NUREG/CR-4967

EGG-2514

Nuclear Plant Aging Research on

High Pressure Injection Systems

Prepared by L. C. Meyer*

Idaho National Engineering Laboratory

EG&G Idaho, Inc.

Prepared for

U.S. Nuclear Regulatory

Commission
I

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NUREG/CR-4967

EGG-2514

RM, R9

Nuclear Plant Aging Research on

High Pressure Injection Systems

Manuscript Completed: July 1989

Date Published: August 1989

Prepared by

L. C. Meyer

Idaho National Engineering Laboratory

Managed by the U.S. Department of Energy

EG&G Idaho, Inc.

P. 0. Box 1625

Idaho Falls, ID 83415

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN A6389

Under DOE Contract No. DE-AC07-761D01570
ABSTRACT

This report represents the results of a review of light water reactor High Pressure

Injection System (HPIS) operating experiences reported in the Nuclear Power Experience Data Base, Licensee Event Reports (LER)s, Nuclear Plant Reliability Data

System, and plant records. The purpose is to evaluate the potential significance of

aging as a contributor to degradation of the High Pressure Injection System. Tables

are presented that show the percentage of events for HPIS classified by cause, component, and subcomponents for PWRs. A representative Babcock and Wilcox plant was

selected for detailed study. The U.S. Nuclear Regulatory Commission's Nuclear Plant

Aging Research guidelines were followed in performing the detailed study that identifies materials susceptible to aging, stressors, environmental factors, and failure

modes for the HPIS.

In addition to the engineering evaluation, the components that contributed to system unavailability were determined and the aging contribution to HPIS unavailability

was evaluated. The unavailability assessment utilized an existing probabilistic risk

assessment (PRA), the linear aging model, and generic failure data.

FIN No. A6389-Nuclear Plant Aging Research on High Pressure Injection Systems

iii
EXECUTIVE SUMMARY

Operating experience of nuclear power plants is

evaluated to determine the significance of service

wear on equipment due to aging and the possible

impact of service wear on safety. The High Pressure

Injection System (HPIS) and those portions of

related systems needed for operation of the HPIS

were selected for detailed study and emphasized in

this report. This research is part of the U.S.

Nuclear Regulatory Commission's (USNRC's)

Nuclear Plant Aging Research (NPAR) Program

and follows the NPAR guidelines.

The NPAR guidelines provided the framework

through which the effect of aging on HPI was studied. The products asked for in the NPAR guidelines

include: an identification of failure modes; a preliminary identification of failure causes due to

aging and service wear degradation; and a review of

current inspection, surveillance and monitoring

methods, including manufacturer recommended

surveillance and maintenance practices. Performance parameters or functional indicators potentially useful in detecting degradation are also

identified and preliminary recommendations are

made regarding inspection, surveillance, and monitoring methods.

A description of the HPIS for a Babcock and

Wilcox PWR is presented based on information

provided by a cooperating utility. The description

provides a general understanding of a HPIS. A

variety of designs exist from the various NSSS vendors. However, they are all similar in that they use

borated water, generally inject into the cold-leg piping, utilize high head centrifugal motor driven

pumps, have motor operated valves and check

valves, and have similar operating environments.

There is some concern that the operational differences and variations in system boundaries would

lead to different conclusions if each type of system

were considered separately. However, this is not

considered likely because of the similarity of equipment and environment.

The HPIS for the B&W type of system is part of

the Emergency Core Cooling System (ECCS) and

has two modes of operation under emergency conditions. The first is the high pressure injection

mode that is necessary to prevent uncovering of the

core for small LOCAs where high system pressure is

maintained, and to delay uncovering of the core for

intermediate size LOCAs. The second provides

long-term core cooling following a LOCA using the

high pressure recirculation mode. Part of the HPIS

is also used during normal operation to provide

reactor coolant pump seal cooling and to maintain

the volume of the reactor coolant system within

acceptable limits. The HPIS interfaces with many

other systems in performing its functions, which

include IE electrical power, service water, instrument air, low pressure injection system, and the

Engineered Safety Feature Actuating System.

In addition to the one plant that was studied in

detail, generic data bases were reviewed for HPIS

failures in PWRs. This review provided larger statistical bases for determining HPIS problems

related to aging. Data sources used include the

Nuclear Power Experience (NPE) data base (prior

to 1986), Licensee Event Reports (LERs) (prior to

1984), Nuclear Plant Reliability Data System

(NPRDS) (prior to 1986), and material from an

operating nuclear plant supplied by a utility

(including personnel interviews). The system

boundaries as defined by each data base were used

as reported.

A review of Nuclear Power Experience (NPE)

shows that the most frequent cause of HPIS component failures is maintenance error, followed by

design error, and mechanical disability. The four

types of components with the highest frequency of

failure for PWRs were valves-35%o; I&C-19%;

pumps-15%; and piping-7%.

Instrumentation and Control (I&C) failures

included the sensors, electronics, and motor control centers for valves and pumps. Instrumentation

accounted for 50% of the I&C failures, valve control 17%, pump control 9%, and the rest were miscellaneous control circuits.

HPI piping failures, after eliminating design,

construction, and maintenance errors (which

accounted for 37%), were primarily due to weld

failures-15%, corrosion-7%, and vibration-5 %.

The rest of the events were spread over many

causes. The HPI pipe sizes varied from I to 14 in.,

depending upon location. Failures were not dominated by any one particular size.

Command faults were the leading cause for valve

failures, reported in LERs. They include electrical

power or any support system that prevents the valve

from performing its intended function. Mechanical

parts failure, seat or disk failure problems, packing

failures (leaks), and foreign material were the most

frequent causes of the basic valve failures. There

were 44 HPI pump failures reported in the LER

data base. Out of these, 22 events were caused by

iv
control, maintenance, and design error. The

remaining events found no single dominant cause

of failure. Only nine failures of HPIS pumps were

potentially aging related.

The NPRDS data followed the same component

failure pattern as NPE. The aging fraction for HPI

based on NPRDS data was 0.213, indicating 21.3%

of system failures in the HPIS were aging related.

Plant data followed the same pattern as NPE for

those events requiring an incident investigation

report. Plant maintenance records listed many

more events. While the top four components with

the highest frequency of failure were the same, the

order placed I&C first; pipe, supports, and nozzles

second; then valves and pumps. These data indicate

that many minor problems associated with I&C

and pipe hangers received corrective maintenance

before major failures occurred.

A review of the electrical standards identified

that operational life is based on accelerated aging

tests. As naturally aged component data becomes

available, the standards should be updated. Standards for mechanical equipment are under review by

ASME Section XI and are supported by the NPAR

research where applicable.

The conclusions for the HPIS study are based on

a review of one plant and generic information from

the various data bases. The plant maintenance

record contains many minor adjustments and

repairs for I&C and pipe hangers, but most of the

major components failures concerned valves and

pumps. Serious piping problems concerning thermal sleeve and nozzle cracking were attributed to

thermal fatigue. The utilities affected have taken

corrective action by redesigning the thermal

sleeves, using warm up lines and enhanced IS&M.

Materials in seals and valve packing deteriorates

with time and results in leaks. Borated water leaks

are potentially serious because of the corrosive

action of boric acid on carbon steel and potential

for loss of pressure boundary.

Approximately 57% of the component failure

lead to system degradation but, because of system

redundancy, only 0.7% caused loss of system function. The failure modes that involve total loss of

system function are, failure to inject cooling water

for emergency operations, and failure to provide

makeup water or seal cooling water for normal

operations.

The specific problems related to aging were:

(a) through-wall cracks occurring in the makeup

nozzle and safety injection line elbow from thermal

fatigue, (b) valves failing to operate due to boron

crystallization, and (c) injection boron concentration diluted from leaking valves.

Inspection and surveillance review has identified

that electrical measurements on pump motors and

valve operators (for MOVs) could be used to detect

aging. Also, that improved inservice testing of

valves is needed to detect aging and assure operability with load. The detection of cracks caused by

thermal fatigue requires enhanced ultrasonic testing methods. In addition, inspection of base metal

in high-stress regions is needed to detect cracks in

those areas.

The HPIS unavailability assessment identified

that motor-operated valves contributed significantly to unavailability of the system for all three

operating modes evaluated. The HPIS pumps were

significant contributors to systems unavailability

for the two of three pumps required mode and the

recirculation mode of operation. The time dependant unavailability assessment showed that the

HPI unavailability was only moderately affected by

aging with a relatively small increase over the operating life.

v
ACKNOWLEDGMENTS

The detailed systems studies are based on material supplied by the Duke Power

Company, Charlotte, North Carolina.

