ML20248F341

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Analysis of Core Damage Frequency: Grand Gulf, Unit 1 Internal Events Appendices
ML20248F341
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 09/30/1989
From: Mary Drouin, Lachance J, Miller S, Shapiro B, Tyrus Wheeler
SANDIA NATIONAL LABORATORIES, SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
CON-FIN-A-1228 NUREG-CR-4550, NUREG-CR-4550-V6R1P2, NUREG-CR-4550P2, SAND86-2084, NUDOCS 8910060347
Download: ML20248F341 (857)


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e NUREG/CR-4550 SAND 86-2084 Vol. 6, Rev.1, Part 2 Ana:ysis 0:?

Core Damage Frecuency:

u,ranc Gul:,,: L_m. -t_

Internal Events A;ppencices Prepared by M. T. Drouin, J. L. LaChance,11. J. Shapiro, S. Miller, T. A. Wheeler Sandia National Laboratories Prepared for U.S. Nuclear Regulatory Commission l

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AVAILABILITY NOTICE I Availability of Reference Materials Crted in NRC Publications  !

l 1

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DISCLAIMER NOTICE This report was prepared as an account of work sponsored by an agency of the United States Govemment.

Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, l expresed or implied, or assumes any legal liability of responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disc!osed in this report, or represents that its use by such third party would not infringe privately owned rights.

NUREG/CR-4550 SAND 86-2084 Vol. 6, Rev.1, Part 2 Analysis of Core Damage Frequency:

Grand Gulf, Unit 1 Internal Events Appendices s

Manuscript Completed: August 1989 Date Published: September 1989 Prepared by M. T. Drouin*, J. L LaChance*, B. J. Shapiro*, S. Miller *, T. A. Wheeler Program Manager: A. L Camp Principal Investigator: W. R. Cramond Team Leader: M. T. Drouin' Sandia National 12boratories Albuquerque,NM 87185

  • Science Applications International Corporation 2109 Air Park Road S.E.

Albuquerque, NM 87106

) Prepared for Division of Systems Research Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN A1228

\

ABSTRACT This document contains the appendices for the accident sequence analysis j

of internally initiated events for the Grand Gulf Unit 1, Nuclear Power i Plant. This is one of the five plant analyses conducted as part of the i

NUREG-ll50 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published

.f in April 1987 as NUREG/CR-4550, Volume 6. It addresses comments from numerous reviewers and significant changes to the plant systems and

procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward Probabilistic Risk Assessment practitioners who need to know how the work was done and the details for use in further studies.

The mean core damage frequency is 4.0E-6 with 5% and 95% uncertainty bounds of 1.7E-7 and 1.2E-5, respectively. Station blackout type accidents (loss of all AC power) dominate the overall results contributing about 97% of the core damage frequency. Anticipated transient without scram accidents contributed another 3%. The numerical results are driven by loss of offsite power, failure of the diesel generators, failure of the steam-driven reactor core isolation cooling system, and common cause failure of the batteries.

iii/iv

i CONTENTS Paee A. SUPPORTING THERMAL HYDRAULIC ANALYSES................A-1 B. SYSTEM FAULT TREES . ................... .......... ,g.1 C. HUMAN RELIABILITY ANALYSIS - DETAILED RESULTS........C-1 D. IMPORTANCE VALUES FOR DOMINANT CUT SETS EVENTS.......D-1 f

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l FOREVORD j This is one of numerous documents that support the preparation of the '

NUREG-1150 document by the NRC Office of Nuclear Regulatory Research.  :

Figure 1 illustrates :he front-end documentation. There are three  !

interfacing programs at Sandia National Laboratories performing ' this I work: the Accident Sequence Evaluation Program (ASEP), the Severe Accident Risk Reduction Program (SARRP), and the Phenomenology and Risk Uncertainty Evaluation Program (PRUEP). The Zion PRA was performed at i Idaho National Engineering Laboratories and Brookhaven National '

Laboratories.

Table 1 is a list of the original primary documentation and the corresponding revised documentation. There are several items-that should

.be noted. First, in the original NUREC/CR-4550 report, volume 2 was to be a summary of the internal ~ analysas. This report was deleted. In Revision 1, Volume 2 now is the expert judgment elicitation covering all plants.

Volumes 3 and 4 include external events analyses for Surry and Peach Bottom. External events for Fequoyah, Grand Gulf and Zion will be )

1 analyzed in follow up studies after NUREG-1150 is published.

