Text

Review of Seabrook Units 1 and 2 Auxiliary Feedwater System Reliability Analysis
	ML20080M435
Person / Time
Site:	Seabrook
Issue date:	02/29/1984
From:	Fresco A, Papazoglou I, Youngblood R; BROOKHAVEN NATIONAL LABORATORY
To:	; Office of Nuclear Reactor Regulation
References
	CON-FIN-A-3933 BNL-NUREG-51723, NUREG-CR-3531, NUDOCS 8402170503
	Download: ML20080M435 (194)
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. NUREG/CR-3531 BNL-NUREG-51723 L

l l Review of the l l Seabrook Units 1 and 2 '

l Auxiliary Feedwater System Reliability Analysis I

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Prepared by A. Fresco, R. Youngblood, l. A. Papazoglou Brookhaven National Laboratory Prepared for U.S. Nuclear Regulatory j Commission l

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a s NOTICE This trJort was prepared as an account of work :ponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability of re- f sponsibility for any third party's use, or the results of such use, of any information, apparatus, '

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NUREG/CR-3531 BNL-NUREG-51723 Review of the Seabrook Units 1 and 2 Auxiliary Feedwater System l

Reliability Analysis Minuscript Completed: October 1983 DIto Published: February 1984 Przpared by A. Fresco, R. Youngblood, l. A. Papazoglou Brookhaven Nathnal Laboratory Upton, NY 11975 Prepared for Division of Ssfety Technoiogy Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission W:shington, D.C. 20555 NRC FIN A3393

l ABSTRACT This report presents the results of a review of the Emergency Feedwater System Reliability Analysis for Seabrook Nuclear Station Units 1 and 2. The objective of this report is to estimate the probability that the Emergency Feedwater System will fail to perform its mission for each of three different initiators: (1) loss of main feedwater with offsite power available, (2) loss cf offsite power, (3) loss of all AC power except vital inetrumentation and control 125 VDC/120 VAC power. The scope, methodology, and failure data are prescribed by NUREG-0611, Appendix III. The results are compared with those obtained in NUREG-0611 for other Westinghouse plants.

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TABLE OF CONTENTS Page ABSTRACT ............................................................... iii LIST OF FIGURES ........................................................ vii LIS1' 0F TABLES ......................................................... vii

SUMMARY

AND CONCLUSIONS ................................................ ix 1.0 INTLODUCTION ...................................................... 1 2.0 SCOPE OF BNL REVIEW ............................................... 2 3.0 MISSION SUCCESS CRITERIA .......................................... 3 4.0 SYSTEM DESCRIPTION ................................... ............ 4 4.1 Con figurati on and Overall Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.1.1 Startup Feed Pump System .............................. 4 4.1.2 Eme rgency Feed wate r Sys tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.2 Component Design Cl assi fi cation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.3 P o we r S o u r c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.4 Instrumentation and Control s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5.0 EMERGENCY OPERATION ............................................... 11 5.1 Lo s s o f Ma i n Fe ed wa t e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 5.2 Lo s s o f O f f s i t e Po we r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 5.3 Lo s s o f Al l AC Po we r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 6.0 TESTING........................................................... 14 7.0 TECHNICAL SPECIFICATIONS .......................................... 15 8.0 ASSUMPTION 5 ....................................................... 17 8.1 Ge n e r al Fa i l u r e Da ta . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 8.2 Treatment of Time Dependent Fail ures . . . . . . . . . . . . . . . . . . . . . . . . . 18

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8.3 Te s t a nd Ma i n te n a n ce Ou ta g e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 8.4 Operator Errors .............................................. 24 9.0 RELIABILITY-ANALYSIS .............................................. 27 9.1 Qualitative Aspects .......................................... 27 9.1.1 Mode o f Sy s tem In i t i a ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 9.1.2 System Contrei Followi ng Initiation . . . . . . . . . . . . . . . . . . . 27 9.1.3 Effects of Test and Maintenance Activities . . .. . . . . . . . . 29 9.1.4 Avail ability of Al ternate Water Suppl ies . . . . . . . . . . . . . . 29 9.1.5 Adequacy and Separation of Fower Sources . . . . . . . . . . . . . . 29 v

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TABLE OF CONTENTS (Cont'd)

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l Page 9.1.6 Common Mode Failures .................................. 29 9.1. 7 Si ngl e Poi nt Fai l u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 9.1.8 Adequacy of Emergency Procedures ...................... 31 9.2 Quantitative Aspects ......................................... 31 9.2.1 Applicant's Use of NRC-Suggested Methodology and Data.. 31 9.2.1.1 Fault Tree Construction and Evaluation ....... 31 9.2.1.2 Failure Data ................................. 34 9.2.2 Applicant's Results ................................... 34 9.2.2.1 Sy stem Un avai l abi l i ti es . . . . . . . . . . . . . . . . . . . . . . 34 9.2.2.2 Dominant Failure Modes and Conclusions ....... 35 9.2.3 BNL Assessment ........................................ 36 9.2.3.1 Fault Trees .................................. 36 l

9.2.3.2 Failure Data ................................. 39

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9.2.3.3 System Unavailabilities ...................... 39 9.2.3.4 Dominant Failure Modes ....................... 41 .

1 9.P.3.5 General Comparison to Dther Plants . . . .. .... . . 42 9.2.3.6 General Comments ............................. 43 REFERENCES ............................................................. 44 APPENDIX A SEABROOK AUXILIARY FEEDWATER SYSTEM FAULT TREES ............ A-1 APPENDIX B FAULT TREE DATA ............................................ B-1 APPENDIX C BACKGROUND INFORMATION PROVIDED BY THE APPLICANT ........... C-1 APPENDIX D Letter to Mr. F. J. Miraglia, Chief - Licensing Branch No.3, Division of Licensing, U.S. NRC from J. DeVincentis, Project Manager, PSNH, "Seabrook Station Emergency Feedwater System Design Changes," SBN-321, September 7, 1982................. D-1 SUPPLEMENT Seabrook AFWS Revised Reliability Assessment for LOOP....... S-1 vi

LIST OF FIGURES Figure Title Page 1 Comparison of Reliability of Seabrook AFWS to Other AFWS Designs in Pl ants Using the Westinghouse NSSS. . . . . . . . . . . . . . . . . x 2 Seabrook Nuclear Station...................................... 45 3 Seabrook Nuclear Station Emergency Feedwater System........... 46 4 Steam Supply fo r Turbi ne-Dri ven EFW Pump. . . . . . . . . . . . . . . . . . . . . . 47 5 Startup Feed Pump Normal Alignment to the Main Feedwater Sy s tem ( Si m pl i fi ed ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 6 Moto r-Driven EFW Pump Cont rol Logi c. . . . . . . . . . . . . . . . . . . . . . . . . . . 49 7 Applicant's WAM Results for Loss of Main Feedwater............ 50 8 Applicant's WAM Results for loss of Offsite Power...........< . 51 9 Applicant's WAM Resul ts for Total Loss of AC Power.. .. . . . . . .. . 53 10- BNL Cutsets - LMFW............................................ 54 11 BNL Cutsets - L00P............................................ 57 12 BNL Cutsets - Top Event No Flow to 3 out of 4 Steam Generators - LMFW Gate AFW.................................... 60 13 BNL Cutsets - Top Event No Flow to 3 out of 4 Steam Generators - LOOP Gate AFW.................................... 62 LIST OF TABLES Table Title Page 1 Unavailabilities of Seabrook AFWS Comparison of Applicant's Results to BNL Assessment..................................... xi 2 Applicant's Summary of Maintenance and Test Unavailabilities.. 64 3 Applicant's Summary of Operator Actions / Failure Probabilities. 65 4 Appl i cant 's AFW System Un rel i abi l ity. . . . . . . . . . . . . . . . . . . . . . . . . . 66 i 5 Applicant's Results - Dominant Contributors to Conditional j Unavailability Loss of Main Feedwater Event................... 67 6 Applicant's Results - Dominant Contributors to Conditional Unavailability - Loss of Main Feedwater/ Loss of Offsite Power L

Event......................................................... 68

! 7 Applicant's Results - Dominant Contributors to Conditional l Unavailability - Loss of Main Feedwater/ Loss of All AC Power.. 69 8 SNL Results - Unavailability of Seabrook AFWS - REF.3 Design Using NUREG-0611 Data LMFW Transient................... 70 9 BNL Results - Unavailability of Seabrook AFWS - Proposed g

Design Using NUREG-0611 Data LMFW Transient. . . . . . . . . . . . . . . . . . . 72 vii

LIST OF TABLES (Cont'd)

Table Title Page 10 BNL Results - Unavailability of Seabrook AFWS - REF.3 Design Using NUREG-0611 Data LOOP Transient................... 74 11 BNL Results - Unavailability of Seabrook AFWS - Proposed Design Using NUREG-0611 Data LOOP Transient................... 76 12 Summary of BNL Assessments.................................... 78 A.1 Sy st em Identi fi cati on Code. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2 A.2 Component Types.............................................. A-3 A.3 Fa u l t C o d e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-5 B.1 Mechanical and Electrical Component Failure Rates............ B-1 B.2 Fault Identifiers for th9 Seabrook Emergency Feed Station.... B-8 B.3 NRC-Supplied Data Used for Purposes of Conducting A Com-parative Assessment of Existing AFWS Designs and Their Potenti al Rel i abi l i ti es. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-34 C.1 Engineering Drawing List for the Seabrook Nuclear Station Emergency and Startup Feedwater Systems. . . . . . . . . . . . . . . . . . . . . . C-3 I

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SUMMARY

AND CONCLUSIONS After the accident at Three Mile Island, a study was performed of the re-liability of the auxiliary feedwater system (AFWS) of each then-operating plant with NSSS designp by Westinghouse. The resu in.NUREG-0611.ll i At the request of the HRC,p Wi theof that study Yankee Atomic were presented Electric

~ Company and the-Public Service Company of New Hampshire, operating license ap-plicag 1 and 2

.AFWS,\gs,haveprovidedtheNRCwithastudyoftheSeabrookUnits l performed using NUREG-0611 as a guideline. BNL has reviewed this study. The BNL conclusions are as follows ("High", " Medium" and " Low" refer to

-the NUREG-0611 reliability scale).

1. For an accident resulting in a loss of main feedwater (LMFW) with offsite power available: The reliab lity of AFWS is in the High range. (Unavailability =1.95x10g/ demand.)

2. For a loss of offsite power (LOOP) resulting in a concurrent loss of mainfeedwater(LMFW): The reliability of the AFWS is in the High range, provided.that systen design changes as described in the Sup-plement of this report are implementod. (Unavailability = 8.6x10-57 demand.)

3.' For a loss of all AC power (LOAC), except for the 125 VOC/120 VAC vital instrumentation and control power systems, resulting in a con-current loss of main feedwater (LMFW): The reliability of the AFWS is in the _ Medium rang'e. (Unavailability = 2.3 x 10-2/ demand.)

The results are summarized in Table 1. Two separate calculations were performed by BNL. The first was based on the Emergracy Feedwater system design as it was reported-in REF.3. The second was based on modifications proposed by the ~applicent and contained in a September 7,1982 letter which appears in Appendix D. Additional modifications were proposed on November 17, 1982 and

.resulted in a third calculation for the unavailability of the system for LOOP condit$ons only. The results of.the two calculations, as well as the third partial one, are given in Table 1 along with the results of the applicant's analysis. A comparison of the Seabrook AFWS reliability to other AFWS designs in plants using the Westinghouie NSSS is shown in Figure 1.

The design mcdifications mentioned above are quite significant and are described in Appendix D and the Supplement to this report. In all of the calculations, it was assumed that the Start-Up Feed Pump will be subjected to the .same or.more stringent Technical Specification outage limitations as the Emergency Feedwater Pumps. If not.'the BNL assessed unavailabilities quoted in this report vill be subject to substantial increase.

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Tiensient Events LMFW LMFW/ LOOP LMFW/ Loss of All AC' Plants low Med High Low Med High Low Med High Seabrook 9g Haddam Nesk G I>

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No. Anna G G ,9 h- Order of Magnitude in Unavailabihty Represented.

' Note: The scale for this event is not the same as that for the LMFW and LMFW/ LOOP.

BNL Assessment - NUREG-0611 Scope.

S Reference 3 Design A Proposed Design

$ Supplemert Design (Nov. 17,1982)

Applicant's Results m Reference 3 Design Figure 1: Comparison of Reliability of Seabrook AFWS to Other AFWS Designs in Plants Using the Westinghouse NSSS.

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Table 1 Unavailabilities of Seabrook AFWS Comparisc.n of Applicant's Results to SNL Assessment APPLICANT'S*

RESULTS BNL ASSESSMENT ** J Transient REF.3 Design REF.3 Design Proposed Design Suppl. Design i

1. LMFW 2.1x10-5 4.5x10-5 1.35x10-5 .____

5.2x10-5 1.8x10-4 1.15x10-4 , 8.6x10-5

2. LOOP

3. LOAC 2.1x10-2 2.3x10-2 2.3x10-2 ____

Using Applicant's Data

- Using.NUREG-0611 Data Note: The Proposed Design refers to the design revisions described in the applicant's letter of September 7,1982 which appears in Ap-pendix D. After the draft version of this report was transmit-t ted to the NRC, the applicant proposed further changes on November 17, 1982. Those changes and their effect on system un-availability for LOOP are described in the Supplement.

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1.0 INTRODUCTION

This report is a review by Brookhaven National Laboratory (BNL) of WLA R-82-02, " Reliability Analysis of the Emergency Feedwater System at the Sea-brook Nuclear Power Station", which was prepared by Wood-Leaver and Associates for Yankee Atomic Electric Company and the Public Service Company of New Ham-pshire.

After the. accident at Three Mile Island, a study was perfomed of the Auxiliary Feedwater Systens ( AFWS) of all then-operating plants. The results obtaiggqforoperatingWestinghouse-designedplantswerepresentedinNUREG-0611.D1 At that time, the objective was to compare AFWS designs; accord-ingly, generic failure probabilities were used in the analysis, rather than plant-specific data. Some of these generic data were presented in NUREG-0611.

The probability that the AFWS would fail to perform its mission on demand was estimated for three initiating events:

(a) loss of sein feedwater (LMFW) without loss of offsite power; (b) loss of main feedwater associated with loss of offsite power (LOOP);

(c) loss of mair feedwater associated with loss of offsite and onsite AC (LOAC).

Since.then, each applicant for an operating license has been required (2) to submit a reliability analysis of the plant's AFWS, carried out in a manner similar to that employed in the NUREG-0611 Itudy. A quantitative criterion for AFWS reliability has been defined by t'-e NIJC)in the current Standard Review Plan (SRP) for Auxiliary Feedwater Systems \4:

... An acceptable AFWS should have an unreliability in the range of 10-4 to 10-5 per demand based or, an analysis using methods and data presented in NUREG-0611 and NUREG-0635. Compensating factors such as other methods of a'ccomplishing the safety functions of the AFWS or other reliaale methods for cooling the reactor core during abnormal conditions may be considered to justify a larger unavailability of the AFWS."

It'should be noted that because of the differences between the applicant's system and the AFWS at most other plants, the applicant has chosen to call his system tha Emergency Feedwater System.

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2.0 SCOPE OF BNL REVIEW data, and scope of NUREG-0611, Appendix ItIII.11(/The BNLobjec-has two major review has been conducted tives:

(a) To evaluate the applicant's reliability analysis of the AFWS.

(b) To provide an independent assessment, to the extent practical, of.the AFWS unavailability.

Unavailability as used in this report has been defined as the "probabil-ity that the AFWS will not perform its mission on demand". The term unavail-ability is used interchangeably with unreliability. Specific goals of this re-view are then:

(a) To compare the applicant's AFWS to the operating plants studied in NUREG-0611 by following the methodology of the latter as closely as possible.

(b) To evaluate the applicant's AFWS with respect to the reliability goal set forth in SRP 10.4.9, i.e., that the AFWS has unreliability in the range of 10-4 to 10-5 per demand, usir.g the above methodology.

The NUREG-0611 methodology and the BNL review specifically exclude exter-nally caused common mode failures such as earthquakes, tornados, floods, etc.,

and internal failures caused by pipe ruptures.

On August 19, 1982, BNL was informed by the NRC that the applicant had proposed certain design changes which are not described in Reference 3. Such changes affect the use of the startup feed pump during loss of offsite power conditions, and also the capability to parform maintenance on valves in the em-ergency feedwater header and supply lines to the steam generators. There-fore, this report describes and refers to the proposed changes to provide a

' comparative assessment of both designs. The term " Proposed Desiga" as used in this report refers to the applicant's proposed changes as described in the August 19, 1982 telephone conversation, and also in the September 7,1982 let-ter f. am the applicant, which appears in Appendix D.

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l 3.0 MISSION SUCCESS CRITERIA As described in REF.3,-for each of the transient conditions a..alyzed, unreliability was defined as the probability of failure of the combined EFW and startup pump system to start and provide feedwater to at least two of the four steam generators prior to the time that the steam generators would boil dry fallowing a reactor trip from full power. The time required to boil away the water in the steam generators is determined by the initial mass of water con-tained in them at the time of trip and the amount of decay heat liberated from the core. For the Seabrook Station, this time would generally be in the range of 35 to 60 minutes following a trip from full power operation; therefore, 30 minutes was selected as a conservative mission time for this reliability study.

At the July 15, 1982 plant visit, it was stated that 200,000 gallons of water are required to be available for the design basis shutdown to Hot Shutdown conditions after maintaining the plant at Hot Standby for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The hot shutdown conditions are the pressure and temperature of the reactor coolant system at which the residual heat removal system may begin operation.

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4.0 SYSTEM DESCRIPTION I 4.1 Configuration and Overall_, Design The Seabrook Auxiliary Feedwater System, as the term is commonly applied to systems which are used for Startup, Hoc Standby and Hot Shutdown, consists of the single pump of the Startup Feed Pump System (SUFPS) and its associated components, all of which are non-safety class. The SUFPS is augmented by the two pumps of the Emergency Feedwater System (EFWS), both of which are safety-class. For the purposes of the analysis, the applicant has defined the com-bination of the SUFPS and the EFWS as the Auxiliary Feedwater System ( AFWS).

The AFWS is described 'in REF.3 as follows:

4.1.1 'Startup Feedpump System The elements of the startup feedpump (SUF) system at Seabrook are shown in Figure 2. The system consists of a single motor-driven pump, P-113, capable of supplying 1500 gpm at 3000 feet of head. The pump takes suction from the Condensate Storage Tank (CST) via the main condensate makeup line. The -

suction line between the pump and CST is equipped with three normally open man-ual isolation valves, V-152, V-143 and V-141 (see Figure 5). The discharge -,

headers from the pump attach to six other feedwater system headers, i.e., the '

main feedwater pump dischacge header, the high pressure feedwater heater outlet header, the condensate pump discharge header, the make-up header fiom the CST, the steam generator recirculation pump discharge header, and the EFW pump discharge header. The pump is also equipped with a recirculation line to the CST for pump protection and testing. Flow through the recirculation line is

,. controlled by a pressure-controlled throttling valve, PCV-4326, that senses pressure at the pump discharge.

With the exception of the main feedwater pump discharge header, the dis-charge from the startup pump is isolated from all feed system headers by at least one normally closed valve. The supply path to the main feedwater pump '

discharge header is normally open but is equipped with a manual gear-operated ~ '

valve, V-100, to allow isolation if necessary. Flow to the EFW pump discharge header is prevented during normal operation by two normally closed manual gear-operated isolation valves, V-163 and V-156.

BNI Comment: In a conference call on July 26, 1982, the applicant stated ILt V-156 would be locked closed with the key under administrative con-trol. V-163 would not be locked closed. Thus, under a LOOP condition, the operator would be required to open a locked-closed valve, V-156, if the EFW header is to be used.

During startup, lubrication of the SUF pump is provided by-a motor-driven L auxiliary lube oil pump, P-161. Operation of the auxiliary lube oil pump is controlled by SUF pump lube oil pressure. When the SUF pump is in the AUTO control mode, startup of the lube oil pump will be followed by start of the SUF pump when sufficient oil pressure is established. Once started, a shaft-driven lubs oil pump located on the SUF pump supplies lubrication and the auxiliary 4

1 lube oil pump is s' cpped. Should the shaft-driven pump fail, the auxiliary oil pump will automatically restart.

in its normal operating mode the SUF pump will start automatically on a trip of both main feed pumps (LMFW) unless a safety injection or high-high steam generator level signal also occurs.

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BNL Comment: At the July 15, 1982 plant visit, the applicant stated that

,',. the basic dcsign philosophy of the SUFPS is tnat the system is used for all normal plant startups and shutdowns and also for most, if not all, LMFW transients. Therefore, the EFWS would not be automatically activated for a LMFW transient unless a safety injection, a low-low steam generator level, or a loss of offsite power (LOOP) signal also occurs, The appli-cant considers LMFW to be a part of normal plant operating conditions.

4.1.2 Emergency Feedwater System The EFWS 1s a standby system which would not be operated during normal plant operation except in case of a loss of the SUFPS during a startup or a shutdown or after a LMFW. The EFWS is automatically actuated upon an Engi-neered Safety Feature (ESF) actuation signal, i.e., a loss of offsite power (1.00P), low-low level in any steam generator, or any safety injection signal.

The system is described in REF.3 as follows:

A schematic of the EFWS at Seabrook is shown in Figure 3. The system con-sists of two pumps, each supplied by individua' suction lines from the CST.

Each pump has a design flow of 710 gpm at a head of 3050 feet and is capable of providing N11 cooling of the Reactor Coolant System in emergency situations.

One pump, P-37B, is driven by an AC motor which is powered by one of the 4160V plant emergency buses. The second pump, P-37A, is steam-turbine driven with steam being supplied from either of two steam generators. Take-off points for the turbine steam supply lines are upstream of the main steam isolation valves, thereby ensuring motive power to the turbine even in the event of steam line isolation. Both pumps are attached to a common return to the CST which is used for pump testing. This return line is isolated during normal operation.

During operation of the EFW r.ystem, both pumps discharge into a common header, which in turn supplies four individual supply lines to each of the four steam generator main feed lines. Each emergency feed line joins its associated main feedwater header downstream of the feedwater isolation valve and outside of containment.

The emergency feedwater supply lines are each equipped with two motor-operated flow isolation valves and a flow limiting venturi. The valves are normally open and are designed to fail "as-is" on loss of power. The valve positions are set such that they assure a minimum of 235 gpm to each steam generator during normal operation with both EFW pumps running. The control systems for the valves are designed to isolate an emergency feed supply line if f flow in the line exceeds a pre-set high flow value. This feature prevents di-5

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version of EfW flow following a line break in any steam generator. A single flow orifice located between the isolation valves in each line provides dif-ferenti al pressure information to the Control equipment for flow measurement.

Two separate flow transmitters are used to provide independent high flow isola-tion signals to each of tne isolation valves. The flow transmitters, control equipment, and motor-operators for the valves up.etream of the flow elements are powerer; by train B emergency electrical buses, while those downstream of the flow elements are powered by train A buses.

Assuming both EFW pumps are running, flow through any EFW line is ;imited to a maximum of 750 gpm by a flow limiting venturi also located between the isolation valves. This flow limitation provides runout protection for the EFW pumps in the event of depressurization of any steau generator. The venturis provide an added benefit in that a pipe break i any steam generator, along with fiilure of both isolation valves in the associated EFW line, will not cause i complete loss of cooling water to the remaining steam generators.

Each supply line is al'so equipped with a non-return valve which prevents the EFW system from being subjected to normal steam generator pressures when the EFW system is not in use.

The pump discharge headers and the common emergency feed ster header are equipped with a total of five isolation valves that are used to segregate vari-ous parts of the system for testing and maintenance activities. These valves are all manual, gear-operated valves that are locked open when the system is in its normal readiness state. 'Tha pump discharge headers are also equipped with check valves to prevent reverse flow through a pump during operation with the pump out of service. Flow diversion through the pump recirculation lines is prevented during normal operation by normally closed manual valves in each re-circulation header. The recirculation lines are also equipped with pressure reducing orifices that will limit flow should the manual valves be left open.

_B_NL Comment:. At the June 23, 1982 meeting at the NRC offices, the ap-plicant stated that the maximum flow through the recirculation lines is 220 gpm. There is no recirculation during the normal operation of the EFWS. . The recirculation lines are used for pump testing. The pump performance at 220 gpm is matched against the manufacturer's performance curve for the TDH at 220 gpm. Any deviations would be noted. There are no provisions for full flow pump testing.

Each pump suction line to the CST contains two manual isolation valves, one in the tank yard (V-154 and V-158), and one in the emergency feed pump building (V-155<and V-159). Both valves are normally open and locked in posi-tion. _

BNL Comment: At the July 15, 1982 plant visit, the applicant stated that the CST has a design capacity of 400,000 gallons and that under virtually all plant conditions, the normal level would be this amount. Of the 400,000 gallons, 200,000 gallons is reserved for the EFWS. The connec-6

tions for the EFWS are at the base of the CST. The SUFP normally draws water from a connection on the main condensate make-up line which is con-nected to the CST above the 200,000 gallon level. However, the SUFP can also draw water from the base of the CST through a nonnally closed, locked valve. This is not shown on Figure 3.

The steam supply lines for the turbine-driven EFW pump are shown in Figure

4. Steam can be supplied to the turbine from either steam generator "A" or "B". . Steam from either steam generator is supnlied to a common header through air-o'perated, fail-open valves, V-127 and V-128, that are actuated by an engi-neered safety feature (ESF) actuation signal, Both valves will open as a re-sult of a loss of offsite power, low-low level in any steam generator, or any safety injection signal, any one of which will also automatically start the motor-driven pump as well.

Each steam supply line is equipped with a check valve, V-94 and V-96, to prevent diversion of steam from the turbine in the event of a pipe break in one of the steam lines. The common supply header to the turbine-driven pump con-tains a normally open manual isolation valve, V-95, used during turbine main-tenance and a spring-loaded mechan. A trip v6ve that closes on turbine overspeed, V-129.

Oil cooling for the turbine-driven Etw pump bearings is provided by an oil cooler supplied directly from the discharge of the turbine-driven pump. The cooling flow is discharged to the common recirculation header.

BNL Comment: See Appendix D for additional design information regarding the air supply for the actuators of V-127 and V-128.

4.2 Component Design Classification The applicant has not specifically identified the design classification of each component, except to say that all components in the SUFPS are non-safety class up to the normally closed, locked manual valve V-156. The entire EFWS, including the CST, is safety-class. There are no makeup lines to the CST, either safety or non-safety class, 'which can provide flow at the same rate as one.of the EFWS pumps, i.e., 710 gpm.

4.3 Power Sources As described in REF.3, emergency electrical power for the EFW and SUFP systems is supplied from both 4160V emergency AC buses and both vital DC in-strument buses. Power for the motor-driven EFW pump is taken from emergency AC bus E-6 and diesel generator 18, while the SUF pump, via operatcr action de-scribed in Section 5.2, can be powered from emergency AC bus E-5 and diesel generator 1A. The auxiliary lube oil pamp used when starting the SUF pump is :

also supplied power by bus E-5 through buses E-52 and E-53. Control power for the motor-driven EFW train is taken entirely from vital DC instrument bus 118.

Control power for the steam-turbine admission valves is supplied from both ESF 7

L __ _ _ _ - - - - - - - - - - - - -

trains, one valve receiving control power from DC bus 11A in train A, and the other receiving power from DC bus llB in train B. There are no AC power de-pendencies in the turbine-driven EFW pump taain. Electrical power for the EFW isolatian valves also comes from the emergency buses, and train separation cri-teria are met for each FFW supply line.

BNL Comment: ~

The SUFPS is norm?'ly supplied power from the non-safety station electrical grid. During a LOOP condition, such power is not available and local manual transfer to the emergency sources is required to operate the SUFP.

4./ Ir.strumentation and Controls As described in REF.3, the control room operitor at Seabrook has avail-able a variety of instrumentation and controls that allow him to monitar and direct operation of Loth the emergency feedwater system and the startup feed pump. The :mportant equipment rc'ative to EFW and SUF system operation is listed below:

Instrumentation Location o Operating status lights for the Control room / remote motor-driven EFW pump, P-378 safe shutdown panel o Pnsition indication lights for Control room / remote both steam admission valves, V-127 safe shutd0wn panel

and V-128, to the turbine-driven EFW pump, P-37A o Suction and discharge pressures Control room / local for both EFW pumps o Flow indication for each emergency Control room / remote feedwater supply line safe shutdown panel o Three narrow-ranga and one wide- Control room / remote range level transmitter in each safe shutdown panel steam generator o Steam pressure in each steam Control room generator o Qual CST level transmitters Control room

Position indication is available at the remote shutdown pancl only for valve V-127.

BNL Comment: In Appernix D, it is stated that V128 will also have position indication on tbt remote panel.

8

l Alarms Location o Trip clarm for motor-driven EFW Control room pump, P-37B o Alarms indicating local operation Control room of either EFW pump o- Low suction pressure alarms for Control room both EFW pumps o Startup feed pump, P-113, trip alarm Control room o Startup feed pump pre-lube pump, Control room P-161,. running alarm o Low and low-low level alarms in Control room each steam generator

-o CST low level alarms Control room o SI actuation alarm Control room o Pump motor bearing and winding Control room tencarature alarms BNL Comment: It is not stated whether this applie's to the SUFP also.

o Emergency feed pump valves misaligned Control room BNL Comment: We assume this applies to V-65, V-67, V-71 and V-73 only, o SUF pump powered from bus E5 Control room Control s Location o Manual / auto controller for Control room /

motor-driven EFW pump switchgear room o Manual / auto controller for Control room / remote turbine-driven EFW pump safe shutdown panel

steam admission valve c Manuai/ auto controller for Control room / remote each EFW flow limiting valve safe shutdown panel o Manual / auto controller for Control room startup feed pump o Manual / auto controller for Control room startuo feed pump prelube pump

0nly stet.m-admission valve V121 can be controlled at the remote shutdown panel, BNL Comment: In Appendix D, it is stated that V128 will also be controllable

. from the remote panel.

9

. _ . -. . .- = .

Automatic Actuation Signals FFW Function o Safety injection signal Starts both EFW pumps o High flow to one S/G Close both EFW isola-tion valves in line with high flow o Low-low level in any steam generator Starts both EFW pumps o Loss-of-offsite power signal Starts both EFW pumps Automatic Actuation Signals FFW Function o Trip of both main . feed pumps Starts SUF-pump *'

o Low' bearing oil pressure at Starts SUF prelube SUF pump pump

This signal 'is prohibited if either a safety injection or steam generator-high-high level signal .is present.

i

[

10

f 5.0 EMERGENCY OPERATION 5.1: Loss of Main-Feedwater In the case of LMF'.l, the SUFP prelube pump receives an automatic signal to start, followed by the SUFP itself vaon tripping of both MFW pumps. The SUFP is normally aligned to the main fee uater nozzles upstream of the main feed-water isolation valves (MFVIVs). See. Figure 5, which is a simplified sketch showing the normal alignment of the.SUFPS to the MFW flowpaths to the steam generators.- The sketch was prepared by BNL based on FSAR Figure 10.4-4, Sh-1,

" Condensate System, P&I Diagram", and rigu-e 10.4-5, "Feedwater System, P&I-Diagram". There are no manual actions required for the SUFP to supply water to the steam generators unless the suction source line from the CST or the dis-charge to the MFW nozzles are unavailable. The connection from the SUFP to the EFW header is left closed. The SUFP initiation will be blocked if there is a concurrent' safety injection or high-high steam generator level signal.

If the SUFP should f ail to operate, the EFWS pumps will be automatically actuated upon low-low level in one or more .of the steam generators. The EFWS pumps and header are not normally used for this transient.

5.2 Loss of Offsite Power - REF.3 Desisn In_the f.ase of LOOP, the EFWS pumps are given an automatic signal to start and no additional actions are required for the pumps to supply water to the EFW header and into the steam generators.