VA
CONTENTS

ABSTRACT ......................................................................... iii

EXECUTIVE SUMMARY . ........................................................... iv

ACKNOWLEDGMENTS . ............................................................ vi

ACRONYMS ........................................................................ xv

INTRODUCTION .................................................................... I

Aging and Plant Safety . ........................................................... I

NPAR Program Goals and Strategy ................................................ I

Phase I Aging Assessment of a PWR High Pressure Injection System ..... ............... 2

SYSTEM FUNCTIONAL DESCRIPTION .............................................. 5

Makeup and Purification System .................................................. 5

Cooling System for RCP Seals ............ ......................................... 7

Emergency Injection Mode of the HPIS ............................................ 7

High Pressure Recirculation Mode ................................................. 7

INTERFACES WITH SUPPORTING SYSTEMS ......................................... 10

Electrical ....................................................................... 10

Service Water . ................................................................... 10

Instrument Air .1.................................................................. 10

ECCS Pump Room Coolers ............ I ...................................I......... 10

Engineered Safety Features Actuating System ........................................ 10

SYSTEM COMPONENTS AND HARDWARE .......................................... 12

Valves ........................................................................ 12

Instrumentation and Control .............. ........................................ 12

HPI Pump A Controls ............. ......................................... 12

HPI Pump B Controls .............. ......................................... 12

HPI Pump C Controls ............. ......................................... 14

Motor-Operated Valves .............. ........................................ 14

Pneumatic-Operated Valves .................................................. 14

HPI Pumps .................................................................... 14

vii
HPI Piping ...................................................................... 14

Nozzles and Thermal Sleeves ...................................................... 14

Piping Penetrations . .............................................................. 14

Pipe Hangers . .................................................................. 14

Snubbers ....................................................................... 14

Tanks .......................................................................... 15

OPERATING EXPERIENCE .............. ........................................... 16

Nuclear Power Experience Data .................................................... 16

LER Data ..................... 19

Pump Failures . ............................................................. 19

System Pumps . ............................................................. 19

Valves ..................................................................... 19

Nuclear Plant Reliability Data System .............................................. 19

Plant Operating Experience from Site Visit and Personnel Interviews ..... ............... 26

Nuclear Maintenance Data Base .............................................. 26

Incident Investigation Reports ................................................ 27

Summary of Operating Experience ................................................. 27

Piping . ................................................................... 29

Snubbers .................................................................. 30

Penetrations ............................................................... 30

Tanks ..................................................................... 30

Pumps .................................................................... 30

Valves ..................................................................... 30

Chemistry Problems ............. ........................................... 30

Microbial Influenced Corrosion ............................................... 30

HPI SAFETY ISSUES AND POTENTIAL AGING PROBLEMS ........................... 31

Locking Out of HPIS Power-Operated Valves ........................................ 31

Inadvertent Actuation of Safety Injection in PWRs ................................... 31

Switch from HPI Mode to Recirculation Mode ....................................... 31

High Pressure Recirculation System Failure Due to Containment Debris ..... ............. 31

Failure of Demineralizer System and the Effect on HPIS .............................. 31

Systems Interaction . .............................................................. 32

viii
AGING ASSESSMENTS FOR THE HPIS ............................................... 33

Preliminary Identification of Susceptibility of Materials to Aging ....................... 33

Stressors for HPIS . ............................................................... 33

Maintenance, Operations, and Testing Stressors ................................. 34

Environmental Stressors ............. ........................................ 34

Electrical Stressors ................ .......................................... 34

Mechanical Stressors ............... ......................................... 34

Functional Indicators that would Aid in Failure Prediction ............................. 34

Methods of Detection and Control of Aging Degradation .............................. 34

Aging Assessment Summary ............. ......................................... 35

REVIEW OF INSPECTION, SURVEILLANCE, AND TESTING .......................... 37

ROLE OF MAINTENANCE IN COUNTERACTING AGING EFFECTS ..... ............... 40

Present Regulations and Guidance ................................................. 40

Current Maintenance Practices .................................................... 40

Normal Operation Precautions ............................................... 40

High Pressure Injection Mode Precautions ..................................... 40

Benefit of Preventive and Corrective Maintenance .................................... 40

Improper Maintenance . .......................................................... 41

Recommendations for Preferred Maintenance Practices ............................... 41

CODES AND STANDARDS . ........................................................... 42

HPIS AGING SYSTEM UNAVAILABILITY ASSESSMENT ............................... 43

PRA and Basic Event Data . ....................................................... 43

Failure Cause Data . ............................................................. 43

Methodology . .................................................................. 43

Assumptions .................................................................... 46

Components that Contribute Significantly to System Unavailability ..... ................ 46

HPIS Unavailability . ............................................................ 48

Estimate for Types of Events that Cause HPIS Unavailability .......................... 51

CONCLUSIONS ..................................................................... 53

ix
Operating Experience . ........................................................... 53

Specific Problems Related to Aging ................................................ 53

Aging Assessment . ............................................................... 54

Inspection, Surveillance, and Monitoring ........................................... 54

Maintenance .................................................................... 54

HPIS Unavailability Assessment ............ ....................................... 54

REFERENCES ...................................................................... 56

APPENDIX A-REACTOR COOLANT MAKEUP SYSTEM, PURIFICATION

SYSTEM, AND COOLANT INJECTION AND RETURN FOR RC PUMP SEALS . ..... A-I

APPENDIX B-LOW PRESSURE INJECTION SYSTEM INTERFACE WITH HPI ..... ..... B-i

APPENDIX C-ELECTRICAL POWER REQUIREMENTS FOR HPIS

COMPONENTS . ............................................................... C-1

APPENDIX D-ENGINEERED SAFETY FEATURES ACTUATING SYSTEM

FOR HPIS ...................................................................... D-l

APPENDIX E-COMPONENT DESIGN INFORMATION ............................... E-1

APPENDIX F-MAKEUP/HPI NOZZLE CRACKING .................................. F-I

APPENDIX G-OPERATING EXPERIENCE DATA .................................... G-l

APPENDIX H-PROBLEMS WITH BORATED WATER SYSTEMS ...... ................. H-i

APPENDIX I-HPIS RISK ASSESSMENT DATA SUMMARIES .......................... 1-1

FIGURES

1. NPAR program strategy ................ .......................................... 3

2. Emergency core cooling system .................................................... 6

3. ECCS with makeup and purification system highlighted ........ ....................... 6

4. ECCS with RCP seal cooling system highlighted ........... .......................... 8

5. ECCS with high pressure injection system highlighted ......... ........................ 8

6. ECCS with high pressure recirculation system highlighted ........ ...................... 9

7. Representative ESF actuated systems .1............................................... ;3

x
8. HPI pump ..................................................................... 15

9. Failure data for HPIS motor-operated valves in 5-year increments ....................... 26

10. Components that contributed significantly to HPIS unavailability for emergency

modes of operation . .............................................................. 47

11. HPIS unavailability vs years showing the effect of aging for the three emergency

operating modes . ................................................................ 50

A-I. Part of the HPIS used during normal and emergency operations ........................ A-4

A-2. Reactor coolant makeup and purification system ..................................... A-5

A-3. RC pump seal return and injection system ........................................... A-6

D-l. Engineered safeguard system .......... I ............................................ D-4

D-2. ESFAS channel for initiating the high pressure injection with related aging and

engineering data . ................................................................ D-5

E-l. HPI pump characteristics . ......................................................... E-5

F-I. Makeup HPI nozzle (new and old design) ........................................... F-4

TABLES

1. HPIS components actuated by the engineered safety features actuating system ..... ....... 11

2. ECCS component failure ranking (NPE) ............................................ 16

3. ECCS failure causes (NPE) .............. .......................................... 17

4. Subcomponents causing ECCS system failures for PWRs (NPE) ........................ 18

5. HPI pump failure causes (LER) ................................................... 20

6. Data on HPI and other chemical volume control pumps (CVCS/HPI) (LER data) ..... .... 21

7. Cause for all system pump failures other than main HPI pumps (LERs) ..... ............. 22

8. Summary of HPIS valve failures ............ ........................................ 23

9. High pressure injection system totals and fractions .................................... 24

10. High pressure injection system component failure category fractions ..... ................ 25

11. Component failures in 5-year increments ............................................ 27

12. HPI plant data . .................................................................. 28

13. IIR summary of causes for HPIS failures ............................................ 29

xi
14. Data base ranking of most troublesome HPI components .............................. 29

15. Summary of aging processes for HPIS .............................................. 36

16. Test frequency for HPIS components ............................................... 38

17. HPIS performance testing ............... ......................................... 39

18. HPIS data for the risk assessment, aging acceleration factors, and component

unavailabilities in 5-year increments for upper and lower bounds ........................ 44

19. Events with significant Fussell-Vesely importance measures ............................ 48

20. Data for HPIS unavailability (one of three HPI pumps required) ........................ 49

21. Data for HPIS unavailability (two of three HPI pumps required) ........................ 49

22. Data for HPIS unavailability (recirculation mode) .................................... 50

23. Probability for type of events causing HPIS unavailability ............................. 52

A-I. Reactor coolant quality . ........................................................... A-7

