Text

Draft Review of Seabrook Units 1 & 2 Emergency Feedwater Sys Reliability Analysis
	ML20082B202
Person / Time
Site:	Seabrook
Issue date:	10/31/1982
From:	Fresco A, Papazoglou I, Youngblood R; BROOKHAVEN NATIONAL LABORATORY
To:	; NRC
Shared Package
ML20079F081	List: ML20079F079; ML20079F278; ML20079G616; ML20079G620; ML20079G748; ML20079G889; ML20079K188; ML20079Q381; ML20082B113; ML20082B121; ML20082B125; ML20082B133; ML20082B142; ML20082B150; ML20082B151; ML20082B202;
References
	FOIA-83-388 NUDOCS 8311210049
	Download: ML20082B202 (190)
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. d, NUREG/CR-ji[2 BNL-l.UR EG-REVIEW OF THE SEABROOK UNITS 1 & 2 EMERGENCY FEE 0 WATER SYSTEM RELIABILITY ANALYSIS

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A. Fresco, R. Youngblood and I. A. Papazoglou Department of Nuclear Energy Brookhaven National Labor 6 tory Upton, New York 11973 October-1982 1

fy12 49 830002 SHOTWEL83-388 PDR 56-W.>4 c50-5H Prepared for U. S. Nuclear Regulatory Commission Washington,_0. C.

20555' Under Interagency Agreement DE-AC02-76CH00016

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ABSTRACT This report presents the results of a review of the Emergency Feedwater Systen 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 miss' ion for each of three different initiators:

(1) loss of main feedwater with offsite power available, (2) loss of offsite power, (3) l'oss of all AC power except vital instrumentation and control 125 VOC/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, i

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TABLE OF CONTENTS Page ABSTRACT...............................................................

iii L I S T OF F I G UR E S........................................................

vii LIST OF TABLES.........................................................

viii

SUMMARY

AND CONCLUSIONS................................................

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1.0 INTRODUCTION

1 2.0 SCOPE OF BNL REVIEW...............................................

2 3.0 MISSION SUCCESS CRITERIA..........................................

3 4.0 SYSTEM DESCRIPTION................................................

3 4.1 Confi gurati on a nd Ove ral l Des i gn.............................

3 4.1.1 St a rt u p Fe ed P ump Sy s t em........................ ;.....

3 4.1.2 Emergency Feedwater ' stem............................

5 4.2 Component De si g n Cl a s si fi c ci o n..............................

8 4.3 Power Sources................................................

8 4.4 Instrumentation and Controls.................................

9 5.0 EMERGENCY OPERATION...............................................

12 5.1 Lo s s of Ma i n Fe edwa t e r.......................................

12 5.2 Loss of Offsite Power........................................

12 5.3 L o s s o f Al l AC P owe r.........................................

14 6.0 TESTING...........................................................

15 7.0 T ECH N IC AL SP E CI F I C AT I O N S..........................................

16 8.0 ASSUMPTIONS.......................................................

19 8.1 Ge n e ral Fa i l u r e Da t a.........................................

19 8.2 Treatment of Time Dependent f441 ures.........................

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8. 3 Test and Maintenance Outages.................................

23 8.4 Operator Errors..............................................

30 9.0 R EL I AB I L I TY AN AL YS I S..............................................

35 9.1 Oualitative Aspects..........................................

35 9.1.1 Mode of System Initiation............................

35 9.1.2 Syst em Control Fol l owi ng Ini ti ati on...................

35 9.1.3 Ef fects of Test and Maintenance Activities............

37 9.1.4 Avail abili ty o f Alternate Water Supplies..............

38 9.1.5 Adequacy and Sepa ration of Powe r Sources..............

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TABLE OF CONTENTS (Cont'd)

Pace 9.1.6 C ommo n Mod e Fa i l u re s..................................

39 9.1.7 Si ng l e Po i n t F a il u r e s.................................

41 9.1.8 Adequa cy of Emergency Procedure s......................

41 9.2 Gu a nt i t a t i v e Aspe ct s.........................................

41 9.2.1 Applicant's Use of NRC-Suggested Methodology and Data..

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9.2.1.1 Fault Tree Construction and Evaluation.......

41 9.2.1.2 Failure Data.................................

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9. 2. 2. Ap pl i c a nt ' s R e s ul t s...................................

45 9.2.2.1 System Unavailabilities......................

45 9.2.2.2 Dominant Failure Modes and Conclusions.......

46 9.2.3 BNL Assessment........................................

51 9.2.3.1 FaultTrees..................................

51 9.2.3.2 Failure Data.................................

54 9.2.3.3 Sy s t em U n av a i l a b i l i t i e s.....................

55 9.2.3.4 Domi n a n t Fa i l ur e Mode s............,..........

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9. 2. 3. 5 ' Ge n e ral Compa ri so n...........................

67 9.2.3.6 Possibl e Reli ability Improvement s............

68 REFERENCES.............................................................

70 APPENDIX A SEABROOK AUXILI ARY FEEDWATER SYSTEM FAULT TREES............

A-1 APPENDIX B F AU L T T R E E D ATA............................................

B-1 APPENDIX C BACKGROUND INFORMATION PROVIDED BY THE APPLICANT...........

C-1 APPENDIX D Letter to Mr. F. J. MiragTra, 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.................

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i LIST OF FIGURES Figure Title Page 1

Comparison of Reliability of Seabrook AFWS to Other AFWS Designs in Plants Using the Westinghouse NSSS 71 2

Seabrook Nuclear Station 72 3

Seabrook Nuclear Station Emergency Feedwater System 73 4

Steam Supply for Turbine-Driven EFW Pump 74 5

Startup Feed Pump Normal Alignment to the Main Feedwater System (Simplified) 75 6

Motor-Driven EFW Pump Control Logic 76 7

Applicant's WAM Results for Loss of Main Feedwater 77 8

Applicant's WAM Results for Loss of Offsite Power 78 9

Applicant's WAM Results for Total Loss of AC Power 80 10 BNL Cutsets'- LMFW 81 11 BNL Cutsets - LOOP 84 12 BNL Cutsets - Top Event No Flow to 3 Out of 4 Steam Generators (Gate AFW) 87 O

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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 29

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3 Applicant's Summary of Operator Act' ions / Failure Probabilities 34

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4 Applicant's AFW System Unreli. ability 46 5

Applicant's Results - Dominant Contributors to Conditional Unavailability Loss of Main Feedwater Event 47 6

Applicant's Results - Dominant Contributors to Conditional Unavailability - Loss of Main Feedwater/ Loss of Offsite Power Event 48 7

Applicant's Results - Dominant Contributors to Conditional Unavailability - Loss of Main Feedwater/ Loss of All AC Power 49 8

BNL Results - Unavailability of Seabrook AFWS - Ref. 3 Design Using NUREG-0611 Data LMFW Transient 56 9

BNL Results Unavailability.-of Seabrook AFWS - Proposed Design Using NUREG-0611 Data LMFW Transient 58 10 BNL Results - Unavailability of Seabrook AFWS - Ref. 3 Design Using NUREG-0611 Data LOOP Transient 60 11 BNL Results - Unavailability of Seabrook AFWS - Proposed Design Using NUREG-0611 Data LOOP Transient 63 12 Summa ry of BNL Assessments --

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LIST OF TABLES (Cont'd)

Table Title Page A.1 System Identification Code A-2 A.2 Component Types A-3 t

A.3 Fault Codes A-5 B.1 Mechanical and Electrical Component Failure Rates B-1 B.2 Fault Identifiers for the Seibrook Emergency Feed Station B-8 B.3 NRC-Supplied C&ta Used for Purposes of Conducting A Com-parative Assessment of Existing AFWS Designs and Their Potential Reliabilities B-34 C.1 Engineering Drawing List for the Seabrook Nuclear Station Emergency and Startup-Feedwater Systems C-4 e

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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 designed by Westinghouse. The results of that study were presented in NUREG-0611.(1) At the request of the NRC,(2) the Yankee Atomic Electric Company and the Public Service Company of New Hampshir', operating license ap-

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e plicants, have provided the NRC with a study of the Seabrook Units 1 and 2 AFWS,(3) pe,' formed using NUREG-0611 as a guideline.

BNL has reviewed this study. The BNL conclusions are as follows ("High", " Medium" and " Low" refer to i

the NUREG-0611 reliability scale).

i 1.

For an accident resulting in a loss of main feedw:ter (LMFW) with offsite power available:- The reliability of AFWS is in the Hich range.

(Unavailability = 1.95 x 10-5/ demand.)

2.

For a loss of offsite power (LOOP) resulting in a concurrent loss of main feedwater (LMFW):- The ' reliability of the AFWS is in the Medium range, provided that operator action is taken to connect the startup feed pump and th'e instrument air compressor to emergency diesel generator 1A.

(Unavailability = 1.15 x 10-4/ demand.)

3.

For a loss of all AC power (LOAC), except for the 125 VDC/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 range. (Unavailability = 2.3 x 10-2/ demand).

Results are summarized in Table 1.

A compartson of the Seabrook AFWS re-liability to other AFWS designs in plants using the Westinghouse NSSS is shown in Figure 1.

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1 Table 1 Unavailabilities of Seabrook AFWS Comparison of Applicant's Results to BNL Assessment i

APPL ICANT ' S =

RESULTS BNL ASSESSMENT **

Transient Ref. 3 Desian Ref. 3 Desian Proposed Design l

1.

LMFW 2.1x10-5 4.5x10-5 1.95x10-5 2.

LOOP 5.2x10-5

'1.8x10-4 1.15x10-4 3.

LOAC 2.1x10-2 2,3x10-2

'2. 3 x 10.

Using Applicant's Data

- Using NUREG-0611 Data i

Note:

See page 2 for a discussion of the proposed design.

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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 performed of the Auxiliary Feedwater Systems (AFWS) of all then-operating plants.

The results obtained for operating Westinghouse-designed plants were presented in NUREG-0611.(1) 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 main feedwater (LMFW) without loss of offsite power; (b) loss of main feedwater associated with loss of offsite power (LOOP);

(c) 1.oss of main 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 study.

A quantitative cri-terion for AFWS reliability has been defined by the NRC in the current Standard Review Plan (SRP) for Auxiliary Feedwatl!7 Systems:(4)'

... An acceptable AFWS should have an unreliability in the range of 10-4 to 10-5 per demand based on an analysis using methods and data presented in NUREG-0611 and NUREG-0635.

Compensating factors such as other methods of accomplishing the safety functions of the AFWS or other reliable 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 the Emergency Feedwater System.

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2.0 SCOPE OF SNL REVIEW The BNL review has been conducted in accordance with the methodology, data, and scope of NUREG-0611, Appendix III.(1)

It has two major objec-tives:

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

(b) To provide an intependent 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 review 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 A'FWS 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 cemand, using 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 f ailures 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 Ref. (3).

Such changes affect the use of the startup feed pump during loss of offsite power conditions, and also the capability to perform maintenance on valves in the emergency 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 "Prdposed Design" 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 from the applicant, which appears in Appendix D.

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3.0 MISSION SUCCESS CRITERIA As described in Ref. (3), for each of the transient conditions analyzed, unreliability was defined as the probability of f ailure of the combined EFW and startup pump system to start and provide feedwater to et least two of the four steam generators prior to the time that the steam generators would boil dry following a reactor trip from full power. The time required to boil away the

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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.

BNL Corenent: 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 af ter 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 re-moval system may begin operation.

4.0 SYSTEM DESCRIPTION 4.1 Configuration and Ov'erall Design The Seabrook Auxiliary Feedwater System, as the term is commonly applied to systems which are used for Startup, Hot Standby and Hot Shutdown, consists of the single pump of the Startup Fe.ed,p ep System (SUFPS) and its associated g

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 3

Condensate Storage Tank (CST) via the main condensate makeup line.

The suction line between the pump and CST is equipped with three normally open manucl iso-lation valves, V-152, V-143 and V-141.

The discharge headers from the pump at-tach to six other feedwater system headers, i.e., the main feedwater pump dis-charge header, the high pressure feedwater heater outlet header, the con-densate pump discharge header, the make-up header from the CST, the steam gen-erator recirculation pump discharge header, and the EFW pump discharge header.

The pump is also equipped with a recirculation line to the CST for pump protec-tion and testing. Flow through the recirculation line -is controlled by a pressure-controlled throttling valve, FCV-4326, that senses " pressure at the pump discharge.

l With the exception of the main feedwater pump discharge peader, the dis-l charge from the startup pump is isolated from all feed systen headers by at least one normally closed valve. The supply path to the'~ main feeuwater pump discharge header is normally olen but is equipped with a manual gear-operated valve, V-in0, 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.

BliL Comment:

In a' conference call on July 26, 1982, the applicant stated that 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 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 lube oil pump located on the SUF pump supplies lubr,1 cation and the auxiliary lube oil pump is stopped.

Should the, shaft-driven pump fail, the auxiliary oil pump will automatically restart.

In its normal operating mcJe the SUF pump will start automatically on a trip of both main feed pumps,(LMFW) unicss a safety injection or hig6-high steam generator level signal also occurs.

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BNL Coment:

At the July 15, 1982 plant visit, the applicant stated that the basic design philosophy of the SUFPS is that 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 is a standby system which would not be operated during normal plet 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-nee.ed Safety Feature (ESF) actuation signal, i.e., a loss of offsite power (LOOP), 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 Figur.e 3.

The system con-sists of two pumps, each supplied by individual 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 full cooling of' the Reactor Co'olant 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 stcam-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 "toTcommon return to the CST which is used for pump testing. This return line is isolated during normal operation.

During operation of the EFW system, both pua. - discharge into a common header, which in turn supolies 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.

i 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 t

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positions are set such that they assure a minimum of 235 spm 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 flow in the line exceeds a pre-set high flow value.

This feature prevents di-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-ferential pressure information to the control e,quipment for flow measurement.

Two separate flow transmitters are used to provide independent high flow isola-tion signals to each of the isolation valves.

The flow transmitters, control equipment, and motor-operators for the valves upstream of the flow elements are powered 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 limited 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 steam generator.

The venturis provide an addea benefit in that a pipe break in any steam generator, along with failure of both isolation valves in the associated EFW line, will not caus.e a complete loss of cooling water to the remaining steam generators.

Each supply line is also 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 feedwater header are equipped with a total of five isolation-velves 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.

The 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.

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BNL 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 manufa'cturer'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 +.he 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-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 normally 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 supplied to a common header through air-operated, fail-open valves, V-127'a'n3 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 whict vill 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 mechanical trip valve that closes on turbine overspeed, V-129.

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Oil cooling for the turbine-driven EFW 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 IB, while the SUF pump, via operator action de-scribed in Section 5.2, can be powered from emergency AC bus E-5 and diesel generator 1A.

The auxiliary lube oil pump 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 llB.

Control power for the steam-turbine admission valves is supplied from both F.SF trains, one valve receiving control power-from DC bus 11A in train A, and the other receiving power from DC bus 11B in train B.

There are no AC power de-pendencies in the turbine-driven EFW pump train.

Electrical power for the EFW

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isolation valves also comes from the emergency buses, and train separation cri-

'teria are met for each EFW supply line.

BNL Comment: The SUFPS is normally supplied power from the non-safety i

station electrical grid. During a LOOP condition, such power is not available and transfer to the emergency sources is required to operate the SUFP.

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4.4 Instrumentation and Controls As described in Ref. (3), the control room operator at Seabrook has avail-

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able a variety of instrumentation and controls that allow him to monitor and direct operation of both the emergency feedwater system and the startup feed pump.

The important equipment relative to EFW and SUF system operation is listed below:

3 Instrumentation Location o Operating status lights for the Control room / remote motor-driven EFW pump, P-378 safe shutdown panel o Position indication lights for Control room / remote both steam admission valves, V-127 safe shutdown 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-range'and one wide-Control room / remote a

range level transmitter in each safe shutdown panel steam generator o Steam pressure in each steam Control room generator o Dual CST level transmitters Control room Alarms Location o Trip alarm for motor-driven EFW.----

Control room pump, P-378 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

Position indication is available at the remote shutdown panel only for.

valve V-127.

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

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Alarms Lo'ca ti on 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 temperature alarms BNL Comment:

It is not stated whethen this applies 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 Controls Location o Manual / auto controller for Control room /

motor-driven EFW pump switchgear room o Manual / auto controller' for

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Control room / remote turbine-driven EFW pump safe shutdown panel

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

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o Manual / auto controller for Control room startup feed pump prelube pump

0nly steam-admission valve V127 can be controlled at the remote. sh.itdown panel.

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BNL Comment:

In Appendix D, it is stated that V128 will also be controlable from the remote panel.

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Automatic Actuation Signals FFW Function o

Safety injection signal Starts both EFW pumps l

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.

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5.0 EMERGENCY OPERATION 5.1 Loss of Main Feedwater In the case of LMFW, the SUFP prelube pump receives an automatic signal to start, followed by the SUFP itself upon tripping of both MFW pumps. The SUFP is normally aligned to the main feedwater nozzles upstream of the main feed-water isolation valves (I1FWIVs).

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 Figure 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 fail 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 noimally used for this transient.

5.2 Loss of Offsite Power - Ref. (3) Design In the case 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"W flowpath connections are upstream of ~

~

the MFWIVs.

The latter are air-operated piston valves which close upon loss of air supply.

The footnote on 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 to the EFW header ' requires several operator ac-tions which must be taken 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 12 l

l v

the non-essential switchgear room, move it co the essential switchgear room, and manually " rack in" the breaker to emergency bus E5.

He must also change the bus transfer switch to the 25 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 closed to prev'ent possible flow diversion of the SUF pump discharge to the condensate tank via the SUFP recirculation line should power be lost to the SUFP recirculation valve (PCV-4326).

The remaining two valves (V-156 and V-163) must be opened 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 fiFWIVs automatically close.

In the duly 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 SUFP can no longer supply water to the steam generators through the main feedwater lines and the EFW header must be used. The entire analysis in Ref. (3) is based on the assumption that for LOOP, the MFWIVs close and therefore several manual valve opeFlflons have to be performed outside the Control Roo'm, as previously described, to align the SUFP to the EFW header.

If the MFWkVs do not close upon LOOP, BNL assumes that the actions re-quired to use the SUFP are the following:

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

(b) Electrical power must be supplied from Diesel Generator 1A to the SUFP and the prelube pump. Also, the instrument air compressor must 13

be manually loaded on to the diesels to allow operation of the tiFW 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, about( ) minutes.

Since the information provided in Appendix D does not specifically men-

~

tion the above, the applicant should verify that (a) and (b) above are correct and complete.

A detailed description of the SUFP activation and control of FW flow through the tiFW flowpaths should be provided.

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, f ail-open valves that are actuated by a LOOP signal, and also since the air supply is lost 'upon LOAC, the turbine-driven pump P-37A will auto-matically start.

No further alignment 0r operator actions are required to start the pump.

It is not stated in Ref. (3) if control of steam generator level could be accomplished under such conditions.

Proposed Design The proposed design changes do not have a very significant effect on the reliability analysis for the case of LOAC.

The only change is in the addition of safety class ~ accumulators for the air operators of the steam supply valves, V-127 and V-128, as per Appendix D.

14

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 through PCV 4326, 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 possib'e flow diversion.

Neither of these test methods changes the configuration of the SUFP; therefore, no test outage was applied (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 f1'ow.

The test flow rate for the 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.

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7.0 TECHNICAL SPECIFICATIONS The proposed Seabrook Technical Specifications (Ref. (5)) for the emer-gency feedwater system are as follows:

LIMITING CONDITION FOR OPERATION 3.7.1.2 At least two independent steam generator energency feedwater pumps and associated flow paths shall be OPERABLE with:

I 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 1

ACTIDN a.

With one emergency feedwater pump inoperable, restore the required emergency feedwater 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 HOT 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' hours and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

SURVEILLANCE RE0UIREMENTS 4.7.1.2 The emergency feedwater system shall be 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 (later) psig at a flow of greater than or equal to (later) gpm.

2.

Verifying that the steam turbine driven pump develop: a discharge pressure of greater than or equal to (later) ~psig at a flow of greater than or equal to (later) gpm wnen the secondary steam sup-ply pressure is greater than (later) psig. The provisions of Specification 4.0.4* are 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, sealed, or otherwise secured in position is in its correct position.

See Specification 4.0.4 next page.

16

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 signal.,"

Specification 4.0.4 Entry into an OPERATIONAL MODE or ot'her 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 Seabrook Technical Specifications (Ref. (5)) for the Conden-sate Storage Tank are as follows:

LIMITING CONDITION FOR OPERATION 3.7.1.3 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 tank 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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

S'RVEILLANCE RE0VIREMENTS J

4.7.1.3.1 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 c'ont1Thed water volume is w.ithin 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 Recommendation 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 .

17-

It should be remarked that 4.7.1.3.1 implies that the EFW pumps have an alternate water source besides the CST itself.

Such an alternate source is not mentioned anywhere in Ref. (3).

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8.0 ASSUMPTIONS The applicant has made the following assumptions in the preparation of the analysi s.

BNL comments are provided both here and in Section 9.0 when necessary:

8.1 General Failure Data Previous analyses similar to the one presentd here 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 A9pendix C) specified a quantitative reliability goal for emergency feedwater systen performance.

For that reason the component failure data presented.in NUREG-0611 were considered to be too general to allow an accurate fault tree analysis of system unreliabi'. ;ty to be performed.

Therefore, it was decided that 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, Reconnended Conponent Failure Rates, May 1974, c)

IEEE-Std 500-1977, Nuclear Reltability Data Manual, and the following reports from the Licensee Event Report (LER) evaluation pro-gram:

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

NUREG/CR-1362, Data Summaries of LERs of, Diesel Generators c) NUREG/CR-1740, Data Summaries of LERs of Selected Instrumentation and Control Components d) NUREG/CR-1363, Data Summaries of LERs of Valves 19

The data values obtained _from the above references are presented in Table 8.1 for each failure mode for which 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 analysi s.

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 enviror. ment to the equipment represented by the data.

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.

BNL Comment:

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

In some instances the data presented in the referenced sources were either too general or the component data were obtained on 11ke 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, revie'w of the LERs c'evealed that sufficient data were avail-able to extract specific component data for motor and turbine driven auxiliary feedwater pumps.

Similarly, the generic values presented in NUREG/CR-1363 f or safety /re-lief valve failure rates were calculated using primary side components (i.e.,

pressurizer relief valves, pump rel,ieflalves, etc.) only.

The components of interest in the Seabrook fault tree were the steam generator safety / relief valves. 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 valve failures since no credit is taken for them in accident analyses.

To compensate for these facts, the values 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 failure 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 diesel gener-ator, as taken from N'UREG/CR-1362 for weekly testing, are 1.0 x 10-2/d for 20

the failure to start mode, and 6.0 x 10-3/hr for the failure to run mode.

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 underes'timated 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 with 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 applicant's design using NUREG-0611 data wherever possible. Therefore, i

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

8.2 Treatment 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-ponents which are required to change pos'ition 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 state upon receipt of the ap-propriate actuation signal.

Time dependent failures are charaqt,erized by the necessity of a component to maintain condition, position or status in order to perform its requird 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 status (e.g., pump breakers do not trip) for the entire mission time prescribed for a particular transient.

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

f The unavailability of a time dependent component is calculated from the hourly failure rate and a mission time for operating components, or a testing 21 l

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interval for standby components. The time interval used is dependent on the testing frequency, the actuation 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 offsite power signal, or on a low-low steam generator water 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 from CS-4255-1 to the pump.

Therefore, for certain f ailure modes of the control circuits, the proper testing frequency would be calculated from the one month testing' interval, i.e.,

t = L30 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 /monthj/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, ten 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 the 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 .

The unavailabiitty of components which are required 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 genarators would boil dry given an insufficient supply of water from the emergency feedwater system.

22 b

r 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 has 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 pr'oper 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, typically at least 30 minutes.

In any case, running failures were not considered in NUREG-0611 because 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 I hour, 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 fu-ll 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.

23

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 low energy lines, and position changes, other than those required f or 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 (Operator Errors).

BNL Comment: Since technically the SUFP and its associated components are not part of the EFWS, the applicant's assumption that there are r.o 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 nomally 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 th'e 'adlicant 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 perfomed 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 maintained, 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-erato m 24

BNL Comment (Cont'd)

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.

Also, 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 manual isolation valves was confirmed in the applicant's lett.er (Appendix D), although the instrumenta-tion of V-125 was not.

Another point to note is that the EFW feed flow isolation valves are only required by the Technical Specifications 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 isolation valves at the discharge of each EFW pump, i.e., V-65 and V-71, because those valves are operated every month to perform testing of the' pumps. Th'e maintenance is modeled on the fault trees such that no adjacent valves are isolated at the same time. BNL assumes 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 majntenance 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, TOP-37A, to be unavailable.

