF-TOTAL Sequences > 1 . E-7 *.
SCENARIO CONTRIBUTION TO CDF
E5. 16.2%
~
IJQ ~ ~ E7 *15.4% . e:;--- E3 3.8% i:,:,t:;,,,,,,,\-- E2 3. 8 % Other 16.5% El3B 6.1% CV1 21."9%
Table 12.1 Quantification of Sequences Greater than Scenario, Window, Sequence ur7/yr Core Damage Frequency (/yr)
Results POS No. E1W2D6 9 1.02-7 E2W2D6 4 2.42-7 E2W2R6 8 1.36-7 E3W2D6 4 2.42-7 E3W2R6 8 1.36-7 E4W1D6 3 3.68-7 E4W2D6 8 1.37-6 E4W2R6 8 7.68-7 E4W3D6 3 2.24-7 E4W3R6 3 1.04-7 E5W1D6 3 3.68-7 E5W2D6 8 1.37-6 E5W2R6 8 7.68-7 E5W3D6 3 2.24-7 E5W3R6 3 1.04-7 E7W1D6 7 2.86-7 E7W1D6 8 1.96-7 E7W2D6 9 8.85-7 E7W2D6 10 3.32-7 E7W2R6 17 7.45-7 E7W2R6 18 3.00-7 E10W2D6 4 1.70-7 E12W1D6 7 1.88-7 E13BW2D6 9 3.63-7 E13BW2D6 10 1.24-7 E13BW2R6 17 2.02-7 CV1W1D6 3 4.98-7 CV1W2D6 4 1.86-6 12-9 NUREG/CR-6144
Results Table 12.1 Quantification of Sequences Greater than 10*7/yr (continued) Scenario, Window, Sequence Core Damage Frequency (/yr) POS No. CV1W3D6 3 3.04-7 CV1W3R6 3 1.40-7 CI2W1D6 2 4.68-7 TOTAL 1.46-5 NUREG/CR-6144 12-10
~-
Table 12.21 Total Core Damage Frequency (/yr)
Scenario CDF (/yr) , Results El 2.96-07 E2 7.04-07 E3 7.04-07 E4 2.98-06 E5 2.98-06 E6 2.46-07 E7 2.84-06 E8 2.18-07 E9 7.42-10 ElO 3.77-07 Ell - E12 2.27-07 E13A 1.01-07 E13B 1.12-06 CTl 9.95-08 CT2 5.28-07 CVl 4.03-06 Nl 3.52-08 N2 1.62-08 N3 - Ml 5.74-09 M3 3.88-08 M4 5.23-08 M5 - M8 8.77-09 M9 - M12 - 1 Events, marked'-' have a core .damage frequency below the cutset cutoff of 1.E-10. 12-11 NUREG/CR-6144
Table 12.21 Total Core Damage Frequency (/yr) (continued) Scenario CDF (/yr) El 2.96-07 M13 9.45-09 MA 4.18-08 MC 1.35-09 MD - MG 4.42-07 MH 1.30-09 NUREG/CR-6144 12-12
sults Table 12.32 Total CDF for Each POS and Scenario Scenario POS CDP (/yr) El D6 1.96-07 El R6 8.67-08 El RlO 1.29-08 E2 D6 4.79-07 E2 R6 2.09-07 E2 RlO 1.30-08 E3 D6 4.79-07 E3 R6 2.09-07 E3 RlO 1.30-08 E4 D6 1.97-06 E4 R6 9.52-07 E4 RlO 5.31-08 ES D6 1.97-06 ES R6 9.52-07 ES RlO 5.31-08 E6 D6 1.63-07 E6 R6 7.55-08 E6 RlO 6.85-09 E7 D6 1.71-06 E7 R6 1.11-06 E7 RlO 1.32-08 ES D6 1.36-07 ES R6 6.56-08 ES RlO 1.58-08 E9 D6 2.64-09 E9 R6 7.42-10 E9 RlO -