C-l. HPIS pumps and valve electrical power requirements .................................. C-4

D-1. Engineered safeguards actuated devices for Channels I and 2 .......................... D-6

E-1. Engineered safeguards piping design conditions ...................................... E-3

E-2. High pressure injection pump data ................................................. E-4

E-3. Active-HPIS and LPIS reactor coolant pressure boundary valves ...................... E-6

E-4. Borated water storage tank ............... ......................................... E-6

E-5. HPIS regulatory requirements and guidelines ........................................ E-7

G-l. Valve failure ranking by function ................................................... G-3

G-2. I&C failures for HPI . ............................................................ G-3

G-3. Causes for pipe failures . ........................................................... G-3

G-4. Causes for pipe support failures ................................................... G-4

G-5. Subcomponents involved in tank problems .......................................... G-4

G-6. Summary of valve failures and command faults for HPIS and CVCS (LERs) ..... ........ G-4

I-l. Basic events for the high-pressure injection/recirculation system: hardware ..... .......... 1-4

1-2. Basic events for the high-pressure injection/recirculation system: human errors .... ....... 1-8

xii
I-3. Basic events for the high-pressure injection/recirculation system: maintenance ..... ....... 1-9

1-4. Fault tree: HPI importance measures report (5 year) .................................. 1-10

1-5. Fault tree: HPI importance measures report (10 year upper bound) ..... ................. 1-11

I-6. Fault tree: HPI importance measures report (10 year lower bound) ..... ................. 1-12

1-7. Fault tree: HPI importance measures report (40 year upper bound) ..... ................. 1-13

I-8. Fault tree: HPI importance measures report (40 year lower bound) ..... ................. 1-14

1-9. Fault tree: HP201 importance measures report (5 year) ................................ 1-15

I-10. Fault tree: HP201 importance measures report (10 year upper bound) ......... ........... I-16

I-11. Fault tree: HP201 importance measures report (10 year lower bound) ......... ........... I-17

I-12. Fault tree: HP201 importance measures report (40 year upper bound) ..... ............... 1-18

1-13. Fault tree: HP201 importance measures report (40 year lower bound) .................. .. 1-19

1-14. Fault tree: HPRI importance measures report (5 year) ................................ . 1-20

1-15. Fault tree: HPRI importance measures report (10 year upper bound) .............. . 1-21

1-16. Fault tree: HPRI importance measures report (10 year lower bound) .............. 1 -22

1-17. Fault tree: HPRI importance measures report (40 year upper bound) ................ . 1-23

I-18. Fault tree: HPRI importance measures report (40 year lower bound) ............. . 1-24

1-19. Fault tree: HPI cut sets quantification report (5 year) ............................... . 1-25

1-20. Fault tree: HPI cut sets quantification report (10 year upper bound) ......... ............ I-25

1-21. Fault tree: HPI cut sets quantification report (10 year lower bound) ................ 1-26

1-22. Fault tree: HPI cut sets quantification report (40 year upper bound) ......... ............ I-27

1-23. Fault tree: HPI cut sets quantification report (40 year lower bound) ............... .1 -28

I-24. Fault tree: HP201 cut sets quantification report (5 year) ............ .. ................. I-28

1-25. Fault tree: HP201 cut sets quantification report (10 year upper bound) ................ . 1-29

1-26. Fault tree: HP201 cut sets quantification report (10 year lower bound) .............. . 1-30

1-27. Fault tree: HP201 cut sets quantification report (40 year upper bound) ................ .1 -31

1-28. Fault tree: HP201 cut sets quantification report (40 year lower bound) ............... . 1-32

1-29. Fault tree: HPRI cut sets quantification report (5 year) ............. .. ................. I-33

xiii
1-30. Fault tree: HPRI cut sets quantification report (10 year upper bound) ..... .............. 1-33

1-3 1. Fault tree: HPRI cut sets quantification report (10 year lower bound) ..... ............... I-34

I-32. Fault tree: HPRI cut sets quantification report (40 year upper bound) ..... .............. 1-34

I-33. Fault tree: HPRI cut sets quantification report (40 year lower bound) ..... ............... I-35

xiv
ACRONYMS

ANSI American National Standards Institute

ASME American Society of Mechanical Engineers

B&W Babcock and Wilcox

BIT Boron Injection Tank

BNL Brookhaven National Laboratory

BWST Borated Water Storage Tank

CE Combustion Engineering

CFS Core Flood System

CM Corrective Maintenance

CVCS Chemical Volume Control System

ECCS Emergency Core Cooling System

EEI Edison Electric Institute

ES Engineered Safeguards

ESF Engineered Safety Features

ESFAS Engineered Safety Features Actuating System

gpm gallons per minute

HHIS High Head Injection System

HP High Pressure

HPI High Pressure Injection

HPIS High Pressure Injection System

HPP High Pressure Pump

HPRS High Pressure Recirculation System

HPSW High Pressure Service Water

I&C Instrumentation and Control

IS&M Inspection, Surveillance and Monitoring

IRRAS Integrated Reliability and Risk Analysis System

xv
IE Classification for electrical power for safety systems

IEEE Institute of Electrical and Electronic Engineers

IGSCC Intergranular Stress Corrosion Cracking

IIR Incident Investigation Report

INEL Idaho National Engineering Laboratory

IS&M Inspection Surveillance and Monitoring

LCO Limiting Condition for Operation

LDST Let Down Storage link

LER Licensee Event Report

LOCA Loss of Coolant Accident

LP Low Pressure

LPI Low Pressure Injection

LPIS Low Pressure Injection System

LPP Low Pressure Pump

LPRS Low Pressure Recirculation System

LPSW Low Pressure Service Water

LW Lower Bound

MCC Motor Control Center

MIC Microbial Influenced Corrosion

MU Make Up

MOV Motor Operated Valve

NPAR Nuclear Plant Aging Research

NPRDS Nuclear Plant Reliability Data Systems

NPE Nuclear Power Experience

O&M Operation and Maintenance

ORNL Oak Ridge National Laboratory

PRA Probabilistic Risk Assessment

xvi
PWR Pressurized Water Reactor

RC Reactor Coolant

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RWST Refueling Water Storage Tank

SI System Injection

SWS Service Water System

UHIS Upper Head Injection System

UP Upper Bound

USNRC United States Nuclear Regulatory Commission

W Westinghouse

xvii
NUCLEAR PLANT AGING RESEARCH

ON HIGH PRESSURE INJECTION SYSTEMS

INTRODUCTION

As part of its responsibilities to protect the public

health and safety, the USNRC is concerned with

the effect aging has on the safety of commercial

nuclear power plants. To meet this responsibility,

the USNRC has developed and implemented a

hardware-oriented research program to investigate

plant aging and the related degradation of components, systems, and structures. This program is

called the Nuclear Plant Aging Research (NPAR)

Program and is being conducted by the Electrical

and Mechanical Engineering Branch of the Division of Engineering of the office of Nuclear Regulatory Research. 1 A complementary program

focusing on pressure vessel, piping, steam generator materials problems, and nondestructive examination methods is being conducted by the Materials

Engineering Branch of the Division of

Engineering.

Aging and Plant Safety

The NPAR Program is investigating how the

aging of components, systems, and civil structures

can affect the safe operation of nuclear power

plants. The United States currently has approximately 100 commercial pressurized and boiling

light water reactors in operation. In the context of

NPAR, aging is defined as the "cumulative degradation that occurs with the passage of time in a

component, system, or structure." The main concern of the NPAR program is that plant safety

could be compromised if aging degradation is not

detected and corrective action taken before there is

a loss of the required functional capability in a

component, system, or structure. Consequently,

aging might result in a reduction in the safety level

achieved by the defense-in-depth approach used to

ensure the safety of domestic reactors. Defense-indepth requires that the public is protected from the

accidental release of fission products by a series of

multiple barriers and engineered safety systems.

Operating plant experience provides examples

where age induced degradation of a key component

has led to a reduction in the capability of a barrier

to prevent the release of fission products. These

examples include degradation of valves and pipe

cracks.

Age degradation can also cause a loss of operational readiness of engineered safety systems. The

engineered safety systems are designed to mitigate

the consequences of failure of a vital component,

system, or physical barrier, such as a loss of main

feedwater or a break in the primary system boundary. These systems are also designed to mitigate the

effects of events ranging from anticipated operational transients such as loss of offsite power to low

probability occurrences such as design basis seismic

events. Failures have occurred in systems such as

the auxiliary feedwater system and in the emergency diesel generators used to supply vital ac

power to the IE Power System.a

Aging can also lead to a higher probability of

common mode failure. Aging can result in wide

scale degradation of a physical barrier or to simultaneous degradation of redundant components.

