ML20072D058
| ML20072D058 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/31/1994 |
| From: | Lambright J, Jeffery Lynch, Ross S, Yakle J SANDIA NATIONAL LABORATORIES, SCIENCE & ENGINEERING ASSOCIATES, INC., SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-L-1923 NUREG-CR-6143, NUREG-CR-6143-V03, NUREG-CR-6143-V3, SAND93-2440, NUDOCS 9408180206 | |
| Download: ML20072D058 (112) | |
Text
NUREG/CR-6143 SAN D93-2440 Vol. 3 Eva ua~ ion of Potentia Severe Accic en::s During Low Powe.r anc Sau~:cown Oaerations a~:
Granc Gud, Eni~: 1 Analysis of Core Damage Frequency from Internal Fire Events for Plant Operational State 5 During a Refueling Outage Piejur etI by J. I ;unhn;'ht. S. Ross, J. I y nch. .l. Lkle Sandia National I.aboratories
()perated by Sandia ('orporation l
Prepared for 11.S. Nuclear Regulatory Conunission i 9408180206 940731 PDR ADOCK 05000416 P PDR i
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l
i NUREG/CR-6143 .
SAND 93-2440 Vol. 3 Evaluation of Potential Severe Accidents During Low Power anc Shutdown Operations at Granc Gulf, Unit 1 Analysis of Core Damage Frequency from Internal Fire Events for Plant Operational State 5 During a Refueling Outage hianuscript Completed: April 1994 Date Published: July 1994 i Prepared by j J. Lambright, S. Ross1, J. Lynch , J. Yakic2 Sandia National Laboratories Albuquerque, Nh187185 Prepared for Division of Safety issue Resolution !
Olrice of Nuclear Regulat 3ry Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC FIN L1923
' Science and Engineering Associates. Inc.,6100 Uptown Illvd. N.E.,
Albuquerque, Nh187110 2 Science Applications International Corporation,2109 Air Park Road S. E., Albuquerque, Nh187106
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NUREG/CR-6143 ii Vol. 3
Abstract This report presents the details of the analysis of core damage frequency due to fire during shutdown Plant Operational State 5 at the Grand Gulf Nuclear Station. Insights from previous fire analyses (Peach Bottom, Surry, LaSalle) were used to the greatest extent possible in this analysis. The fire analysis was fully integrated utilizing the same event trees and fault trees that were used in the internal events analysis.
In assessing shutdown risk due to fire at Grand Gulf, a detailed screening was performed which included the following elements:
a) Computer aided vital area analysis b) Plant inspections c) Credit for automatic fire protection systems d) Recovery of andom failures c) Detailed fire propagation modeling This screening process revealed that all plant areas had a negligible (< l.0E-8 per year) contribution to fire-induced core damage frequency.
Vol. 3 iii NUREG/CR-6143
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I NUREG/CR-6143 iv Vol 3
Contents Page Abstract . .
. . iii List of Figures . . . . .. vii List of Tables VIII Foreword . . . .. IX Acknowledgements .
xi Executive Summary . , , . ES-1 1.0 Introduction 1-1 1.1 The Grand Gulf Shutdown Fire Analysis . 1-1 1.2 Steps in the Analyses . . 1-1 1.2.1 Plant Walkdown and Data Gathering 1-1 1.2.2 Fire Risk Assessment Methodology 11 1.3 References . .
1-3 2.0 Plant Description . .
2-1 2.1 Introduction . . 2-1 2.1.1 Selection and Characterization of Plant Operational State (POS) 5 . . 2-1 2.2 Description of Plant Systems 2-1 2.2.1 liigh Pressure Core Spray System (IIPCS) . . 2-2 2.2.2 Control Rod Drive (CRD) System . . . . 2-2 2.2.3 Suppression Pool Makeup (SPMU) System . 2-2 2.2.4 Condensate CDS) System . . . 2-6 2.2.5 Low Pressure Core Spray (LPCS) System . . 2-6 2.26 Low Pressure Coolant Injection (LPCI) System . . 2-6 2.2.7 Standby Service Water Cross Tie (SSWXT) System .
29 2.2.8 Firewater (FW) System . 29 2.2.9 ResidualIIcat Removal: Suppression Pool Cooling (SPC) System 29 2.2.10 Residual ileat Removal Shutdown Cooling (SDC) System 2 14 2.2.11 Residual 1Icat Removal Containment Spray (CS) System . 2 14 2.2.12 Containment Venting System (CVS) 2 14 2.2.13 Emergency Power System (EPS) . 2-17 2.2.14 Standby Senice Water (SSW) System . 2 21 2.2.15 Emergency Ventilating System (EVS) 2 21 2.2.16 Instmment Air Systems (IAS) 2-21 2.2.17 Altemate Decay lleat Removal (ADlIR) System , 2-26 2.2.18 Reactor Water Cleanup (RWCU) System 2-26 2.2.19 Reactor Recirculation System (RRS) 2-29 2.2.20 Component Cooling Water (CCW) System . 2-29 2.2.21 Plant Senice Water (PSW) System . . 2-33 2.3 References . 2 33 Vol. 3 v NUREG/CR-6143
Contents (Continued)
P_ars 3.0 Grand Gulf Fire Analysis . .
31 3.1 Introduction 3-1 3.2 Fire Locations Analyzed 3-1 3.3 Initiating Event Frequencies . 3-1 3.3.1 Fire Data 3-1 3.3.2 Bayesian Updating of Fire Frequencies . 3-6 3.4 Determination of Fire-Induced Off-Norrnal" Plant States 3-6 3.4.1 Initiating Events . 3-6 3.5 Detailed Description of the Screening Analysis 3-8 3.6 Fire Propagation Modeling Using COMPBRN IIIe 3-9 3.6.1 Introduction . 3-9 3.6.2 Grand Gulf Fire Propagation Modeling 3-11 3.6.3 Fire Zone 1 Al 17, North flallway 93'/103' Elevation of the Auxiliary Building 3-11 3.6.3.1 Discussion 3-11 3.6.3.2 Results 3-13 3.6.4 Fire Zone 1 A201, East IIallway,119' Elevation of the Auxiliary Building 3-13 3.6.4.1 Discussion . 3-13 3.6.4.2 Results . . 3 13 3.6.5 Fire Zone i A211, North IIallway 119' Elevation, Auxiliary Building 3 13 3.6.5.1 Discussion 3-13 3.6.5.2 Results 3 13 3.6.6 Fire Zone 1 A316, North IIallway Area,139' Elevation, Auxiliary Building . 3-13 3.6.6.1 Discussion . . . 3-13 3.6.6.2 Results 3-13 3.7 Barrier Failure Analysis . 3-13 3,8 Recovery Analysis 3-18 3.9 Uncertainty Analysis . 3-18 3.10 Quantification of Unsercened Fire-Induced Core Damage Scenarios and Their Associated Fire Zones 3-18 3.10.1 Introduction . . 3-18 3.10.2 Fire Zone 1 A117 . 3-18 3.10.3 Fire Zone I A211 3-18 3.10.4 Fire Zone I A201 . 3-20 3.10.5 Fire Zone 1 A316 3-20 3.11 Conclusions . . 3-22 3.12 References . . 3-22 NUREG/CR-6143 vi Vol. 3
Contents (Cancluded)
Eue Appendix A Manual Fire Suppression Analysis A-1 A.I Backgound and Analysis . A-2 A.l.1 Fire Brigade Training, Organization and Standard Operating Procedures A-2 A.1,2 Fuel and Fire Development . A-2 A.l.3 Smoke Detector Response (DETACT) A-2 A.I.4 Time From Detection to Alarm , A-3 A.I.5 Fire and Time Related Space Conditions (ASETBX Room Fire Model) . A-3 A.1.6 Locating the Fire . . A-3 A.I.7 Agent Application A-3 A.I.8 Extinguishment . A-3 A.2 Results A-4 A.3 Summary . A-4 Appendix B Grand Gulf Critical Components by Fire Area . B-1 Appendix C Auxiliary Building Shutdown Fire Event Data C1 Figures Pace 2.1 IIPCS System Schematic . 2-3 2.2 CRD System Schematic . 2-4 2.3 SPMU System Schematic . 2-5 2.4 Condensate System Schematic 2-7 2.5 LPCS System Schematic . 28 2.6 LPCI System Schematic . 2-10 2.7 SSW Crosstie System Schematic . 2 11 2.8 Firewater System Schematic 2 12 2.9 SPC System Schematic 2-13 2.10 SDC System Schematic 2-15 2.11 CS System Schematic 2 16 2.12 CVS System Schematic . 2 18 2.13 EPS System Schematic 2-19 2.14 Diesel Generator Cross Tie Schemat;c 2-20 2.15 SSW System Schematic (Page 1 of 2) 2 22 2.15 SSW System Schematic (Page 2 of 2) 2-23 2.16 EVS Systems Schematic for Diesel Generator Rooms 2-24 2.17 EVS Systems Schematic for Safety-Related Pump Rooms 2-25 2.18 IAS System Schematic 2-27 2.19 ADIIR System Schematic . 2-28 2 20 RWCU System Schematic 2-30 2.21 RRS System Schematic 2-31 2.22 CCW System Schematic 2-32 2.23 PSW System Schematic (page 1 of 2) 2-34 2.23 PSW System Schematic (page 2 of 2) 2-35 Vol. 3 vii NUREG/CR-6143
Figures (Concluded)
Pace 3.1 General Plant Layout .
3-5 3.2 COMPBRN Zone Model . 3-10 3.3 Fire Zone 1 A117 Layout . 3-14 3.4 Fire Zone 1 A201 Layout 3-15 3.5 Fire Zone 1 A311 Layout 3 16 3.6 Fire Zone I A316 Layout . . . . 3-17 Tables 3.1 Grand Gulf Fire Zone Descriptions 3-2 3.2 Grand Gulf Fire-Induced Initiating Events Analyzed 3-7 3.3 COMBRN llie Input Parameters . 3-12
'.4 Summary of COMPRN Results 3-12 3.5 Approximate Number of Baniers at a Plant 3-19 3.6 Estimates of Single Barrier Failure Rate . . 3-19 3.7 Accident Sequence Core Damage Frequency Contributors 3-19 3.8 Fire Zone 1 Al 17 Core Damage Equation Terms (Point Estimate Values) 3-19 3.9 Fire Zone I A211 Core Damage Equation Terms (Point Estimate Values) 3-21 3.10 Fire Zone 1 A201 Core Damage Equation Terms (Point Estimate Values) 3-21 3.11 Fire Zone I A316 Core Damage Equation Tenns (Point Estimate Valuce . . 3-22 A-1 Grand Gulf Nuclear Power Plant Zone 1 Al 17 - Small Pool Fire Miscellaneous Equipment Area A-5 A2 Grand Gulf Nuclear Power Plant Zone 1 Al 17 - Large Pool Fire Miscellaneous Equipment Area A-6 A-3 Grand Gulf Nuclear Power Plant Zone I A201 - Small Pool Fire Passage Area . A-7 A-4 Grand Gulf Nuclear Power Plant Zone I A201 - Large Pool Fire Passage Area A8 A-5 Grand Gulf Nuclear Power Plant Zone 1 A211 - Small Pool Fire Miscellaneous Equipment Area . A-9 A-6 Grand Gulf Nuclear Power Plant Zone 1 A211 - Large Pool Fire Miscellaneous Equipment Area . A-10 A-7 Grand Gulf Nuclear Power Plant Zone 1 A316 - Small Pool Fire Miscellaneous Equipment Area . A-11 A-8 Grand Gulf Nuclear Power Plant Zone I A316 - Large Pool Fire Miscellaneous Equipment Area . . A-12 NUREG/CR-6143 viii Vol. 3
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 Intentially occurring only during full power operation. Some previous screening analyses that were perfonned for other modes of operation suggested that risks during those modes were small relative to full l ower operation. Ilowever, more recent studies arxl operational experience have implied that accidents during low power arxl 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 inchules two parallel projects performed by Brookhaven National Laboratory (BNL) aml Sarulia 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, arxl seismic events initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences arx1 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.
The 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 i Volume 1: Summary of Results Volume 2: Analysis of Core Damage Frequency from Internal Events for Plant Operational State 5 During a Refueling Outage Part 1: Main Report Part 1 A: Sections 1 - 9 Part IB: Section 10 Part IC: Sections 11 - 14 Part 2: Ir.ternal Events Appendices A to H Part 3: Internal Events Appendices 1 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 Flomling 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 l Volume 6: Evaluation of Severe Accident Risks for Plant Operational State 5 During a 1 Refueling Outage !
Part 1: Main Report i Part 2: Supporting MELCOR Calculations l
l Vol. 3 ix NUREG/CR-6143 i
Foreword (Continued)
For Surry:
NUREG/CR-6144 - Evaluation of Potential Severe Accidents During Low Power arxl 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 I A: Chapters 1 6 Part 1B: Chapters 7 - 12 Part 2: Intemal Events Appendices A to D Part 3: Intemal 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 Operations 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 b
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- NUIU!G/CR-6143 x Vol. 3 l 1
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Acknowledgements We would like to acknowledge Sharon Daniel and Donnie Whitehead of Sandia National Laboratories who performed the !
computer aided screening. We would further like to acknowledge Mike DiMascio of Solutions Engineering, Inc. who performed the analysis of the fire fighting effectiveness. We would also like to acknowledge the assistance ofJames Owens of Entergy Inc. whose assistance and knowledge during the plant walkdowns were invaluable. Additionally, Dena Wood of Science & Engineering Associates, Inc, deserves special thanks for her assistance in the preparation of this report.
l Vol. 3
Executive Summary A detailed fire risk assessment at shutdown has been criteria. Even for the unscreened initiating events, very performed for Grand Gulf Nuclear Station. Although the few fire zones were found to be applicable because of plant may be in various Plant Operational States (POS) at physical separation criteria. Also, relative to other plants, shutdown it was decided to examine one POS in detail. Grand Gulf utilizes more automatic fire protection systems POS 5 was chosen for this analysis. POS 5 can be entered in critical safety-related areas which in turn reduces the with decay heat as high as 0.9% of full power (34 MW); probability of damage due to a fire. Therefore, taking however, after refueling decay heat will not be higher than credit for physical separation of safety-related functions, 0.16% of full power (6MW) and this is the situation during automatic fire protection systems, lower fire initiating event hydro testing. For non-hydro conditions the plant state can frequencies, and manual fire suppression most init> ting be maintained with Residual Heat Removal (RIIR), or with events at shutdown and many fire zones were screened Alternate Decay Heat Removal System (ADHRS) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from further analysis.
afler shutdown; during hydro testing, shutdown cooling is with Reactor Water Cleanup (RWCU) letdown and Control For those remaining initators and respective fire zones, a Rod Drive (CRD) makeup, but during hydro testing RIIR detailed fire propagation analysis was performed It was and ADURS are also available. Recirculation can be either found that only in very limited areas could fire damage forced or natural. Initial level can be normal when on result in both the initiating event and other fire-related forced recirculation, raised w hen on natural circulation, er failures uhich were required to lead to core damage. Even the vessel can be full as is the case during hydro testing. in these situations other random failures (non fire-related)
Containment can be open or closed. Suppression Pool were also required to lead to core damage. Therefore, when water level can be 18 feet 4 inches,12 feet 8 inches, or taking into account reduction in fire frequency due to the empty if 170,000 gal of Condensate Storage Tank (CST) limited area ofinfluence and other random failures which water is available to IIPCS. Numerous components can be were required to lead to core damage, all remaining fire out of service for maintenance in POS 5; our model scenarios were found to be less than IE-08 per reactor year.
assumes all of train A is unavailable due to maintenance in POS 5. The techniques used in this assessment have made In all areas additional random failures of equipment full use ofinsights gained during the past 15 years in fire { damage not related to the fire itself) were regmred to occur in order to obtain core damage. Adequate separation of risk assessment. This methodology utilized previously equipment (and/or) cabling between redundant functions ;
completed intemal event fault / event tree models. Thus, the level of detail of the fire analysis is consistent with the level and the presence of automatic fire suppression systems had the efTect of reducing core damage frequency for those of detail of the intemal events analysis.
areas.
A detailed screening analysis was performed which showed i most plant areas had a negligible contribution to fire )
induced core damage frequency. For four fire zones, a detailed fire propagation analysis was performed There j were no plant areas which were found to have a contribution to core damage frequency of greater than 1.0E-8/ry.
The fire-induced core damage frequency is low er than other fire risk assessments at power due to a number of factors.
First, this plant operational state represents only three percent of the time at shutdown and shutdown fire frequencies are similar to those at power. This provides an immediate reduction in core damage frequency. Second, even if active electro-mechanical safety-related equipment is damaged by fire, an initiating event may not necessarily occur. For instance, for the loss of TBCW (Turbine Building Cooling Water) initiator to result from fire- related damage, multiple operational pumps must fail. These pumps and their associated cabling have suflicient i separation such that it is highly unlikely that a single f tre could lead to failure of all pumps. Many initiating events at shutdown were screened because of physical separation Vol 3 ES-1 NUREG/CR-6143 i
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- 1. Introduction 1.1 The Grand Gulf Shutdown Fire ouring the initial walkdown. team members visited all
, areas containing safety or support equipment. The two full Analysis days spent on site were considered adequate for this initial visit. At the completion of this initial visit, the following This report describes the shutdown fire risk analysis for the information had been obtained:
Grand Gulf nuclear plant. Although this risk assessment is intended to be complete, budget and schedule a. For each room or compartment containing considerations have dictated the use of past probabilistic essential safety equipment, identification of fire risk assessment (PRA) experience (Refs 1.1,1.2), generic sources (power cables, pump motors, solvents and data bases, and other defensible simplifications to the other transient combustibles, etc..) locations of fire maximum extent possible. barriers, fire / smoke detectors, separation of cable 13esides simplification in terms of cost reduction and minimization of execution time, the simplified fire risk b. Plant layout drawings which formed the basis for assessment described here will also meet the following division of the plant into fire areas.
additional objectives:
- c. A list of key plant personnel to be contacted later
- a. To be consistent with intemal event analyses. The ifspecific questions arose, same event trecs/ fault trees and random, common mode failure, and test / maintenance data are used- Following the initial plant visit, a list of drawings and documentation needed for further study was prepared and h To be transparent. A standard report format sent to the designated plant contacts.
should enable the reader to reproduce any of the results. Subsequent visits to the plant were made by the fire analysis personnel to conduct cable path tracing and
- c. To be realistic. flest estimate data and models are verification and to set up physical models for the fire used as much as possible. Important plant specific propagation analysis.
failure modes have been analyzed.
d To be comparable. The fire analysis is directly comparable with internal event analyses due t 13ased on an update to the nuclear power plant operating common genetic data, common methodology, experience in Reference 1.3, it has been observed that common level of detail, and presentation of typical nuclear power plants will have three to four resuhs-s gnificant fires, on average, over their operating lifetime.