The SUFP cannot supply. water through its usual flowpaths to the MFW supply lines to each steam generator because the flowpath connections are upstream of the MFWIVs. .The latter are air-operated-piston valves which close upon loss of air supply. The footnote ca page 29 of REF.3 implies that this will occur as the air- compressors lose power from the offsite sources. Also, the normal AC power sources are no longer available to the SUFP. - Providing AC power to the SUFP ~and aligning the pump tt, the EFW header requires several operator ac-tions which must ba 'caken outside of the Control Room, under the relatively

-adverse lighting conditions of a loss of offsite power. Such actions are de-scribed in REF.3 as follows:

In. order to provide power to the SUF pump from an emergency AC bus, an operator _ must manually " rack out" the SUF pump breaker, from bus 4 located in the non-essentid switchgear room, move it to the essential switchjear room, and manually " rack in" the breaker to' emergency bus E5. !!e must a:so change the bus transfer switch to the E5 bus position. The breaker has been equipped with built-in rollers to facilitate moving it'from room to room. In addition, the two switchgear rooms are adjacent to each other, minimizing the distance that the breaker must be moved.

Aligning of the SUF pump with the EFW systems also requires an operator (or operators) to change the position of three. manual ~ isolation valves. One of the valves (V-109) must be clo' sed to prevent possibir. flow diversion of the SUF 11 L ___

,~

pump discharge to the condensate tank via the SUFP recirculation line should

~

power be lost to the SUFP redrculation valve (PCV-4326). The remaining two valves .(V-156 and V-163) must be opener' to connect the SUF pump discharge head-er to the EFW system header. Valves V-109 and V-163 are located in the turbine

-hall. Valve V-156 is in the emergency feed pump room, and is locked in the

-closed position (see Section 4.1.1).

Proposed Design Changes On page 29 of REF.3, a footnote states that upon loss of offsite power, the MFWIVs automatically.close. In the July 26, 1982 conference call, BNL

_ questioned the correctness of this statement, since normally MFWIVs'only close upon a steamline or. feedwater line break. In this specific case, if. the MFWIVs do indeed close, the SOFP can no longer supply water to th?. steam generators through the main feedvater lines and the EFW header must be tsed. The entire analysis in REF.5 is based on the assumption that for LOOP, the MFWIVs close and therefore several me.nual valve operations have to be per Funded outside the Control Room, as previously described, to align the SUFP to the CFW header.

If the MFWIVs do not'close upon LOOP, BNL has assumed in its analysis that the~ actions required to use the SUFP are the following:

(a) Since the SUFP is already aligned to the.MFW supply lines, no valve closures are necessary. This includes V-109 in the recirculation line.

(b) Electrical ~ powar must be supplied from Diesel Generator 1A to the SUFP and the prelube pump.- ' Also, the instrument air compressor must be manually loaded on to the diesels.to allow operation of the MFW flow control valves which are used to control steam generator level from the.SUFP. Such electrical power is not available until after the diesels have completed their automatic sequencing of the essen-tial loads.

The'above matter is clarified in the supplement to this report.

5.3 Loss' of All AC Power - REF.3 Design In the case of LOAC, the system is basically reduced to a situation in which the .only available pump is P-37A, the turbine-driven pump of the EFWS.

Power to P-113, the SUFP, and to P.-37B of the EFWS, both motor-driven, will not be available. Since the steam supply valves, V-127 and V-128, are air-operated, fail-open valves that are actuated by a LOOP signal, and also since l

the air: supply is lost upon LOAC, the turbine-driven pump P-37A will auto-matically start. No 'further alignment or operator actions are required to

. start tha pump. It .is not stated in REF.3 if control of steam generator level could be accomplished under such conditions. -

12

... . ,~ .. . . . .

I Proposed Design The proposed design changes do not have a very significant effect on the reliabiiity analysis for the case of LOAC. The only change is in the addition of safety class accumulators for the air operatcrs of the steam suoply valves, V-127 and '/-128, as per Appendix D.

i k

13 m.... . .. _ . . .

6.0 TESTING According to REF.3, the procedure for testing pump P-37A or 37B is to close either manual isolation valve V-65 or V-71 and open manual valve V-67 or V-73 to recirculate emergency feedwater to the condensate storage tank. In the case of the turbine-driven pump, P-37A, only one of the steam supply valves, V-127 or V-128, is used for each test. Therefore, the testing of each valve and its control circuitry is alternated between one test and the next.

The SUFP can be tested in several ways. One method would be through the normally open manual valve V-100 in the line which connects the SUFP discharge to the discharge line of the main feedwater pumps. Another would be to close V-100 and recirculate water to the condensate storage tank tiirough PCV-43?6, which will open automatically on high pump discharge pressure. However, if the SUFP is needed, PCV-4326 will automatically close as pump discharge pressure decreases, thereby eliminating possible flow diversion. Neither of these test methods changes the configuration of the SUFP; therefore, no test outage was cpplied (by the applicant) to the SUFPS.

BNL Comment: As discussed in Section 5.2, the EFWS pumps are tested at a .

. flow rate of 220 gpm, but not at full flow. The test flow rate for the I SUFP has not been specified by the applicant, although it appears that the SUFP can be full flow tested directly into the steam generators by pumping through V-100.

l t

14

7.0 TECHNICAL 3PECIFICATIONS The proposed Seabrook Technical Specifications (5) for the Emergency Feedwater System are as follows:

LIMITING CONDITION FOR OPERATION At least two independent steam generator emergency feedwater pumps and as-sociated flow paths shall be OPERABLE with:

a. One motor-driven emergency feedwater pump capable of being powered from an emergency bus, and

b. One steam turbine-driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

=

APPLICABILITY: Power Operation, Startup, Hot Standby ACTION

a. With one emergency feedwater pump inoperable, restore the required emergency hedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two emergency feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 {

hours. Initiate corrective action to restore at least one emergercy feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS The emergency feedwater system shall oe demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying that motor driven pump develops a discharge pressure of greater than or equal to

psig at a flow of greater than or equal to

gpm. _

2. Verifying that the steam turbine driven pump develops a discharge pressure of greater than or equal to

psig at a flow of greater than or equal to

g greater than

psTg.pm when theofsecon3ary The provisions steam Specification supply 4.0.4** are pressure is not applicable for entry into the HOT STANDBY mode.

3. Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, realed, or otherwise secured in position is in it.s correct position.

These values were not specified at the writing of this report.

- See Specification 4.0.4 next page.

15

-b. At least once per 18 months during shutdown by:

1.- Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of a runout protection test signal.

2. Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an emergency feedwater actuation test sigr.al.

Speci fication 4.0.4. Entry into an OPERATIONAL MODE or other specified con-dition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval, or as otherwise specified.

The proposed Sea'> rook Technical Specifications (5) for the Condensate Storage Tank are as follows:

LIMITING CONDITION FOR OPERATION The condensate storage tank (CST) shall be OPERABLE with a contained water volume of at least 200,000. gallons of water.

APPLICABILITY: Power Operation, Startup, Hot Standby ACTION _

With the condensate storage tark inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the CST to

' OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ar.d in HOT SHUTDOWN within 'the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained watea volume is within its limits when the tank is the supply source for the emergency feedwater pumps.

' Surveillance requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components of the Emergency Feedwater System, including the Condensate Storage Tank, are detailed in REF.2.

.BNL Comment: The applicant's Proposed Technical Specifications comply

..with. Recommcndation GS-1 of NUREG-0611 that -the outage time for one AFW system flow train and essential instrumentation be limited to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and that the subsequent action time by which the plant must be in the HOT SHUTDOWN condition is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

It should be remarked that 4.7.1.3.1 implies that the EFW pumps have an alternate water source besides the CST itsel f. Such an alternate source is not mentioned anywhere in REF.3.

, 16

s.

8.0 ASSUMPTIONS The applicant has made the following assumptions in the pre, the ,

analysis. BNL' comments are provided both here and in Section 9.0 j necessary:

8.1 General Failure Data Previous analyses similar to the ora presented he e that have been con-ducted by other utilities owning plants designed by Westinghouse have generally had as an objective a comparative evaluation of the reliability of a specific emergency feedwater system with generic reliability analyses reported by the NRC staff in NUREG-0611. However, the October 30, 1981 letter to Seabrook (see Appendix C) specified a quantitative reliability goal for emergency feedwater system erformance. For that reason the component failure data presented in NUREG-0611 were considered to be too general to allow an accurate fault trae analysis of system unreliability to be performed. Therefore, it was decid tnat the best failure information available to date would be incorporated in this study. The following section presents that data and the sources from which they were taken.

Table B.1 (Appendix B) presents a compilation of data for various failure modes of different power plant components, both mechanical and electrical. The data were extracted from the following sources:

a) The Reactor Safety Study (WASH-1400),

b) GE-22A2589, Recommended Component Failure Rates, May 1974, c) IEEE-Std S00-1977, Nuclear Reliability Data Manual, and the following repeats from the Licensee Event Report (LER) evaluation pro-gram:

a) NUREG/CR-1205, Cata Summaries of LERs of Pumps

b)- NUREG/CR-1362, Data Summaries of LERs of Diesel Generators c) NURE6/CR-1740, Data Summaries of LERs of Selected Instrumentation and Control Components d) NUREG/CR-1363, Data Summaries of LERs of Valves The data vdlues obtained from the above references are presented in Table B.1 for each failure mode for whieb data from that reference was applicable.

To avoid ambiguity where multiple values are presented for a single failure mode, Table B.1 indicates the recommended value that was used in the fault tree analysis. -In the cases where multiple data values exist, engineering judgment was used to determine the most appropriate data based on similarity of the plant component, function and environment to the equipment represented by the data.

17

i. l

'The unreliabilities calculated exclude any consideration for the causes

"or: probabilities of the specified transient conditions, nor do they consider

~ external common mode failure. initiators such as earthquakes, floods, etc.

RNL Comment: These aspects of the data comply with the NUREG-0611 guidelines.

L In some instances the data presented in the referenced sources were either

, too general or the component data were obtained on like components having dis-

.similar functions. In particular, NUREG/CR-1205 presents component failure

data for pumps by generic classification, namely, running, alternating and ,

standby. However, review of the LERs revealed that sufficient data were avail-able to extract specific component data for mctor and turbine driven auxiliary Teedwater. pumps.

Similarly, the gener!c values presented in NGEG/CR-1363 for safety /re-

-lief ' valve failure rates were calculated using primary side components '(i.e.,

~

(pressurizer relief valves, pump relief. valves, etc.) only. The components of

. interest .in the- Seabrook fault tree were the steam generator safety / relief valves.n A limited amount of data existed in the LERs on secondary safety /re-lief valve failures. Also, it was noted that licensees do not always report' relief valveufailures since no credit is taken for them in accident analyses.

, -To compensate'for these facts, the valdes presented in Table B.1 for safety and relief valve premature opening were calculated using the information available in NUREG/CR-1363 and applying a factor of 5 to the safety valve .Nilure rate '

and a factor of 10 to the relief valve failure rate.

One further point should be mentioned 'as to the conservative bias built

-into some of the data. In particular, the failure rates.of the diegel gener-ator,!as taken from NUREG/CR-1362 for weekly testing, are 1.0 x 10-'/d for ;

. -the ' failure to start mode, and 6,0 'x 10-3/hr for the failure to run mode.

g These failure rates are calculated assuming that all plant diesel generators

, are tested weekly. -However, this does not account for the many starts-of the diesel generators which occur outside.of normal-testing periods. Therefore,

.the number of demands on the diesel -generators is underestimated while, con-

. versely, the number of failures reflects diesel generator failures which occur ,

during all phases of-operations. For those reasons, the failure rates from NUREG/CR-1362~ associated wit! weekly.. testing were considered to be most rep-

.resentative of the diesel failure frequencies to be expected at the Seabrook station. '

BNL Comment: While the applicant's arguments for developing independent data may be very legitimate, the scope of the BNL' review is to assess the

j. applicant's design using NUREG-0611 data wherever possible. Therefore,

[ BNL has made no attempt to verify any of the data used by the applicant.

'P.2 Traatment of Time Dependent Failures Failure rates used in the fault tree analysis are either demand dependent or time dependent. Demand dependent failure rates are applied to static com-L f'

r 18 g

. . . . ----....-.-.___m .-...,.,-..--.~....-.-__-.-,,,.-----,%. - , - - . .-,,.-,m.. - ,

E ponents which are rc,uired to change position or state to perform their re-quired function. Examples are the auxiliary feedwater pumps which are required to start on demand and certain valves, such as the steam turbine inlet valves (V-127 and V-128), which are required to change sta1;e upon receipt of the ap-propriate actuation' signal.

Time dependent failures are characterized by the necessity of a component to aaintain condition, position or status in order to perform its required func-tion. Examples are the auxiliary feedwater pumps which must continue to run once started, valves which must maintain their position (e.g., remain open),

and electrical components which must maintain their statas (e.g., pump breakers do not trip) for the entire mission time prescribed for a particular transient.

Time dependent failures are also characteristic of components which are in a standby condition and whicn could fail prior to operation.

The unavailability of a time dependent component is calculated from the hourly failure rate and a mission time for operating components, or a testing interval for standby components. The time interval used is dependent on the testing frequency, the actuction circuitry employed, and the operational re-quirements of a component for the transient being considered. For example, consider the actuation circuit of the motor-driven emergency feedwater pump (P-378) shown in Figure 6. This pump can be started automatically on receipt of either a safety injection signal, a loss of cffsite power signal, or on a low-low steam generator dater level signal. It can also be started manually from the control room by the operator using manual / auto control station CS-4255-1. The Technical Specifications require that the motor-driven EFW pump be tested every month. During these tests the pump will be started manually from the control room using CS-4255-1. This procedure will also test the integrity of the control circuit fren CS-4255-1 to the pump. Therefore, for certain failure modes of the control circuits, the proper testing frequency would be calculated from the one month testing interval, i.e.,

t = [30 days x 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months / day]/2 = 360 hours0.00417 days 0.1 hours 5.952381e-4 weeks 1.3698e-4 months .

In comparison, the tests of EfW system components to actuate on an auto-matic signal will be performed only every 18 months. The unavailability of a component due to failure to receive an automatic actuation signal would there-fore be calculated on the basis of the following time interval:

t = [18 months x 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months / month]/2 = 6480 hours0.075 days 1.8 hours 0.0107 weeks 0.00247 months .

It is assumed that for the monthly test of the turbine-driven emergency feed pump only one of the two steam admission valves is opened and that these valves are used alternately from one test to the next. Therefore, the control circuitry to the steam inlet valves V-127 and V-128 would each be tested on a bi-monthly interval, and the unavailability of these valves due to failures of

he control system are calculated using the following interval:

t = [60 days x 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months / day]/2 = 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months .

19

The unavailability of components which are raquired to operate or main-tain condition are calculated using the mission time. In the study presented here the mission time is the time in which the steam generators would boil dry given ar insufficient supply of water from the emergency feedwater system.

The _ unavailability of each failure event used in the Seabrook fault tree analysis, defined using the criteria discussed above, is presented in Table B-2 (Appendix B).

BNL Comment: The applicant has performed a commendable action by extend-ing the scope of the analysis to consider time dependent failures. How-ever, the mission time which his been assumed is only the 30 minutes in which the steam generators would boil dry given an insufficient supply of water from the Emergency (or Auxiliary) Feedwater System. There appears to be no logical basis for assuming the boil dry time to be the proper mission time. The proper mission time should be the time interval from the actuation of the AFWS until the plant has achieved hot shutdown con-ditions (generally about.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ), or until offsite power is restored, t M cally at least 30 minutes. In any case, running failures were not considered in NUREG-0611 bec6use such failures are usually small when con-sidered within the relatively short mission time of AFWS required oper-ation. One exception is the running failure rate of diesel generator plants given in WASH-1400, 3 x 10-3/hr. The diesel generators are usu-ally required to operate for only 30 minutes to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the average dura-tion of a LOOP incident. For the above reasons, running or time dependent failures have not been considered in the BNL assessment using NUREG-0611 data.

8.3 Test and Maintenance Outages According to REF.3, the applicant's assumptions regarding test and maintenance outages are as follows:

In addition to a component being unable to accomplish its function due to mechanical or electrical faults, a component may be unable to respond to a sys-tem demand because that component is out of service due to maintenance or test-ing. Technical Specifications limit the time during which some components can be unavailable and the plant still maintained at full power conditions. At Seabrook_one such limit applies to the EFW system. In the event that an emer-gency feedwater pump is disabled, restoration must occur within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be placed in a Hot Standby condition. This 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limit is assumed to apply also to pump discharge isolation valves if they require servicing.

All other components within the emergency feedwater system at Seabrook are assumed to have no time restrictions in relation to plant operation. However, the assumption was made that combinations of components which disable more than one emergency feedwater supply line could not be taken out of service simul-taneously. No maintenance requirements were considered for manually operated valves within the emergency feedwater system since these valves are located in 20

low energy lines, and position changes, other than those required for testing, do not routinely occur between scheduled outages. Unavailabilities of these valves due to maintenance errors during scheduled outages have been considered and will be described in Section 8.4 (0perator Errors).

BNL Comment: Since technically the SUFP and its associated components are not part of the EFWS, the applicant's assumption that there are no time restrictions on all other components within the EFWS does not neces-sarily apply to the SUFPS. However, to exclude the SUFPS from any mainte-nance outage time limitation because it is not a safety class system ap-pears to give an unfair advantage to the Seabrook design when compared to plants which have three safety class auxiliary feedwater pumps. There-fore, BNL has assumed that the SUFPS will be subjected to the same time restrictions as the other AFW systems mentioned.

The applicant's assumptions concerning maintenance of manual valves whose positions would normally never change between scheduled outages is rea-sonable. There are no specific guidelines in NUREG-0611 regarding which types of valves should be assumed to undergo maintenance.

It should also be noted that the applicant is assuming maintenance on the Emergency Feed Flow Isolation Valves as shown in Table 2. In the July 26, 1982 conversation with the applicant, BNL questioned whether maintenance on the motor-operated isolation valves such as V-75, V-87, V-93 and V-81 (see Figure 3) could be performed without also closing one of the EFW header valves such as V-125, V-126 or V-127, as well as one of the EFW pump discharge valves, V-65 or V-71. Simultaneous closure of one of the discharge valves and one of the header valves dramatically increases the system unavailability during _ maintenance. In particular, if V-75 is to be maintair.ed, both the SUFP and TDP-37A would become unavailable due to the closure of V-65 and V-125, leaving only MDP-37A feeding three steam gen-erators.

In the. August 19, 1982 conversation, the NRC indicated that the applicant is now adding a manual isolation valve immediately upstream of V-75, V-87, V-93 and V-81. The manual valves will be designated V-75, V-87, V-93 and V-81, with new numbers assigned to the motor-operated valves. Al so, V-125 will be instrumented in accordance with Regulatory Guide 1.97. BNL inter-prets that as meaning that position indicated in the Control Room will be provided for that valve. The addition of the mantal isolation valves was confirmed in the applicant's letter (Appendix D), although the instrumenta-tion of V-125 was not.

Another point to note is that the EFW. feed ficw isolation valves are only required by the Technical Spccifications to be tested at least once every 18 months during shutdown. Therefore, it is not clear how it would be de-termined that maintenance on those valves is- required and whether it may even be reasonable to assume that no maintenance is performed on those valves during power operation. The applicant does assume maintenance on !

the manual isolatio. valves at the discharge of each EFW pump, i.e., V-65 21 L. __. - __- - - ____ __ _ _____ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

\

and V-71,~because those valves are operated every month to perform testing of the pumps. The maintenance is modeled on the fault trees such that no 3 adjacent valves are isolated at the same. tine. BNL aesumes that only maintenance acts of such a nature that isolation of adjacent valves is not required will be performed. In this regard, the function of the EFW head-er valves V-125,~ V-126 and V-127 should be clarified to determine if the valves will be closed at any time during operation at Power. Startup, Hot Standby or Hot-Shutdown.

In one other. situation, BNL has revised the fault trees to indicate that if maintenance on one of the steam supply valves, V-127 or V-128, is to be performed, then the other valve must also be closed (see Figure 4). It was deemed unreasonable to assume-that maintenance could be performed on active main steam lines without taking such actions. The net effect is that a maintenance act on one of. those valves causes the pump, TDP-37A, to be unavailable.

.The discussian en maintenance and test unavailabilities in REF.3 con-tinues as follows:

Maintenance unavailabilities were calculated from data presented in NUREG/

CR-1635, Nuclear Plant Reliability Data System 1979 Annual Reports of Cumula-tive System and Component. Reliability. This source presents average restora-tion times for various. components and failure modes.- For those components whose outage times are limited by the Technical Specifications, the average restoration time was assumed equal to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if the average time specified by NUREG/CR-1635 was greater.than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The maintenance unavailability ~for a component was then calculated as follows:

Omaint = N x t/T where:- N = number of maintenance acts t = average component restoration time T = total component calendar hours.

Note that this calculation introduces additional conservatism because it as-sumes all maintenance acts are performed while the plant is op< rating at power.

A list of maintenance unavailabilities is presented in Table 2.

BNL-Comment: As previously noted, BNL has made no effort to verify this data.

Additional unavailabilities can Le assigned to emergency feedwater system components due to periodic testing. In particular, the Technical Specifica-tions require that the emergency feedwater pumps be started every month. Re-ferring to Figure 3, the procedure for testing pump P-37A or 37B is to close

.either manual isolation valve V-65 or V-71 and open manual valve V-67 or V-73 to recirculate emergency feedwater to the condensate storage tank.

22

l The startup feed pump can be tested in several ways (:;ee Figure 2). One iethod would be through the normally open manual valve V-100 in the line which connects the startup pump discharge to the discharge line of the main feedwater pumps. Another would be to close V-100 and recirculate water to the condensate storae tank through PCV-4326 which will open automatically on high pump dis-charge pressure. However, if the startup pump is needed, PCV-4326 will auto-matically close as pump discharge pressure decreases, thereby eliminating pos-sible flow diversion. Neither of these test methods change the configuration of the startup feed pump; therefore, no test outage was applied to the startup feed system. One exception to this assumption is discussed in the section on operator actions (Section 8.4).

BNL Comment: The exception referred to above is the supposed necessity of closing V-109 which isolates the SUFP recirculation line to the Conder. sate Storage Tank before the SUFP has been aligned to the EFW header. RFF.3 states that this local manual operator action is required to prevent diversion of flow from the SUFPS because a LOOP could result in PCV-4326 cpening due to loss of air supply. However, in the September 7,1982 let-ter (Appendix D), the applicant states that even if the recirculation line to the CST remains open, the maximum recirculation flow rate possible is insufficient to cause a reduction in the SUFP flow capacity to a level at which mission success is jeopardized. Since the SUFP normal flow capacity is 1500 gpm, while each EFW pump has a capacity of 710 gpm and only one of the EFW pumps is required to achieve a flow rate sufficient for mission success, BNL agrees that flow through the SUFP recirculation line cannot cause insufficient flow from the SUFP.

The test frequency for the emergency system is once per month, and the time interval of the test was assumed to be the average test time for pumps of 1.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months found in Table III 5-1 of WASH-1400.(6) The unavailability due to testing therefore is:

Qtest = 1.4/720 = 2 x 10-3 The test unavailabilities and the components to which they apply are shown in Table 2.

BNL Comment: The above test frequency complies with the requirements of NUREG-0611. . The applicant assumed, realistically, that all of the pump test unavailability appears only in the EFW pumps' discharge isolation valves V-65 and V-71, since those valves are closed to perform the test-ing. As such, only item 6(d) in Table _2 has any contribution due to test-ing shown. Note that the SUFP can be tested by pumping directly into the MFW ficw nozzles during power operation so that its test unavailability is assumed to be zero. Also, the diesel generators have been assumed to be available during testing. BNL cannot verify this at this time.

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The emergency feed ficw isolation' valves have been assumed to have a zero test outage time because the Technical Specifications only require that they be tested at least once every 18 months during shutdown. The steam supply valves are normally closed and testing causes them to assume the open position which makes them available. Steam supply valve V-129 is the turbine overspeed protection valve. It can only logically be tested dur-ing testing of.TD Pump 37-A itself.

8.4 Operator Errors The following discussion of operator errors appears in REF.3 and re-fers to the previous SUFPS and EFWS design:

Operator errors can be divided into two basic types: 1) errors of com-mission, and 2) errors of omission. Errors of commission occur when the operator performs an action which terminates or reverses the normal operation or condition of a component. Examples would be the operator shutting off a running pump or changing the position of a valve.

Errors of omission occur when the operator fails to perform an action which would initiate component operation or place it in its proper operating condition, given that these actions have not oce.urred automatically. Errors of omission also occur when the operator is the prime mover causing a system to function. such as in the proper alignment of the startup feedwater systm to provide oackup emergency feedwater flow. This type of error also includes failure to restore valves to their proper position following maintenance test acts.

A description of all operator actions used in the fault tree analysis and their associated unavailabilities are shown in Table 3. The guidelines of NUREG/CR-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, and NUREG-0611 were used in formulating the unavail-abilities.

As a general rule, errors of commission are assigned a 1 x 10 , and errors of omission a probability of 1 x 10-3. probability of These proba-l bilities age adjusted for abnormal circumstances. For instance, a probability i of 1 x 10- would normally be assigned both to errors of omission by the i operator for actions which can be performed from the Control Room, and to main-tenance restoration acts. However, if the operation must be performed locally (outside of the Control Room) or under potentially adverse conditions, the i failure probability is increased accordingly.

Except for automatic actuation of the lube oil pump (P-161), the startup feedwater system requires manual operation outside of the Control Room for j alig, ment to the emergency feedwater sysem.

B_NL Comment: The analysis of REf.3 assumed for conservatism that the

. SUFP must be aligned to the EFW header for a LMFW transient, although this was never true even under the previous design.

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In the event of Loss of Station Power, the operator must manually trans- -

fer the startup pump breaker from Bus 4 in the non-essential switchgear room to Bus E5 in the essential switchgear room, change the bus transfer switch to the \

EE. bus position, open discharge isolation valve V-156 in the emergency feed pump building, open discharge isolation V-163 and close condensate storage tank recirculation line isolation valve V-109 in the turbine hall. This last action (closing V-109) is necessary to prevent a diversion of flow from the startup system because a loss of station power could result in PCV-4326 opening due to loss of air. The o sumedtobe1x10geratorfailureratesforthefirstthreeactionsareas-

/ demand because these actions, even though assumed to be i covered by energency procedures, may include multiple steps and must be done at different locations. In contrast, the failure probability assigned to the closing of V-109 is assumed to be only 1 x 10-3 Since both V-163 and V-109 are located in the same vicinity, it was assumed that a single operator would be assigr J the task of changing the position of both valves. Therefore, the failure to complete both actions will be dominated by the failure to perform the first, and the failure probability for the second action is more appropri-ately represented by the standard failure rate for errors of omission. Thus, the total failure probability for completing both actions is 1.1 x 10 ',. ,

The leading of the SUFPS on to an emergency bus represents multiple opera-tions (viz., starting the prelube oil pump, moving the pump circuit breaker to the essential switchgear room, starting the SUF pump, etc.). In this case, however, all the concrols necessary to start both the SUF prelube pump and the SUF pump are available on the main control board. The only actions required outside the control rcom are moving of the pump breaker and changing of the bus transfer switch as described earlier. This is likely to be done by one oper-ator following well defined procedures. Therefore, for the purpose of this study, it was judged that a single operator error event could adequately rep-resent failures. in the pump loading process.

In addition to using four operator errors that could result in failure of the 5dF pump system, special consideration was also given to the failure rates applied to thest actions. Even though it is assumed that specific emergency procedures and operator training will be used at Seabrook to ensure proper utilization of the SUF pump during emrgencies, the failure rates used in this study for these operator errors is signifiantly higher (.01/ demand) than would normally be expected for situations where well developed procedures are in place and special operater training is provided. Again, use of the higher values was felt to be necessary to reflect the disparity in locations where actions must be performed and because of the short time ( 30 minutes) avail-able for the actions to be completed.

BNL Cominent: As discussed previously, it is not necessary to close V-109, the SUFP recirculation line isolation valve, for either the previous or proposed designs. BNL assumes that more than one operator would be re-quired to perform the valve manipulations because V 136 is located in the EFW pump area, which is a considerable distance from the SUFF, located in 25

l L

the Turbino Building. In the letter in Appendix D, the applicant states +

that V-156 will be relocated outside of the EFW pump room at ea, but not necessarily significantly closer to the SUFP. It is difficult to verify the correctness of the values for human errors assumed in Table 3 on the basis of NUREG-0611 criteria alone. The operator actions required are far more complex and demanding than those described in Table II!-2 of NUREG-0611 (Table B-3 of this report).

In the proposed design which with raspect to the use of the SUFP is not really a design change but a correction of the assum;tions made in REF.3, it is only necessary to align the SUFP to the EFW header if the MFW flow paths are not available. This will be true for both LMFW and LOOP. For both of- these cases, feedwater flow can now be provided through two sepa-rate flow paths to each steam gene:ator. This is discussed in more detail in Section 5.0.

l l

l 26

9.0 RELIABILITY ANALYSIS t'

9.1 Qualitative Aspects 9.1.1 Mode of System Initi tior.

1. LMFW - The SUFPS is automatically initiated upon trip of both MFW pumps, as described in Section 5.1. Should the SUfPS fail to initiate, the EFWS is automatically initiated upon low-low steam generator level. The SUFP and the EFW pumps can also be manually actuated from the Control Room. Therefore, the applicant complies with Recommendation GL-1 of NUREG-0611 that the AFW system flow be automatically initiated using safety grade equipment and that manual start serve as a backup to automatic AFW initiation.

2. LOOP - Only the EFWS pumps are automatically initiated. The SUFPS can only be initiated after completion of the manual actions previously described.

However, the applicant still complies with Recommendation GL-1 mentioned above.

3. LOAC - The TD pump 37-A is automatically initiated upon loss of power to its air-operated steam admission valves. Since it is aligned to the CST, it is also capable of providing the required AFW flow for at least two hours in-dependent of any AC. power source. Therefore, the applicant complies with Recommendation GL-3 of NUREG-0611, 9.1.2 System Control Following Initiation

1. LMFW - Only the SUFP is normally operating. Steam generator level is maintained by modulating the air-operated flow control valves on the MFW feed lines to the steam generators. REF.3 does not state whether there are provisions for automatic level control, but it is assumed that control can be performed manually from the Control Room.

If, for some reason, the normal water supply to the SUFP, which is above the 200,000 gallon level in the CST, is not avaliable, local manual operator actions can be taken to align the SUFP to the base of the CST by opening locked-closed valve V-142. It can also be manually aligned to the condenser hot well. However, normally there are ao manual or automatic actior.s required to maintain flow from the CST.

2. LOOP - The EFW pumps are automatically initiated so that steam Senerator level control is now maintained by modulating the reaundant AC motor-cperated flow control valves on the EFW feed lines to eact steam generator. If the SUFP is also operating, it will normally feed through the MFW feed lines. If one of the instrument air compressors can also be connected to the diesel generators, then the air-operated MFW flow control valves could be utilized to control steam generator level. See the supplement to this report for an extended discussion of this subject.

The EFW pumps are aligned only to the base of the CST, and no further manipulations are necessary to maintain the suction source. The SUFP suction source control is the same as described for LMFW above.

27 a _ - A

3. LOAC - Only the TD pump P-37A will be operating. No EFW flow control is possible because of loss of power to the AC motor-operated flow control valves.

As in the LOOP case, the pump is aligned to the base of the CST and no further manipulations are necessary, or possible, nor is AC power required, to maintain the suction source.

According to the listing of Instrumentation and Controls in Section 4.4, the CST has redundant level transmitters and low level alarms which are mon-itored in the Control Room. Therefore, the applicant appears to comply with Additional Short Term Recommendation I of NUREG-0611 that the licensee should provide redundant level indication and low level alarms in the Control Room for the AFW system primary water supply to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

The recommendation also states that the low level alarm should allow at least 20 minutes for operator action, assuming that the largest capacity AFW pump is operating. In the Seabrook design, the CST is a safety class component with sufficient capacity (total 400,000 gallons with 200,000 gallons dedicated to the EFW pumps) to supply water to the SUFP or the EFW pumps to cool the re-actor to the Hot Shutdown condition. REF.3 does not state what sources or flow rates are available to the CST, but this subject is discussed in Section 9.1.4.

BNL cannot determine whether any additional operator actions can or should be taken within the 20 minute period. REF.3 also does not state the time available to the operator upon receipt of the low level alarm.

At the July 15, 1982 plant visit, the NRC staff members discussed with the applicant whether the redundant level transmitters and alarms for the CST are safety grade. REF.3 does not state whether there are redundant level transmitters and low level alarms for the upper half of the CST from which the "UFP normally draws suction. The entire subject of the CST instrumentation is currently under review by the NRC staff.