One example of this is a simultaneous degradation

of the redundant valves designed to isolate the reactor coolant lines to a PWR. If this were to occur, a

failure in the piping outside the containment could

lead to an uncontrolled release of the primary coolant and radioactivity outside of the containment.

NPAR Program Goals and

Strategy

1. Identify and characterize aging and service

wear effects associated with electrical and

mechanical components, interfaces, and

systems likely to impair plant safety.

2. Identify and recommend methods of

inspection, surveillance, and condition

monitoring of electrical and mechanical

components, and systems that will be

effective in detecting significant aging

effects before loss of safety function so

a. IE is the classification given for all the electrical power for

nuclear plant safety systems.

1
that timely maintenance and repair or

replacement can be implemented.

3. Identify and recommend acceptable maintenance practices that can be undertaken

to mitigate the effects of aging and to

diminish the rate and extent of degradation caused by aging and service wear.

The NPAR Program uses a two-phased approach

to conduct aging research on the risk significant

components and systems in light water reactors; as

illustrated in Figure 1. The first stage, Phase I

makes use of readily available information from:

public and private data bases, vendor information,

open literature, utility information, and expert

opinion. The Phase I analysis includes a review of

three elements:

1. The hardware design, operating environment, and performance requirements

2. A survey of operating experience

3. The current methods used for inspection,

surveillance, monitoring, and maintenance and for qualifying end-of-life

performance.

The results of the Phase I evaluation include an

identification of actual and potential failure

modes; a preliminary identification of failure

causes due to aging and service wear degradation;

and a review of current inspection, surveillance and

monitoring practices, standards and guides. The

Phase I evaluation is used to decide if a Phase II

evaluation is warranted. If a Phase II evaluation is

needed, recommendations are developed to identify the detailed engineering tests and analyses to be

conducted in Phase II and which will result in

improved industry standards, guides, and

practices.

A Phase II assessment includes developing and

validating advanced inspection, testing, monitoring and maintenance methods. This development

includes both laboratory and field testing to verify

candidate technologies. Phase II may also include

examining and testing naturally aged components

from operating power plants and developing service life prediction models.

With the completion of the Phase I and Phase II

aging assessment research, a technical basis will be

available for use in the regulatory process. The key

end uses are shown in Figure 1. The uses envisioned

for the NPAR program results include: implementing improved inspection, surveillance, maintenance, and monitoring methods; modifying

present codes and standards; developing guidelines

for plant life extension; and resolution of generic

safety issues.

Phase I Aging Assessment of a

PWR High Pressure Injection

System

This report describes the results of an in-depth

Phase I evaluation of the High Pressure Injection

Systems (HPIS) used in light water reactors.

The study was performed at the Idaho National

Engineering Laboratory (INEL) and addresses the

system aspects of the HPIS and the materials susceptible to aging in components associated with the

HPIS. Certain components, such as valves and

pumps, have been extensively studied at other

national laboratories as part of the USNRC aging

and equipment qualification programs.2-4 Operating experience from the generic data bases and

plant records on the HPIS are complemented by

data from these component studies where applicable. Specifically, Phase I NPAR component studies

on valves and pumps are being pursued at the Oak

Ridge National Laboratory (ORNL). The valves

are the HPIS component category with the most

failures. Each motor operated valve and pump is

controlled by a motor control center (MCC). The

MCC (and electric motors in general) are currently

being studied at the Brookhaven National Laboratory (BNL) as part of the NPAR program. Data

from these component studies are also used in supporting the system studies. In addition, the Engineered Safety Feature Actuating System (ESFAS)

aging study performed at the INEL is directly

applicable to the HPIS because it provides the actuating signal for initiation of the HPIS operation

under accident conditions.5

The strategy for this HPIS study follows the

NPAR guidelines. Generic data bases are used to

get statistical data on which HPIS components

have experienced the most failures and the causes

for failures. Plant specific data supplied by a cooperating utility includes design descriptions, drawings, maintenance records, and personnel

interviews. (The plant specific data, of course,

applies to one plant considered to be typical for

those plants that use the HPI pumps for both normal operation to supply makeup water and for

emergency injection.) Information sources used

include: the Nuclear Power Experience (NPE) data

base, Licensee Event Reports (LERs), Nuclear

Plant Reliability Data System (NPRDS), IE

2
Phase I Phase II Utilization

a I~~~~llP~~~t~~~r~~g~~r~~rl i A .W 1W ,(,)~~~~~~~,

* I~~~~~~~~~~~~~~~~~~~~~~~~~~~aa

IPROS. NPRDS. LERs. FucinlInteractions of oferc

NPE. aet. F a Value with Code S a ety

Knowl edgen: e i : hv arl P her rm ne 1 jmpctIsue

K Exps e t 1.5. D Lice National nsing and Standards ) Indicators in a and es Inspection Paraormrln u Standards Experience Par [ Jr I Committees

I i / \ I \ i a

* a C~~~~~~~~~~~~~~~~xonsider-

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* * Extens~~~~~~~~~v once

I / X i~~~~~~~~~~~~~~~~~~~~~~~~ I \ii

Select~~~~~~~~PoslPhsl

Equinao nterimAsssM Comprehesive Aging r cation e n )

Defineperatng Aging Characteri- ofC I Assessment and 

Boundary E -rec xteotlonsl and U Ihods Review ond Appi Recommendations erchn fo r v o an of1. an UM1.. and MMAi

Figure 1. NPcR pmm en-

* I~~~~~~~~~~~~~~~~~~~~~~~

a I~~~~~~~~~~~~~~~~~~~~~~~~~~

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Disvc vem

* ~~~~~~~~Materials

* ~~~~~~~and

Designs

Priortlzaton Exerts Analysis of Ijethads and Applied Research

Pr~~or~~t~zalDesi n Tachmnolo y for , Pat2erVICe eiwad Itrctos Itrcin Study of noldge pcfctos-Iseto xmnto, R er~~a~n wt

Sstm s andu -O Peramtiona - Survilln.ee : Lab Tests. and ot Advanced Industry NRC Staff - Rik Indusryc Paamtrs -Mantnnco Analyses Mehd -Aging Prcie - Ongoing - Artificial/ Accel- i MtosST- 0250-al

Research crated Aging

Figure 1. NPAR program strategy.
notices and bulletins, plant-design information

and specifications, operation and maintenance

(O&M) manuals and procedures, historical

records, site-event records, and site interviews with

maintenance personnel.

The specific objectives of this study are:

1. Evaluate the overall operating experience

to determine if aging-related operational

problems have developed.

2. Use specific examples from a representative Babcock and Wilcox (B&W)

pressurized water reactor (PWR) to illustrate the functions of the HPIS and evaluate specific problems related to aging.

3. Perform screening type aging assessments

of the impact of aging on operability. The

assessment will focus on identifying:

a. Failure modes and causes

b. Materials susceptible to aging

degradation

c. Stressors during operation

d. Functional indicators that would aid

in failure prediction

e. Methods for detection and control of

aging degradation.

4. Review and provide recommendations for

inspection, surveillance, and monitoring

(IS&M), as well as advanced methods for

IS&M.

5. Evaluate the role of maintenance in counteracting aging effects to include the

following:

a. Survey and evaluate currently used

maintenance practices that counteract

aging and service wear effects

b. Evaluate relative benefits of preventive and corrective maintenance

c. Identify potential mechanisms causing component system degradation

through improper maintenance

d. Provide recommendations for preferred maintenance practices.

6. Perform an unavailability assessment of

the HPIS to determine which components

contribute significantly to the HPIS

unavailability and how aging affects

unavailability.

Other work at the INEL related to this HPIS

aging study includes the reactor protection system

aging study,5 the reported failure cause study of

component failures for selected systems, 6 and the

development of technical criteria for use in assessing the residual life of the major light water reactor

components. 7 The reported failure cause work

identified safety systems significantly affected by

aging phenomenon (of which HPIS is included).

Although many component failures were identified

in the reported failure cause work, actual HPIS

failure occurred as a result of only 0.7% of the

component failures. This is due to control channel

redundancy and priority maintenance.

The description of the HPIS for the representative PWR is given first. This is followed by a review

of the operating experience section, which provides

information from the various data bases. HPIS

safety issues and potential aging problems are discussed next followed by aging assessments. Then a

review of HPIS IS&M and the role of maintenance

is discussed. A section on the HPIS unavailability

assessment identifying risk significant components

is the last section just before the conclusions.

4
SYSTEM FUNCTIONAL DESCRIPTION

The HPIS along with the Low Pressure Injection

System (LPIS) and the Core Flooding System (CFS)

collectively form the overall Emergency Core Cooling

System (ECCS), which is designed to prevent core

damage from a loss-of-coolant accident (LOCA).

High pressure injection is necessary to prevent uncovering of the core for small LOCAs, where high system

pressure is maintained, and to delay uncovering of the

core for intermediate sized LOCAs. The HPIS can also

be used to cool the core following a non-LOCA reactor

shutdown (e.g., transient). This mode of HPIS operation would be utilized only if normal and emergency

secondary heat removal via the steam generators cannot be achieved.