The revised NUREC/CR-4551 covers the analysis included in the original NUREG/CR-4551 and NUREC/CR-4700. However, it is different from NUREG/CR-4550 in that the results from the expert judgment elicitation are given in four parts to Volume 2 with each part covering one category of issues.

The accident progression event trees are given in the appendices for each of the plant analyses.

Originally, NUREG/CR-4550 was published without the designation " Draft for Comment." Thus, the final revision of NUREG/CR-4550 is designated Revision 1. The label Revision 1 is used consistently on ' all volumes, including Volume 2 which was not part of the original documentation.

NUREC/CR-4551 was originally published as a " Draft for Comment" so, in its final form, no Revision 1 designator is required to distinguish it from the previous documentation. s There are several other reports published in association with NUREG-1150.

These are:

NUREC/CR-5032, SAND 87-2428, Modelina Time to Recovery and Initiatina, Event Frecuency for Loss of Off-site Power Incidents at Nuclear Power Plants, R. L. Iman and S. C. Hora, Sandia National Laboratories, Albuquerque, NM, January 1988.

NUREC/CR-4840, SAND 88-3102, Methodology for External Event Screenine Ouantffication - RMIEP Methodolorv, M. P. Sohn and J. A. Lambright, Sandia National Laboratories, Albuquerque, NM, July 1989, vii

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3 NUREG/CR-4772, SAND 86-1996, Accident Secuence Evalurition Procram Human Reliability Analysis Procedure, A. D. Swain III, Sandia National Laboratories, Albuquerque, NM, February 1987.

NUREG/CR-5263, SAND 88-3100, The Risk Management Implications of NUREG-1150 Methods and Results, _A. L. Camp et al., Sandia National Laboratories, Albuquerque, NM, September 1989 A Human Reliability Analysis for the ATUS Accident Seouence with MSIV Closure at the Peach Bottom Atomic Power Station, A-3272, W. J.

Luckas, Jr. et al. , Brookhaven National Laboratory, Upton, NY,1986.

A brief flow chart for the documentation is given in Figure 2. Any related supporting documents to the back-end NUREG/CR-4551 analyses are delineated in NUREG/CR-4551. A complete list of the revised NUREC/CR-4550, Revision 1 volumes and parts is given below.

General '

NUREG/CR-4550, Revision 1 Volume 1, SAND 86-2084, Analysis of Core Damare Frecuenev: Methodolorv Guidelines for Internal Events.

NUREG/CR-4550, Revision 1, Volume 2, Part 1 SAND 86-2084, Analysis of Core Damaga Frecuenev: Expert- Judement Elicitation on Internal Lvents Issues - Excert Panel.

NUREC/CR-4550, Revision 1, Volume 2, Part 2, SAND 86-2084, Analysis of Core Damage Frecuenev: Excert Judgment Elicitation on Internal Events Issues - Proiect Staff.

Parts 1 and 2 of Volume 2, NUREG/CR-4550 were published in one binder.

This volume was published in April 1989 and distributed in May 1989 with an incorrect title, i.e., Analysis of Core Damage Frequency from Internal Events: Expert Judgment Elicitation, without the Revision 1 designation.

l The complete, correct title is: NUREG/CR-4550, Revision 1, Voluma 2, SAND 86-2084, Analysis of Core Damage Frequency: Expert Judgment Elicitation on Internal Events Issues.

l Surry NUREG/CR-4550, Revision 1, Volume 3. Part 1, SAND 86-2084, Analysis of Core Damare Frecuenev: Surry Unit 1 Internal Events.

NUREG/CR-4550, Revision 1, Volume 3 Part 2. SAND 86-2084, Analysis of Core Damare Frecuency: Surry Unit 1 Internal Events Aeoendices.

NUREC/CR 4550, Revision 1, Volume 3, Part 3, SAND 86-2084, Analysis of Core Damare Frecuenev: Surry Unit 1 External Events.

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1 Peach Bottom l NUREC/CR 4697, EGG-2464, Containment Ventine Analysis for the Peach Bottom Atomic Power Station, D. J. Hansen, et al., Idaho National Engineeri,ng Laboratory (EGGG Idaho, Inc.) February 1987 NUREG/CR 4550, Revision 1 Volume 4, Part 1. SAND 86-2084, Analysis of Core Damare Frecuenev: Peach Bottom Unit 2 Internal Events.

NUREG/CR-4550, Revision 1, Volume 4. Part 2, SAND 86-2084, Analysis of Core Damare Frecuenev Peach Bottom Unit 2 Internal Events ADoendices.