One of :hr tYsign changes identified in the letter given in Appendix D is the in-stallatise of EFW pur.p minimum recirculation lines. The discharge of one pump will be connected to the suction line of the other pump. This change will minimize the possibility of pump damage if a pump should be actuated with its discharge valve closed. The capacity of the minimum recirculation lines is not stated in Appendix D, but it is assumed that flow to the steam generators cannot be degraded by the recirculation lines.

There appears to be no positiun indication in the Control Roer;. for any of the suction valves V-154,~7-155, V-158 and V-159. Ac.ording to FSAR Fig.

10.4-4 (Sh.1) for the Condensete System, V-154 and V-158 which are adjacent to the CST are locked open. However, FSAR Fig. 6.8-1 for the Emergency Feedwater System does not indicate that the valves adjacent to the EFW pumps, V-155 and V-159, are locked open.

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9.1.3 Effects of Test and Maintenance Activities See Section 8.3 for a detailed discussion of this subject.

9.1.4 Availability of Alternate Water Supplies In the Seabrook design, the EFW pumps are normally aligned to the base of the CST, as discussed in Section 9.1.2. For the reasons aentioned in that sec-tion, there are no design basis alternate EFW supplies to the CST. However, a limited make-up flow rate is available from the Demineralized Water System. At the June 23, 1982 meeting at NRC Headquarters, the applicant stated that some other means could be found to supply a limited make-up flow rate to the CST, such as by use of the Fire Protection System. Hcwever, no makeup exists that can supply water at the flow rate of one of the EFW pumps.

Since the SUFPS is the applicant's primary means of supplying feedwater to the steam generators in case of LMFW, consideration should be given to its sources of supply. As stated in Section 9.1.2, the SUFPS is normally aligned to the upper half of the 400,000 gallon CST. It can also be aligned to the base of the CST by opening locked-closed valve V-142. If the main condenser is not under vacuum, the SUFP can also take suction from the Condenser Hot Well, according to FSAR Section 10.4.12.2.

Therefore, the applicant should provide further information i'n the form of emergency procedures' governing the transfer to alternate water sources ;or the SUFPS. On the basis of the information provided in REF.3, the applicant has not provided adequate emergency procedures for transferring to alternate sources of AFW supply, as described in Recommendation GS-4 of NUREG-0611. The procedures should include criteria to inform the operators when, and in what order, the transfer to alternate water sources should take place.

9.1.5 Adequacy and Separation of Power Sources REF.3 states that a qualitative review of the engineering drawings showed that electrical power sources were found to be sufficiently separated and diverse to prevent dependencies due to power failurcs. It also states that separation of the SUF and EFW pumps provides protection from electrical train common-cause failures due to localized grounding of power supplies.

See also the discussion in Section 4.3 of this report on Power Sources.

9.1.6 Common Mode Failures The following discussion has been taken from REF.3:

The cut-set results from the reliability analysis were also used in con-junction with the system engineering drawings to condect a qualitative review of potential common-cause failure modes of the Seabrook AFW system. During the review, consideration was given to potential dependencies resulting from common location, environment, human interactions, and support equipment for all three 29

AFW pump trains. As a result of this investigation, two potential susceptibil-ities were identified.

The first of the common-cause susceptibilities results from a combined incation and environmental dependency. Because both en gency feedwater pumps are located in the same pump room, conditions which re.att in an extreme envi-ronment in that room can adversely affect both pumps. fa obvious potential source for such an environmental upset are failures associated with the steam turbine-driven pump that cause steam to escape into the pump room. The resul-tant high temperatures and high humidity might result in consequentiai failure of the motor-driven pump. Failure of the two EFW pumps alone are not suffi-cient to fail the AFW system because of

he availability of the SUF pump which is locatcd in the turbine building. However, for the SUF pump to be able to stpply cooling to the steam generators via the EFW piping requires that man-ual isolation valve V-156 be opened. This valve is located in the emergency feed pump room and would be inaccessible in the event of extreme er.vironments in the room.

...[1]n many situations it will not be necessary to align the SUF pump with the EFW system in order to use it for plant cooling. Only in circumrtances where the normal flow path through the main feedwater lines is unavaiiable will this be required. Therefore, even should both EFW pumps fail due to a pump room steam leak and valve V-156 also be inaccessible, the ability to cool the plant will still exist in most circumstances.

Most probable causes (or steam leaks in the pump room of sufficient sever-ity to cause environmental problems are associated with cracks in the pump turbine casing or breaks in the steam supply lines to the turbine. The most likely cause of the main feedwater lir.es being unavailable for supplying cool-ing to the steam generators is a safety injection signal which will cause closure of the main feed isolation valves. Ine probaDility of simultaneous oC-currence of these cvents is small compared to the overall system unavailability predicted by the fault tree analyses. Therefore, this common-cause ruscepti-bility has a negligible effect on the system.

A common cause failure potential often present in systems that incorpcr-ate automatic feedwater line isolation features is the possibility of a faulty calibration procedure causing all isolation setpoints to be improperly ad-justed. As a result, inadvertent closure of all isolation valves can occur

'turing system startup or following system flow perturbations. The design of the Seabrook system avoids this problem by incorporating control logic to in-hibit isolation of more than a single EFW line. Signals denoting the closure of EFW isolation valve in any EFW line will inhibit closure of additional valves in the remaining lines.

No other common-cause susceptibilities were identified which might ad-vernely impact the Seabrook AFW desisn. Electrical power sources were found to be suit'ciently separated and diverse to prevent dependencies due to power failures. With one exception *, all powered valves critical to system oper-ation are of a fail-safe design such that loss of air or loss of powar events

Recirculation valve PCV-4326 on the SUF pump discharge.

30

do not pose threats to system function. With the exception of the location de-pendency noted above, separation of the SUF pump from the EFW pumps provides protection from location depandent effects such as vibration, grit, tempera-ture, impact, explosions, etc. Separation of the SUF and EFW pumps also pro-vides protection from electrical train common-cause failures due to localized grounding of power supplies.

BNL Comn.ent: As noted previously in Section 5.2 of this report, if re-circulation valve PCV-4326 on the SUFP discharge te the CST should fail open, or if manual valve V-109 is not closed by operator action, the re-circulation flow rate to the CST is not large enough to cause insufficient ficw from the SUFP. Also, V-156 will be located outside of the EFW Pump Room area. See Appendix D.

9.1.7 Single Point Failures No single point failures have been found by BNL during the course of re-view of REF.3 or during the plant visit on July 15, 1982.

9.1.8 Adequacy of Emergency Procedures Since the Labrook EFWS is still undergoing design ctanges at the writing of this report, the applicant was not able to provide adequate emergency proce-dures. Such pracedures should be provided in the future.

9.2 Quantitative Aspects 9.2.1 Applicant's use pf N'"'-Suggested Methodology and Data 9.2.1.1 Fault Tree Construction and Evaluation According to REF.3, the applicant's fault trees were developed in the following manner:

The fault tree model used for this study was developed from an existing fault tree created several years ago as part of a " mini WASH-1400" review of the Seabrook Station. In its original form the tree considered the two-pump emergency feedwater system, but did not 'nclude medeling of the startup feed pump. In the initial phases of this study the old fault tree model was re-viewed for accuracy and revised as necessary to properly reflect the current EFW system design at Seabrook. In addition, the logic necessary to model the impact of the SUF system on AFW system reliability was incorporated into the trees. As s result, the fault tree now models the entire "three pump" AFW sys-tem as it currently exists in the Seabrook design. In essence, failures of all components shown in Figures 2 through 4 are now considered by the fault tree model. A logic diagram of the complete fault tree is provided in Appendix A.

31

In addition to component failures, the fault tree also includes logic to consider the effects of failures in interfacing systems on AFW system relia-bility. Examples are failures of the electrical power sources for the EFW and SUF pump and controls, failures of reactor protection system actuation signals, failures at piping interfaces with the main feedwater/corJensate system, fail-ures in the steam generators, and errors by plant personnel while maintaining and operating the system.

BNL Comment: The applicant's fault trees are quite comprehensive and include areas not required by NUREG-0611, e.g., pipe ruptures and running tailures. The human errors have been broken down into the following types as shown on Table A.3, Fault Codes:

0A - Operator Fails to Open/De-energize / Disengage OB - Operator Fails to Close/ Energize / Engage OC - Operator Inadvertently Opens /De-energizes / Disengages /

Leaves Open OD - Operator Inadvertently Closes / Energizes / Engages /

Leaves Closed OE - Operator Fails to Start OG - Operator Fails to leave Running Unfortunately, the fault trees were prepared based on the assumption that the SUFP would only be used by aligning it to tne EFW header for both LMFW and LOOP. The normal fiow path of the SUFP through the MFW feedlines was not modeled into the trees.

In any case, the applicant has exceeded the requirements of NUREG-0611, Figs. III-2 and III-3, concerning the construction and content of the fault trees, The applicant provides the following additional discussion concerning the fault trees in REF.3:

In a general sense, a loss of main feedwater event is the transient for which the auxiliary feedwater system is intended to provide protection. There-fore, the reliability of the AFW system for the LMFW transient can be viewed as a reference against which reliability calculations for the other transients may be compared. The fault tree described in the previous sections and presented in Appendix A was designed specifically for the LMFW event. Evaluations of the other transients were made by modifying this baseline fault tree as described later.

Before discussing the modifications necessary to model these other events, one point of conservatism regarding LMfW events that has been included into the 32

m 4

4 f ault ' tree should be re[terated. ...[F]ault tree modeling of the effects of the startup feed pump on AFW system reliability assumed in all cases that the

. SUF pump.was successful. only when supplying the emergency feedwater header. As a result, all the operator actions required to achieve this goal must be .

successful. This includes manually starting the SUF pump and, if necessary,

y. loading-it on an emergency bus. It also requires the necessary actions to -

. change the. positions of the three valves in the startup pump discharge and cross-tie headers... In many LMFW. transients, however, none of these actions will . be requi red. In ~those transients which result in a trip of the main feed-

. pumps but do not result in either a high steam c 'nerator level or a safety injection signal, the SUF pump will be automatically started and will deliver flow to the steam generators by way of the main feed lines. Therefore, no 'i operator actions are necessary to receive the benefit of cooling from the SUF pump. .Similarly, if these same transients are' accompanied by a loss of the normal. SUF pump power. source, only the actions to load the SUF pump on the emergency bus andLisolate its recirculation line are necessary. . Flow can still be provided to the ' steam generators through the main feedwater lines without additional valve manipulations.* Thus, the fault tree model, by requiring

.the SUF pucp to < supply cooling water via the EFW headers in all cases, provides 4

conservative estimates of reliability for these transients where main feedwater flow paths are still available and in which a safety injection or high steam generator level signal is not generated.

. The LMFW/LOSP transients, impact the-EFW systen in only one way. They eliminate the redundant electrical power sources for both the motor-driven EFW pump and the motor-driven SUF pump. As a-result,.the reliability of both punps is reduced because all single point failures causing loss of the emergency bus

- supplying the pump will also result in loss of the pump. In the case of the startup pump, the necessity of an cperator-action to load the pump to the emer-gency bus is also introduced into the system.

Modeling the loss of offsite power in the fault tree was done by cor. vert-Ling- gates EP 21 -(pg. A-43 of Appendix A) and SUP 21 (Pg. A-38 of Appendix. A) to

~

AND gates, converting gates EPE6 (pg. A-43 of Appendix A) and SUP 19 (pg. A-38 of' Appendix- A), and MOD 4 (pg. A-45 to A-48 of Appendix A) to OR gates, and in-putting _an LOSP. frequency of 0. This has the same effect as inputting an LOSP frequency of ~1.0 in the reference tree hut greatly reduces the computer calcu-

-lations" required to. evaluate the tree. All cut-sets and failure ~ probabilities determined for the modified tree will be conditional on the LOSP event even

, though the code cut-set output will not include specific indication of that fact.

4

Note that if the cause of the power loss is a loss of offsite power, the main feedwater -lines will not be open because of closure of the main feedwater isolation valves on loss of power.

BNL Comment: This is the-footnote which was questioned by BNL and led to

- the system design clarification concerning the use of the SUFPS.

33 u: ,. - _ _ ._ _ _ _ _ _ . _ ~ . . . _ _ _ _ _ _ . - _ _ _

l l

l The total' loss of AC power events ha've a much more drastic effect on AFW system reliability. In essence the system is reduced to a single pump system because both motor-driven pumps become unavailable. Thus, all single point failures disabling the turbine-driven EFW pump result in loss of system func-

. tion.

For the total. loss of AC power events, the fault tree modifications were also more extensive. All tree structures below gates AF127, SUP1, and M004 *

(pgs. A-30, A-41, and A-45 of Appendix A) were eliminated. The net effect is the same as inputting frequency values of 1.0 for both the LOSP event and fail-ure of both diesel generators in that both motor-driven pumps are eliminated from the system. Again code results are conditional on these failures although the conditionality is not reflected specifically in the output.

-BNL Comment: The above methods of modeling LOOP and LOAC are functionally correct and acceptable.

i 9.2.1.2 Failure Data

-' .This subject has been substantially: iscussed in Section 8.0. In sum-

. mary, the applicant has developed ~his own data base resulting in failure proba-bilities generally lower-than those assigned by NUREG-06tl. The one case in which the applicant utilized the exact NUREG-0611 specified data is for test outage time of the EFW pumps.

One significant case where the applicant's data is higher than the NUREG-is for random failure of TDP-37A. The NUREG-0611 data specified 3 x 10-061)/

has assumed demand for 8.4 x 10 / demand. FailuS'

Stt f

t"rbi"*-dr*" pump -while the applicant The applicant's data base is shown in Table B.1.

9.2.2 Applicant's Results

9.2.2.1_ System Unavailabilities The applicant's results as described in 'REF.3 are as follows:

. Computer Codes All qualitative cut-set analyses' and numerical evaluations of unreliability made using the Seabrook AFW system fault tree model were performed by the WAMBAM (7) and WAMCUT (8) computer codes. Versions of these codes' were obtain-ed from the Electric Pcwer Research Institute (EPRI) by PSNH and its service organization, Yankee Atomic Electric Co. (YAEC) for the purpose of conducting this study. Some modifications were required to the codes to reduce their p memory requirements during execution so that they could be run on the CDC-7600 computer at YAEC; however, the modifications only affected the size of the fault. tree that could be analyzed and not the numerical probability calcula-tions or cut-set evaluations performed by the code.

34 4 - , , ., , , , , , , -v-- er - - - . - , - - - - - -m m .

l f, ' Events Analyzed

.Three specific events were analyzed using the Seabrook fault tree. They were:

A loss of main feedwater transient with reactor trip (LMFW)

. o o A loss of main feedwater transient with coincident loss of offsite

,- . power (LMFW/LOSP)-

o : A loss of main- feedwater transient with coincident loss of'offsite -

power and both onsite emergency diesel-generators (LMFW/LOAC).

In all cases, successful operation of the AFW system required that at least two :

~

of-the four plant steam generators be supplied with cooling flow from the AFW o system.

Numerical Reliability Results JA' total of five' cases were analyzed with the Seabrook AFW system fault

' tree model. They were.the LMFW, LMFW/LOSP, and LMFW/LOAC events assuming all three pumps are part of the EFW system, and the LMFW and LMFW/LOSP events 1 -assuming only the two-train emergency feedwater system is used to provide steam

' generator cooling.1 The latter two cases were done to provide a reference for evaluating the ef fect of the SUF pump on overall system reliability. The re-

, sults-of the five cases.are shown in Table 4.

.It is clear:from these results that the Seabrook AFW system easily meets the NRC specified reliability goals when use of the SUF pump is considered for Lthe LMFW transient. Even with afcoincident loss of offsite power, the system 4 . exhibits an unreliability of better than 10-4/ demand. In terms of the re-sults published by the. NRC in. NUREG-0611 for other-Westinghouse plants, the Seabrook AFW system would' fall -into the high, high, and medium categories re-spectively for the LMFW, ^ LMFW/LOSP and LMFW/LOAC' transients.

i I

9.2.2.2 Dominant Failure' Modes and Conclusions i I

The applicant's dominant. failure modes and conclusions as described in REF.3 are as'follows:

Dominant Failures for Three Pump AFW System t

Dominant contributors to availability of the AFW system at Seabrook for

- the three loss of main feedwater/ loss of- power events are shown in Tables 5, 6 E and 7 and Figures 7 to 9. Events are ranked by the magnitude of their contri-bution to system unavailability. It should be noted.that no single' point-35

_ . . , , . , , , . - ,.m..._ _

-, , . , , , , , , ,,. ,mm._-.__-. _ _ - . - _ _ ~ . - ..

failures

were found in either the LMFW or LMFW/LOSP events that would disable the entire AFW system, although, as should be expected, a number of single failures will disable the turbine-driven pump train during a LMFW/LOAC event.

Applicant's Conclusions The results presented in this report lead to the following conclusions:

1. The Seabrook combined auxiliary feedwater system consisting of the two-train emergency feedwater system and the single-train startup feedwater system has an unreliability of 2.1 x.10-5 for a loss of main feedwater event and is well within the range of unreliability specified by the NRC staff in their October 30, 1981 letter to the Public Service Company of New Hampshire (see Appendix C).

2. The unreliability of the Seabrook combincd AFW system during combined loss of main'feedwater/ loss of offsite power events is 5.2 x 10-5, and for a combined loss of main feedwater/ loss of all AC power event is 2.1 x 10-2. These values compare favorably with analyses done for auxiliary feedwater systems at other plants of Westinghouse de-sign.

3. Major contributors to system unreliability generally relate to fail-ures of pumps and to maintenance errors causing pump trains to be inadvertently disabled.

4. No severe common-cause failure susceptibilities were identified for the Seabrook auxiliary feedwater system.

9.2.3 BNL Assessment 9.2.3.1 Fault Trees REF.3 Design The applicant's fault trees were checked for correctness and complete-ness. As noted previously, a detailed analysis of pipe and valve ruptures, in-cluding operator recovery from the effects of the ruptures, has been included on the trees. The trees as constructed do not rule out coincident test or maintenance of more than one pump, either both EFW pumps or the SUFP coincident with one or both EFW pumps. There are also no restrictions incorporated on test or maintenance on any of the valves in the feedlines to each of the steam

0ne single failure exists that will disable the AFW system under any cir-cumstance. That is a failure of the condensate storage tank such that no water is available to the suction of any of the pumps. The probability of such a failure was assumed negligible for the purposes of this study.

36

generators, either in conjunction with pump maintenance or such that 3 out of 4 of the feedlines are isolated due to maintenance. Therefore, the expected contribution within the calculated system unavailabilities due to coincident test or mait n enance of components in violation of the Technical Specifications can be expected to be significant. It should be noted that when using the WAMBAM code, the effects of coincident test or maintenance can not be re-adily identified since the code yields only a numerical result. No listing of specific cutsets is provided as in the WAMCUT code. However, it is only prac-tical to use WAMCUT tc identify the coincident test or maintenance contribu-tions at fairly high minimum probability cutoff points. As the cut off point

.is lowered, the number.of cutsets can run into the thousands and the computer time can exponentially increase with diminishing improvement in the final re-sults.

Taking the above factors into account, BNL evaluated the applicant's fault trees using NUREG-0611 data wherever applicable. In addition to the top event, AFW, the following subgates were evaluated:

1. AF91 (pg. A-31) - No Flow to Supply Header From TDP-37A.

2. AF127 (pg. A-32) - No Flow to Supply Header From MDP-378.

3. SUP1 (pg. A-41) - No Flow to Supply Header From SUFP P-113.

4 The effects of double and triple maintenance outages and also pipe rup-tures and electrical and control wiring faults not included in the NURE -0611 methodology was estimated by running the WAMBAM code to a cutoff for of 10 g0 The WAMCUT code was also used with thetopeventAgWandthesubgateslisted.

a cutoff of 10 , primarily to obtain a listing of the cutsets for the subgates AF91, AF127 and SUP1 so that a breakdown of the hardware f ailures ver-sus test or maintenance outages for each of the subgates could be obtained.

A modification was made to the trees to account for the fact that if main-tenance is performed on one of tha steam admission valves, V-127 or V-128, to TDP-37A, both valves will have to be closed.

Therefore, faults QV1XV12700 and QV1XV12800 were removed from the input of gates AF108 and AF114 (pgs. A-36 and-A-37) respectively and combined into a new OR gate designated AF100TM as an input to gate AF100 (pg. A-35) to logically model the fact that TDP-37A will be inoperable.

Proposed Design After receipt of the August 19, 1982 telephone call from the NRC, BNL de-termined that no new fault trees were required to obtain an accurate answer for the new design configuration. Fig. 5 is a simplified flow schematic of the 4

g 37

SUFP as it will normally be used for the LMFW and LOOP transients (i.e. the SUFP will be aligned to the MFW flow paths), based upon FSAR Fig.10.4-4 (Con-densate Syst. ,P&I Diagram) and Fig. 10.4-5 (Feedwater System P&I Diagram). The normal flowpath of the SUFP discharge is to the discharge of Steam Generator

. Feed Pump'P-32B. The Steam Generator Feed Pumps P-32A and P-32B are intercon-nected by a cross-tie. P-32A normally discharges to Heater E-26A. The heater can be by-passed through valve V8. Similarly P-32B normally discharges to Heater E-26B and the heater-can be by-passed through valve V19. The heater

discharges are then conn.ected to a common Main Feedwater header. The branch connections from the header to the steam generators each contain a motor-operated isolation valve, V28 for S.G. A. , V37 for S.G.B. , V46 for S.G.C. , and V55: for S.G.D. The branch lines are continued on Fig. 3.

As can be seen from Fig. 2, the SUFP is normally aligned to both of the heaters and then to all four steam generators. When considered together with the EFWS, there now exist two virtually independent systems for admitting AFW to the steam generators, i.e., the SUFPS and its normal MFW connections, and the EFWS and its header and branch connections. In the SUFPS, aside from failures in the SUFP itself or ir, the valves leading from the CST, all the cutsets which cause insufficient flow from the SUFPS are at least second-order.

For extmple, if V3 and V14, the heater inlet valves, were both inadvertantly closed due to human error or closed due to plugging, this would be a cause of LMFW if neither one of the heater by-pass valves, V8 and V19 were open. Since the plant is assumed to be at power operation prior to the three transients considered in this report, such a fault should be readily detectable by the operators. Only one of the by-pass valves would have to be opened to allow flow from the SUFPS. Similarly, the inadvertent closure or failure of any valve on the MFW branch connections to the steam generators should be readily detectable by the operators since the plant was at power operation prior to the fault. Valve faults would have to occur in at least 3 out of the 4 branch con-nections in order to cause insufficient flow from the SUFPS. All of the cut-sets discussed above are quantitatively insignificant when compared to the quantitative value of subgate SUP1, No Flow to Supply Header From SUFP P-113 so that No Flow to 3 Out of 4 Steam Generators From the SUFPS Through the MFW Flowpaths can be adequately represented by SUP1, if failure to open of V156 and V163 is omitted.

It should be noted that NUREG-0611 has no failure data for MFW heaters, either due to hardware or test or maintenance. Also, Table B.3 has no pro-visions for human errors in normally operating MFW systems but is oriented rather.to the typically standby nature of AFWS. For all of the above reasons, BNL has determined that additional fault trees for the MFW flowpaths are neither necessary nor practical.

1 38 l

l

l 9.2.3.2 Failure Data As noted previously, the NUREG-0611 data nas been utilized throughout the BNL assessment, to the extent possible, for both the REF.3 Design and the Proposed Design. Test and maintenance of manual valves was assumed to be zero in accordance with the applicant's contentions described in Section 8.3.

One area which should be discussed is the value assigned to operator error in failing to restore a locked manual valve to its proper gosition after test or maintenance. According to Table B.3 a value of 5 x 10- should be as-signed to Operator Inadvertently Leaves Correct Valve in Wrong Position if it has local walk-around and double check procedures associated with it. If it has neither,1 x 10-2 should be assigned. There is no distinction made be-tween locked and unlocked valves. However, the Technical Specifications re-quire surveillance of manual valves only if they are not locked into position.

Therefore, from Table B.3, a value of 1 x 10-2 should be assigned in this case. This obviously seems inconsistent with the intent of the fechnical Specifications. According to Table 3, Section 8.4, the applicant has assumed a value of 1 x 10-3 for such valves in the case of the operator failing to restore a valve to its normal position after maintenance. We find 1 x 10 3 to oe a reasonable assumption if 30 minutes recovery time is available for op-erator corrective actions, and have utilized this value in the BNL assessment.

In the case where recovery is not feasible, such as a pump starting with a suc-tion valve clqsed thereby causing damage to the pump, no recovery is assumed, i.e., 5 x 10-3 Concerning the operator actions required to manually conn the emergency power sources, the applicant has assumed 1 y 10 gct the SUFP to

' for MPB1P1610E, Operator Fails to Start the Startup Prelube Pump, P161, as discussed in Section 8.4. The failure to start the SU7P is represented by failure to start the Prelube Pump. We believe that 1 x 10-2 is not realistic given the complexity of the task under a moderately stressful situation with reduced station lighting as occurs in a LOOP. In the BNL assessment, a value of 3 x 10-2 has been assumed for this operator failure and this becomes a very significant contributo, to the subgate SUP1 "No Flow to Supply Header from SUFP."

In the area of. pump test outages, we assume that since the EFW pumps or the SUFP are already in operation during the testing and there is a 30 minute mission success time, the operators should be able to restore the pumps to l

their normal alignment to allow flow into the steam generators, i.e., outages l

due to testing have been assumed to be negligible.

I 9.2.3.3 System Unavailabilities i

! Both the WAMBAM and WAMCUT computer codes have been utilized as previously l ' described in Section 9.2.3.1. The results are given separately for each tran-sient:

39

1. LMFW a) REF.3 Design The BNL assessment for this event is 4.5 x 10-5 The method of. calculation is shown in Table 8.

b) Proposed Design See Table 9. The BNL assessment for this event is 1.96 x 10-5,

2. LGOP a) REF.3 Design The basic defference between this case and 1(a) is that random failure and maintenance outage of both Emergency Diesel Generators must be considered. The value 3 x 10-2 is added to the hardware failure of gates AF127 and SUP1 for random diesel failure while 6.4 x 13-3 is added to the maintenance failures of those two gates.

See Table 10.

The BNL assessment for this case is 1.8 x 10-4 b) Proposed Design This case is a combination of Case 1(b) and 2(a), above. See Table 11.

The BNL assessment for this case is 1.15 x 10-4

3. LOAC a) REF.3 Design Since this case is essentially a single pump situation, i.e.

TDP-37A, the top event can be approximated quite accurately by neg-lecting failures in the feedlines to the 4 steam generators. The expression for the top event is then:

Top Event = AF91 + V115

= 2.18 x 10-2 + 0.11 x 10-2 = 2.3 x 10-2 where VPl!i = 1 x 10-3 Operator Error 1 x 10-4 Plugging 1.1 x 10-3 40

b) Proposed Design This case is exactly the same as 3(a) above:

Top Event = 2.18 x 10-2 + 0.11 x 10-2 = 2.3 x 10-2 The results of all three cases are summarized and compared to the applicant's results in Table 1 (see Summary and Conclusions) and Table 12.

9.2.3.4 Dominant Failure Modes

1. LMFW a) REF.3 Design The dominant modes are random or maintenance failures of MDF-37B coupled with maintenance errors causing V125 to be closed. This agrees qualitatively with the applicant's results given in Tcble 5.

The WAMCUT results for this case are shown in Fig.12, Sh.1-2. If V125 is closed, flow from both the SUFP and TDP-37A is restricted to Steam Generator A only. Any failure in MDP-37B causes system failure.

The SUFP is connected to the offsite power sources and it receives an automatic initiation signal .

b) Proposed Design In this case, the SUFPS is no longer dependent on the EFW header and the position of V125. The dominant modes are random and maintenance failures of all three pumps. The significance of V125 diminishes greatly. The SUFP is again powered from offsite sources and it receives an automatic initiation signal but its overall failure rate is larger than the failure rate of the safety-class EFW pumps.

2. LOOP a) REF.3 Design The dominant modes are similar to Case 1(a) except that in addition to random or maintenance failures of MDP-37B coupled with maintenance errors causing V125 to be closed, random and maintenance failures of Emergency Diesel Generator 1B are present. This agrees qualitatively with the BNL results. The SUFP must be connected to the Emergency Diesel Generator 1A power source and aligned to the EFW header, but it is functionally redundant to TDP-37A so that failures of MDP-37B and V125 still predominate. The results of the WAMCUT out-put are shown in Fig. 13, Sh. 1-2.

41

b) Proposed Design

. This case is similar to Case 1(b) in that the SUFPS is no longer dependent upon the EFW header and the position of V125. In addition to the random and maintenance failures of all three pumps themselves, random and maintenance failures of Emergency Diesel Generators IA and 18 becosre significant contributors as well as failure to connect the SUFPS to electrical' power sources. Diesel 1A is used to supply power j to the SUFPS.

3. LOAC-In this case, there are no major differences between the REF.3 Design and the Proposed Design which affect the dominant failure mod-es. The dominant modes are maintenance acts on or random failure of

. the TDP-37A itself or on one of the steam admission valves V127 or

, V128.

, The position of V125 is also critical in that if it is left closed due to maintenance error, only Steam Generator A can be sup-plied feedwater, _ violating the mission success criteria. A listing of 1 the cutsets generated by WAMCUT for subgate AF91 which represents "No Flow to the Supply Header From TDP-37A" is shown in Figs.10 and 11.

9.2.3.5 General Comparison to Other Plants The Proposed Design at Seabrook;cnosjsts gf two safety-class EFW pumps and a Startup Feed Pump. The latter is dedicatt-E Mbsesibp to 57, of power opera-tion and for the Hot Standby and Hot Shutdown modes. Many plants have two safety-class motor-driven pumps and a third safety-class . steam turbine-driven pump. In the 'Seabrook design, one of the safety-class pumps is steam turbine-

' driven while the other is motor-driven. A somewhat similar arrangement exists at the Byron /Braidwood plant which has two safety-class AFW pumps and a manually-actuated Startup' Pump which is in series with four Boocter Pumps and four Condensate Pumps. One of the safety-class pumps is motor-driven while the other is diesel-driven.

The Seabrook SUFP exhibits automatic initiation upon trip of both MFW pumps and draws suction from the CWdth:-Qpnal reserve of 200,000 gallons.

It is independent of any Booster or Condensate Pumps. The CST is a safety-

._ class designed tank with sufficient capacity (400,000 gallons) to supply both the EFW pumps and the SUFP ~if all three are operating simultaneously. There l are no manual-actions required, either locally or in the Control Room, to main-tain the suction source to the pumps once the pumps have begun operation, ex-cept if the SUFP _is to be aligned to the base of the CST or to the Condenser Hot Well.

n l The SUFPS is normally aligned to the MFW headers for both the LMFW and LOOP transin ts. The Seabrook design limits or stops all feedwater flow to a l

l i

42

l steam generator undergoing depressurization without compromising the use of each pump, since any one of the pumps can feed all four steam generators.

With the exception of the need to manually connect the SUFP to Emergency Diesel Generator 1A during LOOP and the higher failure rate of the SUFPS it-self, the Seabrook design is comparable in reliability to plants with three safety-class AFW pumps. The use of only the SUFPS for LMFW reduces the number of challenges to the EFWS.

One definite disadvantage in comparison to most other plants is the pres-ence of valves V125, V126 anu V127 on the EFW header. In particular, the inadvertent closure of V125 or V127 can limit the flow from one of the EFW pumps to only one steam generator.

9.2.3.6 General Comments The following aspects of the Seabrook AFWS should be highlighted-

1. Pump Suction Valves The locked open, manual 1 solation valves on the suction line to each EFW pump, V154 and V155 for TDP-37A and V158 and V159 for MDP-378, and to the SUF pump, V141, V143, and V152, do not have Control Room indication or position interlocking with the pumps' start-up circuit. Since they are locked open, they do not require periodic surveillance as per the Tech-nical Specificaitons. If one of the valves is closed and the cor-repsonding pump is actuated, damage to the pump will probably occur since there are no protective pump trips upon low NPSH.

2. Emergency Feedwater Header Valves Closure of valve V125 or V126 or V127 on the EFW header reduces the number of steam generators which are supplied from each EFW pump. However, the applicant has indicated in the Supplement to this report that these valves would only be closed in the event of a pipe rupture in the EFW system.

3. SUFPS Technical Specification Limits Although it is not explicitly stated in the Technical 6 Oifications, BNL's calculated unavailabilities for this EFWS are baseo _a the as-sumption that the SUF pump will be subjected to Technical Specification limits on allowed outage time. If no such limits are placed on the SUF pump, the system unavailabilities will be substantially increased.