The discussion on maintenance and test unavailabilities in Ref. (3) con-tinues as follows:

25

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 ccmponents 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 operating 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 be assigned to emergency feedwater system components due to periodic testing.

In particular, the Technical !?:rifica-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:11 and open manual valve V-67 or V-73 to recirculate emergency feedwater to the condensa'te storage tank.

The startup feed pump can be tested in several ways (see Figure 2). One method 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 26 8

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 Coment: The exception referred to above is the. supposed necessity of closing V-109 which isolates the SUFP recirculation line to the Conder.; ate Storage Tank before the SUFP has been aligned to the EFW header. Ref. (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 opening 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 emergenc'y 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 unavailabili'ty due to testing therefore is:

Qtest = 1.4/720 = 2 x 10-3 The test unavailabilities and the conp'oNnts to which they apply are shown in

~

Table 2.

BNL Coment:

The above t'est 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 I has any contribution due to test-ing shown.

Note that the SUFP. can be tested by pumping directly into the MFW flow 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.

27 n

J BNL Comment (cont'd)

The emergency feed flow 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 4

open position which makes them available.

Steam supply valve V-129 is the

~

turbine overspeed protection valve.

It ca'n only logically be tested dur-ing testing of the TD pump 37-A itself.

I e

t* h l

e 1

1 28 l

i

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g

TABLE 2 Applicant's Summary of Maintenance and Test Unavailabilities Components Maintenance Test Total

1) Motor Driven EFP-37B,

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 8.7x[b4 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 29 O

r h

~

8.4 Goerator 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 conmission 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 of f 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 occurred 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 system to provide backup 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 Dnphasis 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 probability of.

1 x 10 4, and errors of omissio,n a probutility of 1 x 10-3 These proba-bilities are adjusted for abnormal circumstances.' For instance, a probability i

30 e

L.

I of 1 x 10-3 would normally be assigned both to errors of omission by the 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 f ailure 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 alignment to the emergency feedwater sysem.

BNL 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 prev.ious design.

In the event of Loss of Station Power, the operator must manually trans-fer the startup pump breaker from But 4 in the non-essential switchgear room to Bus E5 in the essential switchgear room, change the bus transfer switch to the E5 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 Pa 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 operator failure rates for the first three actions are as-sumed to be 1 x 10-2/ demand because these actions, even though assumed to be covered by emergency 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 onlyTx 10-3 Since both V-163 and V-109 are located in the same vicinity, it was assumed that a single operator would be assigned 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 seccnd action-is more appropri-ately represented by the standard failure rate for e'rrors of omission.

Thus, the total failure probability for completing both actions is 1.1 x 10-2 The loading of the SUFPS on to an emergency bus represents multiple opera-tions (viz., starting the prelube oil pump, moving the pump circuit breaker to 31

the essential switchgear room, starting the SUF pump, etc.).

In this case, however, all the controls 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 room 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 operai;or errors that could result in failure of the SUF pump system, special consideration was also given to, the failure rates applied to these 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 emergencies, the failure rates used in this study for these operator errors }is significantly higher (.01/ demand) than would normally be expected for situations where well developed procedures are in place and special operator 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 ahd because of the sheet time (=30 minutes) avail-able for the actions to be completed.

BNL Comment:

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-156 is located in the EFW pump area, which"is a cons 4derable distance from the SUFP, located in the Turbine Building.

In the letter in Appendix D, the applicant states that V-156 will be relocated outside of the EFW pump room area, 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 III-2 of NUREG-0611 (Table B-3.of this report).

In the proposed design, which with respect to the use of the SUFP is not really a design change but a correction of the assumptions made in Ref.

(3), it is only necessary to align the SUFP to the EFW header if the MFW l

~

l i

32 l

f

1 flow paths are not available.

This will be true for both UtFW anc LOOP.

For both of these cases, feedwater flow can now be provided through two separate flow paths to each steam generator.

This is discussed in more detail in Section 5.0.

t O

e* W 6

me i

t 33

TABLE 3 Acolicant's Summary of Operator Actions / Failure Probabilities 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 sta' tup 1 x 10-2 r

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 in recirculation line to the CST

8) Operator fails to start the startup feed 1 x 10-3 pump (P-113) from the control reorfryiven no automatic actuation signal and exis-tence of emergency procedure

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

10) Operator fails to operate transfer switch

?

on Bus E4 e

34

9.0 RELIABILITY ANALYSIS 9.1 Oualitative Aspects 9.1.1 Mode of System Initiation 1.

LMFW The SUFPS is automatically initiated upon trip of both MFW pumps, as de-

~

scribed in Section 5.1.

If the SUFPS should f' ail 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 is 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.

How-ever, 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 Reconmendation GS-5 of NUREG-0611.

9.1.2 System Control Followinh 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 whethe? there are provisions for automatic level control, but it is assumed that control can be performed manu.

ally from the Control Room.

I 35 l

l l

w If, for some reason, the nonaal water supply to the SUFP, which is above the 200,000 gallon level in the CST, is not available, 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 no manual or automatic actions required to maintain flow from the CST.

~

2.

LOOP The EFW pumps are automatically initiated so that ' steam generator level control is now maintained by modulating the redundant AC motor-operated flow control valves on the EFW feed lines to each steam generator.

If the SUFP is also operating, it will normally feed through the MFW feed lines.

If the in-strument air compressor can also be connected to the diesel generators, it appears that the air-operated MFW flow control valves can be utilized to con-trol steam generator level.

4 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.

3.

LOAC Only the TD pump P-37A will be operating.

No EFW flow control is pos-sible because of loss of power ta 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 Reconmendation 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.

36

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 outage c.f 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~

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 whi.le, 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 2.1 x 10-5 Total unavailability (all cut-sets)'

=

47 b

4 e

9 f

(

TABLE 6 i

APPLICANT'S RESULTS DOMINANT CONTRIBUTORS TO CONDifTDNKL UNAVAILABILITY LOSS OF MAIN FEEDWATER/ LOSS OF OFFSITE POWER EVENT 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 4

causing isolation valve V125 to be closed.

2.

Equipment and operator fauli:s: Equipment failures 5.8 x 10~0 disabling both EFW trains coupled with failure of t

operator to properly align SUF pump with EFW system.

3.

Maintenance faults: Maintenance outages or errors 5.6 x 10-6 I

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 (doub'les):

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 10-6 tenance errors that disable turbine-driven EFW pump train coupled with equipment and operator errors that L

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 1B and failure of operator to properly align SUF pump with EFW system.

8.5 x 10-6 8.

Cut-sets with unavailability values less than - -

10~7.

Total unavailability-(all cut-sets)-

5.2 x 10-5

=

l 48

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 failures.

5.

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

6.

Cut-sets with unavailability values less than 1.1 x 10-5

-5 10 J

2.1 x 10-2 Total unavailability (all cut-sets)

=

e

6 m

0 49

6 i

O f ailures* were found in either the LMFW or LfiFW/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.

i 1

q t

e

..p m

0ne single failure exists that will disable the-AFW system under any circumstance.

That is a failure of the condensate _ storage tank such that no water is available to the suction of any of the pumps. The probabil-ity of such a failure was assumed negligible for the purposes of this

~

study.

!~

50'

(

Applicant 's Conclusions The results presented in tnis 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 HampsSire (see Appendix C).

2.

The unreliability of the Seabrook conbiner AFW system during combined

-5 loss of main feedwater/ loss of offsite power events is 5.2 x 10 and for a combined loss of main feedwater/ loss of all AC power event

-2 is 2.1 x 10

. 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-ciuding 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 EFM pGmps 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 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 maintenance of components in violation of the Technical Specifications 51 i

1

can be expected to be significant.

It should be noted that when using the WAMBAM coce, 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'to 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-suits.

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-37B.

3.

SUP1 (pg. A-41)

- No Flow to Supply Header From SUFP P-113.

The effects of double and triple maintenance outages and also pipe rup-tures and electrical and control wiring -faults not included in the NUREG-0611 methodology was estimated by running the WAMBAM code to a cutoff of 10-10 for the top event AFW and the subgates listed. The WAMCUT code was also used with a cutoff of 10-5, primarily to obtain a listing of the cutsets for the subgates AF91, AF127 and SUP1 so that a breakdown of the hardware failures ver-sus test or maintenance outages for each of the subgates could be obtained.

f A modification was made to the trees to account for the fact that if main-tenance is performed on one of the steam admissioi) 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 l

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 1

l l

52 l

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 System 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 Vl9.

The heater discharges are then connected 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 in the valves leading from the CST, all the cutsets which cause insufficient flow from the SUFPS are at least second-order.

For example, 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 faglg,should be readily detectable by the o'perators.

Only one of the by-pass valves would h. ave 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.

53 O

t It should be noted that NUREG-0611 has no failure data for tiFW heaters, either due to hardware or test or maintenance.

Also, Table B.3 has no pro-visions for human errors in nomally 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 f ault trees for the MFW flowpaths are neither necessary nor practical.

9.2.3.2 Failure Data As noted previously, the NUREG-0611 data has 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 position after test

-3 or naintenance. 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 a'nd double check procedures associated with it.

If it has neither, 1 x 10 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 obvivusly seems inconsistent with the intent of the Technical Specifications.

According to Table 3, Section 8.4, the applicant has assumed a value of 1 x 10 for such valves in, the_ case of the operator failing to restore a valve to its nomal position after maintenance. We find 1 x 10-3 to be 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 closed thereby causing damage to the pump, no recovery is assumed, i.e., 5 x 10.

Concerning the operator actions required to manually connect the SUFP to

-2 the emergency power sources, the applicant has assumed 1 x 10 fog MPBIP1610E, Operator Fails to Start the Startup Prelube Pump, P161, as discussed in Section 8.4 The failure to start the SUFP is represented by 54

failure to start tne Prelube Pump.

We believe that 1 x 10-2 is an under-conservative assumption 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- 'has been assumed for this operator failure and this becomes a very significa.,t contributor t'o the subgate SUP1 -

"No Flow to Supply Header from SUFP."

{

~

In the area of pump test outages, we assudie that since the EFW pumps or the SUFP are already in operation during the testing and there is a 30 minu*.e mission success time, the operators should be able to restore the pumps to their normal alignment to allow flow into the steam generators, i.e., outages due to testing have been assumed to be negligible.

9.2.3.3 System Unavailabilities Both the WAMBAM and WAMCUT computer codes have been utilized as previously described in Section 9.2.3.1.

The results are given separately for each tran-sient:

1.

LMFW a) Ref. (3) Design The BNL assessment for this event is 4.5 x 10-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

-5 10 2.

LOOP a) Ref. (3) Design The basic defference between this case and 1(a) is that random failure and maintenance outage of both Emer.gency Diesel Generators

-2 must be considered. The value 3 x 10 is added to the hardware failure of gates AF127 and SUP1 for random diesel failure while 6.4 x

-3 10 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 55 O

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:

10 AFW = 5.0 x 10-5 at a minimum probability cutof f of 1 x 10 2.

Unavailability of the following subgates from the WAMCUT code at a minimum

-5 probability cutoff of 1 x 10 is (from Fig. 10, Sh. I to 3):

AF91:

No Flow to supply Header From TOP-37A AF91 = 2.18 x 10-2' AF127: No Flow to Supply Header from MDP-37B

-2 AF127 - 1.78 x 10 SUP1:

No Flow to Supply Header From SUFP

-2:

SUP1 = 7.35 x 10 3.

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

AF91 AF127 SUP1 M2 H3 M3 Mi H2 Hi 1.1 x 10-2 1.1 x 10-2 1.2x10-2"].8x10-3 7.0 x 10-2 5.8 x 10-3 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 Sopply Header From TDP-37A, TDP-378, and SUFP":

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

=(H1 + M )-(H2 + M )*(H3+M) 1 2

3 i

56 9

TABLE 8 (cont'd)

Let AFWSH' = H H H123+HHM123+HMHi23+MHH123 where double and triple maintenance and/or test contributions have been j

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:

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

= (2.18 x 107 ) (1.78 x 10- ) (7.35 x 10-)

-5

= 2.85 x 10 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 0.48 x 10-5 DMT

=

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 -

4.5 x 10-AFW'

=

i l

NOTE: The difference between the subgates AF91, AF127 and SUP1 and the sum of-3 + M, is caused by the subtraction their components, e.g., SUP1 < H 3

of intersection terms in the WAMCUT code.

1 57 l-

TABLE 9 BNL RESULTS UNAVAILABILITY OF SEABROOK AFWS PROPOSED DESIGN USING NUREG-0611 DATA LMFW TRANSIENT 1

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:

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

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

+ (AF91)-(SUP1)-(V127) where Ar51, AF127, and SUP1 are as defined in Table 8 cia V125 = Unavailability of V125, V127 = Unavailability of V127.

2.

V125 and V127 are locked open valves without periodic surveillance. 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:

W2T = WIT = 1 x 10-3 OperatorENor 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 (MV01V1560A = 1.0E-02 and MV01V1630A = 1.0E-02). From Table 8, the values of AF91, AF127 and SUP1 are now:

58 I

_.__mm

TABLE 9 (cont'a)

AF91 AF127 SUP1 Mi H2 M2 H3 M3 Hi

-2

-3

-2 1.1 x 10-2 1.1 x 10-2 1.2 x 10 5.8 x 10 5.0 x 10 5.8 x 10" 4.

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

(a)

(AF91)-(AF127)-(SUP1) = H H Hi 2 3 + N Mp 3 +

H 1

2 3 + H M )-(V125)

(b)

(AF127)-(SUP1)-(V125) = (H H23+MH 23 i 3 + M H ')-(V127)

(c)

(AF91)-(SUP1)-(V127) = (H H13+HM 13 5.

Substituting the new value for H3 (a) = 6.60 x 10-6 4 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 1.95 x 10-5 AFW* = (a) + (b) + (c)

=

1.95 x 10-5 AFW*

=

LMFW t

59 l

-r e

TABLE 10 BNL RESULTS UNAVAILABILITY OF SEABROOK AFWS REF. 3 DESIGN USING NUREG-0611 DATA LOOP TRANSIENT 1.

Final answer from th,e WAMBAM code for the top event, AFW, " Insufficient Auxiliary Feedwater Flow to Steam Generators" which includes double and triple test and/or maintenance outage contribut; ions is:

AFW = 2.01 x 10-4 at a minimum probability cutoff of 1 x 10-10 2.

Unavailability of the following subgates from the WAMCUT code at a minimum

-5 probability cutoff of 1 x 10 is (from Fig.11, Sh. I to 3):

AF91:

No Flow to Supply Header From TDP-37A AF91 = 2.18 x 10-2 AF127: No Flow t'o Supply Header From MDP-378 AF127 = 5. 52 x 10-2 SUP1:

No Flow to Supply Header From SUFP SUP1 = 1.18 x 10 3.

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

SUP1 AF91 AF127 H2 M2 H3 M3 Mi Hi

-2 1.1 x 10-2 1.1 x 10-2 4.3 x 10-2 1. 2 x ' 10 -2 1.12 x 10-1 1.2 x 10 4.

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

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

=(H1 + M )-(H2 + M )-(H3+M) 1 2

3 60

.m]

TABLE 10 (cont'd) j Let AFWSH' = H H H123+HHM123+HMHi23+MHH123 where double and triple maintenance and/or test contributions have been eliminated AFHSH' = 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 61

The recommendation also states that the low level alam 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 tQ 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 discussed with the applicant whether the redundant level transmitters and alams 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 SUFP normally draws suction.

The entire subject of the CST instrumentation is currently under review by the NRC staff.

One of the design changes identified in the Appendix D letter is the in-stallation of EFW pump 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 can-not be degraded by the recirculation lines.

There appears to be no_ position indication in the Control Room for any of the suction valves V-154, V-155, V-158 and V-159.

According to FSAR Fig.

10.4-4 (Sh.1) for the Condensate 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.

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

37 e

4 g

t

9.1.4 Availability of Alternate Water Supplies In the Seabrook design, the EFW pumps are normally alignM % une base of the CST, as discussed in Section 9.1.2.

For the reasons mentioned 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 tRC 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. However, 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 in the form of emergency procedures governing the trans'fer to alternate water sources for 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 flVREG-0611.

The procedures should include criteria to inform the operators when, and in what order, the transfer to alternate water sources shouid 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 failures.

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.

38

g'.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 conduct 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 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 location and environmental dependency.

Because both emergency feedwater pumps are located in the same pump room, conditions which result in an extreme envi-ronment in that room can adversely affect both pumps.

An 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 consequential failure of the motor-driven pump.. Failure of the two EFW pumps alone are not suffi-cient to fail the AFW system because of the availability of the SUF pump which is located in the turbine building. However, for the SUF pump to be able to supply 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 environments in the room.

As was noted in Section..., in 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 circumstances where the normal flow path through the main feedwater lines is unavailable will this be required.

Therefore, even should both EFW pumps fail due to a pump room steam leak and valv'e V:156 also be inaccessible, the ability to cool the plant will still exist in most circumstances.

Most probable causes for 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 I

39

likely cause of the main feedwater lines 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.

The probability of simultaneous oc-currence of these events is small compared to the overall system unavailability predicted by the fault tree analyses.

Therefore, this cmmon-cause suscepti-bility has a negligible effect on the system.

~

A common-cause failure potential often present in systems that incorpor-ate automatic feedwater line isolation features is the possibility of a faulty calibration procedure causing all isolation setpoints to be improperly ad-j usted.

As a result, inadvertent closure of all isolation valves can occur during 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-versely impact the Seabrook AFW design.

Electrical power sources were found to be sufficiently separated and diverse to prevent dependencies due to power f ail ures.

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 power events 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 dependent ef fects such as vibration, grit, tempera-ture, impact, explosions, etc.

Separatten of the SUF and EFW pumps also pro-vides protection from electrical train common-cause failures due to localized grounding of power supplies.

Recirculation valve PCV-4326 on the SUF pump discharge.

BNL Comment: As noted previously in Section 5.5 of this report, if re-circulation valve PCV-4326 on the SUFP discharge to 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 flow from the SUFP. Also, V-156 will be located outside of the EFW Pump Room area.

See Appendix 0.

{

40 s

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 Adeouacy of Emergency Procedures Since the Seabrook EFWS is still undergoing design changes at the writing of this report, the appitcant was not able to provide adequate emergency proce-dures.

Such procedures should be provided in the future.

9.2 Ouantitative Aspects 9.2.1 Apolicant's Use of NRC-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 include modeling 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 a result, the fault tree now models the entire "three-pump" AFW sys-tem as it currently exists in the SeabPeck design.

In essence, failures of all canponents 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.

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 pow'er sources for the EFW and SUF pump and controls, failures of reactor protection system actuation signals, failures at piping interfaces with the main feedwater/ condensate system, fail-ures in the steam generators, and errors by plant personnel while maintaining and operating the system.

41 O

BNL Coment:

The applicant's fault trees are quite comprehensive and include areas not required by NUREG-0611, e.g., pipe ruptures and running i

failures. 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 Fa,ils 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 0G - 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 the EFW header for both LMFW and LOOP. The normal flow path of the SUFP through the MFW feedlines was not modeled into the trees.

In any case. the applicant has exce'eded 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 maiMeedwater 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.

ihe 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 1

42

fault tree should be reiterated.

As was noted in Section..., fault tree mod-eling 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, loading it on an emergency bus.

It also requires the neces-sary actions to change the positions of the three valves in the startup pump discharge and cross-tie he3ders as was described in Section....

In many LMFW transients, however, none of these actio'ns will be required.

In th'ase transients which result in a trip of the main feed pumps but do not result in either a high steam generator 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 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 and isolate its recircu-lation 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 pump to supply cooling water via the EFW headers in all cases, provides conservative estimates of reliabil-ity 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 thstEFW system 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 pumps is reduced because all single point failures causing loss of the emergency bus j

supplying the pump will also result in loss of the pump.

In the case of the startup pump, the necessity of an operator acticn to-load the pump to the emer-gency bus is also introduced into the system.

l

Note that if the cause of the power loss is a loss of offsite power, the l

main feedwater lines will not be open because of closure of the main feedwater l

isolation valves on loss of power.

1 BNL Comment:

This is the footnote which was questioned by BNL and led to the sys' tem design clarification concerning the use of the SUFPS.

43 s

Modeling the loss of offsite power in the f ault tree' was done by convert-ing 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 but 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.

The total loss of AC power events have a much more drastic effect on AFW 4

system reliability.

In essence the system is reduced to a single pump system because both motor-driven pumps become unavailable.

Thus, all single point failurer 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 MOD 4 (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 functional-ly correct and acceptable.

9.2.1.2 Failure Data This subject has been substantially discussed 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-0611.

The one case in which the applicant utilized the exact NUREG-0651 specified data is for test outage time of the EFW pumps.

One significant case where the applicant's data is higher than the NUREG-0611 is for random failure of TDP-37A.

The NUREG-0611 data specified 3 x 10-3/ demand for Failure to Start of a turbine-driven pump while the applicant has assumed 8.4 x 10-3/ demand.

The appl _icant's data base, is shown in Table B.I.-

44

y l'

V 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 i

All qualitative cut-set analyses and.numdical evaluations of unreliability

~

madeusingtheSeabrookAFWsystemfaulttreefmodelwereperformedbythe WAMBAM (7) and WAMCUT (8) computfr codes.

Ver$1onsofthesecodeswere'obtain-ed from the Electric Power 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 memory requirements during execution so that they could be run on the CDC-7600 computer at YAEC; however, the modifications only affeci.ed the size of the fault tree that could be analyzed and not the nu.nerical probability calcula-tions or cut-set evaluations performed by the code.

Events Analyz'ed

=

.4*

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

were:

.c A 19Wof main feedwater transient with reactor trip (LMFW) o A lo'ss-of main' feedwater transient with coincident loss of offsite o

power (LMFW/LOSP) o A loss of main feedwater transient with coincident loss of offsite power and both onsite emerggEy diesel-generators (LMFW/t,0AC).

In all cases, successful operation of the AFW system req'uired that at least two of the four plant steam generators be supplied with cooling flow from the AFW system.

Numerical Reliability Results A total of. five cases were analyzed with the Seabrook AFW system fault tree codel.; They were the LMFW, LMFW/LOSP, and LMFW/LOAC events assuming all

1 three pumps are part of the EFW system, and the U4FW and LMFW/LOSP ' events

^

assuming only ti,1e two-train emergency feedwater system is used to provide steam I

/, i/

generator cooling.

The latter two cases were done to provide a reference for.

n}

i}

s:

l

,e' 45 r

evaluating the effect of the SUF pump on overall system reliability. The re-sults of the five cases are shown in Table 4.

TABLE 4

( APPLIC ANT'S)

AFW SYSTEM UNRELIABILITY TRANSIENT 3-PUMP AFW SYSTEM 2-PUMP EFW SYSTEM LMFW 2.1 x 10 2.8 x 10-4 LMFW/LOSP 5.2 x 10-5 1.4 x 10-3 (5.8 x 10-4)

LMFW/LOAC 2.1 x 10-2 2.1 x 10-2 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 the'LMFW transient.

Even with a coincident loss of offsite power, the system exhibits an unreliability of better than 10-4/d emand.

In terms of the re-sults published by the NR.C in NUREG-0611, for other Westinghouse plants, the Seabrook AFW system would f all into the high, high, and medium categories re-spectively for the LMFW, LMFW/LOSP and LMFW/LOAC transients.

BNL Comment:

At the June 23, 1982 meeting at NRC headquarters, the ap-plicant stated that the value of 1.4 x 10-3 for the 2-pump EFW System under the LMFW/LOSo case should be 5.8 x 10-4 9.2.2.2 Dominant-Failure Modes and Conclusions The applicant's dominant failure modes and conclusions as described in Ref. 3 are as follows:

Dominant Failures for Three Pump AFW System Daninant 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 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 46

',P

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 essenti. ally 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 + TITE

= ?.18 x 10-2 + 0.11 x 10-2 2 2.3 x 10-2 where VITE = 1 x 10-3 Operator Error 1 x 10-4 Plugging 1.1 x 10-3 b) Proposed Design This case is exactly the same as 3(a) above:

Top Event = 2.18 x 10-2 + o,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 MDP-37B coupled with maintenance errors causing V125 to be closed. This agrees qualitatively with the applicant's results given in Table 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.

62

TABLE 11 BNL RESULTS UNAVAILABILITY OF SEABROOK 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:

AF127 SUP1 AF91 H1 M1 H2 M2 H3 M3 1.1 x 10-2 1.1 x 10-2 4.3 x 10-2 1. 2 x ' 10 -2 9.2 x 10 1.2 x 10

~

3.