Commercial nuclear power plants have various

designs for HPIS in regard to boundaries, function,

and terminology. A typical Westinghouse 4-loop plant

uses accumulators [sometimes referred to as the Upper

Head Injection System (UHIS)] as the immediate

response system performing an ECCS function if the

reactor coolant system (RCS) pressure drops. When

system injection is called for, the boron injection tank

(BIT) subsystem is valved into the charging system to

supply borated water to the RCS. This BIT injection in

independent of the UHIS and is a HPIS function performed by the charging system to supply borated water

to the RCS. The system injection (SI) signal also starts

the two HPIS pumps and aligns both the HPIS and

charging systems to take suction from the refueling

water storage tanks (RWST). The HPIS, sometimes

referred to as the High Head Injection System (HHIS),

injects borated water to the RCS after the system pressure drops to 1500 psi. Both the HPIS and charging

system can be aligned to the residual heat removal system which takes suction from the containment sump.

Westinghouse 3-loop designs use an accumulator

system for an immediate borated water injection system and uses three pumps that perform the high pressure injection function including the BIT insertion.

One of the pumps is also used for normal charging.

The charging and high pressure injection function is

similar to the B&W system except that Westinghouse

uses separate injection nozzles.

The Combustion Engineering designs have three

HPIS pumps used in the emergency mode only and a

separate charging system. The B&W system which is

exampled in this report consists of three motor-driven

high pressure centrifugal pumps, with two primary

suction and discharge paths. One of the three pumps is

also used for supplying makeup water during normal

operation. The detailed system configuration is shown

in Figure 2. The HPIS and related systems perform the

following functions for the B&W type system:

1. Maintain the Reactor Coolant system (RCS)

inventory during normal operation

2. Maintain proper RCS water chemistry and

purity

3. Control RCS boric acid concentration

4. Provide fill and makeup for the core flood

tanks

5. Provide seal injection water for the reactor

coolant pumps

6. In the event of an RCS accident, provide

high pressure injection of borated water

for emergency core cooling and plant

shutdown

7. Provide long term core cooling following a

LOCA using the high pressure recirculation system and low pressure recirculation

system.

The first four items can be combined into one

system called the makeup and purification system

to maintain the volume of the reactor coolant system (RCS) within acceptable limits during most

modes of plant operation. It also recirculates reactor coolant for purification, addition of chemicals

for the control of RCS corrosion, and the control

of soluble boron concentration for long term reactivity control.

For the purpose of this report, the HPIS configured for the emergency injection mode will be of

primary interest. However, parts of the system are

shared for RC pump seal cooling, RC makeup and

purification, as well as the high pressure recirculation mode. Each of these configurations is briefly

discussed, as necessary, to cover the functions of

the HPIS shared components.

Makeup and Purification System

The makeup function is achieved primarily by a

portion of the HPIS and the coolant storage and

chemical addition systems. Makeup flow is supplied by either pump HPP-A or HPP-B (Figure 3)

and is controlled automatically to balance normal

leakage. Letdown flow from the RCS accommodates small increases in RCS volume due to

inleakage from the seals of the reactor coolant

pumps (RCPs) and variations in RCS temperature.

The system was not designed for emergency

5
Figure 2. Emergency core cooling system.

Sump

Figure 3. ECCS with makeup and purification system highlighted.

7.8159

6
operation; however, it does provide RCS inventory

control during most transient conditions other than

loss-of-coolant accidents. If the system is not capable of meeting the requirements for inventory control after a reactor trip, manual action can be taken

to start additional HPI pumps, establish additional

discharge paths to the RCS, or align the borated

water storage tank (BWST) for assurance of a sufficient suction source. Figure 3 highlights the ECCS

makeup and purification system. More detail system information is given in Appendix A.

Cooling System for RCP Seals

Seal injection flow is provided by the HPI pump

operating to supply normal RCS makeup. These

seals prevent the leakage of reactor coolant between

the shaft and the housing of the RCPs. When the

RCS is at a high temperature, the seals must be

cooled to keep them from warping, to keep the seal

faces from becoming cracked or eroded, and to prevent the O-rings from extruding. The interruption

of cooling flow can result in seal damage, leading

to increased RCS leakage or small break LOCA

conditions.

The ECCS, with the Seal Cooling System highlighted, is shown in Figure 4. Details of the RCP

seal cooling system are given in Appendix A.

Emergency Injection Mode of the

HPIS

The HPIS provides emergency core cooling in

the event of a small break LOCA, and it also provides an alternative means of core heat removal if

the ability to cool via the steam generators is lost.

Figure 5 illustrates the HPIS configuration. For

most event sequences, flow from one pump is sufficient; for special cases, two pumps may be

required. The HPIS is capable of supplying flow at

a relatively high RCS pressure, with a shutoff head

of 2900 psig (other types of systems have shutoff

heads less than normal system pressure). The emergency mode of operation is initiated if the RCS

pressure decreases to 1500 psig or if Reactor Building pressure increases to 4 psig. Under these condi--

tions, the following actions are automatically

initiated:

1. Three isolation valves in the purification

letdown line close (HP-3, HP-4, HP-5),

and two isolation valves in the seal return

line close (HP-20 and HP-21)

2. One inlet valve in each injection line opens

(HP-26, HP-27, HP-409, and HP-410)

3. Two valves in the lines to the borated water

storage tank outlet header open (HP-24

and HP-25)

4. All high pressure injection pumps start.

The emergency high pressure injection flow path

is from the borated water storage tank through the

high pressure injection pumps and into both reactor coolant loops. The emergency mode of operation will continue until manually terminated.

High Pressure Recirculation

Mode

The High Pressure Recirculation System (HPRS)

is one of two systems designed for long-term core

cooling following a LOCA. The other system is the

Low Pressure Recirculation System (LPRS). After

exhaustion of the BWST, the LPRS and HPRS are

used to recirculate water from the containment

sump to the RCS. If the LOCA is large enough, the

RCS will be at a low enough pressure so that only

the LPRS would be required. If the LOCA is small,

however, the RCS will be at a pressure above the

shutoff head of the LPRS pumps and the HPRS

would be required. For small break LOCAs, the

HPRS and the LPRS are both required, because

the HPRS takes its suction from the discharge of

the LPRS. This realignment is shown in Figure 6.

In this configuration, the LPIS takes its suction

from the reactor building sump through valves

LP-19 and LP-20. The discharge for the HPRS is

through the HPIS nozzles into the reactor coolant

system.

Any one pump is capable of providing enough

flow to prevent core damage for those smaller leak

sizes that do not allow the RCS pressure to decrease

rapidly enough to the point where only the LPRS is

required. One high pressure line can deliver

450 gpm at 1800 psig reactor vessel pressure. One

of the three high pressure pumps is normally in

operation and a positive static head of water

ensures that all pipe lines are filled with coolant.

The high pressure lines contain thermal sleeves at

their connections into the reactor coolant pipe to

prevent thermal stressing at the pipe juncture.

All three pumps have self-contained lubrication

and mechanical seal coolant systems tied in with

the Low Pressure Service Water System (LPSW).

7
7.8158

Figure 4. ECCS with RCP seal cooling system highlighted.

Sump 7-8157

Figure 5. ECCS with high pressure injection system highlighted.

8
Sump

7-8160

Figure 6. ECCS with high pressure recirculation system highlighted.

In the representative plant studied, the high

pressure recirculation mode is manually initiated by the following operator actions:

1. BWST supply line valves HP-24, HP-25,

LP-21, and LP-22 are closed when the

BWST low level alarm notifies the operator. All high and low pressure ECCS

pumps are also shut off at this point

2. Containment sump valves LP-19 and

LP-20 are opened and the low pressure

pumps LPPA and LPPB are restarted

3. Valves LP-15 and LP-16 are opened in

order to divert a portion of the LPRS

flow to the high pressure pumps that are

also restarted.

Appendix B contains more information on the

LPIS as it relates to the HPIS.

After initiating the HPRS, the operator continues to

control the system in the recirculation mode. Tb aid the

operator, the following system conditions are monitored and displayed in the control room: the reactor

building sump level, the temperatures of water in the

line from the sump to the low pressure pumps, the low

pressure pump discharge pressure, the flows in the low

pressure and high pressure supply lines to the reactor

vessel, the level in the BWST, and all motor-operated

valve positions.

9
INTERFACES WITH SUPPORTING SYSTEMS

The supporting systems are those systems

required for the HPIS to perform its function. Failure in a support system can affect the operation of

HPIS.

ice water system. The LPSW system also supplies

cooling flow to the heat exchangers of the component cooling system.

