ML20039E178

Auxiliary Feedwater Sys Reliability Analysis Final Rept, Prepared for Toledo Edison Co

ML20039E178
Person / Time
Site:	Davis Besse
Issue date:	12/31/1981
From:	Gross J EDS NUCLEAR, INC.
To:
Shared Package
ML20039E172	List: ML20039E171 ML20039E178
References
02-1040-0195, 2-1040-195, TAC-43516, NUDOCS 8201060613
Download: ML20039E178 (350)

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DAVIS-BESSE UNIT No. 1 A

AUXILIARY FEEDWATER SYSTEM l

j RELIABILITY ANALYSIS 1

W FINAL REPORT i

HEdslVED O[G 2 9193) i NUCLE,1

'*1G lI i

l 1il i

I Prepared by:

EDS Nuclear Inc.

for f

!i Toledo Edison Company l

f December, 1981 Reoort No. 02-1040-0195 l

Revision 1 l

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i 8201060613 6

DR ADOCK 0 00 PDR

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EDS NUCLEAR INC.

NEW YORK REGIONAL OFFICE I

REPORT APPROVAL COVER SifEET Client:

Toledo Edison Company Project:

Auxiliary Feedwater System Reliability Analysis Job Number: 1040-003-671 R* port

Title:

Davis-Besse Unit No. 1 - Auxiliary Feedwater System Reliability Analysis Final Report Report Number: 02_1040_iogs Rev.

o The work described in this Report was performed in accordance with the EDS Nuclear Quality Assurance Program. The signatures belcw verify the accuracy of this Report and its compliance with applicable quality assurance requirements.

Prepared By:

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h oV I fl No l QTXcSS i[9 [8/

Date:

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Reviewed By:

,ucM 7 Adh Date:

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Approved By:

Date__ j /O

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IConcurrenceBy(-

M.///,b Date:i/hh.' /

J Regional Quality Assurance Manager

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REVISION RECORD Rev.

Approval No.

Prepared Reviewed Approved Concurrence Date Revision k

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Report No. 02-1040-1095 Revision 1 Page i TABLE OF CONTENTS I

Page 1.0

SUMMARY

l 2.O SCOPE AND OBJECTIVES 6

2.1 Background

6 2.2 Objectives 6

2.3 Scope of Work 7

3.0 SYSTEM DESCRIPTION 9

3.1 AFWS Safety Function 9

3.2 Pre-TMI System Configuration 10 3.3 Post-TMI System Configuration 11 3.4 Third Train System Configuration 13 I

3.5 Analysis-Based System Configuration 14

3. 6 AFWS Support Systems 15 4.0 METHODOLOGY 29 4.1 System Fault Tree Development 29 I

4.1.1 Fault Trees Developed 29 4.1.2 Fault Tree Methodology 31 4.2 Data Analysis 33 I

4.2.1 Industry Data Review 33 4.2.2 Review of Davis-Besse Experience 34 4.2.3 Recommended Data Base 39 1

4.3 System Unavailability Analysis 40 4.3.1 Quantitative Analysis of 40 Fault Trees I

4.3.2 Uncertainty Analysis 42 4.3.3 Eqportance Ranking 42 I

4.4 Initiating Event Analysis 43 4 4.1 Frequency Estimates 43 4.4.2 Factors Influencing Event 45 Frequencies 4.5 Combined System / Event Analysis 46

I Report No. 02-1040-1095 Revision 1 Page 11 TABLE OF CCNTENTS (continued)

Page 5.0 RESULTS 73 5.1 Relative Unavailability Ranking of AFWS 73 Configurations 5.1.1 Differences Among Event 73 Categories 5.1 2 Differences Among AFWS 73 Configurations 5.2 Significant Contributors to AFWS 74 1

Unavailability 5.3 Results of Combined System / Event Analysis 76 I

5.4 Potential Common Cause Contributors to 77 AFWS Unavailability

6.0 CONCLUSION

S 93

7.0 REFERENCES

95 I

Report No. 02-1040-1095 Revision 1 Page 111 LIST OF TABLES PAGE l-1 Results of ANS Fault Tree Analysis 4

1-2 Overall Figure-of-Merit for AWS Reliability 5

3-1 Stean-Feedwater Rupture Control System Actuation 18 3-2 AN S/ Electric Power System Interfaces 19 4-1 System Safety Functions 47 4-2 Plant System Designator 49 50 4-3 Component Code 4-4 Failure Mode Code 52 4-5 AWS Reliability Study Failure Data 53 4-6 INilure Experience for Davis-Besse Valves 63 4-7 Failure Experience of Davis-Basse Auxiliary 64 Feedwater Pumps 4-8 Failure Experience of Davis-Besse Diesel 65 1

Generators 4-9 Initiating Events Challenging AWS 66 4-10 Initiating Event Frequency Estimates 67 5-1 AWS Unavailability 78 5-2A Significant Contributors to A NS Unavailability:

79 I

Pre-TMI Configuration - Category 1 5-2B Significant Contributors to A N S Unavailability:

80 Pre-TMI Configuration - Category 2 5-2C Significant Contributors to A NS Unavailability:

81 Pre 'DiI Configuration - Category 3 5-3A Significant Contributors to ANS Unavailability:

82 Post-OiI Configuration - Category 1 1

5-3B Significant Contributors to AWS Unavailability:

83 Post-HiI Configuration - Category 2 I

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II Report No. 02-1040-1095 Fevision 1 Page iv LIST OF TABLES (Cont.)

PAGE

.I 5-3C Significant Contributors to A WS Unavailability:

84 Post 'Dil Configuration - Category 3 5-4A Significant Contributors to AWS Unavailability:

85 Third Train Configuration - Category 1 5-4B Significant Contributors to AWS Unavailability:

86 Third Train Configuration - Category 2 lI 1

5-4C Significant Contributors to A W S Unavailability:

87 Third Train Configuration - Category 3 5-SA Significant Contributors to AWS Unavailability:

88 Analysis-Based Configuration - Category 1 5-5B Significant Contributors to AWS Unavailability:

89 Analysis-Based Configuration - Category 2 5-5C.

Significant Contributors to A NS Unavailability:

90 Analysis-Based Configuration - Category 3 5-6 Combined System / Event Analysis Results 91 5-7 Relative Significance of the Pre-D1I - Post-TMI Changes 92 i

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Report No. 02-1040-1095 Revision 1 4

Page v LIST OF FIGURES PAGE 3-1 Pre-TMI AWS Configuration 20 3-2 Pre-TMI Main Steam System Configuration 21 3-3 Post-TMI A WS Configuration 22 4

3-4 Post-DiI Main Steam System Configuration 23 i

3-5 Post-TMI Startup Pump Configuration 24 3-6 Makeup System Configuration 25 1

3-7 Third Train Configuration 26 i

3-8 AWS Analysis Based Configuration 27 3-9 Startup Pump Analysis Based Configuration 28 4-1 Davis-Besse A W Reliability Program Flow Diagram 68 4-2 Symbols Used in Fault Trees 69 4-3 System / Event Analysis Matrix 72 4

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Paport No. 02-1040-1095 Revision 1 Page vi A'I"I'ACHMENTS I

Drawing No.

AWS (Pre-24I) Fault Tree 1040-003-001 flain Steam System (Pre-24I) Fault Tree 1040-003-002 Electrical System (Pre-TMI) Fault Tree 1040-003-003 AWS ( Post-24I) Fault Tree 1040-003-004 Main Steam System (Post-24I) Fault Tree 1040-003-005 Electrical System (Post-TMI) Fault Tree 1040-003-006 Start-up Pump (Post-24I) Fault Tree 1040-003-007 Start-up Pump with Feed and Bleed 1040-003-008 I

AWS (Analysis-Based) Fault Tree 1040-003-009 Main Steam System (Analysis Based 1040-003-010 1

Fault Tree Start-up Pump with Feed and Bleed 1040-003-011 (Analysis Based) l'I I

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I Report No. 02-1040-1095 I

Revision 1 Page 1 1.O

SUMMARY

This report presents results of a reliability analysis of the Davis-Besse Nuclear Power Station, Unit No. 1 auxiliary feedwater system (A W S).

This analysis, performed by EDS I

Nuclear, Inc. for Toledo Edison Company (TECo), is in support of the TECo commitment to the U.S. Nuclear Regulatory Commission (NRC) for a continual review of auxiliary feedwater system reliability. This analysis also provides a comparative I

probabilistic risk assessment basis for various design modifications to further upgrade the AWS for added reliability and performance. The analysis has resulted in the quantification of AWS unavailability, identification of major contributors to system unavailability and recommendations for system modifications to minimize unavailability.

The analysis is based on development of fault trees for the AW S and other plant systems supporting the A W S safety

function, i.e.,

the delivery of cooling water to one or both I

steam generators whenever the main feedwater flow has been interrupted. The fault trees depict the logical relationship between the failure to deliver sufficient feedwater to the steam generators and the basic mechanical, electrical and human factors which may cause an individual system component to fail. Failure data, derived from industry sources and reviews of Davis-Besse plant-specific operating experience, are used to assign probabilities of failure to the basic component failure mechanisms. These basic event probabilities are then propagated through the fault tree, using Boolean I

algebra, in order to derive a probability for failure to achieve the A WS safety function. For the purpose of this report, failure to achieve the AWS safety function is defined as "AWS unavailability", even though the failure may result from failures in other plant systems which support the AWS safety function.

I Initiating events which challenge the AWS can be conveniently categorized as follows:

Category 1 - Events in which the main feedwater flow or reactor coolant system forced circulation is interrupted, but offsite electrical power is available to the plant.

Category 2 - Events in which offsite electrical power to the plant is interrupted.

Category 3 - Seismic events.

Report No. 02-1040-1095 l

Revision 1 Page 2 The AWS unavailability is determined for each of these categories of initiating events. The annual frequencies with which these events occur are estimated from industry experience and from reviews of Davis-Besse operating history.

The frequency of the initiating event is then multiplied by the AWS unavailability. The result is the annual frequency with which the AWS is unavailable when called upon to perform its safety function. This is the overall figure-of-merit used to judge the relative reliability of various AWS configurations.

The following four AWS configurations are considered in this analysis:

The AWS configuration that existed in March,

" Pre-21I" 1979.

" Post-DiI" The AWS configuration that contains DiI-related plant changes, including those I

planned to be implemented in the 1982 refueling outage.

Included in this configuration is a written procedure for fulfilling the AWS safety function using the I

main feedwater startup pump, reactor coolant system makeup pump and power operated relief valve as a backup to the AWS.

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" Third Train" - A potential configuration, which utilizes the main feedwater startup pump, in an altered I

alignment, as a backup third train of auxiliary feedwater.

A configuration which incorporates l1 "Analysf9-I Based" recommendations resulting from this reliability analysis, but does not include the realigned startup pump.

The unavailabilities of each AWS configuration for the three categories of initiating events are summarized in Table 1-1.

The overall figure-of-merit, the annual frequency with which each A WS configuration is unavailable when challenged, is presented in Table 1-2.

I The following conclusions are reached as a result of the analysis:

I The Pre 'IMI configuration unavailability is dominated by potential human errors, primarily in valve misalignment.

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I Report No. 02-1040-1095 Revision 1 Page 3 I

The Post 'IMI configuration incorporates many changes to plant procedures which diminish the likelihood of human errors. The unavailability is then dominated by mechanical failures, primarily associated with motor operated valves.

The third train configuration reduces the AWS unavailability by over an order of magnitude for Category 1 and Category 2 events through the addition of a third I

redundant train.

The analysis-based configuration provides over an order A

of magnitude improvement in AWS availability by addressing those specific mechanical factors and procedural limitations which dominate the Post-TMI results.

Significant improvements in the Davis-Besse AW S reliability have already been achieved since the original NRC requests (1) 1 (2) for a review and upgrade following the Three Mile Island, Unit 2 (TMI-2) event in March, 1979.

Further improvements are planned. Of the alternatives examined in this study, the third train and analysis-based configurations offer the greate; c improvement in AWS reliability. The cost of the third train configuration is relatively high. The analysis-based alternative offers an even greater improvement l

in system reliability, and its associated costs are likely to be relatively low.

The design and procedures modification of the analysis-based configuration are now planned as a means to enhance the Davis-Besse AWS reliability and performance.

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Report No. 02-1040-1095 Revision 1 I

TABLE l-1 Page 4 Results of ANS Fault Tree Analysis I

\\MS Unavailability (per demand) g n

l Category 1 l

Category 2 l

Category 3 ANS Configuration l

Events l

Events l

Events i

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Pre-TMI l

3.3 x 10-2 l

4.1 x 10-2 l 8 8 x 10-2 I

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Post-tiI l

6.6 x 10-4 l

5.5 x 10~3 l 1 9 x 10-2 I

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I Third Train l

4.5 x 10-5 l

1,4 x 10-4 l 1,9 x 10-2 I

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Analysis-Based l

3.3 x 10-5 l

9,3 x 10-5 l 1,1 x 10-2 l

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TABLE l-2 Page 5 Overall Ficmre-of-Merit for AWS Reliability lI l Frequency of AWS Unavailability jI l

When Challenged (yr~1) l AF.fS Configuration l ('Ibtal of all event categories )

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Pre-DiI l

8.2 x 10-2 l

Post-TMI l

3.3 x 10-3 1

Third Train l

2.2 x 10-4 Analysis-Based l

1 4 x 10-4 1

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lI Report No. 02-1040-1095 Revision 1

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Page 6 2.0 SCOPE AND OBJECTIVES

2.1 Background

As part of its review of the Three Mile Island, Ur.it-2 event, the NRC issued, on May 16, 1979, a Confirmatory Order (1) to l

the Toledo Edison Company as part holder of the operating license of the Davis-Besse Nuclear Power Station, Unit No. 1.

This order required, in part, that the licensee review all l

ll aspects of the safety grade auxiliary feedwater system to

5 further upgrade components for added reliability and performance. On Jaly 6,1979, the NRC issued a letter (2) lif ting the above Confirmatory Order, allowing Davis-Besse Unit No. 1 to return to power.