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

(a ) _ (AF91)-(AF127)-(SUP1) = H H H123+HMHi 2 3 +'

(b)

(AF127)-(SUP1)-(V125) = (H N ~+.M H + H M )-(V125) 23 23 23 I 3 + M H )-(V127)

= (HgH '+ H M (c) -(AF91)-(SUP1)-(V127) 13 3

i 4.

Subicituting 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

= 104.86 x 10 -

~

63 l

I

l j

I I

l TABLE 11 (cont'd)

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

(b)

=

(55.76 x 10-4)(11 x 10-4) = 6.13 x 10-6

=

(10.12 x 10 4 + 1.32 x 10 4 + 10.12 x 10-4)(11 x 10-4) f (c)

=

(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 LOOP i

Q e

n h

n l

l 64

[

L

'4

e 7

er y

y

_ ~.

TABLE 12

SUMMARY

OF BNL ASSESSMENTS LMFW LOOP Description Ref. 3 P.oposed Ref. 3 Proposed 1.

TDP-37A AF91 J.18x10-2 2.13x10-2 2.18x10-2 2.18x10-2 1.1x10-2 1.1x t0-2 1.1x10-2 1.1x10-2 '

Hi M1 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 1.2x10-2 1.2x10-2 4.3x10-2 4.3x10-2 H2 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 7.0x10-2 5.0x10-2 1.12x10-1 9.8x10-2 H3 5.8x10-3 5.8x10-3 1.2x10-2 1.2x10 2 M3 4.

HEADER

V125, 1.1x10-3 1 1x103 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 AFW*

1.95x10-5.

1.15x10-4 G5

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 agaiu 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, 1

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 agree?,

~

qualitatively with the BNL results. The SUFP must be connected tc 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-378 and V125 still predominate.

The results of the WAMCUT out-put are shown in Fig. 12, Sh. 3-4.

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 Emerg,ency Diesel Generators 1A and IB become significant contributors as well as failure to connect the SUFPS to electrical power sources. Diesel 1A is used to supply power 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 daninant failure mod-es.

The dominant modes are maintenance acts on or random failure of the TDP-37A itself or on one of the steam admissicn valves V127 or V128.

66 1

1 I

___a

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 the cutsets generated by WAMCUT for subgate AF91 which repres.ents "No Flow to the Supply Header From TDP-37A" is shown in Figs.10 and 11.

9.2.3.5 General Comparison The Proposed Design at Seabrook consists of two safety-class EFW pumps and'

~

a Startup Feed Pump.

The latter is dedicated for' use up to 5% 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 th'e 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 Booster Pumps and four Condensate Pumps.

One of the safety-class pumps is motor-driven while the other is diesel-driven.

geh.

a 4

6 67

The Seabrook SUFP exhibits automatic initiation upon ' trip of both MFW pumps and draws suction from the CST with a normal 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 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.

The SUFPS is normally aligned to the MFW headers for both the LMFW and LOOP transients. The Seabrook design limits or stops all feedwater flow to a steam generator undergoing depressurization without canpranising 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 compahison to most other plants is the pres-ence of valves V125, V126 and 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 Possible Reliability Improvements The Proposed Design incorporatbs"' Mutt of the possible improvements which BNL was going to suggest for consideration. Othe'r possibilities are the fol-l owing:

1.

Pump Suction Valves The locked open, manual isolation valves om the suction line to each EFW pump, V154 and'V155 for TDP-37A and V158 and V159 for MDP-37B, do not appear to have Control Room indication. St..ce they are locked open, they do not require periodic surveillance as per the Technical Specifications.

If one of the valves is inadvertently closed for whatever reason and the EFWS is actuated, damage to the corresponding pump will probably occur sir.ce there are no protective pump trips upon low NPSH.

68

Therefore, consideration should be given to the merits and feasi-bility of providing either Centrol Room indication for these valves or interlocking the EFW pumps so that pump operation will only occur if the valves are open.

The same consideration should be given to the valves on the suction of the SUFP, V141, V143 and V152.

Automatic Loading of SUFP Onto Onsite Sources 2.

In the Proposed Design, it is s'till necessary to move the circuit breaker to the Essential Switchgear Room and perform other operations to provide electrical power to the SUFPS during LOOP., it is also desirable, if not necessary, to power the Inst'rument Air Compressor so that the MFW flow control valves can be operated to control SUFP flow. This should be clarified by the applicant.

Therefore, consideration should be given to the merits and feasibil-ity of adding the; SUFPS and the Instrument Air Compressor to the automati-cally sequenced loads of Diesel Generator 1A.

The intention is to auto-mate the availability of the SUFPS.

3.

Technical Specification Limits on EFW Valves Valves V125, V126 and V127 on the EFW header can reduce the number of steam generators which are supplied from each EFW pump.

Pending clarifi-cation from the ap'plicant on the intended usage of these valves, con-sideration should be given to the merits and feasibility of placing time restrictions on the closure of these valves,during Power Operation, Hot Standby, or Hot Shutdown.

l The appli' cant should also clarify the type of maintenance and the expected frequency for the MOVs on each EFW inlet line to the steam generators.

4.

Technical Specification Limits on the SUFPS Since many other plants have three safety-class AFW pumps which are all under outage restrictions according to their Technical Specification, it is reasonable to expect that the SUFPS at Seabrook should be under Technical Specification limitations.

69 O

REFERENCES 1.

" Generic Evaluation of Feedwater Transients and Small 3reak 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 Operatir.g License Ap-plicants of Nuclear Steam Supply Systems Designed by Westinghouse and Combustion Engineering", dated 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).

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.

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

1 t

% g* %Mm 8

l l

l 70 l

I

Tr.nsient Ev.nts LMFW LMFW/ LOOP LMFW/ Loss of All AC' Pl.nts Low Med High Low Med High Low Med Hegh Seabrook 1.g l

l,,

l, l l L.

I

+

l H.

. ~...

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I.

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.i...

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}w-o,6.. oe r..... ia un.v.ii.biiii, n.,,

t.4.

f

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

BNL Assessment - HUREG-0611 Scope e Reference 3 Design A Proposed Desion Applicant's Resuits i

a Reference 3 Design Figure 1:

Comparison of Reliability of Seabrook AFWS to Other AFWS Designs in Plants Using The Westinghouse NSSS.,

71

g 3y

$I%

q To EFW Header E5" iEEE

\\l56

g LM c

4.

% st y3

) gg 7 To IIP lleater Out.let Ileader P m

g ygg7 EE a

o l

SG Recirc. Pump b PCV4326 I

V100 V158

}

lF.0.

Hakeup Line From Startup Fecd V109 CST V159 PC4326 Pump V128 r

Condensate Pumo h

Oi f

M

'N-ygg

. V163 Discharge l

t

,/

l Lube Oil Press V162 l~ Estabilsed

-~~~~}

Hair Feedwater Pil3 Ileader V152 g

V160 (3 Startup Pump Main Feedwater P161 lleader

] V340 l

1 F V341 PSLli PSLli

'- PS4 PSS SG Recirc. Pump Discharge VISO Cont'd From Figure 5 N, VISI

^

~

If e

From Condensate Cleaning System V344 V343 I

Figure 2: Seabrook Nuclear Station Startup Feed Pump System

ll t

0 l

1 i

M M

1 e

o l

t 2

8 4

8 8

6 l

V F

V v

V 4

t 4

2 4

3 l

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2 4

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Main Steam Line A 9

,_ _ _ _ _. ] -][R ][ V171 I

V127 i

I I

5 V94 1

i ESF Pump Actuation Emergency Feed VI 9 Signal V95 Pump Turbine I

l Z v96 1

4 l

l

h-)(2 )[V172 V128 f

Main Steam Line B e4 Figure 4:

Steam Supply for Turbine-Driven EFW Pump i

~~

4

P O

M o S.G.D

-4

~~

M To S.G.A h

h V23X d}-EV25 SeeFigure3/

a 1

i ef To S.G.B

/3l hJ V19 Heater E-268 (V8 Heater E-26A

[l[3 V3((

V14 c)

L 5

z w

Condensate Startup Feed o

a Storage Pump System T

V141 See Fig. 2 Steam Gen.

Steam Gen.

(

l I

Feed Pump Feed Pump 3_ _ j P-32B P-32A l

I I

b G

[' ~l o

- ir_t l

-W 4

V142 VI43 l

I- - - 8 i

L.C.

I l

Figure 5: Startup Feed Pump Normal Alignment to the Main feedwater System (Simplified)

tt.aass.e : ai#Tt F

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ss me-.sie t it}

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Figure 6: Motor-Driven EFW Pump Control Logic e

~

76

I Syste:n Unavailability Calculated by WAMBAM:

1028 AFW-2.06786E-05.

I Important Cut-sets as Caculated by WAMCUT:

I DJT SETS FOR GATE AFW ONDERED BY FROBABILITY 1.

3. 40'E-06 WB1D37BMG MVD1V1250D 2.

2.40E-06 MPB1D37BME N/DIV1250D 3.

2.OOE-06 MVD1V71BCO NJD1V1.'.30D 4.

1.20E-06 MVD1V1250D

.MCD1D37BMK 5.

1.OOE-06 11/DIV1250D FVH1V1580D 6.

1.OOE-06 MVD1V125CD FUM1V15?OD 7.

9.40E-07

MPB1Ebe rOJ MVD1V1250D 8.

3.40E-07 MPB1D37BMG MVD1V125MD 9.

3.14E-07 MPB1Ib/oMG MPBl i o /AM:.

MVD1V1630A 10.

2.G6E-07 MPB1D37ImG MPB1T37AME MVD1V1560A 11.

2.40E-07 t1/D1V125MD MPB1EG7BME 12.

2.22E-07 MPB1T37AME MPB1D37BME MVD1V1630A 13.

2.13E-07 t9B1T37AMG MPB1EureMG MVD1V1630A 14.

2.02E-07 MPBIT37ME MPB1D37BME MVD1V1560A 15.

2. OOE t1/A1V70BMA MVD1V1250D 16.

2.OOE-07 MUD 1V125MD MVD1V71 BOO 17.

1.94E-07 MPB1T37AhG MPB1Eu edMG t1/D1U156CA 18.

1.85E-07 MPBIT37AME MVD1V 71B00 MVD1V1630A 19.

1.68E-07 MPBlio/AME t1/DIV71 BOO MVD1V1560A 20.

1.66E-07 MPB1D37BMG MPBIT37AME 6IC1FPSAMN 21.

1.50E-07 MPB1T37AMG t+ B1EG7BM MVD1U1630A 22.

1.37E-07 MPB1T37AMG MPB1D37BME MUD 1V1560A i

23.

1.25E-07 MPB1T37AMG t1/DIV71 BOO id/D1V1630A 24.

1.20E-07 MVD1V12*MD MCDID37BMK 25.

1.17E-07 MPB1T37AME MPB1EG7BME 6IC1FPGAMN 26.

1.14E-07 MPB1D37EWG MPB1T37AME MPB1P113ME 27.

1.14E-07 MPB1EureMG MPBlia/AME MPB1P161ME 28.

1.14E-07 MPB1T37AMG idVD1V71 BOO MVD1V1560A 29.

1.12E-07 MPB1T37AMG MPB1EG7BMG 6IC1FPSAMN 30.

1.11E-07 MPB1D37BnG MPB1T37AME MRA1P161MK 31.

1.11E-07 MPB1T37AME MCD1EG7BMK MVD1V1630A 32.

.1.01E-07 MPB1T37 M E MCD1D37BMK MVD1V1560A i

I Figure 7: Applicant's WAM Results for Loss of Main Feedwater l

77

c

\\

, System Unavailability Calculated by WAMSAM:

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

CUT SETS FCR GATE AFW ORDEREIl BY FROBABILITY 1.

1.OCE-05 MUD 101250D RGD11B-ME 2.

3.40E-06 MPB1D37BMG MVD1V1250D 3.

3.OCE-06 MVD1V1250D RGD11B-MG 4.

2.40E-06 MPB1D37BME MUD 1V1250D 5.

2.COE-06 MVD1V71B00 MVD1V1250D 6.

1.20E-06 MVD1V12 SOD MCD1D37BMK 7.

3.OOE-06 MVD1V125MD RGD11B-ME O.

1.OOE-06 MVD3V1250D FVM1V1580D 9.

1.OOE-06 MVD1V1250D FVM1V1590D l

10.

1.OOE-06 MVD1V1250D RCA1A74-MB

~

11.

9.40E-07 MPB1Do<eOO MVD1V12 SOD

' 12.

9.24E-07' "MPB1T37AME' RGD11B--ME MVD1U1630A 8.40E-07 MPB1T37AME RGD11B-ME RGD1IG1AME 13.

14.

8.40E-07 MPB1T37ME RGD11B--ME HVD1V1560A 15.

7.COE-07 MVD1V1250D RGD11B-OO 16.

6.27E-07 MPB1T37AMG RGD11B--ME MVD1V1630A 17.

5.70E-07 MPB1T37AMG RGD11B-ME RGD1EG1AT 18.

5.70E-07 MPB1T37AMG RGD11B--ME MVD1V1560A 19.

4.87E-07 MPB1T37AME RGD11B-ME 6IC1FFSA*

20.

3.40E-07 MPB1Dd/8MG MVD1V125MD 21.

3.36E-07 MPB1U7AME RGD11B-ME MPB1P1136 22.

3.36E-07 MPB1T37ME RGD11B--ME MPB1P161ME 23.

3.31E-07 MPB1T37AMG RGD11B-ME 6IC1FPGA*

24.

3.28E-07 MPB1T37AME RGD11B--ME MPA1P161MK 25.

3.14E-07 MPB1DoreMG MPB11o/AME HVD1V1630A 26.

3.COE-07 MVD101'25PlD RGD11B--MG 27.

2.86E-07 MPB1EG7BMG MPB11o/AME RGD1EG1AN 28.

2.86E-07 MPB1D37BMG MPB1T37AME MVD1V1560A 29.

2.77E-07 M:'B1T37AME RGD11B--MG MVD1V1630A 30.

2.52E-07 MPB1T37AME RGD11B--MG RGD1DGIAME 31.

2.52E-07 MPB1T37Am RGD11B-MG MVD1V1560A 32.

2.52E-07 MPB1T37ME RGD11B--ME RGD1DG1AMG 33.

2.40E-07 MVD1V125MD MPBFD37BM1:.

34.

2.28E-07 MPB1T37AMG RGD118--ME MPB1P113ME 35.

2.2SE-07 MPBluTAMG RGD11B--ME MPB1P161T 36.

2.22E-07 MPB1T37AMG RGD11B--ME MRA1P161MN Figure 8: Applicant's WAM Results for Loss of Offsite Power (Sheet 1) 78

Important Cut-sets as Calculated by WAMCUT:

j 37.

2.22E-07 WB1lo/AT W B1EG7B T MJD1U1630A 38.

2.20E-07 MVD1V65 ADO RGD11B--ME MVD1V1630A 39.

2.13E-07 MPB114/ANG MPB1D37BMG MVD1V1630A 40.

2.02E-07 MPBIT37AME MPB1D37BME RGD1DGIAME 41.

2.02E-07 MPB1#4/AME MPB1EG7BN MVD1V1560A 42.

2.OOE-07 MVA1V70BMA MVD1V1250D l

43.

2.OOE-07 Tf>D1V125MD MVD1V71 BOO I

44.

2.OOE-07 MVD1V65AOO RGD11B--ME RGD1DG1AME

~

45.

2.OOE-07 MVD1V65ACO RGD11B-ME MVD1V1560A 46.

1.94E-07

' MPB1T37MG

' MPB1D37BMG RGD1DG1AME 47.

1.94E-07 MPB1ioeAMG MPBIEG7BMG Pf>D1V1560A 48.

1.8SE-07 MPBl i e/MG RGD11B--MG MVD1V1630A 49.

1.85E-07 MPB1T37 APE MVD1V71 BOO MVD1V1630A 50.

1.71E-07 MPB1T37 MG RGD11B--MG

. RGDIDG1AME 51.

1.71E-07 MPB1to/AMG RGD11B-MG MVD1V156CA 52.

1.71E-07 MPB16 J /MG RGD11B--ME RGD1DG1AMG 53.

1.68E-07 WB1T37 APE MVD1V71 BOO.

RGD1DG1AT 54.

1.6BE-07 MPB1T37ME MVD1V71 BOO MVD1V1560A 55.

1.66E-07 M! 'B1EG7BMG MPBlio/AME 6IC1FPSAtH 56.

1.50E-07 MPB1T37MG -

MPB1D37BME MVD1V1630A 57.

1.44E-07 WB164/AME RGD11B-MG 6IC1FPSAtN i

58.

1.37E-07 MPB1T37MG MPB1D37BME RGD1DG1AME t

59.

1.37E-07 W Blio/AMG MPB1D37BME MVD1V1560A 60.

1.25E-07 MPB114/AMG

- MVD1V71 BOO MVD1V1630A 61.

1.20E-07 MVD1V125MD MCD1EG7BMK 62.

1.17E-07 MPB1T37ME MPB1D37BME 6IC1FPSAMN 63.

1.16E-07 MVD1V65 ADO RGD11B-ME 6IC1F PSAMN I

~

64.~

1.14E-07 MPB1D37BMG MPB1T37AME MPB1P113ME 65.

1.14E-07 MPB1Eu/dMG MPBlio/AME:

WB1P161T 66.

1.14E-07 MPB1T37AMG MUD 1V71 BOO RGD1DG1AME 67.

1.14E-07 MPB1to/AMG MVD1V71 BOO if>D1V1560A 68.

1.12E-07 MPBIT37AMG MPB1D37BMG 6IC1FPSAMN 69.

1.11E-07 MPB1EG7BMG" MPB1 :a/AME MRA1P161Pt<

70.

1.11E-07 MPB1isiAME MCD1D37BMK MVD1V1630A 71.

1.10E-07 FVM1V1540D RGD11B-ME ff>D1V1630A 72.

1.10E-07 FVM1V1550D RGD118--ME MVD1V1630A 73.

1.10E GVD1V93AOD RGD11B-ME MVD1V1630A 74.

1.01E-07 MPBIT37ME RGD11B--MG MPB1P113ME 75.

1.01E-07 MPB1 io /AM RGD11B-MG MPB1P161T 76.

1.01E MPB1T37mE MCDID37bMK RGD1DG1AME i'

77.

1. 01E-07 WB1 t o/AT MCD1Eu /bMK ff>D1V1560A j

78.

1.01E-07 MPB1T37AME RGD11B--ME MCD1P113MN.

1 i

79.

1.01E-07 MPB1T37AME RGD11B-ME MCD1P161?*

Figure 8 Continued

(. Sheet 2) l 79 l

g

l l

I I

System Unavailability Calculated by WAMBAM:

i 702 AFW 2.13190E-02 Important Cut-sets as Calculated by WAMCUT:

CUT SETS FOR GATE AFW ORIERED BY FROBABILITY l

1.

8.40E-03 PFBI i 4. Aff..

I 2.

5.70E-03 MPB1T374G 3.

2. CCE-03 t1/D1V6 TACO 4.

1.OOE-03 M'.il1V1210D 4

5.

1.OOE-03 FNn1V1540D 6.

1.00E-03 Fvd1V15EOD-7.

1.CCE-03 0'JDit.9"AOD 8.

4.20E-04 MPBlio/AOO 9

2 90E-04 MVA1V643,t%

i 10.

1.OOE-04 MVD1V12MD 11.

1.OOE-04 FVM1V154MD 12.

1.OOE-04 FVM1V15 mD 13.

1.OOE-04 GJM1V95AMD 14.

1.OOE-04 Gi>D1V12900 15.

1.OOE-04 tt/DIV6*A OD 16.

1.OOE-04 MVD1V654tD t

17.

8.50E-05 t+>D1~42DTDO M/DIV1260D l

F,tgure 9: Applicant's, WM Results for Total Loss of AC Power 80 e

'EEBRDDK:CMr w ; trwa (7kAIN5 + d 11MUe e tEDPUMP - NUREG"06T1 sc0 i

cui SETS-POPW r e 91

TT H-P50RA81L'ITY7GEMT0 0E%-05 1.

iWO ETITA MVD1VETAT1D 2.

5.00E-04 MVDlV65A00 J.

1.00E O6 oVDIVI29PD 4

1.n0E-04 QVM1V95AND u i-r1Y95KOD v

d.

A.vDEr03 6.

2.50E-03 OVXIV12700 i

t.

2.50 erd 3 QVXIVI2TOD 8.

3 00E-03 MPBIT37AME l

9.

5.60t-u3 MPB1737AOD 10.

1.00E-04 FVM1V155MD A1.

5N or:03 rVMTV15500 12, 1.00E-04 FVM1V154MD 13.

1.00t-u3 rVMTVTS%Vu 14 1.00E-04 MVAlV64AMA I

COT SETS-F0p-G ATE-- ::F91 ORDEREWBY-PROP, ABILITY-1.

57 0E703 MPBIT37A00 2.

5.00E-03 FVM1V15500

(

3.

3700E-1 3.

MPBIT37 AME 4.

2.50E-03 OVXIV12800 l

d.

2MOELO3 uVXIV12700 6.

1.00E-03 FVM1V1540D 7.

1700Er03 uVMIN95ACD 8.

5 00E-04 MV01V65A00 ~

v.

1700E 04 r4VAlv6'4AMA

[

10.

1.00E-04 FVt11 V154hD fl.

AT00Er04 t VM1V155MD 12.

1.00E-04 OVM1V95AMO l'3.

AD O Er0'4 OVD1V129MD 14, 1.00E-04 MVD1V65AMD IST MOMENT =

2.1785E22 Figure 10: BNL Cutsets - LMFW (Sheet l': Turbine-Driven Pump P'-37A).

81

NUREG unii SCOP E--

5EKBROUK:L'*F W I EW5 ( 8*AINb + 3iA M P PLtOPUMF t

w fTH PDOBAT3It-ITMGt.,

IT00E 0w

- COT-SET 5-FnW un6t AFT 2 r 1.

iT 0 0 E;0~4 HVDIV71nito

^

2.

5.00E-04 MVulv71900 ST00E;03 MP;1D37 Brit.

3 5. 54 0 E - 0 3 MP R10371400 4

~

5.

l'.00E:04 FVH1Y159MD 6.

5.00E-03 F V-11 V 1590D 17 0 0 E:.0~4 FVM1V158MD j

7.

" 1 00F-03 FVM1V1590D 8.

13 0E;04 HVA1V70BMA l

v.

10.

3 00E-05 LOSP QG0118--!1E RMFRED7Y PROB ARILITT

.I CUT 5FTS-POR G ATt.

A F l ?.7 U

STR O Er03 tiPPrD37600 1.

2.

5.00E-03 FVM1V15900 3.

ST00E203 r'Po ln37BMF -

4 1.00E-03 FVM1V15800 o.00EJ0'4 MVD~1V71P.0D t

5.

6.

1.00E-04 HVA1V70BMA 1 n0E 04 FVM.rv158Mn i.

8.

1 00E-04 FVt41 V159ND 9,

1 7 0 0 E' O'4 9Valv7taMD-10.

3.00E-05 LOSP DGD118 -PE IST tiOMENT:

1.7647E-07.

- - ~ ~

f.igure 10 Continued Motor-DrivenPumpP-378).

(Sheet 2:

82 d

~

,.- i.. a _

. G

+ d A TOP FtECTPUMP - NUREG-06T17CO E j R

P SETBROOK:CMFT: EFWS",- TREIN5 i

~~ TUT SET S-POM AT t suP1 wrfH-PROBABitPTY TGt.

170'0 E=0 5 i.

I 1.

3700F 02 MPBIP16T0t 2

1 00E-02 MVD1*V1560A iT00Er02 MVD1V1630A 4

5.80E-03 MPB1P11300 l

5.

SDOE 03 MPBIPIT3Mt 6,

5.00E-03 MP91P161ME t.

5. 0:0 E'-13 r V K1'V1520D

}

8, 2.00E-03 MCE1P113MN 9.

1700E03 iVD1V1410D 10.

1.00E-03 FVD1V14300 11 A.00E04

.MPR1 PIT 30G 12.

1.00E-04

'RCA150AEMC 13.

1700Er04 HCATAF4;MC 14 1.n0E-04 RCA1A63-MC 15.

F.00F204 e Vf'1V152MD 16.

1._00E-04 FVO1V141MD l

l'7.

AT00E704 FVD1V143MD 18.

1.00E-04 MVA1V99-MA l

l 19, 3 00E'05' COSP RGD1DGIAME 20.

3.00E-05 RCA1A42-MB PGD10G1AME e17 1700E-05 L.0 SP oCA1A93-0B 22.