9 NUREG/CR-4550, Revision 1, Volume 4, Part 3 SAND 86-2084, Analysis of Core Damare Frecuenev: Peach Botton Unit 2 External Events. ,

l Seouovah NUREC/CR-4550, Revision 1, Volume 5, Part 1, SAND 86-2084, Analvsis of j Core Damare Frecuenev: Seouovah Unit 1 Internal Events. l NUREC/CR-4550, Revision 1. Volume 5, Part 2, SANDB6-2084, Analvsis of Core Damare Frecuenev: Seouovah Unit 1 Internal Events Appendices.

Grand Gulf NUREG/CR 4550, Revision 1, Volume 6 Part 1, SAND 86-2084, Analysis of Core Damare Frecuenev: Grand Gulf Unit 1 Internal Events, NUREG/CR 4550, Revision 1, Volume 6, Part 2. SAND 86-2084, Analysis of Core Damare Frecuenev: Grand Gulf Unit 1 Internal Events Appendices.

ZLED NUREG/CR-4550, Revision 1 Volume 7, EGG-2554, Analysis of Core pamare Frecuenev: Zion Unit 1 Internal Events.

xii

APPENDIX A SUPPORTING THERMAL-HYDRAULIC ANALYSES i

1

Table of Contents fra_ge Section A.1 ATWS Supporting Calculations............................ A-5 A.2 Accident Sequence Timing Supporting Calculations........ A-5 A.3 Success Criteria Supporting Calculations................ A-10 A4 Abbreviations........................................... A-15 A.5 References...............................................

A-15 I

1 1

I 1

l 1

A-2

List of Figures figure Pane A-1 Downcomer Water Level for Vessel Boiloff Calculation.... A-ll A-2 Reactor Vessel Pressure for RCIC Injection Calculation.. A-12 A-3 Reactor Vessel Water Level for RCIC Inj ection  ;

Calculation........................... ................. A-13 A-4 Containment Pressure for RCIC injection Calculation. . . . . A-14 A-5 Downcomer Water Level for CRD Injection Calculation. . . . . A-16 A-3

List of Tables I_abh a Pace A-1 ATWS B W LTAS Runs............. ..... . .... .......... A-6 A-2 Summary of ATWS MSIV Closure Sequence Analyses........ . A-9 i

93 A-4

i I

l A. SUPPORTING THERMAL-HYDRAULIC ANALYSIS i

i This appendix describes the thermal-hydraulic calculations used in  ;

the Grand Gulf probabilistic risk analysis.

support of Both plant-specific calculations and calculations for other plants were used in support of three areas:

o To determine behavior' during an Anticipated Transient Without Scram (ATWS),

o To determine timing of important phenomena, and

)

~o. To help determine success criteria.

The calculations used in all three areas are discussed in the following-  ;

sections.

A.1 ATUS Suonortine Calculations The thermal-hydraulic information used in construction of the ATWS event k tree was obtained . from ATWS code simulations performed for Grand Gulf l and LaSalle and from related studies. The ATWS code simulations were i performed using the B N LTAS code [1] and are given in Table A-1.

Other sources of information are General Electric (GE) generic ATWS- j analyses on boiling water reactors [2), an Oak Ridge National Laboratory {

ATWS Main Steam Isolation Valve (MSIV) closure analysis using Brown's J Ferry as the model plant [3) . and an ATWS study by the Idaho National Engineering Laboratory (INEL) [4). The results _ of these studies are given in Table A-2. It should be noted that, while the thermal-hydraulic calculations for these sources renerally agree, specific results do vary somewhat. This represents ' uncertainties in sequence timing and success criteria which are not explicitly analyzed in this study because of resource limitations.

l A.2 Accident Secuence Timine Suncortine Calculations-BWR-LTAS code calculations were performed to determine the following timing questions: l o When core uncovery begins and when the core is uncovered down to two feet above the Bottom of the Active Fuel (BAF) following loss of coolant injection, and i

o When a high turbine exhaust pressure trip of the Reactor '

Core Isolation . Cooling (RCIC) pump occurs following a loss of containment heat removal.  !

i For the boiloff calculation, the MSIVs were closed and no injection to i the reactor vessel was allowed. The simulation started 120 see after j the initiating event and assumed that the Low Low setpoints were in i effect for automatic actuation of the Safety Relief. Valves (SRVs).