2 Ev~nts, marked'-' have a core damage frequency below the cutset cutoff of lE-10.
12-13 NUREG/CR-6144
Results Table 12.3 Total CDF for Each POS and Scenario (continued) Scenario POS CDF (/yr) ElO D6 4.58-08 ElO R6 2.83-09 ElO RlO 9.46-09 Ell D6 - Ell R6 - ' Ell RlO - E12 D6 2.01-07 E12 R6 2.60-08 E12 RlO - E13 D6 6.84-08 E13 R6 2.88-08
;e E13 RlO 3.88-09 E13 D6 7.40-07 E13 R6 3.67-07 E13 RlO 1.08-08 CT D6 7.27-08 CT R6 2.23-08 CT RlO 4.53-09 CT D6 4.80-07 CT R6 4.78-08 CT RlO 4.64-10 CV D6 2.67-06 CV R6 1.28-06 CV RlO 7.40-08 Nl D6 2.38-08 Nl R6 1.04-08 Nl RlO 9.29-10 N2 D6 1.12-08 N2 R6 4.45-09 NUREG/CR-6144 12-14
Results Table 12.3 Total CDF for Each POS and Scenario (continued) Scenario POS CDF (/yr) N2 RlO 4.56-10 N3 D6 - N3 R6 - N3 RlO - Ml D6 4.24-09 Ml R6 1.50-09 Ml RlO - M3 D6 2.99-08 M3 R6 8.44-09 M3 RlO 3.46-10 M4 D6 2.91-08 M4 R6 2.19-08 M4 RlO 1.33-09 MS D6 - MS R6 - MS RlO - M8 D6 3.91-09 M8 R6 4.98-09 M8 RlO 7.56-10 M9 D6 - M9 R6 - M9 RlO - M12 D6 - M12 R6 - M12 RIO - M13 D6 6.65-09 M13 R6 2.80-09 Ml3 RlO - 12-15 NUREG/CR-6144
Results Table 12.3 Total CDF for Each POS and Scenario (continued) Scenario POS CDF (/yr) MA D6 1.27-08 MA R6 2.71-08 MA RlO 2.00-09 MC D6 1.07-09 MC R6 2.87-10 MC RlO - MD D6 - MD R6 - MD RlO - MG D6 4.40-07 MG R6 2.92-09 MG RlO - MH D6 9.04-10 MH R6 3.98-10 MH RlO - NUREG/CR-6144 12-16
suits
Table 12.4 CDF for Window 1 Scenario CDF (!YR)
El 4.84-08 E2 1.22-07 E3 1.22-07 E4 3.93-07 ES 3.93-07 E6 3.58-08 E7 5.20-07 E8 2.25-08 E9 2.06-09 ElO 4.86-08 Ell - E12 2.05-07 E13 2.65-08 E13 2.01-07 en 3.60-08 CT2 4.70-07 CVl 5.32-07 Nl 7.26-09 N2 3.30-09 N3 - Ml 3.30-09 M3 2.87-08 M4 1.03-08 MS - M8 1.59-09 M9 - M12 - M13 1.29-09 MA 4.15-09 12-17 NUREG/CR-6144
Table 12.4 CDF for Window 1 (continued) Scenario CDP (!YR) MC 7.94-10 MD - MG 4.40-07 MH* 5.96-10 NUREG/CR-6144 12-18
suits Table 12.53 Conditional Core Damage Probability vs. Scenario Scenario Conditional Core Damage Probability El 1.06-01 E2 9.02-02 E3 9.02-02 E4 4.94-01 ES 4.94-01 E6 4.61-01 E7 3.61-01 ES 6.34-01 E9 2.18-03 ElO 5.28-01 Ell - E12 1.93-03 El3A 1.84-03 E13B 6.51-02 CTI 8.46-03 CI2 1.81-03 CVI 4.95-01 Nl 2.05-03 N2 2.13-03 N3 - Nl 4.24-03 M3 3.92-02 M4 3.20-01 MS - MS 7.63-02 M9 -