1.00E-05 RCA1A42-MR ACA1A93-OB j

IST MOMENT =

7.353nE-02 Figure 10 Continued (Sheet 3: Start-up Feed Pump P-113).

D

+

83 9

f

~5EARROOK tn0P': EFWd d I P'ATNS + s T ARTUP-FEEUptJMP - NUR EG-tre i t 3c0 E-P l

l i

- CUT SETS-PORMA'T t' nF91 41-TM-p 0 B'A Biti-T Y-rG E----lT0 0 E -0 5 i

1.

iWVt-us 14VDTV65 Amu 2.

5.00E-04 MVD1V65AOD l

l 3.

A.v'0 erd 4 oVIT1VT29Mu 4

1.00E-04 QVM1V95AMD l

b.

A.uVE703 vvFIV95%Du 6.

2.50E-03 QVX1V12700 T.

c.deEros ovXIV17600 8.

3.00E-03 MPB1T37AME 9.

5.nDETU3 r4PB1T37A00 g

10.

1.00E-04 FVH1V155MO i

11.

5T00Er03 r VM1V1550D i

12.

1.00E-04 FVM1V154MD 13.

iW GE703 rVM1V15400 14 1.00E-04 MVA1V64AMA I

LUT SETS-FOR--GATE AF91 URDEPEL7 Y pR00AOTLITY 1.

5. 8 0Tr0 3 MPSTT37A00 2.

5.00E-03 FVM1V15500 3.

.1. 0 0 E'703 MPBTT3'7AMt 4,

2.50E-03 QVxIV12800 l

d.

2750E303 uVXTV12700 6.

1.00E-03 FVM1V15400 T.

1700E03 oVH1V95400 8.

5.00E-04 MVO1V65AOD l

v.

iW OEuo*

MVAIV64'AMA 10.

1.00E-04 FVM1V154Mi' ~

A1 1T00E-~04 rVM1V155MD 12.

1.00E-04 QVM1V95AMD I

17.

1700E04 OV01V129MO l

14.

1.00E-04 MVD1V65AMD l

IST MOMENT =

2.1785E-02 i

Figure 11: BNL Cutsets - LOOP

(. Sheet 1: Turbine-Driven Pump P-37A).

84

I O

~5E'A73RDOK C00P': EFWS-2--TRTIN5- + S r A5MUP r tEOPUMPvNURES"067T SC PE--

udT SETS-POR-tTATE--AFT 27 W TM-P408W87 L-I TY7GC.

1.00E70s 1.

iWO E70~4 5NITI 07TBFD 2.

5.00E-04 uvD1V71800

.3.

t. nnE207 wCr1T77. H8 4

3.00E-02 RGD116-ME o.

6740E V3 r< G D I T B ~ 0 0 6.

5.00E-03 MP01037BME r.

a.BVt-UJ MPB1U37BC'O l

8.

1.00E-04 FVM1V159MD 9

570Vt-va tVMIVIS900 10 1.00F-04 FVH1V158MD l

IT.

1TODE:03 rVM1715300 12.

1.00E-04 MVA1V70 BHA i

CUT SETS-FOPG#TE-aft 27 O R D F R EO-B Y-P RO B A B I L-l'T Y 1.

3-00E:72 sGD1TB';Mt 2.

6.40E-03 RGD11R--00

{.

3.

b.uoEr63 HPRrD37300 4

5.00E-03 FVH1V1590D b.

5'V0E 03 t4PB1037011t.

l 6.

1.00E-03 FVM1V15AOD i.

1 00E=03 nCATA74=i48 8.

S'.00E-04 MV01V71 HOD 9.

1700E 0%

r4VA1V70BMA 10.

1 00F-04 FVt11V158MD i1.

iT00E;04 rVM1V159MD 12.

1.00E-04 MVD1V71BMD IST M0HENT:

5. 4175 E-O'2 Fjgure.11 Continued (Sheet 2: Motor-DrivenPumpP-378).

85 l

l

5E~A'BR00K t00P :"EFWS' 27R A' INS + STAoTUP FEEDPUMP NUREG-06T1 Scope Cu i SETS-FnR-G-ATE 7UPI vrI'N -7908'A B1 L-17Y7G E.

1700EW 6 1.

3.00E-02 MhB1P1610E c.

3:0 0 L"-12 PGD10Gl'AME 3.

1.00E-02 wCA1A93-0B 4

1, 0 0 E'-12 Mv01V1560A 5.

1.OnE-02 MVn1V1630A h.

o.4'0E 03 xGDTIAe;00 7.

5.80E-03 MPB1P11300 8.

St00E-03 MPR'1P 113 M E 9.

5.00E-03 MPB1P161ME I'O.

5.00E303--

rVM1V.5200 11.

2.00E-03 MCE1P113MN I P..

1700E03 PCATA93;MB 13, 1.00E-03 PCA1AS4-MB F4 iT00E303 3rVDIV1410D 15.

1. DOE-03 FV01V14300 l'6 ~.

1 7 0~0 E 7 0 4 HPR1P1130G 17.

1.00E-04 RC A150 AEf4C 18.

4700E-04 RCA'IAF4;MC

~

19.

1.00E-04 RCA1A63-MC c0.

1 00E'04 rVMIV152MD 21.

1 00E-04 FVD1V141t:0 72.

A Tn 0 E;0'4 FV01 VT43 MO~

23.

1 00E-04 v.V A1V99-M A 1si MOMENT =

1.1766E-01 Figure 11 Continued (Sheet 3:

Start-up Feed Pump P-113).

I e

[

86-le

i O

r~EDPutiP - NUREG-0611 SC PE SEARR00K-LHFW: EFWS 2 TRAINS + STA*7M:

~

WI TH ' P 70B AS ILI TY "* IGE ;-'l. 0 0F-0 7 A F W -~" ~"

CUT SETS ~ron' GATE uv01v1P.500 5.80E-06

A31037900 2."

~ 5'.* 0 0 E;0 6 ' - ~ MV OTV 12 500

~ r VM TV159'00-1.

5.00E-06 MP51037Br4E "VD1V12500 upB1037900~~ 74PB1P 1610E -

3.

1.01E-06 -"'MDB1T37A00 FVM1V19800

~~ 4 1.00E-06 MVDlV12500 0

mpg 1Ple~10E~ 7

,-...8'.70E-G7'~~MP61T37A00~ ~ FVP1V1590 -

5 FVMlV155nD MP81P1610E 6.70E-07 MP81037600 MPR"T 37 A00

' MP 91P16'10E 7.

S.7 0E-07

~MP81O 37 Bht FVM!V159'0D MPB1Ple10E

"~~ 8.

~

7.50E-07 fVM1V15500 FVttTV15500 MPa l P l e~10E ~~~~~

9.

~7.50F MPG 1037BME "P01037800

" 10.

MV01V125"O MPB1 T37 A9E-MPR10't7 A00--'P91P 1010E -

5.AOE-07 11

~12. ~ 5.'22E-07 5.nnE-07 MV0lv12500 uv0tV71900 FVMTV1590D-13.

ST0 0E; 07~-~ MVD1V125MO -

~~ 14 MVDlV125MO MPBID376ME 5.00E-07 FVH1V15900 r4PR1PT610E i

15.

16". --~ S O E- 0 7--~MP R I T 37 AME -

MPd1D179ME MPG 1P1610E l

4.snE-07 HP81T37AME

~ 18. ~4'. 35E-0 7 ~~*MP81037800"* -0VX'Y12 900 ~~MP91 P L610E 17.

OVXIV12700 MPR1P1610E MP81037900 CVXIV12500 --~rVH rV15900 M o a l P 1 6 1 0 E" -

4.35E-07 19.

FvM1V150nD MPR1P1610E 3;75E 20.

3.75E-07 CVx1V12700 3. 7 5 C - 07 ~ ~'MP 810 3 7 F ME-

~"0 V X I V 12 8 00 MP B I P 1610E~~- -

21 3.75E-07 MPB1037BME OVXIV12700 unelP1010E

' 2 2.

~~

2 4, " ~3.36E-07 " ups t 737 A00 -upn1077 S 00

"VD IV 15 60 A ~~

23.

"PG!D37800 MVDlV1030A l

MP81T17A00 MP91 T37 A00-'" ~ FYM1V 15900~--- MV01V 1560 A -

3.36E-07 25, F.VH1vl5900 MVD1V1630A

~2 00E "26.

MPBIT37A00 PPR1037000 '

-FV!11V15500

~MVDiv 1560 A " ~

27.

2.90E-07 MP01037800 I FVMlV15900 MVD1V1630A 2 8. ~-

'2. 9 0 E-0 7 PPG 1037BME - *P81T37 A00-uVoly tS60A

-i 2.00E-07 29, 2.40E-07 vPR1037BME HP81T37A00 MV01V1630A

'3 0.-

~ ~ 2. 9 0 F. - 0 7 ~-

'FV"1V)S900' MVDIV1560A 31.

FVM1V15500-FVH1V1550D FVP11V15900 MV01V1630A

32. "- 2.4 E-07

MP91037 BME

- FVM1V15500 '-~~MVDIV B60 A -~

2.5cE-07 33, MVD1Vlb30A

- 2.50E-07 MP81037BME-M M1V15500" ' MPB 1 D37 800 ~ ~-- "MP 34 2.50E-07 35, 1.95E-07 ' 'MPHIT37A00 FVMlV15800

iP81P1610E l

~36.

MPHIT37A00

-~ MP 91P 1010E -" ' '

37.

1.74E=07

^ MP 81037 600 ~~ ~~ ' FVM1 V15 4 00 MvDiv1S60A

~1.74E-07 39 1.76F-07 HPB1T37AME MP9)D37R00

'38.

~1.74E-07 PPFIT37At1E" - MP01037800 - ~ MVD1V 1630 A OVM1VQSAOD MP81P1610E f

40.

"PH1037000

NP81T37 A00 ~~"VO1 V127.OD, PMPB1P 1610E -"

41.

1.74E-07 1.6BE-07 MPBIT37A00 FVMlV15900 MPn1Pll300

42. "

1 74E-07

^FVH1V15500 MPalP 11300 -~ 1 43.

45.

1.68E-07 MPHlD376ME MPG 1T37A00 MPB1Pll300 l

~~1 6DE-07

~ :iP51D37800 ~

44

-1.6BE-07 MPHIT37A00'~

"PRID17900

-MPalPil3*1E' '

.PRlpt6 DIE MP41037600 46.

1.68E-07 MPH 1T37A00 "- 4001037800 - FV4lV ISP.0D "--

47.

48, 1.6AE-07 MPH 1T37A00 BNL Cutsets - Top Event No Flow to 3 Out of 4 Steam Generators Figure 12:

(Gate AFW) (Sheet 1) 87

i

/

r r

s s'+

49.

1 90E-07 FVP.1 V 15400 FVM1V1590D MPR1P1610E l

~50, 1 ~.'50 E;07*

MP BIT 37 AME FVMTV159FD MV01TtS60A 51.

1.50E-07 e MPG 1T37AME FVM1V15900 HVD1V1630A

-52.

---1.50E QVM1V95A00-Fynijy}{9QQ tippiP1619E 53.

1.50E-07 HPBIT37AME upB10378vE ttV01V1560 A

~ 5 4.--" 1.5 0 E- 0 7 -4P S I T3 7 AtiE

--~"P6310 3 7 BM E--

-MV01Vle30A.

55.

1."0E-07,

FVM1V15500 FvM1V15s00 MPB1P1610E l '. 5 0 E 07 --- 7 P S 103 7 B M E--F yvTV15% 00 MP9tP-1610E A

' 56.

DVM1V95AOD' MP91P1610E 57.

1.50E-07

' uPalO37 AME

- 58.

1. 5 0 E- 0 7 - --F V M 1 V 15 50 0---'1V 0 tV12 7 00 M P 9tP-1-610E 59 1~.'4 5 E 4 4 7 MPRIT37Tc0 VMTV15900 MPRTPrl'3ME 60.

1.45E-07 MPB1T37A00 FVM1V1590*3 MPB1P161ME 61.

1.'5E MPBIT37E00 FVK1VI5900 eVMTV15200 62.

1.45E-07 MP91037800 FVH1V1550D MPB1P113ME i

~~63.~~ ~1.45E-07 -~~'MPRID37B00 FVMTV15500 WPR1Pib1ME-64 1.45E-07 MP81037600 FVM1V1550D FVM1V1520D

~~6 5.

1 35E 07 MPBID37B00 OVXIV175Ou

>WD1TTb50A 66.

1.45E-07 MP91037800 QVXIV12800 MVD1V1630A i

~' 6 7.

"1.65E-07 M P B 10 37 8 00 ~~'- 0 V X 1'V 1'2 7 00 MVD1V1560A 68.

1.45E-07 MP81037800 OVXIV12700 MVD1Vle30A 69.-- '1. 4 5E-0 7 FW11 V 15500 " FVM1 V15 900 -- MP B 1P11300 ----

70.

1.'5E-07 MPB1D37BME FVM1V15500 t1PR1P11300 i

~ ~71. ~ ' " 1. ' 5 E- 0 7 '

.M P a l D 3 7 B M E'~

.. u p B T T 3 7 E 0'O'

[sPS 1 PIT 3$4E 72.

1.45E-07 MP910378HE "PBIT37A00 MPG 1P161"E T3-

,1.45F-07 MP910373PE MPB1,T37AOO FVv1VT5200

~

74

. 1.75E-07 FVM1V15500 FVM1V1590D MPBIP113ME

~7 5.

"1.25E-07' FVP1V15500-~~~FVMTV15900 MP91FL61ME 76.

1.25E-07 FVM1V15500 FVMlV15900 FVM1V15g00

~77.- - "1.?5E-07 ~-"-QVXIV12SOO FVMI V1590D" MVD1Vlb60A 78.

1.25E-07 QVX1v12900 FVMIV1590D tiVD1Vl630A

~7 9. -- -- 1. 7 5 E- 0 7

--~0 V X I V 12 7 0 0 --~ -~ F V M 1 V 15 9 0 0 -

MVD1V1560A 8 0.'

1.75E-07 QVXIV1270.0, _ FVMlV15900 MV01V1630A 81.~~~-'1.P5E-07

~MP S 10 3 7 9 M E-~~~ FV MTV 1550 0 MPR1P113ME-82.

1.25E-07 MPB1D37RME FVM1v15500 MPA1P161HE 83.

- 1.25E-07 MPR1037BME ' --"FVH1V 15500 '---'FVH1V15200-84.

1.25E-07 HP8103TBME OVX1V12800 MVD1V1560A j

85. ~ ~ 1. 25 E-07 MP310 37GliE" "-~ 0VX1 V12900--- MVD1V1630 A i

I 86.

1.25E-07 MPR1037BME OVXIV12700 MVD1V1560A

~87. "

^1. 25E-0 7 - ' MPB1037BME~ ~~"' 0VXIV12700 PVD1V1630A 88.

1.01F-07 MP01T37AhE MPR1D3'7R00 MPR1P11300 IST tiOMENT:

3.8976E-05 CUT TOOK 9.698 SEC$

Figure.12 Continued (Sheet 21

~

88 l

~5ET8[UU GI000: tr W5 T Tri7 ipa. TrAnTg:r p-ggptjgp,, NUREG-0611 SCupt h

- CU T~~5 ET 5-#Tm u. Tt ev w i m'PnDR7isIt ITY.GE-570 OE 07 1.

3.00E-05 MVD1V12500 DGD114--uE 2.

c.40E-06 AiVPlv1250D

.PGII'1R--0U 3.

5.80E-06 HPAID37800 vvD1V17.500

. e 2 r- 0 e wFIT37AUO wUDTTS::Pt MPtTTFTbTot 5.

5.P2E-06 MP41T37A00 AGDlln--ME RGO1DU1AME

).

). 9 0 t- 0 0

<iV D l V 12TOU

VIIIY1390T 7._

5.00E-06

-*4 P H 1 D 3 76 M E MVDtVIP500 8.~

4.C01-06 FVM1V1F500 aGDi14--Mt f4P F1 PIT 1BE 9

4.50E-06 FVH1V15500 DGD110--ME-PGD1DGlA"E 1 i).

J a 6E- 06 uv1TIV T25Fu 7GDTT C;f>E 11 2 70E-4 MPb1T37AME 3GD11R-uE MP91P1610E 12.

c 77'O t - o t)

IGUTT3TEI4L WOTTETIME

.<GDTDG'lC'it 13.

2.7.5E-06 QVXIV12800 A G D i l i---M E MP91P1610E

~ ~14 e.25i~r06 cVXTV17eco uBD115--mt RUD1tGTAT3t 15.

2.25E-06 QVx1V12700 AG0118-uE MPB1Ple10E 16.

d. dE 707) cVXTV127UD DGDTIG=TE RBDTDG-17;M t 17.

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APPENDIX A

/

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

4 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 (seeTableA.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).

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TABLE A.1 5

F.

SYSTEM IDENTIFICATION CODE C - Condensate System i

M - Emergency Feedwater System Q - Steam Supply System R - Electrical Distribution System l

5 - Condensate Storage System 6 - Control / Protection System l

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TABLE A.2 CDMPONENT TYPES b.

II.

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 OB - Piping greater than 1 inch but less than 2 inch OC - Piping greater than 2 inch but inss 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 enun-mw e M9 N

e

4

[

TABLE A'.2 (CONT'D.)

\\..

PB - Centrifugel Pump I.}.

RA - Control, General Purpose TR - Transformer TU - Turbine VA - Check Valve ~

VB - Relief Valve VC - Yacuum Relief VD - Isolation, Shutoff Valve VG - Flow Control s

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h3 0

v~3 s-s!, a E ",i_ s" l.E a e-s-j al5 O,s] s83 -gE gg 2 E. r: 7 gNY$ m 3

== 4 034 .3

)

c w l r'. es-s nas! E 1 ~ 5 m a ET = r* 2 4 Er a 3k u 5-5 Gt E.* 3.W I g < 8 5%

_ sy
W 5'

22 W-ag= -jv =a a 2". E: ' .a s Eg: e 2 Igg-g e= M as is 3230 j z saae 4444 'TN 1 a:g! 3 = _1 A= a.s 5 g V 22:::- E .t u r E 4 0 WW g37; g 5 * - W 3 Eg>g g IR. e E e e 5 2 *5 J 5 ga .a E z --E IG X =g n s.

== 1 75 C 4- "I g A-20 j

k " h t, I a -- I I E" 3 w .o 11"8 g 24 ..3

.ca 85I[ ]g f

g O. ~ g. = -t f"

t

~ 1 s3I a = EN 3 ~ 5: I l. gas g **? W ~ = 2 m{ 1 8-5.- 2 ~ =* =aw e8s 2 2$. 5'3

EEJ, E

a5d =s. sw - "d. OE J s 3-31" 2 535" c2 au 3 w. ~_ N is Os ( 3220 ] 4444 4444 3 g .I', 3; } ~~:t O~- sg=3 On a sa

===, w e 9 6 O e'Y g n ',4 0 ~ gp5$ ~ ESEj E .z e: O: I 2> E l l A-21

" - ~ W l i 1 O.!.: 2 spin O! z. 5 5 ** h a ~ ~ ca B E.h. [ s.-1, ?,E ~ AREE I I zu E Es a

5 g

se = Ek gr -s, g os. gg: 24-

==d.

z
45 d

YEg"; s --Gd syd -o =a ag3 a e-s M s 3 -. = I - em se = a444 aaaa ege n _ B $I E h O n u as UE 5.54 V-I EE!v f -.ss W N H .siv na = _s: v-arg3 E s m-. 1 1 l -T5 I. ~ tm .e I A-22'

a--A il CE taf a = E 6 E rm,I g I a g$*agOs I s i. 6 4 e . - {- 3 a! " 5 R r! j;aII Os in 0 I I I zu .R 1: n , ~. gm c m 4 ed = a 2 ~ n n C E-

t 5*

2 = 8 a =s- =mv e4== I 5'3 g: gr - = _a a a- = e Ens s I!_g er 9 25 5 22:::2 [ = 4444 Lisi b5 I a*EE R ss a r 4s j:tI y 1 a W1._ = s E 4a. mE:q g r== O=- t= as ag[5 I s*=! OE vt-I: i n c>. k, I in Os -l 3 l A-23 f

l l l l l o-e l l 20 = sd C L = %) E n g = CS ~ utty a-2 I a 2 :. EI:"E-EX [m E_E_ ,W E = 5 I lwE ~ re g"gae 34 "E a n-a a.a. 5 HV Es-C wE w s' SilR ~ SS IXi E5 1 EEE ag:E .= $ R Ez = 2 y EU R S ge ' On _ jil8 j -~- as sEtz. I 3 [*==. e > = Ig[2 4 e g $_h s a:c: I gmme I ame gw E i .XX9 = 1 -A-24

g .I = Th y.: ;.v - y. y ]

s. i s

mn = xw.e at A E a S Sa 3 1 g= at a M C (( .f.wps-Yo ? R 'Jt ~ ~.e h g ". " - .YS .Y =OE 5 '3 Os }C-$ E ogu= 'g. = 31 $!

E J5LM

*I 1,,,

498 ~ 3 EE: M d' g a::: xEx 9 Es-u 5 m' =, .%=ES E g3 ',Ft.4.., -r - .W 3~g h s';; gS=A 3 -::::e r : [*. h. A2 u;. 1" ..w.. ..'.;?f.- ,e 'W*W 3 p I w...- V-gCI=- [ .=us ..?,-- A-25. r l

i 1 i ) l 1 1 g.- -LS C ~ "3 E E i .g

s

^ = U =: =g R Ca hh ~ = si s ss - ' c .-Ed - 8 ~ W 4 N E $~ ~litu s!-!

5

$6LM !.- S ~ll.l au' a ed .m e n s a. ,1 9 EEE .4 R ES$; 2 O)' re-4 e=s. %..=.15 13 l \\ -..e- % Y w.*3 IgI1El"' E u "wEr l mza- {J5S ns-8'- BOOo E mang y A-26

O 9 9 O =i a => = 9 unne VE as I E gg = E E O S E. C w " A 2: M = -NY S w 4 WEE a-a = g,.o y 31"y

~

g-E 'ME JZtn EU{ g -E ~ E ~ O S E. w - o N 4 m g 5 $2 5 ge ena -05gv2 ~' t=~2 .2 R 1: 6 e F s=s-3'e - E- 'g

s. m g=

e 9 9 Y sin Y a:4: I 'g b tus tw sa-2=x y e A-27

6 h e on.e m b4 en e.e 'e N.e

e

.mo 9 e

O N

eg > a.e 44 I I N S e 3 es,e su O e. em

== a.m t e. = 3m Ue Ie e qs .g e m a,

== eg a e> e I en. >= te. d G. R = O d St W. te. Og eE "EI EE7 7 a. .o te > Su. .d e. E 4. tR ap

== tem -~ b E< ~ g g a. N U- @ o O e ~ ag >e e U Be 's" *r .= E'S f, E O =.=* .e e

e y

e e 3 2 O g 32 / ~ ~ ~ .e = e a. m e o 4 k. N tm. ~ s e 4 . o N - a l I

e Silh m.

N.mo ~ >EM g l ' a.c y ;. S i 11 i i 'l' i I c ,e 1 i 1 -l 4 i i A-28

b* up. e .m. mw s. N . u ~ mea Z $e 4 ~ 4d R .~ 7 g g MIR /. .i. e = e s> -e 44 an-a t 5 ~ gg i W"* N ew

C i

d .nn om* BR bs 4 K5 c ten N =e i b j k 2 5 ?~ c$"=. = b Io l Edu f "E 5 i -s gy O. a.: L- .o N N me I s2 E R. g s ~ s g 4 9[

d i
- mg*h TN i

t t g w i I ~ .e t mem a 1 sa m ? u l' L Y 4 es s, g [ e -g n c ne-- Og=3 g 4 E EoY' i W .? L f rx 1: ' f, o A-29

O 9 2 2S T is

g

=._ = 23 2 4 aS i e u e 3 up 8 5 5 a3 4 ~ S sO 3 .I = O tud & &= g N - me II = d a. I d at '

s a

g SYE E *? 5.6 = 4 -25 4 2 aBE f ,.O ~. 23 2 7 - .(** oW 43 4 u - -s-. $6 sh = g B2 = / ~ f( ew e

b O

4 V - <- 4 3 4 i 4 3g w k 8 s. e E N .e W =. = N man q -. :- g =.e. = g O ? e 4 e5 2 By6 3 4 e e* O-6 g t a 3-I A-30

= 3r E ~ 2

h l

a J =E f UNh5Ol $S E f,==0: i =r in =. ...a .-_= w = ..= -o 3 bye 5 2 0= 2 BE*C E l 3-2 2 =E e ~ Eg=.g I [04 g-8 "U E. 20 5 . Z2 tty m =*ri .? = 4 m .$s

22 2g_W" I

=8a..- a -a = 2 = = EEEE = 2 4 o E <

g =

-. ~

= 2 Dl-- 837 O c

0 a s=

d5 g 2h2 'O' D 'Sggi tztiv . ~.. c. .s-s ggag

==ce ~~ L85. s G__E : ~ i EEEE .= /, = g 777,. B5ga 5555 w=-o

==

gggg r 3 B. W: E .. _da g g fes i

ED 2

E =3 B50" ]_ =at EMW" B 2 S. W E y 2

- s 2{h

~ = E-' m 9Et .2 Kaa K 222 98 = g ggg EEi A-31 e __m ._l_

e ,ME s v am !== is^ _ *

s. g g e.$^Ju s

U-x-

s. r a_

i .sa__2 s ed. .s-N*sg - = .l.1 a s.= g s_ _ su - an g5E. E=u ~ M 8 sa,g_ g 2 "'._4 Detn m 1 O .s_n gg=st_

==33 = 2 4M EEEE 4 l R EEEE = g

5t Ea gY?