The safety evaluation attached to that letter indicated that

.l the NRC would at some future time require system diversity i

!W through the installation of an additional 100 percent capacity motor operated auxiliary feedwater pump, or an alternative acceptable to the staff.

In reviewing the NRC's intended purpose for such a modification, and relating it to the magnitude of the cost impact, TECo determined that a quantification of the relative i=

risk reduction actually provided by such a modification is appropriate.

2.2 Objectives The overall objective of this analysis is to evaluate the reliability of the Davis-Besse Unit No. 1 auxiliary feedwater system in delivaring feedwater to one, or both, steam generators whenever main feedwater is interrupted or whenever

'g reactor coolant system forced circulation is interrupted.

'g Each of four potential AWS configurations, identified in Section 2.3, is evaluated.

Specific objectives with respect to the evaluation of each A WS configuration are:

I to determine the AWS unavailability for various categories of plant initiating events, to identify the most significant contributors to AWS t

unavailability, when challenged.

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Report No. 02-1040-1095 Revi31on 1 Page 7 The objectives of the comparative evaluations of the four AWS configurations are:

to establish an overall figure-of-merit with which to judge the relative reliabilities offered by the four AWS configurations, to determine cost-effective system modifications to upgrade the AWS reliability and performance.

Formal fault tree techniques, as discussed in Section 4.0 are utilized in achieving these analysis objectives.

I 2.3 Scope of Work This reliability analysis examines four potential AWS I

configurations. These are:

" Pre-TMI" Configuration The AWS and other plant equipment as configured prior to implementation of TMI-2 related plant modifications.

I

" Post-TMI" Configuration The AW S and other plant equipment as configured I

sub;equent to implementation of certain TMI-2 related, and other, plant modifications.

It includes those modifications planned to be implemented through the 1982 I

Davis-Besse refueling outage.

It also includes a written pre edure to fulfill the AWS safety function using the ma!

feedwater startup pump, the reactor coolant system maktap pump and the power operated relief valve (PORV) as I

a backup to the AWS.

" Third Train" Configuration The same A WS configuration as for the Post-TMI case, except that a third manually initiated motor driven main feedwater startup pump would be aligned to supply

'I auxiliary feedwater flow, without the necessity for reactor coolant system makeup flow and steam venting via the PORV.

"7.nalysis-Based" Configuration The same AWS configuration as for the Post-TMI case, I

except that certain recommended system modifications, resulting from this study, are assumed to be implemented.

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Report No. 02-1040-1095 Revision 1 Page 8 The first two configurations are analyzed to demonstrate relative improvement to ANS reliability resulting from system I

modifications already planned or implemented by TECo.

The third configuration represents a system designed to address explicitly the NRC concerns with respect to A WS reliability as outlined in references (1) and (2).

The fourth configuration represents a system designed to address the most significant contributors to the Post 'IMI system unavailability, as determined from a comprehensive evaluation of system reliability.

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3.0 SYSTDi DESCRIPTION 3.1 AFUS Safety Function The AFWS is designed to provide coolant to the secondary side I

of the steam generators whenever the main feedwater flow has been interrupted or to establish natural circulation whenever the reactor coolant system forced circulation has been interrupted. This is necessary to maintain adequate core I

cooling and prevent fuel damage.

In the Post-TMI I

configuration, there are two ways in which this safety function can be met:

1.

The AFWS can deliver full capacity flow from at least one of the redundant AFWS turbine-driven pumps to one steam I

generator. The water delivery to the steam generator (s) l1 must begin within ten (10) minutes of the initial loss of main feedwater or loss of forced circulation.

The water delivery must continue until the reactor coolant system I

cools down and is depressurized to the point where the decay heat removal system can be operated.

I 2.

The main feedwater startup pump can be manually started and aligned to deliver coolant to either steam generator.

The present capacity of the startup pump is not sufficient for complete decay heat removal.

I Therefore, the manual initiation of feedwater via this path must be accompanied by the manual opening of the power operated relief valve (PORV), initiation of primary I

coolant makeup flow (through at least one makeuo pump) and isolation of the reactor coolant system letdown line.

In this mode, partial reactor coolant heat removal is obtained by venting fluid from the primary system I

through the PORV.

The makeup flow is necessary to prevent excessive reactor coolant inventory loss until the high pressure injection pumps can provide emergency core cooling.

In this mode, the safety function is accomplished if all actions are initiated within thirty (30) minutes of the initial loss of feedwater. The systems must function until the operating conditions for I

the decay heat removal system are reached.

Emergency procedures for this second approach exist in the Post-TMI configuration only for the situation in which offsite electrical power is available at the plant site.

Emergency procedures for the second approach are presently planned for the added situation in which offsite electrical power is not 1

I available at the plant site.

The extension to the emergency procedures is credited in the analysis-based configuration.

The combination of the startup pump, makeup pump and PORV is referred to as the " feed and bleed" method throughout the balance of this report.

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Revision 1 Page 10 I

The first of the above methods is the anticipated technique for fulfilling the AWS safety function. The second method is designed only as an emergency backup in the unlikely event that the first method is unsuccessful.

3.2 Pre-TMI System Configuration The Pre-TMI AWS is illustrated in Figure 3-1.

The system consists of two independent trains, each containing:

one steam-driven auxiliary feedwater pump, AC powered motor operated valves, I

crossover piping which allows the pump to supply water to either steam generator, redundant water supplies.

The primary sources of auxiliary feedwater are the condensate storage tanks (CST) with a minimum water supply of 250,000 gallons. Should this supply fail, plant procedures call for I

the manual transfer of the AWS pump suction to the fire protection system. The service water system provides an automatic safety grade backup to the other two supplies. The service water system is connected to the AWS through motor I

operated valves, which are initially aligned shut.

They receive an open signal on a low pressure condition at the AWS pump inlet, as measured by redundant pressure switches.

The auxiliary feedwater pumps are both driven by steam f rom the main steam generators. Normally, steam generator 1 provides steam to AWS pump turbine 1, and steam generator 2 I

provides steam to AWS pump turbine 2.

However, in the event of low pressure in one steam generator, the unaffected steam generator can provide steam to both turbines through crossover I

paths, as illustrated in Figure 3-2.

Normally, the motor operated steam admission valves are aligned closed. They receive an open signal from redundant Steam and Feedwater Rupture Control System (SFRCS) channels on low steam generator level, loss of four reactor coolant pumps or high main feedwater differential pressure. The SFRCS actuation logic for the valves is explained in Table 3-1.

Individual valves would subsequently close on a low pressure signal at the turbine inlet, a low pressure signal at the AW pump suction, or a low pressare signal fron one steam generator. The I

turbine contains a trip throttle valve which closes on a turbine overspeed signal. A turbine governor valve is used to control turbine speed.

It is controlled automatically or manually from the control room through a DC powered motor.

The exhausts from both turbines come together and are vented through a common silencer.

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Report No. 02-1040-1095 Revision 1 Page 11 The AW pumps are self-cooled and have minimum flow protection through a normally open recirculation line.

In addition, I

there is a normally closed test line connected to the pump discharge. Steam generator level is controlled at low steam generator pressures through the closing of the motor operated pump discharge valves. These valves are AC powered and initially closed. Additional motor operated valves downstream of the pump discharge direct the auxiliary feedwater flow to the steam generators. These valves are AC powered, are I

initially closed, and receive open/close signals from SFRCS.

Normally, A N pump 1 would supply the water to steam generator 1 and AFW pump 2 would supply water to steam generator 2.

In the event of a steam generator isolation, crossover paths are available so that both pumps would supply water to the remaining active steam generator. The motor operated valves I

at the stecm generator auxiliary feedwater inlet nozzles are normally open, and would only close on a steam generator low pressure isolation signal.

Prior to the TMI-2 event, no procedures existed for using the main feedwater startup pump, in conjunction with the " feed and bleed" procedure, as an alternative method for fulfilling the I

AFWS safety function. As a result, no credit has been taken for this backup success path in evaluating the Pre-TMI AFWS configuration.

3.3 Post-TMI System Configuration The Post-TMI configuration represents the originally planned y

configuration of the AFWS at the end of the 1982 zafueling outage.

It incorporates a number of design improvements over the Pre-TMI configuration. Flow diagrams for the Post-TMI AFUS and main steam configurations are shown in Figures 3-3 I

and 3-4.

Major differences between the Pre-TMI and Post-TMI configurations are:

I 1.

The Post-TMI configuration has diverse electric power sources for motor operated valves.

Certain valves on train 1 'AF-360, AF3870 and the main stream turbine admission valve MS-106) are powered off DC power I

suppliis.

The remainder are AC powered.

2.

The turbine exhausts are redundant and seismically qualified. The plugging of the exhaust pipe / silencer is no longer a common cause failure for both AFWS trains.

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1 Report No. 02-1040-1095 Revision 1 Page 12 l

3.

Administrative procedures have been implemented to lock in position all manual valves and local control stations and hand wheels for motor operated valves in the auxiliary feedwater supply paths, the recirculation line, the test line and main steam supply paths. This reduces the probability for human error in misaligning remotely I

operated and remotely indicated manual valves.

4.

The turbine admission valves now have automatic dual level control, with the option for manual control.

5.

An emergency procedure has been implemented to manually start and align the main feedwater startup pump to provide feedwater to the steam generators in the event that both trains of the ANS fail. This procedure includes the feed and bleed procedure for relieving fluid I

through the PORV while maintaining makeup flow to the reactor coolant system. This procedure effectively provides a diverse and redundant third train of AWS.

The feedwater startup train consists of a single AC powered pump, which is supplied from three water sources, and which discharges to either steam generator.

The water sources are, firat, two deaerator storage tanks and, secondly, the CST.

The fire protection system is available as a backup water supply should these two sources fail.

To initiate the startup train the operator performs the following operations:

block the SFRCS signal and open either, but not both, of the main feedwater stop valves W-601 or N-612 1

(operation performed from the control room),

block the SFRCS signal and open either, but not both, of the main feedwater startup control valves SP-7A or SP-7B (operation performed from the control room),

manually open the startup pump discharge valve m-106 (operation performed locally),

manually start the startup pump (operation performed from the control room).

Report No. 02-1040-1095 Revision 1 Page 13 I

A flow diagram for the startup pump is shown in Figure 3-5.

I In addition to the startup pump the operator must initiate the feed and bleed operation. This consists of manually opening a PORV and its block valve, and operating the reactor coolant system makeup pumps. The PORV and block valves are controlled I

from the control room.

Normally, the PORV is aligned closed and the block valve aligned open. The makeup system is illustrated in Figure 3-6.

The system consists of two trains of pumps discharging through a common pipe to the reactor coolant system. The makeup water tank provides a water supply of 4,480 gallons, after which the water supply is automatically switched to the borated water storage tank.

A I

motor operated valve provides the switchover function on a low level signal from the makeup tank.

In normal operation, one train of makeup is assumed to be operating at all times.

In the feed and bleed mode of operation, however, the normally open reactor coolant system letdown line must be isolated to prevent additional loss of primary coolant inventory. The I

isolation of the letdown line involves manually closing a motor operated valve from the control room, a routine procedure with all reactor trip conditions. The feed and bleed procedure in the Post-TMI configuration applies only to 1

I situations in which offsite electrical power is available at the plant.

It is not credited for initiating events in which offsite power is assumed to be lost.

3.4 Third Train System Configuration The third train configuration examined in this study consists I

of an independent, manually initiated train of auxiliary f eedwater in parallel with the two present AFWS trains.

Manual initiation is required so as to prevent excessive I

feedwater flow in the anticipated event that the two safety grade steam-driven auxiliary feedwater trains function as designed. The third train would be started only if both of the steam driven trains failed.

The third train flow diagram is shown in Figure 3-7.

The train consists of a single AC powered motor driven pump, supplied from three water sources, discharging into either of the steam generator auxiliary feedwater inlet nozzles. The pump is considered to be the main f eedwater startup pump, upgraded in flow capacity such that the feed and bleed operation is unnecessary. The time requirement for initiating auxiliary feedwater via the third train is 10 minutes from the initiating event.

The water supplies would be the same as for I

the present startup pump.

However, the discharge piping would be rerouted to bypass the feedwater heaters and discharge directly into the A WS steam generator inlet nozzles. Either of two AC powered

Report No. 02-1040-1095 I

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motor operated valves, normally isolating the train from the steam generators, wquid be manually opened from the control room when the pump in started.

In all other respects the third train configuration is I

identical to the Post-T!E configuration.

3.5 Analysis-Based Configuratio!.

This configuration is based on results of the reliability analysis of the Post-TMI configuration. The Post-TMI configuration is found to be most susceptible to failures of I

motor operated valves (MOV) to open/close on demand and to the inability to implement the feed and bleed procedure following loss of offsite power events. The analysis-based I

configuration represents the presently planned AFWS configuration at the end of the 1982 refueling outage. It incorporates several design modifications as well as improvements to the feed and bleed procedure. These I

y additional system modifications include the following:

1.

The speed switch control for the pump discharge MOVs AF-360 and AF-388 is eliminated and the valves are normally aligned and locked open.

2.

The MOVs AF-3870 and AF-3872 are normally aligned and I

locked open.

3.

All four turbine steam admission valves, including the I

valves in the crossover paths, open on an SFRCS signal.

In this case, both turbines are supplied with steam from both steam generators through parallel paths.

In the j

event of a steam generator isolation due to low steam i

generator pressure the isolated steam generator discharge I

valves close and the steam supply system to the turbine is identical to the Post-TMI configuration.

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4.

Flow indication is temporarily installed in both AFW pump l

minimum recirculation lines during surveillance testing.

l This permits flow testing of pumps to be performed without opening valves AF-21 and AF-23 (for pump No. 1) j or AF-22 and AF-23 (for pump No. 2).

Thus, an auxiliary feedwater pump remains available if the AFNS is I

challenged during a surveillance test.

e 5.

The startup pump discharge valve FW-106 is locked open and a check valve placed between the pump and E4-106.

6.

The startup pump bypass valve FW-102 is locked closed.

Report No. 02-1040-1095 Revision 1 Page 15 7.

Redundant steam generator pressure and level control room indicators are provided for each steam generator. The startup pump feedwater flow indication is upgraded.