Reactor vessel pressure was relieved by a single SRV cycling between 1,033 psia and 926 psia. Since there 'was no source of injection to the i

A-5

Table A ATWS BWR-LTAS Runs DESCRIPTION RESULTS/ CONCLUSIONS / NOTES Plant /Run: Grand Gulf:

o Standby Liquid Control (SLC) o One Control Rod Drive (CRD) starts injecting at three pump injects from time zero, minutes.

o High Pressure Core Spray o All systems other than SLC (HPCS) begins to inject at  !

actuate automatically. ~85 seconds due to low reactor vessel water level, o Automatic Depressurization System j (ADS) is inhibited.

o AT 150 seconds core inlet ilow is less than 5% of full o RCIC System available, power flow; thus boron 3 mixing is zero (no power i reduction).

o HPCS and one CRD pump succeed.

o Reactor vessel water level o Intercompartmental leakage area is at Top of Active Fuel present between containment and (TAF) at 155 seconds.

drywell equivalent to 2,600 cfm at 3 psid. o With ADS inhibited, the low pressure systems are never available for inj ection.

Injection' from HPCS and one CRD ' pump is not sufficient to raise the level. above TAF.

Plant /Run: LaSalle Run 1:

o Feedwater available, o One CRD and HPCS inject at

-1 minute.

o HPCS and RCIC available.

o RCIC injects at ~2 minutes, o Low Pressure Core Spray (LPCS) and Low Pressure Coolant Injection o Boration starts and HPCS and (LPCI) available. RCIC trip at ~3 minutes.

o Condensate Booster Pumps (CBP) o Level' control using feed-available, water pump starts in ~3.5 minutes (Level 540").

o One CRD pump available, o No suppression pool cooling initially available.

o Operator injects boron at suppression pool temperature of 1200F, A-6

F Table A-1 ~

ATUS BWR-LTAS Runs (Continued) l DESCRIPTION RESULTS/ CONCLUSIONS / NOTES l Plant /Run: LaSalle Run 1: (Concluded) I 1

o Operator uses only CRD pump for injection once boration commences.

o ADS is inhibited, o Operator actuates single element control of feedwater pump when level falls to TAF and restores level to 540".

o Initiate suppression pool cooling at 120 seconds after commencement of 540" level restoration.

Plant /Run: LaSalle Run 2:

o Feedwater available. o One CRD and HPCS inject at

-1 minute.

o HPCS and RCIC available, o RCIC inj ects at -2 minutes.

o LPCS and LPCI available, o Boration starts and HPCS and o Condensate Booster Pumps (CBP) RCIC trip at ~3 minutes.

available.

o Level control using feed-o One CRD pump available, water pump starts in ~3.5 minutes (Level 540"),

o No suppression pool cooling initially available. o Boration complete, level raised to 540" at -24 o Operator injects boron at minutes.

suppression pool temperature of 1200F, o Operator uses only CRD pump for injection once boration commences.

o ADS is inhibited.

o Operator actuates single element control of feedwater pump when level falls to TAF and restores I level to 540".

o Initiate suppression pool cooling at 120 seconds after commencement of 540" level restoration.

A-7

Table A-1 ATWS BWR-LTAS Runs (Concluded)

DEMRIPTION RESULTS/ CONCLUSIONS / NOTES Plant /Run: LaSalle Run 3: j i

o No operator action, o One CRD pump and HPCS inject j at ~1 minute.

o No feedwater.

o RCIC injects at ~2 minutes, o No suppression pool cooling.

o ADS at ~3.5 minutes, o HPCS and RCIC available, o CBP injects at ~4 minutes.

o LPCS and LPCI available, o LPCS and LPCI inject at ~5 o CBP available, minutes, o HPCS and LPCS fail at ~5 o One CRD pump available, minutos from high suppres-sion pool temperatures, o ADS available, o RCIC fails at ~17.5 minutes l from high containment pressure.

o At ~26.5 minutes, drywell pressure is greater than 100 psia, closing the SRVs and failing LPCI and the CBPs.

o The peak pressure for this sequence is 1075 psia. The oscillations are such that there are 200 seconds between peaks.

1 1

A-8

Table A-2 Summary of ATWS MSIV Closure Sequence Analyses INEL CALCU1ATIONS FOR BROWN'S FERRY SLC (Minimum Effectiveness) Level-Control, No Depressurization, and No Pool Cooling o SLC initiated at 2 minutes.

o Operator reduces level to TAF beginning at 2 minutes.

o Sufficient ' boron inj ected to shutdown reactor at -16 minutes, q operator raises water level resulting in insertion of boron into core.

o Maximum pool temperature reaches 1730F at ~17 minutes. ,

o Depressurization of the plant to stay within suppression pool heat- I capacity temperature limits not performed.