Instrument Air

Electrical

Electric power is supplied to the three HPIS

pumps by three independent 4160 volt buses. The

motor-operated valves (MOVs) also require ac

motive power. Emergency power is available to critical components in the event that the normal power

source is lost. Control power for the HPI pumps is

provided from the dc power system. Control power

for all other electrical components is derived from

the same source as the motive power. The HPIS

components and their power supplies are listed in

Appendix C.

Following a loss-of-coolant accident, assuming a

simultaneous loss of normal power sources, the emergency power source and both the LPIS and HPIS will

be in full operation within 25 seconds after actuation.

All calculations for the representative plant studied

have assumed a 25 second delay from receipt of the

actuation signal to start of flow for both the HPI and

LPI systems. Upon loss of normal power sources

including the startup source and initiation of an engineered safeguards signal, the 4160 volt engineered safeguards power line is connected to the emergency power

source. The emergency unit will start up and accelerate

to full speed in 23 seconds or less. An analysis has

shown that by energizing the HPI and LPI valves

(which have opening times of 14 seconds and 15 seconds respectively at normal bus voltage) and pumps at

less than 100% voltage and frequency, the design injection flow rate (HPI - 450 gpm, LPI - 3000 gpm) will

be obtained within 25 seconds.

Service Water

Cooling water for the HPI pump motors is provided by the LPSW system. Backup cooling flow

can be made available by local manual action from

the elevated storage tank of the high pressure servThe instrument air system supplies motive power

to the number of valves required for HPIS to function in the normal makeup and RCP seal cooling

modes. Upon loss of instrument air, all the pneumatic valves transfer to or remain in the closed

position with the exception of control valve

(3HP-31), which opens fully for seal injection flow

to the RCPs.

ECCS Pump Room Coolers

New plants (0-5 years old) for all four U.S. NSSS

vendors have pump room coolers. Older plants

(5-15 years old) may or may not have pump room

coolers. Plants older than 15 years do not have

them.

Engineered Safety Features

Actuating System

The HPIS is one of the Engineered Safeguards

(ES) Systems that is automatically actuated by the

ESFAS. The aging study on ESFAS is covered in

Reference 5.

For normal operation, automatic control signals

are supplied for two flow control valves. They are

the normal makeup flow control valve, HP-120,

and valve HP-31 for RC pump seal flow control (see

Figure 2).

During emergency conditions ES signals are provided to components in the HPIS that must change

state. The components required for function of the

HPIS receive signals from the ESFAS when the

RCS pressure is low (1500 psig), or when the reactor building pressure increases to 4 psig. The

ES-actuated HPIS components as well as the ES

channel doing the actuation are listed in Table 1.

Appendix D describes the ESFAS system for the

HPIS in greater detail.

10
Table 1. HPIS components actuated by the engineered safety features actuating system

Component

Pump HPP-A

Pump HPP-B

Pump HPP-C

MOV HP-24

MOV HP-25

MOV PH-26

MOV PH-27

MOV HP-20

AOC HP-21

MOV HP-3

MOV HP-4

AOV HP-5

ES Actuation

Channel

1

1&2

2

1

2

I

2

1

2

5

6

2

Normal Channel

On/off

On/off

Off

Closed

Closed

Closed

Open

Open

Open

Open

Open

Open

ES Status

On

On

On

Open

Open

Open

Open

Closed

Closed

Closed

Closed

Closed

II
SYSTEM COMPONENTS AND HARDWARE

All HPIS piping and hardware components are

made of stainless steel which is resistant to corrosion from boric acid used in the borated water. The

major components are discussed in this section.

More detailed aging related design information on

these components is given in Appendix E.

Valves

For the purposes of this report, a valve is defined

as the valve body and all its internal parts, the valve

operator (motor, solenoid, hand wheel, etc.), and

any limit and torque switches mounted on the valve

body or operator needed to make the valve function. The HPIS uses many types of valves and valve

operators including motor-operated valves (MOV),

pneumatic-operated valves, solenoid-operated

valves, manual-operated valves, check valves, and

safety relief valves. The component NPAR studies

have extensively covered aging of valves (see References 2 and 3).

Valve failures and how they affect system operations is an important consideration in this research.

The failure modes that can affect system operation

include failure to open, failure to close, internal

leakage, external leakage, and plugged. General

failure mechanisms that exist independent of valve

type include normal wear, excessive wear, corrosion, foreign material contamination, and excessive

vibration. Control valves such as the HPI flow control valves can also have a failure mode in which

they fail to operate as required. They are designed

to constantly change position during operation. A

system failure occurs when a valve has lost the ability to control system parameters.

Types of valves are discussed further in

Appendix E and valve failures are covered in a later

section on Operating Experience.

Instrumentation and Control

In general, there are three variations of the engineered safeguards control stations as shown in Figure 7. The first station is a controller and monitor

for a motor-operated valve automatically operated

by an ESFAS signal. Feedback information on the

valve position is provided by a limit switch. The

second station is a pump motor control station that

has inputs from either the ESFAS or main control

room. Instrumentation for the HPI pumps

includes local indication of pump discharge pressure and suction pressure. Discharge flow through

the pump crossover lines, discharge pressure, and

low pressure are also provided to the control room.

The third type of control station is one that controls

an air operated pilot valve to control the pneumatic

valve. Two stations are shown on this drawing

meaning that the valve can be controlled from two

different locations. The "C" in the triangles on

these drawings indicates computer monitoring for

control room display.

The operator can override the automatic flow

control on HPI by adjustment of the flow controller so the flow rate may be reduced to match the loss

of coolant from the vessel when small line breaks

occur.

HPI Pump A Controls. The HPI pump A is normally controlled from the control room with a

manual four position (start-run-off-auto) switch.

In the start position, power is supplied to the pump

circuit breaker closing coil and pump starts. In auto

position, the pump automatically starts on low seal

injection flow, or loss of voltage on the main feeder

bus as sensed by the main feeder Bus Monitor.

When in the off position, the control switch energizes the circuit breaker trip coil and the pump

motor is de-energized.

In the event of a reactor accident, the ESFAS

channel I will automatically start HP injection

pump A regardless of its manual control switch

location. If the control switch was in the off position when the start occurred, the pump will continue to run after the ESFAS signal is reset until the

manual pump control switch is cycled out of the off

position and back.

After an ESFAS signal is initiated, the automatic

control can be overridden by pressing the manual

pushbutton in the control room. This override can

be done only when a safeguards trip signal, or test

signal is present. When in the manual mode, the

pump control then functions as if no ESFAS signal

is present. Control is returned to the automatic

mode for safeguards when the auto pushbutton is

operated in the control room or if the ESFAS trip

signal is cleared at the safeguards cabinet.

HPI Pump B Controls. The controls for pump B

are essentially the same as those provided for pump A

except that in the event of an accident it receives an

automatic start signal from both ESFAS channels 1

12
ES command

!Engineered 4

Safeguards n C

Control Station Auto &

It ~~~~~~~~~~~~~~Contro l( 

l ~~~~~control ndica tonUs~~~~~Nra Control)

l ~~~~~controlIndication 4

_Safeguards A, cond sol_4 |Control Station Auto coto im

l Indication I Indicallon I Il Indlcatlon I4 Inialo

I_____

7*1904

Figure 7. Representative ESE actuated systems.

13
and 2. In order to take manual control of pump B

when an ESFAS signal is present, both channels I

and 2 must be put into the manual mode.

HPI Pump C Controls. The controls for HPI

pump C are essentially the same as for pump A

except that pump C is not used for normal operation. In the event of a reactor accident, HPI

pump C receives an automatic start signal from

ESFAS channel 2.

Motor-Operated Valves. The ESFAS provides

automatic operation for the valves listed in Table 1.

After an ESFAS signal is initiated, the automatic control signal can be overridden by manual pushbutton.

While in the manual mode the valve is controlled as if

no ESFAS signal is present. Control is returned to the

automatic mode for safeguards when the auto pushbutton is operated or when the ESFAS trip signal is

cleared at the safeguards cabinet. Manual controls are

provided in the control room and safeguards cabinet.

Pneumatic-Operated Valves (HP-5 and HP21). The pneumatic-operated valves are controlled by

a 125 Vac solenoid operated valve that controls the air

line operating the pneumatic valve. Manual control is

provided in the control room and automatic operation

is provided by ESFAS channel 2. A manual override

similar to the MOV and HPI pump controls is provided. Interlocks in the control circuit will close HP-5

in the event of an excessively high letdown temperature

or low instrument air pressure. Similarly, valve HP-21

is interlocked to close on low instrument air or if all

four of the individual RC pump seal return valves are

closed.

HPI Pumps

The HPI pumps are vertical multiple-stagecentrifugal pumps with mechanical seals. The wetted

parts of the pumps are stainless steel. Figure 8 is a picture of a HPI pump. Operation of the HPI pumps

requires lubricating oil cooling and pump seal cooling.