8 The makeup system valve MU-33 and startup feedwater control valves SP-7A and SP-7B cre controlled from the station nitrogen system or local nitrogen bottles and are therefore available following a loss of offsite power.

I 9.

The original feed and bleed procedure is modified to better reflect the steps necessary to implement the feed 1

and bleed operation.

Improved descriptions of various parameter responses enhances the ability to recover from incorrect operator actions. The revised procedure format will be similar to the Abnormal Transient Operating Guidelines emergency procedures.

10.

The feed and bleed procedure is extended to the situation in which offsite electrical power is unavailable at the plant site.

3.6 AEWS Support Systems For the purpose of this study the makeup system, PORV and main feedwater startup pump train are considered part of the AFWS, since they directly support the AFWS safety function. Other plant systems indirectly support the AFWS as well.

These include:

electric power system I

SFRCS service water system fire protection system station nitrogen system.

1 Of these support systems, the reliability analysis results are only impacted significantly by the electric power system.

The electric power system is, therefore, considered explicitly in the reliability analysis. The impacts of other systems are conservatively estimated (as discussed in section 4.3.1) and found to be generally insignificant.

The importance of the electric power system is based upon this system providing the electric power for valves, motors and I

pumps throughout the AFWS.

These power supplies can be categorized as:

powered from essential AC-buses powered from non-essential AC-buses powered from DC panels

Report No. 02-1040-1095 Revision 1 Page 16 I

The DC panels are normally powered by battery chargers powered from essential AC buses.

In the event of a bus failure, however, the DC panels are backed up by battery power supplies. As a result, the DC panels have relatively high reliability. The essential AC buses are powered by the turbine generator (through the auxiliary startup transformer),

offsite power sources or a diesel generator.

For events challenging the AFWS it is assumed that a turbine I

generator trip has occurred and that this power source is unavailable.

For events in which offsite power is assumed to be lost, the essential buses must, therefore, be powered from I

a diesel generator. One diesel generator powers the "Cl" bus while the other powers the "Dl" bus.

The startup pump is powered from bus D2 which can be powered from either diesel generator.

In the Post-TMI configuration, selected non-essential buses can be fed from the diesel generators through operator action (control room operation).

For other non-essential buses the power supply is limited to offsite power following a turbine generator trip.

Table 3-2 lists the interfaces between the electrical power I

system and the AFWS and shows the ultimate power supplies to individual AFWS components.

Major differences between the Pre-TMI configurations are:

1.

In the Pre-THI configuration a ground fault on any of the Essential Motor Control Centers would cause a loss of one of the two redundant electrical systems. This has been modified in the Post-TMI configuration with the installation of ground-fault detectors to trip the individual breakers on all loads attached to an essential bus.

(Note that this is not a TMI-related plant modification, but was undertaken by TECo to upgrade the reliability of plant electrical systems. )

2.

Davis-Besse has an automatic switching system that changes the plant's electric source from onsite power (main generator) to offsite power in the event of a turbine trip.

There is a 30 second time-delay between the turbine trip and the generator trip. When the generator trips there is automatic transfer of the plant's electrical source from the auxiliary transformer to the startup transformer.

In the Pre-TMI configuration there was a possibility that the generator 345 KV breakers could be manually opened before the 30 see time delay and thereby fault the entire switching system by not allowing it to switch to offsite power.

Procedures

I l

Report No. 02-1040-1095 Revision 1 Page 17 have been added to insure that there are no actions done until the 30 see time delay and automatic switching is completed. Also, if the 345 KV breakers are opened for I

any reason other than a degraded offsite power source, a fast dead transfer to the offsite source will occur in the Post-MI configuration.

3.

In the Pre-TMI configuration there was no way to know if there was a ground fault on one of the D.C. MCC Essential Buses. The Post-MI configuration includes a load fault detection. system to correct that situation.

(This change is not being planned for completion in the 1982 refueling i

outage, but will be completed later.)

I I

I I

I

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R; port No. 02-1040-1095 R: vision 1 TABLE 3-2 Page 19 AFWS/ Electric Power System Interfaces l Electric Power l Power Supply l Ultimate AEUS Component l

Supply l

Type l

Source 1

l l

AF-360 (Pre-TMI) lMCC-EllE l Essential AC l Diesel Generator #1 AF-360 (Post-TMI) lDC Panel DlP l Essential DC l Battery IP I

AF-3870 (Pre-TMI) lMCC-EllD l Essential AC l Diesel Generator #1 AF-3870 (Post-TMI) lDC Panel DlP l Essential DC l Battery IP AF-3869

[MCC-EllE l Essential AC l Diesel Generator #1 AF-388 lMCC-F12A l Essential AC l Diesel Generator #2 AF-3872

! ACC-F12B l Essential AC l Diesel Generator #2 AF-3871 lMCC-F12A l Essential AC l Diesel Generator #2 SW-1382 lMCC-E12A l Essential AC l Diesel Generator #1 SW-1383 lMCC-FilC l Essential AC l Diesel Generator #2 MS-106 (Pre-TMI) lMCC-EllC l Essential AC l Diesel Generator #1 MS-106 (Post-TMI) lDC Panel DlU l Essential DC l Battery 1N MS-106A lMCC-E12B l Essential AC l Diesel Generator #1 MS-107 lMCC-Fila l Essential AC l Diesel Generator #2 MS-107A lMCC-FilB l Essential AC l Diesel Generator #2 ICS-38A l DC Panel D2P l Essential DC l Battery 2P I

ICS-38B l DC Panel DlP l Essential DC l Battery IP AV-1 lMCC-F13 lNon-Essential AClDiesel Generator #1 & 2 AV-3 lMCC-F13 lNon-Essential ACl Diesel Generator #1 & 2 Startup Pump l Bus D2 lNon-Essential ACl Diesel Gcnerator #1 & 2 Makeup Pump #1 l Bus Cl l Essential AC l Diesel Generator #1 Makeup Pump #2 l Bus D1 l Essential AC l Diesel Generator #2 FW-601 lMCC-FilD l Essential AC l Diesel Generator #2 FW-612 lMCC-EllC l Essential AC l Diesel Generator #1 MU 3971 lMCC-EllD l Essential AC l Diesel Generator #1 FW-786 lMCC-EllD l Essential AC IDiesel Generator #1 FN-790 lMCC-F12A jEssential AC l Diesel Generator #2 FW-460 lMCC-F3 2A l Non-Essential AClOffsite Power MU-2B lMCC-EllB l Essential AC l Diesel Generator #1 PORV lDC Panel DBP l Essential DC l Battery 2P y

Block Valve lMCC-E16 l Essential AC l Diesel Generator #1 I

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I Report No. 02-1040-1095 I

Revision 1 Page 29 4.0 Methodology The reliability analysis is based on the development of fault trees for the AFWS and its support systems.

Probabilities for basic events appearing in the fault trees are assigned based I

on reviews of industry reliability data sources and Davis-Besse plant-specific experience. A Boolean manipulation computer code is then used to determine the AFWS unavailability.

The unavailability of each AFWS configuration I

is determined in this manner.

The AFWS unavailability is dependent on the plant initiating I

event which causes the AFWS to be challenged. The relative differences in AFWS unavailability for the four configurations are also dependent on the specific initiating event.

In order to develop an overall figure-of-merit for each AFWS configuration, the annual f requencies for various initiating events are estimated, again based upon industry and plant-specific operating experience. The frequency of the initiating event is multiplied by the AFWS unavailability for that initiating event.

The results are summed for all initiating events. The result is the overall annual frequency with which the AFWS is unavailable when challenged. This is I

the overall figure-of-merit for each AFWS configuration.

The analysis can, therefore, be descritad in the following phases:

system fault tree development I

data analysis system unavailability analysis initiating event analysis combined system / event analysis The interrelationship among these analysis phases is illustrated in Figure 4-1.

4.1 System Fault Tree Development 4.1.1 Fault Trees Developed Fault trees are constructed for four AFWS configurations:

l1 Pre-E!I Configuration Post-D1I Configuration Third-Train Configuration Analysis-Based Configuration l1 l11

Report No. 02-1040-1095 Revision 1 Page 30 The fault trees contain as the " top event", the failure of the AFWS to perform its safety function. The specific safety functions are described in Section 3.1.

In addition to the fault trees for each AFWS, subtrees are developed iar other plant systems which support the AFWS safety function. These I

subtrees are limited to only those parts of the support systems which directly affect the AFWS safety function.

Support system subtrees which are developed are:

electric power system main steam system main feedwater system and startup pump makeup and purification system power operated relief valve The attachments include fault trees and subtrees arranged as follows:

AWS ( Pre-24I) Fault Tree Main Steam System (Pre-TMI) Fault Tree Electrical System (Pre-TMI) Fault Tree AWS (Post-O(I) Fault Tree Main Steam System (Post-21I) Fault Tree Electrical System (Post-24I) Fault Tree Start-up Pump (Post-TMI) Fault Tree Start-up Pump with Feed and Bleed I

AFWS (Analysis-Based) Fault Tree Main Steam System (Analysis-Based) Fault Tree 1

Startup Pump with Feed and Bleed ( Analysis-Based)

Subtrees for other systems supporting the AFWS safety function are not developed. Such systems include:

fire protection system service water system

- SFRCS

- station nitrogen system (1

The reliability of these systems is found to have a lesser impact on the achievement of the AFWS safety function.

I Conservative estimates of the unavailability of these systems are used in the quantitative analysis of the AFWS fault trees as discussed in Section 4.3.1.

Report No. 02-1040-1095 Revision 1 Page 31 4.1.2 Fault Tree Methodology For each system f ault tree or subtree, the safety function of the system is first defined, and failure criteria applied. An absolute determination for failure criteria is made for I

systems where a reduction in capacity leads to failure or unavailability of that system. A list of systems, safety functions and failure criteria assumed in this reliability study is presented in Table 4-1.

Note that a 10 minute AFWS I

actuation criteria has been arbitrarily assumed, for conservatism, in this analysis.

I Detailed fault trees, or subtrees, are then constructed using the methodology and symbology of WASH-1400(5) and IEEE-352(6). The construction of the detailed fault trees, or subtrees, is done in a rigorous, systematic manner, I

considering every component and event which could contribute to the failure of the system. Quantitative judgements about the likelihood of failure of a compon.nt are not made during I

the detailed fault tree construction phase. The following mechanisms for failure are included in the fault trees:

Pre-Existing Faults

- outages for test and maintenance I

- demand faults for initially inactive components

- failure mechanisms which are dependent on the duration of the standby period, and which will cause failure on demand for initially inactive components

- pre-existing human errors, e.g.: maintenance faults Faults Occurring During Mission

- failure of an active component to change its state, e.g.; a demand fault

- failure of a component to continue operating I

- human errors of omission human errors of commission Only pre-existing faults which would not be detected in normal I

plant operation are included, e.g.

pre-existing faults in active systems (other than test and maintenance outages) are not considered.

I I

Fepcrt No. 02-1040-1095 Revision 1 Page 32 In developing detailed system fault trees, single passive failures and double active failures are considered within a single process flow path (e.g.

within a single train of t multiple train system). A single failure is the failure of one element within a process flow path which causes the failure of the required flow path function. A double failure is the combined failure (either random failure or dependent failure) of two elements within a process flow path which causes the failure of the flow path function. A passive failure is breach of a fluid pressure boundary or blockage of a process flow path in a fluid systems or short circuit, loss of electrical charge or loss of ability to conduct electricity due to physical defects in electrical systems.

An active failure is a malfunction, excluding passive failures, of a component which relies on movement to complete its intended function upon demand.

Examples of active failures include the I'

failure of a powered valve to move to its correct pcsition; or the failure of a pump, fan, or diesel generator to starts or f ailure of a circuit breaker, relay, or solenoid to change I

position when energized. Human errors (acts of commission or omission) are considered active failures.

In constructing fault trees the following rules are applies:

- The fault tree is developed to the level where acceptable failure data exist.

- Components and basic events are coded using an eight character nomenclature as shown belew:

A Mi 001A F

l I

i i

System Code l

l l

l Failure Mode (See Table---l l

l l

Code (See 4-2) l l

l--Table 4-4) l I

Component Code l

l Descriptive

( See Table 4-3 )-----l l---Nomenclature Parts of the fault tree which are only applicable to I

specific plant conditions (e.g.; loss of offsite power events) are combined through a gate with the " house" logic symbol.

The symbology shown in Figure 4-2 is used.

In addition, the following assumptions are made in developing I

fault trees:

I

I Report No. 02-1040-1095 I

Revision 1 Page 33

- The plant is assumed to be in a normal operating condition at 100% power at the time of the initiating event.

Pre-existing faults in active plant systems (e.g.; one train of the makeup and purification system) are not considered, since such faults, if present, would have been readily detected and corrected.

I

- Operator action to recover from a faulted condition is only credited when the operator has sufficient instrumentation to detect the fault and a written procedure directing his recovery action.

The probability for failure to take the recovery action is discussed in Section 4.2.

- Component alignments, as shown on plant P& ids and electrical drawings, are assumed for the initial plant I

configuration.

However, the possibility of misalignment is considered when such misalignment would contribute to system failure, and when such misalignment might not be detected in normal plant operation.

- Spurious human acts of commission, such as taking an incorrect action when there is no indication that action is I

required, or acts of sabotage are not considers 4.

4.2 Data Analfsis I

Data on the probabilities for failures or unavailabilities of basic events are necessary for quantification of the fault trees. Such data cons!.st of system and component failure data and human error data.

Industry data sources and Davis-Besse plant-specific operating experience have been reviewed to develop a recommended data base for this analysis.

4.2.1 Industry Data Review Sources for failure rate data are listed in Section 7.0.

These sources contain summaries of recent nuclear power industry experience for electrical and mechanical components generic to the industry.

In some cases, the data is I

supplemented by experience with similar components in other industry applications. A major data source for nuclear power industry component reliability is the Reactor Safety Study, WASu-1400(5). This study contains reliability data on most components found in nuclear power plant safety systems. It is based on compilations of many reliability data sources avaialble at the time of the study, in 1975.