CE CALCU1ATIONS FOR BWR/6 MARK III SLC, No Level Control, No Depressurization, and Pool Cooling o SLC initiated at 2 minutes.  ;

o Suppression pool cooling begins at 11 minutes. '

o Maximum suppression' pool temperature of 1670F reached at 20 minutes, o Maximum containment pressure of 6.9 psia.

l l

I A-9

reactor vessel, the downconer level fell steadily until it reached 491.0" at 555 see as indicated in Figure A-1. At that point, the upper pool dump timer started because of low reactor vessel water level, and 1,800 see later, at 2,350 sec, the upper pool began to dump its  :

inventory to the suppression pool. The upper pool emptied in 240 sec.

Meanwhile, the level reached the top of the active fuel, 366", at 1,685 sec, and the 240 level, 2 ft above the BAF, at approximately 4,065 sec.

l For the RCIC injection calculation, the MSIVs were closed and the only source of inj ection to the reactor vessel was from the RCIC turbine driven pump. The simulation began 120 see after the initiating event and assumed the Low Low setpoints were in effect for automatic actuation of the SRVs. This allowed reactor pressure to be relieved by a single SRV cycling between 1,033 psia and 926 psia. The RCIC system was configured for control by the operator, controlling the level between the initiation setpoint of 500.0" and a trip setpoint of 550.0". The i RCIC system injected to the reactor vessel seven times before the vessel I was depressurized at 9,700 see as indicated in Figure A-2, due to exceeding the suppression pool heat capacity limit. Reactor pressure I was controlled to a target level of 150 psia, high enough to allow the I RCIC system to continue in operation. After reactor depressurization, the collapsed water level fell to approximately 390" and then recovered because of prolonged RCIC injection flow as shown in Figure A-3. After level recovery, the RCIC system was used to keep the level between 500.0" and 550.0". Level was maintained until RCIC system isolation at 35,000 sec (9.7 hrs), because of a high turbine exhaust backpressure of 39.7 psia at the RCIC turbine exhaust pipe inlet (See Figure A-4). j With loss of inj ection at 35,000 sec (9.7 hrs), the reactor vessel repressurized, with pressure relief via a single SRV. The reactor vessel collapsed water level' fell rapidly, reaching the TAF, 366", at approximately 40,070 sec (11.1 hrs). The collapsed level reached 2 ft above the bottom of the active fuel, 240", at approximately 47,965 see (13.3 hrs).

A.3 Success Criteria Sucoortine Calculations A BWR-LTAS calculation was performed to determine if use of the CRD system in the enhanced flow mode (defined as both pumps operating) can cool the core. The MSIVs were closed and only injection by the two CRD system pumps to the reactor vessel was allowed. The simulation started i 120 see after the initiating event and assumed that the Low Low i

setpoints were in effect for automatic actuation of the SRVs. Reactor vessel pressure was relieved by a single SRV cycling between 1,033 psia and 926 psia. It was assumed that both CRD pumps were in operation from the start of the simulation, 120 see into the accident, and that they were pumping through a flowpath so configured to allow delivery of 238 gpm at 1,103 psia. Since the reactor vessel pressure was cycling between 926 psia and 1033 psia, actual CRD system flow to the reactor vessel was between 290 gpm and 260 gpm. That flow rate is sufficient to remove approximately 1.2% rated power. With the decay heat curve installed in BWR-LTAS, that point would be reached after approximately 1.3 hrs. The reactor vessel level fell steadily, with the 366" level A-10

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(top of active fuel) being reached at 2,320 sec, 635 see later than for the pure boiloff calculation discussed in Section A.2. The water level shown in Figure A-5 began to recover at around 6,000 see (1.67 hrs).

The 20 min difference between the time when decay power reached 1.2% and the time the water level recovery began is likely due to the release of the energy from the reactor vessel and its internals to the coolant.

Reactor vessel water level continued to recover slowly until 10,380 sec (2.9 hrs), when the suppression pool heat capacity limit was reached.

(It should be noted here that the upper pool began dumping to the suppression pool at 2,650 see and completed 240 sec later). At that point the reactor vessel was depressurized using the eight automatic system depressurization system SRVs, and blowdown to ambient containment pressure was accomplished. However, with the reduction in reactor vessel pressure and the enhanced nature of the CRD pump flow, (two CRD pumps in operation), the CRD system tried to deliver more flow than the condensate storage tank could provide, resulting in tripping of the  ;

pumps due to low CRD pump suction pressure.