. Previous probabilistic risk assessments (FRAs) at power 1.2 Steps in the Analyses have shown that fires are ollen a significant contributor to the overall core damage frequency (Ref.1.4), contributing 1.2.1 Plant Walkdown and Data Gathering anywhere from 7 percent to 50 percent of the total core damage frequency (considering contributions from T'.e Grand Gulf shutdown fire analysis began with a plant internal, seismic, flood, fire, and other events).
visit in June 1991. The initial visit served as the basis for Because of this potentially significant core damage the initial plant information request submittal. Prior to the contribution, it is important that fires be examined in detail.
first plant visit, the fire analysis team was briefed by the An overview of the simplified fire PRA methodology, intemal events systems analyst as to the general character which is outlined in Reference 1.5, is as follows:
of safety systems, support systems, system success criteria and critical interdependencies identified to date-A. Initial Plant Visits The team consisted of the following personnel:
13ased on the intemal event and seismic analyses, the PRA Project Manager - J. Lambright (Sandia National general location of safety-related wmponents of the Laboratories) systems ofinterest is known. The plant visit provides the Team Leader - J. Lambright analyst with a means of verifying the physical arrangements Fire Propagation Analyst S. Ross (SEA,Inc.) in each of these areas. The analyst completes a fire zone Fire 13rigade Analyst - M. DiMascio (Solutions checklist which aids in the screening analysis and in the Engineering,Inc.) quantification of tisk.
Vol 3 1-1 NUREG/CR-6143
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l Introduction The second purpose of the initial plant visit is to confirm detailed fire propagation modeling was with plant personnel that the documentation being used is accomplished.
in fact the best available information and to get clarification about any questions that might have arisen in a review of C. Final Quantification the documentation.
After the screening analysis has climinated all but the Also, a thorough review of fire fighting procedures is probabilistically significant fire areas, quantification of conducted. This review is performed to detennine the dominant cut sets is completed as follows:
probability of manual suppression in any given length of time for all critical plant areas. The results of this analysis 1. The temperature response in each fire area for are given in Appendix A. each postulated fire is determined.
IL Screening 2. Fire fragilities are computed. The latest version of ,
the fire growth code COMPBRN (Ref.1.6) was It is necessary to select those fire locations within the used to calculate fire propagation and equipment power plant having the greatest potential for producing damage.
risk-dominant accident sequences. The objectives of kication selection are somewhat competing and should be 3. The probabilities of barrier failure for all balanced in a meaningful risk assessment study. The first remaining combinations of adjacent fire areas are objective is to maximize the possibility that all impodant assessed. A barrier failure analysis was conducted locations are analped, and this leads to the consideration of for those combinations of two adjacent fire areas a potentially large number of candidate locations. The which, with or without additional random failures, second objective is to minimize the efTort spent in the remained after the screening analysis. The quantification of event trees and fault trees for fire k> cations methodology to assign barrier failure probability that turn out to be unimportant. A proper balance of these to the fire area combinations is described in objectives is one that results in an ideal analysis alk) cation Section 3.7.
of resources and efliciency of assessment.
- 4. A recovery analysis is performed. In a similar l The screening analysis is comprised of: fashion as in the internal events analysis, recovery of non-fire related random failures was addressed.
- 1. Identification of potentially important fire areas. Appropriate modifications to recovery l Fire areas which have either safety-related probabilities were made as described in Section equipment or power and control cables for that 3.8.
equipment were identified as requiring further analysis. This group of fire areas is briefly 5. An uncertainty analysis is perfonned to estimate described in Section 3.2. All critical safety error bounds on the computed fire-induced core components within these fire areas are given in damage frequencies. The IRRAS code (Ref.1.7)
Appendix 11 is utilized in the uncertainty analysis as described in Section 3.9. Since all remaining fire areas were
- 2. Screen fire areas for probable fire induced found to have a core damage frequency initiating events. Determination of the fire contribution of <l .0E8/y, no uncertainty analysis frequency fbr all plant kications and determination was perfonned for this study.
of the resulting fire-induced initiating events and "off-normal" plant states are delineated in Sections In Section 3.10, a detailed description of all fire scenarios 3 4 and 3.5, respectively. which survived the initial screening process and their associated fire areas is given. Descriptions of all factors
- 3. Screen fire areas both on order and frequency of used in the final quantification of these fire areas are cut sets. A cut set is defined as a minimal delineated.
combination of fire-related and random failures i which lead to core damage.
- 4. Each fire area remaining is numerically evaluated and culled on frequency. The screening methodology (Section 3 5) describes how reduction of the initial group of h3 cations from Section 3.2 to the fbut remaining which underwent NUREG/CR-6143 1-2 Vol. 3
Introduction 1.3 References 1.5 M. P. Bohn and J. A. Lambright, Procedures for 1.1 U.S. Nuclear Regulatory Commission,' Reactor the Extemal Event Core Damace Freauency Risk Reference Document. NUREG-1150, Vol.1, Analyses for NUREG-1150. NUREG/CR-4840, Febg 1987. SAND 88-3102, Sandia National Laboratories, Albuquerque, NM, December 1990.
1.2 U.S. Nuclear Regulatory Commission, Egactor Safety Study: An Assessment of Accident Risks .
1.6 V. Ilo, et al., COMPBRN lHe: An Interactive inJJ S. Nuclear Power Plants. WASW1400, Computer Vode for Fire Risk Analysis. EPRI NUREG-75/014,1975.
NP-7282, May 1991.
1.3 W. T. Wheelis, Users' Guide for a 1.7 K. D. Russell et al., Intecrated Reliabih.ty and Risk Personal-Computer-Based Nuclear Power Plant Analysis System GRRAS) Version 4.0.
Fire Data Base. NUREG/CR-4586, SAND 86-NUREG/CR-5813, EGG-2664 EG&G Idaho, 0300, Sandia National Labo atories, Albuquerque, Januaq 1992.
NM, August 1986.
l.4 J. A. Lambright, et al., Fire Risk Sconine Study-Current Perception of Unaddressed Fire Risk Issues. NUREG/CR-5088, SAND 88-0177, Sandia National Laboratories, Albuquerque, NM, December 1988.
Vol. 3 'l -3 NUIEG/CR-6143 t , ,
- 2. Plant Description 2.1 Introduction (5) POS 5 consisting of: OC 4 (T 5 200 degrees F) and OC 5 until the vessel head is off The Grand Gulf Nuclear Station has one Boiling Water Reactor (BWR) of 1250 megawatts (electrical) capacity and (6) POS 6 consisting of: OC 5 with the head off is housed in a Mar k Ill containment. The plant is operated and level raised to the steam lines by Entergy Incorporated, is located in Port Gibson, Mississippi, and began commercial operation in July 1985. (7) POS 7 consisting of: OC 5 with the head otT, the upper pool filled, and the refueling transfer
- "
- P""'
2.1.1 Selection and Characterization of Plant Operational State (POS) 5 Grand Gulf can be in POS 5 with a variety of conditions that afTect the availability of mitigating equipment should This section discusses the POSs at Grand Gulf Nuc! car an accident occur. POS 5 can be entered with decay heat as Station and describes the characteristics of POS 5 which high as 0.9% of full power (34 MW); however, after was selected for this analysis. A POS is a plant condition refueling decay heat will not be higher than 0.16% of full for which the status of plant systems (operating, standby, power (6MW). This is the situation during hydro testing.
unavailable) can be specified with sufficient accuracy to For the non-hydro situation the plant state can be with model subsequent accident events. A POS is not identical RIIR, or with ADIIR 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> afler shutdown; during to a mode or operating condition, but is defined based on hydro testing, shutdown cooling is provided by the operating conditions. The operating conditions at Grand combination of RWCU letdown and Control Rod Drive Gulf are defined as follows: (CRD) makeup, but during hydro testing R1IR and ADIIR are also available. Recirculation can be either forced or (1) OC 1, Power Operation: hkide Switch in Run. natural. Initial level can be normal when on forced Any Temperature recirculation, raised when on natural circulation, or the vessel can be full as is the case during hydro testing.
(2) OC 2, Startup: Mode Switch in Startup/Ilot Containment can be open or closed. Suppression Pool Standby, Any Temperature water level can be 18 feet 4 inches,12 feet 8 inches, or empty if 170,000 gal of CST water is available to IIPCS.
(3) OC 3,Ilot Shutdown: Mode Switch in Numerous components can be out of service for Shutdown, Temperature Greater than 200"F maintenance in POS 5; our model assumes all of train A is unavailable due to maintenance during operation in POS 5.
(4) OC 4, Cold Shutdown: Mode Switch in Shutdown, Temperature 200*F or Lower (5) OC 5, Refueling: Fuelin Vessel with Head 2.2 Description of Plant SySteniS Detensioned or Removed, Mode Switch in . .
Shutdown or Refuct, Temperature 140*F or The followm.g sections provide the system descriptions and lower' system models of the major front hne and support systems 1 identified as important to mitigating the effects of fire-The POSs are then delimed based upon the operating induced plant transients during POS 5 (Ref. 2.1). Event conditions stated above as follows: tr es and component fault trees were developed by the ;
internal events analysts. This study utilizes the same event (1) POS 1 consisting of: OC 1 and OC 2 with trees, fault trees, and accident sequences developed during pressure at rated conditions (about 1000 psig) the intemal events analysis to ensure consistency (Ref. 2.2).
and thermal power no greater than 15% rhe discussion of the systems that follow melude:
(2) POS 2 consisting of: OC 3 from rated pressure a. A brief functional description of each system with to 500 psig reference to the one-line diagrams that were developed to mdicate which components were (3) POS 3 consisting of: OC 3 from 500 psig to included in the model.
where R1IR/SDC is initiated (about 100 psig)
- b. Interfaces and dependencies between the fronth.ne (4) POS 4 consisting of: OC 3 with the unit on systems and the support systems.
RHR/SDC i
I Vol 3 21 NUREO/CR-6143
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Plant Description Two discharge paths are provided for the CRD pumps. The 2,2.1 liigh Pressure Core Spray System first path is through the Hydraulic Control Units' (IICUs)
(IIPCS) cooling headers. Flow is controlled by one of two air perated control valves. When instniment air is lost, the in POS 5 the function of the HPCS system is to provide wntrol valves fail "as is". The second path is through the coolant makeup to the reactor vessel in order 9 tuamtain IICU charging headers. This path is upstream of the proper water level and/or flood the reactor. The HPCS mntr i valves and fails open on loss of air. Ilowever, with system consists of a single train with motor- >perated valves both CRD pumps running and the reactor at nominal and a motor driven pump. Suction is taken ' rom either the pressure, the second discharge path restricts flow, by means Condensate Storage Tank (CST) or the suparession pool.
f an wince, to approximately 165 gpm. This flow rate is fnjection to the reactor vessel is via a spra' ring mounted assumed m, su0icient for core cooling and thus no credit is inside the core shroud. The pump is capable of delivering taken for this discharge path.
550 gpm against a reactor pressure of 1177 psig and a full How of 7115 gpm against a reactor pressure of 200 psig.
Normally one CRD pump is nmning with the suction and The total maximum pump run out flow is 9100 apm. A discharge valves to the standby pump being open. Should simplified schematic of the 1IPCS is provided t,y Figure the operator be required to realign the CRD system as a 2.1. Maior system components are represented with valves s urce f early high pressure mjecuon, the standby CRD shown in their nonnal standby position.
pump must be placed into operation and one air operated control valve must be fully opened to achieve suflicient The 1IPCS system is automatically initiated and controlled D w to We reacta wsscL However, operator intervention is required to throttle flow to prevent the 1IPCS injection valve from opening and The CRD success criteria require that both pumps be closing in response to the reactor vessel level. The operator running and the IICU cooling header discharge path be may also be required to manually stad the system, if an available when CRD is required at the start of the accident automatic start failure occurs. as the only coolant makeup source. However, when c lant makeup has been provided for a period of time and The success criterion for the HPCS system is injection of then lost, only one CRD pump is required.
flow to the reactor vessel.
CRD Pump A is powered from Division 14160 V AC Bus The 1IpCS system major dependencies are DC control q AA with control and actuation power supplied by power for initiating the actuation relay logic and HPCS Divismn i 125 V DC Bus 11 DA. CRD Pump B is pump breaker, AC power for operating the HPCS pump and p wered from Division 2 4160 V AC Bus 16AB with valves, and iIPCS pump room cooling. ,
control and actuation power supplied by Division 2125 V
"" " O' 2,2.2 Control Rod Drive (CRD) System The CRD hydraulic system was modeled as a backup 2,2,3 Suppression Pool Makeup (SPMU) source of high pressure mjection.
System The CRD pumps take suction from the condenser hotwell makeup / reject line. Makeup to the condenser hotwell is The SPMU system provides water from the upper provided by the CST. Excess condensate from the containment pool to the suppression pool following a condenser is rejected to the CST by the condensate system.
From the condenser hotwell makeup / reject lines, water I.OCA. Water which gravity flows from the upper containment pool to the suppression pool is of suflicir-flows to the CRD pumps through a pump backwash suction filter and one of two pump suction filters A simplified quantity to keep the uppermost drywell vents covered titt schematic of the CRD system is provided by Figure 2.2. most conceivable accidents.
The SPMU system consists of two lines which penetrate the The CRD pumps, together, can achieve a Dow rate of side walls in the separator storage area of the upper approximately 238 ppm with the reactor at 1103 psia. Each containment pool. These lines are routed down to the pump is provided with a minimum flow line which recirculates 20 gpm back to the CST. This minimum suppression pool on either side of the steam tunnel. A simplified schematic of the SPMU system is provided by flow line prevents the pump from operating at shutotT head u here the pump would overheat and possibly be damaged. Figure 2.3.
2-2 yot 3 NUREG/CR-6143
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-Vol. 3 2-5 NUREG/CR-6143 i
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Plant Description ne pml makeup lines each have two normally closed, during normal plant operation. The systems required to motor-operated butterfly valves in series. All motor- maintain condenser vacuum are not required since the operated valves are powered from onsite emergency power condensate system is unaffected by loss of condenser sources maintaining divisional separation and redundancy. vacuum. In addition, parts of the condensate system (e g.,
the low pressure heaters) are not required to function. The The upper pool is dumped by gravity llow when the valves condensate pumps are powered by non safety 4.16 kV receive a divisionally separate but simultaneous signal to buses. Power to condensate system motor-operated valves open. The open signal for each valve division is generated is also provided by non-safety related buses (480 V). The in any of three ways: instrument air system is required to supply air to the condenser makeup valve and also to open the feedwater a By low-low suppression pool level, providing a startup valve. Makeup to the condenser is provided by the LOCA signal has been generated or the condensate and refueling water storage and transfer system.
Emergency Core Cooling System (ECCS) has been manually initiated This ensures adequate 2.2.5 Low Pressure Core Spray (LPCS) water volume in the suppression pool to keep the System suppression pool vents covered for all break sizes.
The function of the LPCS System is to provide coolant to
- b. By a timer, thirty minutes aller a LOCA s.ignal has the reactor vessel during accidents in which vessel pressure been generated. This ensures an adequate long is low. The LPCS system is a single train system consisting term heat smk is available regardless of break s.ize.
of motor-operated and manual valves and a motor-driven
. pump. The LPCS pump is rated at 7115 gpm with a
- c. By manual initiation, provided a LOCA signal .is discharge head of 319 psig. The LPCS pump takes water present or the ECCS has been manually imtiated. from the suppression pool through strainers located 10 feet above the suppression pool floor. A simplified schematic In addition, m order to actuate each SPMU valve by any of de LPCS is provided by Figure 2.5.
the three methods hsted above, the mode selector handswitch for each division must be in AUTO position The LPCS system is automatically initiated and controlled.
and the reactor mode switch must not be in REFUEL The operator may be required to manually start the system p sition if an automatic actuation failure occurs.
The SPMU system requires electrical power for operation. The success criterion for the LPCS system is injection at The two redundant SPMU lines are cach powered from rated flow to the reactor vessel.
separate emergency electrical buses. The Tram A valves are powered lly emergency AC Division 1 MCC 15B21 , The LPCS system major dependencies are DC control while the 'l ram B valws are pu ed by AC Division 2 power for initiating the actuation relay logic and LPCS MCC 16B41. The initiaton log r - fram A and B is pump breaker, AC power for operating the LPCS pump and powered by 125 V DC Divisions i and 2, respectively. valves, and LPCS pump room cooling.