3 Events, marked '-' have a core damage frequency below the cutset cutoff of l.E-10.
12-19 NUREG/CR-6144
Res Table 12.5 Conditional Core Damage Probability vs. Scenario (continued) Scenario Conditional Core Damage Probability M12 - Ml3 1.91-01 Ma 2.34-03 MC 3.41-03 MD - MG 6.20-02 MH 1.42-03 NUREG/CR-6144 12-20
Table 12.6 4 Conditional Core Damage Probability Vs. Scenario and POS Conditional Scenario POS Core Damage Probability El D6 1.34-01 El R6 1.26-01 El RlO 2.03-02 E2 D6 1.17-01 E2 R6 1.09-01 E2 RlO 7.28-03 E3 D6 1.17-01 E3 R6 1.09-01 E3 RlO 7.28-03 E4 D6 6.27-01 E4 R6 6.36-01 E4 RlO 3.85-02 ES D6 6.27-01 ES R6 6.39-01 ES RlO 3.95-02 E6 D6 5.84-01 E6 R6 5.75-01 E6 RlO 5.63-02 E7 D6 4.15-01 E7 R6 5.74-01 E7 RlO 7.37-03 E9 D6 1.82-02 E9 R6 1.08-02 E9 RlO - ES D6 7.56-01
4 Events, marked '---' have a core damage frequency below the cutset cutoff of l.E-10.
12-21 NUREG/CR-6144
Results Table 12.6 Conditional Core Damage Probability Vs. Scenario and POS (continued) Conditional Scenario POS Core Damage Probability E8 R6 7.74-01 E8 RlO 2.01-01 ElO D6 5.67-01 ElO R6 6.35-01 ElO RlO 5.54-02 Ell D6 - Ell R6 - Ell RlO - E12 D6 3.26-03 E12 R6 8.96-04 E12 RlO - E13 D6 2.37-03 E13 R6 2.12-03 E13 RlO 3.08-04 E13 D6 8.22-02 E13 R6 8.65-02 E13 RlO 2.75-03 en D6 1.18-02 CTl R6 7.67-03 CTl RlO 1.68-03
'Cl2 D6 3.13-03 CI2 R6 6.59-04 CI2 RlO 6.93-06 CVl D6 6.27-01 CVl R6 6.37-01 CVl RlO 3.98-02 Nl- D6 2.65-03 NUREG/CR-6144 12-22
Results Table 12.6 Conditional Core Damage Probability Vs. Scenario and POS (continued) Conditional Scenario POS Core Damage Probability Nl R6 I 2.46-03 Nl RlO 2.36-04 N2 D6 2.86-03 N2 R6 2.69-02 N2 RlO 5.82-03 N3 D6 - N3 R6 - N3 RlO - Ml D6 5.97-03 Ml R6 4.49-03 Ml RlO - M3 D6 5.78-02 M3 R6 3.48-02 CM* M3 RlO 1.53-03 M4 06 3.40-01 M4 R6 5.42-01 M4 RlO 3.55-02 MS D6 - MS R6 - MS RlO - M8 D6 6.49-02 M8 R6 1.75-01 M8 RlO 2.87-02 M9 RlO - M9 R6 - M9 RlO - M12. D6 - 12-23 NUREG/CR-6144
Results Table 12.6 Conditional Core Damage Probability Vs. Scenario and POS (continued) Conditional Scenario POS Core Damage Probability M12 R6 - M12 RlO - M13 D6 2.56-01 M13 R6 2.29-01 M13 RlO - MA D6 136-03 MA R6 6.12-03 MA RlO 4.89-04 MC D6 5.13-03 MC MC MD MD R6 RlO D6 R6 2.92-03 MD RlO - MG D6 1.18-01 MG* R6 1.66-03 MG RlO - MH D6 1.89-03 MH R6 1.76-03 MH RlO - NUREG/CR-6144 12-24
,suits Table 12.75 Conditional Core Damage Probability for Window 1 Scenario WINDOW 1 Conditional Core Damage Probability El 2.65-01 E2 2.38-01 E3 2.38-01 E4 1.0 E5 1.0 E6 1.00 E7 1.00 E8 9.97-01 E9 1.14-01 ElO 7.10-01 Ell -
E12 2.66-02 E13A 7.36-03 E13 1.79-01 CT1 4.67-02 Cf2 2.46-02 CV1 1.00 Nl 6.47-03 N2 7.32-03 M8 - Ml 3.72-02 M3 4.43-01 M4 9.64-01 M5 - M8 2.12-01 M9 - 5 Events, marked'-' have a core damage frequency below the cutset cutoff of 1.E-10. 12-25 NUREG/CR-6144
Results Table 12.7 Scenario Conditional Core Damage Probability for Window 1 WINDOW 1 Conditional Core Damage Probability M12 - Ml3 3.9-01 MA 3.54-03 MC 3.06-02 MD - MG 9.40-01 MH 9.94-03 NUREG/CR-6144 12-26