43

REFERENCES

- 1. " Generic Evaluation of Feedwater Transients and Small Break' Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants", NUREG-0611,

, U.S. NRC, (January,1980).

2. Letter from D.F. Ross, Jr., U.S. NRC, to "All Pending Operating License Ap-

-plicants of Nuclear-Steam Supply Systems Designed by Westinghouse and Combustion ~ Engineering", (March 10,1980).

3. " Reliability Analysis of the Emergency Feedwater System at the Seabrook Nuclear Power Station", B.A. Brogan, R.E. Land, L.E. Peters, Jr., and A.E. -

Tome, Jr., Wood-Leaver and Associates Report No. WLA-1-R-82-02 (June,

1982). ,

r 4. " Auxiliary Feedwater System (PWR)", U.S. NRC Standard Review Plan 10.4.9, Rev.-2, NUREG-0800, (July,1981).

5. Seabrook Station-Final Safety Analysis Report-Technical Specifications.

6. '" Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants - Appendix 3 and 4:' Failure Data", U.S. Nuclear Regulatory Commission, WASH-1400 (NUREG 75/014), (October,1975). _

~

7. " User's Guide for the.WAMBAM Computer Code", F.L. Leverenz and H. Kirch, EPRI Research Project 217-2-5, Key Phase Report (January,1976).

8. "WAMCUT, A Computer Code for Fault Tree Evaluation", R.C. Erdmann, F.L.

Loverenz and H. Kirch, EPRI Research Project 767-1, NP-803 (June,1978).

i i

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44

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.i=r :.-e.s.

' 2--.F g

.: Ix,s I{

Y

=

'f'stf-,.q.3" @M

-r em b 3' .; I ! lD t

,~ g :

a

*' Sa** '7 m: ."..' W.*tm-

=e : -*13 - *' "

/s'Pw'3.y,,,:ry

c. .au a i ci.e s ,ius ~ -.

.j s,..an : tec I, l & '; ,. .l. i

,,.. . e., u.o. m . _ L .' Y,.e.: .. .

ii . i

.5..-'~ _ ,, # iw o

........ - 4 l M'1! 4

..:... . . = - g D-'o o !,,,,,... ll RtMi o

i.. .in . . %

,i.pa..... 31

............ , v . .. . . . , , Iii .

.a.

, a ... .., 1 .2 3.' 9 L j g%n,,

'e fq:

i~m E % ' -

4.

5 !.}e Il 15}3 $ l e . s.s hg pt tp - .g g , %Z.g 4 il -

Eb .r ~'

.d.,

4 I <-....,...,o S

l 3 .oe .C- 3 8:n..o 0 . n.i-I '! 535 SSE

1 I P" I /* I f **

, , = , 1, ,

'a='n... .g..m w.m

% ., - f; g :se...e> N "

soci . .. e n, es ers's c s' c* * ?s3

. ....r

,... . <ei.:/

cs 4755 2 afSwGt _

l . F. 3 - . S h e.

Figure 6: Motor-Driven EFW Pump control Logic 49

l l

System Unavailability Calculated by WAMBAM:

1029 AFW 2.06786E-05 Important Cut-sets as Caculated by WAMCUT:

CUT SETS FCR GATE AFW ORDERED BY FROBABILITY

1. 3.40E-06 WB1DT/BMG M/D1V1250D

2. 2.40E-06 MPB1D37BME MVD1V1250D

3. 2.OOE-06 MVD1V71BCO MVD1V1250D

4. 1.20E-06 MVD1V1250D MCDID37BMN

5. 1.OOE-06 MVD1V1250D FVM1V1580D

4. 1.OOE-06 MVD1V1250D FVM1V1590D

7. 9.40E-07 MPB1D37 BOO MVD1V1250D

8. 3.40E-07 MPB1D37BMG N/DIV125MD

9. 3.14E-07 W B1D37BMG MPB1 U7AME MVD1V1630A

10. 2.86E-07 MPB1D37BMG MPB1T37AME MVD1V1560A

11. 2.40E-07 M/D1V125MD MPB1IG7BME

12. 2.22E-07 MPB1T37AME MPB1D37BME MVDIV1630A

13. 2.13E-07 W B1T37ANG NPD1D37BMG MUD 1V1630A

14. 2.02E-07 MPB1T37t#iE MPB1D37BME MVD1V1560A

15. 2.OOE-07 MVA1V70BMA MVD1V1250D

16. 2.OOE-07 MVDIV125MD MVD1V71 BOO

17. 1.94E-07 MPB1T37AMG MPB1D37BnG N/D1U1MOA

13. 1.35E-07 MPB1T37AME MVD1V71 BOO MVD1V1630A

19. 1.6GE-07 MPB1T37AME MVD1V71 BOO MVD1V1560A

20. 1.66E-07 MPB1D37BM3 MPB1T37AME 6ICIFPSAMN

21. 1.50E-07 MPB1T37ANG Mr B1EG7BM MVD1V1630A

22. 1.37E-07 MPB1T37ANG MPB1D3?BME MVD1V1560A

23. 1. 2"E-07 MPB1T37AMG N/D1V71 BOO M/DIV1630A

24. 1.20E-07 MVD1V125ND MCD1LG7DMK

25. 1.17E-07 WB1T37 ate MPB1TG7BME 6IC1FPSAMN

26. 1.14E-07 MPB1D37IWG MPB1T37AME MPB1P113ME

27. 1.14E-07 MPBIEG7BMG MPB1T37AME WB1P161ME

28. 1.14E-07 MPB1T37AMG MVD1V71 BOO MVD1V1560A

29. 1.12E-07 WB1T37AMG MPB1EG7BMG 6IC1FPSAMN

30. 1.11E-07 MPB1D37BMG MPBIT37AME MRA1P161MK

31. 1.11E-07 WB1T37AME MCD1EG7DMK M/D1U1630A

32. 1.01E-07 MPB1T37AME MCD1D37BMK MVD1V1560A figure 7: Applicent's WAM Results for Loss of Main Feedwater l 50 c --- .- -

i System Unavailability Calculated by WAPEAM:

1028 AFlJ 5.21742E-05 Important Cut-sets as Calculated by WAMCUT:

CUT SETS FCF: GATE AFW ORDERED.BY PROBABILITY

1. 1.COE-05 MVD1V1250D ROD 11B-ME

2. 3.40E-06 MPB1D37BMG MUD 1V1250D

3. 3.OOE-06 MUD 1V1250D RGD11B-MG

4. 2.40E-06 MPB1D37BME MVD1V1250D

5. 2.OOE-06 MVD1V71 BOO MVD1V1250D

6. 1.20E-06 MVD1V12 BOD MCD1D374MN

7. 1.OOE-06 MVD1V125MD RGD11B-ME

8. 1.OOE-06 MVD1V1250D FVM1V1580D

9. 1.OOE-06 MVD1V1250D FVM1V159CD

10. 1.OOE-06 MVD1V1250D RCA1A74-MB

11. 9.40E-07 WB1Du dOO MVD1V1250D .

12. 9.24E MPBIT37AME RGD11B--ME MVDIV1630A

13. 8.40E-07 WB1T37AME RGD11B-PE RGDIEG1AME

14. 8.40E-07 MPB1T37ME RGD11B--ME MVDIV1560A

15. 7.OOE-07 MVD1V1250D RGD11B-00

16. 6.27E-07 MPBIT37AMG RGD11B--ME MVD1V1630A

17. 5.70E-07 WB1TJ7AMG RGD11B-PE RGD1DG1AME

18. 5.70E-07 MPB1T37AMG RGD11B--ME MVDIV1560A

19. 4.87E-07 MPB1T37AME RGD11B-ME 6IC1FPSAtN

20. 3.40E-07 MPB1D3/eMG MVD1V125MD

21. 3.36E-07 MPB1T37AME- RGD11B-ME MPB1P113PE

22. 3.36E-07 MPBIT37AME RGD11B--ME MPB1P161ME l 23. 3.31E-07 MPB1T37AMG RGD11B-ME 6IC1FPSAFN i 24. 3.28E-07 MPB1T37ME RGD11B--ME MPA1P161MK

25. 3.14E-07 MPB1EG7BMG MPB1T37AME MVD1U1630A

26. 3 OOE-07 MVD1V125HD RGD11B--MG

27. 2.86E-07 MPB1EG7BMG MPB1T37AME RGD1EG1 APE

) 28. 2.86E-07 MPB1D37BMG MPBIT37AME MVD1V1560A l 29. 2.77E-07 WB1TJ7AME RGD11D-MG MVD1V1630A i 30. 2.52E-07 MPB1T37AME RGD11B--MG RGD1DGIAME

31. 2.52E-07 MPB1T37AME RGD11B-MG MVD1V1560A

32. 2.52E-07 MPBIT37ME RGD11B--ME RGD1DG1AMG

! 33. 2.40E-07 MVD1V125MD MPB1D37BME l 34. 2.28E-07 MPB1T37AMG RGlil l B--ME MPP1P113ME

35. 2.2GE-07 MPB1T37ANG RGD11D-ME MPB1P161PE

36. 2.22E-07 MPB1T37AMG RGD11B--ME MRA1P161Mh l

l Figure 8: Applicant's WAM Results for Loss of Offsite Power (Sheet 1) 51 c

i i

p I l Important Cut-sets as Calculated by WAMCUT: l l

37. 2.22E-07 WB1T37AE WB1D37BE MUD 1V1630A

38. 2.20E-07 MVD1V65AOO RGD11B--ME MVD1V1630A r

l 39, 2.13E-07 WB1T37ANG - MPB1EG7BMG MVD1V1630A

40. 2.02E-07 MPBIT37AME MPB1D37BME RGD1DGIAME

41. 2.02E-07 MPB1T37ANE WB1EG7BE MVD1V1560A

42. 2.OOE-07' .MVA1V70BMA MVD1V1250D

43. 2.OOE-07 MVD1V125MD MVD1V71 BOO

' 44. 2.00E-07 MVD1V65AOO RGD11B---ME RGD1DG1AME

45. 2.OOE-07 MVD1U65 ADO RGD11B-E MVD1V156CA

46. 1.94E-07 MPBIT37ANG MPB1D37BMG RGD1DG1AME

47. 1.94E-07 WB1T37AMG MPBIEG7BMG MVD1V1560A

48. 1.88E-07 MPB1T37AMG RGD11B--MG MVD1V1630A

49. 1.85E-07 MPB1T37AE MVD1V71 BOO MVD1V1630A

50. 1.71E-07 MPB1T37AMG RGD11B--MG RGDIDG1AME

51. 1.71E-07 MPB1T37ANG RGD11B-MG MVD1V15604 l

52. 1.71E-07 MPB1T37ANG RGD11B--ME RGD1DG1AMG

53. 1.60E-07 W B1T37AE MVD1V71 BOO RGD1DG1AE

54. 1.68E-07 MPB1T37AME MVD1V71 BOO MVDIV1564]A

55. 1.66E-07 MPB1EG7BMG MPB1 m AME 6IC1FPSA m .

56. 1.50E-07 MPB1T37MG MPB1D37BME MVD1V1630A

57. 1.46E-07 WB1T37AN RGD11B-MG - 6IC1FPSAtH

58. 1.37E-07 MPB1T37AMG MPB1D37BME RGD1DG1AME

59. 1.37E-07 WB1T37ANG MPB1EG7BME NJD1V1560A

60. 1.25E-07 MPB1T37ANG MVD1V71 BOO MVD1V1630A

61. 1.20E-07 P1/DIV12SMD MCD1D3^7BMK

62. . 1.17E-07 MPB1T77ME MPB1D37BME 6IC1FPSAMN

! 43. 1.16E-07 MVD1V65ACO RGD11B-ME 6IC3 F PSAMN

64. 1.14E-07 MPB1D37EMG MPB1T37AME MPB1P113ME

65. 1.14E-07 WD1EG7BMG MPB1D7AME WB1P161E

66. 1.14E-07 MPB1T37AMG MVD1V71 BOO RGD1DG1AME

67. 1.14E-07 WB1T37AMG MVD1U71 BOO f1>D1V1560A 68.- 1.12E-07 MPB1T37f#tG MPB1D37BMG 6IC1FPSAMN

69. 1.'11E-07 MPB1EG7BMG MPB1T37AME MRA1P161PK

70. -1.11E-07 MPB1T37AME MCD1D37BMK MVD1V1630A

71. 1.10E-07 FVM1V1540D RGD11B-ME t1/D1V1630A

72. 1.10E-07 FVH1V1550D RGD118--ME MVD1V1630A

73. 1.10E-07 CNDIV95AOD RGD11B-ME MVD1V1630A

74. 1.01E-07 MPBIT37AME RGD11B--MG MPB1P113ME

75. 1.01E-07 MPB1T37AE RGD11B-MG MPB1P161T

76. 1.01E-07 MPB1T37AME MCD1D37BMK RGD1DGIAME

77. 1.01E-07 WB1T37AT MCDIEG7BMN MVD1V1560A

78. 1. 01E--07 MPB1T37AME RGD11B--ME MCDIP113MN

( 79. 1.01E-07 W B1T37 ate RGD11B-ME MCD1P161Mb Figure 8(Continued) (Sheet 2) 52

System Unavailability Calculated by WAMBAM:

702 AFW 2,13190E-02 Important Cut-sets as Calculated by WAMCUT:

CUT SETS FOR GATE AFW ORDERED BY Ff<0BABILITY

1. 8.40E-03 ffB1T37Ati

2. 5.70E-03 MPBIT374G

3. 2.OOE-03 t1/D1V65AOO

4. 1.00E-03 MVD1V12tOD

5. 1.OOE-03 FVn1V1540D

6. 1.00E-03 FVH1V1550D

7. 1. CK)E-03 GJDIV95AOD

8. 4.20E-04 MPB1T37AOO 9 2.OOE-04 MVA1V64Am

10. 1.00E-04 iIVD1V12mD

11. 1.COE-04 FVM1U154MD

12. 1.00E-04 FVM1V25MD

13. 1.OOE-04 G/M1V95AMD

14. 1.OOE-04 GVD1V12900

15. 1.OOE-04 ff>D1V65AOD

16. 1.00E-04 MV181V6541D

17. 8.SOE-05 i+>D1421400 Pf>DIV1260D Figure 9; Applicant's WAM Results for Total Loss of AC Power 53

WBRDOK;tMFP.~EFWS- 27RITNS T ST ARTUP FEEDPUMP - NUREG-05T1 St0Pti CUT SETTPOR GWTW91 wtTH-Pc0RABTL'1'TY TG M T00E%ts le iT00 Errr4 MVDIVESIMD

2. 5 . 0 0 E-> 0 4 MVD1V65A00

3. 1. 0 0 E--~0 <, qVDTVT29MD

4. 1.00E-04 QVM1V95AHD 1.uBE-ud uVTf1V95K00 d.

6. 2.50E-03 OVXIV12700 7 2T90E703 OVX~IV12800

8. 3 00E-03 MPBIT37AME

9. 5. 8;)E;0 3 MPBIT37 A00

10. 1.00E-04 FVM1V155MD

11. 5700E ~03 rVM1V15SDD 12, 1.00E-04 FVM1V154MD 17 . IT00E03 r VMI VTS4DD-

14. 1.00E-04 MVA1V64AHA CUT 7ET S-F0p-G A T E-MF91 ORDERED BY-PHORABIt1TY

1. 5780E;03 MP61T37A00 E. 5.00E-03 FVM1V15500

s. 3700t:03 NP81T37AME

4. 2.50E-03 OVXIV12800

5. 2.50E-03 -QVXIV12700

6. 1.00E-03 FVM1V15400

7. iT00E;03 OVMI'V95AOD

8. 5 00E-04 MV01V65A00

v. 1.-00E;04 MVAIV6*AMA

10. 1.00E-04 FVM1V154MD l'1 . 1700E 04 FVM1V155MD

12. 1.00E-04 OVM1V95AM0

13. 1700E;0C OV01V129MD i

14. 1.00E-04 MVD1V65AMD IST MOMENT = 2.1785E-02 Figure 10: BNL Cutsets - LMFW l Sheet 1: No Flow from Turbine-Driven Pump P-37A. l l

54 l

5EABROOK;t. T Fw. EFW5 / TRKTNS T STARTUP FEEDPUMP - NUREO-0611 st0PE-CUT SET 5-FOR-GATt AFT 2r g r-TH po0BABIl-1TT TGt.. 1700E 0w 1.

1. 0 0 E- 0 4-----~ M V D I V 71 n M D -~-~-- -~ ~~~--

2. 5.00E-04 MVD1V71900 --

3. S';00E 7tP91037BME 4 5.40E-03 MP R 1037 E* 00

5. 1~;00E 04 FVH1V159MD'

6. 5.00E-03 F V>t i V 1590D ' - ~ ~ -

7. 1. 0 0 F- O'4-~~ FVM I V 15 8 M D-- --

8. 1.00F-03 FVM1V15800 MVAIV70BMA~~~

9. ~l~.~00E-04 10, 3.00E-05 LOSP 4GDilB--ME CUT ~ SETS-POR GATE AF127 OROFRED BY-~ PROB ARILIT T IT SWROEr03 HPP1037P00 ---~

2. 5.00E-03 FVM1V1590D

3. 5~0 0E 03 MP91037BME'--~

4 1.00E-U' FVM1V1580D 5 5'.00E-04 MVD1V71' ROD ~ ~~~

6 1.00E-04 MVA1V70BMA

7. 1700E;04 FVM1V158MD

8. 1.00E-04 FVM1 V159ND 9, 170 0 E 40'4 MVD I V71TIMD

10. 3.00E-05 LOSP DGD118--ME IST MOMENT: 1.7647E-02 figure lo (Continued) BNL Cutsets-LMFW Sheet 2: No Flow from Motor-Oriven Pump P-378.

55

I I

TETBROOK;CMFW EFWS~2 TRKINS~F STARTUP-~FEEDPUMP' 6~NUREG-06T1 SCO P E-

- CUT- SET 5-POR' G ATt--"5UP1 wTf H-PROB A8It.ITY TGFITO0E705

1. 3.00F;02 MPB1916 Tot

2. 1 00E-02 MVD1'V1560A

3. .i.00E;02 MVD1V1630A 4 _5.80E-03 MP81P11300

% ST0 0 E;0'3' rA P G 1 P1'13Mt

6. 5.00E-03 MP81P161ME

r. ST00E;03 eVH1vl520D

8. 2.00E-03 MCE1P113MN

v. 1700E-03 FVD1V14100

10. 1.00E-03 FVD1V14300

11. 1700E;04' MPR1P1130G

12. 1.00E-04 RCA150AEMC 1700E;04 H C ATAF4;MC-

~

l3.

14 1.00E-04 RCA1A63-MC

~15 . 1.0CE;04 FVFivl52ND

16. 1._00E-04 FV01V141MD

17. iT00E;04 FVDIV143MD

18. 1.00E-04 MVA1V99-MA

19. 3.00E 05' L O S P -~~ R3D1DGIAME

20. 3.00E-05 RCA1A42-M8 PGD1DG1AME el. 1.00E-05 COSP' DCA1A93-OB

22. 1 00E-05 RCA1A42-M8 ACA1A93-08 IST MOMENT = 7.3536E-02 Figure 10 (Continued) BNL Cutsets-LMFW Sheet 3; No Flow from Start-up Feed Pump P-113.

56

V ARR00K-Tn0PTEFWd <' I FATNS7-~STARTUP TEEDPtJMP - NUREG=neali dCOP E-

" TUT SETS-POR-0ATMF9i WPP"OB'AB11.-fiY- 7GF 170 0Eh6

1. 1.0DELT4 MVDIV65 Amu

2. 5.00E-04 MVD1V65AOD

a. 1.00t-u4 uYDIVT2"9Mu 4 1.00E-04 QVMIV95AMD uY!TIV95T0u

d. 1.uVE 03

6. 2 50E-03 QVXIV12700

7. 2TEUEr03 OVXIV12000 8 3.00E-03 MPB1T37AME

9. 5.tSVE;V3 etPUTT37A00

10. 1.00E-04 FVH1V155MD

11. 5.00E;03 rVM1V1550D

12. 1.00E-04 FVM1V154 tid

13. 1700E703 FVM1V1540D 14 1.00E-04 MVA1V64AMA LUT SETS-TOR-~G ATE-- AF91 vRDEPEfrRY pR00ABIETTY

1. ST80E;03 MPBIT37A00

2. 5 00E-03 FVM1V15500

3. 3.00F;03 MPBIT37AMt

4. 2.50E-03 OVX1v12800

n. 2.50E303 wVXIV12700-

6. 1.n0E-03 FVM1V15400

7. 1700E;03 OVMIV95A00

8. 5.0fiE-04 MVD1V65AOD

v. 170 0 E 0 4- MV A1 V64 AM A-"-

10. 1.00E-04 FVM1V154MO

41. 1700E-04 FVM1V155MD

12. 1.00E-04 QVM1V95AMD

13. 1.~ 0 0 E;0 4 -QV01Vl?.9MD

14. 1.00E-04 MVD 1V6'i AMD 1GT MOMF,NT: 2.178SE-02 Figure 11: BNL Cutsets - LOOP Sheet 1: No Flow fr a Turbine-Driven Pump P-37A.

57

SEKBR00K;t00P : EFWS' 2-7RNINSTSTA5?TUPTEEOPUMP-~ ~NUREG-06T17CO PE-

- CUT 7ETS-POR-GATPArt27 d rT H-P R O R'A BI L-I TY--G E'. 1.00E=0;

1. 1700E;U4 MVU1971BMD

2. 5.00E-04 HVDiv7180D

3. i.00E;03 HCK1T7WaM8 4 3.00E-02 RG0118--ME

3. 6T4 D E'- Q a x001TB-w00

6. 5.00E-03 MP01037BME

r. 5.80E-03 uPB1037800 -

t .J 8. 1.00E-04 FVM1V159MD

9. ST00E;03 FVMIV1590D

10. 1.00F-04 FVM1V158MD fl . 1TnDE 03 FVid1V15B00

12. 1.00E-04 MV41V70BMA CUT SETS-FOPG ATE-AF1~27 ORDFRED-0 Y-PRO B AB IL-I T Y

1. 3:00E=92 u G D11 g=a M t.

2. 6.40E-03 RGD11R--00

3. 5 80E;0'3 HPR1037B00 4 5.00E-03 FVM1V15900

5. 570 0 E20 3- MPBI D 37 BME-

6. 1.00E-03 FVM1V15AOD

r. 1700E703 RCATA74 ?48-

8. 5*00E-04 MVD1V71H00

9. IT00E;04 MVA1V70BMA

10. 1.00F-04 FVM1V158MD

11. 1T00E-04 FVM1V159MD <

12. 1.00E-04 MVD1V71BMD IST MOMENT: 5.4175E-02 Figure 11 (Continued) BNL Cutsets-LOOP Sheet 2: No Flow from Motor-Driven Pump P-37B.

58

4

~SEABROOK; LOOP :"EFWS ?~ TR AIfJSYSTA5) TUP ~FEEOPUMP' - NUREG-06'11 scope cur SETS-FnR-GATE M VP1 WTT W P!)08 A B I L-1 T Y- TG ET--1T0 0 Evo s

1. 3.00E-02 MPB1P1610F

e. 3700C;02 RGD1DGl'At4E--

3 1.00E-02 RCA1A93-0B 4 1.00Ew02 MV01V1560A

5. 1.00E-02 Mvn1V1630A 6 6T40E;03 RGD11AG00

7. 5.80E-03 MPHIP11300 -

8. St00E'-03 MPR1P113ME

9. 5.00E-n3 MPR1P161NE f0'. 5 00E303- FVM1V15200

11. 2.00E-03 MCE1P113MN

12. 1.'00E-03 PCA1A93-MB-

13. 1.00F-03 RCA1A54-MB -

14 T;00E-03 rVD1V14100

15. 1.90E-03 FV01V14300 l' 6 . ~ 1.00E~04 - HPR1P1130G

17. 1.00E-04 R C A150 A Et4C

~

l'8 . "~ ~ T .~0 0 E - 0 4 RCA1AF4-MC'