355. M37 O g3:2 s -t5 "c' Q e s s.1 son 2.s : 2 =~~~ C ~ m

t.s a==

M=se Este = ssEs 1 s.... a ~~~~ 7777 m!51 = 5555 o e a e6 en wwd. ~ Ed.g 8 p ~ "I '. w 3 G fi 2 g _ eis = --e a E B5 " g Eg ~ 5M2 3 3 ~ g3 A /- IN \\ SZin 1 I%I _S 98 a ;;; \\ s l

C 553 Eff 1

i A-32. t

M 9 e 4 E. -oa. kb N W smina -5 W -em

8.=

g E s =.. t 9 E ~2 m E' 5& 5. y 3,=2 E.* 3=4 2"3 5 5 8" g~ b t 5. .:5 v 25d R 5 0,* P 2#5 4;5 = ts"3 3 5 S =I M s -f I ex.s ~. L*En g-- =. t 4!? O- -r! O 5-gI g g,,, e = = = / f k E e 1

v. !.

= E e l l 5 W N

@q

? ?"* / l O 4%

d ! E i. R .g!g_"oe = s a.? U =_ ^ Y-2

2 w

I { mk l 2 ~~< s.. s8, i 2587 E83 g 3_ 5 a X- = i = I I oe s k" ~ ^ := ---e =sa=l g5 V = k a wa E wsi O=E -. = = '85 t k e~ E E c6 : s m S E,4 8 .o . : ~1 t ~< ~ ,= z "g5 3 k' /, n. ? Kin a sg .e g .=_*= e-g =.,, = 2 = s= .a. 2: n E 3 .30 t km I I tttn E. __v : 23c5 1{:s- ..gz d .?,,,=.,= ,ga 3 as g.- g .E.== ww ME = M ttta R ~~ "2 '2 g g< sg fg ? 4 St ~ 5

g:EI 4

8 40 E. 1--gX" EE A-34 ~

l = mt 5a J n

e 1

=1 D=- n 3_

iIrh IEiO!.

i -.x-a. ,a -

n. u s

r a .w-21!g 0s = -e e t 3.,. g

,s.

1 5:.s.0_ Oa !=iiOa g s ,I.. 8 g = .e= s s 2;i 3 -i 1

rgaBre-Og l

1:58 0i ~ l' =! s rg 5 o_ n s = = = = 1 a = a-s,3 g G.-.!=EnOm - != g Os 4..- r 5 's=s - n== s ru 5 m 4 = = n=. C*g., = y to!# S ? = . E.3 -a B5 R3RR a e=== -ir*E eg:- a-gggg- 'a=i' a g s!. I-

==, ag: ~ .~ __ 9 - E_t_t = =- j SE=4 22: ( i ESE ~ ~ ~ A-35.

e e D ,!.Af= "EM 5 + =- -= e. ~t ,i -.n. = .E o -g ( maer e-a-3

2*E e

~ s N. 25W3 .=., sa-s i .a 1 - ! =<. ~ 5-2 5 eis K_ b %e" $#d l ,2cEE a x 5 _,, ca e A 2 a 3 8 '. t- - g R =~, 2 g . *W .5 2 2 l,,

gb

E 8 tit a 5

t Et ~!-$ g ~ jAas EE_ .g S E 58 (1 9 E E = .~- = ;,. : c 13-A L-etta .5 5 E;;;g 358 hi I g --R 2 . g' l E E3W l3: E 1 -rs .a t

_s l

As Mi I "sEb 35 E ~- - 1 3: W g =,a

ht

$3)$ 1 2 = s a5 = A-36

' gh 6 0 1 ~d=s t-I = E 2323 g = r = gey = - s2 322 3 T ~5 -IE T { Tg 1 I = EgOg = a = 5-2 ge E. rs. 3:=a = e g m. - zz ~ 5 .-eng = 3 am ,= ~= 2 .= E 4 , a !E~n O E*@ E o a! O5 ags u n mn g e-il es ~~ s a

== = y 3 58 / 2 3 5 k"$E E g =l=; I. 2 u .n n

ins, y

- =

= s-a us,i ,11 a AST gd tttn ~1 =5 2. 3!1 i t=* a===a - "h8 5555 ~' I in Q~ g. 1 A-37 9 ~-

1 \\ 9.* 58 - $$ lie.Cn E =.=. /

g. ^E gz =5 S

s,=s Us 4 d3-3 am s e e. ' !s! I

g. ^I g=r.n.

e es g Jg_

n

a,ss u- ,u us g EEI5sOE

==5. 3 = t g. - Sw I -.o I s=as us c =- = E..- m. suf R es gg -===.

g. s== e 5

.-= x-Er n= b k. s .~ ~ .g . go s. a 5 s.il r_. lE5") E W 3 t_ g sus a N E!:s ss52 4.8 a se g gggs ii. ...m a = m y g .h gk 'k!j l gggg / T WSE. 4422 " 4 R~~" E 22_44 R2.3 gg 5555 + ~ a8. =51 OS 2: r, ( A-38 g. -w.. -..ys.

M t \\ f o W

0e5 a

E 1 h.- 5".3 I =""8 2 l 1**l5 g I =52* 2 1 B=#5 a 2355 2 ~ "=2 -E .2 2 M 3 E

== si c i-a e x s a R* N' j =. ~ .I". 252 W l"="I S.. ^I e=.- s=" V a. E St=" "5l M = -== = W. =8 Y .y "= = p

=

4-r 23 8 = n!g y. -I y E W !""l h - =5 .g n s

g

= 4 4 = ]= c. .2 L 23 _,c sm 2* E2E ]e= j=ia E g = _= R =.-

- E MS R

S 2 a:2 yrs ( s,. t sge em .2 W2E gs= , gia 8== _== i All A-39

g 9M Sh6 W " 3 E ~ kh 2 -- = s. o-f t#% .= E E=-= a Ws2* 2 C3 8 E X - 53 2 E=4s =sas 5 suas 5 = = W 8 'O g .mMg Ea = = = ans 4

==s y

== s W W Ev= x ~ 3 5 9 *' ~ = s-o = = = f

s V

i = [ <Es g ~ g= V 1 s = / 3 Y = wn"l" 2 ~2 g Eu

-g vs-2 m ~:.

Vg -s = s==s .E E,5 0 SaB "Ig $x 4 "e \\. sil O: 5.a 2 s o e A-40

i im I i gsve _ = = = = agagg , w =s-n.2 5..

g c

=

E EI:s= iEEEE -s. 85,3y 37((3EE E'EE555] 85 jus E E ri a 8,~' liffffff u i s_ g -*W i saw a s. s:g-g -e .m M = s==IsvOs E5 R -h= 8 -= = ante 3IE 1 5555 ~ s!= sass k s -s o E-Ee s

-=

a s-a 5 =- sa-e== n, e :s.2 e-s 84 s ss-a- g333 -e= - a --:= w ,e s: sass q::: / ngs s s- =4. s s =$_e a rz es R sev - ass . ~= - - =

--s man eeg E

=.: .!.-s a -a -=rg=. =- w 6E-g g s,

n EE=E-3R]RRE

- al'! = = = s.~ s=*

===e

111111 a= E' s g na -v av .E s _ I k 3 _ _ _ 3_ 3_ M e er 8 T r sE 22 -=E'.I. ___I_I_ s:s t-E A.41 a X-II 81118

6 0 ed @W M

==u u =. .sV.

e. c =

g 1 g T=

g5 E

1 k ss-E l w E ~ a. m ag. s g fe-Y L 5 E-N- l :a. R {1x= ~- I. O' W w ~ E 85 m 25s OE Es! = a

-=

,E = k l 'E 5 Ess Q=E _ :g i-M w m. E De M - ~ 2 :l4 t =ws5. r a 2 8815 i s =3 =5: 8 E*- g I o s*- a w E a d ., = ~

3 9

4 .305 = = =. V ovats g .a g:::a. g i E W 5 E -!=. E 3 5=I S

"C y

/ e i* g ~~

==s -wv -s -- y" =

3 M

E==Es .:a=

Is y

-=3 g -- -- g g t Zix 51 9 e -31 5 . 4s; oi =- = g o,. g E E e s., = l n.: 4 E4 d 5 3 0,32 g a:Wj 3 En2" E

5 y

= -3E A-4 2.

6 8 5 i ij l h 5 ag Y l MwN d Ih* I se t

=

1 -.w. b G = a-es! =g. U=1 ~ g i "M. 5 I~ a i = W h ="o Y 5=2 = i f2" .= 7 d8' E = I:E$ ' UId" V~ d t ".g 2 I j a A i 2 ~ g 3" $ E o y 2 4

e!

mais n! 4 a

===a n, vg saga.

=

3 = =~ s toe G.af a W52 i / R$2 4 E i ~

L 2

i, w sg= ts aa a_ 50I g gg h. i a ,5"' V SG V~ $~ s $ 2-- i a t i I h I2WI A4RM =

== EE .B E 2-E 3344 .93: ^m =a Us 2 RE*= $555 =

M

== gS3-F RU C" 83 i M2g .:... l =s * -1 2344 E'*. i 2 355E R.- 2

==

,e s-J A-43

6 G w O g . s a p 2 25 5

= 5 2

O C=.a ~ = -- = 5 w Q G E e R ~ ~ ~ h h I a O ~ w u tea s-S-5 - e a .g e, M e v vs-- a u

===- Eu 4 Erom x Ev.= 3% D 5 C148 I C& 3 C1 -3 3 = I bo e. = 0 G e 9 e 9 /" S 4 e A-44

90du IEE a ~E 2 E ~~ s M' E Ed hkh

===. Id C 2 N~ a 5 -se 52 2 ~ ~*N 4-o g ae Y g h 2a: e 4 o 5 ~ ~" '5% B If5 [kJ d 2 -a 225

=r--

Y<eN~ ss - g 3 3 f8 + g = e "7 m= E MY W E"$ N*d g f = .= Y a t 28:" WW* D i = 84; 3 V. ~ ' ~ ~ -

,g; res o=-

g h-6m* m er n3 tw) TUI G E =: =: : = E-EEE c ~ E s8 e e - O A-45

O O 11088 RRR ig-S-E g;g as: b ~ EEE e n. E a Ed C-E 3 23 E 5 = ?. E n.a tad x s*M C 5 " '. --a 5 O$Y E T N I w w gs $g. g W Oa

-=

o.5B a- -. g {- M a i w CJ' oa. = g f w E w S dg D e-e W E taf 4 e.d > 3t 2 s :w R = -E : Ex: s.x~. ag-a a b dw k O X _, i ~" ten CE .g - a xx,g er

c. =.

r ~~~ j .,. E w E... l EEE 9 e e E .r - 4

O l aua R g{3 ..M E ~~~ 2 ?.ES I..EE w 2 R 3 es se; y E59 Os 25-ase E N 2 l5 5 I W5% = 22=, ~ bE N EEC seniu =*2: .g ~ t n p.- 8.1" E. I. (- $5W </ N5 o~

g g

~ Wes BW Em. = 5 g 25.- 52: I-T- ww BET 2.- d8E Sd5

== = are g 1 . "Jk - s i e tZW3 E e-R EWG ~ 53 E 20 0M AgS EEC an-D6 m. l l l l 1 i l A-48 l

9 e

4...sm== 4

? 1 l l l 1 APPENDIX B FAULT TREE DATA 6 2, m ' e e ' s>g* h 9 9. G 9 e

g.

e b. e e S

0 1 b l b A, ,b i. ,t. ~. ~. ,4. ,,,,,b ,b ,b ~ .o.. ~. . o

a. o.

a. e N ~ ~ . _. a. .e.- b o. . ~. ~ ~.. ~ - 4. ~ - ~ ~. 5k a ~. c ~. .o. o e. o. y 3 ~ ~. -~~ 3 g. 6 k-4. , R. ',.E. 2 b. , m . = = ), ~. - ~. - ~ s a. a o ..N

a. a %

Na %e. n. a-a-o a . _a .= "4"4 .s nw w.4m"A e o 4 5 ~ = ~ ~ e s u lE .t w E-6 g .s a. =a a .N. 3 c. .e., e. n. oa a ~.,,

r

..I 2 2 2. 22

== 3 ~.l.. v . ~ A st J4 ~ a n u { w O o a i = b b 8 a 4 i ,b ,6 ,6 ?'. -4 at .% a u S w .. e eo-o .e.. .o 4.2 s v g g oo o s l 8 l 3 b b b b b *. b b b = A A A 2 A . *.,,, A O. 3'* N m. eg, m. en. b.%.a. m.e m. p% % g ,%3N M. M. M g g g e. 4 L "E .O. C.O. O. O. .O. O O. O .O. 3 O .O. .O. .O. .O .O .O. .O. O. .O. e Is .. as.. . as e F.F. e. e. h. b b t had di G. e. g g g g g g 5,, c .E } ~ e E e. b y u e.

C es es es es e.

e. tem g.. h. b 4. b t. e. es e. e. 3 3. O u u. O .e .a .a 3 3 3 a. e es e b em b ** = .b a.. m 1 O ed a a.e e e 4 e =. e.s S b a. te. h ts. b te. h t to. ln & g a. 4 e 9 e g . te. - E te. E le. te. GA. __g_. 4 A 2 3 b -E 3 ~ e Ab

e. a Wl A 4.#

w t .b m. =. O O a. es ed g g, s. g te. G. u G. O G. G. t.f E =3 i eE 3 \\ O e. .U E w k 3 -1 G. O G. D B-1 nv.-. e

a j 5 ' A o i 1, 12., 1,. 1.,., ,,s .t % .,. 4, .%.d %..s. . ~. ~.. .- ~.s...N s. m. m. ~ e.., .O a.s... ..G..O. .O D .O. DO O. .O O O. ,. ~. M O .O. .C. .O. O DO e 2 25 2.2 Os 2 . s

=

. O. J. ~........... O. O... e en.* g .fl ed mN .P N W 9. ~.. 8't n l 4 b b b b b 4 5. A A =, E 5, 4 4,, a,. 2 g.,

.e
=.%

D.%. N.N.6 P.% *, % g, W 84 W W G 84. % P W @ 4.e e e e e O e e e e a ed .O. O. O .O. .Oi= O. O. .O. O..O. .O. .C O. .O. O. O. d 4s W M W st 88 EB. H 4 4 4 B. W Et SB g5 W W F. O. O. ift .D. if.t St. N. if.t N. W d eD g aN Tw e.= m s.t =< t

8 4

I, e 6 ,b,b

- ,l.

M ,= e e w f 'd W ** .O. .O. ist .t e e e a e O. O. m .. C W sg O. g E , au O. N. u.s Ob W g =d. d bD .a ,.,s -s 2,, .= , =, 3 y e e s.t M M g M 94. e e e I .O. .G. O. O. O et u . as n G W.ft I

3 e s J

W s... e 1 I, 8 l ,b ,J, b ,b .*in. %.*==J.%.n,g N N % z m e. e e. .m e e.v v. e a I e e - e e 'a O. O. .O. O .O..O. .O. O .O. O. .O.. t as.. as.. m m. .t, i i. 8 .E t = 3 - 6 .3 = =- f

I '.

I. 5 E' r E.;. .&. 2 I.2 4' 5.t e t 4' 5 t' *; I* 2 I * **' 2" 8w e w i.=. .o o .e .a .o. .e..s..o. i 8.

8.. e.e 8 8

o..e.

-*= = -88 .o. 8. 8..e .e. i. t... .. I=. =. = 3% =. 3 = ' * ' "" =. =. i k = 6. =

I.'- =w.

J"t .R" =. = J"t I ..-.= w w w w m. ) t 1 lMa-g 2 3 - ~ 2 2 m e T. l. S -h ese $g $_ ~g .I E b. t t 84" w 5-M" .2 .2 .O .== 44 44 e e .aA G. ft ~ '. ' / N'

g

= i-3 5 =8 = 5 3 b 8" D D B-2 ~ -... --- - -.c

"g u 6 6 t (a (

== .s s s s,s,s .~n w w ~ s ~..: ~. ~ k .a. o

m. -

e o..a o eo .o. e o.o =. 3 s - =. = . =. m - *: = m. ~ ~ . =a a

g; 3'=,4-s 1,

i,, A,. k E. ~. ~. ~. ~.

a..s

, s., .s.s.s.- o a. o o o

e..

oe o e..o.. e .o..o. o.

7.,

=

= =

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T [ \\ s IAst! 8.1 pt(MANICAL AND [LICIRlCAL COMPONENT FAltuRC ftAlt5 FAllupt AAI( W45H 1400 f.t het 1((E10(t Agt,ggggjg1D 1 FAltimE swa twa ( AllG0fti COMPON(RT MDf Electrical Cable Olstributtoa rower: Open circuit 9.1 a 10'y/hr 9.8 a 10'#/hr (Cont'd) Copper Short to ground 1.7 a 10.gshr 1.1 a 10 6the 4 Short to power 8.0 a 10/hr 1.0 a 10 /hr i Alweina l Open circuit i I.1 a 10/hr 1.3 a 10/br 3.9 a 10'8/hr 3.9 a lo/hr Short to ground Short to power 3.5 a lo-6/hr 1.5 a 10/hr Control Non circuit 9.1 a 10*I/hr 9.8 a 10'#/hr Copper Short to ground 2.4 a 10/hr 2.4 s 10 Ar ~8 to Short to power l.0 a 10'0/hr l.0 m 10'"/hr eus Terutnal Boards Open I I a 10*I/hr -6 3.3 a 10/hr 3.3. s IO /hr Short I a 10-e/hr 1.4 a 30/hr 1.4 a 10/hr IIstroentation he say Cell falls to I a 10/d .4 a 10-6/hr I a 10 /d and Controls Call falls to open 3 a 10 7/hr .08 a 10-6/hr 3 s 10-F/hr Temperature Tails to operate 1.4 s 10 /hr 1.4 s 10-6/hr Senslag Device (31 Degraded operation 6.6 a 10*I/hr 6.6 s IO'I/hr ~ Temperature Element Fall to operate 1.8 s 10-6/hr I.s a 10-6/hr Degraded operation l.2 a 10'"/hr 1.2 m O'0/hr Temperature falls to operate 3.4 a 10'#/hr 3.8 s 10'#Ihr Transmitter 4 Degraded operation 3.6 s 80'I/hr 3.6 s 10 /hr [3] Sensing Device includes switch, monitor, sensor, and transmitter i t' i 9'

/ I l IAst! 8.1 wruAk tf al Aan filflgfral rrespohfht FAttnar saff t l FAltunt AAlt WA518 1400 Gl NaC 11((.500 ALEffi[.8MD FAltuRE swa pWe CAffGORT COMP 0NENT Moor 'lastrumentation Tesverature ' Failed to operate 2.3 a 10/hr I.F s 10*#/d 8.2 a 10*I/d I" I M ' and Control 5purious operation 1.4 a 10'I/d B.4 a 10'#/d (font'd) Degraded operation 4.5 a 10*I/d 4.8 s 10*#/d ,.33 a 10/hr Jil.]./hr. Falls closed Pressure Sensing Falls to operate F.l a 10*Ithe 6 s 10'#/hr 6 s If thr j Degraded operation 8.3 a 10'8/hr 3.7 a 10'*/hr' 3.7 s 80/hr Pressure Elonent Falls to operate -5 I.I s 10'0/hr 3 I.I 10 /hr Pressure Transelt. Falls to operate 1.2 m 10*I/hr 9.2 a lo'I/hr ter Degraded operation 6.3 s 10*y/hr 6.3 a 10*I/hr Pressure Switch Falls to caerate' I a 10 /d 2.0 s 10'I/d 2.0 a 10'#/d - tzs Spurlous operation 4,3 a 10'*/d 4.s a io.e, j Degradedoperatlh 5.7 a 10-8/d __1.Z_a_10-ejd flew Sensing Device Fall to operate 5.9 s 10*#/hr 4.8 s 10' /hr 4.8 a 10/hr 4 Degraded operation 2.9 s 10 thr 2.5 a 10-6/hr 2.5 a 10 6/hr Flas tienent Fall'to operate 3,3 a 10-I/hr 3.1 a 10-I/hr Degraded operation 1.0 s 80' /hr 3,4 a 10*Ifhr flow Controller, Fall to operate 4.2 s 10-6/hr 4.2 s 10/hr flow Transmitters fal? to operate. 1.4 s 10-6fn, g,,, gg 6/hr Degraded operation 1.4 s 80ar 1,4 a 80/br Flow Switches Fall to operate 4.2 s 10~8/hr 1.3 s 10 /d 1.3 a 10'8/d 5purious operation 1.2 a 10/d 1.2 s 10/d Degraded operetton 1.7 s 10'0 Lielt 5=ltch Fall to operate 3.4 a lo'*/d -/d - - - -1.7 a 10'8/d _______..._. _ 4 3.4 s 10/d Itenmal Switch Fall to transfer I a 10'I/d i I a lo-5, j t w I 9

[X / e TAatt 8.1 M(HANICAL ANO (t[CIRl(Al, C0MPONiki FAltuRE'8ATIS fAltUNE RAtt WA5N.1400 CE haC 1(([-190 _EL2tlL40[0_. FAltimE SWil FWR CAllCDRV COPW'ON(ti WDr 11st wentation tevel Sensing Fall to operate 3.1 a 10thr 2.0 a 10 6/hr 2.6 a 10'I/hr 'I 'I* jg", '4 Degraded operation 5.7 a 10-6/hr 5.2 m to-6/hr 5.2 s 10 /hr 4 level Element 3,3 m 10 6/hr 3.9 s 10 the tevel fressaltte'r Fall to operate I l.4 a 50-6/hr 1.4 s 10/hr f Degraded operation I.1 s 80'"/hr 1.1 a 10-6/hr level Switch Fall to operate 3 a 10*I/d 3 a 10*i/d spurlous operations 3.4 s 10'8/4 3.4 a 10'8/d De' graded operations 4.5 a 10'I/d 4.5 s 10'#/d level Controller Fall to operate I a 10/hr I s 10'8/hr Y 4 5purious operations j i a 10 /hr I a 10'0/br f 2 a 10/hr 2 a 10-6/hr Degraded operation E/5 Converter Fall to operate 4.2 s 10'0/hr g,3 m go-6/hr Square Root Con-Fall to operate 4.3 s 10' /hr urter 4.2 a 10' /hr peer Supply f all to operate 4.2 a 10' /hr 2.0 a 10 6/hr 2.8 s 10/hr Degraded operatloa 1.9 a 10-6/hr I.9 a 10/hr Selld State: to= power Falls to Function I a 10'6thr I a 10-6/hr Falls shorted I a 10'#/hr I a 10/hr High power Falls to Function 3 s 10-6/hr 3 a 10'8/hr I s 10',0 Falls shorted /hr I a 10-6/hr forgse Switch Falls to operate I a 10**/d I a 10**/d Switch Contacts Norwelly open I a 10'#/hr sultches fall I a 10.y/hr to close Normally closed 3 s 10'8/hr 3 a 10'8/hr sultcles fall to clase Short across con. I a 10/hr i s 10 /hr 4 tar r a 9 t V