Several more recent NRC-sponsored data summaries (7,8,9) document the reliability experience of common nuclear power plant components, specifically valves, pumps and diesel generators.

l

Report No. 02-1040-1095 Revision 1 Page 34 These summaries are based upon Licensee Event Reports (LERs) through which safety system malfunctions are reported to the NRC.

IEEE-500(10) represents a thorough compilation of electrical component reliability data.

Reference (13) contains summaries of nuclear power plant equipment I

mallunctions as reported through the Nuclear Plant Reliability Lata System (NPRDS).

Failure data from these sources have been reviewed and tabulated in Table 4-5.

The table lists the recommended value for failures (expressed either as a failure rate - units of inverse time, or as a f ailure probability - dimensionless).

Note that the recommended value is not necessarily the mean value of the data sources reviewed; in fact, the recommended values are generally the highest of the reported values.

Also listed, when available, is the uncertainty factor representing a measure of the spread in the reported data.

The uncertainty factor is defined as the square-root of the ratio of the maximum reported value divided by the minimum reported value.

The uncertainty factors are rounded to the nearest half decade. Where only one data source is given an uncertainty factor of 10 is assumed unless the data source reported the data spread. The data in the table are presented by component I

type, e.g.; motor operated valves, and by failure mechanism, e.g.; failure to open on demand.

Human reliability data consist of human errors of commission and omission.

Errors of omission include omitting steps from written plant procedures during routine operations (e.g.;

maintenance), during emergency operations (emergency I

procedures) and during attempts to recover from a faulted condition.

Errors of commission similarly include those committed during routine operation and those committed during I

the course of the accident in attempting to mitigate the accident. The primary sources for estimatin the probabilities of human errors are WASH-1400( ) and a recent Ill) of human reliability in nuclear NRC-sponsored study power plant operations.

Table 4-5 also lists f ailure probabilities, and uncertainty I

factors, associated with various types of human errors. The values listed are from the above two sources.

4.2.2 Review of Davis-Besse Experience Where possible, the generic data sources have been

,g supplemented by analysis of failures experienced at g

Davis-Besse, Unit No. 1.

This analysis is based upon a review of LERs for Davis-Besse from the time it began commercial operation until February, 1981.

Due to the rather limited I

Report No. 02-1040-1095 Revision 1 Page 35 I

data base, this analysis concentrated on components and failure mechanisms which occur relatively more frequently in nuclear power plants and which could have a more significant impact on the AFWS reliability analysis results. The components included in this plant-specific analysis are:

- valves

- auxiliary feedwater pumps

- diesel generators

- test and maintenance outages

- human factors analysis for the feed and bleed operation, valves Failure rates for Davis-Besse motor operated valves, air operated valves and check valves have been determined.

Failure mechanisms are failure to open on demand and leakage (for check valves only).

Failure rates arn computed by dividing the total number of f ailures reported in the LERs by the total number of valve demands (for failure to open/close on demand) or the total number of operating valve-hours (for leakage).

LER's are limited in terms of the plant systems in which f ailures are reported.

In this analysis, the following six I

plant safety systems are considered:

auxiliary feedwater system I

- main steam system

- containment spray system high pressure injection system low pressure injection system chemical volume control system Five of these systems were considered by the NRC in the I

development of their generic data base (Reference 7).

The sixth system, main steam, was considered in this analysis since it is directly pertinent to this program.

Table 4-6 summarizes results of the analysis. The total population of valves in the system is listed along with the total valve demands, total operating hourt and total valve failures.

In computing the number of demands placed on valve s, it is assumed that valves are only operated during testing and that the minimum testing schedule contained in the Davis-Besse Technical Specifications (12) is used.

The I

resultant failure rate is thereby conservatively estimated.

While this estimate is conservative, the procedure used is consistent with that used in the data analysis of all U.S.

I operating reactors considered in Reference (7).

I Report No. 02-1040-1095 Revision 1 Page 36 The failure of Davis-Besse motor operated valves to open/close on demand, computed in this manner, is a factor of three greater than the same failure probability computed for all operating U.S.

reactors, as reported in Reference (7).

The Davis-Besse plant-specific value is used in the quantitative analysis of the AFWS fault tree.

This failure mechanism turns I

out to be a major contributor to AFWS unavailability, as discussed in Section 5.3.

Most of the motor operated valve failures are attributable to torque switches and limit switches being out of adjustment.

Auxiliary Feedwater Pumps This analysis includes failures of the auxiliary feedwater pumps and/or turbine to start and to continue operation.

The probability of f ailure to start is computed by dividing the I

total number of reported failures by the total number of attempts to start the pumps.

It is assumed that each pump is only started for monthly testing, and that there is one demand of each pump per test.

Failures to continue operation are generally attributable to faults occuring during its standby period. The failure I

probability is calculated from the total reported failures, divided by the total standby hours for the pumps. Results are shown in Table 4-7.

During the period covered by the data reviews, there are a total of three demand failures of the AFWS pump / turbine. All three of these failures occured during the first year of I

commercial operation of the plant.

Faulty speed control relays were the primary cause of the failures and the relays were replaced with relays capable of operating under design conditions. No subsequent failures of this type have occured since 1977.

Since this type of failure appears to be associated with the plant " burn in" period, it is felt that generic industry failure probabilities are more appropriate to be used for analyzing the plant in its present phase of operation. There has recently been a fourth demand failure of the AFRS pump / turbine.

It's cause is unrelated to the earlier reported f ailures and, while it is not included in this data review, it would not alter the conclusion that the Davis-Besse AFWS pump / turbine demand faults are consistent with reported industry average failure data.

The Davis-Besse experience in failure of turbine driven pumps to continue operation is in agreement with the generic data reported in Reference (8), so again, the generic data are used in the reliability analysis.

I Report No. 02-1040-1095 I

Revision 1 Page 37 I

Diesel Generator Diesel generator failures reported in the Davis-Besse LERs can be categorized as failure to start, failure to stabilize and failure to continue operating.

The failure to start includes I

actual failures of the diesel generator to start on demand.

In computing a failure probability, only demands during monthly testing of the diesel generators are considered. This results in a conservatively high estimate of the demand I

failure rate since other diesel generator demands (e.g.;

demands imposed by other system tests) have not been included in the calculation, although any failure occurring during such I

demands are included. Failure to stabilize includes faults which prevent the diesel generator from operating for more than a very. short time af ter starting. These failures are generally due to pre-existing faults occurring during the standby period. The failure rate for this mechanism is computed from the total standby hours for the diesel generators.

Failure to continue operating includes faults occurring as an actual result of running the diesel generators. The failure rate for this mechanism is computed from the total operating hours logged for the diesel generators. Table 4-8 summarizes the Davis-Besse diesel generator reliability experience.

In the analysis of the electric power system fault tree a probability of failure to start and stabilize is computed.

This probability is the sum of the probability for failure to start on demand and the probability for failure to stabilize, which is calculated by multiplying the failure to stabilize failure rate by one-half the mean test interval.

In general, the Davis-Besse diesel generators appear to have I

experienced slightly higher f ailure rates than the reported industry averages contained in Reference (9).

Many of the diesel generator failures have occurred as a result of faults I

in the turbochargers.

TECo plans to improve the diesel generator reliability by installing new high capacity turbochargers and modifying the lube oil system for the turbochargers. While these changes are planned for tne 1982 outage and should significantly improve the diesel generator reliability, their quantitative impact on the reliability is not known.

Therefore, the higher failure probabilities I

computed from past Davis-Besse experience are used in this reliability analysis.

I Report No. 02-1040-1095 I

Revision 1 Page 38 I

Test and Maintenance Outages Test and maintenance outages for the AWS and diesel generators are computed in the same manner as reported in WASH-1400(5). This calculation is dependent on the plant-specific frequency of testing and the maximum time allowed by the Technical Specifications (12), during which a component can be out for maintenance while the plant is in operation.

The unavailability for a component being in maintenance ist fm tm 720 Qm

=

where f is the frequency per month at which maintenance I

m is performed and t is the average outage time per m

maintenance act (expressed in hours).

The unavailability for testing is ft tt 720 Qt

=

where f is the testing frequency per month and t is t

t the average time per test (expressed in hours).

For the I

AWS and diesel generators, the Davis-Besse Technical Specifications (12) limit craintenance outages to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before the plant must be shutdown and specify monthly test intervals. An hourly test duration is assumed.

The I

frequency of maintenance acts is taken as.22 acts / month, and the mean duration of the maintenance is taken as 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> for the AWS and 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> for the diesel generators.

The maintenance values are developed in WASH-1400, Appendix III, Section 5(5),

Human Factors Analysis The human factors probabilities used in the fault tree analysis for the feed and bleed operation are computed using NUREG/CR-1278.

In each instance, the operator action contained in the procedure is analyzed and compared to specific events in NUREG/CR-1278.

It was necessary in many cases to make assumptions concerning the operator system, I

since perfectly analogous examples do not exist.

These assumptions are documented in this calculation.

A typical list of assumptions is as follows:

1. Remotely operated and locally operated valves are treated under the generic category of " Manual Valves" I

Report No. 02-1040-1095 Revision 1 Page 39 I

2.

No recovery f rom error is assumed unless a specific 1

control room annunicator is available.

3.

Operator errors are assumed to be consietent with i

populational stereotypes.

For the purpose of fault tree analysis, one number, representing the probability of failure for that operator action, is indicated. It must be noted, however, that this I

single entry is a composite of unique errors that, when combined, form the operator error probability shown in the fault tree diagram. For example, the f ault tree entry I

" Operator Fails to Open Valve" consists of the following components:

1.

Operator omits step in written procedure or 2.

Operator selects wrong valve from grouped system or 1

3.

Operator operates valve incorrectly multiplied br 4.

Operator stress factor (moderate) 4.2.3 Recommended Data Base The results of the plant-specific data evaluation and the I

review of generic data are presented in Table 4-5.

Also listed are recommended values for use in this reliability study.

The recommended values are generally based on the following prioritization.

Whenever possible, plant-specific data are used.

The highest value of recent generic data sources, (References 7, 8,

9,

13) is used.

The human reliability data of Reference (11) are used, since this represents an expansien of the earlier work reported in Reference ( 5).

Also shown in the table are uncertainty factors on the data.

These are determined by taking the maximum variance of the tabulated data sources. Some data points are discarded if they vary from the mean value by more than a factor of 100 (In all cases, such values are smaller than the mean value so that, in no case, are reported high f ailure rates discarded).

Such values are not considered in determiaing maximum variances. The uncertainty factor is then rounded to the nearest half decade.

If only one data source is available, an uncertainty factor of 10 is assigned.

l t

t

I Report No. 02-1040-1095 Revision 1 Page 40 4.3 System Unavailability Analysis The system unavailability analysis includes quantitative analysis of the fault trees, the uncertainty analysis for AFWS unavailability, and the importance ranking of fault tree basic events in contributing to AFUS unavailability.

4.3.1 Quantitative Analysis of Fault Trees Each AFWS configuration fault tree is analyzed for each category of initiating event.

This analysis results in the qualitative determinatica of minimal cut sets for the f ault I

tree and the quantitative determination for the point estimate for the probability of the top event. All support system fault trees, except that for electric power, are evaluated as part of the overall AFUS fault tree.

The electric power system fault tree is evaluated separately.

Probabilities for failure of the electric power interfaces with the A WS fault tree (see Table 3-1) are computed separatcly and values inserted into the AFUS fault tree.

In cases where the dominant failure mode for separate electric I

power supplies is actually a common failure, these interfaces are treated as the same basic event in the AFWS f ault tree.

For example, with loss of offsite power, the dominant failure mode for failure of MCC-E12A and MCC-E12B is the failure of diesel generator #1 to start and continue running. This is treated as a single event wherever MCC-E12A and MCC-E12B interf ace with the AFWS f ault tree.

The WAMCUT computer code (14) is used in the fault tree analysis.

This is a Boolean manipulation computer code which deterrmines the probability of occurrence of the top event I

(and any specified intermediate events) in the fault tree.

It also identifies the minimal cut sets of the fault tree.

I Since the AIUS fault trees developed in this study are very detailed, many thousands of minimal cut sets exist.

In order to lunit computer running time and to avoid exceeding the capacity of the code, the code has an input minimum probability cutoff. Any cut set whose probability is less than the cutoff value is discarded from the calculations. So as not to eliminate any potentially significant cut sets, this I

minimum probability cutoff is generally selected to be three orders of magnitude (1000 times) less than the probability of the top event of the tree.

In a few cases, excessive computer time requirements dictate a minimum probability cutoff of not less than 500 times smaller that the top event probability.

Report No. 02-1040-1095 Revision 1 Page 41 Mean values for basic event failure probabilities are input to the WAMCUT code.

These are assigned from the recommended data column of Table 4-5.

Where failure rates are given in this table (units of inverse time) the failure rate is multiplied by either one-half the mean test interval or the mission time, as appropriate.

For all initiating events, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AFWS I

mission time is assumed. This is based on a conservative estimate of the time required for auxiliary feedwater prior to setting the decay heat removal system in operation. Also, the AFWS reliability analysis is based on the assumption of non-repairable component failures (except for human recovery actions not actually requiring repair of the fault). Twenty four hours after a plant initiating event, repair of components in plant safety systems would almost certainly be initiated.

For operation of the diesel generators, a ten hour mission time is assumed. This is based on the review of offsite electric power restoration experience reported in Reference (5).

Once offsite power is restored, the diesel generators would no longer be required as an electric power source.

For certain categories of events, various components may have I

an unavailability of unity.

For events involving loss of offsite power, all components not powered off the diesel generators have an unavailability of unity. Also, for seismic events, all non-seismically qualified equipment is assumed to be unavailable. One exception to this general assumption is the Pre-TMI configuration in which the common turbine exhaust silencer is not seismically qualified. This is a potential common failure mode for both AFWS trains.

However, the failure mode is not the rupture of the silencer and exhaust pipe, but its becoming plugged to an extent that steam cannot be exhausted. A probability of.01 (uncertainty factor of I

ten) is arbitrarily assigned to this event.