Table 12.8 Parameters of Uncertainty Distribution For Total Core Damage Frequency (1/yr) 500 Trials Results Parameter Value (/yr)
Point Estimate 1.68-5 Mean 2.19-5 (l 3.75-5 Sample minimum 2.47-7 Sample maximum 3.63-4
12-27 NUREG/CR-6144
Res Table 12.9 Quantiles of Uncertainty Distribution for Total Core Damage Frequency (/yr) Quantile (%) Value (/yr) Quantile (%) Value (/yr) 5 1.44-6 60 1.28-5 10 2.22-6 70 1.85-5 20 3.34-6 75 2.25-5 25 4.01-6 80 2.73-5 30 5.03-6 90 4.77-5 40 6.99-6 95 7.55-5 50 9.14-6 NUREG/CR-6144 12-28
13. PLANT DAMAGE STATE ANALYSIS In this section a brief summary is presented on the development and identification of plant damage states for Level 2 analysis. This process gathers the dominant core damage sequences and cutsets into plant damage states (PDSs).
The plant damage state analysis involved the identification of detailed PDS categories using a seven state indicator. The PRA model discussed in the previous sections assessed the severity of accident consequences in terms of core damage frequency. In order to assess the severity of accident consequences, the degree to which containment systems remain operable must also be determined. 13.1 Definition of Plant Damage State Indicators A total of seven PDS indicators were used to identify the unique plant damage states. The seven indicators address the following issues: Time of accident initiation Status of AC power Human error Status of RCS at onset of core damage Status of ECCS Status of recirculation spray system Status of RWST e above PDS indicators question the status of various safety systems and their actual state is indicated by various ers indicating the possible status of these variables. Table 13.1 lists a more detailed description of the indicators and their possible status with the appropriate letter designations. 13.2 PDS Analysis, Rules The PDS analysis was based on the previous event tree analysis. There were 151 cutsets above the cutoff of 10.s (1734 cutsets above the 10-10 cutoff). The 151 cutsets comprise about 90% of the CDF, and they were used to build the rules for cutset-PDS relationships. These rules were then applied to all the cutsets. The rules were developed and applied in the IRRAS code. The code automatically assigned the fire cutsets into the appropriate PDS according to the specified rules. The final cutset assignment was manually verified to insure full consistency of the results. A total of 59 different PDSs were obtained. The most dominant PDSs are designated as 2FNLRRR, 2YNLUUN and lYNGUNN. These POSS comprise about 50% of the total CDF. 13.3 PDS Uncertainty Analysis Uncertainty analysis were performed for each of the PDSs utilizing a Monte-Carlo sampling routine in the IRRAS code. The calculations for each PDS were done independently not accounting for effects represented by the sharing of many basic events through the PDSs. The point estimate results and the uncertainty results are listed in Table 13.2. 13-1 NUREG/CR-6144
- J,?""111 ~