19. 1.00E-04 RCA1A63-MC e 0 T ~~-9 . 0 0 E- 0 4" FVHIV152MD=

21. 1.00E-04 FV01V141 TID - - -

~~~2 2 '.-~ ' 1.' 0 0 F - 0 4- - F V D 1 V I 4 3 M O

23. 1.00E-04 MVAlv99-MA IST MOMENT = 1.1766E-01 Figure 11 (Continued) BNL Cutsets-LOOP Sheet 3: No Flow from Start-up Feed Pump P-113.

59

I SEA 8 ROOK-LMFW: EFWS 2~ TRAINS +

STFTUP FEEDPUMP~ 'NUREG-0611 SCOPE

~ CUT SETS"FOR' GATE AFW~~ ~ ~~~ WI TH' PROB A BII I TY~.GE'.-140 0E-0 7

1. 5.80E-06 MP81037800 MVDIVIP500

-"2.~ " 5.00E;06~' "MVD) V 12500~-- FVH1V1590D-

~ - ~ ~ ~

3. 5.00E-06 MPB1D378hE "VDIV12500 4 ~'1.01E-06 MOBIT37 A00 ~~~ MPBI D37800-- ^ MPB1P 1610E ~

5. 1.00E-06 MVDIV12500 FVM1V1580D

~6 .' ~8.70E-07 * 'MP61T37A00 ~ ~ FVM1V1590D- mpg 1 P 1610E ""

7. ~~

8.70E-07 NP81D37600 FVM1V155nD MP81P1610E

~8. 8. 7 0 E- 0 7 ~ ~ hP81D37BhE MPRIT37A00 MP91P16'10E -

9 7.90E-07 FVM1V15500 FVM!V15900 MPB1P1610E

- 107 '7.50F-07~ MPRID37BME FVMTV15500- MPR1Plb10E -~ ~

11. 5. ROE-07 MVD1v125MD MP01037R00

~'12. - 5. 2P E- 0 7 ~~* MP81 T 3 7 A M E"~-- MP R 10 3 7 8 00--'"~MP P 1 P l e 10 E ~-~'

13. 5.0nE-07 MVD1V12500 MVDIV71900

~ ~ - - -

^14. "-~~ 5 . 0 0 EL O F- '-~ MVD 1 V 12 5 MD FVH 1 V 15900--

15. 5.00E-07 MVD1V125MD MPB I D37FeiE

--) 6. 4. 5 0 E-07'~~ MPR I T3 7 AME~~-FVM1 V 15000 MPR1Pt610E-- ~

17. 4.50E-07 MPBIT37AME MP810379ME MPR1P1610E

~ 18 .' --~ 4 '. 3 5 E- 0 7 " ~'MP81D37800-" DVU V12800 M P 81 P 1610 E"--""

19. 4.35E-07 MPB1D37800 OVXIV12700 MPR1P1610E

' 2 0. ' ~ ~ 3. 7 5E-0 7 cVXIV12800 ryMTV19900 M p B 1 P 1610 E' -~

21, 3.75E-07 OVXIV12700 FVM1V15900 HPR1P1610E

'22. -~3.75C-07 ~ MPRID37FME - ~0VXI V12800 MPB1P1610E- -

23. 3.75E-07 hP91037HME OVY1V12700 MP81P1610E 2 4 '. '"3. 3 6 E-0 7 - uPRI T37 A00 ~~ '"PR1037800-- "VD 1V IS60 A--~~~

25 3.36E-07 MPGIT37A00 MPG 1037900 MVD1Vl630A

26. '2.00E-07'" MPB1 T37 A00- FvM1vl 5900'-- ~ MV01V 1560 A 27 2.90E-07 MPB1T37A00 FVH1V1590D MVD1V1630A

28. 2.9nE-07 MPH 1D37n00 '-' FVM1V15500 - ~ MVD1V LS60 A '

29, 2.onE-07 MP01037800 FVPt1V1550D MVD1V1630A

30. 2.40E-07~ PPR1037BME ~~ ~~MP0iT37 A00 ---"MVD1V l560 A -

31. 2.90E-07 MPR1037BME MP81T37A00 MV01V1630A

32. 2.40E-07 FVM1V1550D' FVM1V15900~~ ~MVD1VIS60A

~

33, 2.50E-07 FVH1V1550D FVM1V15900 MVO1V1630A 34 2.50E-07 MPR1037BME - FVM1V15500 ~ ~MVD1V1560A

35. 2.50E-07 MPB1037BME FVH1V15500 MVD1V1630A

36. .1.95E-07 MPH 1T37A00 MP91037800 ~ MPR1P11300 -

37. 1.74E-07 MPH 1T37A00 .FVM1V15800 HPBIP1010E 38 1.74E-07 VPR1037800 FVM1V15400 MP91Plo10E 39 1.74F-07 MPH 1T37AME UP41037R00 MVD1V1560A

40. 1.74t-07 MPB1737At1E- MP01D37800 MV01V1630A

41. 1.74E-07 "PH1037000 OVH1V95AOD MPR1Pl610E 42, 1.74E-07 MP81T37A00~ -"V01V12700 ~ MPB1P1610E -

43. 1.60E-07 MPRIT37A00 FVM1V19900 MPR1P11300 1.6PE-07 44 MPG 1D37H00 FVH1V15500'- HPA1Pil300 45, 1.68E-07 MPH 1037BME MPBIT37A00 MPG 1P11300

46. 1.68E-07 MPH 1T37A00 ' "P91D37000 MP81P113ME

47. 1.68E-07 MPH 1T37A00 MPRID37800 MPB1Pl61aE 48 1.68E-07 MPn1T37A00 '

Mn01037800 -- FV91V16 POD Figure 12: BNL Cutsets - Top Event No Flow to 3 out of 4 Steam Generators -

LMRI Gate ARI (Sheet 1) 60

l

49. 1 90E-07 FVM1V15400 FVMlV15900 MPR1Pl610E 50.~ ~1.90E ~ MPB 1 T 37 AME~"- -FVM l V15900--- MV01V 1560 A---

51. 1 50E-07 MP91T37AME FVM1V15900 MVDlV1630A

52. 1.60E407 -QVH1V95AOD' --FV41V15900 -MPp1P1610E

53. 1.50E-07 HPBIT37AME "PB1037RvE tivDivl560A 54 - 1.5 nE-07 -- MPB1 T 37 AME -MPB 1D37 BME--MV0 lV 16 30 A ----

55. 1.50E-07 FVM1V15500 FVMlV15 MOD MP81P1610E

56. ~ ~ 1.50E-07 ~ ~ MPG 10378ME - -FV'41.V15400 -MP81Plb10E

57. 1.50E-07 MPR1037RME QVM1V45AOD MP81Pl610E

- 5 8. ~ 1.50E-07 - FVM1V 15500 ---MVD ' V1270D - MP91P1610E 5 9. ~ ~~~ 'l .'4 5 E - 0 7 ~~ ~~).iP R I T 3 7 K00 FVMTV1590D MP R1 P 1'13 ME

60. 1.45E-07 MPRIT37A00 FVM1V15900 MPR1P1binE

~ 61. ~ 1.45E-07 ~ ~MPBI T 37 A00 FVMTV 15900- FV41V15200

62. 1 45E-07 MP91037000 FVM1V1550D MP81P113ME

- 63. 1 45E-07-~ 'MPR1037R00 - FVMIV15500 MPR1P16'IME - ~

64, 1.45E-07 MPB1D37800 FVM1V15500 FVM1V1520D

65. ~ ~1.45E-07' ~ ~MPBTD37800 O V X 171250(T- MVDTV1560A i 66. 1.45E-07 MP91037800 OVXIV12800 MVDivl630A

~^67. l.45E MPB1D37800 - ~ 0VX 1 V12700 ' MVD1V 1560 A - ~-~ ~

68. 1 45E-07 MPG 1037800 DVXIV12700 MVD1Vib30A

~69 1.45E-07' FVH1V1550D"~~ FVHIV15900 ^ ~ MP81P11300'-~~~

70. 1.45E-07 MP91037BME FVM1V15500 MPRlPll300 "

'71. 'l.45E-07~ ~ MP01037BME uPB1T37A00- MPB1P113ME

72. 1.45E-07 MP91037BMC "P01T37A00 MPR1P1619E

~73. "-1.45F-07 " MP910373ME ~ ~~MPBl.T37 A00 FVY1V1'520D 74 l.P5E-07 FVM1V15500 FVM1V1590D MPB1Pil34E

~?5. l .Pe r-07' F V M 1 V 155 0 0 - - r yM i y 13 9 00 --- u p q 1 P 1614 E---- --

76. 1.PSE-07 FVM1V15500 FVMlV15900 FvM1V15200

77. ' l.PSE-07 QVXIV12800~ ~ FVMIV15900 - ~~MV01V1560A'

78. 1 25E-07 QVXIV17.800 FVMIV1590D MVD1Vl630A

-" 7 9 ~. ~ " '1.75E-07 ~ OVXIV12700~ FVM1 V15900 ~ ~~- MVD1V 1560 A ~

80. l.PSE-07 QVXIV12700 FVM1V15900 MV01V1630A MPR1Pil3ME -

81. 1.?5E-07 HPR1037BME ~~ ~ ~FVM1V1550D -

82. l.PSE-07 MPUID37RME rVH1V1550D MPn1P161HE

83. 1.2SE-07 ttPRID37BME- FVM1V15500 ~ --' FVM1VlB200- ' - ~ ~

84. 1.25E-0T MP81037PME OVXIV12800 MVD1V1560A

85. 1.25E-07 MPA1077BME OVX1V12900 MVDIV1630A- ~~-

86. 1.25E-07 MPR1037RME OVXIV12700 MVD1V1560A

' 87 1.25E-07 MPG 1037BME O V X I V 12 7 00 ~~~ M V D I V 16 3 0 A - -~~~~

- 8 8. 1 01F-07 MPBIT37ANE MPR1D37H00 MPR1Pil300 IST MOMENT: 3.897bE-05

-CUT TOOK 9.498 SECS

~

Figure 12(Continued)(Sheet 2) 61

,.\.

5EKBR60K;EonPT EFWS- 2 TRXINd

+ sTARTUP TEEDPtJMP 7 NUREG"06'11- SCUPE-

" TUT SET 5 M P UATt ** wi M PRDRKBIt. TTY .oE. V.40E-07-~"

1. 3.00E-05 MVD1V12500 AGD11R--uE

2. c.40E-06 MVDTV1250D P GITITs--O u

3. 5. ROE-06 MPRID37000 MVDIV12500 4 n.22E;0'6 MPRTT37AUO wGDTITi;;ME MPq1P16T0t

5. 5.72E-06 MP41T37A00 PGD118--4E RGO-10G1 AME

6. 3.UTE-06 HVolv12500 F VE((Y}39pQ 7 5.00E-06 MPH 1D378ME MVn1V12500

a. 4 90E;l)6 FVRRT5500 RGD 11 f1;7Mt hPF1 PIT 10E-9 4.50E-06 FVH1V1550D AGD)10--ME PGD1DG)A*eE l'0. E. UOT;IF6 MVDTVT25Rv wGDTIRC ME

11. 2.70E-06 MP81T37AME DGD11R--uE MPs1P1610E Iz. d7Vt-un lWBTT37%ML CUDTIIT -NE H GDTDOW1E--

13. 2.25E-06 QVXIV12800 DGD11H--ME MP91P1610E

~14 d . 25Er0'6 cVXTV17800 RGDT1 W E AGD rDGTAME~

15. 2.25E-06 OVXIV12700 RG0110- nE MPB1P1610E T6. 2. dt:E;0B cVXTV127UD PGDTITF WME- RBD1DGtAME-

17. 1.74E-06 MPRIT37A00 RG0118--ME RCA1A93-0B

~TA. 2./4E';U6 FTPGITTEKDD wGDTIB n ME 1VD1V1567TA~~

19. 1 74E-06 f.PHIT37A00 AGD11R--ME MV01V1630A N. 1.5WE;06 ;-VM1V1550D ' WEDT1B;;ME RCKIT137 00-

21. 1.90E-06 FVM1V15500 AGD118--9E MVD1V1560A

72. 1.w0F;06 FV 6 '5500 "00110--Mc MVDIvre30 A-

23. 1.11E-06 MPRIT37A00 RG0118--NE RG011A -00 74 1 11E;00 tiPRIT37%eu wnDT10--00 MPBTP16 TOE ~

25. 1.11E-06 MPalT37A00 00D119--00 RGDIDG1AME

26. 1-~0 TE70~6 =1PiTIT37'A 00 MPRTD37PUO L1PRTP tbTot 27 1.01E-On itPRIT37A00 MP81037000 RGD1DG1AME

28. T.01Er06 HPBIT37%00 AGOTIS--ME MPR1P1-1300-

29. 1 00E-06 ftVDIV1250D FVM1V19900

~~J0. h00t:0n gvDIVr2300 4 C7tTA74 M6 -

31. 9.60E-07 FV41V1550D DGD119--00 NPB1P1610E

72. v~60Er07 FVMIVISSOD RGITrig 00 RGDIDG1 AliE-

33. 9.60F-07 FVH1vl5500 4GDl19--vE RG011^ -00

-34 9 T O'O E - 0 7 f VMIV154tD-- RGDT1Ti-sME MPFr1Prb10E-

35. 9.00E-07 FV:41V15400 PGD110--ME RG01DGlAME

~36. cr0E 07 r1PTTIT37 APE OG011R--ME A C A T-A 9 3 : 0 H -

37. 9.00E-07 MPRIT37AfiE AGD11H--ME MV01VIS60A 3B. v;00F~07 QPFIT37 apt 9001 In;aE MVnTV1630A-

39. 9.00E-07 OVM1VQ5AOD PGD114--HE MP91Plul0E 7, 0~. 97 DOE 07 GVMIV95'AOD DGDTIB;;ME RG01001AME-

41. 8.70F-07 t1PB1T37A00 FVr41V15900 MPBIP1610E T2. BU 0E;07 H PRTT37AUD FeH1719000 RGD10GrA9E ~

43. 8.70E-07 MP81037H00 FVM1V15500 MPf11P 1610E

- 4'4 . 8 470E;0T HPillD37B00 'FVMTV 19500- RG0'1DG TAME ~

45. H.70F-07 HPH1037BME *81T37A00 MPR1Pl610E 4T F ~8.70E;07 - e,PR1037BMt 7 P B I T 37 A G O---~ R GD I D G l'A V E~

47 H.7 0 E-07 FVMIV1550D QGD11P--4E taPR 1P 11300

-4B. 8770F:07 MPBIT37AV0' RGDTIT wHE 11P R 1 P 1'1351E -~

49 0.70E-07 MPBIT37A00 CGD114--nE MPHIPIDINE l ~ 5'0 .~ -' B .7 0 E-07 --"uPITI T 3 7 A00- oGD 119--ME FVil1 V 15200-Figure 13: BNL Cutsets - Top Event No Flow to 3 out of 4 Steam Generators -

LOOP Gate AFW (Sheet 1) 1 62

51. 7.90E-07 FVH1V15500 AGD118--ME MPR1P1134E

2 2 . 7 . ST,E%T17 tVfi1VTSSCD nGDlTU~;ME MPB1Pitring

. 53. 7 90E-07 FVM1V15500 RG0110--t1E FVM1V15200 54 77511E 07 uVTIViccuo wGDTint TIE RCATAv3-os

55. 7.50E-07 QVXIV12800 AGD11H--ME MVD1V1560A

56. ( . 5 0 t.- 0 7 uVXTVT2FDu wGDTIT1;;ME MVDivib30'A

57. , 7.50E-07 OVXIV12700 DG011R--mE ACA1A93-08

~5 8 . 7 30 E;TT7- OVXTV1'2700 4GDr19;w E - MV01VfS60A

59. 7799E;07 r7VXIVT270u eGDTIT:::yE r.1VDtVto30A

60. 7.90E-07 F Vt11 V15500 Fyrt t V15900 MPB1P1610E '~~

61^. 7;90E707" FV'f1V15500 FV4 W15 90D-"-~~ RGD1001~A ME

62. 7.50E-07 MPRID37HME FVMt.V15500 MPR1P1610E

63. i.50'E 07 MPBID370NE FVM1V155GD RG01001AME~

64 6.40E-07 MVD1V125HO PGD11R--00 ~-~ ~~ " ~ ~

-~6 5 . STR O E70 7 " M V D I VT2 5VD" M P G 1 D 3 7 R 00 '~ "-

ri 6. 5.76E-07 MPB)T37AME DGD11H--oE 9G0114 -00 67 . 5776E;07 NPRTT37 Af4E RGn r18 200-- itPR1P1'610E

68. 5.76E-07 PIPBIT37AME RGD116--00 RCD 10G1AME

~6 9 S'.'72E 07 MPBTT37 At'E ~HPG'D37R00-' 'tiPR1Plel'qE--

70. S.?2E-07 MP81T37AME MP810"l7800 PG010GlAME ~~

71. - 7.P2E;07 ~MP ETI T37 AMt. nGD110 -ME ~~ HP91P11300-

72. 5.00F-07 HV01V12500 "Vo!V71 ROD ~ - --

T3. ST00ELO7 HV Dl V T25MD FVM1V19900-74 5.00E-07 MV01V125HD MPHID37 bee 1,407/E-ne IST MO*1ENT:-- ~ . _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ _ . . _ _ _ _ _ . _ _ _ .

CUT TOOK 6.391 SF.CS Figure 13: (Continued) (Sheet 2) 63

l I

TABLE 2 APPLICANT'S

SUMMARY

OF MAINTENANCE AND TEST UNAVAILABILITIES l

Components Maintenance Test, Total

1) Motor Driven EFP-378 4.2 x 10-4 N/A 4.2 x 10-4

2) Turbine Driven EFP-37A N/A 9.4 x 10-4 Pump contribution 4.2 x 10-4 Turbine contribution 5.2 x 10-4

3) Startup Feed Pump P-113 4.2 x 10-4 N/A 4.2 x 10-4

4) Lube Oil Pump P-161 5.0 x 10-4 N/A 5.0 x 10-4

5) Diesel Generator 7.0 x 10-4 N/A 7.0 x 10-4

6) Valves a) Emerg. Feed Flow 8.5 x 10-4 N/A 8.5 x 10-4 Isolation Valves (4214,4224,4234, 4244,75,87,93, 81) b) Steam Supply 8.7 x 10-4 N/A 8.7 x 10-4 Valves V-127, V-128 c) Steam Supply 1.0 x 10-4 N/A 1.0 x 10-4 Valve V-129 d) Manual Isolation 9.3 x 10-6 2.0 x 10-3 2.0 x 10-3 Valve V-65, V-71 64

- _ ~ - . .. . __. -_

TABLE 3 APPLICANT'S SUMARY 0F OPERATOR ACTIONS / FAILURE PROBAGILITIES Operator Action / Error Failure Probability

1) Operator fails to open either Steam 5 x 10-3 Supply Valve V-127 or V-128 given -

failure to open automatically

2) Operator fails to close an isolation 5 x 10-3 valve which fails to close automatically

3) Operator fails to close Emergency Feed- 9 x 10-1 water System manual isolation valve to isolate rupture in header

4) Operator fails to restore valve to normal 1 x 10-3 :

position after maintenance

5) Operator inadvertently blocks actuation 1 x 10-4 signal, turns off running pump, shuts an isolation valve or fails to restore valve given indication of improper positioning

-6) Operator fails to open V-156 in startup 1 x 10-2 4

feed pump discharge line and align pump to emergency power within 20 minutes

. 7) Operator fails to open V-163 in startup -1.1 x 10-2 feed pump discharge line and close V-109 1 in recirculation line to the CST

8) Operator fails to start the startup feed 1 x 10-3 I

- pump (P-113) from the control room given '

no automatic actuation signal and exis-tence of emergency procedure

9) Operator fails to properly transfer 1 x 10-2 breaker for SUF pump to bus E5

10) Operator fails to operate transfer switch ?

on Bus E4 65

TABLE 4 APPLICANT'S AFW SYSTEM UNRELIABILITY RESULTS i

TRANSIENT 3-PUMP AFW SYSTEM 2-PUMP EFW SYSTEM LMFW 2.1 x 10-5 2.8 x 10-4

LMFW/LOSP 5.2 x 10-5 1.4 x 10-3 (5.8x10-4)

LMFW/LOAC 2.1 x 10-2 2.1 x 10-2 BNL Comment: At the June 23, 1982 meeting at NRC headquarters, the applicant stated that the value of 1.4 x 10-3 for the 2-pump EFW System under the LMFW/LOSP case should be 5.8 x 10-4 s

i 1

66

. -. .- . . - - - . . . . . . . _ _ . ,- = - _ - ,

TABLE 5 APPLICANT'S RESULTS

. DOMINANT CONTRIBUTORS TO CONDITIONAL UNAVAILABILITY LOSS OF MAIN FEEDWATER EVENT CONTRIBUTION TO EVENT UNAVAILABILITY

1. Equipment and maintenance faults: Failures 7.0 x 10-6 preventing motor-driven EFW pump from functioning coupled with maintenance errors causing isolation valve V125 to be closed.

2. Maintenance faults: Maintenance outa'ge of motor- 3.0 x 10-6 driven EFW pump train coupled with maintenance

. errors causing isolation valve V125 to be closed.

3. Maintenance faults: Maintenance errors causing 2.0 x 10-6 isolation valve V125 to be closed and the motor-driven EFW train to be inoperable.

4. Equipment and operator faults: Equipment failures 1.9 x 10-6 in both EFW trains coupled with failure of operator to properly align SUF pump with EFW system.

5. Equipment faults: Equipment failures disabling 9.0 x 10-7 motor-driven EFW pump train and isolation valve V125.

6. Equipment faults: Equipment failures disable all 7.3 x 10-7 three pump trains.

7. Equipment, maintenance and operator faults: Equip- 5.9 x 10-7 ment failure in one EFW train while other EFW train out of service coupled with failure of operator to properly align SUF pump with EFW system.

.8. Cut-sets with unavailability values less than 4.4 x 10-6 1 x 10-7 Total unavailability (all cut-sets) = 2.1 x 10-5 67

TABLE 6 APPLICANT'S RFSULTS DOMINANT CONTRIBUTORS TO CONDITIONKL UNAVAILABILITY LOSS OF MAIN FEEDWATER/ LOSS OF OFFSITE POWER EVENT

. 4 CONTRIBUTION TO EVENT UNAVAILABILITY

1. Equipment and maintenance faults:. Failures prevent- 2.1 x 10-5 ing either diesel generator 1B or motor-driven EFW pump from functioning coupled with maintenance errors causing isolation valve V125 to be closed.

2. Equipment and operator faults: Equipment failures 5.8 x 10-6 disabling both EFW trains coupled with failure of operator to properly align SUF pump with EFW system.

3. Maintenance faults: Maintenance outages or errors 5.6 x 10-6 disabling motor-driven EFW pump train coupled with maintenance errors causing isolation valve V125 to be closed.

4. Equipment faults (triples): equipment failures 7.0 x 10-6 disable all three pump trains.

5. Equipment faults (doubles): Equipment failures 2.0 x 10-6 disabling motor-driven EFW pump train coupled with failure of valve V125 to remain open.

6. Maintenance, equipment, and operator faults: Main- 1.8 x 10~0 tenance errors that disable turbine-driven EFW pump train coupled with equipment and operator errors that disable both the remaining EFW pump train and the SUF pumps.

7. -Maintenance, equipment, and operator faults: Main- 3.3 x 10-7 tenance errors that disable turbine-driven EFW pump train coupled with failures of diesel-generator 18 and failure of operator to properly align SUF pump with EFW system.

8. Cut-sets with unavailability values less than 8.5 x 10-6 10-7 .

Total unavailability (all cut-sets) = 5.2 x 10-5 68

TABLE 7 APPLICANT'S RESULTS' DOMINANT CONTRIBUTORS TO CONDITIONAL UNAVAILABILITY LOSS OF MAIN FEEDWATER/ LOSS OF ALL AC POWER CONTRIBUTION TO EVENT UNAVAILABILITY

1. Equipment faults: Failure of turbine-driven EFW 1.4 x 10 -2 pump to start or continue running once started.

2. Maintenance faults: Maintenance errors causing 4.1 x 10-3 turbine-driven EFW train to be inoperable.

3 Maintenance faults: Turbine-driven EFW train out 2.5 x 10-3 of service for maintenance.

4. Equipment faults: Miscellaneous single valve 7.0 x 10-4 frilures.

5. Maintenance faults: Miscellaneous multiple main- 8.5 x 10-5 tenance errors causing turbine-driven EFil train to be inoperable.

6. Cut-sets with unavailability values less than 10" . 1.1 x 10-5 Total unavailability (all cut-sets) = 2.1 x 10-2 69

TABLE 8 BNL RESULTS UNAVAILABILITY OF SEABROOK AFWS REF. 3 DESIGN USING NUREG-0611 DATA ,

LMFW TRANSIENT

1. - Final answer from the WAMBAM code for the top event, AFW, " Insufficient

" ' Auxiliary Feedwater Flow to Steam Generators" which includes double and triple

. test and/or maintenance outage contributions is:

AFW = 5.0 x 10 -5 at a minimum probability cutoff of 1 x 10-10,

2. Unavailability of the fo probability cutoff of 1 x 10' glowing subgates Fig. from the WAMCUTI to 3): code at a minimum is (from 10, Sh.

'AF91: No Flow to supply Header From TDP-37A l AF91 = 2.18 x 10-2

.AF127:

. No Flow 'to Supply Headgr from MOP-37B AF127 = 1.78 x 10-SUP1: No Flow to Supply Header. From SUFP SUP1 = 7.35 x 10-2

3. Contributions to unavailability of subgates AF91, AF127 and SUP1 separated into Hardware and Maintenance (or. Test) are:

! AF91 AF127 SUP1 H1 =:1.1x10-2 H2 = 1.2x10-2 H3 = 7.0x10-2 M1 := 1.1x10-2 M2 = 5.8x10-3 M3 = 5.8x10-3 where H refers to failures due to random equipment failures and human errors and M refers to outages caused by maintenance and test acts.

4. Define a new top event, AFWSH, "No Flow to Supply Header From TDP-37A, TDP-378, and SUFP":

AFWSH = (AF91)-(AF127)-(SUP1) 1

=(H1 + M1 )-(H2 +- M 2 )-(H3+M) 3 i

70

i TABLE 8 (cont'd)

Let AFWSH' = H H Hi23+HHM123+HMHI23+MHH123 where double and triple maictenance and/or test contributions have been eliminated.

AFWSH' = 9.24 x 10-6 + 0.76 x 10-6 + 4.47 x 10-6 + 9.24 x 10-6 AFWSH' = 2.371 x 10-5

5. The unavailability of AFWSH including the double and triple maintenance and/or test contributions is:

AFWSH = (AF91)-(AF127)-(SUP1)

= (2.18 x 10-2) (1.78 x 10-2) (7.35 x 10-2)

= 2.85 x 10-5

6. To estimate the contribution of double maintenance and/or test actions, subtract AFWSH' from AFWSH:

DMT = AFWSH - AFWSH' = 2.85 x 10 2.37 x 10-5

-5 DMT = 0.48 x 10

7. To obtain the final answer for the top event, AFW corrected to eliminate double maintenance and/or test actions, AFW', subtract DMT from AFW:

~

AFW' = AFW - DMT = 5.0 x 10 0.5 x 10~5 AFW' = 4.5 x 10-5 NOTE: The difference between the subgates AF91, AF127 and SUP1 and the sum of their components, e.g., SUP1 < H 3 + M3 , is caused by the subtraction of intersection terms in the WAMCUT code.

71

TABLE 9 BNL RESULTS UNAVAILABILITi 0F SEABROOK AFWS PROPOSED DESIGN USING NUREG-0611 DATA LMFW TRANSIENT

1. Refer to Fig. 2, 4 and 6. The SUFP is assumed to be supplying feedwater through the MFW system and the EFW pumps supply feedwater through the EFW header. The top event AFW*, " Insufficient Auxiliary Feedwater Flow to Steam Generators From the SUFPS and the EFWS", is approximated by the following expression:

AFW* = SUP1-[(AF91 + V125)-(AF127) + (AF91)-(AF127 + V127)]

= (AF91)-(AF127)-(SUP1) + (AF127)-(SUP1)-(V125)

+(AF91)-(SUP1)-(V127) where AF91, AF127, and SUP1 are as defined in Table 8 and V125 = Unavailability of V125, V127 = Unavailability of V127.

2. V125 and V127 are locked open valves without periodic surveillan5e. For the reasons discussed in Section 9.2.3.2, they are assigned 1 x 10- for the operator inadvertently leaving them in the wrong position. Therefore:

VT2T = VITl = 1 x 10-3 Operator Error 1 x 10-4 Plugging 1.1 x 10-3

3. Since it is no longer necessary for the operator to open V163 or V156, the failure rates for the operator failing to open these two valves can be sub-tracted from H3 of SUP1 as shown in Fig. 10, Sh. 3 (MVD1V1560A = 1.0E-02 and MVD1V1630A = 1.0E-02). From Table 8, the values of AF91, AF127 and SUP1 are now:

72

l l

TABLE 9 (cont'd)

AF91 AF127 SUP1 Hi = 1.1x10-2 H2 = 1.2x10-2 H3 = 5.0x10-2 Mi = 1.1x10-2 M2 = 5.8x10-3 g3 = 5.8x10-3

4. Separating AFW* into Hardware Failures and n'aintenance (or Test) Failures:

(a) (AF91)-(AF127)-(SUP1) = H H H123+HMHi23+

(b) (AF127)-(SUP1)-(V125) = (H H23+MH 2 3 + H23 M )-(V125)

(c) (AF91)-(SUP1)-(V127) = (H Hi3+HM i 3 + M 13 H )-(V127)

5. Substituting the new value for H3 (a) = 6.60 x 10 6 + 3.19 x 10 6 + 0.76 x 10 6 + 6.60 x 10-6

= 1.72 x 10 5 (b) = (6.0 x 10 4 + 2.9 x 10 4 + 0.7 x 10 4)-(11 x 10 4)

= (9.6 x 10 4)(11 x 10-4) = 1.06 x 10-6 (c) = (5.5 x 10 4 + 0.64 x 10 4 + 5.5 x 10 4)-(11 x 10 4)

= (11.64 x 10 4) (11 x 10-4) = 1.28 x 10-6

6. Therefore AFW* = (a) + (b) + (c) = 1.95 x 10 5 AFW* = 1.95 x 10-5 LMFW 1

1 73

TABLE 10 BNL RESULTS UNAVAILABILITY OF SEABROOK AFWS REF. 3 DESIGN USING NUREG-0611 DATA LOOP TRANSIENT

1. Final answer from the WAMBAM code for the top event, AFW, " Insufficient Auxiliary Feedwater Flow to Steam Generators" which includes double and triple test and/or maintenance outage contributions is:

-4 AFW = 2.01 x 10 at a minimum probability cutoff of 1 x 10-10

2. Unavailability of the fol probabilitycutoffof1x10gowingsubgatesfromtheWAMCUTcodeataminimum is (from Fig.11, Sh. I to 3):

AF91:

No Flow to Supply Hgader From TDP-37A AF91 = 2.18 x 10-AF127: No Flow to Supply Header From MDP-37B AF127 = 5.52 x 10-2 SUP1:

No Flow to Supply Hgader From SUFP SUP1 = 1.18 x 10-

3. Contributions to unavailability of subgates AF91, AF127 and SUP1 separated into Hardware and Maintenance (or Test):

AF91 AF127 SUP1 Hi = 1.1x10-2 H2 = 4.3x10-2 H3 = 1.12x10-1

! M1 = 1.1x10-2 M3 = 1.2x10-2 M3 = 1.2x10-2

4. Define a new top event, AFWSH, "No Flow to Supply Header From TDP-37A, TDP-378, and SUFP":

i AFWSH = (AF91)-(AF127)-(SUP1) l = (H1 + M 1 )-(H2 + M 2 )*(H3+M) 3 l

l 74

TABLE 10 (cont'd)

Let AFWSH' = H H H123+HHM123+HMHi23+MHH123 where double and triple maintenance and/or test contributions have been eliminated AFWSH' = 5.30 x 10-5 + 0.568 x 10-5 + 1,478 x 10-5 + 5.30 x 10-5 AFWSH' = 12.65 x 10-5

5. The unavailability of AFWSH including the double and triple maintenance and/or test contributions is:

AFWSH = (AF91)-(AF127)-(SUP1)

= (2.18 x 10-2) (5.52 x 10-2) (1.18 x 10-1)

AFWSH = 14.20 x 10-5

6. To estimate the contribution of double maintenance and/or test actions, subtract AFWSH' from AFWSH:

DMT = AFWSH - AFWSH' = 14.20 x.10 12.65 x 10-5 DMT = 1.55 x 10-5

7. To obtain the final answer for the top event, AFW corrected to eliminate double maintenance and/or test actions, AFW', subtract DMT from AFW:

AFW' = AFW - DMT = 20.1 x 10 1.6 x 10-5 AFW' = 1.8 x 10-4 LOOP l

l i

75

TABLE 11 BNL RESULTS UNAVAILABILITY OF SEAGROOK AFWS PROPOSED DESIGN USING NUREG-0611 DATA LOOP TRANSIENT

1. Refer to Table 8 and 10. Again the expression for AFW* is:

AFW* = (AF91)-(AF127)-(SUP1) + (AF127)-(SUP1)-(V125)

+ (AF91)-(SUP1)-(V127)

2. . As in the Proposed Design for the LMFW transient, it is no longer necessary for the operator to open V156 or V163 so that the failure rates for

.those events can again be subtracted from H3 of SUPl. The values of AF91, AF127 and SUP1 are now:

AF91 AF127 SUP1 Hi = 1.1x10-2 H2 = 4.3x10-2 H3 = 9.2x10-2 M1 =-1.1x10-2 M2 = 1.2x10-2 M3 = 1.2x10-2

3. Separating AFW* into Hardware Failures and Maintenance (or Test) Failures:

(a) (AF91)-(AF127)-(SUP1) = H H H123+HMHI23+HHM123+MHH122 (b) (AF127)-(SUP1)-(V125) = (H H23+MH 2 3 + H23 M )-(V125)

(c) (AF91)-(SUP1)-(V127) = (H H13+HM I 3 + M 13 H )-(V127) 4 .- Substituting the new value for H 3.

! (a) = 43.52 x 10 6 + 12.14 x 10 6 + 5.68 x 10 6 + 43.52 x 10-6 l = 104.86 x 10 6 l

t

[

76

TABLE 11 (cont'd)

(b) = (39.56 x 10-4 + 11.04 x 10 4 + 5.16 x 10-4)(11 x 10-4)

= (55.76 x 10-4)(11 x 10-4) = 6.13 x 10-6 (c) = (10.12 x 10-4 + 1.32 x 10-4 + 10.12 x 10-4)(11 x 10-4)

= (31.68 x 10-4)(11 x 10-4) = 3.48 x 10-6 AFW* = (a) + (b) + (c) = 1.15 x 10-4 AFW* = 1.15 x 10-4 L

. LOOP '

i l

l 77

TABLE 12

SUMMARY

OF BNL ASSESSMENTS LMFW LOOP LOAC Description REF.3 Proposed REF.3 Proposed REF.3 Proposed

1. TDP-37A AF91 2.18x10-2 2.18x10-2 2.18x10-2 2.18x10-2 2.18x10-2 2.18x10-2 Hi 1.1x10-2 1.1x10-2 1.1x10-2 1.1x10-2 1.1x10-2 1.1x10-2 M1 1.1x10-2 1.1x10-2 1.1x10-2 1.1x10 2 1.1x10-2 1.1x10-2

2. MDP-37B AF127 1.78x10-2 1.78x10-2 5.42x10-2 5.42x10-2 _ _

H2 1.2x10-2 1.2x10-2 4.3x10-2 4.3x10-2 _ _

M2 5.8x10-3 5.8x10-3 1.2x10-2 1.2x10-2 _ ,

3. SUFP SUP1 7.35x10-2 5.35x10-2 1.18x10-1 9.8x10-2 _ _

s H3 7.0x10-2 5.0x10-2 1.12x10-1 9.2x10-2 _ _

M3 5.8x10-3 5.8kl0-3 1.2x10-2 1.2x10-2 _ _

4. HEADER V125, 1.1x10-3 1.1x103 1.1x10-3 1.1x10-3 1.1x10-3 1.1x10-3 VALVES V127

5. WAMBAM AFW 5.0x10-5 - 2.0x10-4 - - -

6. TOP EVENT AFW' 4.5x10-5 - 1.8x10-4 -

2.3x10-2 _

AFW* - 1.95x10-5 - 1.15x10-4 2.3x10-2 i

l

[

i l

APPEliDIX A l

i SEABROOK AUXILIARY FEEDWATER SYSTEM FAULT TREES The following is a guideline for interpreting the basic fault l identifiers used in the attached Seabrook EFW fault tree and in fault l identifier Table B-2. !

Each fault identifier consists of 10 alphanumeric characters of the form:

X-XX-1-XXXX-XX The first character identifies the system to which the component belongs (see Table A.1). The second and third characters identify the component type (TableA.2). The fifth through eighth characters are for component identi-fication and the last two characters identify the fault codes (Table A.3).

A-1

TABLE A.1 l

SYSTEM IDENTIFICATION CODE C - Condensate System M - Emergency Feedwater System Q - Steam Supply System R - Electrical Distribution System 5 - Condensate Storage System 6 - Control / Protection System l

l l

A-2

i TABLE A.2 COMPONENT TYPES BA - Batteries BC - Battery Chargers CA - Circuit Breaker CB - Contactor CC - Controller CD - Starter CE - Switch EC - Electrical Conductors GD - Diesel Generator HX - Heat Exchanger IC - Instrument Controller ID - Sensor / Detector / Element - Pressure IE - Sensor / Detector / Element - Temperature IF - Sensor / Detector / Element - Flow IG - Sensor / Detector / Element - Level IH - Sensor / Detector / Element - Radiation IP - Power Supply IX - Instrument Error MA - AC Motor MD - DC Motor 0A - Piping less than 1 inch in diameter DB - Piping greater than 1 inch but less than 2 inch OC - Piping greater than 2 inch but less than 3 inch OD - Piping greater than 3 inch but less than 4 inch OE - Piping greater than 4 inch but less than 6 inch 0F - Piping greater than 6 inch but less than 8 inch OG - Piping greater than 8 inch but less than 10 inch OH - Piping greater than 10 inch but less than 12 inch OI - Piping greater than 12 inch but less than 16 inch DJ - Piping greater than 16 inch but less than 24 inch OK - Piping greater than 24 inch but less than 36 inch OL - Piping greater than 36 inch A-3

TABLEA'.2(CONT'D.)

PB - Centrifugal Pump RA - Control General Purpose TR - Transformer TU - Turbine VA - Check Valve VB - Relief Valve VC - Vacuum Relief VD - Isolation. Shutoff Valve VG - Flow Control l

1 l

l A-4

l-TABLE A.3 FAULT CODES CL - Cooling Failure LB - Lubrication Failure MA - Fails to Open/De-energize / Disengage MB - Fails to Close/ Energize / Engage MC - Fails to Remain Closed /De-energize / Disengaged MD - Fails to Remain Open/ Energized / Engaged ME - Fail to Start MG - Fail to Run MJ - Leak / Rupture / Electrical Short Circuit MK - Open Circuit ML - Overload MM - Underload MN - No Signal /No Input MO - Spurious Signal OA - Operator Fails to Open/De-energize / Disengage OB - Operator Fails to Close/ Energize / Engage .

OC - Operator Inadvertently Opens /De-energizes / Disengages / Leaves Open OD - Operator Inadvertently Closes / Energizes / Engages / Leaves Closed OE - Operator Fails to Start OG - Operator Fails to Leave Running OH - Calibration Error 00 - Out of Service l