2.2.4 Condensate (CDS) System The DC power is provided by Division i 125 V DC Panel I E124Bl. Power for the L.PCS pump is provided by Credit for the condensate system as a low pressure injection Division 14160 V AC Bus 15AA, and power for the valves system is taken in this study. The condensate system has is provided by Division 1480 V AC MCC 15B11.
i three main condenser units, three condensate pumps, three condensate booster pumps, three strings of four low 2.2.6 Low Pressure Coolant Injection (LPCI) pressure heaters, a condensate drain tank and associated l valvea, piping, instrumentation, and controls. The System condensate system supplies water to the reactor vessel The function of the LPCI system is to provide coolant to through the feedwater startup valve AV513. A simplified schematic of the condensate system for use as a low the reactor vessel danng accidents m which system pressure is low. The LPCI system is one mode of the RIIR pressure injection system is shown in Figure 2.4. The system and, shares components with other modes. The success criteria for the condensate system in POS 5 is one LPCI system is a three tram system consisting of motor-of three condensate pumps operating with a flow path to the perated valves and motor-driven pumps. The three pumps reactor through the feedwater start-up flow control valve.
are each rated at 7450 gpm. Trains A and B each have two heat cxchangers in series downstream of the pump. Train C The system dependencies for the condensate system as a ,
is injection dedicated and has no heat exchangers. Cooling low pressure injection system are less demanding than NUIUiG/CR-6143 2-6 Vol. 3
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Plant Description
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Plant Description water flow to the heat exchangers is not required for the three train system consisting of one motor driven pump and LPCI mode. A simplified schematic of the LPCI system is two diesel-driven pumps. The pumps can each provide provided in Figure 2.6. 1500 gpm at 125 psig. The pumps feed into a common header that supplies water to the fire hoses and sprinkler The LPCI system is automatically initiated and controlled systems. The pumps take suction from two 300,000 gallon Ilowever, operator intervention may be required to water storage tanks. Any pump can take water from either manually realign and stan the system in POS 5 since the tank. The fire hoses are connected, via an adapter to individual RIIR Trains A and B could be aligned for either various test connections in the auxiliary building. These shutdown cooling or ADIIRS, and Train C could be aligned connections feed into various injection systems and water for ADI1RS, Given any of these configurations, the can be injected through the systems' injection valve. T1.e associated train would not automatically initiate for LPCI pumps are located in the firewater pump house. A operation simplified schematic of the firewater system is provided in Figure 2.8.
The success criterion for the LPCI system is injection of flow from any one pump to the reactor vessel. The firewater system, when used for injection, must be manually initiated and controlled. The operator is required The LPCI system major dependencies are DC control to align the system and to start the pumps.
power for initiating the actuation relay logic and RIIR pump breakers, AC power for operating the RIIR pumps The success criteria for the firewater system is injection of and valves, RIIR pump cooling, and RIIR pump room flow from any one pump.
cooling.
The two diesel-driven firewater pumps have no outside The DC power to Train A is provided by Division 1 125 V interfaces or dependencies, each pump has self-contained DC for Trains B and C it is provided by Division 2125 V batteries that provide it with staning power. The electric DC. Power for RIIR Pump A is provided by Division 1 motor-driven pump requires balance of plant AC power.
4160 V AC Bus 15AA. Power for RIIR Pump B and LPCI Pump C is provided by Division 2 4160 V AC Bus 15AB. 2.2.9 Residual Heat Removal: Suppression All pumps require pump cooling.
Pool Cooling (SPC) System 2.2.7 Standby Service Water Cross-Tie The function of the SPC system is to remove decay heat (SSWXT) System from the suppression pool'during an accident. The SPC system is but one mode of the RilR system and, as such, The SSW cross-tic system is used to provide a coolant shares components with other modes.
makeup source to the reactor vessel during accidents in w hich nonnat sources of emergency injectmn have failed The SPC system is a two train system consisting of The SSW cross-tic system is comprised of Train B of the motor-operated valves and motor driven pumps. Both SSW system and Train B of the LPCI system. trains have two heat exchangers in series downstream of the pump. Each pump is rated at 7450 gpm. Cooling water The SSW cross-tie system uses SSW Pump B to inject flow to the heat exchanger is required for the SPC mode.
water into the reactor via the LPCI system Train B injection The SPC suction source is the suppression pool. A lines. A simplified schematic of the SSW cross tic system simplified schematic of the SPC (RIIR) system is provided is provided in Figure 2.7. Major system components are by Figure 2.9. Major system components are shown with shown in their normal standby position with valves shown. valves shown in their normal standby operating position.
The SSW eross-tie system has no automatic actuation. The system must be manually aligned and manually actuated. The SPC system is manually initiated and comrolled. The operator is required to align the system and to start the The dependencies for this system are the same as those for pumps. In POS 5 the RIIR system configuration is SSW Train B (Section 2.2.14) and LPCI Train B (Section accident sequence dependent, i.e., RIIR Train B can be 2.2.6). aligned in (1) Standby LPCI mode, (2) SDC mode, or (3)
ADlIRS. The operator action to align for SPC is dependent 2.2.8 Firewater (FW) System n the prior RIIR configuration.
The firewater system was modeled as a backup source of The success criterion for the SPC system is injection of low pressure injection as well as a source for automatic 11 w fr m any ne pump / heat exchanger train to the 8"PP' 88I " P I' and manual fire suppression. The firewater system is a Vol. 3 2-9 NUREG/CR-6143
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Plant Description The SPC system major dependencies are DC control power 2.2.11 Residual 11 eat Removal: Containment for actuation, AC power for operating the RIIR pumps and Spray (CS) System valves, RIIR pump coolmg, and RIIR pump room cooling.
The function of the CS system is two-fold: (1) to suppress The DC power to Train A is provided by Division 1 125V pressure in the containment during an accident and (2) to DC; for Train B it is provided by Division 2125 V DC. remove Dssion products from the containment atmosphere Power for RI1R Pump A is provided by Division 14160 V following core damage. The CS system is one of the AC Bus 15AA. Power to RilR Pump B is provided by modes of the RIIR system and shares components with Division 2 4160 V AC Bus 16AB Both pumps require other modes.
pump cooling.
The CS system is a two loop system consisting of motor-operated valves and motor-driven pumps. There are 2.2.10 ResidualIIcat Removal: Simtdown two heat exchangers in series per loop. Each pump is rated Cooling (SDC) System at 7450 gpm. Cooling water Dow to the heat exchanger is required for CS when used to suppress pressure in the containment. The CS suction source is the suppression The ftmetion of the SDC system in POS 5 is to remove decay heat during shutdown and during accidents in which pool. A simplified schematic of the CS system is provided reactor vessel integrity is maintained. The SDC system is by Figure 2.11. Major system components with valves are one of the males of the RIIR system and shares shown in their normal standby operation position.
components with other modes.
The CS system is automatically initiated and may be controlled. Ilowever, operator intervention is required to The SDC system is a two train system consisting of motor-operated valves and motor driven pumps. Both manually realign and start the system In POS 5 RIIR Train trains have two heat exchangers in series dov i stream of the B can be aligned for either shutdown cooling or ADHRS.
Given these configurations, the associated train would not pump Each pump is rated at 7450 gpm. Coohng nater Dow to the heat exchanger is required for the SDC mode. automatically initiate and spray.
The SDC system suction source is one recirculation pump's The success criterion for the CS system is injection of flow suction line. A simplified schematic of the SDC (RIIR) from any one pump / heat exchanger train to the spray ring.
system is provided by Figure 2.10. Major system components are shown with valves in their normal standby The CS system major dependencies are DC control power operating position.
for actuation, AC power for operating the RHR pumps and in POS 5 cither SDC or ADIIRS is in operation removing valves, R1IR pump cooling, and RIIR pump room cooling.
decay heat. If ADI1RS is in operation one Train of SDC is in standby. From standby, the SDC is manually aligned The DC power to Train A is provided by Division 1 125 V and started when placed in operation. Note that for the DC and for Tr ain B it is provided by Division 2125 V DC.
Power for RilR Pump A is provided by Division 14160 V POS 5 analysis Train A of R1IR has been assumed to be AC Bus 15AA. Power to RilR Pump B is provided by unavailabic due to maintenance.
Division 2 4160 V AC Bus 16AB. Both pumps require The success criterion for the SDC system is injection of pump cooling.
flow from any one pump / heat exchanger train to the reactor vessel. 2.2.12 Containment Venting System (CVS)
The SDC system major dependencies are DC control power When suppression pool cooling and containment sprays for actuation, AC power for operating the RIIR pumps and have failed to reduce primary containment pressure, the valves, RIIR pump cooling, and RI1R pump room cooling. CVS is used to prevent a primary containment pressure limit from being exceeded.
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The DC power to Train A is provided by Division 1 125 V DC, and for Train B it is provided by Division 2125 V DC. The vent path used is a 20-inch diameter purge exhaust line Power for RHR Pump A is provided by Division 14160 V which is part of the containment ventilation and filtration AC Bus 15AA. Power to R1IR Pump B is provided by system. This line includes four air-operated dampers which Division 2 4160 V AC Bus 16AB. All pumps require are normally closed All four fait closed on loss of air.
pump cooling.
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Plant Description Two of the dampers are closed by a containment isolation For Divisions I and 2, when a loss of normal power signal signal. The other two are closed by the standby gas occurs, the diesel generators automatically start and connect treatment system initiation. The CVS discharges to the roof to the associated ESF bus if no other source of power is of the auxiliary building. A schematic of the CVS is shown available. To prevent overloading the diesel generator in Figure 2.12. w hen no alternate source is available, unnecessary loads are shed from the associated bus and those loads required for The venting pucedure requires containment venting when plant safety are sequenced onto the bus. For Division 3, the pressure exceeds 17.25 psig. Venting requires that the when a loss of normal power occurs, the diesel generator operator jump the isolation relays for each damper and then will start and automatically close on the bus when at speed open them (they are located on back panels in the control and voltage.
room) The actual venting procedure can only be initiated by order of the emergency director. If Divisions 1 and 2 dicsci generators fail to power their buses, power can be supplied to certain Division 1 or 2 Containment venting requires instrument air for opening the loads from the HPCS diesel generator. This is air-operated dampers. The dampers also require power accomplished by isolating the normal Division 3 laads from from emergency AC Divisions 1 and 2 for operation of the the dicsci and connecting either the Division 1 or 2 loads to solenoids. the HPCS diesel generator. The electrical equipment that is involved in accomplishing this is shown in Figure 2.14.
Instrument air to the auxilian building is isolated by a 1.OCA signal (high drywell pressure); and instrument air to The ESF 125 V DC system includes three divisions, each the containment is isolated by a containment isolation consisting of two battery chargers which normally supply signal. Since both signals are expected to be present during the load and a bank of batteries which functions as a a venting situation, the operator must also restore backup. Divisions I and 2 of the ESF DC (Buses 11DA instrument air to successfully vent. and 1 i DB, respectively) system supply the majority of the ESF loads. Both are rated at 1600 amperes. Division 3 2.2.13 Emergency Power System (EPS) (Bus i 1 DC) is dedicated to the IIPCS system and is rated at 100 ampere hours.
The EPS consists of the AC and DC power divisions required by all systems (except firewater) needed to The battery chargers normally supplying power to each mitigate postulated accidents. This includes Balance or ESF bus are silicon controlled, rectifier type chargers rated Plant (BOP) and Engineered Safety Feature (ESF) buses. at 400 amperes,125 V DC. The ESF battery chargers Both ESF AC and DC power are divided into three separate maintain the terminal voltage of the associated batteries above a minimum of 1.75 volts per cell. Either charger can divisions. Two of the divisions (1 and 2) are for the majority of the ESF and the third (3) is dedicated to the restore the batteries from this minimum voltage to their IIPCS system and its required support systems. The EPS is fully charged state within eight hours under normal plant shown in Figure 2.13. operating conditions.
1 he ESF AC divisions normally receive power fmm one of Each ESF DC battery bank consists of sixty lead-calcium three ofTsite sources through ESF (34.5 LV/4.16 kV) In type cells connected in series to produce the rated output of addition to the normal supply fmm the ESF transformers, 125 V DC. Each ESF battery bank can supply the required each ESF 4.16 kV bus has a standby diesel generator which DC loads for cleven hours aller a loss of AC power if is available to supply bus loads upon a loss of nonnal AC unnecessary loads are shed.
power. These diesels may be started manually or automatically. The diesels supplying Divisions 1 and 2 Each diesel generator has six subsystems required for its buses are rated at 7000 kW and start on a loss of normal operation: (1) fuel oil subsystem, (2) air starting subsystem, AC power to the associated bus, low reactor level of-l 50 (3) lube oil subsystem, (4) jacket water cooling subsystem, inches, or high dqwell pressure of +2 psig. The diesel (5) combustion air intake, exhaust and crankcase supplying Division 3 buses (rated at 3300 kW) is ventilation, and (6) standby generator excitation subsystem.
exclusively for the IIPCS and stants on a loss of normal AC All of these subsystems are normally treated as part of the power, low reactor water level (-42 inches), and high diesel generator. Ilowever, some of these other subsystems dryuell pressure signal of 42 psig. For Divisions 1 and 2, are dependent on operation of other systems.
the transfer of power from normal to backup or emergency power supplies is controlled by the load shedding and sequencing system.
Vol. 3 2-17 NUREG/CR-6143
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Plant Description 2.2.14 Standby Service Water (SSW) System generator startup and stop on diesel generator shutdown.
The damper opens on the same signals. A heating coil has The function of the SSW system is to provide heat removal been provided to maintain the minimum required air temperature during cold weather. The ventilation system from plant auxiliaries that require cooling water during an for the diesel generato, rooms is shown in Figure 2.16.
emergency shutdown of the plant.
The SSW system is made up of three independent trains. The cooling and emergency ventilation systems for the Each t.ain consists of a motor-driven pump, motor-operated safety-related pump rooms are shown schematically in valves, and heat exchangers. Train C is dedicated to Ge Figure 2.17. Each safety-related pump room is provided with one f ai-capacity fan-coil unit to prevent the room 1IPCS system-temperature from exceeding I 50*F during pump operation.
SSW Pumps A (unavailable for POS 5) and B are vertical, The SSW system provides cooling water for the fan-coil centrifugal pumps, each with a 12,000 ppm capacity. SSW units. The units start automatically when the associated Pump C is also a vertical, centrifugal pump, but with only a 1300 2,m capacity. Eacn pump takes water from the ECCS pump starts.
cooling ower basins, circulates water through the heat During normal plant operation, the safety-related pump exchangers for each load, and returns the water to the rooms and penetration rooms are maintained at a slight
, % through a motor-operated discharge valve. Each tm t.as its own discharge valve. A simplified schematic negative pressure with respect to the corridors by the fuel of the SSW system is provided in Figure 2.15.
handling area ventilation system. Supply air is provided fmm the auxiliary building ventilation system. Air is drawn from the safety-related pump rooms and discharged The SSW system is automatically initiated and controlled.
Ilowever, operator intervention is required to manually start by the fuel handling area exhaust fans to the fuel handling area vent. The success criterion for the diesel generator the system gisen an auto-start failure.
rooms requires operation of the fan with the damper The SSW system major dependencies are DC control power opened.
for initiating the actuation relay logic, and AC power for
%ccess of the ECCS pump roorn systems involves operating the SSW pumps and valves. The pumps are operation of the fan-coil units with associated cooling by self-cooled.
the Standby Senice Water System.
The DC power to Trains A, B and C is provided by the Division 1 125 V DC, Division 2125 V DC, and Division 3 It is assumed that failure of the EVS would fail operating 125 V DC buses, respectively. Power for SSW Pump A is diesel generators in filleen minutes. The low pressure provided by Division 14160 V AC Bus 15AA. Power for ECCS pumps are assumed to fail within four hours after loss of the associated room cooling. IIPCS is assumed to SSW Pump B is provided by Division 2 4160 V AC Bus fail within twelve hours afler loss of room cooling.
16AB. Power to SSW Pump C is provided by Division 3 480 V AC Bus 17B01.
The ECCS pump room coolers are all cooled by the Standby Service Water system. Power for the fans for the 2.2.15 Emergency Ventilating System (EVS) RCIC, ITCS, and RIIR-A pump room is provided by AC Division 1. The fans for RiiR-B and RIIR-C pump rooms The objvetive of the EVS is to maintain suitable are p wered by Dmsion 2. The HPC5 pump room fan is temperatures in safety related equipment rooms to preclude p wered by iIPCS dedicated AC Dm,ston 3.
component failures.
The EVS cools the following: (1) standby diesel genera:or 2.2.16 Instrument Air Systera GAL rooms, (2) senice water pump rooms, (3) pump rooms for the RilR, RCIC, I!PCS, and ifCS pumps, and four rooms The IAS provides a pn.umatic supply to support operation containing electrical switchgear. The senice water pump of safety related equipment.
room system is assumed not to be required. Three
- independent subsystems, one per diesel generator room, The Instmment Air System for each unit consists of one each having 100% capacity, are provided for the emergency full-capacity, muhistage, packaged centrifugal compressor, diesel generator rooms to maintain a design temperature of complete with inlet filter, inlet air controller, and 120*F. Each diesel unit is provided with a fan damper aftercooler. The compressor has a receiver and a system connected to the respective dicsci engineered safety regenerative desiceant air dyer. The Unit 2 IAS features bus. The fan is controlled to start on diesel compressor, receiver, and dryer are present and operational.