The HPI pump cooling is accomplished via the LPSW

systems where heat generated in the pump lubricating

oil and seals is removed via heat exchangers. The HPI

pump data is given in Appendix E.

than those encountered during emergency operation. Pipe sizes range from I to 14 in. Design pressures and temperatures for piping are given in

Appendix E.

Nozzles and Thermal Sleeves

The high pressure injection/makeup (HPI/MU)

nozzles with thermal sleeves are located on all four cold

legs of the reactor coolant piping. They provide emergency core cooling and normal makeup flow to the

primary coolant system. In general, one or two of the

lines are used for both HPI and MU, while the remaining nozzles are used for HPI alone. The thermal sleeves

are incorporated into the nozzle assembly to provide a

thermal barrier between the cold HPI/MU fluid and

the hot HPIS nozzle. This prevents thermal shock and

fatigue of the nozzle. See Appendix F for a discussion

of nozzle cracking.

Piping Penetrations

The reactor building penetrations for HPI lines

associated with the representative plant studied

include the following: the letdown from the reactor coolant system to the demineralizers is a 2.5 in.

line with remotely controlled valves on the inside

and outside of containment for isolation control;

two (1 in.) nozzle warming lines with check valves

for isolation control; a normal makeup inlet line

(4 in.) having a check valve on the inside of containment; two seal injection lines that supply RC

pump seal cooling (2.5 in.) with a check valve

inside of containment; and the emergency injection

line (4 in.) with a check valve on inside of containment. The penetrations must be able to maintain

reactor building pressure seal. Reactor coolant isolation under accident conditions is maintained by

valves on the inside and outside of the

penetrations.

Pipe Hangers

Pipe hangers are rigid carbon steel supports for

the various HPI piping. Their purpose is to resist

the dead weight loads of the piping and water.

Snubbers

HPI Piping

The HPI piping is stainless steel and designed for

normal operation. The normal operating system

temperature and pressure requirements are greater

Snubbers move freely at low acceleration and

lock up at higher acceleration, to provide support

for seismic and other dynamic loads. They may be

mechanical or hydraulic.

14
Tanks

The borated water storage tank is the source of water

for the emergency HPIS. It is a 350,000 gallon coated

carbon steel tank. For the representative plant studied

it is considered part of the low pressure injection system because it also supplies borated water for LPIS

and containment spray systems.

Figure 8. HPI pump.

15
OPERATING EXPERIENCE

The HPIS operating experience is based on

information from generic data bases, plant records

from the representative PWR studied, and other

sources such as USNRC information notices and

bulletins for specific problems. Information was

taken from the NPE, LER, and NPRDS generic

data bases. These data bases were adequate for

drawing general conclusions about the ECCS with

emphasis on the HPIS. The system boundaries

used were those indentified by these data bases.

The discussion of each data base identifies the

components or the subsystem covered by that data

base. Although the components and subsystems

that support the HPIS function are generally

included as part of the HPIS, actual high pressure

exists only in that part of the system from the HPI

pumps to the injection nozzles. All the components

that are included in the HPIS are important to its

service and therefore, are included in this study.

Nuclear Power Experience Data

The NPE automated retrieval system was developed and introduced by the S. M. Stoller Corporation at Boulder, Colorado. This system contains

information on nuclear plant components available

from the public domain. The index and key words

are computerized, allowing a rapid search of the

system for specific articles with titles and reference

numbers to hard copy volumes. The system is

updated quarterly and is a convenient source for

obtaining generic information on problem areas.

The NPE data is summarized in this section. More

detailed data from NPE on the various HPI and

associated ECCS components are given in Appendix G.

- The NPE data base includes the following ECCS

subsystems: safety injection, high head injection,

upper head injection, boron injection, recirculation phase, containment spray, and accumulator

tank. All the above systems are part of the HPIS

except containment spray. The containment spray

system could not be conveniently removed, thus, to

indicate the data summaries from the NPE also

include containment spray, they are called ECCS

summaries. The components in these subsystems

are similar in that they have similar operating environments and must be compatible with borated

water.

For U.S. Nuclear Power Plants, from startup

through 1986, there were 1552 articles on the ECCS

for PWRs. Component failure events listed in order

of frequency of occurrence for these PWRs is given

in Table 2. Valve failures are listed most frequently

(35°%), followed by I&C (19%), pumps (15°%), and

pipes (7%). Pipes in this case include nozzles and

penetrations.

The ECCS failure causes are listed in Table 3.

The top three causes in order of frequency of occurrence were maintenance error, design error, and

mechanical disability. This was followed with. 4°0%

each for local I&C, set point drift, chemistry out of

spec, and subcomponent sticking. The failure

causes that are considered potentially aging related

and identified in Table 3 account for 28% of the

failures. The cause is identified only as potentially

aging related because the root cause can not always

determined from the reported events.

After eliminating human and procedure errors, a

breakdown of subcomponents causing ECCS failure is given in Table 4 for PWRs. Moving internal

parts for valves, pumps, and motors accounted for

15°10 of the problems. Just a little over half of these

were caused by valve stem or disc seating problems.

Instrumentation accounted for 12% of failures and

control 11°0% of failures. The drive sources for most

of the valves and pumps are electric motors which

accounted for 66% of drive/actuator failures.

Table 2. ECCS component failure ranking

(NPE)

PWR

Component (%0)

Valves 35

Instrument and control 19

Pumps 15

Pipe 7

Electrical 4

Heat exchanger 2

Other 18

Number of events 1,552

16
Table 3. ECCS failure causes (NPE)

PWR

Cause .(No)

Maintenance error. 28

Design error 13

Mech. disabilitya- 10

Local I&C failurea 4

Setpoint drifta 4

Chemistry out of spec 4

Sub-comp. stickinga 4

Short/grounda 3

Weld failurea 3

Blockage 3

Other 24

Number of events 1,552

a. Potentially aging related failures in 28%.

: 17
Table 4. Subcomponents causing ECCS system failures for PWRs (NPE)

Subcomponent

Moving Internal

Stem/disc seating

Blade/impeller

Coupling shaft

Bearing/bushing

Piston/diaphragm

Instrumentation

Percent (a) Subcomponent Percent (a)

15 Drive/Actuator 8

(52)

(9)

(16)

(14)

( 9)

12

Pneumatic air

Hydraulic

Motor

Other

Chemistry

Relay/breakers

Wire cable

Connectors

Seal/gasket

Fitting/flange

(10)

(3)

(66)

(21)

Bistables

Transmitters

Sense lines

Indicator

(33)

(25)

(25)

(17)

7

6

5

5

5

4

Control

Limit SW

Solenoid valve

Torque SW

Other

11 External Support

(14)

(1 1)

(23)

(52)

Mounting/fastener

Body/casing

Other misc.

4

4

2

12

a. Number in "parentheses" is 5o of number in left column heading.

18
Electrical subcomponents-relays, breakers, cable,

and connectors-accounted for 16%.

LER Data

The Code of Federal Regulations (10 CFR 50.72

for occurrences before 1984 and 10 CFR 50.73 for

events after January 1, 1984) require nuclear

power plants to report significant events to the

USNRC. The pre-1984 LER data base has been

used as a source of reliability data. Events reported

to the LER system after January 1, 1984 are only

those that are, or lead to, safety-significant events.

If a component fails and can be replaced within the

time constraint of the limiting condition for operation (LCO), no LER is required. This limited

reporting would not be expected to provide an adequate representation of failure experience; therefore, only the LERs prior to 1984 were used in this

aging study. Data from licensee event reports on

HPI pumps and valves from LERs are presented in

this section. The LER data covers HPI and CVCS

valves and pumps for B&W, W, and CE plants.

Pump Failures. A total of 44 events associated

with HPI pumps were reported in the LERs for all

PWRs from January 1, 1972 to September 30,

1980.8 For the HPI pumps, the leading failure

cause was control malfunction with 14 failure

events out of a total of 44. The next most frequent

causes were maintenance and design error with four

events each. This is followed by three events each

for operation error, failed internals, and unknown.

These HPI pump failure causes are summarized in

Table 5.

The failure mode experienced most often was

does not start for 24 of the 44 events. This was followed by does not continue to run (8 events) and

loss of function (7 events). Most of the problems

(25) were discovered by performance testing.

TWelve were discovered by normal operations (See

Table 6).

System Pumps. The system pump category covers all HPIS/CVCS pumps apart from the main

HPI pumps, (i.e., charging pumps, makeup

pumps, etc.). There were 130 events reported where

the leading cause was seal or packing failure (29),

followed by control malfunction (13), loss of pressure boundary (12), and drive train failure (9).

Table 7 ranks these causes by number of events.

Valves. A summary of HPIS valve failure from

LER data for 1976 through 1980 is given in

Table 8.9 About 42% of the valve failures were

classified as potentially aging related. In this category, mechanical controls (parts failed or out of

adjustment) were the most frequent with 8%, followed by seat or disk failure 6.9%, and packing

failure 5%. It is interesting to note that command

faults (faults due to power source, controls or supporting systems) caused 32% of HPIS valve failures. See Appendix G for additional LER data on

valves.