For failures of support systems for which explicit fault trees are not developed, the following order-of-magnitude unavailabilities are assigned:

System Unavailability fluid systems (service water

.01 system, fire protection system) a specific single channel of SFRCS

.001 ll station nitrogen supply to any 9.6 x 10-5 lg single valve (involves passive failures only) l11 lIl

Report No. 02-1040-1095 Revision 1 Page 42 I

These values are based on judgments formed from various reliability studies on fluid systems and safety grade I

electrical control systems.

In all cases, an uncertainty factor of ten is applied to these unavailabilities.

It should also be noted that the fire protection system has an unavailability of unity for loss of offsite power events and seismic events.

I 4.3.2 Uncertainty Analysis The standard deviation of the fault tree top event unavailability, due to data uncertainties, is determined I

through a mcments calculation. First moments (mean values) and second moments for the fault tree basic event probability distribution function are input to the WAMCUT code. The code computes the resultant top event Tirst moment (point value) and second moment. The standard caviation of the top event, is then computed from the relatioc4 hip 2

M2 - M1 0

where M2 is the top event second moment and My is the top event first moment. The assumed form for the probability distribution functions of basic events is a log-uniform distribution. The second moment calculation, in conjunction with the large nunber of events contributing to the top event, tends to make the top event standard deviation insensitive to the assumed probability distribution fnnction. The limits of the distribution are the mean value multiplied and divided by the uncertainty f actors listed in Table 4-5.

since the recommended values are greater than the true mean values (which could be compiled from the various data sources), this procedure tends to bias the standard deviation tcwards higher unavailabilities. The true + 7 value reported here is, therefore, too large while the - c value may actually be somewhat larger.

However, the intent here is not to develop I

absolute confidence limits, but rather to evaluate relative changes in AFWS unavailability and to develop a qualitative measure of the uncertainty in the results.

Since the same bias is used in all cases, the results can be compared in relative terms.

4.3.3 Importance Ranking The importance ranking is used to judge the relative significance of basic events in contributing to the unavailability of the AFWS.

Events with a high importance measure are more significant in contributing to AFWS failure.

Such a ranking is most useful in evaluating various means to improve AFWS availability.

Report No. 02-1040-1095 Revision 1 Page 43 There are several measures ef importance used in reliability studies. The measure employed here is the Fussell - Vesely measure (4).

This measure, applied to a single basic event, is defined as the total probability for the occurrence of a cut set containing 'he event, divided by the probability for I

the occurrence of the top event.

Basically, this is a measure of the sensitivity of the top event (AWS unavailability) to small changes in the unavailability of a basic event. The Fussell-Vesely measure can be applied to generic categories of I

events, e.g.,

the failure of motor operated valves to open on demand, which is a feature utilized in this study.

The importance rankings in this study are computed through hand calculations, using the minimal cut set identification and probabilities generated by WM4 CUT.

4.4 Initiating Event Analysis Estimates for frequencies of initiating events requiring I

actuation of the AEWS are developed. Initiating events considered in the analysis are listed in Table 4-9.

The events are grouped into three event categories which differ in their assumed availabilities for certain plant components, equipment and systems. These event categories are:

Category 1:

Events in which main feedwater flow is lost, but offsite electrical power and non-seismically qualified equipment are available.

Category 2:

Events in which offsite electrical power is assumed to be lost, but non-seismically qualified equipment is available.

Category 3:

Events in which offsite electrical power and all non-seismically qualified equipment are assumed to be unavailable.

Table 4-9 shows the categorization of initiating events challenging the AFWS.

4.4.1 Frequency Estimates Initiating event frequency estimates are developed from reviews of generic industry sources, and are supplemented f rom reviews of Davis-Besse operating experience. The primary generic data sources are References (5) and (15).

Davis-Besse experience is summarized in L~.ms and unit trip reports.

I Results of these reviews are tabulated in Table 4-10.

Report No. 02-1040-1095 Revision 1 Page 44 Loss of Main Feedwater Reference (15) cites B&W reactors as experiencing this type of transient slightly less frequently than other vendors' PWRs.

There is evidence of a "burnin" period associated with this type of transient, with a 50% increase in the f requency during the first two years of plant operation. The generic PWR frequency listed in Table 4-10 is the frequency af ter this two-year period of operation. The Davis-Besse experience is I

in agreement with the generic PWR f requency, s7 the generic value is used in the reliability analysis.

Steam Generator overfill Davis-Besse has experienced three steam generator overfill e v'e nt s, but none of these actuated the AFWS.

The generic I

value is therefore assumed.

Small Break in RCS Reported events in this category include control rod leakage, primary system (primarily pump seal) leakage, pressurizer I

leakage and opening of the pressurizer safety or relief valve. Davis-Besse has experienced one initiating event of this type, which occurred at less than 10% power during the first month of operation.

This event is not considered to be I

representative of post "burnin" operation, so the generic PWR frequency is assumed.

Loss of Forced RCS Circulation This event includes the loss of all reactor coolant pump forced circulation as tha initiating event.

It does not I

include loss of of f site power at the initiating event, which would also result in loss of forced RCS circulation. The generic data indicates that this type of initiating event is I

relatively infrequent. Davis-Besse has not experienced a complete loss of RCS circulation as an intiating event.

Davis-Besse did, however, experience a partial loss of forced RCS circulation (two loop flow) which resulted in low steam generator level and AFWS actuation. The " initiating event" in this instance is considered to be the partial loss of RCS circulation.

Since the frequency for partial loss of forced RCS flow is expected to be an order of magnitude greater than for tctal loss of flow, and siace a partial loss of flow may result eventually in AFWS actuation, the larger Davis-Besse based frequency is used in this analysis.

Report No. 02-1040-1095 Revision 1 Page 45 Loss of Offsite Power There have been three loss of offsite power events at I

Davis-Besse, however, one occurred during the initial power assentation and a second occurred as a result of power transfer logic which has since been modified. The Davis-Besse experience since this modification was implemented agrees well with the generic experience of all PWR's reported in Reference (15).

The generic value is, therefore, used.

'Ibrnado The frequency estimate for tornadoes is taken from the I

Davis-Besse FSAR.

No incidents have been recorded. This value is in general agreement with the similar frequency estimate of Reference (5), which is presented as a conservative upper bound for the entire Eastern U.S.

It is assumed in this analysis that a tornado would cause a loss of offsite power.

Its impact on the AFWS and other support systems is assumed to be negligible, since the buildings housing support systems are designed to withstand the effects of a tornado.

Eatunquake The only source for this event is the estimate contained in I

Reference (5).

The value cited is the frequen:y of earthquakes in the Eastern U.S. which result in ground accelerations greater than 0.lg.

4.4.2 Factors Influencing Event Frequencies There have been many Davis-Besse design modifications, either implemented since the TMI-2 event or planned to be implemented by the 1982 refueling outage, which may affect not only the AFWS reliability, but also the frequency with which the AFWS may be challenged. Modifications which reduce the frequency of initiating events challenging the AFWS may actually contribute more to the overall plant reliability and the diminishing risk then do modifications intended to upgrade the I

AFWS availability.

Unfortunately, there are generally insufficient or no plant operating data available with which to quanitfy the reduction in challenges to the AFWS resulting from these modifications.

This is true not only for Davis-Besse experience, but also for overall industry experience, since many plants have implemented significant changes in design and operation as a result of the TMI-2 event.

Report No. 02-1040-1095 Revision 1 Page 46 The initiating event frequencies used in this analysis are, therefore, generally based on " Pre-TMI" plant operating I

experience. The frequencies are, however, applied uniformly in developing overall figures-of-merit for all of the AFWS configurations.

(As discussed in Section 4.5, the overall I

figure-of-merit is defined as the annual frequency with which AFWS is unavailable when challenged).

The resultant figures-of-merit for the " Post-TMI", " Third Train" and " Analysis-Based" configurations are, therefore, conservatively high; the figure-of-merit for the " Pre-TMI" configuration contains less conservatism.

Since the same I

conservaitvely high frequencies are applied to each of these configurations, however, the relative differences in the results are indicative of the relative benefits attainable from each configuration.

It should be noted, that significant I

reduccions in the annual frequency with which AFWS is unavailable when challenged, could be attained through modifications to plant design and operations which are not directly linked to the AFWS.

4.5 Combined System / Event Analysis Results of the AFNS reliability analysis are combined with initiating event frequencies in order to derive an overall figure-of-merit for each AEWS configuration. The figure-of-I merit is derived as follows:

A matrix is constructed listing each AFWS configuration and each initiating event, as shown in Figure 4-3.

The frequency for each initiating event is multiplied by the AFWS unavailability for that event.

The products are summed for each AFWS configuration.

The resultant figure-of-merit is the annual frequency with which the AFNS is unavailable when challenged.

i

I R: port No. 02-1040-1095 Revision 1 Page 47 System Safety Functions I

System Safety Function l

Failure Criteria for Study I

AWS (Pre-mI) l Primary ~ system decay heat l1.

less than full capacity flow from l removal l

at least one pump train, or l2.

flow to steam generator (s) delayed I

l l

more than 10 minutes from iniciating l

l event, or l

l3.

all AWS flow is interrupted during l

l required mission time.

I I

I AWS (Post-MI l Primary system decay heat 11.

less than full capacity flow from I

and Analysis-l removal l

at least one pump train, or Based) l l2.

flow to steam generator (s) from AWS l

delayed more than 10 minutes from l

l initiating event, or l

l3.

all AWS flow is interrupted during l

l required mission time.

I LAND l

l1.

full flow from startup rn 0 delayed l

l more than 30 ninutes, or l

l2.

full flow from one makeup pump to l

I primary system delayed more than 30 l

l minutes, or l

13.

letdown line not isolated at reactor l

l trip, or l

l4.

less than full discharge from one l

l PORV within 30 minutes, or l

l5.

feed and bleed procedure is inter-l l

rupted prior to HPI initiation, or l

16.

startup pump flow is interrupted l

l during required mission time.

I l

l l

f w

-e-

I R: port No. 02-1040-1095 Rsvision 1 9

TABLE 4-1 (Cont.)

System Safety Functions System l

Safety Function l

Failure Criteria for Study I

I AWS (Third Train)l Primary system decay heat l1.

less than full capacity flow from l removal l

one AWS pump train, and less than l

l full capacity flow from startup l

l pump, or l

12 flow to steam generator (s) delayed I

l l

more than 10 minutes from initiating l

l event, or l

l3.

all AWS flow including startup pump l

l flow is interrupted during required l

l mission time.

I I

Electric Power Provide AC or DC power to ll.

inability to supply rated load to System lAW S components l

AW S components.

I I

i l

Main Steam System l Provide steam to AWS pump l1.

inability to provide sufficient l turbines l

steam to maintain full AWS pump l

l flow.

I I

i I

I I

I Report No. 02-1040-1095 Revision 1 Page 49 TABLE 4-2 1'lant System Designator A

Auxiliary Feedwater System C

Condensate System E

Electrical System F

Fire Protection System M

Main Steam System S

Service Uater System P

Feedwater System I

I Report No. 02-1040-1095 Revision 1 I

S TABLE 4-3 Page 50 Component Code I

Mechanical Components Diesel DL Valve, Check CV Filter or Strainer FL Valve, Hydraulic Operated HV Flow Element FE Valve, Manual XV Gaa Bottle GB Valve, Motor Operated MV Nozzle NZ Valve, Pneumatic Operated AV I

Orifice OR Valve, Relief RV Fipe PP Valve, Safety SV Pump PM Valve, Solenoid Operated KV Tank TK Valve, Stop Check DV Tubing TG Vent VT Turbine TB I

I

I Report No. 02-1040-1095 Revision 1 TABLE 4-3 (Cont.)

Page 51 Component Code Electrical Components Battery BY Relay RE I

Battery Charger BC Relay or Switch Contact CN Bus BS Switch, Pressure PS Cable CA Switch, Temperature TS Circuit Breaker CB Switch, Torque QS I

Control Switch CS Transformer, Power TR DC Power Supply DC Transmitter, Flow TF Flow Switch FS Transmitter, Level TL 1

I Fuse FU Transmitter, Pressure TP Generator GE Transmitter, Temperature TT Ground Switch GS Wire WR Inverter (solid state)

IV I

Level Switch ES Limit Switen LS Motor MO Motor Starter MS I

J l

t i

i i

II i

)

Report No. 02-1040-1095 i

Revision 1

)

Page 52 TABLE 4-4 1

2 Failure Mode Code Closed C

i j

Does Not Close K

1 m

Does Not Open D

Does not Start A

Exceeds Limit M

Leakage L

l Loss of Function F

1 4

Maintenance Fault Y

i Open O

Open Circuit B

1 Operational Fault X

Plugged P

1 Rupture R

Short Circuit Q

t Short to Ground S

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In tiele case, the luwer end of the NUMEC/CD-127a uncestalnty T el.l e 13-1 amem.ed.

e-a lxand was accosas. ended. primari ty beecause it seemed likely that t aae f

ite l Inn.3 Li st with checkoff.

r eq.si s tement to lock a valve $n place woesid dsaw more attention to O

E7)

Asaus.n o con t s ol a e s e functionally qsonipeJ.

(a) the violat ion of populat ion.a t atescot yg e assie.ed.

It s geosition than almgaly taggini) St.

In addition, the cite.1 median value of 89)

Ai; tion after

3. minutes.

. 001 w.no aans e closely s elated to the f a!!ure t o lock the O

410) (ncestataty tand le anew.e.g.

valve t han to the act ual valve g=>sition.

y 4

I R; port No. 02-1040-1095 Revision 1 I

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E E

E TABIE 4-5 Koeit )

AtwS ELE.l ABILITT 5111DY Fall 3JPE DAT As 6,ltJMAN F AC1U5tS

.. tnicessainty Factor I

i I

I I

I l

I I

I I

I I

I I

t*1 1 *al l **l I **l I **l nuoto 1278 i *al le.co ended I aal I

I I

I I

l_ i 1

i l _._1 I

I i

__l Ili,ta:mieTlow s cinicistar Autest r.!!ure n,-le 1

I I

I I

I I

I I

i 1

1 I

I I trie-.mi. ting saintenance i

i i

i i

i l

i l

i i I = to-2 l 1

,lo-a g3 g i

luros i. not m.e r ve.1 I n l

l l

1 l

l l

l l

l l

1 i

i i

l liuuttne in.pection I

i 1

1, I

I I

t l

l L

i 1

i i I

I 1

1 I _ _I r

l l

l l

i I

l JJ J

.I 1 m to-4 1

1 Ia to-4 Ito i I Annunciated Damplay of an l

l l

I

)

l l

1 1

I l

1 l

l 1

I I

l' I

t IALno,m.t m ltston i

i i

i i

i i

i i

i i

l I

i i

IFalluse to hubpond to an l

I I

l_

i l

i i

1

_1 i

I 1

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I I

I I

I I

l I

i i

l l

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3 m to-3 llo i lunat rive A.munc t a t ion.