Vol. 3 2-21 NUREG/CR-6143
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Plant Description c
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NUREG/CR-6143 2-22 Vol. 3
Plant Description PS-178
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Vol. 3 2 23 NUREG/CR-6143
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Plant Description Instrument air is distributed to the plant aner drying to a 2.2.17 Alternate Decay IIcat Removal dew point of-40*F and filtration of particles 0 9 microns (ADHR) System and larger. In addition, each instrument or group of instruments has a pressure regulating valve with an integral The function of the ADlIR system is to provide an attemate filter k>cated in its instrument air supply. The instrument inethod of decay heat removal during cold shutdown and air systems of Units 1 and 2 can be cross-connected by
, refueling when maintenance is being performed on the openmg two air-operated valves from the control room. RIIR shutdown cooling loops or associated support One instrument air compressor can supply all instrument air systems' demands with the other compressor as a backup. The air-operated cross-connect valves fail open upon loss of air The ADIIR system consists of components common to the to their operators, thereby assuring instrument air supply t RIIR system including the RIIR common suction line, fuel Unit 1 from either instrument air system.
pool cooling and cleanup piping, and the RIIR Train C
. LPCIinjection header. Components exclusive to ADIIR The Service Air System (SAS) is also arranged as an include two ADIIR pumps, two heat exchangers, associated automatic backup supply to the instrument air system piping, valves, instmmentation and controls, through a control valve that opens upon reduced hne pressure in the instrument air system. The backup The ADIIR can operate in four main modes of which one is connection is upstream of the mstrument air dryers. Credit modeled for POS 5. In POS 5, ADIIR is used in the for the SAS is taken m this study and included under the reactor vessel cooling mode via RIIR B. During the reactor lAS discussion vessel cooling mode, ADIIR draws water from the existing
. RilR common suction line. The reactor coolant is then The SAS consists of two full-capacity, multistage, . pumped from the reactor recirculation loop through valves packaged centrifugal compressors, each complete with F066A and F006A or valves F066 and F006B to the ADlIR mlet filter, mlet air now controller, anercooler, and . pumps, then to the heat exchangers and back to the reactor receiver. A simplified schematie of the IAS and SAS is vessel via RIIR C LPCI injection line. A schematic of the provided in Figure 2.18 ADIIR system is shown in Figure 2.19.
The IAS supplies clean, dry, oil-free air to EVS air valves, Control for the ADilR system is remote manual from the the CRD control system, condensate system valves' control room. Flow and temperature indications are containment venting air valves, main steam isolation provided in the control room for ADilR heat exchangers valves, RW CU AOVs, FW AOVs, PSW AOVs, and the while m. dis.idual manual control of pump operation with SRV valves (a nitrogen system backs up the IAS supply to pump running status lights is provided.
these valves).
The success criterion for the ADlIR system is to provide The success criterion for the IAS is that either of the IAS " "* * '* **
"E compressors or one of the SAS compressors must supply air to system pneumatic loads. The ADHR major dependencies are AC Division I and 2 p we f r md r paa a 6A and M6B Failure of the IAS does not directly fail any related safety resp ctively, and BOP AC bus 141IE for the ADlIR pumps systems' and motor operated valve F424. Plant Service Water
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Cooling requirements of the IAS and SAS air compressors and aftercoolers are normally supplied by the Turbine 2.2.18 Reactor Water Cleanup (RWCU)
Building Cooling Water (TBCW) system. In the event of offsite power failure, the SSW system cools the air System compressors and anercoolers.
The function of the RWCU system is to provide continuous The Unit i IAS air compressor is powered from emergency purification of reactor water to reduce the fouling of heat AC Division 2. The Unit 2 IAS and the SAS air transfer surfaces, minimize compressors are powered from non-safety buses. secondary sources of radiation, and maintain water clarity Following a loss of offsite power, standby onsite power is during refueling. The system also acts as a decay heat provided to the Unit 1 IAS air compressors to replenish removal system and reduces stratification in the reactor compressed air storage as required. vessel by providing a discharge path for excess water to eithei the condenser hotwell or to the radwaste system.
NUREG/CR-6143 2-26 Vol. 3 4 l
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Plant Description Water is drawn into the RWCU system from both from 25% to full flow. These valves are mechanically recirculatico lines and the vessel drain. The flow is then restrained to prevent operation beyond these limits.
pumped to t reties of three regenerative heat exchangers.
From the rgenerative heat exchangers, the w ater is routed The RRS major dependencies are the Component Cooling to the non regenerative heat exchangers and the filter- Water (CCW) system for pump cooling and AC power for demineralizers and finally to the reject lines of either the operating the recirculation pumps and velves.
radwaste or the main condenser. See Figure 2.20 for a Dow diagram. Major system components are shown in their The recirculation pumps are driven by squinel cage standby configurations. induction motors. At full speed the pumps are powered
. from 4160V AC BOP buses 1IHD fbr loop A and 12 HE Operation of the RWCU system is controlled from the main for loop B. At 25% rated speed the pumps are powered control room. T he outboard isolation valve will close from separate 15Hz motor-generator sets. The motor-automatically to prevent damage to the filter demmerahzers generator sets are in turn powered from 4160V AC BOP resins when the outlet temperature of the non-regenerative buses 13 AD for loop A and 14 AE for loop B. ,
heat exchangers is too high.
The motor-operated valves in the RRS are powered from The success criterion for the RWCU in POS 5 is (1) the 480V AC BOP MCC 11851. The flow control valves are discharge of reactor water to either the main condenser or hydraulic-operated valves and are mechanically limited radwaste system in order to maintain proper level in th from going more than 25% closed. The hydraulic units for reactor vessel, and (2) to remove decay heat from the vessel these valves are powered from 480V AC BOP MCC during hydro testing.
1 IB51. The Dow control valves fail as is on loss of power to the hydraulic power units. Therefore, loss of power at The RWCU system's major dependencies are AC Division BOP MCC 11B51 would not necessanly fail the RRS if the 1 and 2 power for the RWCU pump suction and system is already operating and properly aligned.
containment isolation MOVs. BOP AC power is supplied to the RWCU pumps, heat exchanger and filter demineralizes bypass valves, instmment air for the system's 2,2.20 Component Cool.ing Water (CCW,)
blowdown valves to the main condenser and radwaste, and System CCW for cooling the non-regenerative heat exchangers and pumps The function of the CCW system is to provide a closed cooling loop between certain plant auxiliary components 2.2.19 Reactor Recirculation System (IIRS) and the PSW system or the SSW system. The CCW system consists of: (1) three pumps each with a 50% capacity; (2)
The function of the Reactor Recirculation System (RRS) is three heat exchangers each with a 50% capacity;(3) a 550 to provide mixing of water in the downcomer region and gal capacity surge tank; and (4) a 50 gal. capacity forced circulation through the reactor core. The RRS chemical additwn tank. Each pump is of a single-stage, prevents stagnation and stratification of the core region horimntal, centrifugal, double-suction design powered by a causing possible transition boiling or film boiling and high 480V AC,100 hp electric motor. Each pump has a 1987 temperatures for the claddmg of the fuel nx!s The RRS gpm D w rate. Each heat exchanger is cooled by the plant insures nucleate boiling for optimum heat transfer, servic water system and is of a straight tube, single pass, improved efliciency, and lowering of fuel rod cladding counterflow design with a 1987 gpm capacity. Figure 2.22 temperatures. In the event of a RRS failure, the reactor shows the CCW system diagram.
w ater level can be raised to initiate natural circulation through the core region. During nonnal peration, two of the three pumps are operating with the third in standby, with both isolation The RRS consists of two loops external to the reactor vessel velves open (the check valve prevents backDow). The (RRS Pump A is unavailable in POS 5). Each loop standby pump auto starts on a low pressure signal of 100 psig. Two of the three heat exchangers are operating with contains an electric motor-driven centrifugal pump, two the third isolated by two manual valves from the rest of the electric motor-driven gate valves (for isolating the pump),
and a hydraulic flow contml valve. A simplified schematic system. The surge tank is connected to the system, but the of the RRS system is shown in Figure 2.21. chemical addition tank is generally not connected until needed for addition ofchemicals.
The Flow Control Valves (FCV), one in each loop, are hydraulically operated and are located immediately afler The Component Cooling Water (CCW) system supplies the discharge of the pump and before the discharge gate cmling to the following loads:
valve. The Dow control valve has a range of operation Vol. 3 2 29 NUREG/CR-6143 l
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Plant Description
. Recirculation pump seals and motor bearings, These eight pumps all feed the common PSW header to supply the loads.
. RWCU pumps and non-regenerative heat exchangers, The PSW system uses this untreated water to cool various heat loads and provide makeup to various treated water
. CRD pump coolers, systems. The PSW supplies cooling water to the CCW heat exchangers, the Steam Jet Air Ejection (SJAE)
- Fuel pool heat exchangers, intercondensers, the ADIIR heat exchangers and A/C, and the Turbine Building Coohng Water (TBCW) heat
. Post accident sample cooler, and exchangers. The PSW provides makeup for the following treated water systems: (1) the Circulating Water system;
- Drywell equipment drain sump cooler. (2) the SSW system; and (3) the Firewater System and the Makeup Water system.
The CCW success criteria requires that two of the three CCW pumps and two of the three CCW heat exchangers be The PSW system's major dependencies are 4160 V,480 V, in operation. During LOSP, one pump operation can and 120 V AC power and the Instrument Air System (IAS).
successfully cool essential loads if the non-essential loads (fuel pool heat exchangers and Reactor Water Clean Up 2.3 Refererices non-regenerative heat exchangers) properly isolate.
2.1 System Energy Resources, Inc., Grand Gulf Undated The CCW system's major dependencies are: (1) the PSW Final Safety Analysis Report.1992.
system; (2) the SSW system;(3) 480V AC power; (4) 120V AC power;(5) 125V DC power; and (6) Instrument D. Whitehead, et. al., Evaluation of Potential Severe 2.2 Air. The three heat exchangers all use PSW system water Accidents D_gine Low Power and Shutdown for cooling, with the SSW system as a backup source of Oncrations at Grand Gulf. Unit 1. Analysis of Core cooling water in the event of LOSP or loss of PSW. The Damace Frecuency from Internal Events for Plant SSW system as also a backup cooling water source for the Operation State 5 Durine a Refitelinn Outacc. Main Fuel Pool heat exchangers during a LOSP when the Fuel iienon. NUREG/CR-6143, Vol. 2, Part 1, SAND 93-Pod heat exchangers are isolated from the CCW system 2440, Vol. 2, Part 1, Sandia National Laboratories, Cooling from both the PSW and SSW systems is inhibited Albuquerque, NM,1994.
if a Loss Of Coolant Accident (LOCA) signalis present. A LOCA signal can consist of either a high drywell pressure or a low reactor vessel level signal. CCW pumps A, B, and C are powered from buses 11BDS,16BB3, and 12BE2, respectively. The 120V AC power is used to control the surge tank level, RWCU isolation valve control, all alarms, test circuits, and motor-operated valve heaters. The 125V DC power is used as control power for the CCW pumps, the power to auxiliary relays for alarms, and the fuel pool heat exchanger A valves. The instrument air operates the air operated valves in the system.
2.2.21 Plant Service Water (PSW) System The function of the PSW System is to provide cooling to various plant heat exchangers including the CCW heat exchangers.
The PSW system is an open-loop system which uses the Mississippi river bank radial wells as its source and the discharge basin and circulating water pit as its two sinks.
See Figure 2.23 (1 of 2 and 2 of 2) for the PSW diagram.
The PSW system draws water from four radial wells on the bank of the Mississippi river. Each well provides water to the suction of two pumps in parallel. The pumps are 4 stage,5000 gpm centrifugally driven by a 500 hp motor.
Vol. 3 2-33 NUREG/CR-6143
Plant Description N (I F002A-N N
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N'IREG/CR-6143 2-34 Vol. 3 i
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Vol. 3 2-35 NUREG/CR 6143
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1
- 3. Grand Gulf Fire Analysis Introduction inf rmation are needed: (1) the number of fire incidents 3.1 that have occurred in specific compartments at shutdown during commercial operation, and (2) the number of The objective of this analysis is to estimate the contribution
. compartment years that the nuclear mdustry has of fire-induced events to core damage frequency during accumulated. Most of the data for the first part comes from plant shutdown This is a report of the analysis of such risk bM Nu h performed for the Grand Gulf plant m Plant Operational Insurers (ANI), although other sources are also used, e g.,
l State (POS 5). Iqunng this state of plant operation, the the U.S. Nuclear Regulatory Commission.
plant is bemg mamtamed within the required parameters by operation of the RHR system in the SDC mode of While the NRC requires the reporting of fires that, in some j operation. For a complete description of how the POSs g .gp gg were defined, see Section 3.2 the Grand GulfInternal stringent requirements in the sense that all fire events 1: vents Analysis (NUREG/CR-6143)(Ref. 3.1).
resulting in a property loss clairn must be reported.
Compartment years at shutdown are computed by adding 3.2 Fire Locations Analyzed the age of all compartments (within a certain category of compartments) of units that were in commercial operation in this section, the plant fire zones that were analyzed are by the end of December 1989. The age is defined as the hsted and discussed Table 31 provides a comprehensive time at shutdowTt between first commercial operation and listing of the fire zones and a brief physical description of the end of December 1989 (or date of decommissioning).
each. All other fire zones not contained in Table 3.1 were ,
climinated from the analysis because they did not contain Events were only included if they occurred dunng cither vital equipment or cabling for safety- related shutdown. Eight general areas are typically found in equipment. Figure 31 presents a general plant layout nuclear power plants. These are (1) the control room,(2) drawing which illustrates main buildings where the fire cable spreading room, (3) diesel generator room,(4) zones in Table 3.1 are located. reactor building,(5) turbine building,(6) auxiliary building, (7) electrical switchgear room, and (8) battery Identification of the fire zones ofinterest, and room. In most plants, the first three areas, and the cleeirical charactenzation of the boundaries or baniers between switchgear room and battery room, are single compartments zones was accomplished by a comprehensive analysis of while the other three are typically large buildings.
plant layout drawings and the Grand Gulf Fire IIazards Analysis (Ref. 3.2) supplemented by plant walkdowns. The The fire events and operating years for the eight plant areas walkdowns were conducted to verify the selections made were obtained by using an update to the fire data base and to obtain clarifications where adequate information was developed by Wheelis (Ref. 3.6).
not available from documentation. Typically, fire zones are bounded by 3-hr rated fire barriers, fire doors, fire-resistant To determine operating years for electrical switchgear walls, and fire dampers The subdivision of the plant into rooms and battery rooms, auxiliary building operating years fire zones is also accomplished by the division of the were doubled A survey of all U.S. light water reactors building into ditTerent levels (elevations). In the cases indicated that there is an average of 2.25 trains of where open gratings separated levels, the rooms at the emergency switchgear and their associated batteries per difTerent elevations were considered part of a single fire plant. Ilowever, it is known that some plants such as Surry zone. locate both trains of their emergency switchgear in one fire area. So it was assumed that an average number would be 3.3 Initiating Event Frequencies two per plant fbr both types of rooms.
To obtain fire area-specific initiating frequencies, a 3.3J Fire Data partitioning method is required. Partitioning is a process in which the analyst subdivides the frequency of fire Data on fires in commercial Light Water Reactors have ccurrence fr m a large building (e.g., auxiliary building) been analyzed in several studies (Refs. 3.3,3 4,3.5).
am ng specWic rmms or nm zones whin that building.
Although these studies have been done independently, they Als , fmther partitioning can occur witlu,n a specific room have some common characteristics. For example, almost r ama. One method of partuomng is accomplished by all studies have used Licensee Event Report (LER) data ratioing the areas of fire areas withm a building. The from the Nuclear Regulatory Commission (NRC). All have , ,
assumptim here is that the frequency of fire occurrence is reported the overall frequency of fires of approximately dependent only upon the amount of area a fire area 0.16 incidents per reactor year on a plant wide basis. To c ntams. Another method of partitioning examines determine fire initiating event frequencies, two kinds of Vol 3 3-1 NUREG/CR-6143 i
. = . .
Fire Analysis Table 3.1 Grand Gulf Fire Zone Descriptions Fire 7mne Building Elevation Physical Description 1A101 Auxilian 93' & 103' Passage lA102 Auxiliary 93' RIIR A IIcal Exchanger Room 1A103 Auxiliary 93' RIIR A Pump Room 1A105 Auxiliary 93' RIIR B Pump Room 1A106 Auxiliary 93' R1IR B 1Icat Exchanger Room lA109 Auxiliaq 93' ilPCS Pump Room lAl10Cl Containment 135'4" Electrical Containment Penetration Area i IAI10C3 Containment 135'4" Electrical Containment Penetration Area !