Plant Damage State Anal Table 13-1 Plant Damage State Definitions PDS Indicator Status Designator I I I Time of Accident Initiation 1- Window 1 2-Window 2 3 - Window 3 4 - Window 4 Status of AC Power Y - Available U - Nonrecoverable blackout B - Recoverable blackout (using off-site power) F - Loss of 4 kV Bus Human Error N - No human error or nonrecoverable human error D - Diagnosis error Status of RCS at Onset of Core Damage A - Action error L - Low pressure G - 5% probability that pressure is high Status of ECCS U - Hardware Failure R - Recoverable if human error, LOSP, or 4 kV recovered C - Failure of recirculation Status of Recirculation Spray System R - Recoverable U - Nonrecoverable Status of RWST Y - Injected R - Recoverable, not injected N - Nonrecoverable, not injected NUREG/CR-6144 13-2
Table 13-2 Plant Damage States Results of Uncertainty Analysis PDS POINT MEAN 5% MEDIA 95% N 1 lUNGUU 8.88E- 6.97E- 2.20E- 2.12E-08 2.60E-07 N 08 08 09 2 lUNGUU 3.88E- 2.70E- 6.96E- 2.84E-11 1.0SE-09 y 10 10 13 3 lYAGRRR 9.06E- 9.03E- 1.09E- 2.09E-09 323E-08 09 09 10 4 lYAGRRY 1.96E- 1.68E- 1.14E- 2.56E-10 5.97E-09 09 09 11 5 lYAGRU 5.37E- 5.98E- 1.29E- 1.35E-08 2.30E-07 R 08 08 09 6 lYAGRU 1.85E- 1.34E- 4.24E- 1.33E-11 4.27E-10 y 10 10 13 7 lYDGRR 5.95E- 5.41E- 3.37E- 1.08E-10 2.02E-09 R 10 10 12 8 lYNGRRR 1.llE- 8.23E- 1.06E- 5.40E-12 3.33E-10 10 11 13 9 lYNGRRY 3.0SE- 2.0lE- 5.17E- 2.llE-11 7.87E-10 10 10 13 10 lYNGRU 7.35E- 7.34E- 4.73E- 1.28E-09 2.97E-08 R 09 09 11 11 lYNGRU 2.77E- 2.88E- 2.02E- 4.37E-09 9.75E-08 y 08 08 10 13-3 NUREG/CR-6144
Table 13-2 (continued) Plant Damage State Ana Plant Damage States Results of Uncertainty Analysis PDS POINT MEAN 5% MEDIA 95% N 12 2UNLUUN 1.05E- l.21E- 2.65E- 2.98E-07 4.47E-06 06 06 08 13 2UNLUUY 7.67E- 6.05E- 5.70E- l.48E-08 2.15E-07 08 08 10 14 2YALRRR 231E- 2.09E- 2.03E- 4.88E-09 7.880E-08 08 08 10 15 2YALRRY 3.09E- 2.92E- 3.59E- 6.55E-10 8.53E-09 09 09 11 16 2YALRUR 1.98E- 1.89E- 6.15E- 6.19E-08 7.79E-07 07 07 09 17 2YALRUY 1.14E- 1.03E- 7.91E- 2.0lE-09 5.03E-08 08 08 11 18 2YDLRRR 133E- 1.16E- 2.64E- 3.53E-10 4.57E-09 09 09 11 19 2YDLRUR 3.49E- 322E- 3.92E- 7.46E-11 l.14E-09 10 10 12 20 2YNLRRR 3.18E- 3.llE- 1.99E- 4.52E-10 l.13E-08 09 09 11 21 2YNLRRY 8.0lE- 7.97E- 5.17E- 1.0SE-10 2.84E-09 10 10 12 22 2YNLRUR 1.02E- 1.27E- 9.58E- 1.86E-09 4.22E-08 08 08 11 NUREG/CR-6144 13-4
Table 13-2 (continued)
Plant Damage States Results of Uncertainty Analysis PDS POINT MEAN 5% MEDIA 95% N 23 2YNLRUY l.85E- l.59E- l.97E- 3.SOE-09 6.73E-08 08 08 10 24 3UNLUUN 151E- l.84E- 359E- 4.82E-08 5.79E-07 07 07 09 25 3UNLUUY l.08E- 9.83E- 2.19E- 2.81E-07 3.92E-06 06 07 08 26 3YALRRR 135E- l.99E- l.96E- 2.60E-09 3.70E-08 08 08 10 27 3YALRUR l.94E- l.90E- 6.47E- 7.0lE-08 6.79E-07 07 07 09 28 3YDLRRR 1.07E- 9.42E- 2.14E- 2.85E-10 3.68E-09 09 10 11 29 3YNLRRR l.82E- l.89E- 1.72E- 8.87E-12 439E-10 10 10 13 30 4UNLUUN 1.89E- l.99E- 1.02E- 2.34E-07 7.30E-06 06 06 08 31 4UNLUUR 4.88E- 5.21E- l.94E- 5.09E-10 l.23E-08 09 09 11 32 4YALRRR 1.lOE- 6.98E- 4.47E- 9.37E-10 2.86E-08 08 09 11 33 4YALRUR l.65E- l.59E- l.93E- 2.56E-08 5.76E-07 07 07 09 13-5 NUREG/CR-6144