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TAett 0.1 eEtiaantc4 Amp EttCielCAL Coleontui FAILWAE Mlts FAILUAf AATE WA98-1400 SE MAC lift.taA sarn marmara FAILUAE OuS put CATEGORT COMPONENT sent (1ectrical Cable Olstr h tlon power: Open circuit 9.1 a 10'I/hr 9.1 s 80'I/hr Copper Short to ground 1.7 a 10-4the 1.7 a 104 /hr short to poner 4 1.0 a 10 /hr 1.0 a 10'0/hr Alwinus Open circuit 1.1 a 10'0/hr 1.1 a M'0/hr Short to ground 3.9 a 10'ON 3.9 a 50-63, Short to poner 4 1.5 m 10 the 1.5 a 10-4/hr Control: Open circett 9.1 a 80'I/hr 9.1 a 10*I/hr Copper Short to ground 2.4 a 10-6Ar 2.4 a 10-6/hr cp Short to power 4 e 1.0 a 10'Othe 5.0 a 10 /hr m _ . . . .. . . . - - - Terminal Goerds open I a 10*I/br J.3 a 10'0/hr 3.3. a 104the Short I a 10'8/hr 4 1.4 a 10-6/hr 1.4 a 30 /hr

                                                                                                                                                                                                                                             -- .. -     . . . .      .                -.. . . -             . - . ~ . - - - - - - -         ~~

12struentation Aelay Cosi falls to I a 10'*/d .4 a 10 6/hr I a le /d and Controls *P*** , Coll falls to open 3 a 10'y/hr .00 a 10'0/hr 3 a 10'I/hr 4 1.4 a 194 /hr Temperature Fatis to operate 1.4 a 10 /hr Sensing Devica (3) Degraded operation C.6 a 10'I/hr 6.6 a 10*I/hr Temperature Elemens Fall to operate 1.0 a 104 /hr 1.0 a 104 /hr 4 Degraded operation . 1.2 a 10 /hr 1.2 s 104 /hr Temperature Falls to operate 3.0 a 10*Ishr P.0 a 10'#/hr I'**"'II'# Degraded operation 3.6 a 10*I/hr 3.6 a M /hr [3] 5enslag Device includes sultch, monitor, sensor, and transmitter

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TABLE 8.1 MCatullCAL AAB ELECtelCAL COWOutNF FAILISE'AEft$ 1 FAILISE RATE 1s4548-1400 GE I hC Ittf-laa merammenmara FAltimE sin Fim CATEGORT C390NEg7 emar lastrumentatten level sensing fall te aperate

                        ~ 'a                                                                   3.1 a 104 /hr          2.6 a 10-4/hr gc r id                          w          o,e ati-                                     ..,a  ie w             .. a .4/hr 2.5 a 30*'/hr i.2 a =%

t m l Element 3,g a 194 /hr 3.g a M4the Level Transmitter fall to operate 1.4 a 10-8/hr 1.4 a 50'0/tr Degraded operetten 1.1 a 144 /hr 4 1.1 a M thr Level 5mitch fall to operate 3 a IS*I/d 3 a W I/d Spurious operetiens 3.4 a le-8/d 3.4 a 194 /d Degraded operations 4.5 a 10'I/d 4.5 a W I/d Level Controller fall to operate Y I a le /hr I a 184 /hr 5puricus operettens

 "                                                                                                                                      I a W /hr     I a 104 /hr Degraded operation                                                                                2 a 10/hr   2 a 184 /hr E/5 Converter      fall to operate                                    4 4.2 a 10 /hr                                                         4.2 x 10' /hr   -

Square Acet Con- Fall to operate verter 4.2 a 10 /hr 4.2 a le /hr Power Supply fall to operate 4 4.2 a 19 /hr 2.8 a 104 /hr 2.3 a 10-4/hr Degraded operation 1.3 a 304 /hr 4 I.g a 10 /tr Sette State: ! Low Power falls to function I a 104 /hr I a 104 /hr falls Shorted I a 10 the I a 104 /hr High poner falls to function 3 a 104 /hr 3 a 104 /hr Falls shorted I a 10}6 /hr I a 104 /hr Tor p 5mitch falls to operate I a LO'4/d I a 10'4/d Sultch Contacts flernally open I a 83'I/hr sultches fall I a 10*I/hr to close i harsally closed 3 a 10'8/hr 3 a 10'0/hr switches fall } to close l Short across con-tM i t I a 10'8/hr I a 10 M i I

l

TABLE 8.2 FAULT IDENTIFIERS FOR TE SEASR00K EERGENCY FEED STATION DESCRIPTION LMAVAILABILIT'Y FAILURE RATE ID ER S.8.A. Relief Valve Calibration QV81SGAR9H Valve 6.0 x 10- 7 1.2 x 10-6/hr S.G.D. Relief Valve Calibration 4 QV81SGDR9H Shift 6.0 x 10~ 7 1.2 x 10 /hr 5.G.C. Relief Valve Calibration QV81SGCR$H Shift 6.0 x 10- 7 1.2 x 10-6/hr S.G.8. Relief Valve Calibration QV81SGRR9H Shift 6.0 x 10- 7 1.2 x 10-6/hr S.G.A Relief Valve Control Wiring QV81SGARM Reversed 4.2 x 10-8 8.4 x 10-8/hr S.G.D Relief Valve Control Wiring QIX1SGDRM Reversed 4.2 v. 10-8 8.4 x 10-8/hr S.G.C Relief Valve Control Wiring QIX1SGCRM Reversed 4.2 x 10 4 8.4 x 10-8/hr S.G.8 Relief Valve Control QIX1SG8Rm Wiring Reversed 4.2 x 10 ~0 8.4 x 10-8/hr Turbine Driven Pump P-37A M m T37 E 5.7 x 10-3 Fails to Run 5.7 x 10-3/d Motor Driven Pump P-378 l

MP810378MG Fails to Run 3.4 x 10-3 3.4 x 10-3/d - Isolation Valve V-30 In Feedwater l MVD1V30AMJ Supply Line to S.G.A Ruptures 5 x 10-8 1 x 10-7/hr i Check Valve V-29 In Feedwater Supply Line To S.G.A Ruptures MVA1V29AMI 5 x 10~9 1 x 10-8/hr Stop Check Valve V-76 In Aux. MVA1V29AMJ Feed. Supply Line A Ruptures 5 x 10'9 1 x 10-8/hr ! Flow Control Valve FV-4214 In l MVD14214MJ Aux. Feed. Supply Line A Ruptures 5 x 10-8 1 x 10~7/hr l- Stop Check Valve V-94 In Aux. MVA1V940MJ Feed. Supply Line D Rupture 5 x 10-9 1 x 104 /hr Flow Control Valve FV-4244 In MVD14244MJ Aax. Feed. Supply Line D Ruptures 5 x 10-8 1 x 10-7/hr i Stop Check Valve V-88 In Aux. MVA1V88CMJ Feed. Supply Line C Rupture 5 x 10~9 1 x 10-8/hr 8-8 I

TABLE 8.2(Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE. RATE ID ER Flow Control Valve FV-4234 In Aus . MVD14234MJ Feed Supply Line C Ruptures 5 x 10-6 1 x 10-7/hr Stop Check Valve V-82 In Aux. MVA1V828MJ Feed. Supply Line 8 Ruptures 5 x 10-9 1 x 10-8/hr Flow Control Valve FV-4244 In MVD14224MJ Aux. Feed. Supply Line B Ruptures 5 x 10-8 1 x 10-7/hr [ . Isolation Valve V-57 In Feedwater MVD1V57DMJ Supply Line D Ruptures 5 x 10-8 1 x 10-7/hr Check Valve V-56 In Feedwater MVA1V56DN Supply Line D Ruptures 5 x 10-9 1 x 10-8/hr Isolation Valve V-48 In Feedwater MVD1V48CMJ Supply Line C Ruptures 5 x 10-8 1 x 10-7/hr Check Valve V-47 In Feedwater MVA1V47CMJ Supply Line C Ruptures 5 x 10 4 1 x 10-8/hr Isolation Valve V-39 In Feedwater MVD1V39BMJ Supply Line B Ruptures 5 x 10-8 1-x 10-7/hr Check Valve V-38 In Feedwater MVA1838BMJ Supply Line B Ruptures 5 x 10-9 1 x 10-8/hr Manual Valve V-75 In Aux. Feed. MVM1V75AMJ Supply Line A Ruptures 1 x 10-8 2 x 10-0/hr Gear Driven Valve V-65 In P-37A VMDIV65AMJ Discharge Line Ruptures 5 x 10-9 1 x 10-8/jr Manual Valve V-152 In Feedwater MVH1V152MJ Recire. Line A Ruptures 1 x 10-8 2 x 10-8/hr Gear Driven Valve V-156 In Start ~ MVD1V156MJ up Feed Pump Discharge Line Rup. 5 x 10-9 1 x 10-8/hr Flow Control Valve FCV-510 In MVD1V510MJ Feed. Supply Line A Ruptures 5 x 10-8 1 x 10-7/hr Manual Valve V-87 In Aux. Feed. MVM1V87DMJ Supply Line D Ruptures 1 x 10-0 2 x 10-8/hr Manual Valve V-153 In Feedwater MVM1V153MJ Recirc. Line D Ruptures 1 x 10 4 2 x 10-8/hr Flow Control Valve FCV-540 In ' MVD1V540MJ Feed. Supply Line D Ruptures 5 x 10-8 1 x 10-7/hr l B-9

TABLE B.2 (Continued) FALA.T IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION j j h DESCRIPTION LMAVAILABILITY FAILURE RATE Manual Valve V-93 In Aux. Feed. MVM1V9304J Supply Line C Ruptures 1 x 10 4 2 x 10-8/hr Manual Valve V-154 In Feedwater i MVM1V154MJ Recire. Line C Ruptures 1 x 10 4 2 x 10-8/hr Flow Control Valve FCV-530 In EV1V530MJ Feed. Suppif Line C Ruptures 5 x 10 4 1 x 10~7/hr Manual Valve V-81 In Aux. Feed. MVM1V81tlMJ Line 8 Ruptures 1 x 10 4 2 x 10-8/hr Gear Driven Valve V-71 In P-378 MVDIV718HJ Discharge Line Rupture 5 x 10-8 1 x 10-8/hr Manual Valve V-155 In Feedwater MVM1V155MJ Recirc. Line B Ruptures 1 x 10-8 2 x 10-8/hr ; Flow Control Valve FCV-520 In MVD1V520MJ Feed. Supply Line B Ruptures 5 x 10-8 1 x 10-7/hr Check Valve V-64 In P-37A Dis-MVA1V64AMJ charge Line Ruptures 5 x 10"I 1 x 10-8/hr Check Valve V-70 In P-35 Dis-MVA1V708MJ charge Line Ruptures 5 x 10-8 1 x 10-8/hr Valve V-129 In P-37A Turbine QVD1V129MJ Steam Inlet Line Ruptures 5 x 10'I 1 x 10-8/hr Manual Valve V-95 In P-37A Tur- i QVM1V95AMJ bine Steam Inlet Line Ruptures 1 x 10-8 2 x 10-8/hr Check Valve V-94 In Steam Supply i QVA1V94AMJ Line A to P-37A Ruptures 5 x 10~9 1 x 10-8/hr Check Valve V-96 In Steam Supply QVA1V96BMJ Line A to P-37A Ruptures 5 x 10~I 1 x 10-8/hr Steam Supply Line A Control Valve QVXIV127MJ V-127 Ruptures 5 x 10-8 1 x 10-7/hr Steam Supply Line A Manual Bypas! QVM1V171MJ Valve V-171 Ruptures 1 x 10-8 2 x 10-8/hr Steam Supply Line S Control QVXIV128MJ Valve V-128 Ruptures 5 x 10-8 1 x 10-7/hr Steam Supply Line 8 Manual By-QMV1V172MJ pass Valve V-172 Ruptures 1 x 10~8 2 x 10~8/hr 8-10

TABLEB.2(Continued) FAULT IDENTIFIERS FOR THE SEA 8R00K EERGENCY FEED STATION l .. DESCR M ION MAMILm FA E RATE IDE ER Manual Valve V-155 In P-37A FVM1V155MJ Suction Line Ruptures 1 x 10~0 2 x 10-8/hr Manual Valve V-154 In P-37A 4 FW11V154MJ Suction Line Ruptures 1 x 10 2 x 10-8/hr Manual Valve V-159 In P-27A FVM1V159MJ Suction Line Ruptures 1 x 10-8 2 x 10-8/hr Manual Valve'V-158 In P-37A FVM1V158MJ Suction Line Ruptures 1 x 10-8 2 x 10-8/hr Isolation Valve V-30 In Feedwater MVD1V39AM Supply Line A Fails To Close 9 x 10 9 x 10-4/d Flow Control Valve FV-4244 In Aux, MVD142442 Feed Supply Line D Fails To Close 2 x 10~3 2 x 10-3/d Flow Control Valve FV-4234 In Aux, MVD14234M Feed. Supply Line D Fails To Close 9 x 10 4 9 x 10_4/d Flow Control Valve FV-4224 In Aux, MVD142242 Feed. Supply Line D Fails To Clost- 9 x 10~4 9 x 10~4/d Isolation Valve V-57 In Feedwater MVD1V57DM ' Supply Line D Fails To Close 9 x 10-4 9 x 10~4/d Flow Control Valve FV-4214 In Aux.

MVD14214MB Feed. Supply Line A Fails To Close t 9 x 10 2 x 10~3/d Isolation Valve V-48 In Feedwater MVD1V48CM Supply Line C Fails To Close 9 x 10-4 9 x 10~4g Isolation Valve V-39 In Feedwater MVDIV39BM Supply Line C Fails To Close 9 x 10-4 9 x 10-4/d l

Flow Control Valve FCV-510 In (. MVD1V510MB Feed. Supply Line A Fails To Close 9 x 10~4 9 x 10-4/d Flow Control Valve FCV-540 In MVD1V5402 Feed. Supply Line D Fa11s to Close 9 x 10_4 9 x 10-4/d Flow Control Valve FCV-530 In MVD1V530MB Feed. Supply Line C Fails To Close 9 x 10~4 9 x 10 4/d Flow Control Valve FCV-520 In -4 i MVD1V520MB Feed. Supply Line B Fails To Cicsa 9 x 10 9 x 10-4/d Isolation Valve V-30 In Feedwater MVD1V30AMC Supply Line A Fails To Remain C1. 2.3 x 10-7 4.5 x 10~7/h - B-11 i

     - ----,,        .- -- _ - _ _ - - _ _ _ - .                        - - _ .- -              - - - - - - ~ . - - - -          , - , - - - - - -        --- ---- ,

l l FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE IDS $IEn

              ! Flow Control Valve FV-4244 In Aux.

MVD14244MC , j,$u /]ineDFailsTo 6.0 x 10-8 1.2 x 10-7/hr Flow Control Valve FV-4234 In Aux. MVD14234MC j, []ineCFailsTo 6.0 x 10 1.2 x 10-7/hr Flow Control Valve FV-4224 In Aux. Feed. Supply Line B Fails'To 6.0 x 10-8 1.2 x 10-7/hr MVD14224MC Remnin Closed Iso'ation Valve V-57 In Feedwater Supply Lini D Fails To Remain 2.3 x 10-7 4.5 x 10-7/hr MVD1V57DMC Closed Flow Control Valve FV-4214 In Aux. Feed. Supply Line A Fails T 6.0 x 10-8 1.2 x 10-7/hr MVD14214MC Remain Closed Isolation Valve V-48 In Feedwater Supply Line C Fails To Remain 2.3 x 10-7 4.5 x 10-7/hr MVD1V48 CMC Closed

.                Isolation Valve V-39 In reedwater Supply Line B Fails To Remain       2.3 x 10~7     4;5 x 10~I/hr MVDIV398MC Closed riow controi vaive tcv-lou in Feed. Supply Line A Fails To       8.5 x 10-8     1.7 x 10-7/hr MVD1V510MC Remain Closed Flow. control Valve FCV-540 In MVDIV540MC     Feed Sg1 Line D Fails To           8.5 x 10-8     1.7 x 10-7/hr Flow Control Valve FCV-530 In                                                   ,

Feed. Supply Line C Fails To 8.5 x 10-8 1.7 x 10-7/hr MVD1V530MC Remain Closed Flow Control Valve FCV-520 In Feed. Supply Line B Fails To -8 MVD1V520MC 8.5 x 10 1.7 x 10-7/hr Remain Closed I Flow Control Valve FV-4214 Fails MVD14214MD To Remain Open 8 x 10-8 1.6 x 10-7/hr Flow Control Valve FV-4244 Fails MVD14244MO To Remain Open 8 x 10-8 1.6 x 10~7/hr Flow Control Valve FV-4234 Fails MVD14234MD To Remain Open 8 x 10-8 1.6 x 10-7/hr Flow Control Valve FV-4224 Fails MVD14224MD To Remain Open 8 x 10-8 1.6 x 10~7/hr Isolation Valve V-65 Fails To MVD1V65AMD Remain Open (Plugged) 1 x 10~4 1 x 10_4/d Steam Supply Inlet Valve V-129 QVD1V129MO Fails To Remain Open 2.3 x 10-7 4.5 x 10~7/hr I B-12 l

l TABLE 8.2(Continued) FAULT IDENTIFIERS FOR THE SE4 BROOK EERGENCY FEED STATION DE5CRIPTION UNAVAILASILITY -FAILURE RATE I ER Isolation Valve V-125 Fails To MVD1V125ft Rerisin Open (Plugged) 1 x 10-4 1 x 10~4/d Isolation Valve V-126 Fails To MVD1V126pm Remain Open (Plugged) 1 x 10-4 1 x 10'4/d Isolation Valve V-127 Fails To MVD1V127MD Remain Open (Plugged) 1 x 10~4 1 x 10 /d Isolation Valve V-71 Fails To 4 MVD1V71BMD Remain Open (Plugged) 1 x 10 1 x 10 /d Steam Supply Line A Flow Valve QVXIV127MA V-127 Fails To Open 9 x 10 4 9 x 10 /d 4 Steam Supply Line B Flow Valve QVXIV128MA V-128 Fails To Open 9 x 10,4 9 x 10"4/d Steam Supply Line A Flow Valve QVXIV127MD V-127 Fails To Remain Open 2.3 x 10-7 4.5 x 10~7/hr Steam Supply Line B Fails To QVXIV128My Remain Open 2.3 x 10-7 4.5 x 10~2/hr Stop Check Valve V-76 Fails To MVA1V76AMA Open 2 x 10-4 2 x 10-4/d Stop Check Valve V-94 Fails To MVA1V94DMA Open 2 x 10,4 2 x 10,4/d Stop Check Valve V-88 Fails To MVA1V88CMA Open 2 x 10-4 2 x 10-4/d Stop Check Valve V-82 Fails To MVA1V82BMA Open 2 x 10-4 2 x 10~4/d MVA1V64AMA Check Valve V-64 Fails To Open 2 x 10~4 2 x 10-4/d Steam Supply Line A Check Valve QVA1V94AMA V-94 Fails To Open 2 x 10~4 2 x 10 4/d Steam Supply Line B Check Valve j- QVA1V96BMA V-96 Fails To Open 2 x 10,4 2 x 10,4/d MVA1V70BMA Check Valve V-70 Fails To Open 2 x 10-4 2 x 10-4/d Valve V-75 In Aux. Feed Supply MVD1V75AMD Line A Plugged 1 x 10~4 1 x 10,4/d B-13

j l TABLEB.2(Continued) l FAULT IDENTIFIERS FOR THE SEA 8R00K EIERGENCY FEED STATION DESCRIPTION LMAVAILABILITY ID T ER FAILURE RATE Valve V-87 in Aux. Feed Supply MVD1V870 E Line D Plugged 1 x 104 4 1 x 10 /d , l Valve V-93 In Aux. Feed Supply MVD1V93C2 Line C Plugged 1 x 10 4 4 1 x 10 /d Valve V-81 In Aux. Feed Supply MVD1V818MD Line.8 Plugged 1 x 10 4 1 x 10 4/d Manual Valve V-95 In Turbine QVM1V95AMD Steam Supply Line Plugged 1 x 10-4 1 x 10-4/d Manual Valve V-155 In P-37A FVM1V1552 Suction Line Plugged 1 x 10-4 1 x 10'4/d Manual Valve V-154 In P-37A 4 FVM1V154MD Suction Line Plugged 1 x 10 1 x 10 /d i Manual Valve V-159 In P-378 FVM1V159MD Suction Line Plugged 1 x 10-4 1 x 10-4/d Manual Valve V-158 In P-378 FVM1V158MO Suction Line Plugged 1 x 10-4 1 x 10-4/d Feedwater Supply Line Ruptures 4 M9J1606BMJ Between V-20 and V-30 5 x 10-II 1 x 10-10/hr Feedwater Supply Line Ruptures MpJ1606AMJ Between V-30 and S.G.A 5 x 10'II 1 x 10-10/hr Aux. Feed. Supply Line A Ruptures MSE1614AMJ ptweenV-76andMainSupplyLine 5 x 10'II 1 x 10-10/hr Aux. Feed. Supply Line A Ruptures MBE1614BMJ Between FV-4214 and V-76 5 x 10~11 1 x 10-10/hr Feedwater Supply Line D Ruptures ~ M9J1609AMJ Between V-57 and S.G.D 5 x 10-11 1 x 10-10/hr Aux. reea suppiy Lme U Rup6urus Between V-94 and Main Feed. MpE1617AMJ Supply Line D 5 x 10-11 1 x 10-10/hr Feedhter Supply Line C Ruptures MSJ1608AMJ 8etween V-48 and S.G.C 5 x 10'11 1 x 10-10/hr Aux. Fced. Supply Line C Ruptures MBE1616AMJ D N d 8 and c Main Feed. 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line C Ruptures M9E16168MJ Between FV-4234 and V-88 5 x 10~II 1 x 10-10/hr B-14

FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE IDE ER Feedwater Supply Line 8 Ruptures M9J1607AMJ 8etween V-39 and S.G.B 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line 8 Ruptures MSE1615AMJ '",d2andMainFeed. n 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line 8 Ruptures MSE16158MJ Between V-4224 and V-82 5 x 10~Il 1 x 10-10/hr Feedwater Supply Line D Ruptures M9E16098MJ Between V-56 and V-57 5 x 10-11 1 x 10-10/hr Feedwater Supply Line C Ruptures M9J16088MJ Between V-47 and V-48 5 x 10~11 1 x 10 -10 /hr Feedwater Supply Line B Ruptures M9J1607BMJ 8etween V-38 and V-39 5 x 10~11 1 x 10~10/hr Aux. Feed Supply Line A Ruptures M9E1614CMJ Between V-75 and FV-4214 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line A Ruptures Between V-75 and Aux. Feed Supply M9E1614 W Header 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Header Ruptures M9G1613AMJ Between V-17.5 and Reducer 5 x 10~11 1 x 10~10/hr aux. reea rump P-a/a uisenarge Piping Ruptures Between V-65 M9F1610AMJ and Supply Header 5 x 10-11 1 x 10-10/hr reeawater Mecirc. une a nuptures Between Aux. Feed. Supply A and M9F1606AMJ V-152 5 x 10-11 1 x 10-10/hr startup Feed Pump Discharge Line Ruptures Between V-156 and Aux. M9F1632AMJ Feed. Supply Header 5 x 10-11 1 x 10-10/hr Feedwater Recirc. Line A Ruptures M9E1606BMJ Between V-152 and Main Feed Sup- 5 x 10-11 1 x 10-10/hr nly Iins A Feedwater Supply Line A Ruptures M9J1606CMJ Between FCV-510 and V-29 5 x 10~11 1 x 10~10/hr Aux. Feed Supply Line D Ruptures M9E1617CMJ Between V-87 and FV-4224 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line D Ruptures Between V-87 and Aux. Feed M9E1617DMJ Suoolv Header 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Header Ruptures M9G1613BMJ Between V-87 and V-126 5 x 10~11 1 x 10~10/hr B-15

TABLEB.2(Continued) i FAULT IDENTIFIERS FOR THE SEABROOK EMEAGENCY FEED STATION l I'D ER-DESCRIPTION ' UNAVAILABILITY FAILURE RATE Feedwater Recirc. Line D Ruptures

x. Feed. Supply Line D MBE1609AMJ y5 S x 10'11 _1 x 10-10/hr Feedweter Recirc. Line D Ruptures Between V-153 and Main Feedwater M1609BMJ Supply Line D 5 x 10-11 'l x 10-10/hr Feedwater Supply Line.D Ruptures -

MpJ1609CMJ Between FCV-540 and V-56 5 x 10~I1 1 x 10-10/hr Aux. Feed. Supply Line C Ruptures MBE1616CMJ Between V-93 and FV-5234 5 x 10~11 1 x 10-10/hr Aux, reen supply Line L Muptures Between V-93 and Aux. Feed. MpE1616DMJ Supply Header 5xIdTll 1 x 10-10/hr

              . Aux. Feed Supply Header Ruptures Between V-126 and V-127 mpg 1613CMJ                                            5 x 10~11             1.x 10-10/hr reeawater necirc. une c nupturbs Between Aux. Feed. Supply Line MBE160BAMJ C and V-154                             5 x 10-11             1 x 10-10/hr Feedwater Recir. Line C Puptures           '

MpE1608BMJ Between V-154 and. Main Feedwater _5 x 10-11 1 x 10-10/hr h nniv Iins c Feedwater Supply Line C Ruptures ~ MpJ1608CMJ Between FCV-530 and V-47 5 x'10-11 1 x 10-10/hr Aux. Feed Supply Line B Ruptures MpE1615CMJ Between V-81 and FV-4224 5.x 10 -11 1 x 10-10/hr Aux. Feed Supply Line D Ruptures MBE1615DMJ Between V-81 and Aux. Feed Supply 5 x 10'11 1 x 10~10/hr i Aux. Feed Supply Header Ruptures e MSG 1613DMJ Between V-127 and Reducer 5 x 10-11 1 x 10-10/hr l Feed. Pump P-37B Discharge Line l MSF1612AMJ Ruptures Between V-71 & Reducer 5 x 10~11 1 x 10-10/hr i reeawater Necirc. Line b Muptures- , Between Aux. Feed Supply Line B i MPE1607AMJ and V-155 5 x 10-II 1 x 10-10/hr - l Feedwater Recire. Line B Ruptures l I MBE1607BMJ Between V-155 and Main Feeduater 5 x 10~II 1 x 10-10/hr 4.- n - c..no17 ! Feedwater Supply Line B Ruptures . , l MpJ1607CMJ Between FCV-520 and V-38 ' - 5 x 10-'I 1 x 10-10jg l Feed. Pump P-37A Discharge Line - MSF1610BMJ Ruptures Between V-64 and V-65 5 x 10-II 1 x 10-10/hr - B-16 l

FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILA8ILITY FAILURE RATE IDE ER Feed Pump P-37B Discharge Line 5 x 10~11 1 x 10-10/hr

M9F16128MJ -Ruptures Between V-70 and V-71 Feed Pump P-37A Discharge Line 10 M9F1612CMJ Ruptures Between P-37A and V-64 5 x 10~11 1 x 10 /hr

;                               Feed Pump P37A Recirc. Line Rup-tures Between' V-67 and Pump        5 x 10-10        1 x 10-9/hr M901610AMJ          Discharge Line Turbine Steam Supply Line Ruptures Q9E1449AMJ          Between V-95 and V-129              5 x 10-11        1 x 10-10/hr Turbine Steam Supply Line Rupture Q9E1449BMJ          Between V-129 and Turbine           5 x 10~11        1 x 10-10/hr e          Turbine Steam Supply Line Ruptures Q9F1449AMJ          Between Tee and V95                 5 x 10~11        1 x 10~10/hr steam supply Line 5 MupT.ures Between Reducer and Turbine         5 x 10-11        1.x 10-10/hr Q9F1109AMJ          Inlet Line Tee

5 team supply Line A Ruptures Between Reducer and Turbine 5 x 10-11 1 x 10-10/hr Q9F1009AMJ Inlet Line Tee.

                    -           steam 5upply Line d Muptures                                ~10 Q9E1143AMJ        -Between V-128 and V-96               5 x 10~11        1 x 10     /hr Steam Supply Line A Ruptures Q9E1043AMJ          Between V-127 and V-94              5 x 10~11        1 x 10-10/hr r                              ' Steam Supply Line A Ruptures Q9E1043BMJ          8etween V-94 and Reducer            5 x 10~11        1 x 10-10/hr Steam Bypass Exit Line Ruptures Q9A1036AMJ          Between V-171 and Ste:m Supply. 5 x 10-10        1 x 10-9/hr I ins 11 Steam Supply Line A Ruptures 09E1042AMJ          Between Reducer and V-127           5 x 10~11        1 x 10-10/hr Steam Bypass Inlet Line Ruptures Q9A1035AMJ Between Steam Supply Line A         5 x 10-10        1 x 10~9/hr s             and V-171~
              -                 Steam Supply Line A Ruptures
          Q;)F1008AMJ.        Between Main Steam Line A and       5 x 10~11        1 x 10~10/hr Reducer Q9K1SLIAMJ          Main Steam Line A Ruptures          5 x 10-11        1 x 10-10/hr Steam Supply Line B Ruptures        5 x 10-11 Q9F1143BMJ Between V-96 and Reducer                             1 x 10-10/hr l

B-17

7._. i TABLE 8.2 (Continued)- FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION

                   .o DESCRIPTION                 tmAVAILABILITY    FAILURE RATE I                ER Steam Bypass Exit Line Ruptures QSA1136AMJ            8etween V-172 and Steam Supply            5 x 10-10        1 x 10-8/hr Steam Supply Line 8 Ruptures Be-
                 @E1142AMJ              tween Reducer and V-128                  5 x 10-11        1 x 10-10/hr steam Bypass Inlet Line Ruptures QSA1135AMJ            8etiseen V-172 and Steam Supply           5 x 10-10'       1 x 10-I/hr Line 8                                .

steam Supply L1ne 5 Ruptures QSF1108AMJ 8etweer. Main Steam'Line 8 and 5 x 10-11 1 x 10-10/hr Reducer-QSKISL18MJ Main Steam Line'8 Ruptures 5 x'10-11 1 x 10-10/hr Feed Pump P-37A' Suction Line ' F9G1081AMJ Ruptures Between V-15 and Pump 5 x 10-11 1 x 10-10/hr Ynla > Feed Pump P-37A Suction Line F9G10818MJ Ruptures Between V-154 and V-155 5 x 10-11 1 x 10-10/hr Feed Pump P-37A Suction.Line F9G1081CMJ Ruptures Bet:veen Condensate Tank. 5 x 10-11 1 x 10-10/hr and V-1M Feed PunipP-378 Discharge Line MBE1612CMJ Ruptures Between P-378 and V-70 5 x 10-gi 1 x 10-10/hr Feed Pump P-378 Recirc.' Line M901612AMJ Ruptures Between V-73 and Pump 5 x 10-11 1 x 10-9/hr Discharae Line Feed Pump P-378 Suction Line F9G1082AMJ Ruptures Between Y-159 and '5 x 10-11 1 x 10-10/hr Pump Inlet Feed Pump P-378 Suction Line F9G1062BMJ Ruptures Between V-158 'and V-159 5 x 10-11 1 x 10-10/hr Feed Pump P-378 Suction Line - , F9G1082CMJ Ruptures Between Condensate Tank 5 x 10-11 , 1 x 10-10/hr

ma v.iu I Turbine Driven Feed P.c.p-37A MP81T37AME Falls To Start 8.4 x 10-3 8.4 x 10-3/d Motor Driven -Feed Pump P-378 MP81D378ME Fails To Start 2.4 x 10-3 2.4 x 10-3/d Turbine Driven Feed P-37A Out
MP81T37A99 of Service 4.2 x 10-4 Motor Driven Feed Pump P-378 Out MP81037899 Of Service ,

9.4 x 10,4 8-18

    +---e-           -,r, -w-___

l TABLE 3.2(Continued) FAULT IDENTIFIERS FOR THE SEA 8 ROOK EERGENCY FEED STATION I ER DESCA M I M MAMILIU FAME RAM Flow Control Valve FV-4214 Out MVD14214N of Service 8.5 x 10 4 MVD1V75AN Valve Out of Service 8.5 r. 10 4 Flow Control Valve FV-4244 Out MVD14224N of Service 8.5 x 10 4 MVD1V87DN Valve V-87 Out of Service 8.5 x 10 Flow Control Valve FV-4234 Out MVD14234N of Service 8.5 x 10 4 MVDIV93CN Valve V-93 Out of Service 8.5 x 10 4 , Flow Control Valve FV-4224 Out i MVD14224N of Service 8.5 x 10 4 MVD1V81BN Valve V-81 Out of Service 8.5 x 10d MVD1V65AN isolation Valve V-65 Out of Servico 2.0 x 10~3 i Steam Supply. Valve V-129 Out cf QVD1V129N Service 1.0 x 10 4 QMVIV95AN Manual Valve V-95 Out of Service 0.0 Steam Supply Valve V-127 Out of QVXIV127N Service 8.7 x 10,4 Steam Supply Valve V-128 Out of QVXIV128N Service 8.7 x 10'4 P-37A Suction Valve V-155 Out of FVM1V155N Service 0.0 p-37A Suction Valve V-154 Out of FVM1V154N Service 0.0 Isolation Valve V-125 Out of MV01V1299 Service 0.0 Isolation Valve V-126 Out of MVD1V126M Service 0.0 B-19

l TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE ID ER Isolation Valve V-127 Out of MVD1V12799 Service 0.0 Isolation Valve V-71 Out of MVD1V71B99 Service 2.0 x 10-3 P-378 Suction Valve V-159 Out FVM1V15999 of Service 0.0 P-378 Suction Valve V-158 Out FVM1V15809 of Service 0.0 Operator Fails to Open Steam QVXIV1279A Supply Valve V-127 5 x 10-3 5 x 10-3/d Operator Fails to Open Steam QVXIV1289A Supply Valve V-128 5 x 10~3 5 x 10-3/d Operator Fails to Close Feedwater IWDIV39ABB Supply Line A Isolation Valve V-30 5 x 10-3 5 x 10-3/d Operator Fails to Close Flow MVD1422498 Control Valve FV-4224 5 x 10-3 5 x 10-3/d Operator Fails to Close Flow -3 MVD1422498 Control Valve FV-4234 5 x 10-3 5 x 10 /d Operator Fails to Close Flow MVD1424498 Control Valve FV-4244 5 x 10-3 5 x 10-3/d Operatt,r Fails to Close Feedwater MVDIV57D9B Supply Line D isolation Valve V-57 5 x 10-3 5 x 10-3/d Operator Fails to Close Flow MVD1421498 Control Valve FV-4214 5 x 10-3 5 x 10-3/d Operator fails to Close Feedwater MDV1V48C98 Supply Line C Isolation Valve V-48 5 x 10~3 5 x 10-3/d Operator Fails to Close Feedwater MVD1839898 Supply Line B Isolation Valve V-39 5 x 10-3 5 x 10-3/d Operator Fails to Close Feedwater MVD1V52098 Flow Control Valve FCV-510 5 x 10-3 5 x 10-3/d Operator Fails to Close Supply MVD1V12508 Header Isolation Valve V-125 .9 .9 Operator Fails to Close Feedwater MVD1V54008 Flow Control /alve FCV-540 5 x 10-3 5 x 10-3/d B-20

TABLE 8.2(Continued) FALA.T IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION IDE ER ESCRIPT M MAMRm FAME MTE Operator Fails to Close Supply MVD1V12698 Header Isolation Valve V-126 .9 .9 . Operator Falls to Close Feedwater MVD1V53098 Flow Control Valve FCV-530 5 x 10~3 5 x 10~3/d 1 Operator Fails to Close Supply mV1V12798 Header Isolation Valve V-127 .9 .9

Operator Falls to Close Feedwater ' MVD1V52098 Flow Control Valve FCV-520 5 x 10-3 5 x 10-3/d Operator Fails to Close P-378 MVD1V65A98 Discharge Isolation Valve V-65 .9 .9 Operator Falls to Close P-37B MVD1V71898 Discharge Isolation Valve V-71 .9 .9 Operator Fails to Restore Manual MVM1V1529C Valve V-152 1 x 10-3 1 x 10-3/d Operator Fails to Restore Manual MVM1V1539C Valve V-153 1 x 10~, 1 x 10-3/d Operator Fails to Restore Manual MVM1V154pc Valve V-154 1 x 10-3 2 x 10-3/d Operator Fails to Restore Manual MVM1V1559C Valve V-155 1 x 10-3 1 x 10-3/d 61CIAFSA9C Operator Defeats Train A"S" Signal 1 x 10 1 x 10-4/d 6ICIAFSB9C Operator Defeats Train B"S" Signal 1 x 10 4 1 x 10-4/d i Operator Inadvertently Closes MVD1421490 Flow Control Valve FV-4214 1 x 10~4 1 x 10-4/d operator Inadvertently closes MVD1V75A90 Valve V-75 1 x 10-4 1 x 10'4/d Operator Inadvertently Closes MVM1424490 Flow Control Valve FV-4244 1 x 10~4 1 x 10 4/d F Operator Inadvertently Closes MVD1V87D90 Valve V-87 1 x 10'4 1 x 10 4/d Operator Inadvertently Closes MVD1423490 Flow Control Valve FV-4234 1 x 10'4 1 x 10,4/d B-21

i L TABLE B.2 (Continued) 1 L FAULT IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION L I ER KSCMM MAMRIU FAI N M Operator Inadvertently Closes MVD1V93C90 Valve V-93 1 x 10 4 1 x 104 /d Operator Inadvertently Closes 1 x 104 /d MVD1422400 Flow Control Valve FV-4224 1 x 10_4 ! Operator Inadvertently Closes l MVD1V81890 Valve V-91 1 x 10 4 4 1 x 10 /d

Operator Fails to Restore P-37A