= o.c FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION J DESCRIPTION UNAVAILABILIT'Y FAILURE RATE IDEh ER S.G.A. Relief Valve Calibration ,QVBISGAROH Valve 6.0 x 10- 7 1.2 x 10-6/hr i S.G.D. Relief Valve Calibration QVB1SGDROH Shift 6.0 x 10- 7 1.2 x 10-6/hr ~ S.G.C. Relief Valve Calibration QVBISGCROH Shift 6.0 x 10~7 1.2 x 10~9/hr S.G.B. Relief Valve Calibration QVBISGRROH Shift 6.0 x 10- 7 1.2 x 10-6/hr S.G.A Relief Valve Control Wiring QVSISGARMM Reversed 4.2 x 10 8.4 x 10-8/hr S.G.D Relief Valve Control Wiring QIX1SGDRMM Reversed 4.2 x 10-8 8.4 x 10-8/hr l S.G.C Relief Valve Control W) ring 1 QIX1SGCRMM Reversed 4.2 x 10 8.4 x 10-8/hr i -8 I S.G.B Relief Valve Control QIX1SGBRMM Wiring Reversed 4.2 x 10 8.4 x 10-8/hr -8 e. s.- Turbine Driven Pump P-37A rails to Run 5.7 x 10-3 5.7 x 10-3/d MPBIT37AMG Motor Driven Pump P-37B MPB1037BMG Fails to Run 3.4 x 10-3 3.4 x 10-3/d Isolation Valve V-30 In Feedwater MVD1V30AMJ Supply Line to S.G.A Ruptures 5 x 10-8 1 x 10-7/hr Check. Valve V-29 In Feedwater MVA1V29ARJ Supply Line To S.G.A Ruptures ,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 MVD14214MJ Aux. Feed. Supply Line A Ruptures. 5.x.10-8 'l x 10" /hr Stop Check Valve V-94 In Aux. l MVA1V94DMJ Feed. Supply Line D Rupture 5 x 10~9 1 x 10-8/hr Flow Control Valve FV-4244 In l MVD14244MJ Aax. Feed. Supply Line D Ruptures 5 x 10-8 1 x 10-7/hr Stop Check Valve V-88 In Aux. MVAIV88CMJ Feed. Supply Line C Rupture 5 x 10-9 1 x 10-8/hr B-8 ..., _.. ~,,,

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILAB'ILITY FAILURE. RATE IDEh ER Flow Control Valve FV-4234 In Aux. l MVD14234MJ Feed Supply Line C Ruptures 5 x 10-8 1 x 10-7/hr Stop Check Valve.V-82 In Aux. MVA1V82BMJ Feed. Supply Line B Ruptures 5 x 10 1 x 10-8/hr i Flow Control' Valve FY-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 10-0 1 x 10-7/hr Check Valve V-55 In Feedwater MVA1Y56DMJ Supply Line D Ruptures 5 x 10-I 1 x 10 /hr Isolation Valve V-48 In Feedwater MVD1Vt.8CMJ Supply Line C Ruptures 5 x 10-8 1 x 10-7/hr Check Valve V-47 In Feedwate'r MVA1V47CMJ Supply Line C Ruptures 5 x 10-9 1 x 10-8/hr Isolation Valve V-39 In Feedwater MVD1V39EMJ Supply Line B Ruptures 5 x 10-8 1 x 10-7/hr Check Valve V-38 In Feedwater MVAlB38BMJ 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 10-8/hr Mr 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 Irr Feedwater -8 MVM1V152MJ Recire. Line A Ruptures .1 x 10 2 x 10-8/hr Gear Driv'en Valve V-156 In Start ~ MVDIV156MJ 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. Fee d. MVM1V87DMJ Supply Line D Ruptures 1 x 10-8 2 x 10-8/hr Manual Valve V-153 In Feedwater MVM1V153MJ Recire. Line D Ruptures 1 x 10 2 x 10-8/hr .i '1 ]( l L-Flow Control Valve FCV-540 In MVD1V540MJ Feed. Supply Line D Ruptures 5 x 10-8 1 x 10-7/hr ,i B-9 --....e,-,,.,,, -..-,.n,

TABLE B.2 (Continued) FAULT ZDENTZFZERS FOR THE S2ASR00K EMERGENCY FEED STAT 30N DESCR E ON MAMM FAMRE WE ID ER Manual Valve V-93 In Aux. Feed. 1 x 10-8 2 x 10-8/hr MVM1V93CMJ Supply Line C Ruptures Manual Valve V-154 In Feedwater 1 x 10-0 2 x 10-8/hr MVM1V154MJ Recire. Line C Ruptures Flow Control Valve FCV-530 In l MDY1V530MJ Feed. SuppTf Line C Ruptures 5 x 10-8 1 x 10-7/hr i Manual Valve V-81 In Aux. Feed. i,. 10-8 2 x 10-8/hr MYM1VSISMJ Line B Ruptures Gear Driven Valve V-71 In P-37B 5 x 10-9 1 x 10-8/hr MVD1V71SMJ Dischar9e Line Rupture Manual Valve V-155 In Feedwater 1 x 10-8 2 x 10-8/hr MVM1V155MJ Recirc. Line B Ruptures Ficw Control Valve FCV-520 In 5 x 10-8 1 x 10-7/hr l MVD1V520MJ Feed. Supply Line B Ruptures Chert Valve V-64 In P-37A Dis-9 5 x 10-9 1 x 10 /hr MVA1V64AMJ charse Line Ruptures Check Valve V-70 In P-388: Dis- -9 1 x 10-8/hr MVA1V70BMJ charge Line Ruptures 5 x 10 Valve V-129 In P-37A Turbine -9 QVDIV129MJ Steam Inlet Line Ruptures 5 x 10 1 x 10-8/hr Manual Valve V-95 In P-37A Tur- -8 8 QVM1V95AMJ bine Steam Inlet Line Ruptures 1 x 10 2 x 10 /hr Check Valve V-94 In Steam-Supply . -9 1 x 10-8/hr QVA1V94AMJ Line A to P-37A Ruptures ,5 x 10 Check Valve V-96 In Steam Supply QVAlys5BMJ Line A to P-37A Ruptures 5 x 10-9 1 x 10-8/hr Steam Supply Line A Control Valve 5 x.10-8 1 x 10-7/hr QVX1V127MJ V-127 Ruptures Steam Supply Line A Manual Bypas: 1 x 10-8 2 x 10-8/hr QVM1V171MJ Valve V-171 Ruptures Steam Supply Line B Control 5 x 10-8 1 x 10-7/hr QVX1Y128MJ Valve V-128 Ruptures k' Steam Supply Line B Manual By-1 x 10-8 2 x 10-0/hr ' ' ~ OMV1V172MJ pass Valve V-172 Ruptures \\ l B-10 -e ---ev-w,,

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FAULT ZDENTIFlERS FOR THE SEABROOK EMERGENCY FEED STATZON DESCRIPTION UNAVAILABILITY FAILURE RATE IDE ER i Manual Valve V-155 In P-37A FVM1V155MJ Suction Line Ruptures 1 x 10-8 2 x 10-8/hr Manual Valve V-154 In P-37A F'M1V154MJ Suction Line Ruptures 1 x 10-8 2 x 10-8/hr V 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 FVM1V155MJ Suction Line Ruptures 1 x 10'O 2 x 10-8/hr Isolation Valve V-30 In Feedwater 4 MVD1V30AMS Supoly Line A Fails To Close 9 x 10 9 x 10 4/d Flow Control Valve FV-4244 In Aux, MVD14244MB Feed Supply Line D Fails To Close 2 x 10-3 2 x 10-3/d Flow Control Valve FV-4234 In Aux, 4 MVD14234M3 Feed. Supply Line D Fails To Close 9 x 10 9 x 10_4/d Flow Control Valve FV-4224 In Aux, fy MVD14224MB Feed. Supply Line D Fails To Close-9 x 10~4 9 x 10-4/d Isolation Valve V-57 In Feedwater MVDIV57DMS Supply Line D Fails To Close 9 x 10 9 x 10 /d Flow Control Valve FV-4214 In Aux, MVD14214MB Feed. Supply Line A Fails To Clost 9 x 10'4 2 x 10-3/d Isolation Valve V-48 In Feedwater MVDIV48CMB Supply Line C Fails To Close 9 x 10-4 9 x 10-4g Isolation Valve V-39 In Feedwater MVDIV39BMS Supply Line C Fails To Close 9 x 10 9 x 10-4/d 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 MVD1V540MB Feed. Supply Line D Fails 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 MVD1V520MB Feed. Supply Line B Fails To Close 9 x 10-4 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 Ii d i -.. - ~. ~ ~..~- ~ - -

E FAULT ZDENT!F1ERS FOR THE SEABROOK EMERGENCY FEED STATZON g;p!gg DESCR M ION U W AI N m FA M RATE Flow Control Valve FV-4244 In Aux. Feed. Supply Line D Fails T 6.0 x 10-8 1.2 x 10-7/hr MVD14244MC Ram,n cle==d c'y Control Valve FV-4234 In Aux. 1.2 x 10-7/hr Feed. Supply Line C Fails T 6.0 x 10 MVD14234MC R msin Closed Flcw Control Valve FV-4224 In Aux. Feed. Supply _Line B Fails' T 6.0 x 10-8 1.2 x 10-7/hr MVD14224MC Remin Closed Isolation Valve V-57 In Feecwater Supply Line' D Fails To Remain 2.3 x 10-7 4.5 x 10-7/hr MVD1V57DMC Closed Flow Controi 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 Feecwater Supply Line C Fails To Remain 2.3 x 10-7 4.5 x 10-7/hr MVD1V48 CMC Closed isciation valve V-2 in reecwater Supply Line B Fails To Remain 2.3 x 10-7 4 5 x 10-7/hr MVD1V39BMC Closed riow controi vaive rLv-lou in Feed. Supply Line A Fails T 8.5 x 10-8 1.7 x 10-7/hr MVD1V510MC (- Remain Closed C Flow. Control Valve FCV-540 In MVD1V540MC ed S 1 Line D Fails T 8.5 x 10-8 1.7 x 10-7/hr Flow Control Valve FCV-530 In Feed. Supply Line C Fails T 8.5 x 10-8 1.7 x 10-7/hr MVD1Y530MC Remain Closed Flew Centrol Valve FCV-520 In Feed. Supply Line B Fails T 8.5 x 10-8 1.7 x 10-7/hr MVD1V520MC Re:.aln Closed Flow Control Valve FV-42dTails To Remain Open 8 x 10-8 1.6 x 10-7/hr MVD14214 MD F1ow Control Valve FV-4244 Fails MVD14244MD To Remain Open 8 x 10-8 1.6x16-7/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 -7 MVD14224MD To Remain Open 8 x 10-8 1.6 x 10 /hr ~~ Isolation Valve V-65 Fails To q/d MVD1V65AMD Remain Open (Plugged) 1 x 10,,4 1 x 10 Steam Supply Inlet Valve V-129 QVD1V129MD Fails To Remain Open 2.3 x 10-7 4.5 x 10-7/hr B-12

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEASROOK EMERGENCY FEED STATION D SCRIPTION UNAVAILABILITY FAILURE RATE IDE ER i Isolation Valve V-125 Fails To 4 MVD1V125MD Remain Open (Plugged) 1 x 10 1 x 10 /d Isolation Valve V-126 Fails To I MVDIV126MD Remain Open (Plugged) 1 x 10-6 1 x 10 /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 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 4/d Steam Supply Line B Flow Valve I QVXIV12SKA V-128 Fails To Open 9 x 10 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 2.3 x'10~7 4.5 x 10-7/hr QVXIV128MD Remain Open i Stop Check Valve V-76 Fails To MVA1V76AMA Open 2 x 10~4 2 x 10 /d 4 Stop Check Valve V-94 Fails To MVA1V94DMA Open 2 x 10_4 2 x 10 /d 4 Stop Check Valve V-88 Fails To MVAIV88CMA Open 2 x 10-4 2 x 10 /d 4 Stop Check Valve V-82 Fails To MVA1V82BMA Open 2 x 10~4 2 x 10-4/d t~ MVA1V64AMA Check Valve V-64 Fails To Open 2 x 10-4 2 x 10 /d 4 Steam Supply Line A Check Valve QVA1V94AMA V-94 Fails To Open 2 x 10'4 2 x 10 4/d i I. Steam Supply Line B Check Valve QVAIV96BMA V-95 Fails To Open 2 x 10~4 2 x 10,4/d I .[i MVA1V70BMA Check Valve V-70 Fails To Open 2 x 10'4 2 x 10 /c Valve V-75 In Aux. Feed Supply MVD1V75AMD Line A Plugged 1 x 10,4 -1 x 10 /d 4 ~ B-13

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE I D ER Valve V-87 In Aux. Feed, Supply 4 MVD1V87DMD Line D Plugged 1 x 10 1 x 10 4./d Valve V-93 In Aux. Feed Supply MVD1V93CMD Line C Plugged 1 x 10_4 1 x 10_4/d Valve V-81 In Aux. Feed Supply MVD1V81BMD Line B Plugged -1 x 10_4 1 x 10 /d 4 Manual Valve V-95 In Turbine QVMIV95AMD Steam Supply Line Plugged. 1 x 10_4 1 x 10 4/d Manual Valve V-155 In P-37A FVM1V1:55MD Suction Line Plugged 1 x 10-4 ' 1 x 10-4/d Manual Valve V-154 In P-37A FVM1V154MD Suction Line Plugged 1 x 10-4 1 x 10-4/d Manual Valve V-159 In P-37B FVM1V159MD Suction Line Plugged 1 x 10 4 1 x 10_4/d Manual Valve V-158 In P-37B -4 FVM1V15SMD Suction Line Plugged l 'x,10 ,1 x.,10-4/d Feedwater Supply Line Ruptures M0J1506BMJ Between V-20 and V-30 5 x 10-11 1 x 10-10/hr Feedwater Supply Line Ruptures MDJ1506AMJ Between V-30 and S.G.A 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line A Ruptures M0E1614AMJ Between V-76 and Main _ Supply Line 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line A Ruptures M0E1514BMJ Between FV-4214 and V-76 5 x 10-11" 1 x 10-10/hr Feedwater Supply Line D Ruptures M0J1609AMJ Between V-57 and S.G.D 5 x 10-11 1 x 10-10/hr Aux. reeo suppiy une v Rupcures Between V-94 and Main Feed. 5 x 10-11 1 x 10-10/hr M9E1617AMJ Supply Line D Feedwater Supply Line C Ruptures MSJ160BAMJ Between V-48 and S.G.C 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line C Ruptures M0E1616AMJ Between V-88 and Main Feed. 5 x 10-11 1 x 10-10/hr Sunelv Lin C u Aux. Feed Supply Line C Ruptures MBE1616BMJ Between FV-4234 and V-88 5 x 10-11 1 x 10-ID/hr B-14 \\

<a- -s a.a u. -....a s a w a-FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION i ~ 10 ER DESCRIPTION UNAVAILABIOTY FAILURE RATE C Feedwater Supply Line 8 Ruptures ' I M9J1607AMJ Between V-39 and S.G.B 5 x 10~11 1 x 10-10/hr Aux. Feed. Supply Line B Ruptures M9E1615AMJ b ", d2 5 x 10-11 1 x 10-10/hr n Aux. Feed. Supply Line B Ruptures M9E1615BMJ Between V-4224 and V-82 5 x 10-11 1 x 10-10/hr Feedwater Supply Line D Ruptures I M9E1609BMJ Between V-56 and V-57 ~ 5 x 10-11 1 x 10-10/hr Feedwater Supply Line C Ruptures M9J1608BMJ Between V-47 and V-48 5 x 10-11 -10 1 x 10 /hr Feedwater Supply Line B Ruptures M0J1607BMJ Between V-38 and V-39 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line A Ruptures M0E1614CMJ Between V-75 and FV-4214 ~ l 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line A Ruptures M0E1614CMJ Between V-75 and Aux. Feed Supply 5 x 10-11 1 x 10-10/hr Header Aux. Feed Supply Header Ruptures M0G1613AMJ Between V-125 and Reducer 5 x 10-11 1 x 10-10/hr Aux. Peea rump e-4/A uiscnarge M0F161 AMJ piping Ruptures Between V-65 5 x 10-11 1 x 10-10/hr and Supply Header reecwater Macirc. une a xuptures M0F1606AMJ Between Aux. Feed. Supply A and 5 x 10~11 1 x 10-10/hr V-152 startup Feed Pump Disenarse,_une-M0F1632AMJ Ruptures Between Y-156 and Aux. Feed. Supply Header 5 x 10-11 1 x 10-10/hr l Feedwater Recire. Line A Ruptures M0E1606BMJ Between V-152 and Main Feed Sup-5 x 10-11 1 x 10-10/hr n, 1 v f4a a Feedwater Supply Line A Ruptures M0J1606CMJ Between FCV-510 and V-29 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line D Ruptures M0E1617CMJ Between V-87 and FV-4224 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line D Ruptures M0E1617DMJ Between V-87 and Aux. Feed Sucolv Header 5 x 10~11 1 x 10-10/hr (( ' Aux. Feed. Supply Header Ruptures M0G1613BMJ Between V-87 and V-126 5 x 10-11 l 1 x 10-10/hr B-15

FAULT IDENTIFZERS FOR THE SEABROOK EMERGENCY FEED STATZON i IDEN ER A FAILURE RATE (' Feedwater Supply Lir.e B Ruptures M0J1607AMJ Between V-39 and S.G.B 5 x 10-11 1 x 10-10/nr Aux. Feed. Supply Line B Ruptures Be en 2 and Main Feed. M0E1615AMJ 5 x 10-11 1 x 10-10/hr q Aux. Feed. Supply Line B Ruptures M0E1615BMJ Between V-4224 and V-82 5 x 10-11 1 x 10-10/hr-Feedwater Supply Line D Ruptures M0E1609BMJ Between V-56 and V-57 5 x 10-l. 1 x 10-10/hr Feeddater Supply Line C Ruptures M0J160SBMJ Between V-47 and V-48 5 x 10-11 1 x 10 /hr -10 Feedsater Supply Line B Ruptures M0J1607BMJ Between V-38 and V-39 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line A Ruptures MDE1614CMJ Between V-75 and FY-4214 5 x 10-11 1 i 10-10/hr Aux. reec. Supply Line A Ruptures M0E1614CMJ Between V-75 and Aux. Feed Supply 5 x 10-11 l x 10-10/hr Header Aux. Feed Supply Header Ruptures M0G1613AMJ Between V-125 and Reducer 5 x 10-11 1 x 10-10/hr nux. reec eump e-4/A uiscnarge M0F1610AMJ Piping Ruptures Between V-65 ' 5 x 10-11 1 x 10-10/hr , and Supply Header reecwa ter necirc. Line n nuptures M0F1606AMJ Between Aux. Feed. Supply A and 5 x 10-11 1 x 10-10/hr V-1G2 startup reea Pump utscnarge.Qne M0F1632AMJ Ruptures Between V-156 and Aux. Feed. Supply Header 5 x 10-11 1 x 10-10/hr Feedwater Recire. Line A Ruptures M0E1606BMJ Between V-152 and Main Feed Sup-5 x 10-11 1 x 10-10/hr n, t y i%a Feedwater Supply Line A Ruptures M0J1606CMJ Between FCV-510 and V-29 5 x 10-11 1 x 10-10/hr Aux. Feed Supply line D Ruptures M0E1617CMJ Between V-87 and FV-4224 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line O Ruptures M0E1617DMJ Betwet.n V-87 and Aux. Feed Supolv Header 5 x 10-11 1 x 10-10/hr j (' Aux. Feed. Supply Header Ruptures ( M0G1613BMJ Between V-87 and V-126 5 x 10-11 1 x 10-10/hr B-15

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION ID ER DESCRIPTION UNAVAILABILITY FAILURE RATE Feecwater Recire. Line D Ruptures .g M9E1609AMJ Between Aux. Feed. Supply Line D 5 x 10-11 1 x 10-10/hr and V-153 Feedwater Recire. Line D Ruptures i ME1609BMJ Between V-153 and Main Feedwater Suooly Line D 5 x 10-11 1 x 10-10/hr Feedwater Supply Line.D Ruptures M0J1609CMJ Between FCV-540 and V-56 5 x 10-11 1 x 10-10/hr Aux. Feed. Supply Line C Ruptures M0E1516CMJ Between V-93 and FV-5234 5 x 10-11 1 x 10-10/hr Aux. reec suppiy Line L Kuptures M0E1616DMJ Between V-93 and Aux. Feed. Supply Header 5 x l'-Il 1 x 10-10/hr 0 Aux. Feed Supply Header Ruptures M0G1613CMJ Between V-126 and V-127 5 x 10-11 1 x 10-10/hr reecne;er necirc. Line c nuptures MDE1608AMJ Between Aux. Feed. Supply Line 5 x 10-11 1 x 10-10/hr C and V-154 Feedwater Recir. Line C Ruptures . NDE1608BMJ Between V-154 and. Main Feedwater 5 x 10-11 1 x 10-10/hr e.. -1v t w c Feedwater Supply Line C Ruptures M0J1608CMJ Between FCV-530 and V-47 5 x 10-11 1 x 10-10/hr Aux. Feed Supply Line B Ruptures ~ M2E1615CMJ Between V-81 and FV-4224 5 x 10-33 1 x 10-10/hr Aux. Feed Supply Line B Ruptures M0E1615DMJ Between V-81 and Aux. Feed.. Supply 5 x 10711 1 x 10-10/hr Aux. Feed Supply Header Ruptures M0G1613DMJ Between V-127 and Reducer 5 x 10-11 1 x 10-10/hr Feed. Pump P-37B Discharge Line M0F1612AMJ Ruptures Between V-71 & Reducer 5 x 10-11 1 x 10-10/hr. reecwater xecirc. Line 5 Muptures M9E1607AMJ Between %uxrFeed Supply Line B 5 x 10-II 1 x 10-10/hr and V-155 Feedwater Recire. Line B Ruptures. M0E1607BMJ Between V-155 and Main Feedwater 5 x 10-II 1 x 10-10/hr

e.,q.,w i < m. =

Feedwater Supply Line B Ruptures M0J1607CMJ Between FCV-520 and V-38 5 x 10-11 1 x 10-10/hr i ~ Feed. Pump P-37A Discharge Line M0F1610BMJ Ruptures Between V-64 and V-65 5 x' 10-11 1 x 10-10/hr B-16 _,, _ _ ~.... _ -,.., _ _ _. - - - -... _ _,.,.,. ~,. _,,..,. ...--m.,.