A.4 Abbreviations

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l ADS Automatic Depressurization System l ATWS Anticipated Transient Without Scram l

BAF Bottom of Active Fuel {

BWR Boiling Water Reactor CBP Condensate Booster Pumps {

CRD Control Rod Drive i GE General Electric HPCS High Pressure Core Spray '

INEL Idaho National Engineering Laboratory LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray MSIV Main Steam Isolation Valve RCIC Reactor Core Isolation Cooling SLC Standby Liquid Control SRV Safety Relief Valve TAF Top of Active Fuel A.5 References

[1] R. M. Harrington and L. C. Fuller, BWR-LTAS: A Boiline Water Reactor Lont-Term Accident Simulation Code, Oak Ridge National Laboratory, NUREG/CR-3764, ORNL/TM-9163, February 1985.

[2] Assessment of BUR Mitigation of ATUS (NUREG 0460 Alternate No.3),

General Electric, NEDO-24222, 80NED021, CLASS I, February 1981.

[3] R. M. Harrington and S. A. Hodge, ATWS at Browns Ferry Unit One -

Accident Secuence Analysis, Oak Ridge National Laboratory, NUREG/CR-3470, ORNL/TM-8902, July 1984.

{4] R. J. Dallman, et al., Severe Accident Secuence Analysis Procram -

Ant icinated Transient Without Scram Simulations for Browns Ferry Nw br Plant Unit 1, Idaho National Engineering Laboratory (EG6G 1delm , NUHEC/CR-4165, EG6G-2379 (Draft), February 1985.

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APPENDIX B SYSTEM FAULT TREES i

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l

Table of Contents Figure fa&E B-1 High Pressure Core Spray System Fault Tree.............. B-3 B-2 Reactor Core Isolation Cooling System Fault Tree........ B-13 B-3 Control Rod Drive System Fault Tree..................... B-30 B-4 Standby Liquid Control System Fault Tree.............. . B-61 B-5 Suppression Pool Makeup System Fault Tree............... B-82 B-6 Automatic Depressurization System Fault Tree............ B-94 B-7 Condensate System Fault Tree............................ B-127 B-8 Low Pressure Core Spray System Fault Tree............... B-129 B-9 Low Pressure Coolant Inj ection System Fault Tree. . . . . . . . B-140 i B-10 Standby Service Water Cross-Tia System Fault Tree....... B-182 i B-ll Firewater System Fault Tree............................. B-189  ;

B-12 Suppression Pool Cooling System Fault Tree.............. B-197 l B-13 Shutdown Cooling System Fault Tree...................... B-218  !

B-14 Containment Spray System Fault Tree..................... B-245 B-15 Containment Venting System Fault Tree................... B-263 i B-16 Emergency AC Electrical System Fault Tree...............-B-267 )

B-17 Emergency DC Electrical System Fault Tree............... B-280 l B-18 Diesel Cenerator Cross-Tie Fault Tree................... B-290 l B-19 Standby Service Water System: Diesel Generator Cooling  ;

Faul t T r e e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B - 2 9 4 i B-20 Standby Service Water System: Residual Heat Removal j Heat Exchanger Cooling Fault Tree....................... B-312 l B-21 Standby Service Water System: High Pressure Core l Spray Room Cooling Fault Tree........................... B-317 B-22 Standby Service Water System: Residual Heat Removal Room Cooling Fault Tree................................. B-319 B-23 Standby Service Water System: Low Pressure Core Spray Room Cooling Fault Tree............. ................... B-320 ]

B-24 Standby Service Water System: Reactor Core Isolation Cooling Room Cooling Fault Tree......................... B-325 B-25 Standby Service Water System: Residual Heat Removal Pump Cooling Fault Tree................................. B-327 B-26 Standby Service Water System: Common Element Fault Trees................................................... B-334 B-27 Emergency Ventilation System: Diesel Room Fault Tree... B-346 B-28 Emergency Ventilation System: High Pressure Core Spray, Reactor Core Isolation Cooling, and Low Pressure Core l Spray Rooms Fault Trees................................. B-353 B-29 Emergency Ventilation System: Low Pressure Coolant Injection Room Fault Tree. . . . ........................... B-357 B 30 Instrument Air System Fault Tree........................ B-361

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