1 A l 10D3 Containment 161'10" Containment Cooler lAl12 Containment 100 9' Drywell Area lAl14 Auxiliary 93' & 103' Fan Coil Area 1A117 Auxiliary 93' & 103' Miscellaneous Equipment Area IA120 Auxiliary 93' CCW Pump and iIcat Exchanger Area 1Al28 Auxiliary 93' R1IR A IIcat Exchanger Room 1A129 Auxiliary 106' RI1R H IIcal Exchanger Room 1A201 Auxiliary 119' Passage 1A202 Auxiliary 119' RIIR A IIcat Exchanger Room 1A203 Auxiliary 119' Ri!R A Piping Penetration 1A204 Auxiliary 119' Access to RCIC Room 1A205 Auxiliary 119' RIIR B Piping Penetration 1A207 Auxiliary 119' Electrical Switchgear Room Div II 1A208 Auxiliary 119' Electrical Switchgear Room Div I IA210 Auxiliary 1 I 5' RWCU Recirculation Pump B Room 1A211 Auxiliary 119' Misc. Equipment Area NUREG/CR-6143 3-2 Vol. 3
Fire Analysis Table 3.1 Grand Gulf Fire Zone Descriptions (Continued)
Fire Zone Building Elevation Physical Description 1A215 Auxiliary 119' Fan Coil Area 1A222 Auxiliary 119' Motor Control Center Area IA301 Auxiliary 139' Corridor IA302 Auxiliary 139' Corridor iA305 Auxiliary 140' Main Steam Tunnel 1A307 Auxiliary 139' RIIR B IIcat Exchanger Room l
l 1A308 Auxiliary 139' Electrical Penetration Room Div.11 l 1A309 Auxiliary 139' Electrical Penetration Room Div. I 1A313 Containment 135'4* CRD1Iydraulic Control Area lA314 Auxiliary 139' Passage IA316 Auxiliary 139' North Passage Area 1A318 Auxiliary 139' Electrical Penetration Room 1A322 Auxiliary 139' Centrifugal Chiller Area IA401 Auxiliary 166' Passage IA403 Auxiliary 166' Passage 1A407 Auxiliary 166' Motor Control Center IA410 Auxiliary 166* Motor Control Center IA414 Containment 170' RWCU IIcat Exchanger Room 1A417 Auxiliary 166' Misc. Equipment Area 1A420 Auxiliary 166' Misc. Equipment Area lA428 Auxiliary 166' Passage 1A539 Auxiliary 185' Cable Space Divi & 11 1Mi10 SSW Pump Ilouse 133' SSW Pump llouse Div 1 2Mi10 SW Pump Ilouse 133' SSW Pump llouse Div 11 ID310 Diesel Generator 133' Div I Diesel Generator Vol. 3 3-3 NUIGG/CR-6143
Fire Analysis Table 3.1 Grand Gulf Fire Zone Descriptions (Concluded)
Fire Zone Building Elevation Physical Description OC116 Control 93' Ilot Machine Shop OC202 Control 11l' Div i Switchgear Area OC207 Control 11l' Div I Battery Room OC208 Control 11l' Emergency Ilot Shutdown Room Div II OC208A Control 11l' Emergency Hot Shutdown Room Div I OC209 Control 11l' Div III Battery Room )
OC210 Control i1l' Div III Switchgear Room I OC211 Control 11l' Div II Battery Room OC215 Control 148' Lower Cable Spreading Room OC402 Control 148' Corridor OC403 Control 148' Computer & Control Panel OC407 Control 148' Instrument Motor Generator OC409 Control 148' Electrical Space OC503 Control 166' Control Room Area OC504 Control 174'6" Suspended Ceiling above Instr Rack OC601 Control 177' Viewing Gallery OC702 Control 189' Upper Cable Spreading Room OC703 Control 190' Control Cabinet Area Div I OM101 Fire Water Pumphouse 133' Diesel Driven Fire Pump Room A OM103 Fire Water Pumphouse 133' Diesel Driven Fire Pump Room B ITI18 Turbine Building 93' TBCW Pump Area IT132 Turbine Building 93' East Corridor IT214 Turbine Building 113' Motor Control Center Area IT325 Turbine Building 133' FiltersInstr. Rack Area IT327 Turbine Building 133' Steam Instr. Rack Area IT404 Turbine Building 166' Ilatches Area NUREG/CR-6143 3-4 Vol. 3
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- 1. Computer Aided
- 2. Credit for Automatic FPS Coverage
- 3. Recovery of Random Failures
- 4. Plant Inspection determined initiator could not be induced by fire
- 5. Detailed fire propagation modeling Vol. 3 3-7 NUREG/CR-6143
Fire Analysis T,o: Less of all Component Cooling Water Tm: less of Makeup (CRD) The loss of the CCW pumps would prevent the use of either This event involves the loss of CRD which is a two pump recirculation pump, CRD pumps, FPCCU and RWCU. The system. One CRD pump is sufEcient for level control, but success criteria for the CCW pumps is that 2 of 3 pumps two CRD pumps are required to provide adequate makeup and 2 of 3 heat exchangers be in operation. for steaming the core. To: less of IE 4160 V AC Bus B T,,3: Iess of Recirculation Pump This event involves the loss of IE 4160 V AC Bus B and its This event involves the loss of Recirculation pump B. associated loads. For this event, we screened out loss of non-lE buses as initiating events. E: n Loss of Common RIIR Suction Line Ta: less of IE 125 V DC Bus B This event involves the loss of either the inboard and/or l outboard MOV(s) in the common RIIR suction line. l The loss of a DC bus generally falls into one of two I categories. They are: 1) failure to provide DC power on E ,: 3 Less of Operating RIIR-Shutdown Cooling demand as characterized by the loss of charger output, and System
- 2) operational, test, or maintenance errors resulting in the loss of DC power during normal plant operation. The first This event involves major interruptions in the operating category can be fire-induced while the second category is SDC system These events essentially cause a long term random. loss of SDC frorn the previously operating train (s).
The principal cause of failure for the first category involves 3.5 Detailed Description of the operation of the DC power system with one or more . batteries unable to provide sufEcient power to the bus if Screen.ing Analysis battery charger output is lost. A comprehensive screening analysis is required to reduce The loss of non-1E buses was screened out as initiating the mimber of potential fire-induced scenarios to only those events. which have the potential to be probabilistically significant contributors to core damage frequency. Tar less ofInstrument Air The screening analysis was composed of the following four This event involves loss of the service and instmment air steps: compressors or their associated power supplies and cabling. Step 1. Identification of Relevant Fire Areas Ty: Inadvertent Overpressurization (Makeup Greater Than letdown) Fire areas containing equipment or cables associated with safety-related systems which mitigate the effects of the For POS 5 the Tm, initiator is essentially loss of Reactor unscreened fire-induced "off-normal" plant states were Water Cleanup system (RWCU) (letdown) and failure of identified. All other fire areas were climinated from further the operator to stop CRD or control pressure. analysis. This resulted in the identification of relevant fire areas which are listed in Table 3.1. Tng Inadvertent Pressurization sia Spurious IIPCS Actuation Step 2. Further Screen Fire Areas Based on Vital Area Analysis This event involves the spurious actuation of the High Pressure Core Spray system and subsequent inadvertent The remaining fire areas were subjected to a vital area pressurization. analysis (location mapping) of components and cables (both control and power cables) that were kicated within these Tm,: Inads ertent Overfill via LPCS or LPCI areas. This information was integrated into the PRA with the IRRAS computer code (Ref. 3.8), which was then used This event involves the inadvertent overfill of the vessel via to solve all front line systems and solve all of the identified the Iew Pressure Core Spray or the Iew Pressure Core sequences injection system. NUREG/CR-6143 3-8 Vol. 3 Fire Analysis Fire occunence frequency for each area was set to 1.0 and, Step 4. Confirmatory Plant Visit given a fire, all components within that area were assumed to fail. The output of this analysis is a series of accident For the fire areas surviving the screening process to this sequences expressed in tenns of cut sets involving single or point, all scenarios are then associated with equipment multiple fire areas as well as random failures (i c., not fire fire-related failures that were identified during the plant Iclated). walkdown. A scenario can be thought of as a combination of one or more fire-related equipment failures within a fire Cut sets u hich required three or more fire zones were area with or without additional non-fire-related (random) eliminated This was deemed appropriate since these cut failures outside of the fire area. These failure combinations sets imply the failure of two or more fire barriers or large must minimally lead to core damage. Each fire area can fires of sufficient duration that there is a high likelihood of have one or more scenarios depending on the equipment suppression Cut sets which contained two fire zenes were combinations which might fail due to the fire in that screened on the following two criteria: (1) no adjacency particular area. Additional plant visits were conducted to between areas and (2) no penetrations in the adjacency determine which of the postulated scenarios were valid between areas An additional screening step to be based on cable or equipment locations within a particular perfbrmed on the cut sets u hich contained two fire zones is fire area Past experience with fire code calculations, numerical culling with barrier penetration failure set to a discussed in the following section, and fire testing results screening value of 01. It is known from the analysis of provide the basis for assessing the validity of scenarios. many fire barriers that typical random failure rates are on For example, if a scenario required the fire-related failure the order of 10 2 to 104 Therefbre, this screening value has f cabling for components A and B and it could be shown been set high enough to ensure that potentially important that these cables were always separated by greater than 40 fire area combinations are not lost. 11 in an area of suflicient size to preclude buildup of a hot Truncation of cut sets at a random faihire probability of 10" gas layer, or at least one of the component's cabling was in was perforrned. When all fire-related factors are taken into a 3-hr rated fire wrap, then that scenario was eliminated account, the frequency of any scenario is at least three from further analysis. orders of magnitude less than this random failure truncation probability. 3.6 Fire Propagation Modeling using important infbrmation gained from the analysis of these cut COMPHRN Hle sets was identification of the remaining plant kwations w here area-to-area bamers needed to be analyzed. 3.6.1 Introduct,on i ihminant cut sets which contain aJjacent fire areas were analyzed for barrier failure in the quantification process. The COMPBRN fire growth code (Ref. 3.9) was used to calculate fire propagation and equipment damage. COMPBRN was developed specifically for use in nuclear Step 3. Cull Fire Areas on Frequency and Non-power plant fire PRAs. The code calculates the time ,yn y required to damage critical equipment given that a fire has Cut sets not climinated in the first two sercening steps are started. This failure time is then used in conjunction with requantified with fiie area specific initiating event plant specific infonnation on fire suppression to obtain the frequencies that were calculated utilizing the methodology probability that a given fire will cause equipment failure described in Section 3.3. which leads to core damage before the fire can be suppressed. l Additionally, operator recove:Y of non-fire-related random ~ failures was analyzed. For screening purposes only, all COMPBRN employs a quasistatic approach to simulate the short term (less than 24 hr) recovery actions (of non-fire pr cess of fire during the preflashover penod in an faihires) were increased from their respective internal enclosed space. COMPBRN uses an area model which breaks the fire environment into three sub-areas: events probabilities by a factor of five to allow for the additional confusion of the fire situation occurring in flame / plume, hot gas layer, and ambient (see Figure 3.2). conjunction with other random failures. Where recovery Simple fire and heat transfer models and correlations are actions were long term (greater than 24 hr), no employed to predict the thermal envirorunent as a function modifications to intemal event probabilities were deemed f time. The thermal response of various targets in the fire necessary. It was assumed that aner 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the fire scenario is modeled to predict the amount of time required would be extinguished and any spurious signals will have f r a fire to damage or ignite critical equipment. The terminated in open circuits critical equipment is generally taken to be a cable tray canying cables necessary for safe shutdown of the plant, Vol. 3 3-9 NUREG/CR-6143 i 1 Z m h COMPBRN MODELING g a 3-Q s-e U I CABLE o o e e e TRAYS e e o e e e e o e e e e e e o e HOT GAS m og LAYER E I n Y 'J e o e e e r O w ? L o 5 \ /
- c z \ /
U 2 \ /
- PLUME
\ / 8, \ / CONVECTION SL g j AND RADIATION \ / MODELED \ / FIRE Fire Modeled as a Cylinder
- Cable Trays Discretized into Fuel Elements
< - Hot Gas Layer Effects 9-l u _ - _. = . - - . . - - - . - _ . . Fire Analysis although other critical components, such as pumps, may be The COMPBRN results are shown in Table 3.4 for the modeled. cntical areas. A number of scenarios were considered for many of these areas. In most cases, a " zone ofinfluence" 3.6.2 Grand Gulf Fire Propagation Modcling was determined for the equipment and fire sizes modeled. In other words, the fire location was varied in the Walkdowns of the Grand Gulf nuclear plant were COMPBRN models to determine the maximum distance the conducted to obtain the cale input information required to fire could be away from the critical equipment and still perform the COMPHRN IIIe fire damage assessment cause damage. The times to damage increase exponentially calculations. This information includes the kication of as the fire distance increases. The numbers given in Table critical equipment and cable trays, separation between 3.4 represent the combination of greatest distance and redundant trains, types of cable present, and any shielding longest times to damage. Using these results, the floor area or fire barrier material that may be present. in which a fire would have to occur to damage critical cables can be estimated. An area ratio can then be Both small and large fires were postulated in the calculated by dividing this area by the total floor area of the calculations for Grand Gulf. A small fire was assumed to room, fire area, or building (as appropriate). This be 2 ft (0.61m) in diameter and consist of I gal (3.8 I) of reduction factor can then be multiplied by the initiating frequency to estimate the frequency of fires which occur in oil. A larFe fire was assumed to be 3 ft (0 91 m) in diameter and consist of 10 gal (381) of oil. Analysis of a a critical portion of a given room. data base of transient combustible fuel sources found at It should be noted that a small fire, except for zone of nuclear power plants (Ref. 3.9) indicates that oil sources mauence cases, does not yield damage in many of the less than or equal to 1 gal (3.81) were fbund approximately amas. Pn,or expenence with COMPBRN shows that a 70 percent of the time. Oil sources larger than this were smaH fim must be very close to its target to yield damage. found approximately 30 percent of the time. A similar 1.arge fires, however, can and do yield damage m most of partitioning between small and large quantities in terms of heat content (BTU or KJ) can be made for other credible the amas. In Jarge open rooms such as the auxiliary buildmg corndor, the larger fire must still be within about 3 transient combustible sources such as solvents or trash paper. Again, analysis indicates that a 70/30 partitioning it h rizontally of the target cable trays (assuming typical tray heights). The major exception is in small closed rooms between small and large fuel sources is appropriate (within m which a hot gas layer rapidly develops. In such cases, + 10 percent). It can also be shown that 10 gal (381) of oil ihunds any large solvent or trash paper combustible source the hot gas layer effects become quite significant. in tenns of heat content and is, therefbre, an appropriate It has been found in past experience with COMPBRN and upper bound on transient combustible fuel source size. in some of the simulations for Grand Gulf that the COMPBRN results can be quite sensitive to fires located Cable insulation ignition and damage thresholds are . adjacent to walls which are in close proximity to the target currently not well known. For this study, a cable insulatmn cable trays. Using the typical model of the wall as one ignition temperature of 773 K (932 F) was assumed along section results in unrealistic radiative heat fluxes from the with a damage temperature of 623 K (662'F). For the larg I wall to the cable trays ofinterest. For these cases, the wall fire simulations, these thresholds are not as critical to the was divided into several sections to more realistically fire damage time calculations because of the intensity of th calculate the wall thennal response. flames. The uncertainty in the flammability parameters along with the uncertainty in other parameters wa3 The following subsections discuss the approach and results considered by the fire experts in interpreting the for the scenarios analyzed in each of the critical fire areas. COMPBRN results and in constmeting the probability Scenarios can be thought of as unique equipment distributions on time to damage. combinations which must sustain fire damage. All entries in Table 3.4 which are blank indicate no predicted fire A list ofinput parameters for the COMPBRN calculations damage. is shown in Table 3.3. These parameters were selected based on past fire analyses at commercial nuclear facilities 3,6.3 Fire Zone 1 A117, North IIallway to represent typical qualified cable msulation. It was assumed, based on cable tray inspection in some critical fire 9F/10T Elevation of the Auxiliary areas that the cabling in all the areas ofinterest at Grand Build,mg 4 Gulfincluded typical brands of nuclear qualified cable. f Because of the gomi flame resistance properties of these 3.6.3.1 Dheuulon cables, no self-ignited (electrically initiated) cable tray fires wme postulated. The scenario in this area calls for damage to two cable trays and involves the loss of CRD pump A and a R1IR valve. Vol. 3 3-1i NUREG/CR-6143 l Fire Analysis Table 3.3 COMPIIRN IIIe Input Parameters Cable Insulation Parameters Density 1715 kg/m' Specific 1Icat 1045 J/kg-K Thermal Conductivity 0.092 W/m-K lleat of Combustion 1.85-2.31E-7 J/kg Combustion Efficiency 0.6-0.8 Critical Temperature Piloted Ignition 773K Spontaneous Ignition 773K Damage 623K Surface Controlled Burning Rate 0.0001-0.0075 kg/m' S 2 Burning Rate Radiation Augmentation 1.86E-7 kg/J-m Radiative Fraction 0.3-0.5 Smoke Attenuation Factor 1.4 Reflectivity 0.1-0.3 Oil Parameters Density 906 ig/m' Specific IIcat 2 l00 J/kg-K lleat of Combustion 4.67E7 J/kg Combustion Elliciency 0.9 Surface Controlled Burning Rate 0.06 Radiative Fraction 0.3-0.5 Mass of Oil 3.4-34.0 kg Table 3.4 Summary of COMPIIRN Results I.arce Fires Small Fires Fire Zone Location Time Area Ratio Time Area Ratio 1Ai17 93'/103' Elevation 15 min 4OE-2 ---- ---- 1A201 119' Elevation 5 min. 1.7E-2 ----- --- 1 A211 Ge :n 1) 119'lilevation 2 rnin 3.0E-2 3 min. 3.0E-3 1 A211 (Scen 2) 119' Elevation 2 min. 3.0E-2 3 min. 3.0E-3 1 A316 (Scen 1) 139' Elevation 2 min 4.0E-2 3 min. 7.0E 3
- 1 A316 (Scen 2) 139' Elevation 2 min. 4.0E-2 3 min. 7.0E-3 I
NUREG/CR-6143 3-12 Vol. 3 Fire Analysis The closest point of approach of both cable trays to each Scenario 2 - Cable trays along the north wall other is 3 6 m or 11.8 A. This hallway area is part of a larger halhvay area that surrounds containment on this The target cable trays for this scenario include ora from the elevation (see Figure 3.3). Since this area is so large, no previous scenario and an additional cable tray which is 311 hot gas layer was modeled wide and also is situated approximately 211 from the wall. These cable trays an Iso separated by approximately 2.5 3.6.3.2 Results A. Since these cable trays are located in a large open hallway, no hot gas layer was modeled. No small fire was found to damge to both cable trays simultaneously. Ilowever, a large ti,e caused damage to 3.6.5.2 Results both trays in 15 minutes. i For both scenarios, a small fire was found to cause damage l 3.6.4 Fire Zone I A201, East liallway,119' to both cable trays in approximately 3 minutes. A large fire caused damage to both trays in approximately 2 minutes. Elevation of the Auxiliary Buiiding 3.6.4.1 Discussion 3.6.6 Fire Zone I A316, North IIallway Area, 139' Elevation, Auxiliary Building The fire scenario in this area calls for the damage to two cable trays containing cabling for a RIIR valve and 3.6.6.1 Discussion recirculation pump B. The cable tray containing the RHR valve is approximately 9.4 fl o!T the floor with another There are two scenarios which could lead to core damage. cable tray situated 3 A below it. The cable tray with recirc Both scenarios involve cable trays in the same general area. , pump B cabling runs perpendicular to and above the cable tray with the RI1R valve. The closest point of approach for Scenario 1 - Cable trays along the north wall these cable trays is approximately 4 A. This area is part of The target cable trays for this scenario run from the floor to a larger hallway area v hich surrounds the containment on the ceiling along the north wall in this hallway area, where this elevation (see Firare 3 4). T) crefore, a hot gas layer was not modeled. partial spn,nkler coverage protection is provided. This first scenario for this area involves two cab!c trays separated by 2.5ft Both of these trays are i 11 wide and are situated 3.6.4.2 Results approximately 2 A from the wall. Since these trays are 1 ated in a large open hallway (see Figure 3.6), no hot gas For this scenario a small fire wa.s found not to damage to layer was modeled. both cable trays. A large fire caused damage to both trays in 5 minutes. Scenario 2 Cable trays along the north wall 3.6.5 Fire Zone 1 A211, North IIallway 119 The target cable trays for this scenario include one from the Elevation, Auxiliary Building previous scenario and an additional cable tray which is 3 ft wide and also is situated approximately 2 ft from the wall. 3.6.5.1 Discussion These cable trays' arc also separated by approximately 2.5 ft Since these cable trays are located in a large open , There are two scenarios which could lead to core damage. hallway, no hot gas layer was modeled. Both scenarios involve cable trays in the same general area. l
- 3.6.6.2 Results Scenario 1 - Cable trays along the nonh wall l For both scenarios, a.m0 fire was found to cause damage The target cable trays for this scenario run from the floor to to both cable trays in appimnely 3 minutes. A large fire the ceiling along the north wall in this hallway area, where caused damage to both tr n s m approximately 2 minutes.
partial sprinkler coverage protection is provided. This first scenario for this area involves two cable trays separated by . 2.5 it Both of these trays are 1 A wide and are situated 3.7 Barrier Failure Analysis approximately 2 fl from the wall. Since these trays are Iccated in a large open hallway (see Figure 3.5), no hot gas In the unscreened cut sets where a potential for banier L layer was modeled. failure was identified, barrier failure probability was estimated using barrier failure rates developed as part of the i Vol. 3 3-13 NUREG/CR-6143 TURBINE BLDG. DR-IA101 (CORRIDOR ITl32) Fire Analysis gl g STAIR NO. IA01 mumm 5 EA{N _=i:: !I i EEDL
- -5!!