Table 13-2 (continued) Plant Damage State Anal Plant Damage States Results of Uncertainty Analysis PDS POINT MEAN 5% MEDIA 95% N 34 4YDLRRR 133E- 1.06E- 1.17E- 7.41E-12 3.97E-10 10 10 13 35 4YNLRRR 1.62E- 1.25E- 2.14E- 756E-11 4.38E-09 09 09 12 NUREG/CR-6144 13-6
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Westinghouse Savannah River Co. (2) Attn: Kevin O'Kula Jackie East Safety Technology Section 1991 S. Centennial Ave., Bldg. 1 Aiken, SC 29803 Donnie Whitehead Department 6412, MS: 0747 Sandia National Laboratories P.O. Box 5800 Albuquerque, NM 87185-0747 Keith Woodard PLG, Inc. 7315 Wisconsin Ave. Suite 620 East Bethesda, MD 20814-3209 John Wreathall John Wreathall & Co. 4157 MacDuff Way Dubin, OH 43017 M. K. Yeung University of Hong Kong Mechanical Engineering Dept. Polfulam HONG KONG Bob Youngblood Brookhaven National Laboratory Department of Nuclear Energy Bldg. 130 Upton, NY 11973 Carlo Zaffiro A.N.P.A. Directorate for Nuclear Via Vitaliano Brancate, 48 I-00144 Rome ITALY Dr. X. Zikidis Greek Atomic Energy Conun. N.R.C.P.S. "Demokritos" GR-153 10 Agia Paraskevi Attiki GREECE Dist-14
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (2-89) (Assigned by NRC. Add Vol., Supp,, Rev., NRCM 1102, end Addendum Numbers, if any,) 3201, 3202 BIBLIOGRAPHIC DATA SHEET (See instructions on the reverse) NUREG/CR-6144 LE AND SUBTITLE BNL-NUREG-52399 Evaluation of Potential Severe Accidents During Low Vol.3 Part 1 Power and Shutdown Operations at Surry, Unit 1: 3. DATE REPORT PUBLISHED Analysis of Core Damage Frequency from Internal MONTH YEAR Fires During Mid-loop Operations-Main Report July 1994
4. FIN OR GRANT NUMBER L1922

5. AUTHOR(S) 6. TYPE OF REPORT Z. Musicki, T ..L. Chu, V. Ho1, Y.-M. Hou 1, J. Lin 1, J. Yang, N. Siu2

7. PERIOD COVERED (Inclusive Dates/

8. p ER FORM I NG ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, provide name and mailing address.}
Brookhaven National Laboratory 1PLG Inc., 4590 MacArthur Boulevard, Newport Beach, CA 92660-2027 Upton, NY 11973 2Mn; Cambridge, MA, currently at EG&G, Idaho Falls, ID 83415
9. SPONSOR I NG ORGAN IZA Tl ON - NAME AND ADDRESS (If NRC, rype "Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address.)
Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission
11. ABSTRACT (200 words or Jess/
During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful.
12. KEY WO RDS/D ESC R l PTOR S (List words or phrases that will assist researcher.; in locating the report,) 13. AVAILABILITY STATEMENT Surry-1 Reactor-Reactor Shutdown; Surry-1 Reactor-Risk Assessment; Unlimited Surry-2 Reactor-Reactor Shutdown; surry-2 Reactor-Risk Assessment; 14, SECURITY CLASSIFICATION Failure Mode Analysis, Reactor Accidents, Reactor Core Disruption, (This Page/
Reactor Start-up, RHR Systems, Systems Analysis, Thermodynamics, Unclassified Sandia National Laboratories * (This Repon/ Unclassified
15. NUMBER OF PAGES

16. PRICE NRC FORM 335 (2-89)
Printed on recycled paper Federal Recycling Program}}

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