II FAULT XDENTIFIERS FOR THE SEASROOK EMERGENCY FEED STATZON DESCRIPTION UNAVAILABILITY FAILURE RATE IDEh ER l Feed Pump P-37B Discharge Line M9F1612SMJ Ruptures Between V-70 and V-71 5 x 10-11 1 x 10-10/hr 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 Recire. Line Rup-tures Between V-67 and Pump 5 x 10-10 1 x 10-9/hr M001610AMJ Discharce Line Turbine Steam Supply Line Ruptures Q0E1:49AMJ Between V-95 and V-129 5 x 10-17 1 x 10-10/hr Turbine Steam Supply Line Rupture: Q0E14493MJ Between V-129 and Turbine 5 x 10-11 1 x 10-10/hr Turbine Steam Supply Line Rupture! QDF1449AMJ Between Tee and V95 5 x 10-11 1 x 10-10/hr stean suppiy une o nupwres Berneen Reducer and Turbine 5 x 10-11 1.x 10-10/hr QDF1109AMJ Inlet Line Tee 5 team Suppiy Line A Ruptures QQF1009AMJ 98Ud**". Reducer and Turbir.e 5 x 10-11 1 x 10-10/hr 2 Inlet Line Tee i steam supply Line o kuptures Q0E1143AMJ Between V-128 and V-96 5 x 10-11 1 x 10-10/hr ~ i Steam Supply Line A Ruptures Q0E1043AMJ Between V-127 and V-94 5 x 10-11 1 x 10-10/hr Steam Supply Line A Ruptures Q0E1043BMJ Between V-94 and Reducer 5 x 10-11 1 x 10-10/hr Steam Bypass Exit Line Ruptures QOA1036AMJ Between V-171 and Steam Seppiy. 5 x 10-10 1 x 10-9/hr ,1 Iino a Steam Supply Line A Ruptures 00E1042AMJ Between Reducer and V-127 5 x 10-11 1 x 10-10/hr i ~ Steam Bypass Inlet Line Ruptures Q0A1035AMJ Between Steam Supply Line A 5 x 10-10 1 x 10-9/hr i.' and V-171 Steam Supply Line A Ruptures QQF1008AMJ Between Main Steam Line A and 5 x 10-11 1 x 10-10/hr Reducer Q0KISLIAMJ Main Steam Line A Ruptures 5 x 10-11 1 x 10-10/hr k Steam Supply Line 8 Ruptures 5 x 10-11 1 x 10-10/hr QOF1143SMJ Between V-96 and Reducer i i B-17 l - - ~. e .-,.m .,_._.y_,.,,_,,_-._.m_y.. .._.,m ,.,,.,_,,m

FAULT ZDENTIFZERS FOR THE SEABROOK EMERGENCY FEED STATZON DESCRIPTION } UNAVAILABILITYFAILURE RATE (- ID-I ER I Steam Bypass Exit Line Ruptures -10 -9 Q0A1136AMJ Between V-172 and Steam Supply 5 x 10 1 x 10 /hr Une ? Steam Supply Line B Ruptures Be- -11 Q9E1142AMJ tween Reducer and V-128 5 x 10 1 x 10-10/hr Steam Bypass Inlet Line Ruptures Q0A1135AMJ Between V-172 and Steam Supply 5 x 10-10 1 x 10-9/hr Line B 5 team Suppiy Line 6 Ruptures QQF1108AMJ Between Main Steam Line B and 5 x 10-11 1 x 10-10/hr Reducer 00K1SL1BMJ Main Steam Line B Ruptures 5 x 10-I1 1 x 10-10/hr Feed Pump P-37A Suction Line FDG1081AMJ Ruptures Between V-15 and Pu=o 5 x 10-11 1 x 10-10/hr T13 4 Feed Pump P-37A Suction Line F0G1081BMJ Ruptures Between V-154 and V-155 5 x 10-11 l'x 10-10/hr Feed Pump P-37A Suction Line F0G1081CMJ Ruptures Between Condensate Tank 5 x 10-11 1 x 10-10/hr nd V-154 Feed PumpP-378 Discharge Line M0E1612CMJ Ruptures Between P-37B and V-70 5 x 10-II 1 x 10-10/hr Feed Pump P-37B Recirc. Line MOD 1612AMJ Ruptures Between V-73 and Pump 5 x 10-11 1 x 10-9/hr Discharce Line Feed Pump P-37B Suction Line F0G1082AMJ Ruptures Between V-159 and 5 x 10-11 1 x 10-10/hr Pumo Inlet Feed Pump P-378 Suction Line F0G1082BMJ Ruptures Between V-158 and+159 5 x 10-11 1 x 10-10/hr Feed Pump P-37B Suction Line F0G1082CMJ Ruptures Between Condensate Tank 5 x 10-11 1 x 10-10/hr m ad V-1 cA Turbine Driven Feed Pump-37A MPB1T37AME Fails To Start 8.4 x 10-3 8.4 x 10-3/d MotorTriven -Feed Pump P-37B MPBID37BME Fails To Start 2.4 x 10-3 2.4 x 10-3/d Turbine Driven Feed P-37A Out l MPBIT37A00 of Service 4.2 x 10-4 Motor Driven Feed Pump P-378 Out MPB1D37800 Of Service 9.4 x 10 4 B-18 l

TABLE S.2 (Continued) + l l FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION l 1 AVAINILM FAILWE WE ( IDE.N ER Flew Control Valve FV-4214 Out MVD1421400 of Service 8.5 x 10 MhD1V75A09 Valve Out of Service 8.5 x 10 Flow Control Valve FV-4244 Out MVD1422409 of Service - 8.5 x 10 'MVD1V87000 Valve V-87 Out of Service 8.5 x 10-4 Flcw Control Valve FY-4234 Out MVD1423400 of Service 8.5 x 10-4 MVD1V93C00 Valve V-93 Out of Service 8.5 x 10~4 Flcw Control Valve FV-4224 Out-MVD1422400 of Service 8.5 x 10-4 MVD1V81300 Valve V-81 Out of Service 8.5 x 10d MVD1V55A00 Isolation Valve V-65 Out of Service 2.0 x 10-3 Steam Supply Valve V-129 Out of QVD1V12900 Service 1.0 x 10_4 GMVIV95A00 Manual Valve V-95 Out of Service 0.0 Steam Supply Valve V-127 0 @.of QVXIV12700 Service

8. 7 x 10-4 Steam Supply Valve V-128 Out of QVX1Y12200 Service 8.7 x 10-4 P-37A Suction Valve V-155 Out of FVM1V15500 Service O'. 0 -

P-37A Suction Valve V-154 Out of FVM1V15400 Service 0.0 Isolation Valve V-125 Out of MVD1V12500 Service 0.0 { Isolation Valve V-126 Out of MVD1V12500 Service 0.0 B-19 ^ ^ ~ ~ ~ ~

TABLE 3.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION 1 1 DESCRIPTION UNAVAILABILITY FAILURE RATE ID ER Isolation Valve V-127 Out of MVD1V12799 Service. 0.0 f 5 Isolation Valve V-71 Out of 2.0 x 10-3 MVD1V71B09 Service P-37B Suction Valve V-159 Out FVM1V15999 af Service ~ 0.0 P-37B Suction Valve V-158 Out FVM1Y15800 of Service 0.0 Operator Fails to Open Steam 5 x 10-3 5 x 10-3/d QVXIV1270A Supply Valve V-127 Operator Fails to Open Steam 5 x 10-3 5 x 10-3/d QVXIV1280A Supply Valve V-128 Operator Fails to Close Feedwitter MVD1V30A03 Supply Line A Isolation Valve V-30 5 x 10-3 5 x 10 /d Operator Fails to Close Flow 5 x 10-3 5 x 10-3/d ( MVD142240B-Centrol Valve FV-4224 Operator Fails to Close Flow 5 x 10~3 5 x 10-3/d MVD142240B Control Valve FV-4234 Operator Fails to Close Flow MVD142440B Control Valve FY-4244 5 x 10~3 5 x 10-3/d Operator Fails to Close Feedwater -3 MVD1V57008 Supply Line D isolation Valve V-57 5 x 10-3 5 x 10 /d Operator Fails to Close-F4cw MVD142140B Control Valve FV-4214 5 x 10-3 5 x 10-3/d Operator Fails to Close Feedwater MDV1V48 COB Supply Line C Isolation Valve V-48 5 x 10-3 5 x 10-3/d Operator Fails to Close Feecwater MVD1839808 Supply _Line B Isolation Yalve V-39 5_x 10-3 5 x 10-3/d \\ Operator Fails to Close Feedwater MV01V5Z00B Flow Control Valve FCV-510 5 x 10-3 5 x 10-3/d Operator Fails to Close Supply MVD1V1250B Header Isolation Valve V-125 .9 .9 Operator Fails to Close Feedwater MV01V54008 Flow Control Valve FCV-540 5 x 10-3 5 x 10-3/d ~ B-20 l .n. +,. -, - -., - - -w

TABLE B.2 (Continusd). i FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION IDE ER ^ ^ Operator Fails to Close Supply MVD1V1269B Header Isolation Valve V-126 .9 .9 i Operator Fails to Close Feedwater MVDIY5309B Flow Control Valve FCV-530 5 x 10-3 5 x 10-3/d Operator Fails to Close Supply ~ MOV1Y12709 Header Isolation Valve V-127 .9 .9 Operator Fails to Close Feedwater MVD1V52003 Flow Control Valve FCV-520 5 x 10-3 5 x 10-3/d Opera +wr Fails to Close P-37B MVD1V65A0B Discharge Isolation Valve V-65 .9 .9 Operator Fails to Close P-37B MVD1V71S08 Discharge Isolation Valve V-71 .9 .9 Operator Fails to Restore Manua'l MVM1V1520C Valve V-152 1 x 10-3 1 x 10-3/d Operator Fails to Restore Manual MVM1V1530C Valve V-153 1 x 10-3 1 x 10-3/d Operator Fails to Restore Manual i MVM1Y1540C Valve V-154 1 x 10-3 2 x 10-3/d 4 Operator Fails to Restore Manual MVM1V1550C Valve V-155 1 x 10-3 1 x 10-3/d i 6ICIAFSAOC Operator Defeats Train A"S" Signal 1 x 10-4 1 x 10-4/d 6ICIAFSB9C Operator Defeats Train B"S" Signal 1*x 10-4 1 x 10 /d Operator Inadvertently Closes MVD1421400 Flow Control Valve FV-4214 1 x 10-4 1 x 10 /d operator inaovertently closes MVD1V75A00 Valve V-75 l'x f0-4 1 x 10 /d 1 Operator Inadvertently Closes MVM1424400 Flow Control Valve FV-4244 1 x 10-4 1 x 10-4/d Operator Inadvertently Closes MVD1Y87D00 Valve V-87 1 x 10-4 1 x 10 /d Operator Inadvertently Closes MVD1423400 Flow Control Valve FV-4234 1 x 10-4 1 x 10 /d B-21 i ,__.__...-.x

ginunueo) ~' iant: s.4 FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCR M ION MAM M FAI M E RATE ( ID ER Operator Inadvertently Closes MVD1V93CDD Valve V-93 1 x 10,4 1 x 10,4/d Operator Inadvertently Closes MVD142240D Flow Control Valve FV-4224 1 x 10 4 1 x 10 /d 4 Operator Inadvertently Closes MVDIV81 BOD Valve V 1 x 10,4 1 x 10 4/d Operato'r Fails to Restore P-37A MVM1V65 ADD Isolation Valve V 1 x 10-4 1 x 10~4/d Operator Fails to Restore Manual QVDIV95 ADD Valve V-95 1 x 10-3 1 x 10-3/d Operator Fails to Rcstore P-37A FMV1V1550D Suction Valve V-155 1 x 10-3 1 x 10-3/d Operator Fails to Restore P-27A FVM1V1540D Suction Valve V-154 1 x 10-3 1 x 10-3/d Operator Fails to Restore pply MVD1V1250D Header Valve V-125 1 x 10~3 1 x 10~3/d Operator Fails to Restore. Supply MVD1V12500 Header Valve V-126 1 x 10~3 1 x 10~3/d i Operator Fails to Restore Supply MVD1Y1270D Header Valve V-127 1 x 10~3 1 x 10-3/d Operator Fails to Restore P-37B 4 MVDIY71800 Discharge Isolation Valve V-71 1 x 10 1 x 10'/d Operator Fails to Restore-+ 37B FVM1V1590D Suction Valve V-159 1 x 10-3 1 x 10-3/d Operator Fails to Restore P-378 PiM1VI.5800 Suction Valve V-158' 1 x 10~# 1 x 10-3/d Operator Fails to Start Motor MPB1D37B0E - Driven Pump P-37B' 1 x 10-3 1 x 10-3/d Operator Turns Off Turbine Driven 4~ MPB1T37A0G Pump P-37A 1 x 10 1 x 10 /d Operator Turns Off Motor Driven MPB1037B0G Pump P-37B 1 x 10~4 1 x 10 /d ( Circuit Breaker to Motor Driven ,L MCA2037BMK Pump P-37B Open 1.5 x 10-6 4.2 x 10-3/hr B-22 t -

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE IDE ER Control Circuit to Steam Supply QRA2V127MK Valve V-127 Open 6.5 x 10,4 - 9.1 x 10-7/hr h Control Circuit to Steam Supply 4 i QRA1V128MK Valve V-128 Open 6.5 x 10 .9.1 x 10-7/hr Train A Control Circuit to Motor MRA1A37BMX Pump P-37B Open 5.9 x 10-3 9.1 x 10-7/hi Train B Control Circuit to Motor MRAIB37BMX Pump P-37B Open 5.9 x 10-3 9.1 x 10~7/hr i Flow Control Valve FCV-510 Flow MCE1V510MK Control Switch Open 1.5 x 10-8 3 x 10-8/hr F1ow Contr01 Valve FCV-540 F1ow MCE1V540MK Control Seitch Open 1.5 x 10-8 3 x 10~8/hr Flow Control Valve FCV-530 Flow MCE1V530MK Control Switch Open 1.5 x 10 3 x 10-8/hr Flow Control Valve FCV-520 Flow ( 1.5 x 10-8 3 x 10-8/hr MCE1V520MK Control Switch Open Ster.m Supply Valve V-127 Switch QCE1V127MK Open 2.2 x 10-5 3 x 10-8/hr P-373 Motor Controller Circuit 4 HCE1037BMK Open 3.3 x 10 9.1 x 10~7/he MCKID37BMX P-37B Motor Starter Circuit Open 1.2 x 10-3 1.2 x 10-3/d l Turbine Driven Feed Pump P-37A MPBIT37ALB Lubrication Failure 0.0 Motor Driven Feed Pump P-37B MPBID37BLB Luabrication Failure 0.0 LOSP Loss of_ Station. Power 7 x 10-6 1.4 x 10-5/hi STA1TX25MJ Condensate Storage Tank Ruptured 5 x 10-11 1 x 10-10/hr 1 6IC1FWIAMN No Train A Feedwater Isolation 5.8 x 10-3 5.8 x 10-3/d o c19e>T No Signal From FE-4224 to Flow MIF14224MN Control Valve FV-4224 2.0 x 10-3 3.1 x 10-7/hr B-23

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEASROOK EMERGENCY FEED STATION ID ER DESCRIPTION UNAVAILABILITY FAILURE RATE No Signal From FE-4234 to Flow MIF14234MN Control Valve FV-4234 2.0 x 10-3 3.1 x 10 /hr i No Signal From FE-4224 to Flow MIF14244MN Control Valve FV-4244 2.0 x 10-3 3.1 x 10-7/hr No Sgal From FE-4214 ~ to Flow 2.0 x 10-3 3.1 x 10-7/hr Cont Valve FV-4214 MIF14214MN No Train 3 Feedwater Isolatier. SIC 1FWSMN 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 Frem 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 x 10-7/hr No Signal From FE-520 to Flow (. MIF1V520MN Control Valve FCV-520 1.5 x 10-7 3.1 x 10-7/hr Q - 3 No Signal From Safety Injection 6ICIAFSAMN Signal Train A 5.8 x 10-3 5.8 x 10-3/d 1 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 8 MVD14214MO Valve FV-4214 1.2 x 10 1.2 x 10-0/d Spurious Signal to Flow Control. MVD14244M9 Valve FV-4244 1.2 x 10-0 1.2 x 10-8/d Spurious Signal to Flow Control i MVD14234MB Valve FV-4234 1.2 x 10-0 1.2 x 10-8/d t Spurious Signal to flow Control MVD14224M9 Valve FJ-4224 1.2 x 10-8 1.2 x 10-8/d REC 118--MJ 125 DC Bus 11B Shorts to Ground 3.5 x 10-8 7 x 10-8/hr REC 1E512MJ 460 V AC Bus E612 Shorts to Ground 3.5 x 10-8 7 x 19-8/hr RECIE51-MJ 480 V AC Bus E51 Shor's to Ground 3.5 x 10-8 7 x 10-8/hr-t B-24

w--

~ w -, +,. - .,w+-- ,,e m,, ,---,-.----,,m. g-

~ ,6 mO&.C O.4 FAULT ZDENT1FfERS FOR THE SEABROOK EMERGENCY FEED STATION - DESCR M ION UNAVAIUB1 W FAI M E RATE IDE ER Diesel Generator DG-1B Circuit 1 x 10-3 1 x 10-3/d RCA1A74-MB Breaker A74 Fails to Close Crossover Circuit Breaker DN6 ~ RCA1DN6-MB Fails to Close 1 x 10-3' 1 x 10-3/d ~RCA2A61--MC Circuit Breaker A61 Open 2.1 x 10-9 4.2 x10-9/hr RCA1A62--MF Circuit Breaker A62 Fails to Close 1 x 10-3 'l x 10-3/d RCA10N4-MC Circuit Breaker DN4 Open 2.1 x 10-9 4.2 x 10-9/hr RCA1DNS-MC Circuit Breaker DNS Open 2.1 x 10-9 4.2 x 10-9/hr RCA1DAl-MC Circuit Breaker DA1 Open 2.1 x 10-9 4.2 x 10-9/hr RCA1A06-MC Circuit Breaker ADS 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 Circuit Breaker A41 Shorts t -8 RCA1A41-MJ 3.5 x 10 7 x 10-8/hr Ground RCA1A42-MB Circuit Breaker A42 Fails 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 Circuit Breaker A61 Shorts to 3,' x 10-0 7 x 10-8/hr RCA1A61-MJ S Ground. Circuit Breaker A62 Shorts t 3.5 x 10-8 7 x 10-8/hr RCA1A62-MJ Ground rc t Breaker DN5 Shorts t 3.5 x 10-8 7 x 19-8/hr RCA1DN5-MJ Circuit Breaker DA1 Shorts t 3.5 x 10-8 7 x 10-8/hr RCA1DAl-MJ Ground B-25

TABLE B.2 (Continued) FAULT 1DENTIFlERS FOR THE SEABROOK EMERGENCY FEED STAT 10N 4 FAULT [ IDENTIFIER DESCRIPTION UNAVAILABILITY FAILURE RATE Circuit Breaker AD6 Shorts t 4 RCA1AD6--MJ 3.5 x 10 7 x 10-8/hr Ground RCA52-MJ Circuit Breaker AS2 Shorts to Ground 3.5 x 10_g 7 x 10,g/hr l ' Circuit Breaker A75 Shorts to 3.5 x 10-8 7 x 10 /hr -8 RCA1A75-MJ Ground RBC118--MC Battery Charger 1B Opens 2.1 x 10-9 '.2 x 10-9/hr 4 RBC118-MJ Battery Charger 1B Shorts t 3.5 x 10-8 7 x 10-8/hr Ground RSA113--MJ Battery IB Shorts to Ground 3.5 x 10-9 7 x'10-8/hr RSA110--MJ Battery ID Shorts to Ground 3.6 x 10-b 7 x 10-8/hr RBA118--MM Battery l8 Undercharged 1.3 x 10-6 3 x 10 /hr -6 hI; ~ RBA11D-- m Battery 1D 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 Transfomer S-2B Shorts 5 x 10-7 1 x 10-6/hr RTR'lX-2BMC Transformer S-2B Opens 5 x 10-7 1 x 10-6/hr RTRIX-5CMJ Transformer X-SC Shorts 5 x 10-7 1 x 10-6/hr RTRIX-3BMC Transfonner X-38 Opens 5 x-10-7 1 x 10-6/hr i RTRIX-3BMJ Transformer X-3B Shorts 5 x 10-7 1 x 10-6/hr RG0118--ME Diesel Generator DG-1B Fails to Start 1.0 x 10-2 1.0 x 10 /d ~2 e{el Generator DG-1B Fails 3.0 x 10-3 6.0 x 10-3/hr RGD118--MG B-26 ~ - - - - - - - - - - - ~ ~ - - - - - - - - - - - - - - -

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILIBILITY FAILURE RATE F ID ER ( Diesel Generator DG-1B Out of RGD118--09 7 x 10 Service Diesel Generator DG-1A Out of 11A--M 7 x 10 Service 4 MHX1TBRAMJ S.G.A Steam Line Rupture 5 x 10-11 1 x 10-10/hr t fS.G.Shhell Rupture 5 x 10-11 1 x 10-10/hr MHX1SHRAMJ MHX1TBRDMJ S.G.D Steam Line Rupture. 5 x 10-11 1 x 10-10/hr j MMX1SHRCMJ 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 I MHXISHRCMJ S.G.C Shell Rupture 5 x 10-11 1 x 10-10/hr ~. MHX1TBRBMJ S.G.B Steam Line Rupture ' 5 x 10-11' 1 x 10-10/hr l MHXISHRBMJ S.G.B Shell Rupture 5 x 10-11 1 x 10 /hr -10 MPB1T37ACL Turbine Oriven Feed Pump P-37A 0.0 Cooling Loss MPB1D37BCL Motor Driven Feed Pump P-97B Cooling Loss .0.0 MTUlTD-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 +n nnon Control 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 QVL1Y127MA Steam Supply Valve V-128 Fails 1.4 x 10-6 1.4 x 10-6/d tn Onon Solenoid Valve on Steam Supply QLVIV12BMA Valve V-128 Fails to Open 1.4 x 10-6 1,4 x 10-6/d B-27

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION (_ IDE. ER A A FA M RE W E Safety Valve V-6 on Steam Line QVB1V06AMC A Opens 5 x 10-6 1 x 10-5/hr 4 Safety Valve V-7 on Steam Line QiSIV07AMC 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 -6 QVB1VOSAMC A Opens 5 x 10 1 x 10-5/hr i Safety Valve V-9 on Steam Line QVB1V09AMC A Opens 5 x 10-6 1 x 10-5/hr l 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 5 x 10-6 1 x 10-5/hr QVB1V51DMC r. D Opens i l k' Safety Valve V-52 on Steam'Line QVB1V52DMC 0 Opens 5 x 10-6 1 x 10-5/hr QVB1V53DMC Safety Valve V-53 on Steam Line 5 x 10-6 1 x 10-5/hr 0 Ocens Safety Valve V-54 on Steam Line QVB1V54DMC D Ocens 5 x 10-6 1 x 10-5/hr Relief Valve on Main St'eimTine { QVBISGDRMC D Opens 5 x 10-6 1 x 10-5/hr Safety Valve V-36 on Steam Line l QVB1V36 CMC C Opens 5 x 10-6 1 x 10-5/hr l l Safety Valve V-37 on Steam Line 1 QVB1V37 CMC C Opens 5 x-10-6 1 x 10-5/hr Safety Valve V-38 on Steam Line QVB1V38 CMC, C Opens 5 x 10-0 1 x 10-5/hr Safety Valve V-39 on Steam Line QVB1V39 CMC C Opens 5 x 10-6 1 x 10-5 r /h h ~ Safety Valve V-40 on Steam Line QVB1V49 CMC C Opens 5 x 10-6 1 x 10-5/hr B-2B

~ TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION DESCRIPTION UNAVAILABILITY FAILURE RATE IDEh ER Relief Valve on Main Steam Line I j QVBISGCRMC C Opens 5 x 10-6 1 x 10-5/hr I i Safety Valve V-22 on Steam Line QVB1V22SMC B Opens 5 x 10-0 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 6 QVB1V25BMC B. Opens 5 x 10 1 x 10-6/hr Safety Valve V-26 on Steam Line QVB1V25BMC B Opens 5 x 10-6 1 x 10-6/hr Relief Valve on Main Steam Line 6 ,1 x 10-5/hr QVBISGBRMC B Opens 5 x 10 Loss of Voltage on S.G.A Relief 9ECISGARMM Valve Control.ler 7 x 1.7-7 1.4 x 10-6/hr Loss of Voltage onS.G.B Relief DECISGBRMM Valve Controller 7 x 10-7 1.4 x 10-6/hr Loss of Voltage on S.G.A Relief 9ECISGCRMM Valve Controller 7 x 10-7 1.4 x 10-6/hr, Loss of Voltage of 5.G.D Reliei 9ECISGDRMM Valve Controller 7 x 10-7 1.4 x 10-6/hr MVDIV8700B Operator Fails to Close V-87 5 x 10-3 5 x 10-3/d I MVDIV93 COB Operator Fails to Close V-93 5 x 10-3 5 x 10-3/d MVDIV81898 Operat5i Fails to Close V-81 5 x 10-3 5 x 10-3/d ~ MVD1V75ABB Operator Fails to Close V-75 5 x 10-3 5 x 10-3/d l l MVD1V87DMB Valve V-87 Does Not Close 2 x 10-3 2 x 10-3/d MVD1V93 CMS Valve V-93 Does Not Close 2 x 10-3 2 x 10-3/d B-29 -~ ,,nn, ..-m .e,, .,e , ~, ,wr_...,v, ,,.,,,-g., -m ...w -,,,,,.--..,,,,-,-,w,-,c- -.-.,,-w.,---

TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION ( DESCRIPTION UNAVAILABILITY FAILURE RATE IDE ER MVD1V81BMB Yalve V-81 Does Not Close 2 x 10-3 2 x 10-3/d MY,01V75AMB Valve V-75 Coes Not Close 2 x 10-3~ 2 x 10-3/d MVD1V87DMC Valve V-87 Fails to Remain Closed 6 x 10-8 1.2 x 10-7/hr MVD1V93 CMC Vahe 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 MIF1V87CMN 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 MIFIV81BMN Valve V-81 Fails to Receive Signal

7. x 10~3 2 x'IO-3 _d

/ MIF1V75AMN Valve V-75 Fails to Receive Signal 2 x 10-3 2 x 10~3/d Rupture of 6" Line Between V-99 M0F1618CMJ and Tee 5 x 10-11 1 x 10-10/hr Rupture of 6" Line Betwein~V-163 M0F1618BMJ and Tee 5 x 10-11 1 x 10-10/hr Rupture of 6." Line Between V-163 M0F1618AMJ and V-156 5 x 10-11 1 x 10-10/hr Rupture of 6" Line Between P-113 M0F1618DMJ and V-99 3 x'10-11 1 x 10-10/hr Rupture of 4" Line Between Dis-M0E1631AMJ charge Pipe and FW V-156 5 x 10-11 1 x 10-10/hr Rupture of 4" Line Between Dis-M0E1631BMJ charge Pipe and FW V-159 5 x 10-11 1 x 10-10/hr I ^ Rupture of 4" Line Between Dis-M0E1631CMJ charge Pipe and FW V-162 5 x 10-11 1 x 10-10/hr l B-30