I I

l l

1 1

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i 1 l l

1 1

I I r mo.. mm ve. Lut c!ven h

laie Acts..ted i

1 l

l 1

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1 4

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In.me.. m.ove,t,ut ctven i

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5,to-a ile ;

os in.at w. Ann nclatione i

i l

I i

i l

i I

i 1

1 I

I I 1

I I

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h l

I I

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i l

l l

l l

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l i

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lese A:tivatud Ir.s ture to compond to an I

i i

i i

i l

1 l

l i 5m 10-2 1

1 5 m lo-2 flo i lun-A.uiunc t. tea Dr. play of I I

I L

l l

l l

1 l

I (to) 1 8

l l

i i

lan Ahnom I mndition I

i i

i l

i i

i n

I I

i ll l

1 I

I I

I I

I I

I l-1 I

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i l

i 1

1 I

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I I~ l i

t r.nuse to noi.esly thee i

l i

l i

l 1

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5e to-a i 3l Iws t t ten noceduse.

I i

1 I

i i

i i

i i

i i l i

i h

l I

I I

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t i

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l I

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i i

)

I i

l i

i i

i I I I

i i

i i

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l I

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l i

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j i

l I

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i l

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i

_I n, oa l_ _

gn q tamy OH isletAN El4 holts e lit) Ikaste for giot tot estimatea

> fi developed NURAG/CR-3278 was also sosprenelble for

(!! Amt.asing no double-check proce.luse.f atiuse of oposator to detect error of human factors analysis in WASH-1400.

herefore, it was assussed 2 Z Ps o ject team th4t (2) Falluse to psoperly align value I O

omleelun ITable 2o-16).

that HUR14/CP-1278 reps enesited a refineevnt of the W ASH-4 400 e.osk g

llet, Table 13-1.

he one es-Assuming writ ten procedus en w/o double-check, ahost an.$ prob.alallities f asem thle ref erence ese retmsaled.

O (Il WASH-1400 e.guat es t his wit h t ellune to f ollow an a.hatentets ative psocedure.

ception is in the case of manual valves with lock anJ-chain cosi-14) taqqing cunat itut es following an adannist sa-lower esul of the NUPEG/Ch-1278 uncestalnty t i ves l>e ocedure t ha t inclu.$es a fosas o8 checkoff. Shost Blat values f euen trol.

In t his came, the (S) Fus eau htG/C R12 78, locking or la.asul w.is secwn. led, primarily lascause it seeemed likely that the s

Table 13-1 assumed.

n o.gia i s s smient to lock a valve In place tvunald dsaw unnae attention tu In ad.Iltisen, the cited niedlan o

(b) tams I.tet with checkoff.

L t s tun t t loss t han a lapl y t agging i t.

e o.out was anose closely relate.l to the f alluso to lock the (7)

Ase.mone contsola age functiorially geouged.

v.alue of (u) tes violatleen of lopulattunat st o s eot y g=s a s siesed.

valve th.an to t.he actual valve position.

g s

(9) A:ston alter 30 minutes.

n.)

us.ce t. t,

.,4 i, e.e.

e.l.

i Report No. 02-1040-1095 Revision 1 Page 63 TABLE 4-6 Failure Experience for Davis-Besse Valves lI l Total Valve l l

l l

Population l Total Demands l-l Failure l In Selected l Or Operating l';ttal Reported l Probability /

Hours l

Failures l

Rate IComponent/ Failure Mechanism l Safety Systems!

I I

i l

Motor Operated Valves:

l l

l l

Failure to open/close l

28 l

l l

on demand l

l 546 Demands l

8 l 1.5 x 10-2 l

l I

1 Air Operated Valves:

l l'

l l

.l Failure to open/close l

l l

l s

on demand

[

14 l

182 Demands l

7 l 3.8 x 10-2 l

1 1

I i

Chack Valves:

l l

l l

Failure to open on l

43 l

l l

dimand l

l 559 Demands l

0 l

1 I

I I

Leakage

[

11.3 x 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br /> l

3 l2.3 x 10-6hr-1 i

i i

4 I

n I

Report No. 02-1040-1095 Revision 1 Page 64 Failure Experience of Davis-Besse Auxiliary Feedwater Pumps l Number l l Total l

l of l Total Demands / l Feported l Failure Failure Mode l Pumps l Standby Hours l Failures Probability / Rate Failure to start l

2 l

84 Demands l

3 l

3.6 x 10-2 (since commercial l

l l

l operation l

l l

l I

I I

I Failure to start l

2 l

52 Demands l

0 l

(after one year of l

l l

l operation) l l

l l

l l

l l

4 hrs l

3 l4.9 x 10-5hr-1 Failure to continue l 2

l6.1 x 10 operating l

lper pump l

l 1

1 I

l 1

i 4

I J

I 1

. _ -_z_.

.m...

,~,_.,,,,___..r_,

Report No. 02-1040-1095 Revision 1 TABLE 4-8 Failure Experience of Davis-Besse Diesel Generators I

l Number of l

' Ictal l Total l

l Diesel l

Demands /

l Reported l Failure Failure Mode l Generators Hours l Failures l Probability / Rate Failure to Start l

2 l84 Demands l

1 l1.2 x 10-2 l

l 1

1 Failure to l

2 160595(1) hrs l

3 14.9 x 10-Shr-1 (1)

Stabilize l

l l

l l

1 I

I Failure to Con-l 2

l 767 hrs l

5 l6.5 x 10-3hr-1 (2) tinue operating l

l l

l l

l l

l l

1 I

I Failure to Start l

l l

l3 x 10-2 and Stabilize l

l l

l I

l l-l (1) per hour of standby (2) per hour of operation I

I I

I I

I

I Report No. 02-1040-1095 Revision 1 Page 66 TABLE 4-9 Initiating Events Challenging AWS II l Offsite Power j Non-Seismic Equip-Event l Availability l ment Availability l

l Category 1 l

l l

1 Loss of main feedwater

]

Yes l

Yes small break in RCS l

Yes l

Yes Steam generator overfill l Yes l

Yes Loss of forced RCS Cir-l Yes l

Yes culation l

l I

I I

Category 2 l

l l

1 Loss of offsite power l

No l

Yes Tornado l

No l

Yes i

I Category 3 l

l I

i i

Earthquake i

No l

No I

i 4

I 4

s I

4 i

I i

4

I Report No. 02-1040-1095 Revision 1 I

TABLE 4-10 Initiating Event Frequency Estimates

)

I Frequency (yr-1)

I l

l Davis-Besse l Recommended Initiating Event l Feference (5) l Peference (15)l Experience l Frequency I

I I

l

'I Loss of Main l

30 l

.70 l

.67 l

67 Feedwater l

l 1

l l

l I

1 Steam Generator l

l

.95 l

(3) l 95 Overfill i

l l

l 1

I I

l Small Break LOCA l 1.0 x 10-3 l

.17 l

(4)

I 17 in RCS l

l l

l l

1 I

I Loss of Forced l

l

.04 l

.3 l

3 RCS Circulation l l

l l

1 I

I I

Loss of Offsite l

.2 l

.32 l

.31 l

32 I

Power l

l l

l l

l l

l Tornado l

1.0 x 10-3 l

l6.3 x l1-4 (2)(1) l6.3 x 10-4 I

l l

1

'I Earthquake l

4.3 x 10-3 l

l (1)

[4.3 x 10-3 I

I i

1 (1) No events reported (2) Davis-Besse FSAR estimate (3) No events reported which activated the UWS (4) No events reported after " Burn in" period I

a f

4

..-._,..m-

.--__,-__,.-._m--

I:1 0 111111 4 - 1 DAVI5-BE5SE AfWRHIABIll1YPROGRAM ROW DMGRAM l

~IAULT T2EE NVELOPAf!UT ifAulT T2EE3 f0R fou2 Afd mafiGu2AT10HS

• FAulIT2iE5FOR IMTEQFAC&

5YSTEAtS ctsTEAlfUNAVANA8RI1YMAlYM i

• AlldisIAL CUTJET5 e.* UMAVAll AolllTIES [021002

.-s p,fyT E

V

]

• QUANTITATIV4 mad 3/S

- o-VT2/AUTOR6 N

• /stp02TANCE PAus.'njq UMAVAllAC/l17/

DATA AUALYol6 it1statuto 5ysTEy/twxTMAt/M

• WAtP0AIENTFAllu2E EATES

~*

• Efl ATIV4 FRERulMC/ES OF e-emtPLETED SY6 TEM FAMU24/lVEUT AfW FAllU2ES MtQutWFOR EACH AtitRNATIVE

• HusHU FACTORS AWALYS/$

• BAST $ FO' R AllMcAtivT3 04 AFW

/d/TIATIAIG E}'tWTAWALYS/S NS/64 ALT 42 NATIVES i

• FCEQulHCIES WEVENTS CHAlltuGIMG ASW

• IFFECT W DESIGU 04/

EV6HT FREQUENCIES 4

2 ;i'8 mn j

"E 2

?

E E

e

I Report No. 02-1040-1095 Revision 1 FIGURE 4-2 Page 69 Symbols Used in Fault Trees Events Circle -- Basic Event Diamond -- A fault event that is not resolved any further. Though this is not a basic event, it is :onsidered as if it were one in the analysis since it is not resolved any further, either due to lack of failure data at further resolution, or no further resolution is required f or the particular analysis.

I I

circle within a diamond -- A fault event that is treated like a basic event. The reliability / availability characteristics of this event are

/j calculated separately by a spearate fault tree analysis, and inserted in the main fault tree as if it were a basic event.

Double diamond -- An i=portant undeveloped basic event that requires i

further development.

M House -- An event that is normally I

expected to occur (probability of occurrence = 1), or never to occur (probability of occurrence = 0).

It can be used as a " switch" to turn "CN" or "CFF" parts of the tree.

i i

Report No. 02-1040-1095 ev sion 1 FIGURE 4-2 (Cont.)

Page 70 Symbols Used in Fault Trees I

Events I

l Rectangle --

I 1.

An intemediate event that is resolved further, or 11.

'The top event.

i I I

I Gates i

"AND" gate.

"OR" gate.

lI Combination gate h

"NOT" -- The small circle indicates

,l "NOT".

The bigger dotted circle

,E represents the basic event A which is "NOTed".

Together they represent the complement of A.

f OR gate with N inputs (listed), used in strcamlined format of the j

f simplified fault trees.

1N W)T 1

np&r N

!I

I Report No. 02-1040-1095 Revision 1 FIGURE 4-2 (Cont.)

page 71

_Svmbols Used in Fault Trees Other l

Transfer in -- The subtree below triangle is drawn elsewhere.

(This l

I is a convenience used in drawing

W large fault tr6es.)

l Transfer out -- The subtree drawn below the triangle belongs elsewhere.

f W

I \\

This complements the " transfer in" triangle, and an index number within

)

g the triangle indicates the correct ig match.

I ll A

7 Multiple transf ers in il

!I 1

1 Multiple transf ers out 1

i i

1 II

.i

!I i

I i

1 1

I Report No. 02-1040-1095 Revision 1 I

Page 72 FIGURE 4-3 System / Event Analysis Matrix I

I Initiating l AEWS Configuration l

l l

Event l

1 l

2 l

3 l

4 l

I i

1 1

I I

I l

Event 1 l frequency lx lfrequency lx lfrequency lx l frequency lx I

1 luavailability 1 lunavailability 2 lunavailability 3 l unavailability 4 l l

1 I

I I

I i

1 1

I I

I l

Event 2 Ifrequency 2x Ifrequency 2x Ifrequency 2x l frequency 2x l

l l unavailability 1 junavailability 2 lunavailability 3 l unavailability 4 l l

l I

I I

I I

I I

I I

I I

etc.

l l

l 1

l I

I I

I I

I I

I I

I I

I I

1 1

I I

I I

I I

I I

I I

I I

I I

I I

I la tal I

l l

l l Frequency l

l l

l l

lwith which l

l l

l l

I lAFWS is l

l l

l l

lunavailable i I

I I

I Iwwn l

I I

I I

l challenged i 1

l l

l

'I l

I I

I I

I i

I

~

I I

I I

I Report No. 02-1040-1095 I

Revision 1 Page 73 5.0 Results 5.1 Relative Unavailability Ranking of AMfS Configuration The unavailability of the Anis for each category of initiating I

event is summarized in Table 5-1.

These results are based on quantitative computer analysis of detailed fault trees for I

each Anis configuration.

In general, the results in Table 5-1 indicate that design and procedures modifications implemented since the TMI-2 event have improved the reliability of the Anis by about an order of magnitude. The inclusion of the analysis - based system -

improvements or a diverse third train of Anis results in about 1

another order of magnitude reduction in Anis unavailability.

5.1.1 Differences Among Event Categories For a given Anis configuration, differences among the Anis I

unavailabilities for different evsnt categories are due to assumptions regarding the availability of systems and components. The system availability is highest (lowest I

unavailability) for Category 1 events. With the assumption of loss of offsite power (Category 2 events), the potential unavailability of one or both diesel generators contributes additionally to the Anis unavailability. The large difference I

between Category 1 and Category 2 events results for the Post-TMI configuration is due primarily to the assumed absence of the feed and bleed procedure as a backup means to support the Amis safety function.