5
- Eif HPCS IP AREA .
DR-IAl004 IAl23 (GRATING) 9 ]( STAIR NO. m IA08 E
- 5 ELEV. g :ijg NO. 3 d :Ei
- =0 DR-IA106 STAIR NO. 5 IAl0 Y5h MISC,
"" $ EE :, EQUIP. E _i AREA 9 .i!E lAll7 (FLOOR) DR-!Al09 2 lll
- n
((
- E STAIR NO.
IAl3 = Ek I Figure 3.3 Fire Zone IA117 Layout NUREG/CR-6143 3-14 Vol. 3 + %b i T. DR-IA201 g EXHAUST g BLOWERS , 6' W ; DR-IA203 EAST PASSAGE lA201 T- , DR-1A202 DR-IA204 FLOOR EL 119'-0*
- [D .
i RHR HEAT RHR"B" ACCESS RHR"A" RHR HEAT S EXCH. *B" PIPING TO PlPING EXCH. "A" { IA206 ROOM RCIC ROOM IA202 y lA205 ROOM IA203 E lA204 1 Z c W T. a-8 N, i a j. Fire Analysis um t' f DIVISION 11 SWITCHGEAR ROOM lA207 k DR-IA207 1 SRM/lRM DRIVE ! CONTROLCABINET \ ' g MISC. E EQUIP. ' AREA FREIGHT ELEVATOR X ,E NORTH PASSAGE IA211 E] DR IA211 =E STAIR NO. g5 E) lAl0 Q S -m u .y. Figure 3.5 Fire Zone I A211 Layout NUREG/CR-6143 3-16 Vol. 3 EMI I Fire Analys.is CORRIDOR IA301 = ~ = Hl..ta.-.. ~=~ RHR ' HEAT EXCH. *B' IA307 PIPE CHASE - - 3 REMOVABLE CONC. COVER EL 142'-6" 'E DRIA307 - DIV. Il @ @ SWITCHGEAR ROOM CABLE CHASE lA308 f E E MOTOR CONTRO E CENTER AREA E NORTH E PASSAGE FREIGHT ELEVATOR EM E IA316 g] HS-16C/Hi U me w2 < ! DR-IA309 STAIR NO. l DR IA308 M' ~T -" I ,_ / j .DR-IA311 /~ ~ STM NO. 52 q /q -. g DR-1 A310 ^=l STAIR NO. E LOADING ~ IA!9 DOCK EL.137-4* RAILROAD STAIR NO. ACCESS E lA20 IA325 5 EL.133'-O' PLATFORM EL.150' O' ::- I I I Figum 3.6 Fire Zone 1 A316 Layout Vol. 3 3-17 NUREG/CR-6143 l I 1 Fire Analysis Risk Methods Integration and Evaluation Program per reactor year after recovery of non fire related failures, (RMIEP) analysis (Ref. 3.10). Barriers were grouped into and barrier failure probabilities were applied. Core damage three types: (1) fire doors, security doors, watertight frequencies for the following scenarios can be found in doors, and fire curtains,(2) fire dampers and ventilation Table 3.7. dampers, and (3) penetration seals and fire walls. The data base contains 628 records from start of construction on any 3.10.2 Fire Zone 1A117 give plant to the end of June 1985. The number of barriers of each type at a plant is required to estimate the rate at Frre Zone 1 Al 17 is a hallway area located on the 93'/103' which a specific component fails. The number is not elevation of the auxiliary building Fire-related failures in known precisely for each plant, but a nominal figure that this zone were found to lead to initiating event TLM51I has been estimated for each barrier type is given in Table (Loss of Makeup-CRD). Fire-related damage failed all or 3.5. parts of the following systems: The statistical uncertainty of each estimate, reficcting a. CRD sampling variation and plant-to-plant variation,is b. RIIR represented by 90 percent confidence bounds. These estimates and confidence bounds are given in Table 3.6 Additional random failures were required to lead to core where units of both estimates and bounds are failures / year. damage and are represented by the term Enian,. After allowing for credit for recovery of random failmes In the case of Fire Zone l Al 17 only a large fire was found during the screening process, all fire scenarios which to be capable of damaging the critical equipment. required fire-related failures in adjacent fire zones were Therefore, there are associated terms for area and severity climinated from further consideration- ratios for only a large fire. 3.8 Recovery Analysis The core damage equation for Fire Zone 1 Ai 17 is as follows: For those remaining cut sets which survived the screening process and where the COMPHRN code predicted fire @. = 1,. f airn, Q i (to ) E ai4n,Alf fu damage would occur, recovery of random failures and credit for extinguishment of the fire before the COMPBRN Where: predicted time to fire damage was applied. $ = fire-induced core damage frequency for Fire Zone Recovery of random failures (non-fire related) has been 1 Ai 17, treated as in the internal events analysis All operator recovery actions that were used in the internal events 1,, = auxiliary building fire frequency, analysis were assessed for use where appropriate on the remaining cut sets. fairn, = area ratio of Fire Zone 1 A117 to that of the auxiliary building, 3.9 Uncertainty Analysis Q i(ta) = probability that the fire will not be manually An uncertainty analysis was not performed since none of suppressed before the critical components are the scenarios analyzed resulted in core damage frequencies damaged, greater than 1.0E-8 per year, f 4i = area ratio within Fire Zone 1 A i 17 where a large lire
- ^" d"** 8 'h* "I'i ^' * *P """ *"d 3.10 Quantification of Unscreened Fire Induced Core Damage Scenarios fn = severity ratio ror a large fire.
and Their Associated Fire Zones . Table 3.8 gives the values of all terms m the core damage equation (except for random failures) for Fire Zone 1 Ai 17. 3.10.1 Introduct,on i This section describes the fire scenarios and their associated Fire Zones which survived the computer-aided screening pg g j ;, , g .
- 3 9, process. All other fire zones were assessed to be below 10
&vation of the auxiliary building. Fire related failures in NUREG/CR-6143 3-18 Vol. 3 Fire Analysis Table 3.5 Approslmate Number of Barriers at a Plant Type Nominal 1 150 2 200 3 3000 Table 3.6 Estimates of Single Barrier Failure Rate l Barrier Barrier / Estimate 5 Percent 95 Percent Type Unit Confidence Confidence Bound Bound i 150 7.4E-3 0.0 2.4 E-1 2 200 2.7E-3 0.0 2.2E 1 3 3000 1.2E-3 00 3.7E-2 Table 3.7 Accident Sequence Core Damage Frequency Contributors Point Estimate Core Scouence Fire 7ene Damare Freauency Urv3 TIA1511* 1 A117 Auxiliary BuildingIIallway Area,109' Elevation <l x 10 T!.M511 1 A211 Auxiliary Building i tallway Area,119' Elevation < 1 x 10~' TIA1511 1 A316 Auxiliary Building IIallway Area,139' Elevation < l x 10 TRPT5** 1 A201 Auxiliary 13uilding Ilallway Area,119' Elevation <1x10' TRPT5 1 A21 i Auxiliary 13uilding ilallway Area, i19' Elevation <1 x 10 ' TRPTS 1 A316 Auxiliary Buildine llallway Area.139' Elevation <l x 10
- Loss of Makeup (CRD) i
" Loss of Recirculation Pump i l Table 3.8 Fire Z<me 1 Al17 Core Damage Equation Terms (Point Estimate Values) i A, 1.8E-2 l f mn, 4.0E-2 Qi(t)o 0.9 fu 3.2E 3
- f. , 03 Vol. 3 3-19 NUREG/CR-6143
Fire Analysis this vone were found to lead to either initiating event 3.10.4 Fire Zone 1 A201 TRPT5 (loss of Recirculating Pump) or initiating event TLM51I(Ioss of Makeup-CRD). Fire-related damage Fire Zone 1 A201 is a hallway area located on the 119' 1 failed all or pasts of the following systems: elevation of the auxiliary building. Fire-related damage I failed all or parts of the following systems: l
- a. CRD
- b. RIIR a. RRS
- e. RRS b. RIIR Additional random failures were required to lead to core Additional random failures were required to lead to core damage and are represented by the terms E,iun, and damage and are represented by the term Eniuoi.
Ri A2l l B' ! In the case of Fire Zone 1 A201, only a large fire was found In the case of Fire Zone i A211 both a small and a large fire ) to be capable of damaging the critical equipment. i were found to be capable of damaging the critical Therefore, there are associated terms for area and seventy l equipment. Therefore, there are associated terms for area ratios for a large fire only. I and severity ratios for both a small and a large fire. The core damage equation for Fire Zone I A201 is as The core damage equation for Fire Zone I A211 is as follows: follows: ) @,,, = A,,, f uuoi Q 3 (to) Eniuoi fu f31 !
- .,.= A. . f mun Q : (to)I fu fu + faif ul [Enuni+
Eniunal Where: 1 $,,, = fire-induced core damage frequency for Fire Zone 1A201, $,,, = f induced core damage frequency for Fire Zone A,,,, = auxiliary building fire frequency, fmuoi = area ratio of Fire Zone 1 A201 to that of the A= = auxiliary building fire frequency
- auxiliary building, fuun = area ratio of Fire Zone 1 A211 to that of the Q3 (to) = probability that the fire will not be manually auxiliary building, suppressed before the critical components are probability that the fire will not be manually
- E '
Q 2(to) = suppressed befbre the critical components are fu = area ratio within Fire Zone I A201 where a large damaged, fire can damage the critical components, and fu = area ratio within Fire Zone 1 A21 I where a large f= xverity ratio for a large fire. fire can damage the entical components, fn = severity ratio for a large fire Table 3.10 gives the values of all terms in the core damage equation (except for random failures) for Fire Zone 1 A201, ik = area ratio within Fire Zone i A21I where a small fire can damage the critical componenis, and fu = severity ratio for a small fire. 3.10.5 Fire Zone I A316 Fire Zone 1 A316 is a hallway area located on the 139' elevation of the auxiliary building. Fire-related failures in Table 3.9 gives the values of all terms in the core damage this zone were found to lead to either imtiating event equation (except for random failures) for Fire Zone 1 A211. NUREG/CR-6143 3-20 Vol. 3 Fire Analysis Table 3.9 Fire Zone I A211 Core Damage Equation Terms (Point Estimate Values) 1,, 1.8E-2 f 4,cii 3.lE-2 Q3(to) 0.95 fu 4.8E-3 fs, 0.3 Table 3.10 Fire Zone I A201 Core Damage Equation Terms (Point i Estimate Values) 1,, 1.8E-2 fuui 1.7E 2 Q,(t) o 0.9 fu 3.0E-2 f 31 0.3 m. TRPTS (Loss of Recirculation Pump) or initiating event 4,, , x,,fu u,, g ,(to ))[ru f,, + r,,r,,)[3,,u, ,+ TLM511(l.oss of Makeup-CRD). Fire-related damage failed all or parts of the following systems: Emini al
- a. CRD
- b. RilR Where:
- c. RRS
@,,,, = fire-induced core damage frequency for Fire Zone Additional random failures were required to lead to core 1A316 damage and are represented by the term I,icia and I, , u i ... A , = auxiliary building fire frequency, fuui. = area ratio ofFire Zone 1 A316 to that of the in the case of Fire 7ene I A316 only a large fire was auxiliary building capable of damaging the critical equipment. Therefore, there are associated terms for area and severity ratios for a Q. (to) = probability that the fire will not be n anually large fire only. suppressed before the critical compcnents are damaged, The core damage equation for Fire Zone 1 A316 is as fu = area ratio within Fire Zone 1 A316 where a large follows: tire can damage the critical components, Vol. 3 3 21 NUREG/CR-6143 Fire Analysis f,i = severity ratio for a large fire, b. Even if active electro-mechanical safety-related equipment is damaged by fire, an initiating event fm = area ratio within Fire hine 1 A211 where a small fire may not occur. For instance, for the loss of can damage the critical components, and Turbine Building Cooling Water (TBCW) initiator to result from fire related damage, multiple fu = severity ratio for a small fire. operational pumps must fail. These pumps and their associated cables haw suflicient separation Table 3.1 i gives the values of all terms in the core damage such that it is highly unlikely that a single fire equation (except for random failures) for Fire h>ne 1 A316. could lead to failure of all pumps. As a result, many at shutdown initiating events were screened
- 3. I I Conclusions due to physical separation criteria. Even for unsercened initiators, very few fire zones were fed to be applicable due to physical separation.