MVM1V65A90 - Isolation Valve V-65 1 x 10 4 1 x 10_4/d Operator Fails to Restore Manual

! QVD1V95A90 Valve V-95 1 x 10-3 1 x 10-3/d l Operator Fails to Restore P-37A FMV1V15590 Suction Valve V-155 1 x 10-3 1 x 10-3/d j Operator Fails to Restore P-27A FVM1V1549D Suction Valve V-154 1 x 10-3 1.x 10-3/d 1 Operator Fails to Restore Supply

MVD1V1259D Header Valve V-125 1 x 10~3 1 x 10-3/d j

Operator Fails tc Restore Supply MVD1V1259D Header Valve V-126 1 x 10~3 1 x 10-3/d Operator Fails to Restore Supply MVD1V12790 Header Valve V-127 1 x 10~3 1 x 10-3/d Operator Fails to Restore P-37B MVD1V71890 Discharge Isolation Valve V-71 1 x 10 4 4 1 x 10 /d l l Operator Fails to Restore P-37B FVM1V1599D Suction Valve V-159 1 x 10~3 1 x 10-3/d Operator Fails to Restore P-378 FVM1V15890 Suction Valve V-158 1 x 10-3 1 x 10-3/d Operator Fails to Start Motor MPB1D37BSE - Drisen Pump P-37B 1 x 10-3 1 x 10-3/d

Operator Turns Off Turbine Driven l- MPB1T37A9G Pump P-37A 1 x 10~4 1 x 10_4/d l

Operator Turns Off Motor Driven MPB1037B9G Pump P-37B 1 x 10 4 4 1 x 10 /d Circuit Breaker to Motor Driven MCA2037BMK Pump P-37B Open 1.5 x 10-6 4.2 x 10-9/hr B-22

l- TABLE B.2 (Continued) , l l FAULT IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION

                                 . DESCRIPTION                          tmAVAILABILITY      FAILURE RATE I            ER Control Circuit to Steam Supply QRA2V127MK       Valve V-127 Open                                  6.5 x 10 4          9.1 x 10~7/hi Control Circuit to Steam Supply QRA1V121MK       Valve V-128 Open                                  6.5 x 10 4          9.1 x 10~7/hi Train A Control Circuit to Motor IBtA1A37BMK      Pump P-37B Open                                   5.9 x 10-3          9.1 x 10~7/hi Train B Control Circuit to Motor IGLA18378K      Pump P-378 Open                                   5.9 x 10-3          9.1 x 10'7/hi Flow Control' Valve FCV-510 Flow                                                     i MCE1V510MK      Control Switch Open                                1.5 x 10-8         3 x 10-8/hr Flow Control Valve FCV-540 Flow 3       MCE1V540MK      Control Switch Open                                1.5 x 10-8         3 x 10-8/hr Flow Control Valve FCV-530 Flow MCE1V530Mr. Control Switch Open                                1.5 x 10-8         3 x 10-8/hr Flow Control Valve FCV-520 Flow MCE1V520MK      Control Switch Open                                1.5 x 10-8         3 x 10-8/hr Steam Supply Valve V-127 Switch                                                       '

QCE1V127MK Open 2.2 x 10'5 3 x 10-8/hr P-37B Motor Controller Circuit MCE1037BMK Open 3.3 x 10 d 9.1 x 10-7/hr MCK1D37BMK P-378 Motor Starter Circuit Open 1.2 x 10-3 1.2 x 10-3/d Turbine Driven Feed Pump P-37A MPBIT37ALB Lubrication Failure 0.0 Motor Driven Feed Pump P-37B MPBID378LB Lusbrication Failure 0.0 LOSP Loss of Station Power 7 x 10-6 1.4 x 10-5/hi STli1TK25MJ Condensate Storage Tank Ruptured 5 x 10-11 1 x 10-10/hr 61CIFWIAMN No Train A Feedwater Isolation 5.8 x 10-3 5.8 x 10-3/d Cinnal No Signal From FE-4224 to Flow MIF14224MN Control Valve FV-4224 2.0 x 10-3 3.1 x 10-7/hi' 8 . 'B-23

          . _ . ,                                                                .                   .             --   ..       .. _~.   -.    ,

k- .

. TABLEB.2(Continued)

FAULT IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION DESCRIPTION I fER UNAVAILABILITY FAILURE RATE No Signal From FE-4234 to Flow MIF14234M Control Valve FV-4234 2.0 x 13-3 3.1 x 10~7/hr No Signal From FE-4224 to Flow MIF14244M Control Valve FV-4244 2.0 x-10-3 3.1 x 10-7/hr FE-4214 to Flow MIF14214M a FV-4214 2.0 x 10-3 3.1 x 10-7/hr No Train B Feedwater Isolation 61C1FW8MN Signal 5.8 x 10-3 5.8 x 10-3/d No Signal From FE-510 to Flow MIF1V510MN Control Valve FCV-510 1.5 x 10-7 3.1 x 10'7/hr No Signal From FE-540 to Flow MIF1V540MN Control Valve FCV-540 1.5 x 10-7 3.1 x 10~7/hr No Signal From FE-530 to Flow MIF1V530MN Control Valve FCV-530 1.5 x 10~7 3.1 x210-7/hr , No Signal From FE-520 to Flow MIF1V520MN Control Valve FCV-520 1.5 x 10-7 3.1 x 10~7/hr No Signal From Safety Injection 6ICIAFSAM Signal Train A 5.8 x 10-3 5.8 x 10-3/d No Signal From Safety Injection 6ICIAFSBNM Signal Train B 5.8 x 10-3 5.8 x 10-3/d Spurious Signal to Flow Control MVD14214M9 Valve FV-4214 1.2 x 10 -8 1.2 x 10~8/d Spurious Signal to Flow Control [ MVD14244M9 Valve FV-4244 1.2 x 10~0 1.2 x 10-8/d j Spurious Signal to Flow Control MVD14234M9 Valve FV-4234 1.2 x 10-8 1.2 x 10-8/d Spurious Signal to Flow Control MVD14224M9 Valve FV-4224 1.2 x 10-8 1.2 x 10-g/d REC 118--MJ 125 DC Bus 118 Shorts to Ground 3.5 x 10 4 7 x 10-8/hr REC 1E612MJ 460 V AC Bus E612 Shorts to Ground 3.5 x 10-8 7 x 19~0/hr REC 1E61-MJ 480 V AC Bus E61 Shorts to Ground 3.5 x 10-8 7 x 10-8/hr B-24

       -r         . - - - -       , .,.     .-     ,-,m - _--- - . _ _ - . . , - _ . _ _ - , _ - ,

                                    ;     . .. 8MIC #*'

FAULT IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE ID ER

 ,            g        Diesel Generator DG-1B Circuit                                                       1 x 10-3                                                 1 x 10-3/d Breaker A74 Fails to Close Crossover Circuit Breaker DN6 RCA1DN6-M             Fails to Close                                                                       1 x 10-3                                                 1 x 10~3fd RCA2A61--MC           Circuit Breaker A61 Open                                                             2.1 x 10~I                                               4.2 x10~9/hr RCA1A62--MF           Circuit Breaker A62 Fails to Close 1 x 10-3                                                                                                  -1 x 10-3/d RCA1DN4-MC            Circuit Breaker DN4 Open                                                             2.1 x 10-8                                               4.2 x 10~9/hr RCA1DNS-MC            Circuit Breaker DNS Open                                                             2.1 x 10~9                                               4.2 x 10~I/hr RCA1DAl-MC            Circuit Breaker DA1 Open                                                             2.1 x 10~9                                               4.2 x 10~9/hr 4

RCA1AD6-MC Circuit Breaker AD6 Open 2.1 x 10~9 4.2 x 10~9/hr RCA1AD2-MC Circuit Breaker AD2 Open 2.1 x 10~9 4.2 x 10~9/hr RCA1A41-MC Circuit Breaker A41 Open 2.1 x 10~9 4.2 x 10~9/hr RCA1A41-MJ Circuit Breaker A41 Shorts to 3.5 x 10-8 7 x 10-8/hr Ground RCA1A42-MB Circuit Breaker A42 Fatis to Close 1 x 10-3 1 x 10-3/d Circuit Breaker A42 Shorts to RCA1A42-MJ Ground 3.5 x 10-8 7 x 10-8/hr RCA1A61-MJ Circuit Breaker A61 Shorts to 3.5 x 10-8 7 x 10-8/I.r Ground i RCA1A62-MJ Circuit Breaker A62 Shorts to 3.5 x 10-8 7 x 10-8/hr Ground RCA1DNS-MJ Circuit Breaker DNS Shorts to 3.5 x 10 4 7 x 19"I/hr Ground l RCA1DAl-MJ r t Breaker DA1 Shorts t 3.5 x 10-8 7 x 10-8/hr B-25

                                                                                 .z  .
    - , , , , ,      -       -          - - , - - - . - - - - . - - - - - . - -_        - - - - . - - - - - - - - - - - . - - , - _ , , - - - - - - - - . - - - -              --e --- - - - - - - - - -

l TABLE B.2.(Continued) FAULT IDENTIFIERS FOR THE SEA 8 ROOK EERGENCY FEED STATION ID ER DE N T M MANIW FARURE M Circuit Breaker AD6 Shorts t RCA1AD6--N Ground 3.5 x 10 4 7 x 10-8/hr RCA W 2-N Circuit Breaker AS2 Shorts to Ground 3.5 x 10 4 7 x 10-8/hr Circuit Breabr A75 Shorts to RCA1A75-MJ Ground 3.5 x 10-8 7 x 10-8/hr

      'RBC118--MC    Battery Charger 18 Opens             2.1 x 10~I      4.2 x 10'9/hr RBC11B--MJ Battery Charger 18 Shorts to Ground                               3.5 x 10-8      7 x 10-8/hr RBA11B--MJ   Battery 18 Shorts to Ground           3.5 x 10-8      7 x 10-8/hr RBA110--MJ   Battery 10 Shorts to Ground           3.6 x 10-8      7.x 10-8/hr RBA118--m    Battery IB Undercharged               1.3 x 10-6      3 x 10-6/hr RBA11D--M     Battery ID Undercharged               1.5 x 10-6      3 x 10-6/hr RTRIX-5 CMC   Transformer X-Sc Open                 5 x 10-7        1 x 10-6/hr RTRIX-2BMJ     Transfonner S-28 Shorts               5 x 10-7        1 x 10-6/hr

_c RTRIX-2BMC Transformer S-28 Opens 5 x 10~7 t 1 x 10-6/hr RTRIX-5CMJ Transformer X-5C Shorts 5 x 10-7 1 x 10-6/hr RTRIX-3BMC Transfonner X-3B Opens 5 x 10~7 1 x 10-6/hr i RTRIX-3BMJ Transfonner X-38 Shorts 5 x 10~7 1 x 10-6/hr L 1 RG011B--ME Diesel Generator DG-1B Fails i to Start 1.0 x 10-2 l 1.0 x 10-2/d ) e Generator DG-1B Fails RG0118--MG 3.0 x 10-3 6.0 x 10'3/hr I B-26

L TA8LE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEA 8 ROOK EERGENCY FEED STATION i DESCRIPTION UNAVAILABILITY FAILURE RATE ID ER J Diesel Generator DG-18 Out of 7 x 10 RG0118--99 Service es Generator DG-1A Out of 7 x 10-4 RGD11A--99 g NIX 1TBRAMJ S.G.A Steam Line Rupture 5 x 10-11 1 x 10-10/hr MHX1SHRAMJ fS.G.S} hell Rupture 5 x 10-11 1 x 10-10Mr MHX1TBRDMJ S.G.D Steam Line Rupture. 5 x 10-11 1 x 10-10/hr MHX1SHRDMJ S.G.D Shell Rupture 5 x 10-11 1 x 10-10/hr MHX1TBRCMJ S.G.C Steam Line Rupture 5 x 10-11 l'x 10-10/hr MHX1SHPCMJ S.G.C Shell Rupture 5 x 10-11 1 x 10-10/hr MHX1TBRBMJ S.G.8 Steam Line Rupture 5 x 10-11 1 x 10-10/hr MMX1SHRBMJ S.G.8 Shell Rupture 5 x 10-11 1 x 10-10/hr Turbine Driven Feed Pump P-37A 0.0 MP81T37ACL Cooling Loss Motor Driven Feed Pump P-378 MPB1D378CL 0.0 Cooling Loss MTU1TD-2ME Turbine Fails to Start 0.0 Control Relay on Solenoid Valve QRA1C127MA to Steam Supply Valve V-127 Fails 3.1 x 10'6 3.1 x 10~6/d tn nnan Cont'r ol Relay on Solenoid Valve QRA1V128MA to Steam Supply Valve V-128 Fails 3.1 x 10-6 3.1 x 10~6/d en nn.n Solenoid Valve on Steam Valve to QVLIV127MA Steam Supply Valve V-128 Fails 1.4 x 10 -6 1.4 x 10-6/d to Onen Solenoid Valve on Steam Supply QLV1V128MA Valve V-128 Fails to Open 1.4 x 10~6 1,4 x 10~6/d B-27

TABLE 8.2 (Continued) FAULT IDENTIFIERS FOR THE SEA 8 ROOK EERGENCY FEED STATION ID ER MA In FAME WE Safety Valve V-6 on Steam Line QVB1V06AMC A Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-7 on Steam Line

QV81V07AMC A Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-10 on Steam Line QVB1V10AMC A Opens 5 x 10~6 1 x 10-5/hr Safety Valve V-8 on Steam Line QVB1V08AMC A Opens 5 x 10~6 1 x 10-5/hr Safety Valve V-9 on Steam Line QVB1V09AMC A Opens 5 x 10-6 1 x 10-5/hr Relief Valve on Main Steam Line QVBISGARMC A Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-50 on Steam Line QVB1V50DMC D Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-51 on Steam Line QVB1V51DMC D Opens 5 x 10~0 1 x 10-5/hr Safety Valve V-52 on Steam Line QVB1V520MC D Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-53 on Steam Line QVB1V530MC D Ooens 5 x 10-6 1 x 1.0-5/hr Safety Valve V-54 on Steam Line QVB1V54DMC D Opens 5 x 10-6 1 x 10-5/hr Relief Valve on Main Steam Line QVB1SGDRMC D Cpens 5 x 10-6 1 x 10-5/hr Safety Valve V-36 on Steam Line QVB1V36 CMC C Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-37 on Steam Line QVB1V37 CMC C Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-38 on Steam Line j QVB1V38 CMC C Opens 5 x 10-6 1 x 10-5/hr i Safety Valve V-39 on Steam Line i QVB1V39 CMC C Opens 5 x 10-6 1 x 10-5/hr i Safety Valve V-40 on Steam Line QVB1V49 CMC C Opens 5 x 10-6 1 x 10-5/hr i B-28

l TABLE B.2 (Continued) l FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEF.D STATION i DESCRIPTION UNAVAILABILITY FAILURE RATE IDE ER Relief Valve on Main Steam Line QVBISGCRMC C Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-22 on Steam Line QVB1V228MC B Opens 5 x 10~6 1 x 10-5/hr Safety Valve V-23 on Steam Line QVB1V23BMC B Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-24 on Steam Line QVB1V24BMC B Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-25 on Steam Line QVB1V25BMC B Opens 5 x 10~6 1 x 10-6/hr Safety Valve V-26 on Steam Line QVB1V26BMC B Opens 5 x 10-6 1 x 10-6/hr Relief Valve on Main Steam Line QVBISGBRMC B Opens 5 x 10-6 1 x 10-5/hr Loss of Voltage on S.G.A Relief SECISGARM Valve Controller 7 x 17-7 1.4 x 10~0/hr Loss of Voltage onS.G.B Relief SECISGBRM Valve Controller 7 x 10-7 1.4 x 10-6/hr Loss of Voltage on S.G.A Relief 9ECISGCRM Valve Controller 7 x 10~7 1.4 x 10~0/hr

Loss of Voltage of 5.G.D Relief DECISGDRM Valve Controller 7 x 10~7 1.4 x 10-6/hr MVD1V87098 Operator Fails to Close V-87 5 x IC-3 5 x 10-3/d MVD1V93C98 Operator Fails to Close V-93 5 x 10~3 5 x 10-3/d MVD1V81B98 Operator Fails to Close V-81 5x17 3 5 x 10-3/d

                                                                               -3 MVD1V75ABB    Operator Fails to Close V-75          5 x 10-3      5 x 10 /d MVD1VS7DMB     Valve V-87 Does Not Close            2 x 10-3      2 x 10~3/d MVD1V93CMB     Valve V-93 Does Not Close            2 x 10-3      2 x 10-3/d B-29

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION 4 DESCRIPTION UNAVAILABILITY FAILURE RATE ID ER MVD1V81BMS Valve V-81 Does Not Close 2 x 10-3 2 x 10-3/d I M@lV75AMB Valve V-75 Does Not Close 2 x 10-3 2 x 10-3/d MVD1V870MC Valve V-A7 Fails to Remain Closed 6 x 10-8 1.2 x 10-7/hr MVD1V93 CMC Valve V-93 Fails to Remain Closed 6 x 10-8 1.2 x 10~7/hr MVD1V81BMC Valve V-81 Fails to Remain Closed 6 x 10-8 1.2 x 10~7/hr MVD1V75AMC Valve V-75 Fails to Remain Closed 6 x 10-8 1.2 x 10-7/hr MIFIV87DMN Valve V-87 Fails to Receive Signal 2 x 10-3 2.x 10-3/d , MIF1V93CMN Valve V-93 Fails to Receive Signal 2 x 10-3 2 x 10-3/d MIF1V81BMN Valve V-81 Fails to Receive Signal 2 x 10-3 2 x'10~3/d f MIF1V75AMN Valve V-75 Fails to Receive Signal 2 x 10-3 2 x 10-3/d Rupture of 6" Line Between V-99 M9F1618CMJ and Tee 5 x 10~11 1 x 10-10/hr Rupture of 6" Line Between V-163 MSF1618BMJ and Tee 5 x 10~11 1 x 10-10/hr Rupture of 6." Line Between V-163 M9F1618AMJ and V-156 5 x 10~I1 1 x 10-10/hr Rupture of 6" Line Between P-113 MSF1618DMJ and V-99 5 x 10~11 1 x 10-10/hr Rupture of 4" Line Between Dis-M9E1631AMJ charge Pipe and FW V-156 5 x 10~11 1 x 10-10/hr l Rupture of 4" Line Between Dis-M9E1631BMJ charge Pipe and FW V-159 5 x 10~11 1 x 10-10/hr i Rupture of 4" Line Between Dis-l M9E1631CMJ charge Pipe and FW V-162 5 x 10~11 1 x 10-10/hr { B-30 l

TABLE B.2 (continued) FAULT IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION FAULT DESCRIPTION UNAVAILABILITY FAILilRE RATE IDENTIFIER Rupture of 6" Line Between Dis-M901625AMJ charge Pipe and PCV-4326 5 x 10-11 1 x 10-10/hr Rupture of 16" Condensate Line FST1'079AMJ Between Tank and V-141 5 x 10-11 1 x 10-10/hr Rupture of 24" Condensate Line , FSK1080AMJ Between V-143 and V-141 5 x 10-11 1 x 10'IO/hr Rupture of 24" Line Betwen F9K1080BMJ Suction Line and V-142 5 x 10~11 1 x 10-10/hr Rupture of 20" Suction Line FSJ1080CMJ Between V-143'and Tee 5 x 10-11 1 x 10-10/hr Rupture of 20" Suction Line F9J1080DMJ Between Tee and V-145 5 x 10~11 1 x 10-10/hr Rupture of 8" Line Between Tee F9G1080AMJ and V-340 5 x 10~11 1 x 10-10/hr Rupture of 8" Line Between V-340 F9F1080BMJ and V-152 5 x 10~I1 1 x 10-10/hr Rupture of 8" Line Between V-152 F9G1080CMJ and P-113 5 x 10-11 1 x 10-10/hr MVAIV99AMJ Rupture of V-99 5 x 10-9 1 x 10-8/hr MVP1V163MJ Rupture of V-163 5 x 10-9 1 x 10-8/hr MVM1V156MJ Rupture of FWV-156 1 x 10-8 2 x 10-8/hr MVM1V159MJ Rupture of FWV-159 1 x 10-8 2 x 10 '"/ii, Rupture of Bypass Inlet Line F9F10900MJ Between Tee and V-341 5 x 10-10 1 x 10-10/hr

           - FVM1V152MJ        Rupture of V-152                                1 x 10-8          2 x 10-8/hr FVA1V340MJ        Rupture of V-340                                5 x 10~9          1 x 10-8/hr FVM1V142MJ        Rupture of V-142                                1 x 10-8          2 x 10-8/nr 1

B-31

 -- - e..,     -

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR TME SEABROOK EERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE IDE ER i FVM1V341N Rupture of V-341 1 x 10-8 2 x 10-8/hr MVD14326N Rupture of PCV-4326 5 x 10-8 1 x id-7/hr FVD1V145MJ Rupture of V-145 5 x 10 -9 1 x 10-8/hr i FVA1V343MJ Rupture of V-343 5 x in-9 1 x 10-8/hr Rupture of Bypass Outlet Line F9G1080EMJ Between V-341 and Tee 5 x 10-10 1 x 10-10/hr Rupture of Bypass Outlet Line F9G109FMJ Between V-344 and Tee 5 x 10-10 1 x 10-10/hr Rupture of 8" Line Between V-344 F9F1080GMJ and V-343 5 x 10-10 1 x 10-10/hr FMJ1V344MJ, Rupture of V-344 1 x 10-8 2 x 10-8/hr MVM1V162MJ Rupture of FWV-162 1 x 10-3 2 x 10-8/hr 4 Rupture of P-113 Suction Iso- . FVD1V143MJ 1ation Valve V-143 5 x 10*I 1 x 10-8/hr Rupture of P-113 Suction . FVD1V141MJ Isolation Valve V-141 5 x 10-9 1 x 10-8/hr I

                                                                                                            -4 MVA1V99-MA               Check Valve V-99 Fails to Open                        2 x 10-4          2 x 10 /d i

Manual Isolation Valve V-152 Fail s FVM1V152MD to Remain Open (Plugs) 1 x 10-4 1 x 10-4/d Isolation Valve V-143 Fails to FVD1V143MD Remain Open 1 x 10-4 1 x 10-4/d , Isolation Valve V-143 Falls

i. FVD1V141MD to Remain Open 1 x 10-4 1 x 10-4/d MVD1V1639A Operator Fails to Open V-163 1 x 10-2 1 x 10-2/d MVD1V1560A Operator Fails to Open V-156 1 x 10-2 1 x 10-2/d B-32

TABLEB.2(Continued) FAULT IDENTIFIERS FOR THE SEABROOK EERGENCY FEED STATION DESCRIPTION IRAVAILABILITY FAILURE RATE I ER FVM1V15290 Operator Fails to Restore V-152 1 x 10-3 1 x 10-3/d FVD1V14390 Operator Fails to Restore V-143 1 x 10-3 1 x 10-3/d FVD1V1419D Operator Fails to Restore V-141 1 x 10-3 1 x 10-3/d . MPB1P11399 P-113 Out of Service 4.2 x 10 MPS1P16199 P-161 Out of Service 5 x 10-4 MPS1P113ME P-113 Fails to Start 4 x 10-3 4 x 10-3/d MP81P161ME P-161 Fails to Start 4 x 10-3 4.x 10-3/d Circuit Breaker 1E42 Shorts RCA1E42-MJ to Ground 3.5 x 10-8 7 x 10-8/hr Operator-Fails to Close Circuit RCA1A93-98 Breaker A-93 1 x 10-2 1 x 10-2/d Circuit Breaker A-93 Fails RCA1A93-MB to Close 1 x 10-3 1 x 10-3/d Diesel Generator 1A Circuit RCA1A54-MB Breaker A-54 Fails to Close 1 x 10-3 1 x 10-3/d Diesel Generator A Fails RG01DGIAME to Start 1.0 x 10-2 1.0 x 10-2/d RGD1DGIAMG Diesel Generator 1 A Fails to Run 3.0 x 10-3 6.0 x 10-3/hr 61CIFPSAMN No Actuation Signal Generated 5.8 x 10-3 5.8 x 10-3/d Startup Feedpump Suction Isolatio-i FVD1V14399 Valve V-143 Out of Service 1.2 x 10~4 Startup Feedpump Suction Isola-FVD1V14199 tion Valve V-141 Out of Service 0.0 B-31 - - - . .--~_ -__ _ _____- - - - _ - . - - . . __

i I l I l TABLE B.3 NRC-SUPPLIED DATA USED FOR PURPOSES OF CONDUCTING A LUMPARAilVL A55L53 MENT OF EXISTING AFWS DESIGN 5 AND INEIR POTENiiAL RELIABILITIES Point Value Estimata of Probability of* Failure on Denand I. Comoonent (Haniware) Failure Data , a. Yalves: Manual Valves (Plugged) ~1 x 10-4 Check Valves ~1 x 10-4 Motor-Operated Valves

                   -   Mechanical Components                                  ~1 x 10-3
                   -   Plugging Contribution                                  ~1 x 10-4 Control Circuit (Local to Valve) w/Quartarly Tests                                  ~6 x 10-3 w/ Monthly Tests                                   ~2 x 10-3

b. Pumos: (1 Punp)

Mechanical Components ~1 x 10-3 Centrol Circuit

                  -    w/ Quarterly Tests                                     ~7 x 10-3
                  -    w/?4onthly Tests                                       ~4 x 10-3

c. Actuation Locic ~7 x 10-3 i

l

irror f actors of 3-10 (up and dcwn) abcut such values are not unexpected for basic data uncartsf aties.

i L l B-34

                                              .      - . .      . - - . -      . ~ . - - - - . _. - _ . .-

TABLE B.3 (Cont'd) II. Test and Maintenance Outage Contributions:

a. Calculational Approach

1. Test Outage Q

RST ( hrs / test) ( tests / year) hrs / year

2. Maintenance Outage

Q MAINT. (0.22)( hrs /maint. act)

/ZO

b. Data Tables for Test and Maint. Outages

SUM 4ARY OF TEST ACT DURATION Calculated Range on Test Mean Test Act Component Act Duration Time, hr Duration Time, Dt , hr Pumps 0.25 - 4 1.4 Valves 0.25 '2 0.86 Diesel s 0.25.- 4 1.'4 Instrumentation 0.25 - 4 1.4 LOG-NORMAL MODELED MAINTENANCE ACT DURATION Calculated Range on Maintenance Mean Maintenance Act Component Act Duration Time, hr Duration Time, 0t , hr Pumps 1/2 - 24 7 1/2 - 72 19 Valves 1/2 - 24 7 Diesels 2 - 72 21 Instrtmentation 1/4 - 24 6 Note: These data tables were taken from the Reactor Safety Study (WASH-1400) for purposes of this AFW system assessment.

Where the plant technical 3pecifications placed limits on the outage duration (s) allowed for AFW system trains, this tech spec limit was used to estimate the mean duration times for maintenance. In general, it was found that the outages allowed for maintenance dominated those contributions to AFW systen unavailability fran outages due to testing. l 4 B-35

TABLE B.3 (Cont'd) III. Human Acts & Errors - Failure Data: Estimated Human Error / Failure Probabilities Modifying Factprs & Situations' With Valve Position With Local Walk-Around & . W/0 Either Indication in Control Room Double Check Procedures Point Value Est Est. on Point Value Est' Est on Point Yalue 'Est On Error l Error Estirate Error Factor Factor Factor

a. Acts & Errors of A Pre-Accident Nature

1. Valves Mispositioned During Test /Maint (a) Specific Single Valve Wrongly

, cm Selected out of A Population L of Valves Durir.g Conduct of a cn Test or Maintenance Act (X No. 1 y 10

                                                                           -2    1 y, ?
                                                                                                          -2 I X 10 y1                      10-2 gI of Valves in Population at Choice)        Td                     20        2           X       ' 10                Y        10 (b) Inadvertently Leaves Correct                          4                                3                          2 Valve in Wrong Position                     5 x 10               20          5 x 10              10         10              10
                                                                                                            ~3                             ~3

2. More than one valve is affected 1 x 10 ~4 20 1 x 10 10 3 x 10 10 (coupled errors)

3. Miscalibration of Sensors / Electrical Relays

                                                                                                                                       -2 (a) One Sensor / Relay Affected                 -                     -

5 x 10~3 10 10 10 i (b) More than one Sensor / Relay 3 3 Affected - - 1 x 10 10 3 x to 10

s Time Actuation Needed Estimated Failure Estimated Failure Overall Estimated Prob. for Primary Prob. of Jther Estimate Error Factor Operator to (Backup) Control of Failure on Overall Actuate AFWS Rs. Operator to Probability Probability Actuate AFWS

b. Acts & Errors of a Post-Accident Nature

1. Manual Actuation of AFW system from Control Room f

(a) Considering " Dedicated" Operator 15 min 2x10f 2x10'l 10 to Actuate AFW system and Possible 15 min. 1 x 10,4 0.5 (mod. dep.) 5 3 10 4 10 . Backup Actuation of AFWS 30 min. .5 x 10 .25 (low dep.) 10 10 W

 .E y

(a) Considering "Non-Dedicated" Operator to Actuate AFW system 5 min. 15 min. 5x10ll I xx 10 0.5 (mod. dep.) 5x10ll 10 10 10 h

s and Possible Backup 30 min. 5 10 3 . 25 (low dep.) 5 33

                                                                                                            '10        10           q Acutation of AFW system i

3

APPENDIXJ BACKGROUND INFORMATION PROVIDED BY THE APPLICANT BACKGROUND The action plan developed by the NRC in response to the accident at the Three Mile Island Unit-2, NUREG-0737, required (Item II.E.1.1.) that all op-erating nuclear power plants or plants applying for operating licenses conduct a reliability analysis of the auxiliary feedwater (AFW) system. The analysis is to be performed using event-tree and fault-tree logic techniques and is in-tended to evaluate the potential for system failure during a variety of loss of main feedwater transients. The primary purpose of the reliability evaluation is to identify potential failures resulting from human errors, common causes, single-point vulnerabilities, and outages due to test and maintenance. The stated purpose of the recommendations associated with TMI Action Plan ItemII.E.lgwastodecreasetheunreliabilityofAFWsystemstowardsagoalof 10-4 to 10- per demand for loss of main feedwater and loss of offsite power transients. As a result of reliability evaluations performed both by the NRC staff (NUREG-0611) and-various operating license applicants, it was deemed by the staff that three AFW pumps were necessary to achieve the desired unreli-ability goal assuming all other AFW system safety criteria are met. There-fore, the current staff position is that applicants for operating licenses must include at least three AFW pumps in their plant design, and each pump must be capable of providing to the steam generators at least the minimum flow neces-sary for decay heat removal following a loss of offsite power. Also, a minimum of two of these pumps and their associated trains must be safety grade. On October 30, 1981, the NRC staff informed the Public Service Company of New Hampshire, (PSNH) of the staff position regarding AFW system reliability, and questioned the ability of the Seabrook Nuclear Station auxiliary feedwater system to meet the specified reliability goals. The Seabrook AFW systeml consists of a two-pump safety grade emergency feedwater (EFW) system and a non-safety grade "startup feed pump" that may operate in parallel with the em-ergency feed pumps. The staff's concern, as stated in the October 30 letter, related primarily to the perceived inability to power the third " start-up" pump from the emergency AC buses. PSNH replied to the staff's concerns by letter on December 4,1981. 2 In this response it was noted that provisions were included in the Seabrook de-sign to allow the startup feed pump to be powered from an emergen y bus if necessary. With this provision it is the position of PSNH that the Seabrook EFW system design meets the requirements of the October 30 letter from the staff. 1 In this report the term " auxiliary feedwater", or AFW, system as applied to Seabrook means the combined emergency feedwater and startup pump systems. C-1

PURPOSE i The purpose of this study was to perform a reliability analysis of the Seabrook EFW system considering the use of the startup feed pump as a third source of emergency feed water and to demonstrate using the results of the ( study.the validity of PSNH's position, i.e., that the required reliability goals as specified by the NRC staff are met by the existing Seabrook AFW sys-tem design. In addition, this study was intended to identify for PSNH and the NRC-staff any dominant faults affecting the AFW system reliability under the loss of main feedwater/ loss of power transient conditions specified by the staff in NUREG-0611. The techniques used to achieve these objectives were the logic modeling methods specified by NUREG-0737. SCOPE The EFW system design evaluated by this study is that described in Sec-tion 6.8 of the Seabrook Nuclear Station Final Safety Analysis Report (FSAR) and further described in system description document SD-1M. The design of the startup feed pump system is described in system description document SD-1Q. The primary sources of specific design information about both the systems des-cribed in these documents were facility P and I drawings and logic diagrams. A listing of all drawings used in the course of this study is paovided in Table C-1. The transient conditions under which the AFW system reliability was determined are those outlined by the NRC staff in NUREG-0611, i.e., o Loss of main feedwater with reactor trip; o Loss of main feodwater with coincident loss of offsite station power; o Loss of main feedwater with coincident loss of all station AC power. 1 i ? p l 2 Letter No. SBN 198, T.F. H4.4.98 to Mr. Frank J. Miraglia from Mr. John DeVincentis. i i C-2