(The original emergency procedures y

for using the startup pump in conjunction with primary coolant system feed and bleed apply only to the situation in which offsite power is available). With the additional assumption of loss of all non-seismically qualified equipment (Category 3 events), twu of three AMIS water supplies are assumed to be unavailable. The potential unavailability of the remaining third water supply increases the Amis unavailability significantly.

5.1.2 Differences Among Amis Configurations As discussed in Section 5.2, the Pre-24I configuration unavailability is dominated by human factors. Failure of the diesel generators contributes to Anis unavailability for Category 2 events. For Category 3 events, the plugging of the common turbine exhaust line adds to the Pre-TII configuration unavailability.

lI Report No. 02-1040-1095 Revision 1 Page 74 I

Modifications to this exhaust line, implementation of the backup feed and bleed procedure, improvements in the control j

of other human factors and the design modifications discussed lm in Section 3.3 have improved the Post-24I configuration availability significantly. Dominant f ailures in the Post-U4I g

configuration analysis are mechanical failures. This suggests g

that further improvements to the A NS should address mechanical failure mechanisms in order to increase the overall system reliability.

The implementation of a third independent train of AWS is one method for addressing these mechanical failures. The Third Train configuration examined in this study provides over an I

order of magnitude improvement in AWS reliability. For Category 3 events, however, this Third Train provides no improvement in ANS reliability, since the Third Train itself is assumed not to be seismically qualified.

The analysis-based design and procedural modifications provide an alternative means to address the significant contributors I

to the Post-TMI A N S unavailability. The dominant Post-TMI configaration contributor, the f ailure of Mov to open/close on demand, becomes relatively unimportant with the analysis-based I

configuration valve alignment changes.

Improved procedures 1

for locking manual valves in the start-up pump train enhance the relisaility of the feed and blead procedure as an emergency backup to the A WS.

The extension of the feed and I

bleed procedure to loss of offsite power events, in conjunction with the analysis-based configuration design changes, provides almost two orders of magnitude improvement I

in AWS reliability for category 2 events. The analysis-based configuration design changes provide some improvement for Category 3 events, as well.

The uncertainty analysis results for cach configuration indicate that the standard deviation is the same order of magnitude as the point estimate (mean value) of the I

unavailability. The confidence in the calculated order of magnitude of the unavailability is, therefore, high.

It is our belief that the unavailability value shown in Table 5-1 I

are valid for making judgements about the relative reliabilities of each A W S configuration.

5.2 Significant contributors to AW S Unavailability Tables 5-2, 5-3, 5-4, 5-5 present results of the importance ranking of events in contributing to the unavailability of the AWS for the Pre-24I, Post-04I, Third Train and Analysis-Based I

I configurations respectively. The importance ranking is applied to categories of similar events (e.g. ; failure of motor operated valves to open/close on demand). Results are I

shown for each category of initiating event.

I Report No. 02-1040-1095 Revision 1 Page 75 The dominant f ailure mechanism for the Pre-DiI configuration is human factors primarily misalignment of locally operated I

and locally indicated valves. Also, improper manual throttling of the turbine admission valves is a relatively large contributor to the unavailability of this AWS configuration. Failures of motor operated valves and the turbine driven pumps are the major mechanical factors contributing to unavailability. For Category 3 events, the I

failure of the service water system to supply water to the A W S pumps and the potential plugging of the exhaust line are additional significant mechanical factors.

Implementation of locking procedures on manual valves and motor operated valves in the AW S have reduced the significance of these human f actors in the Post-til AWS.

I Mechanical factors have the greatest importance in this configuration. The dominant mechanical factor is the failure of motor operated valves to open/close on demand. The failure of the turbine driven pumps to start or to be unavailable due I

to tesc/m.iaintenance are also significant. Human factors are still significant for Category 1 events.

These are associated with the feed and bleed procedure, as a backup to the two AWS trains. For Category 3 events, the failure of the service water systen is an additional significant mechanical factor.

The Third Train configuration has the same failure mechanisms I

as the Post-TMI configuration with regard to the cwo AWS trains. The same mechanical factors, therefore, appear in Table 5-4.

These AWS failures must occur in conjunction with a failure of the Third Train in order to fail to achieve the AWS saf ety function. The major failure mode for the Third Train include both mechanical factors (failure of the startup pump) and human factors (failure to manually initiate I

feedwater flow via the Third Train).

Significant contributors to the Analysis-Based configuration I

AWS unavailability include the mechanical failure or unavailability due to maintenance of the AWS pumps. The failure of MOVs are significant only for the backup feed and i

I bleed procedure.

For Category 2 events the failure of Diesel Generator No. 1, which provid.es power to !!OVs in the letdown line and the makeup pump suctioc; line, becomes a significant contributor to the failure of the feed and bleed procedure.

I I-I

I Report No. 02-1040-1095 Revision 1 Page 76 I

It should be noted that additional motor-operated valves exist which could be used to manually isolate the letdown line.

Closure of these valves is not included in the normal letdown I

line isolation procedure. Thus, credit has not been taken for this additional means of letdown line isolation in the ova.rall evaluation, although it is expected that the operator would I

utilize these valves, if necessary.

5.3 Results of Combined System / Event Analysis Table 5-6 presents results of the combined system / event analysis for each of the four AWS configurations examined in this study. The Analysis-Based configuration has the lowest I

figure-of-merit (lowest frequency for AWS unavailability when challenged) followed in order by the Third Train configuration, the Post-MI configuration and the Pre-WI y

ccnfiguration. The Analysis-Based configuration and the Third Train configuration offe over a factor of ten improvement over the Post-TMI configuration. A significant difference in Table 5-6 is between the Pre-mI configuration and the I

Post-M I configuration. This is due to the system design modifications already impir.mented and the feed and bleed procedure which reduces the frequency for loss of the AWS safety function in Category 1 events.

Category 1 events are estimated to be significantly more frequent than Cat.egory 2 or Category 3 events. Although the I

AWS reliability is greatest for Category 1 events, the greater frequency of such events makes Category 1 the major contril '. tor to the c.aerall figure-of-merit for the Pre-mI, Third Train and Analysis-Based configurations. Category 2 events are the greatest contributors to the Post-MI configuration's figure-of-merit.

This is due primarily to the lack of emergency procedures (e.g., the feed and bleed procedure) as a backup to the A WS in these events.

I I

I I

1

Report No. 02-1040-1095 Revision 1 Page 77 Table 5-7 illustrates the relative significance of the Pre-TMI

- Post-MI changes. The overall significance of each change is defined as the reduction in the overall f requency with which the Am s is unavailable when challenged (total for all categories) assuming that only that change is made to the I

Pre-M I configuration. The most significant changes are attributed to the implementation of administrative procedures to lock in position all manual valves, the utilization of the emergency " feed and bleed" procedure and the dual level I

controls on the turbine admission valves. The existence of the auxiliary feedwater flow indication in the control room had no effect on the analysis. The significance.is, therefore, considered negligible.

5.4 Potential Cormon Cause Contributors to ANS Unavailability 1

A single failure or event may cause more than one component of the AWS to fa'.1.

This commonality in failure may result from shared power supplies, cooling water sources, operator actions, harsh environment and external causes (e.g.; fire, missiles, etc.).

Where possible, such common cause contributors are explicitly modeled in the fault tree analysis. Such modeled common cause factors include electric I

power supplies and cooling water supplies. Other factors are not explicitly modeled. The quantitative impact of these potential common causes on the AN S unavailability is not assessed.

I I

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I

Report No. 02-1040-1095 Revision 1 TABLE 5-1 Page 78 A WS Unavailability I

Event l

AWS l

ANS l Standard Category l Configuration l Unavailability i Deviation l

l l

1 l Pre-mf l3.3 x 10-2 l1.2 x 10-2 l Post-TMI 16.6 x 10-4 l3.3 x 10-4 I

l Third Train l4.5 x 10-5 l2.7 x 10-5 l Analysis-Based 13.3 x 10-5 l2.0 x 10-5 7

I I

I I

I i

1 2

l Pre-mI l4.1 x 10-2 l1,4 x 10-2 JPost-MI 15.5 x 10-3 l2.2 x 10-3 l Third Train l1.4 x 10-4 l1 2 x 10-4 l Analysis-Based l9.3 x 10-5 16.7 x 10-5 7

l l

l l

l 1

3 l Pre-MI l8.8 x 10-2 l2.6 x 10-2 l Post-mI l1.9 x 10-2 l1.2 x 10-2 JThird Train l1.9 x 10-2 l1.2 x 10-2 4

l Analysis-Based l1.1 x 10-2 l1,1, x 10-2 l

l l

1 i

I I

I I

Report No. 02-1040-1095 Revision 1 TABLE 5-2A Page 79 Significant Contributors to AFWS Unavailability PRE-TMI Configuration Category 1 Events Importance Mechanical Factors:

motor operated valves fail to open on demand

.42 i

- failure of turbine driven pump to start

.10 l

- failure of turbine driven pump to continue operating

.01

- turbine driven pump in test / maintenance

.08 i

l Human Factors:

- valve misalginment - recirculation line

.El

- valve misalignment - test line

.02

- improper throttling of turbine admission valves

.15 I

I I

l

Report No. 02-1040-1095 Revision 1

,l 3

TABLE 5-2B page 80 Significant Contributors to AFWS Unavailability J

PRE-TMI Configuration Category 2 Events Importance Mechanical Facdcrs:

motor operated valves fail to open on demand

.37 l

l

- failure of turbine driven pump to start

.08

- failure of turbine driven pump to continue operating

.01 1

vuybine driven pump in test / maintenance

.06 Electric Power Factors:

- failure of diesel generator te start

.035 failure of diesel generator to continue operating

.075 Htman Factors:

valve misalignment - recirculation line

.74

- valve misalignment - test line

.02

- improper throttling of turbine admission valves

.13 I

ill I

I Raport No. 02-1040-1095 Revision 1 TABLE 5-2C Page 81 Significant Contributors to AFWS Unavailability PRE-TMI Configuration Category 3 Events Importance Mechanical Factors:

- motor operated valves fail to open on demand

.30 failure of turbine driven pump to start

.04 failure of turbine driven pump to continue operating

.004

- turbine driven pump in t,st/ maintenance

.04

- failure of service water system

.24

- plugging of turbine exhaust line

.11 Electric Power Factors:

- failure of diesel generator to start

.031

- failure of diesel generator to continue operating

.067 Human Factors:

valve misalignment - recirculation line

.45

- valve misalignment - test line

.01

- improper throttling of turbine admission valves

.08 I

I

I Report No. 02-1040-1095 TABLE 5-3A Revision 1 Page 82 Significant Contributors to AWS Unavailability Post-TMI Configuratior Category 1 Events Importance I

Mechanical Factors:

- motor operated valves fail to open on demand

.89 failure of turbine driven pump to start

.29

- failure of turbine driven pump to continue operating

.03 turbine driven pump in test / maintenance

.18 Startup Pump with Feed and Bleed Mechancial Factors:

- motor operated valves fail to open on demand

.02 isolation valve on letdown line fails to close

.10 on demano

- air operated valves fail to open on demand

.05 PORV fa.ils to open on demand

.10

- failure to startup pump to start

.02

- failure to startup pump to continue operating

.005

- failure of makeup tank water supply

.05 Human Factors:

operator fails to isolate letdown line

.09

.03

- operator fails to start startup pumps I

- operator operates PORV incorrectly

.05

- valve misalignments - borated water storage line 0.09

- valve misalignments - valve ml02 (startup pump train) 0.35 I

l Report No. 02-1040-1095 l

Revision 1 TABLE 5-3B Page 83 i

Significant Contributors to AFWS Unavailability Post-TMI Configuration Category 2 Events Importanc e Machanical Factors:

- motor operated valves f all to open on demand

.86 failure of turbine driven pu=p to start

.25 f ailure of turbine driven pump to continue operating

.03

~

turbine driven pump in test / maintenance

.18 Electric Power Factors:

I f ailure of diesel generator to start

.038 failure of diesel generator to continue operating

.082

~

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I

I Report No. 02-1040-1095 Revision 1 Page 84 TABLE 5-3C Sicnificant Contributors.to AFWS Unavailability I

Post 'IMI Cenfiguration Category 3 Events Importanc e

(

Mechanical Factors:

- mots operated valves f all to open on demand

.42 I

- f ailure of turbine driven pu=p to start

.09 f ailure of turbine driven pump to continue operating

.01 turbine driven pump in test / maintenance

.07 failure of service water system

.54 Electric Power Factors:

- failure of diesel generator to start

.029 failure of diesel generator to continue cperating

.062 I

I I

I

Report No. 02-1040-1095 Revision 1 TABLE 5-4A Page 85 Significant Contributors to AEWS Unavailability

)

Thii'd Train Configuration Category 1 Events Importance 2

Mechanical Factors:

- motor operated vavles fail to open on demand

.89

}

- failure of turbine driven pump to start

.28 1

- failure of turbine driven pump to continue operating

.01

- failure of startup pump to start

.29 failure of startup pump to continue operating

.06

{g l3

- turbine driven pump in test / maintenance

.19 Human Factors:

l

- failure of operator to start startup pump

.60

!I lI lI lI 1

l II

!I

!I lI

I Report No. 02-1040-1095 Revision 1 TABLE 5-4B Page 86 l

l Significant Contributors to AFWS Unavailability Third Train Configuration I

Category 2 Events Importance Mechanical Factors:

motor operated valves fail to open on demand

.80 failure of turbine driven pump to start

.19

- failure of turbine driven pump to continue operating

.01

- failure of startup pump to start

.04

- turbine driven pump in test / maintenance

.14 Electric Power Factors:

- failure of diesel generator to start

.35

- failure of diesel generator to continue operating

.52

- failure of bus D2

.04 Human Factors:

I

- failure of operator to start startup pump

.16 I

I I

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m-l Report No. 02-1040-1095

=

Rt: vision 1 TABL1; 5-4C Page 87 Significant Contributors to AWS Unavailability Third Train Configuration I

Category 3 Events Importance Mechanical Factors:

- motor operated valves fail to open on demand

.42 failure of turbine driven pump to start

.09 failure of turbine driven pump to continue operating

.01

- turbine driven pump in test / maintenance

.07

- failure of service water system

.54 Electric Power Factors:

- failure of diesel generator to start

.029

- failure of diesel generator to continue operating

.062 I

I I

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I

Report No. 02-1040-1095 TABLE 5-5A Revision 1 page 88 Significant Contributors to AFWS Unavailability Analysis-Based Configuration i=

Category 1 Events Importance f

1 Mechanical Factors:

motor operated valves fail to open on demand 0.019 failure of turbine driven pump to start 0.69

- failure of turbine driven pump to continue operating 0.07 turbine f. riven pump in test /maintenince 0.48 Startup Pump with Feed and Bleed Mechanical Factors:

- motor operated valves fail to open on demand 0.03

- isolation valve on letdown line fails to close on 0.25 I

- air operated valves fail to open on demand 0.06 1

demand PORV fails to open on demand 0.26

- failure of startup pump to start 0.05 I

- failure of startup pump to continue operating 0.01 Human Factors:

valve misalignments - borated water storage line 0.11

- operator fails to isolate letdown line 0.08 operator fails to start startup pumps 0.03 I

operator operates PORV incorrectly 0.08 I

I

'I Report No. 02-1040-1095 TABLE 5-5B l

Page 89 Significant Contributors to AFWS Unavailability Analysis-Based Configuration

( 1 agory 2 Events Importance Mechanical Factors:

motor operated valves fail to open on demand 0.01 failure of turbine driven pump to start 0.77 failure of turbine driven pump to continue operating 0 08 I

turbine driven pump in test / maintenance 0.41 Startup Pump with Feed and Bleed Mechanical Factors PORV fails to open on demand O.08

- isolation valve on letdown line fails to close 0.08 on demand 1

failure of startup pump to start 0.008 Electric Power Factors:

failure of diesel generator to start 0.24

- failure of diesel generator to continue operating 0.52 Human Factors:

I

- operator fails to isolate letdown line 0.02 operator fails to start startup pump 0.03

- operator operates PORV incorrectly 0.02 I

I I

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Report No. 02-1040-1096 Revision 1 TABLE 5-SC Page 90 5

Significant Contributors to AEMS Unavailability Analysis-Eased Configuration Category 3 Events Importance 1

i Mechanical Factorst

- motor operated valves fail to open on demand 0.18

- failure of turbine driven pump to start 0.09 failure of turbine driven pump to continue operating 0 01

- turbine driven pump in test / maintenance 0.05

- failure of service water system 0.89 1

W l

Electric Power Factors:

- failure of diesel generator to start 0.002

- failure of diesel generator to continue operating 0.005 il 1,

,!I f

lI L

h ll

Rsport No. 02-1040-1095 R*. vision 1 TABLE 5-6 Page 91 Combined System / Event Analysis Results I

l l

l Frequency of AEWS l

l l

l Unavailability (yr-1) l I

l Frequency l

l l

l Initiating Event I

(yr-1) l Pre-MI l

Post-MI l Analysis-Basedl Third Train i

I I

I l

l l

1 l

l Category 1 l

l l

l l

1 l

l l

1

- loss of main feedwater l

.67 l2.2x10-2 14.4x10-4 12.2x10-5 l3 0x10-5

- steam generator overfill l

.95 13.1x10-2 16.3x10-4 13.1x10-5 l4,3xio-5

- small break in RCS l

.17 l5.6x10-3 l1,1xio-4 15 6x10-6 17 7x10-6

- loss of fe-ced RCS l

l l

l l

circulation l

.3 19.9x10-3 12.0x10-4 l9 8x10-6 l1,4xio-5 I

i l

I i

l I

I I

Category 1 'Ibtal l

16.9x10-2 ll.4x10-3 16 8x10-5 19 4x10-5 I

I I

I I

I I

I I

l Category 2 l

l l

l l

l 1

l l

1

- loss of offsite power l

.32 l1.3 10-2 ll.6x10-3 12 9x10-5 l4 5x10-5 1

- tornado l6.3x10-4 12.6 10-5 13.5x10-6 15.8x10-8 l8 8x10-8 I

I 1

I I

i l

I Category 2 2tal l

l1.3x10-2 11.8x10-3 l2.9x10-5 l4.5x10-5 I

I I

I I

I ategory 3 I

I i

l I

C l

l l

l l

l l

l l

l l4.3x10-3 13 8x10-4 l8.2x10-5 14.8x10-5 18.2x10-5 I

- earthquake I

I I

I I

I 1

1 l

l Category 3 htal l

13.8x10-4 l 8. 2x10-5 l4.8x10-5 18.2x10-5 I

I I

I I

i i

l i

i Frequency with which l

l l

l l

AFWS is unavailable l

l l

l l

wh n challenged l

l 8. 2x10-2 13.3x10-3 l1,4xio-4 12.2x10-4 (Total for all l

l l

l l

l l

l l

l 1 categories) l l

i l

l I

I

l Rsport No. 02-1040-1095 Rsvision 1 TABLE 5-7 Pagm 92 l

Relative Significance of the Pre-TMI - Post-TMI Changes 1

overall Significance

• 1 1.

Valves on Train 1 (AF-360, Ar-3a70, and the 4 4 x 10-4 Main Steam Turbine Admission Valve MS-106) are powered off DC power supplies.

2.

Turbine exhausts are redundant and seismically 4.0 x 10-5 qualified.

3.

Administratiec procedures have been imple-6.6 x 10-2 mented to lock in position all raanual valves I

and local control stations and hand wheels for motor operated valves.

4.

Turbine adr.ission valves have automatic dual 1.0 x 10-2 I

level contrcl, with option for manual control.

5.

Auxiliary feedwater flow indication now Negligible I

exists in the control room.

6.

An emergency procedure has been implemented 5.9 x 10-2 I

to manually start and align the main feedwater startup pump to provide feedwater to the steam generators in the event that both trains of I

AFWS fail.

This also includes the feed and bleed procedure.

I

• The "overall significance" is defined as the reduction in the overall frsquency with which the AF4S is unavailable when challenged (total for I

all event categories) assuming that only that change is made to the Pre-TMI Configuration.

I I

l P

Report No. 02-1040-1095 Revision 1 Page 93 ll 6.0 Conclusions 1

The methodology discussed in this report provides a useful tool with which to assess the relative reliability of various potential AFWS configurations at Davis-Besse Unit No. 1.

The results may be used as input to AFWS design decisions. Such I

decisions include system level decisions (i.e.; the relative benefits offered by the various AFWS configurations examined) and component level decisions (i.e.: relative importance of individual components in contributing to system failure).

Component level judgments can be the basis for developing improved A1WS designs, not explicitly considered in t'ais a nalysis.

Major conclusions resulting from this study include the following:

The Pre-TMI AFWS configuration has a relatively high unavailability. This is due largely to potential human factor failure mechanisms.

The design and procedures modifications, originally 1

planned or implemented at Davis-Besse Unit No. 1 subsequent to the TMI-2 event, effectively address the major failure mechanisms found in the Pre-TMI configuration analysis. The Post-TMI configuration reliability is over a factor of ten greater than the I

Pre-THI configuration reliability.

The major contribution to the Post-TMI AFWS unavailability is the failure of motor operated valves to open/close on demand. Mechanical failures associated with the turbine driven pumps are of lesser importance.

The reliability of the Post-TMI " feed or bleed" method to 1

provide backup auxiliary feedwater is not high, by itself, because of human factors. The calculated unavailability on demand is 0.14.

However, this backup system does provide an additional measure of reliability to the already highly reliable AFWS.

The inclusion of a third independent train of AFWS offers an order of magnitude improvement in AFWS reliability.

The reliability of the feed and bleed procedure can be improved through additional administrative controls and explicit instructions for parameter response verification. The unavailability of the feed and bleed 1

method with these improvements (Analysis-Based configuration) is 0.06 for Category 1 events.

I I

1 Report No. 02-1040-1095 Revision 1 Page 94 The AFWS design modifications which are part of the Analysis-Based configuration enhance the AFWS reliability, by themselves, nearly as much as the addition of a third 1

I train. The Analysis-Based design modifications, in conjunction with the procedural changes, enhance the AFWS reliability more than does the addition of a third train.

The overall figure-of-merit for the AFWS is dependent not only on the system reliability, but also on the frequency with which it is challenged.

Improvements in the I

figure-of-tarit can be achieved through plant design and procedures modifications which would reduce the frequency of challenges to the AFWS (primarily Category 1 events).

Such improvements may have a greater impact on the AFWS figurc-of-merit than do AFWS design modifications. There have been many such improvements made at Davis-Besse Unit No.1 since the THI-2 event, but their impact on this analysis has not been quantified due to lack of sufficient performance data.

I The use of plant-specific data may have a significant impact on reliability analysis results. Where conservative, Davis-Besse specific data are used in this reliabilty analysis.

Il

I Report No. 02-1040-1095 Revision 1 Page 95 7.0 References 1.

Nuclear Regulatory Commission order in the matter of The Toledo Edison Company and the Cleveland Electric Illuminating Company Davis-Besse Nuclear Power Station, Unit No.1 dated May 16, 1979.

l 2.

Letter from Harold R. Denton (NRC) to Lowell E.

Roe (TECo) dated July 6, 1979.

3.

" Auxiliary Feedwater System Diversity Study" Bechtel Associates Professional Corp., August 22, 1980.

4.

"The IMPORTANCE Computer Code", Lawrence Livermore Laboratory, March 14, 1977.

I 5.

" Reactor Safety Study - An Assessment of Accident Risks in U.S.

Commercial Nuclear Powr Plants: Appendix III",

WASH-1400, U.S. Nuclear Regulatory Cocaission, 1975.

6.

" Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection System,"

IEEE-352, Institute of Electrical and Electronic Engineers, 1975.

7.

" Data Summaries of Licensee Event Reports of Valves at U.S.

Commercial Nuclear Power Plants", NUREG/CR-1363, Volumes 1 - 3, 1980.

8.

" Data Summaries of Licensee Event Reports of Pumps at U.S. Commercial Nuclear Power Plants", NUREG/CR-1205, 1980.

9.

" Data Summaries of Licensee Event Reports of Diesel Generators at U.S. Commercial Nuclear Power Plants",

'UREG/CR-1362, 1980.

10.

"IEEE Guide to the Collection and Presentation of Electrical, Electronic cnd Sensing Component Reliability Data for Nuclear Power Generating Stations', IEEE-500, Institute of Electrical and Electi cnic Engineers, 1977.

11.

" Handbook of Human Reliability Analysis with Emphasis on No" o r Power Plant Applications", NUREG/CR-1978.

12.

Da< s-desse Unit No. 1 Technical Specifications.

I 13.

" Nuclear Plant Reliability Data System 1978 Annual Reports of Cumulative System and Component Reliability",

NUREC/CR-0942, 1979.

I

lI Report No. 02-1040-1095 Revision 1 Page 9,6 14.

"WAMCUT, A Computer Code for Fault Tree Evaluation",

Electric Power Research Institute Report EPRI NP-803, I

June, 1978 15.

"AMS: A Reappraisal, Part III, Frequency of Anticipated I

Transients", EPRI NP-801, Electric Power Research Institute, July, 1978.

16.

EDS Calcu?ations Electric Power System Fault Tree - Post MI, 1040-003-002-1, Rev. 1 Electric Power System Fault Tree - Pre-MI, 1040-003-002-2, re- ' 0 Auxiliary Feedwater System Fault Tree - Post-mI, 1040-003-003-1, Fev. 1 Auxiliary Feedwater System Fault Tree - Pre-mI, 1040-003-003-2, Pav. O Startup Pump as Third Train, I

1040-003-003-3, Rev. 0 1

Main Steam System Fault Tree - Post-MI, I

1040-003-004-1, Pev. 1 Main Steam System Fault Tree - Pre-MI, 1040-003-004-2, Rev. O Reliability Data - Human Errors, 1040-003-005-1, Rev. O Calculation of Failure Probabilities for Davis Besse Unit 1, 1040-003-005-2, Rev. 1 Failure Data Base, 1040-003-005-3, Fe v.

O Human Error Probability - Feed and Bleed,

/-

1040-003-005-4, Re v.

O V

Event Frequency Data Base, 1040-003-006, Re v. O Derivation of Second rioment, 1040-003-007, Re v.

O i

1

1 Report No. 02-1040-1095 Revisien 1 Page 97 I

WAMCUT Analysis - AWS, Post-MI, Category 1, 1040-003-008-1, Rev. O WAMCUT Analysis - AWS, Post-MI, Category 2, 1040-003-008-2, Re v.

O WAMCUT Analysis - AWS, Post-MI, Category 3, 1040-0 03-0 08-3, Rev. O WAMCUT Analysis - AWS, Pre-mi, Category 1, 1040-003-009-1, Rev. O I

WAMCUT Analysis - AWS, Pre-MI, Category 2, 1040-003-nn9-2, Rev. O WAMCUT Analysis - AWS, Pre-MI, Category 3, 1040-003-009-3, Re v. O WAMCUT Analysis - AWS, Third Train, Category 1, 1040-0 03-010-1, Rev. O WAMCUT Analysis - A WS, Third Train, Category 2, 1040-003-010-2, Re v.

O WAMCUT Analysis - Electric Power, Post-mI, Category 1, 1040-003-011-1, Fev. O I

1 vfM1 CUT Analysis - Electric Power, Post-mI, Category 2, 1040-003-011-2, Rev. O WAMCUT Analysis - Electric Power, Pre-MI, Category 1, 1040-003-012-1, Rev. O WAMCUT Analysis - Electric Power, Pre-m I, Category 2 1040-003-012-2, Rev. O

~

Importance Calculation, 1040-003-013, Re v.

1 WAMCUT Analysis - Post-MI Start-up Pump with Feed and Bleed, 1040-003-015, Fev. O I

Startup Pump with Feed and Bleed Fault Tree, 1040-003-016, Re v.

1

I Report No. 02-1040-1095 Revision 1 Page 98 Significance of Pre 'IMI - Post TMI changes, 1040-003-018, Re v. O

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AUXILIARY FEEDWATER SYSTEM '

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i AUxtttARY FEEDWATER SYSTEM PRE-TMI

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