The overall fire-induced core damage frequency during POS 5 for the Grand Gulf Nuclear Station wa= found te be less than 1.0E-8/ry. The fire-induced core damage
- c. Relative to other plants, Grand Gulf utilizes more frequency at Grand Gulf Nuclear Station was found to be automatic fire protection systems in critical safety-lower than other at power fire risk assessments due to the related areas which in turn reduces the probability following reasons:
ofcore damage due to a fire. a- The plant operational state analyzed represents , only three percent of the time at shutdown and fire ) frequencies at shutdown are similar to those at I power. This provides an immediate reduction in core damage frequency. Table 3.11 Fire Lme I A316 Core Damage Equation Terms (Point Estimate Values) ).,, 1.8E-2 f am. 4 0E-2 Q.(t)o 0.9 fu 1.0E-2 f, 3 0.3 fm 7.0E-3 fn 0.7 NUREG/CR-6143 3 22 Vol. 3 Fire Analysis 3.12 References 3.6 W. T Wheelis, Users' Guide for a Personal 3.1 D. Whitchead, et al Evaluation of Potential C mputpr-Based Nuclear Power Plant Fire Data Severe Accidents Durine I,ow Power and Shutdown Operations at Grand Gulf. Unit 1. ILa_se, NUREG/CR-4586, SAND 86-0300, Sandia Analysis of Core Damace Freauency from Internal National Laboratories, Albuquerque, NM, August Events for Plant Oneration State 5 During a 1986. Refueline Outare. Main Report. NUREG/CR-3.7 R. L Iman, Modeline Time to Recovery and 6143, Vol. 2, Part 1, SAND 93 2440, Vol. 2, Part Initiatine Event Freauency for Loss of Off-Site I, Sandia National Laboratories, Albuquerque, Power Incidents at Nuclear Power Plants. NM,1994. NUREG/CR-5032, S AND87-2428, Sandia National Laboratories, Albuquerque, NM, January 3.2 System Energy Resources, Inc., Grand Gulf Nuclear Station Fire Uazards Analysis.1990. 1988. 3,8 K. D. Russell et al., Intevrated Reliability and Risk 3.3 Public Sersice Company of New Ilampshire and Analysis System (IRRAS). Version 4.0, Yankee Atomic Electric Company, Seabrook Station Probabilistic Safety Assessment. Section NUREG/CR 5813 EGO-2664, EG&G Idaho, 9.4, December 1983. 3.9 V.110 et al., COMPHRN IIIe: An Interactive 3.4 Philadelphia Electric Company, Severe Accident @ uter Code for Fire Risk Analysis.EPRI-NP-Risk Assessmeut Limerick Generatine Station. 7282, May 1991. Chapter 4, Main Report, Report #4161, April 1983. 3.10 J. A. Lambright, et al., Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods 3.5 J. A. Lambright, et al., lire Risk Sconine Study: Current Percention of Unaddressed Fire Risk Intecration and Evaluation Procram fRMIIT). Issues. NURI!G/CR 5088, SAND 88-0177, Sandia Internal Fire Anaivsis. NUREG/CR-4832, Vol. 9, SAND 92-0537, Vol. 9, Sandia National National Laboratories, Albuquerque, NM, December 1988. Laboratories, Albuquerque, NM, March l993. Vol. 3 3 23 NUREG/CR-6143 APPENDIX A MANUAL FIRE SUPPRESSION ANALYSIS l l l l l l %>l 3 A-1 NUREG/CR-6143 Appendix A A.1 11ackgound and Analysis The drills were all unannounced, the only person given A.l.1 Fire lirigade Training, Organization advanced waming of the drill being the shill superintendent. The times were modified slightly based on and Standard Operating Procedures the location of the analped spaces within the plant and the I cation of the equipment cages relative to the spaces. 'Ihe Fire Drigade at Grand Gulfis a well trained and organized bngade by all standardt it appears that the brigade has the complete and active backing of A.I.2 Fuel and Fire Development management at the phtnt. The management support is shown by the participation of plant and shift superintendent The fuel used for the analysis is lubricating oils. Two in the brigade as part of the conunand structure and as scenarios were analyzed, they are a spill of approximately actual members of the brigade. filly kilograms over an area of 0.67 square meters and five kilograms over an area of 0.29 square meters. The fires The training program for all members is well organized develop from ignition to full involvement in sixty seconds with regular meetings and drills Any member not meeting and continue at that level throughout the scenarios unless the standards set by the plant for participation in the other combustibles in the room become involved. In those training is not included in the minimum manning levels set instances the heat release rates are adjusted upward to fbr the plant. This suspension is continued until the reflect the other material contribution. In all scenarios, requisite training is complete. Record keeping for all information on the fire development through nine hundred aspects of the brigade is excellent. The records include fire seconds is used in modeling All scenarios are predicted to i drills, quaderly, annual and biannual training, daily be terminated prior to this time. j manning levels, meeting attendance, classes, live fire experience and NRC required information A.I.3 Smoke Detector Response (DETACT) The shill superintendent is the chief of the brigade on each An estimate of the time for the response of the smoke shill The plant supervisor is the brigade leader on cach detectors in the space was determined using a model shitt The operators are given a list of all brigade members developed at the National Institute of Standards and on each shift to ensure manning levels are adequate prior to Technology called DETACT (D}JIector AC_[ivation). It is previous shill members leaving The normal manning level a routine included in a collection of models/ routines called for the brigade is between 8 and 10 members. The FPETOOI minimum manning level is 5 qualified members and the maximum is about i 5. During outages these levels may The input for the model is the time heat release rate growth double. All brigade members are from the operations side parameters, room height, detector spacing and properties of the plant. Total brigade membership is in the range of and ambient temperature. The output is the heat release 100 people Approximately ten percent of the members are rate necessary for detector activation and the time of also volunteer or call firefighters. activation. In a fire incident, the brigade members are omcially The model is designed for unconfined ceilings as found in notified of an alarm by the pubhe address system. All the spaces analyzed. Prior to activation of the detector members also cany radios which is the preferred method of there should be no significant buildup of hot gas below the conununication for the members. Upon receiving the alarm ceiling. Due to the rapid fire growth used in the analysis, all members report to one of the equipment cages such a heat buildup is expected in the early stages soon strategically located throughout the plant to suit up. The after detector operation. In a completely open space, the members then respond to the incident area. The brigade devices may have responded faster than indicated by this leader follows this same procedure instead of responding routine, however, the majority of the detectors were directly to the incident site for initial size up and attack blocked by cable trays located below them. The response planning. This is viewed as a weakness in the standard time index is normally a measurement of the delay of the operating procedure. Time used by other members to suit device in responding to a flow of hot gas due to its mass, up could be used by the leader to locate the fire and plan material, and configuration. To account for the time for the the initial attack. Although the time savings in the smoke to penetrate the cable trays, a higher response time incidents analyzed would be minimal saving in other index than normally used for smoke detectors wns utilized. locations may be substantial In modeling the smoke detector response, a first order The times used for fire brigade response in the analysis are approumation of the response of the smoke detector was based on records of actual fire drills conducted at the plant made by assuming that a temperature rise of about 15C for NUREG/CR-6143 A2 Vol. 3 l Appendix A a response device with a response time index of 10 The input for the model is the time and heat release rate approximates a condition at which the smoke detector growth parameters, room geometry, fire location, ambient would respond The program does not account for the temperature and the heat loss fraction (the amount of heat travel time from the generation of the smoke at the point of losses to the space enclosure as a function ofits geometry). combustion to the detector. Since devices are relatively The output is the depth of the smoke layer and the close to the fire source this a negligible factor. temperature of the smoke layer, both as a function of time. A.I.4 Time From Detection to Alarm ASET is a zone model w hich means it divides the room or space being analyzed into two zones. The upper zone is the The time to alarm is based on the requirements for alarm smoke layer and the lower zone is the ambient air layer. system in the National Fire Protection Association Being a zone model the output presents the layers as being standards for alarm transmission. The maximum allowable two distinct homogeneous layers. In reality the temperature time for a device activation to result in an alann is more a function of the distance from the ceiling of the transmission and receipt at the control / annunciator k> cation space with the highest temperatures closest to the ceiling. is ninety seconds. This time w as used as the maximum A similar condition applies to the smoke with it being estimated time in the analysis. The minimum time used thickest near the ceiling and the density of the smoke was thirty secondt decreasing with distance from the ceiling. l A.I.5 Fire and Time Related Space A.I.6 Locating the Fire l Conditions (ASETHX Room Fire The estimate of time to locate the fire is a function of the M0d'O size of the space, the configuration, time of fire brigade arrival and the predicted conditions within the space at that I.ocating the fire, agent application and extinguishment are time. The spaces analyzed are large, open, uncluttered all alTected by the conditions within the spaces during the areas with good visibility throughout. Any difficulty in period of time they are being accomplished. To estimate locating the fire will be a function of the smoke hvel in the the conditions within the spaces, the fires used to do the space. For most scenarios the smoke level was high COMPHRN analyses were effectively placed in the spaces enough that it had little impact on loc 6g ee fire. using a computer model. The fires used were severe by Therefore, the time associated with this phase is minimal. most standards due to the rapid growth m the heat release rate (from 0 KW to 735 KW and 1650 KW in 60 seconds) The use of these fires account for the rapid response of the .1 Agent Apph. cat. ion detection system and the rapid deterioration of conditions in the spaces in a relatively short time. The estimate of the time to agent application is based on the location of rnanual suppression equipment relative to the incident area. As the standpipe system outlets are located An estimate of the time related conditions within the space was determined using a model developed at the National at regular intervals throughout the plant and fire Institute of Standards and Technology called ASIIT extinguishers are readily available m all areas, the time to @vailable Safe Egress Iime). It is a routine included in a agent apphcation is agam mmimal. The same time was , used for all scenarios wittun each space as it is assumed a collection of models/ routines called FPETOOI.. 1 portion of the bngade will be laying lines or retrieving fire ASET is a simple mathematical model for estimating the extinguishers while other members are k)cating the fire. rise in temperature and the descent of the fire produced hot Therefore, additional time fer tius phase is built into the gas layer in the room of fire origin. It considers conditions previ us phase. up to the point that the room approaches flashover but does not predict the conditions following the occunence of A. I .8 Extinguishment 11ashover. ASET uses a simple point source entrainment and filling calculation that requires the user estimate the The fires modeled are relatively large with regard to heat average portion of the energy puxluced by the fire that is release rate, but are relatively small in area. Based on this lost from the flame or smoke layer. ASET has a fixed condition and the fact a small pool fire can be easily assumption that 35% of the energy m the llame is lost by extinguished by well trained personnel, times used for this radiation The user is required to enter a value representing phase are again minimal. Although the time to l the total heat loss fraction. The choice made has a major extinguishment is reported, once agent application begins impact on the calculated temperature of the hot gas laver the efket of the fire is going to be greatly diminished as the and a lesser impact on the calculated position of the tbttom cooling / smothering alTect of the agent will quickly reduce of the layer. the heat release rate associated with the fire. Vol. 3 A-3 NUREG/CR-6143 i Appendix A A.2 Results The following tables detail the results of the analysis for the fire brigade manual firefighting response to a fire incident in the Grand Gulf Nuclear Power Plant. The tables present three levels of response. The variability between the levels is diminished from similar analyses done lbr other plants due to the following three items; (1) the fire brigade is located at the plant instead of several miles away, (2) standpipe outlets for suppression activities are located conveniently throughout the plant and (3) all areas analyzed are covered by an automatic fire alarm system annunciated at a constantly attended location with area smoke detection provided at the ceiling. A.3 Summary The analysis of the manual suppression of fire incidents in several areas at the Grand Gulf Nuclear Power Plant was based on several factors that impact the ditTerent phases of any fire incident. These factors include the fuel involved, ignition, fire growth, detection, alarm, manual suppression i response, manual suppression equipment location, and the l ability of the fire brigade. Information on all these factors is provided below but a review of several factors that most impacted the analysis is provided herewith The Grand Gulf Nuclear Power Plant is a well protected plant when compared to the industry as a whole. All main plant areas surveyed were provided with smoke detection, standpipe outlets within the area or nearby and fire extinguishers and many areas had either full or panial automatic sprinkler protection. The plant has a well stalTed and very well trained fire brigade. Training for the brigade , is conducted on site in classes or in the training building capable oflive fire and smoke situations. This ensures higher than average participation by all members in actual fire fighting activities. Based on the high level of protection and the quality of the fire brigade, the times to manual suppression are much less than that which would be expected in an average plant. None of the maximum scenario times exceed filleen minutes. NUREG/CR-6143 A-4 Vol. 3 w.-y , Appendix A Table A 1 Grand Gulf Nuclear Power Plant Zone I Al17 Small Pool Fire Miscellaneous Equipment Area AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDPIPE OUTLET LOCATION: Within Space CXTINGUISIIER LOCATION: Within Space Event / Phase Description Cumulative Time (Seconds) Minimum Maximum Aversge
- 1. Detection 66 120 93 2, Alarm 96 210 153
- 3. Fire Brigade response 246 570 408
- 4. Locate fire 256 590 423
- 5. Agent application 316 650 483 6 Extinguishment 346 710 528 NOTE: The cumulative time starts at the time of established buming, any time associated with a smoldering ignition has been ignored.
Vol. 3 A-5 NUREG/CR-6143 Appendix A Table A-2 Grand Gulf Nuclear Power Plant Z4me I A117 - Large Pool Fire Miscellaneous Equipment Area AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDPIPE OUTI.ET LOCATION: Within Space EXTINGUISilER LOCATION: Within Space Event / Phase Description Cumulative Time (Seconds) Minimum Maximwn Average
- 1. Detection 35 70 53 2, Alarm 65 160 113
- 3. Fire 13rigade response 215 520 368
- 4. Incate fire 235 580 408
- 5. Agent application 295 640 468
- 6. Extinguishment 325 700 513 NOTE: The cumulative time starts at the time of established burning, any time associated with a smoldering ignition has been ignored i
I i i NUREG/CR-6143 A-6 Vol. 3 l i ~ .__ _ ._ _. _ _ _ _ . . _ l i 1 l Appendix A l l I Table A-3 Grand Gulf Nuclear Power Plant Zone I A201 - Small Pool Fire Passage Area i AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Autornatic Sprinkler Protection STANDPIPE OUTLET I,0 CATION: Within Space EXTINGUISIIER1,0 CATION: Within Space Event / Phase Description Cumulative Time (Seconds) Minimum Maximum Average
- 1. Detection 45 90 68 2, Alarm 75 180 128
- 3. Fire Brigade response 225 540 383
- 4. I,ocate fire 235 555 395
- 5. Agent application 295 615 455
- 6. Extinguishment 325 675 500 NOTE: The cumulative time starts at the time of' established buming, any time associated with a smoldering ignition has been ignored Vol. 3 A-7 NUREG/CR-6143
1 Appendix A Table A-8 Grand Gulf Nuclear Power Plant Zone I A201 Large Pool Fire Passage Area AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDP!PE OUTLET LOCATION: Within Space EXTINGUISilER LOCATION: Within Space l Event / Phase Ikscription Cumulative Time (Seconds) \ Minimum Maximum Average l 1. Detection 27 60 44 2, Alarm 57 150 104
- 3. Fire 13rigade response 207 510 359
- 4. Locate fire 227 600 414
- 5. Agent application 287 690 489
- 6. Extinguishment 317 780 549 NOTE: The cumulative time starts at the time of established buming, any time associated with a smoldering ignition has been ignored.
I NUREG/CR-6143 A-8 Vol. 3 __ ______.____.m.m _ _ - - _ . - - _ _ _ . _ - . _ _ _ _ - - - _ _ _ - - - - - _ _ . _ - _ _ _ _ - _ _ _ _ _ _ - _ - _ , _ _ _ _ _ _ _ _ _ _ _ -__-__ -. - _ . _ . Appendix A Table A-5 Grand Gulf Nuclear Power Plant Zone I A211 - Small Pool Fire Miscellaneous Equipment Area AUTOMATIC Dl!TECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDPIPE OUTLET LOCATION: Within Space !!XTINGUISIIER LOCATION: Within Space Event / Phase Description Cumulative Time (Seconds) Minimum Maximum Average
- 1. Detection 45 90 66 2, Alarm 75 180 128
- 3. Fire Brigade response 225 540 383
- 4. Locate fire 235 555 395
- 5. Agent application 295 615 455
- 6. Extinguishment 325 675 500 NOTE: The cumulative time starts at the time of established buming, any time associated with a smoldering ignition has been ignored.
Vol. 3 A-9 NUREG/CR-6143 Appendix A Table A-6 Grand Gulf Nuclear Power Plant Zone I A211 - Large Pool Fire Miscellaneous Equipment Area AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDPIPE OUTLET 1.OCATION: Within Space EXTINGUISIIER LOCATION: Within Space Event / Phase Description Cmnulative Time (Seconds) Minimum Maximum Average
- 1. Detection 27 60 44 2, Alarm 57 150 104
- 3. Fire Brigade response 207 510 359
- 4. locate fire 217 570 394
- 5. Agent application 277 630 454
- 6. Extinguishment 307 690 499 NOTli: The cumulative time starts at the time of established buming, any time associated with a smoldering ignition has been ignored.