1 TABLE B.2 (Continued) FAULT IDENTIFIERS FOR THE SEASROOK EMERGENCY FEED ST,ATION FAULT d DESCRIPTION UNAVAILABILITY FAILURE RATE i IDENTIFIER Rupture of 6" Line Between Dis-M001625AMJ charge Pipe anc ?CV-4326 5 x 10-11 1 x 10-10/hr I Rupture of 16" Condensate Line F0T1'079AMJ Between Tank and V-141 5 x 10-11 1 x 10-10/hr Rupture of 24" Condensate Line F0K1080AMJ Between V-143 and V-141 5 x 10-I1 1 x 10-10/hr i Rupture of 24" Line Between F0K1080BMJ Suction Line and V-142 5 x 10-11 1 x 10-10/hr Rupture of 20" Suction Line F0J1080CMJ Between V-143'and Tee 5 x 10-11 1 x 10-10/hr Rupture of 20" Suction Line F0J1080DMJ Between Tee and V-145 5 x 10-11 1 x 10 '0/hr Rupture of 8" Line Between Tee F0G1080AMJ and V-340 5 x 10~11 1 x l'0-10/hr Rupture of 8" Line Between V-340 ~ l0 F0F1080BMJ and V-152 5 x 10~11 1 x if /hr Rupture of 8" Line Between V-152 F0G1080CMJ and P-113 5 x-10-11 1 x 10~IU/hr _~ MVA1V99AMJ Rupture of V-99 5 x 10-9 1 x 10 /hr 8 5 x 10-9 1 x 10-8/hr MVP1V163MJ Rupture of V-163 MVM1V156MJ Rupture of SIV-156 .1.x 10-8 2 x 10-8/hr Hvt*1V159MJ Rupture of FWV-159 1 x' 10-8 2 x 10-8/hr Rupture of Bypass Inlet Line F0F10900MJ Between Tee and V-341 5 x 10-10 1 x 10-10/hr 1 FVM1V152MJ Rupture of V-152 1 x 10-8 2 x 10~9/hr FVA1V340MJ Rupture of V-340' 5'x 0-9 1 x 10-8/hr rL ~1'x.10 2 x 10-8/hr 8 FVMIV142MJ Rupture of V-142 ~, c. N. B-31 s -- f.-

~ TABLEB.2(Continued) FAULT IDEhTIFIERS FOR THE SEABROOK EMERGENCY FEED STATION ) 1 k -IDE ER D SCRIPTION UNAVAILABILITY FAILURE RATE i ] - -- FVM1V341MJ Rupture of y-341 1 x 10~0 2 x 10 /hr 4 i MVD14326MJ Rupture of PCV-4326 5 x 10-8 1 x 10 /hr 4 1 FVD1Y145MJ Rupture of V-145 5 x 10-9 3 x 10~"/hr FVA1V343MJ Rupture of V-343 5 x 10~9 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 l Rupture of Bypass Outlet Line FDG1090FMJ Between V-344 and Tee 5 x 10-10 1 x 10-10/hr s i Rupture of 8" Line Between V-344 F0F1080GMJ and V-343 5 x 10-10 1 x 10-10/hr b h FMJ1V344MJ, Rupture of V-344 1 x 10-8 2 x 10~9/hr MVM1V162MJ Rupture of FWV-162 1 x 10-8 2 x 10-8/hr Rupture of P-113 Suction Iso-FVD1V143MJ 1ation Valve V-143 5 x 10~9 I x 10-8/hr Rupture of P-113 Suction FVD1V141MJ Isolation Valve V-141 5 x 10~9 1 x 10-9/hr - - ~ ~ MVA1V99-MA Check Valve V-99 Fails to Open 2'x 10-4 2 x 10 /d 4 Manual Isolation Valve V-152 Fail s 4 4 FVM1V152MD to Remain Open (Plugs) 1 x 10 1 x 10 /d Isolation Valve V-143 Fails to FVD1V143MD Remain Open 1 x 10 4 1 x 10-4/d Isolation Valve V-143 Fails FVD1V141MD to Remain Open 1 x 10-4 1 x 10 /d 4 s MVDIV1630A Operator Fails to Open V-163 1 x 10-2 1 x 10-2/d l MVD1V1560A Operator Fails to Open V-156 l'x 10-2 1 x 10-2/d B-32 w -.--r, ,-..,r.,-v

FAULT ZDENT8FIERS FOR THE SEASROOK EMERGENCY FEED STATION FAULT DESCRIPTION UNAVAILABILITY FAILURE RATE IDENTIFIER FVM1V15290 Operator Fails to Restore V-152 1 x 10-3 1 x 10-3/d Fb1V1439DOperator Fails to Restore V-143 1 x 10-3' 1 x 10~3/d FVD1V54190 Operator Fails to Restore V-141 1 x 10-3 1 x 10-3/d MPB1P11399 P-113 Out of Service 4.2 x 10-4 MPS1P16199 P-151 Out of Service 5 x 10'4 MPS1P113ME P-113 Fails to Start 4 x 10-3 4 x 10-3/d MPS1P161ME 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 1 x 10-2 1 x 10-2/d RCA1A93-98 Breakar A-93 Circuit Breaker A-93 Fails RCA1A93-MB to Close 1 x 10-3 1 x 10-3/d Diesel Generator 1A Circuit RCA1AS4-MB 3reaker A-54 Fails to Close 1 x 10-3 1 x 10'3/d Diesel Generator A Fails 1.0 x 10-2 1.0 x 10-2/d RGDIDG1AME to Start ~ ~ ~ ~ RGDIDG1AMG Diesel Generator 1 A Fails to Run 3.0 x 10-3 6.0 x 10-3/hr 6IC1FPSAMN No Actuation Signal Generated - 5.8 x 10-3 5.8 x 10~3/d Startup Feedpump Suction Isolatica 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 i 1 l B-33 9 r, ,-,y

TABLE B.3 NRC-SUPPLIED CATA USED FOR PURPOSES OF CONDUCTING A wnrAAAiin. A:55e.55m.ai of r./us 1iG Fds DE51Gss Ano inEIA rote.siiAL AEi.1A5it.ITIES Point Yalue Estima'a i of Probabilt'y of* Fa11ure on Denand I. Cermenent (Hardware) Failure Data 9 a. Valves: Manual Val.ves (Plugged) 1 x 10 4 - Check Valves 1 x 10-4 Met:r-Operated Valves Mechanical Cc=ponents 1 x 10-3 Plugging Contribution 1 x 10-4 Ccatrol Circuit (Lecal to Valve) w/Quartarly Tests 6 x 10-3 l w/Menthly Tests 2 x 10-3 I b. P t= s : (1 Pt=o ) 1 x 10-3 Mechanical Ceepenents C:ntrol Cireutt w/ Quarterly Tests 7 x 10-3 w/Montaly Tests 4 x 10-3 c. ktuation Lecic 7 x 10'3 l 3 ) irror f ac crs of 3-10 (up and dcwn) abcut such value's are not = unexpectec f r basic data uncartainties. e s B-34

g e

TABLE B.3 (Cont'd) II. Test and Maintenance Cutace Contributions: a. Calculational Approach 1. Test Outage Q ( hrs / test) ( tests / year) ~ TEST nrs/ year 2. Maintenance Outage Q (0.22)( hrs /maint. act) MAINT. /20 I b. Data Tables for Test and Maint. Outages

SUW.ARY OF TEST ACT DURATION Calculated i

Range on Test Mean Test Act Component Act Duration Time, hr Duration Time, t. hr-D Pumos 0.25 - 4 1.4 Yal'ves 0.25 ~2 0.86 Diesels 0.25.- 4 1.4 Instrurentation 0.25 - 4 1.4 LOG-NCRMAL MODELED MAINTENANCE ACT DURATION 4 Calculated ~ ~ Range on Maintenance Mean Maintenance Act Component Act Duration Time, hr Duration Time, t, hr j D Pumps 1/2 - 24 7 1/2 - 72 19 Yalves 1/2 - 24 7 Diesels 2 - 72 21 Instrumentation 1/.4 - 24 6 - <~ Nota: These data tables were taken from the Reactor Safety Study (WASH-1400) for purposes of this AFW systen assessment.. I Where the plant technical specifications placed limits on the outage duration (s) allcwed for AFW system trains, this l 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 AFV system unavailability from outages due to testing,. B-35 i O P g

TABLE B.3 (Cont'd) III. Human Acts & Errors - Failure Data: Estimated Human Error / Failure Probabilities ..__,Hodifying Factprs,& Situations' With Valve Position With local Walk-Around & W/0 Elther ' Indication in Control Room Double Check Procedures Polht Valus Est Est. on Pol'nt Value Est Est. on PointYa15e~~ 'Est On Error l Error Estimate Error Factor Factor Factor a. Acts & Errors of A Pre-Arefdent Nature 1. Valves hispositioned During Jest /Hafnt (a) Specific Single Valve Wrongly of Valves During Conduc) tion Selected out of A Popul on g of a V2 -2 -2 cN Test or Haintenance Act (X No. I g 10 X 1 l y 10 g 10 g1 of Valves in Population at Choice) 75 X 20 2 X 10 I 10 (b) Inadvertently teaves Correct -4 -3 Valve in Wrong Position 5 x 10 20 5 x 10 10 10'2 10 2. Morethadonevalveisaffected 1 x 10'4 20 .1 x 10' ' 10 3 x 10'3 10 (coupled, errors) 3. Hiscalibration of Sensors / Electrical Relays -3 -2 5 x 10 10 10 10 (a) One Sens,or/ Relay Affected (b) Here than one Sensor / Relay

1 x 10 10 r3 x 10 10 Affected 3

3 1 0

~~

~ ' ' ~

TABLE B.3 (Cont'd) istimated Failure r P reb. for Primar~v i Time Actuatien Ocerator to Actuata i Needed AFVS Cemoonents l a) Acts and Errors of a Post-Accident Nature 1. Manual actuation of 5 min. ~5 x 10-2 AFVS frem Cor. trol 15 min. 1 x 10-2 Reem. Considering ~ 20 mi'n. ~5 x 10-3 "non-dedicatad" overstar to actuata AF45 and possible backup actuation of AFWS. e 4 l i 1 6, 9 B-37 5 4

APPENDIX C 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 perforned using event-tree and fault-tree logic techniques and is in-tended to evaluate the potential for sys. tem 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, ;ommon causes, single-point vulnerabilities, and outages due to test and maintenance. The stated purpose of the recommendations associated with TMI Action Plan Item II.E.11 was to decrease the unreliability of AFW systems towards a goal of 10 4 to 10-5 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 al'1 other AFW sys' tem 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 ) l system to meet the specified reliability goals. The Seabrook AFW system consists of a two-pump safety grade emergency feedwater (EFW) system and a 1 In this report the term " auxiliary feedwater", or AFW, system as applied to Seabrook means the conbined emergency feedwater ano startup pump systems. C-1 3 ~ " * "

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.

2 In PSNH replied to the staff's concerns by letter on December 4,1981 this response it was noted th-t provisions were included in the Seabrook de-sign to allow the startup feed pump to be powered from an emergency 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. PURPOSE The purpose of this study was to perfom 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 provided in Table C-1. 2 Letter No. SBN 198, T.F. H4.4.98 to Mr. Frank J. Miraglia from Mr. John DeVi ncenti s. C-2 o k.

~ 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 feedwater with coincident loss of offsite station power; o Loss of main feedwater with coincident loss of all station AC power. O m eegdD h - e e 08 m o G e e C-3 l l I

TABL~. C.; Engineering Drawing List for the Seabrook Nuclear Station Emergency and Startup Feedwater Systems Drawing Title Number Mechanical System P&I Diaarams: i 1. Main Steam System (Sheet 1) 9763-F-202074

2.. Emergency Feedwater System 9763-F-202076

3. ' Condensate System 9763-F-202077 4.

Feedwater System - 9763-F-202079 4 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 i 8. Miscellaneous Buildings Compressed 9763-F-202108 Air Headers Electrical System One Line Diagrams. 4 I 1. Unit Electrical Distribution 9763-F-310002 9763-F-310007 I 2. 4160V Switchgear Bus 1-E5 i i 3. 480V Unit Substatian Buses ~ 9763-F-310013 1-E51 and 1-E52 i 4. 125VDC and 120VAC Instrument Buses 9763-F-310041-i S. Turbine Building 480V Motor Control 9763-F-310046 i i Center 1-E523 ~'" Logic Diagrams: 1. Symbols 9763-M-503100 l I 2. FW-Start-up Feed P-113 9763-M-503580 l 4 3. FW-Prelube P-161 For 9763-M-503581 l Start-up Feed P-113 Sht 1 4. FW Emerg Fd P-37A Steam 9763-M-503584 4 Supply Viv (MS-V128) Train B 5. FW Emerg Fd P-37A Stm Supply 9763-M-503585 Viv (MS-V127) Train A 6. FW-Emerg Feed P-37B 9763-M-503586 C-4

TABLE C.1 (Cont'd) 7. FW-Emerg FW Bypass /Inop 9763-M-503599 Status Alarm 8. MS-Trip & Throttle Valve 9763-M-503672 Y-129

9. jFW-Emergency Valves 9763-M-504152 1

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 Diacrams: 1. FW-Start-up Feed P-113 9763-M-506480 & Prelube Pmp P-161 2. FW-Feed Pump P-32B 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-37B 9763-M-506499 TE-4271 & TE-4347 6. MS Supply To Emerg Fd Pmp 9763-M-505555 Turbine Isol Viv 7. FW-Emerg Feed Pump P-37A 9763-M-507043 Discharge Flow 8. FW-Emerg Feed Pump P-37B 9763-M-507044 9. FW-Emerg FW Valve FV-4214 9763-M-507056

10..FW-Emerg FW Valve FY-4224

. 9763-M-507057 l 11. FW-Emerg FW Yalve~FV-4234 9763-M-507058 12. FW-Emerg FW Valve FY-4244 9763-M-507059

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14. Prelube Pump 1-P-161 Legend 9763-M-310844 SHCN1b

~ & Switch C-5

TABLE C.1 (Cont'd) 15. Prelube Pump 1-P-161 9763-N-310844 SHCNic Cable Schematic FSAR Drawings: 1. IFunctional 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 l Pumps Startup 6. Separation of Instrument and Control Power Figure 8.3-3 Sources ~ s e

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AFPtNUlx U ~ D-LJ PLJBLIC SERVICE SEABRogK STATION sns_n, o% ( Companyof New Hampshw e 1671 Worcester Rood Fromincham, Mossochusetts 01701 (617). E72 8100 September 7, 1982 SBN-321 T.F.- H 4.4.98 B 7.1.2 United States Nuclear Regulatory Com=ission Wa shington, D. C. 20555 Attentio n: 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) PSNE 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 Instrheentation and ( Controls Systems Branch (ICSB); A-K", J. DeVincentis to F. J. Miraglia

Subject:

Seabrook Station Emergency Feedwater Syste= Design Changes

Dear Sir:

~ During the Staff review of the Seabrook Station Emergency Feedwater - System (EW), a nu=ber of design changes have bean reco== ended and are being implemented. These design changes are based on :he review of th,'e Emergency Feedwater System Reliability Analysis..[_RJe erenc'e (b)) and also bring the-Seabrook design into co=pliance with the latest Standard Review Plan. These design changes will be incorporated into a revisio'n to the Final Safety Analysis Report as soon as the final details are established.' The following describes design changes which are presently being implemented relative to the Seabrook EN System. An attached simplified. sketch is included for clarification of some of cl}e design changes. " 1. A continuous'*mrnimum' flow recirculation path will be provided from each EW pump's discharge to the condensate storage tank via the opposite. pump's suction line. This recirculation path will assure a continuous flo'w through ac 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 ( perf ormance testing. D1 l %,, r 'r r I

.r. ~ United States Nuclear Regulatory Commission September 7, 1982 i At tentio n: Mr. Frank J. Miraglia, Chief Page 2 2. Redundant, safety grade flow isolation valves will be provided in each EW branch supply line to each steam g' enerator. Safety grade controls will be provided at both the main contr'ol board and remote shutdown locations for these valve,s. Further information relative to this modification can be found in Reference (c). 3. Manual isolation valves will be provided upstream of each pair of flow isolation valves to sach steam generator.. These manual isolation valves will per=it isolation of any EFJ flow isolation valve while retaining the availability of both EW pumps and the Startup Feedwater pump. 4. Safety grade, Seismic. Category I air accumulators will be provided as a back up air supply for the actuators of both main steam supply valves (MS-V127 and MS-V128) to the turbine-driven EW pump, P.-3 7A. h se accumulators will be' sized to proy,ide at least two, complete valve operations plus maintain the valves closed for a minimum of 'four hours. This safety grade air supply will upgrade the reliability of these valves consistent with the Class 1E controls presently utilized in the design. 1 5. The Startup Feedwater-(SUF) pu,mp discharge valve to the EW header, W-V156, will be relocated out of the EW Pump Room. This will assure the ability to cross-tie the SUF pump to the EFW System should _a series of potential failures render both EW pumps inoper'ab'le and the EW Pump Roo= inaccessible. 1 Additionally, during both our in-hobse and your Staff reviet ef ~ Reference (b), three areas were found which should be clarifigd et corrected. ~ First, on Page 12 of Reference (b), an asterisk notes that only one of the steam admission valves (MS-V127) to the' turbine-driven EW 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.gnsure the ability to stari and/or .stop'the turbine-driven EW pump from either the main cont'rol bor ra or the remote shutdown panels. ~ Second, on Page 15 of Reference (b), relative to the manual valve , realignments required to provide SUF pump flow to the EW header, it states that the SUF pump recirculation isolation valve (FJ-V109) must be closed to prevent a diversion of pump flow to, the Condensate Storage Tank (CST) should the recirculation flow control valve (PCV-4326) fail open. What was not. considered, however, is that the capacity of the SUF pump.is significantly greater than that of an EW pump. At a TDH equivalent to the design rating of the EW pump, the SUF pump has a flow capacity greater than an EN pump, even when maximum flow is diverted back to the CST through the recirculation valve. Therefore, it is unnecessary to close valve W-V109 to ensure sufficient flow f rom the SUF pump to the steam generators. This is one less manual action necessary for this ( operation. Third, on Page 29 of Reference (b), _a note on' the bottom of the page indicates that a loss of off-site power will result in'elosure of the D-2

T O- ~ United States Nuclear Regulatory Commission September 7, 1982 Attention: Mr. Frank J. Miraglia, Chief Page 3 b mala feedwater isolation valves. Th'is. note is incorrect - the main feedwater isolation valves will not close due to a loss of of f-site power. Additionally, it should be noted that the loss of off site power does not result in a loss of control of the main feedwater regulating valves nor the main feedwater regulating bypass valves. The result is, the SUF pump can be utill:ed to supply feedwater to the steam generators during a loss of of f-site power event without the need of manual. valve alignments to provide. flow through the EW System. Flow from the SUF pu=p to the steam generators can be accomplished utilizing the normal main Feedwater System. It is hoped that the above information will assist your Staff in their evaluation of the Seabrook Station Emergency Feedwater System and preparation of the Safety Evaluation R.eport. If further information is necessary, please feel free to contact us. Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY s O b ~ J. DeVincentis . Pr,oject Manager PA/kae cc: Mr. Robert Jaross, Argonne National Laboratories m - ef w O h u O m D L O e D-3 e

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1 ? SEABRN STAM LJ PUBLIC SERVICE lD engi omc., Companyof New Hampehu o 1671 Worcewer Road Framinoham, MosIachusetts 01701 (617) - 872 - 8100 December 1, 1982 SBN -394 T.F. B7.2.3 ) Uhited States Nuclear Regulatory Co:nission Washington, D. C. 20555 Attention: G. W. Knighton, Chief Licensing Branch No. 3 Division of Licensing

References:

(a) Construction' Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) Seabrook Station Environmental Report - Operating License Stage, as revised

Subject:

Supplemental Information on the' Use of Steel H-Frame Structures along the Seabrook to Newington Transmission Line Right-of-Way

Dear Sir:

In response to a verbal request by your Mr. L. Wheeler, we are providing the following additional infor=ation concerning Applicants use of steel H-frame structures along the Seabrook to Newington transmission line right-of-way, rather than wooden H-frames as mentioned in the Seabrook Station Final Environmental Statement-Construction Permit Stage (FES-CP). The reasons for this change in H-frame structural materials were briefly discussed in Applicant's response to ER-OLS RAI 310.2 (Reference (b)). Line #369, the 345 kV transmission line that connects Seabrook Station to Newington Station, was constructed with 80 steel tangent structures, rather than those of wood as mentioned in the Seabrook FES-CP. These steel ' structures are located at regular intervals along the transmission right-of-way from Seabrook to Newington, except for the stretch across the Hampton Marsh and from the Portsmouth Rotary into Newington Station where single steel poles were used as explained in the Seabrook Station FES-CP. At the time of the CP proceedings, PSNH testified that wood structures vould be used for Seabrook related transmission line construction, and that I the alternative to wood was the use of galvanized steel lattice towers. Not only did PSNH believe there was a visual impact advantage to wood over steel f lattice structures, but the need for concrete foundations for steel was a l further environmental disadvantage. Years later,.then these transmission line material specifications were being reviewed, it became evident that new concepts in steel structure design, together with the use of weathering steel, permitted further reduction in environmental impact beyond the degree committed. At that time, PSNH felt the change from wood to steel was not a l N 21201 PDR ADOCK 05000443 D PDR { .a. n:= w, u.m:-

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l United States Nuclear Regulatory Commission December 1,1982 Attention: G. W. Knighton Page 2 step backwards, rather it represented an improvement in the overall environmental acceptability of the structures. The advantages of the ) weathered steel simple H-frame supports over similar structures of wood are as ) follows: 1. The use of steel permitted installation of H-frames with less bulk, making the structures appear less massive, hence reducing their visual impact. A comparison of critical dimensions of steel vs. wood is as follows: S teel Wood Typical diameter of poles 14.5 inches 18 inches Dimensions of crossams 4.5 x 4.5 inches 5.25 x 9 0 inches Height of structure and length of crossarms Same Same 2. The steel poles were set into the ground just as wood poles would be, thereby avoiding foundation work. 3 The life of steel structures is equal to o'r greater than wood with less maintenance. Wood structures must be inspected periodically and occasionally. require ground-line treatment with wood preservatives. Wood pole maintenance would increase the travel over the right-of-way. 4. The weight of wood structures is about twice that of steel thus constru'ction could be accomplished with smaller, lighter equipment. 1 5. The color of the steel structures is very much like that of treated wood, so their appearance is quite similar (see enclosed photos). In sununary, the use of steel for tangent structures on the l Seabrook-Newington line was a change from that explained in PSNH's testimony before NRC. However, it is PSNH's opinion that the change was totally in i keeping with the spirit of the commitment to minimize the environmental impact of this transmission line. This expanded discussion on the use of steel vs. wood H-frame structures on the Seabrook-Newington transmission line will be included in the next revision to the Seabrook Station ER-OLS. I very truly yours, John Devincentis Prcject Manager l l RAM /fsf Enclosures cc: Atomic Safety and Licensing Board Service List . m:T;T *m A.7 % s- ...>.>;=* L AD..'

..a.... O ASLB SERVICE LIST Philip Ahrens, Esquire Assistant Attorney General Department of the Attorney General Augusta, ME 04333 Representative Beverly Hollingworth Coastal Chamber of Commerec 209 Winnacunnet Road Hampton, NH 03842 William S. Jordan, III, Esquire Harmon & Weiss 1725 I Street, N.W. Suite 506 Washing ton, DC 20006 E. Tupper Kinder, Esquire Assistant Attorney General Office of the Attorney General 208 State House Annex Concord, NH 03301 Robert A. Backus, Esquire 116 Lowell Street P.O. Box 516 l Manchester, NH 03105 Edward J. McDermott, Esquire i Sanders and McDermott Professional Association 408 Lafayette Road l Hampton, NH 03842 Jo Ann Shotwell, Esquire Assistant Attorney General Environmental Protection Bureau Department of the Attorney General One Ashburton Place,19th Floor l Boston, MA 02108 ^ r* ?: g, 'a ,"***=7_ P J" i .e 4e, =. =. * - f.,

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ML20082B202

Contents

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SUMMARY

1.0 INTRODUCTION

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1.0 INTRODUCTION

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1.0 INTRODUCTION

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