TABLE C.1 Engineering Drawing List for the Seabrook Nuclear Station Emergency and Startup Feedwater Systems i Drawing Title Number l Mechanical System P&I Diagrams:

1. MainSteamSystem(Sheet 1) 9763-F-202074 ,

1

2. Emergency Feedwater System 9763-F-202076 !

3. Condensate System 9763-F-202077

4. Feedwater System 97C3-F-202079

5. Compressed Air System, Key Plan 9763-F-202105

6. Compressed Air System 9763-F-202106

7. Turbine Building Compressed Air Headers 9763-F-202107

8. Miscellaneous Buildings Compressed 9763-F-202108 Air Headers Electrical System One Line Diagrams:

1. Unit Electrical Distribution 9763-F-310002

2. 4160V Switchgear Bus 1-E5 9763-F-310007

3. 480V Unit Substation Buses 9763-F-310013 1-E51 and 1-E52 4, 125VDC and 120VAC Instrument Buses 9763-F-310041

5. Turbine Building 480V Motor Control 9763-F-310046 Center 1-E523 Logic Diagrams:

1. Symbols 9763-M-503100

2. FW-Start-up Feed P-113 9763-M-503580

3. FW-Prelube P-161 For 9763-M-503581 Start-up Feed P-113 Sht 1

4. FW Emerg Fd P-37A Steam 9763-M-503584 l Supply Viv (MS-V128) Train B

5. FW Emerg Fd P-37A Stm Supply 9763-M-503585 Viv(MS-V127)TrainA

6. FW-Emerg Feed P-37B 9763-M-503586 I

C-3

TABLE C.1 (Cont'd)

7. FW-Emerg FW Bypass /Inop 9763-M-503599 Status Alann i

8. MS-Trip & Throttle Valve 9763-M-503672 I V-129

9. %-EmergencyValves 9763-M-504152 l

10. FW-Emergency Valves 9763-M-504155

11. FW-Valve-V148 '

9763-M-504156

12. FW-Prelube P-161 For 9763-M-504157 Start-up Fd P-113 Sht 2 Control loop Diagrams:

1. FW-Start-up Feed P-113 9763-M-506480

       & Prelube Pmp P-161

2. FW-Feed Pump P-328 9763-M-506481' Speed Control & Disch

3. FW-Emerg Feed Pump P-37A 9763-M-506497 (Turbine Driven)

4. FW-Emerg Feed Pump P-37B 9763-M-506498 Discharge Flow

5. FW-Emerg Feed Pump P-378 9763-M-506499 TE-4271 & TE-4347

6. MS Supply To Emerg Fd Pmp 9763-M-506555

[ Turbine Isol Viv l

7. FW-Emerg Feed Pump P-37A 9763-M-507043 Discharge Flow

8. FW-Emerg Feed Pump P-378 9763-M-507044

9. FW-Emerg FW Valve FV-4214 9763-M-507056

10. FW-Emerg FW Valve FV-4224 9763-M-507057

11. FW-Emerg FW Valve FV-4234 9763-M-507058

12. FW-Emerg FW Valve FV-4244 9763-M-507059

13. Start-up Feed Pump 1-P-113 9763-M-310844 SHCNid Prelube Pump 1-P-161

14. Prelube Pump 1-P-161 Legend 9763-M-310844 SHCN1b

       & Switch c.4

TABLEC.1(Cont'd)

15. Prelube Pump 1-P-161 9763-M-310844 SHr"Ic Cable Schematic FSAR Drawings:

1. functional Diagrams-Reactor Trip Signals Figure 7.2-1 Sheet 2

2. Functional Diagrams-Pressurizer Trip Figure 7.2-1 Sheet 6 Signals

3. Functional Diagrams-Steam Generator Trip Figure 7.2-1 Sheet 7 Signals

4. Functional Diagrams-Safeguards Actuation Figure 7.2-1 Sheet 8 Signals

5. Functional Diagrams-Auxiliary Feedwater Figure 7.2-1 Sheet 15 Pumps Startup

6. Separation of Instrument and Control Power Figura 8.3-3 Sources C-5

l APPENDIX D s ua m srA m Engineedng office: IPUBLIC SEAVICE Companyof NewHampshire 1671 Worcester Road Framinoham, Massachusetts 01701 (617) - 872 - 8100 September 7, 1982 SBN-321 T.F. H 4.4.98 8 7.1.2 United States Nuclear P.egulatory Commission Washington, D. C. 20555 Attention: Mr. Frank J. Miraglia, Chief Licensing Branch No. 3 Division of Licensing Re ferences: (a) Construction Permit CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNH Letter, dated August 27, 1982, " Reliability Analysis of the Emergency Feedwater System", J. DeVincentis to F. J. Miraglia (c) PSNH Letter, dated July 27,1982, " Response to Requests for Additional Information (RAIs) from Instrumentation and Controls Systems Branch (ICSB); A-K", J. DeVincentis to F. J. Miraglia

Subject:

Seabrook Station Emergency Feedwater System Design Changes

Dear Sir:

During the Staff review of the Seabrook Station Emergency Feedwater System (EW), a number of design changes have been recommended and are being implemented. These design changes are based on the review of the Emergency Feedwater System Reliability Analysis (Referenc'e (b)) and also bring the Seabrook design into compliance with the latest Standard Review Plan. These design changes will be incorporated into a revision to the Final Safety Analysis Report as soon as the final details are established. The fc11owing describes design changes which are presently being implemented relative to the Seabrook EW System. An attached simplifLed cketch is included for clarification of some of the design changes.

1. A continuous' minimum flow recirculation path will be provided from each EW pump's discharge to the condensate storage tank via the opposite puinp's suction line. This recirculation path will assure a continuous flow through an EW pump should flow to all four steam generators be reduced below that necessary to prevent pump damage. The original recirculation path will be retained for use during periodic pump performance testing.

D-1

        .Unitsd Statss' Nucisar Reguintory con insica                                  Sep'tesber 7, 1982 Actcution: Mr. Frank J.'Mireglia, Chief _                                    ;Page 2

2. Redundant,' safety grade flow isolation valves will be provided in each EFW branch supply-line to each stema generator. -Safety grade controls will be provided at both the main contr'ol-board and reinote shutdown locations for these valves. Further,information relative to this modification can:be found in Reference (c).

3. Manual isolation valves will be provided upstream of each pair of ficw isolation valves to each steam generator. These manual isolation valves will permit isolation of any EFW f'.ow isolation valve while rettising the availability of both EFW pumps and the Startup Feedwater pump.

4. Safety grade, Seismic . Category I air accumulators will be provided as a b

back-up air supply for the actuatora-of both main. steam supply valves (MS-V127; and MS-V128,) to the turbine-driven EFW puaj,, R-37A. These accumulators will be sized to proy,ide at 1gast twq complete valve operetions plus maintain the valvea closed for a minhaua of Tour hours. This safety grade air supply wi11' upgrade the reliability of these valves-i consistent with the Class 1E~ controls presently utilized in the design.

5. The Startup Feedwater-(SUF) pu,ap discharge valve to the EFW header, FW-V156, will be relocated out of the EFW pump- Room. This will assure ~ ,.

the ability to cross-tie the SUF pump to the EN System should ' a " series of potential failures render both EFW pumps inoperable and the EFW Pump _ Room inaccessible.- , Additionally, during both our in-house and your Staff review of

       . Reference (b), three areas were found which should be clarifigd or corttected.                   -

First, on Page 12 of Reference (b) an asterisk notes that only one of the steam admission valves (MS-V127) to the' turbine-driven EFW pump can be controlled from the remote shutdown panels. In conjunction with modification #4 listed above, Class <1E controls for the other steam admission valve (MS-V128) will also, be provided at the remote shutdown location. These modifications will unsure the ability to , start and/or stop'the turbine-driven EFW pump f rom either, the: main control board or the remote shutdown panels. ' Second, on Page 15 of Reference (o), ' relative to the manual valve realignments required to provide SUF pum'p . flow to the EFV header, it states that the SUF pump recirculation isolation valve (FW-V109) must be ' closed to prevent a diversion of pu:ap flow to the Condensate Storage Tank (CST) 'should the recirculation flowtontrol valv'c (PCV-4326) fail open. What was not, considered, however, . is' that the capacity of the SUF pump.is significantly greater than that of an EFW pump. At a TDH equivalent to , the design rating of the EFW pump; the SUF pump has a flow capacity greater than an EFW pump, even when maximum flow is diverted back to the CST through the recirculation valve. Therefore, it is unnecessary to close valve FW-V109 to ensure sufficient flow from the SUF purnp to the steam generators. This is one less manual action necessary fgr this operation. Third, on Page 29 of Reference (b), .a note on the bottom of the page indicates that a loss of off-aite power will result'in' closure of the 0-2 l

   -~-      _ _-                                                        -

e , St: United States Ndelear Regulatory Cossission Saptesber 7,1982 Mr. Frank J. Miraglia, Chief Page 3

          ,Attent io n s.

ma'In fe'edwater isolation valves. This. note is incorrect - the main feedwater. isolation valves will not close due tu a loss of of f-site C power. Additionally, it should. be _ noted that the loss of of f-site power does not result in a loss of control of the main feedwater regulating valves not the main feedwater regulating bypass valves. The result is,

                . th'e SUF pump can be utilized to supply feedwater to the steam generators
                 'during a loss of off-site power event without the need of manual . valve
   .              alignments to provide flow through the EFW System. Flow from the SUF
                 . pump to the steam generators can be accomplished utilizing the normal a'             main Jeedwater System.
                'It is hoped that the above information will assist your Staf f in their ovaluation of the Seabrook Station Emergency Feedwater System and preparation
         - of the Safety Evaluation R port.e If further information is necessary, please feel free to contact us.

, Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY J. DeVincentis Project Manager PA/kac cc: Mr. Robert Jaross, Argonne National Laboratories

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                            } YANKEE ATOMIC ELECTRIC                              PLA   B/23/82 TITLE. b lMPl.lflED OKCTCH COMPANY                                                       OF THE CHECKED g 8Y NUCLEAR SERVICES DIVISION                                   </                         O E'ASROOK           bTATIOh!

owG. NO. AU'Nj3 N {fW $ysygg

SUPPLEMENT Seabrook AFWS Revised Reliability Assessment for Loss of Off-Site Power (LOOP) ~.n

BROOKHAVEN NATIONAL LABORATORY MEMORANDUM DATER December 2, 1982 To: I. A. Papazogiou FROM: R. Youngblood/A. Fresco Aeder LUBJECT: Seabrook AFWS Revised Reliabi ity Assessment for LOOP In a conference call on November 17, 1982, R. Anand of the NRC-ASB and J. Tsao of the NRC-RRAB informed us of some design changes proposed by the ap-plicant which are intended to improve the AFWS reliability under Loss of Off-site Power conditions. We were then requested to re-evaluate the unreliabil-ity for the new conditions. This memo is divided into three parts. Part A is a description of the design changes, Part B is a justification for the human error unreliability data assumed to apply to the new conditions, and Part C is the actual quantitative results for AFWS unreliability for the proposed design changes of Part A utilizing the data of Part B. Part C also contains the segments of the original draft BNL assessment wnich are affected by the design revisions. A. Seabrook AFWS Proposed Design Changes as of November 17, 1982

1. Locked-open manual valves V-163 and V-156, which isolate the SUFP from the EFW header, will be converted to motor-operated valves cperable from the Control Room.

2. Two separate breakers will be installed to eliminate the need to man-l ually transfer the SUFP breaker from Bus 4 in the Non-Essential l

Switchgear Room to Bus E5 in the Essential Switchgear Room. The oper-

ator will still need to change the bus transfer switch to the ES position.

Subsequent discussion with the applicarit on November 22, 1982 clari fied certain issues concerning the use of the MFW flow paths to supply water to the steam generators from the SUFPS:

1. The MFW regulating valves FCV-510, FCV-520, FCV-530 and FCV-540 are air-operated, fail-closed valves. Upon LOOP, the control signals will tend to close the valves upon reduction of the steam flow rate from the steam generators.

S-1

2. The valves require air from any one of the Instrument / Service Air Com-pressors to allow the operator to open them from the Control Room.

There are (3) 100% capacity Instrument / Service Air Compressors. At any given time, one of the compressors is connected to Emergency Diesel Generator Train A and another to Train B. One is operating us-ing non-emergency on-site power and the other is on standby. The third compressor is not powered except when it is used to replace one of the other two compressors.

3. Upon LOOP, the compressors are automatically loaded onto the Diesel-Generators at the end of the 2-minute sequence of loads.

The following was also discussed:

4. The only intended usage of the EFW header locked-open manual valves, V-125, V-126, and V-127 is for isolation in the event of a pipe rup-ture. There are no test or maintenance acts which require closure of those valves during Normal, Hot Standby, or Hot Shutdown operation.

(Although it is not within the scope of the NUREG-0611 analysis, the feasibility of using these valves to isolate a rupture is very limited because they are locked-open and manually-operated. For a full guil-lotine rupture with both EFW pumps operating, flow would emanate from both directions, severely restricting local operator actions. In the applicant's own report, which did consider pipe ruptures, a probabil-ity of 0.9 was assumed for the operator failing to isolate a rup-ture).

5. The addition of the Seismic Category I, safety-class air accumulator on the air supply to the Main Steam Admission Valves, V-127 and V-128, for the turbine-driven EFW Pump P-37A has no effect on the automatic initiation of the pump.

B. Quantification of Human Error Unreliability The task of the present section is to quantify the human error con-tribution to the failure probability of the pump, given LOOP. Recent work indicates that human errors of omission are dominated by " cognitive" failures; this idea is discussed below and applied to the present task. Errors of commission are more difficult to treat, and are not included here. It is felt I that neglect of errors of commission is characteristic of the AFWS reviews which have been performed up to now. The scenario of interest is the following. There has been a LOOP (loss of offsite power), so that the normal source of AC power to the startup feedpump (SUFP) is not available. Ordinarily, the two pumps of the EFWS (Emergency Feedwater System) should be available; one is steam turbine driven, and the other is electric motor driven and is aligned to diesel-backed AC. However, in the scenario considered here, both EFWS pumps are unavailable, for whatever reason. If the operators manually align the SUFP to a diesel-backed l S-2

bus (and if there are no more failures), it can provide feedwater. The human failure of interest is failure to align the SUFP to a diesel-backed bus. A useful approach to this problem is afforded by the Operator Action Tree System. Full descriptions of this are given in Refs.1 and 2. The general idea is summarized as follows. Given a situation which calls for operator ac-tion within a specified time, one divides the available time into three peri-ods:

1) the time it takes for cues (meters, alarms, etc.) to become available to the operators, calling for action;

2) the time available for cognitive processing of the cues (the time available for planning and decision making);

3) the time required to implement the decisions reached in Phase 2.

The general idea of the 0AT method is that cognitive errors in Phase 2 are the dominant contribution to failure to perform a necessary act, and that the single most important parameter determining the failure probability in Phase 2 is the available time. For purposes of the present analysis, a 30-minute period has been adopted as the time after LOOP within which feedwater must be provided. It should be evident within a minute or so whether the EFWS is operating, so Phase 1 can essentially be neglected. Phase 3 (manual alignment of the SUFP to a diesel-backed bus) requires manipulation of breakers outside the. control room. It is this procedure which has recently been streamlined. It is believed that 15 minutes is ample time for this procedure. The arguments which place all the error in Phase 2 seem to require that an ample time frame be assigned to Phase 3; otherwise, procedural errors will begin to contribute substantially to Phase 3, while at present they are implicitly being assumed to be easily ) reccverable. Based on a tour of this facility, our impression is that someone l who knows what to do can easily accomplish the task in 15 minutes. This [ leaves 15 minutes for Phase 2, the decision interval. The OAT method goes on to suggest a cognitive probability corresponding to the 15-minute intervals. At this point, however, since the control-room

   " thinking" phase has been separated from the out-of-control-room " acting" phase, we can consult NUREG-0611 for an applicable error probability. In Table III-2, we find that failure of an operator and his backup to actuate the AFWS within 15 minutes is assigned 5x10-3         From the arguments given above, we would apply this directly to the case at hand: for Phase 2, then, the error probability is 5x10-3, Above, it was argued that " cognitive" errors dominate the overall failure probability. On the other hand, it is widely believed the 10-2 is the basic error rate for procedural acts without prospect of error recovery. Therefore, S-3

r Phase 3 will contribute significantly unless there are real opportunities for error recovery. From the previous discussion, it is clear that "real opportun'ty for. recovery" requires ample time to perfom the operation and personnei who are sufficiently trained in the operation to be able to diagnose their own errors. Since.it will be quickly apparent whether the operation is successful, therg appears to be' abundant opportunity for recovery. On this basis, taen,10-J is assigned as the probability of the failure of trained personnel to complete this action within 15 minutes of being told to perform it. The total failure probability is therefore 6x10-3; 5x10-3 for "cogni-tive"-errors in Phase 2, and 1x10-3 for procedural errors in Phase 3. Er-

    .rors of commission have been neglected, ar.d the low error probability assigned
    'to Phase 3 is based on there being ample time for error recovery, etc. Note that hardware failure of the breaker involved has previously been included in the analysis.

References

1. J. dreathall, " Operator Action Trees: An Approach to. Quantifying Operator Error Probability During Accident Sequences", NUS Report #4655, NUS Corp.,

July 1982.

2. R. E. Hall, J. Wreathall, and J. Fragola, " Post Event Decision Errors Operator Action Tree / Time Reliability Correlation", NUREG/CR-1605 (BNL-h0 REG-51601), Brookhaven National Laboratory, November 1982.

C. dalculation'of 'the Seabrook AFWS Unreliability for the Revised Design The objective here is to calculate the AFWS unreliability for LOOP utilizing the human _ error data resulting from the discussion in the previous section.- The following portions of the BNI. assessment are affected: a) Figure 11-Sheet 3: BNLCutsets-LOOP (Start-upFeedPumpP-113),p.86. b).. Table 11: BNL Results-Unavailability of Seabrook AFWS Proposed Design Using NUREG-0611 Data-LOOP Transient, p.63. c) Figure 1:_ Comparison of Reliability of Seabrook AFWS to Other AFWS Designs in Plants Using the Westinghouse NSSS, p.71. Referring to Figure 11, Sheet-3, one of the dominant cutsets for gate SUP1, "No Flow to Supply Header From SUFP P-113", is MPB1P1610E, " StartSUFPPre-LubePumpP-161",whichwasassessedat3x10gperatorFailsto . This value is about 25 sion,6x10goftheoverallunavailabilityofSUPl. will now become the new value for MPB1P1610E. From the previous discus-S-4

i r l l In T 3 is thereby reduced from 9.2x10-2 to 6.8x10-2.able 11, the value for HSubstituting the latter value into the equations (a), (b), a l l (c),weobtain: AFW* = 8.6x10-5 LOOP In the assessment of Table 11, it was assumed for reasons explained in the BNL report that failures of the HFW flowpaths are a negligible contribution to the unreliability of the AFWS. However, in the ensuing discussions among BNL, NRC, and the applicant, it has been determined that the MFW regulating valves FCV-510, 520, 530 and 540 fail closed upon loss of air supply and that one of the Instrument / Service Air Compressors is required to open them as mentioned earlier. Even so, loss of the MFW flowpaths still requires multiple failures (e.g., failure to open 3 of 4 valves, or multiple compressor failures), which are substantially less likely than single failures of the SUFP train. In addition, the capability would remain to open valves V-163 and V-156, which will now be motor-operated from the Control Room, to allow the SUFP to supply Thus, the result obtained of 8.6x10-{eedwaterthroughtheEFWheader. will not be significantly affected by the unavailability of the MFW flowpaths. Revised copies of Figure 11, Sheet 3, Table 11, and Figure 1 are attached. i l L S-5

       ~SEABR'00K t00P:"EFWSTTRKINS + STA TtJP                D   FEEDPUMP-~~ NUREG-06T1- SCopt CUT 7ETS-Fng-0-ATMyP1                            gT-T H -p tr o g A BI L-I TY T'G ET--170 0 EP30 5

1. 6.00E-03 MPBIP1610E

g. 3700fM02 RGD10Gl'AME^

3. 1.00E-02 RCA1A93-00 4 1.00Er02 MV01'V1560 A

5. 1.00E-02 Mvnly1630A

n. o.4'0E Da gGDT1WJ:00

7. 5.80E-03 MPB1P11300

8. St00E'-03 rtPIT1P113ME

9. 5.00E-03 MPR1P161ME rVM1vl5200

10. 5.00E303--

11. 2.00E-03 MCE1P113MN

12. 1700E03 FCElA'93 MB

13. 1.00E-03 .RCA1A54-MB 14 G00r:03 rV01V14100

15. 1. DOE-03 FV01V1430D 1W. 1 01E704~ HPR1P1130G

17. 1.00E-04 RCA150AEHC l'a.

iT00E 04 RCA1AF4~MC

19. 1.00E-04 RCA1A63-MC 407 1 00E'04 rVM1V152MD

21. 1.00E-04 FVD1V141MD

72. 1700E 04 FVDI VT43NO~

23. 1.00E-04 MVA1V99-HA IST MOMENT = 9.3E-02 Revised Figure 11 (Continued) BNL Cutsets - LOOP (Sneet 3: no Flow From Start-up Feed Pump P-113).

S-6

TABLE 11 (REVISED) BNL RESULTS UNAVAILABILITY OF SEABROOK AFWS PROPOSED (SUPPLEMENT) DESIGN USING NUREG-0611 DATA LOOP TRANSIENT

1. Refer to Table 8 and 10. Again the expression for AFW* is:

AFW* = (AF91)-(AF127)-(SUP1) + (AF127)-(SUP1)-(V125)

                                  + (AF91)* (SUP1)-(V127)

2. As in the Proposed Design for the LMFW transient, it is no longer necessary for the operator to open V156 or V163 so that the failure ratos for those events can again be subtracted from H3 of SUPl. The values of AF91, AF127.and SUP1 are now:

AF91 AF127 SUP1 Hy = 1.1x10-2 H 2 = 4.3x10-2 H3 = 6.8x10 -2 M1 = 1.1x10

              -2                 M 2 = 1.2x10-2              M3 = 1.2x10 -2

' 3. Separating AFW* into Hardware Failures and Maintenance (or Test) Failures: (a) (AF91)-(AF127)-(SUP1) = H H H123+HMHI23+HHMy 2 3I2 + N H "3 (b) (AF127)-(SUP1)-(V125) = (H H23+MH2 3 + H 23N ) (y125) (c) (AF91)-(SUP1)-(V127) = (H H13+HMI 3 + My3 H ) (V127)

4. Substituting the new value for H 3.

(a)a 32.16 x 10-6 + 8.98 x 10-6 + 5.68 x 10-6 + 32.16 x 10-6

           =   78.98 x 10-6 S-7

TABLE 11 (cont'd) (b) = (29.24 x 10-4 + 9.16 x 10-4 + 5.16 x 10-4)(11 x 10-4)

          = (42.56 x 10-4)(11 x 10-4) = 4.68 x 10-6
    -(c) = ( 7.48 x 10-4 + 1.32 x 10-4 + 7.48 x 10-4)(11 x 10-4)
          = (16.28 x 10-4)(11 x 10-4) = 1.79 x 10-6 AFW* = (a) + (b) + (c) = 1.15 x 10-4 AFW* = 8.55 x 10-5 LOOP S-8

Tremant twents LMFW LA*FW/ LOOP LMFW/Lov of Att AC* Plants Low Med High Low Med High Low Med High Seabrook .g_ .a , g, Haddam Neck 9 iD qp San Onofre S e t> - ---g> Prairie Island 4> d> qp Salem (H4 (HD G Zion 9 3 tp Yankee Rowe e e ip Troian 4> G qp indian Point 4> $ t> Kewanee $ D t>

H. B. Robinson 4p e

Cook 0 e ap Twkey Pt. 9 9 O Farfey e e ap $ wry G G No. Anna e e e h order of Magnituda in Unavailability Represented.

Note: The scale for this event is not the sarne as that for the LMFW and LMFW/ Loop.

BNL Assessment - NUREG-0611 Scope e Reference 3 Design AProposed Design $ Supplement Design (Nov. 17,1982) Applicant's Results mReference 3 Design Figure 1: Comparison of Reliability of Seabrook AFWS to Other AFWS Designs in Plants Using the Westinghouse NSSS. . e. . com,.mn rmn,,c orna , i,u . a n _2, , , u 2. 3_g '0"" U.S. NUCLEA7 REZULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET UR G 3b[ BNL-NUREG-51733

4. TITLE AND SUBTITLE LAdd Volume Na,if apprepriate) 2. (Leave bimk)

Review o he Seabrook Units 1 and 2 Auxiliary Feedwater 3. RECIPIENT'S dESSION NO. System Reli ility Analysis f

7. AUTHOR (S) 5. DATE RE[RT COMPLETED Fresco A., Youngb od R., Papazoglou I.A. " d[ber 11 %

9. PERFORMING ORGANIZATs NAME AND MAILING ADDRESS (lactude I/p Code) DA[ REPORT ISSUED Brookhaven National L ratory h ruary IIM Upton, NY 11973 g,,,,y,,,,,

f 8. (Leave Nank) D MAILING ADDRESS (Include I,p Code)

12. SPONSORING ORGANIZ ATION NAME p Division of Safety Technology Office of Nuclear Reactor Regu tion 11. FIN No.

U.S. Nuclear Regulatory Coninissi A-3933 Washington, DC 20555

13. TYPE OF REPORT

[E RIOD COVE RED (inclusive dates) Technical Report

15. SUPPLEMENTARY NOTES 14. (Leave .2/mk/

16. ABSTR ACT #00 words or less)

This report presents the results of a r te f the Emergency Feedwater System Reliability Analysis for Seabrook Nuclear Station its nd 2. The objective of this report is to estimate the probability that the Eme ency Fe water System will fail to perform its mission for each of three different initiato : (1)lo of main feedwater with offsite power available, (2) loss of offsite powe , (3) loss o all AC power except vital instrumentation and control 125 VOC/120 VAC power. The scope, me lodology, and failure data are prescribed by NUREG-0611, Appendix III. The esults are com ed with those obtained in NUREG-0611 for other Westinghouse plants.

17. KEY WORDS AND DOCUMENT A LYSIS 17a. DESCRIPTOR

Reliability Analysis Auxiliary Feedwater Sy tent Seabrook Nuclear Powe Plant Units 1 and 2 Pump and Valve Failu Rates 17b. IDENTIFIE RS/OPEN-EN ED TERMS J

19. SE CUR TY SS (Th<s reporr/ 21 NO OF PAGES

18. AVAILABILITY STATEf1ENT Unlimited 20. SeCuRi ry CLASS (TNs pap / 22 PRICE tinelatsifind s NEC FORM 335 mau

UNITED STATES souunctess m t NUCLEA] CET ULATT."Y CIMMISSION FOS' AGE & f t($ P".10 WASHINGTON, D.C. 20555 $$Up c Ptnuiru g OFFICIAL BUSINESS PENALTY FOR PRIVATE USE,8300 120555078877 1 1AN US NRC ADM-DI V 0F TIUC POLICY f. PUB MGT BR-PDR NUREG h-501 hASHINGTON CC 20S55}}

ML20080M435

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