l NUREG/CR-6143 A-10 Vol. 3 Appendix A Table A-7 Grand Gulf Nuclear Power Plant Zone I A316 - Small Pool Fire Miscellaneous Equipment Area AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDPIPE OUTIEl'l.OCATION: Within Space EXTINGUISIIER LOCATION: Within Space Event / Phase Description Cumulative Time (Seconds) Minimum Maximum Average
- 1. Detection 47 96 72 2, Alarm 77 186 132 i
- 3. Fire Brigade response 257 606 432 l
- 4. Locate fire 267 666 467
- 5. Agent application 327 726 527
- 6. Extinguishment 367 786 572 NOTE: The cumulative time starts at the time of established burning, any time associated with a smoldering ignition has been ignored Vol. 3 A-11 NUREG/CR4143
Appendix A Table A-8 Grand Gulf Nuclear Power Plant Zone I A316 - Large Pool Fire Miscellaneous Equipment Area AUTOMATIC DETECTION: Smoke Detection AUTOMATIC SUPPRESSION: Partial Automatic Sprinkler Protection STANDPIPE OUTI.ET I,0 CATION: Within Space EXTINGUISIIER 1,0 CATION: Within Space > livent/ Phase Description Cumulative Time (Seconds) Minimum Maximum Average
- 1. Detection 30 60 45 2, Alarm 60 150 105
- 3. Fire Brigade response 240 570 405
- 4. l.ocate fire 250 630 440
- 5. Agent application 310 690 500
- 6. Extinguishment 370 750 560 NOTil: The cumulative tirne stads at the time of established burning, any time associated with a smoldering ignition has been ignored-l l
NUREG/CR-6143 A 12 Vol. 3 _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ - .__._____-_m -- __s l APPENDIX B GRAND GULF CRITICAL COMPONENTS BY FIRE ZONE Vol. 3 B1 NUREG/CR-6143 .- -- _ - - - - . . _ - ==. - .. - - - - -- .- - Appendix B This appendix lists the major mechanical equipment or cables In most of the rooms, the major effect of the fire is to damage associated with that equipment for cach of the dominant fire or destroy the vital cables passing through the room. /onc3. Fire 7;me Critical Components or Cables for Critical Components IA101 LPCS MOV SA, RIIR MOV 96, RIIR MOV 94 1A102 RiIR A Ileat Exchanger, LPCI MOV 48A IA103 R1IR A Pump, RIIR MOV 24 A, LPCI MOV 48A IA105 RIIR H Pump, RIIR Pump 213, I PCI MOV 48B 1A106 RIIR B IIcat IIxchanger, LPCI MOV 48B 1A109 IIPCS Pump, IIPCS Breaker 2 1A1IOCi RIIR MOV 42B,IIPCS Breaker 2 iAl10C3 Recirc MOV 67B, SPMU MOV IB & 2B, Recirc MOV 23B 1 Al 10D3 SPMU MOVs 1B & 2B, CS MOV 28B 1All2 RIIR MOV 9B lAl14 I PCS MOV 5A lA117 ADIIRS Pump A, ADIIRS Pump B, SSW Pump C, CCW Pump IB, R1IR MOV 96, RIIR MOV 94, 1 PCI MOV 242, CRD Pump 1 A, R1IR Pump 2C, CRD Pump IB, RIIR Pump 2C 1A119 1.PCS MOV 5A IAl20 CCW Pump 1B, CCW Pump i A, CCW Pump 1C IA128 LPCI MOV 48A IAl29 I.PCI MOV 48B 1A20i Recirc Pump B, Recire MOV 67A, Recirc MOV 23A, FWS MDP, CCW Pump 1 A, LPCI MOV 48A RWCU Pump 1B, RWCU Pump I A,IA Comp (IB), CRD Pump 1 A, RIIR MOV 8A,RIIR MOV 2a, RIIR MOV 42A, SDC MOV 53 A, CS MOV 28A, SDC MOV 6A, RCIC Logic A&B, LPCS Pump 1 A, RIIR MOV 24A, RIIR MOV 8A, RilR Pump A CS MOV 28A, Unit 1 IA Compressor IB 1A202 RIIR MOV 24A 1A204 RIIR MOV 8A, RWCU Pump 1 A lA205 CS MOV 288, LPCI MOV 48B, ilPCS MOV 4, Ri!R MOV 24B, SDC MOV 6B, SDC MOV 53B, RilR MOV 24B, SDC MOV 6B, SDC MOV 53B,l.PCI MOV 48B,IIPCS MOV 4 1A207 Electrical Switchgear Room Division II, RilR MOV 24B, RIIR MOV 42B, SDC MOV 6B, RIIR Pump 2C, SDC MOV 53B, R1IR MOV 9B RIIR MOV 96, RIIR MOV 94, CS MOV 28B, LPCI MOV 48B 1A208 lilectrical Switchgear Room Division I, R1IR MOV 8A, IIPCS Breaker 2, RIIR MOV 24A, LPCI MOV 48A 1A210 RWCU Recirculation Pump Room. RIIR MOV 8A NUREG/CR-6143 B-2 Vol,3 Appendix B Fire 7ame Critical Components or Cables for Critical Components 1A211 Recirc Pump B, CCW Pump 1B, ADIIRS Pump A, ADIIRS B, SSW Pump C , RIIR MOV 96, R1IR MOV 94, CS MOV 28B, LPCI MOV 242, LPCI MOV 488, LPCI MOV 48A, CCW Pump 1C,IIPCS MOV 4, RWCU Pump 1B, IA Comp (1B), RIIR MOV 8A, RIIR MOV 2A, RIIR 140V 248, RIIR L MOV 42B, RHR MOV 42A, SDC MOV 53 A, SDC MOV 6B, RIIR Pump 2C, RIIR Pump 2B, SDC J MOV 53B,IIPCS Breaker 2, SDC MOV 6A, RCIC Logic A&B,IUIR MOV 98, LPCS MOV SA, j R1IR MOV 24A, CRD Pump 1B, SSW Pump 1 A, RIIR MOV 8 A, RIIR Pump A, R1IR MOV 248, RIIR MOV 42B, SDC MOV 6B, RIIR Pump 2B, SDC MOV 53B, RIIR MOV 98, LPCS MOV SA, RilR MOV 24A, CRD Pump IB, Unit 1 IA Compressor IB 1A215 CS MOV 28A, LPCS Pump I A, FWS MDP, CCW Pump 1 A, LPCI 48A, RWCU Pump 1 A, Unit 1 IA Cornprenor IB 1A219 LPCS MOV 5A 1A220 f.PCS Pump i A, LPCS MOV SA, LPCI MOV 242 IA221 RilR Pump 2C, CCW Pump IB,I.PCI MOV 242 lA222 Recire MOV 67A, Recire MOV 23A, ADIIRS Pump A, ADIIRS Pump B, CCW Pump IB, CCW Pump 1 A,I.PCI MOV 242, CCW Pump 1C, RWCU Pump IB, RWCU Pump 1 A, RHR Pump 2C, LPCS MOV SA, RIIR MOV 8 A, LPCS Pump 1 A, LPCS MOV 5A 1A301 FWS MDP, CCW Pump 1 A, RWCU Pump i A IA308 RIIR MOV 42B, RIIR MOV 9B, Recirc MOV 23B, CS MOV 28B 1A309 CS MOV 28A l 1A31i Recirculation MOV 23B, CS MOV 28B 1A316 Recire Pump A, Recire MOV 67B, ADIIRS Pump A, ADI1RS Pump B, SPMU MOV IB, SPMU MOV 2B, Reciic MOV 23B, SSW Pump C, CCW Pump IB, CS MOV 28B, LPCI MOV 242, LPCI MOV 48A, CCW Pump IC, RWCU Pump 1B, R1IR MOV 8A, RIIR MOV 42A, SDC MOV 53A, RIIR Pump 2C, SDC MOV 6A, RCIC Logic A&B, LPCS Pump 1 A, LPCS MOV SA, RIIR MOV 24A, SSW Pump 1 A, LPCS MOV 5 A IA401 FWS MDP, CCW Pump 1 A. RWCU Pump 1 A lA417 RIIR MOV 8A, RCIC Logic A&B,l.PCS Pump i A, LPCS MOV 5A, SSW Pump 1 A, LPCI 48A lA539 Recire Pump A, Recire MOV 67A, Recire MOV 23 A, SPMU MOV 1 A, SPMU MOV 2A, CCW Pump 1 A, LPCI MOV 48A, SSW Pump A, CRD Pump 1 A, RIIR MOV 8A, RlIR MOV 2A, RIIR MOV 42A, SDC MOV 53A, CS MOV 28A, SDC MOV 6A, RCIC Logic A&B, LPCS Pump 1 A, LPCS MOV SA, RHR MOV 24A, SSW Pump i A. RIIR MOV 8A, RIIR Pump A CS MOV 28A, LPCS MOV SA, R1IR MOV 24 A 1MI10 SSW Pump C, SSW Pump A 2Mi10 SSW Pump B, SSW MOV SB, SSW MOV 1B 2 Mil 2 SSW MOV SB 0C116 IIPCS Breaker 2 OC202 SPMU MOV 1B, SPMU MOV 2B, LPCI MOV 48A, SSW Pump A, CRD Pump 1 A, RIIR MOV 8A, RilR MOV 2A, RHR MOV 42A, SDC MOV 53 A, SDC MOV 6A, LPCS Pump 1 A, RilR MOV 24 A, SSW Pump 1 A, RilR MOV 8A, RilR Pump 2A, RIIR MOV 42A, SDC MOV 53 A, SDC MOV 6A, 1,PCS Pump 1 A. SSW Pump A Vol. 3 13 3 NUREG/CR.6143 Apperulix 13 Fire 7ene Critical Components or Cables for Critical Components OC207 SDC MOV 53 A OC208A CRD Pump i A RIIR MOV 8A, SDC MOV 53A, SDC MOV 6A, SSW Pump 1 A, LPCI MOV 48A, SSW Pump A OC208 RIIR Pump 2 A, RIIR Pump 2B, SSW MOV SH OC209 CRD Pump 1B, SSW Pump IH OC210 Division 111 Switchgear, CRD Pump 1 A, RIIR MOV 24H, RilR MOV 42B, SDC MOV 6H, RIIR Pump 2B, SDC MOV 53B,1IPCS Breaker 2, RIIR MOV 9H, CRD Pump IB, SSW Pump IB, SSW MOV 5H, SSW Pump C 0 IPCS), RIIR MOV 96, RIIR MOV 94,IIPCS MOV 4 OC211 Div.11 Hattery Room - l OC215 SSW MOV 513, SSW MOV I B, CCW Pump 1B, R1IR MOV 96, RI1R MOV 94, LPCI MOV 48H,1 A COMP (l B), RIIR MOV 24H, RIIR MOV 42B, RJIR Pump 2C, RIIR Pump 2B, SDC MOV 53B, RIIR MOV 9H, CRD Pump 1H, SSW Pump iB, RilR MOV 24H, RIIR MOV 42B, SDC MOV 53B, RilR MOV 9H, CRD Pump IH, Unit 1 IA Compressor IB OC402 RIIR Pump 2C, RI1R Pump 2B, SDC MOV 53B, CRD Pump 1H, SSW MOV 5H, Recirc MOV 23B, THCW Pump 1H, CCW Pump 1 A, THCW IC Pump, CS MOV 28H, LPCI MOV 48B, COND Pump 3B, COND Pump 3C, CCW Purnp IC, SAS 1 A Unit 1 Service Air Compressor OC407 CRD Pump 113 OC409 ADIIRS Pump A, CRD Pump iB OC504 Recirc Pump A, Recire MOV 67A, Recirc MOV 23A, FWS MDP, CCW Pump 1 A, COND Pump 3 A, FW MOV 1, RWCU Pump I A, SAS Comp l A, RWCU Pump IH l OC702 CRD Pump i A R1IR MOV 8A, RI1R Pump 2A, RIIR MOV 42A, SDC MOV 53A, CS MOV 28A, SDC MOV 6A, RCIC Logic A&ll, LPCS Pump i A, LPCS MOV 5A, THCW Pump i A CCW Pump 1 A, LPCI MOV 48A, COND Pump 3A, FW MOV 1, SSW Pump A, SAS l A Unit i Service Air Compressor l OC703 CRD Pump i A, RIIR Pump 2A, RilR MOV 42A, SDC MOV 53A, CS MOV 28A, SDC MOV 6A, RCIC 1 ogic A&B, l.PCS Pump i A, LPCI 48A OM101 Diesel Driven Fire Pump A OM102 Diesel Driven Fire Pump H FWS MDP OMI19 SAS l A Unit i Service Air Compressor, Unit 2 IA Compressor PS2Cl, Unit i IA Compressor 1B ITl18 THCW Pump 1 A, T13CW Pump IB IT132 Recire Pump B, Recire Pump A, MOV AV504, COND Pump 3 A, COND Pump 3B, COND Pump 3C IT214 TBCW Pump lH, TBCW Pump IC, SAS COMP 1 A IT218 Air Comp 3HN, THCW Pump iH. THCW Pump 1C, COND Pump 3A, COND Pump 313, COND Pump 3C, I A Comp Unit 2, SAS l A Umt 1 Service Air Compressor, Unit 2 lA Compressor P52Cl IT219 THCW Pump lH, THCW IC Pump COND Pump 3B, COND Pump 3C IT220 COND Pump 3 A, Unit 2 lA Compressor PS2Cl IT224 COND Pump 18. COND Pumn 3C NURl!O/CR-6143 B-4 Vol. 3 1 , l l Appendix B ' l Fire 7xme Critical Components or Cables for Critical Components IT226 MOV AV504, TBCW Pump i A, TBCW Pump IB, TBCW Pump 1C, COND Pump 3A, COND Pump 3B, COND Pump 3C IT306 MOV AV504 IT308 FW MOV I IT315 FW MOV 1 IT317 FW MOV 1 IT322 Recirc Pump B, Recire Pump A, MOV AV504, TBCW Pump i A, TBCW Pump 1B. TBCW Pump 1C, Cond Pump 3 A, SAS COMP 1 A, SAS I A Unit i Service Air Compressor IT323 TBCW Pump 1 A, COND Pump 3A IT324 TBCW Pump 1 A, TBCW Pump iB TBCW Pump IC, COND Pump 3 A, SAS COMP 1 A IT325 Recirc Pump B, Recirc Pump A, MOV AV504, TBCW Pump 1 A, COND Pump 3 A, SAS COMP 1 A IT327 Recire Pump B, Recire Pump A, MOV AV504, TBCW Pump 1 A,TBCW IB, TBCW Pump 1C, COND Pump 3 A, FW MOV I, COND Pump 3B, COND Pump 3C, SAS COMP 1 A l IT403 TBCW Pump 1B TBCW Pump 1C, SAS I A Unit 1 Senice Air Compressor IT404 TBCW Pump I A, COND Pump 3 A, FW MOV 1, SAS COMP 1 A IT502 Recirc Pump A, FW MOV I Vol. 3 B5 NUREG/CR-6143 Appendix C AUXILIARY BUILDING SIIUTDOWN FIRE EVENT DATA Vol. 3 C-1 NUREG/CR-6143 Appendix C Presented below is a listing and short description of fires This information is used in Section 3.3 of the main report to w hich have occurred at shutdown in nuclear power plants calculate the fire initiating event frequency for the auxiliary auxiliary buildings thru December 1989. building. Fire Event Data-Auxiliary Building Fires Plant Name Date of Plant Status Fire Type Remarks Occu rrence Trojan 3/4n6 Ilot Shutdown Insulation A short caused by breakers not properly engaging caused the ignition of insulation. Millstone 2 3/2406 Ilot Shutdown Motor Control Fire resulted from arcing of a supply Center lead. Extinguished by de-energizing. Millstone 2 11/15n6 Ilot Shutdown Relav--MCC Relay Orc in motor control center. Three Mile Island i i/22D6 Cold Shutdown MCC A misaligned stab connected to a MCC 2 branch breaker associated with the nuclear service pump stader. _ Peach Bottom 7/2907 Cold Shutdown Relay Improper installation of relay contact arm retainers. Arkansas Nuclear 8/1668 1 lot Shutdown Pump Motor I. PSI pump motor on fire (being ueal One i for shutdown cooling) due to incorrect installation of motor bearings resulting in shoning of rotor with the stator. Ilatch 1 11/23/81 Cold Shutdown Relay Insulation breakdown caused Ere in a reactor low-low RPS relay. NUREG/CR-6143 C-2 Vol. 3 i Kiyoharu Abe Ephraim Asculai Dept. of Reactor Safety Research Division of Nuclear Safety Nuclear Safety Research Center Wagramestrasse, 5 Tokai Research Establishment P.0, Box 100 JAERI A-1400 Wien Tokai-mura, Naga-gun AUSTRIA Ibaraki-ken, JAPAN Vladimar Asmolov Head, Nuclear Safety Department Sarbes Acharya I. V. Kurchatov Institute Department of Energy of Atomic Enegry NS-1/FORS Moscow, 123181 Washington, DC 20585 RUSSIA Dr. Ulvi Adalioglu J. de Assuncao Cekmece Nukleer Arastraima ve Cabinete de Proteccao & Egitim Merekezi Seguranca Nuclear P.K. 1 Ministerio da Indusstria Havaalani/ ISTANBUL Ave. de Republica 45-6 TURKEY 1000 Lisbon PORTUGAL \ Dr. Eng. Kiyoto Aizawa Senior Engineer H.P. Balfanz, Head Reactor Eng. Dev. 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Ortiz, 6400 MS9018 Central Technical Files, M. K. Yeung University of Hong Kong 8523~2 Mechanical Engineering Dept. MS0899 Technical Libmry, 7141 (5) ul a2" MS0619 Technical Publications, 7151 ggg ) Bob Youngblood Brookhaven National Laboratory Department of Nuclear Energy Bldg. 130 Upton, NY 11973 l Carlo Zaffiro 1 A.N.P.A. l Directorate for Nuclear i Via Vitaliano Brancate, 48 ) I-00144 Rome ITALY Dr. X. Zikidis Greek Atomic Energy Comm. N.R.C.P.S. "Demckritos" , GR-153 10 Agia Paraske.vi 1 Attiki GREECE Dist-14 NRC FORM 335 U.S. NUCLE AR REGUL ATORY COMMISSION 1. REPOR T NUM8tR r 1107, Numtiers w.m BIBLIOGRAPHIC DATA SHEET rs,, jm,,m t,om oa m, yarn / NUREG/CR-6143
- 2. T rr L E AND SUB O T LE S AND93-2440 Evaluation of potential Severe Accidents During Low power Vol. 3 and Shutdown Operations at Grand Gulf, Unit 1 3. DATE REPORT PUBLjSHED Analysis of Core Damage Frequency from Internal Events for um" l
"a" plant Operational State 5 During a Refueling Outage .mly 1994 UN OR GRANT NUMBER Lt923
- 5. AU THOR (S) 6. ? YPE OF REPORT J. Lambright, S. Ross I, J. Lynch ,I J. Yakle t Technical
- 7. PE R lOD COV E R E D tractus >ve oarous o ereetT " ~, ,m,ns.
- 8. ,,
PL. R Fri 0RMIN,,,G, na ss .ORG .a ANil A 1l0N - N AM E AND ADDR ESS III NMC, prov& Diwrunn, ostoce or Resion. u.5 Nuckor Repute a y commhsoon, and maismo atrkws. ' I Sandia National Laboratories Science and Engineering Associates Scier <. ? or-(,lica tions Albuquerque, NM 87185 Inc. International Corp. 6100 Uptown Blvd. NE /l0' Air park Rd SE Albuquerque, NM 87110 _,_,, jlbuquerque, NM 87106
- 9. SVONSon tNG ORGANilATlON - NAME AND ADDRESS tirNRC, type " Sam as ntewe", dconteseror. provkle V*C r 4 a PF Reykm, U S Nuckar Regulatory Commissen,
$ PHI n dding d4Afrsr33,j Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
- 10. SUPPLEME NTARY NOTES 11 ABS TR AC T (2M warns na kul This report presents the details of the analysis of core damage frequency due to fire during shutdown plant Operational State 5 at the Grand Gulf Nuclear Station. Insights from previous fire analyses (peach Bottom, Surry, LaSalle) were used to the greatest extent possible in this analysis. The fire analysis was fully integrated utilizing the same event trees and fault trees that were used in the internal events analysis.
In assessing shutdown risk due to fire at Grand Gulf, a detailed screening was performed which included the following elements: a) Computer-aided vital area analysis b) plant inspections c) Credit for automatic fire protection systems i d) Recovery of random failures i e) Detailed fire propagation modeling ! t This screening process revealed that all plant areas had a negligible (<1.0E-8 per year) contribution to fire-induced core damage frequency.
- 12. A L T WOR DS/DE SCH:PI OR S It ur wordt or pareers test *nt assist reweerneru ht torefmg the rerporr. t 13. AV A6LAssLil y bl A f tMLNI probabilistic Risk Assessment, Low power and Shutdown Unlimited Operations, Boiling Water Reactor '( se cum i v c'ea noN
/ tan Pavel Unclassified 1 '0"n"cTa$sified
- 16. NUMBER OF PAGES
- 16. PRICE
.NRCFOHM336 (2 891 _____ - -_ A l l l on recycled paper Federal Recycling Program NUREG/CR-6143, Vol.'3 ' EVALUATION OF POTENTIAL SE\ERE ACCIDENTS DURING 1.OW PO%ER ' -JULY 19941 AND SIIUTDO\W OPERATIONS AT GRAND GULF, UNIT 1: INTERNAL FIRE EVENTS UNITED STATES NUCLEAR REGULATORY COMMISSION *"((I[o7,"y*o y
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