Nuclear Plant Reliability Data

System

The NPRDS was developed by the Equipment

Availability Task Force of the Edison Electric Institute (EEI) in the early 1970s under the direction of

the American National Standards Institute

(ANSI). The NPRDS was maintained by the Southwest Research Institute under contract to the EEI

through 1981. Since January 1982, the NPRDS

has been under the direction of the Institute of

Nuclear Power Operation. The components covered in the NPRDS for HPIS are announciators,

circuit breakers, safety function instruments,

motors, pumps, valves, and valve operators. For

Westinghouse plants upper head injection, the

Boron Injection Tank along with associated valves

is also included. For B&W plants, filters are

included as well as letdown, purification, and also

CVCS components common to the CVCS and

HPIS systems.

The NPRDS data for all Westinghouse and

Babcock and Wilcox plants were compiled for the

HPIS and the aging fraction determined. Combustion Engineering data was unavailable from

NPRDS at the time these data were compiled.

Aging fraction is the ratio of aging-related failures

to total number of failures. The B&W systems

included in this sort were the letdown purification

and makeup systems and high pressure injection

system. The Westinghouse systems included were

the High Pressure Injection System and Upper

Head Injection Subsystem. The results of this sort

are shown in Table 9. The overall aging fractions

for 1036 failures is 0.213. This means that 21.3%

of the failures were aging related. The HPIS components ordered by aging fractions for categories of

design, aging, testing, and human is shown in

Table 10. Valves caused the most failures, followed

by valve operators and instrumentation. These

19
Table 5. HPI pump failure causes (LER)

Number

of

Item Events

1

2

Cause

Electrical/mechanical 14

Control malfunction maintenance error 4

3

4

5

6

7

8

9

10

11

12

Design error

Operation personnel error

Failed internalsa

Unknown

Drive train failurea

Foreign material

Testing

Extreme environment

Loose fastenera

Loss of pressure boundarya

Improper clearance

Seal failurea

Bearing failurea

Total

4

3

3

5

2

2

1

1

I

1

13

14

15

1

1

I

44

a. Potentially aging related.

three categories accounted for approximately 77qo

of HPI component failures in HPIS.

The NPRDS system effect code identifies the

effect on the system caused by the component

failure. The codes were taken directly from the

NPRDS failure records. The NPRDS has five system effect categories and are defined as follows:

* Loss of System Function-A component

failure that, singularly results in the system

being unable to perform its intended

function (i.e., all trains, channels, etc.,

inoperable).

* Degraded System Operation-The system

is capable of fulfilling its intended function, but some feature of the system is

impaired.

* Loss of Redundancy-Loss of one system

functional path.

* Loss of Subsystem/Channel-A partial

loss of system functional path.

* System Function Unaffected-Failure did

not affect the operation of the system.

The fractions for each system effect are shown in

Table 9. Approximately 57% of the failures lead to

system degradation, but only 0.7% caused a loss of

system functions.

The motor-operated valve data from the nuclear

plant reliability data system for Westinghouse

plants had enough events (56) to show an aging

trend (shown in Figure 9) when the data is plotted

20
Table 6. Data on HPI and other chemical volume control pumps (CVCSIHPI) (LER Data)

(a) Mode (b) Type of Event (c) Activity (d) Class

Pump type U A B C D R C S T U V N M N R T U D T U

HPI 4 1 24 7 8 2 2 6 10 10 0 1 2 12 3 25 2 29 9 6

All other CVCS/HPI pumps 2 43 16 18 33 50 3 17 2 4 2 3 3 89 0 13 5 39 50 22

(a) Failure mode (c) Activity resulting in discovery

Code Description Code Description

A Leakage/rupture M During maintenance

B Does not start N During normal operations

C Loss of function R During records review

D Does not continue to run T During testing

U Unknown U Unknown

(b) Type of Event

Code . (d) Event Classification

Failure Command Description Code Description

S Nonrecurring, not common cause D Demand

R T Recurring, not common cause T Time (continuous operation)

C U Nonlethal common cause U Unknown

V Recurring nonlethal common cause

N Lethal common cause

W
Table 7. Cause for all system pump failures other than main HPI pumps (LERs)

Number

of

Cause Events

Seal failurea 29

Control malfunctiona 13

Loss of pressure boundarya 12

Drive train failurea 9

Maintenance personnel error 7

Failed internalsa 6

Shaft/coupling failurea 5

Bearing failurea 4

Operating personnel error 4

Extreme environment 3

Excessive weara I

Other 16

Total 109

a. Potentially aging related.

22
Table 8. Summary of HPIS valve failuresa

Failure Mechanisms Percent

Potentially Aging Related

Mechanical controls 8

(Parts failed or out of adjustment)

Seat or disk failure 6

Packing failure 5

Pilot valve failure 3

Torque valve failure 3

Motor operator failure 3

Leaking/ruptured diaphragm 3

Normal wear 3

Seal gasket failure 2

Limit switch failure I

Excessive wear 1

Solenoid failure I

Corrosion I

40

All Other

Command faults 32

Personnel operations 6

Personnel maintenance 4

Construction 3

Design 3

Other and unknown 12

60

a. Data source is LER.

23
Table 9. High pressure injection system totals and fractions

Failure Category Totals

Design failures 122

Aging failures 221

Test and maintenance failures 83

Human related failures 27

Other failures 583

Total 1036

Failure Category Fractions

Design fraction 0.118

Aging fraction 0.213

Test and maintenance fraction 0.080

Human related fraction 0.026

Other fraction 0.563

System Effect Totals

Loss of system function 7

Degraded system operation 197

Loss of redundancy 138

Loss of subsystem/channel 251

System function unaffected 443

Total 1036

System Effect Fractions

Loss of system function fraction 0.007

Degraded system operation fraction 0.190

Loss of redundancy fraction 0.133

Loss of subsystem/channel fraction 0.242

System function unaffected fraction 0.428

24
Table 10. High pressure injection system component failure category fractionsa

Component Total Design Aging Testing Human Other

Relay I - 1.000 - - -

Support 32 0.156 0.375 0.031 0.031 0.406

Filter 6 - 0.333 0.500 - 0.167

Heat exchanger 9 0.444 0.333 - - 0.222

Valve 307 0.127 0.326 0.085 0.020 0.443

Pump 86 0.105 0.314 0.116 0.047 0.419

Instrumentation 18 - 0.278 - - 0.722

recorder

Valve operator 161 0.081 0.211 0.143 0.031 0.534

Circuit breaker 43 0.163 0.209 0.070 0.047 0.512

Instrumentation: 141 0.106 0.113 0.071 0.007 0.702

transmitter

Heater 36 0.111 0.111 0.111 0.167 0.500

Instrumentation: 19 0.158 0.105 - - 0.737

controller

Instrumentation: 153 0.124 0.039 0.007 - 0.830

switch

Accumulator 4 - - - 0.250 0.750

Motor 9 0.111 - 0.111 0.111 0.667

Pipe 5 0.400 - 0.200 - 0.400

Instrumentation: 6 0.167 - - 0.833

computation module

a. Components ordered by aging fractions.

25
'u

161

co,

a)

E

z3

12 i

a Total number of failures-

. Aging caused failures

8

4

i I - - - - - - - - - - - - - - - - - - --

5 10 15 Years in service

9-6104

Figure 9. Failure data for HPIS motor-operated valves in 5-year increments (data from nuclear plant

reliability data system).

in five-year increments. This also appears in

Table 11 along with data for check valves, manual

valves, and HPIS pumps. The check valves and

manual valves showed no trend on aging failures.

The HPIS pumps had a significant increase in the

number of failures in the 10- to 14.9-year period

for both aging and other causes.

Plant Operating Experience from

Site Visit and Personnel

Interviews

An operating B&W plant was visited and site personnel interviewed. Included in the interviews were

the HPI maintenance supervisor, I&C supervisors,

and an electrical specialist. Detailed I&C drawings

on HPIS were reviewed with plant personnel, as

well as computer printouts of corrective maintenance (CM) requests, incident investigation reports

(IIR), and test procedures for HPIS. All maintenance work is initiated with a maintenance work

request and when the work is finished, a description of the problem and corrective action is written

on the request sheet. These CM requests are then

filed for record purposes.

Nuclear Maintenance Data Base. Corrective

Maintenance summaries were taken from the

Nuclear Maintenance Data Base. This plant data

system summarizes all CM reports as a one-line

summary, work required reference number, and

date. Any channel or component found deficient

during implementation of calibration and testing

procedures would have a CM request written to correct the problem.

The CM request records were computerized by

the utility starting October 23, 1980 and these were

the ones reviewed. Microfilm records for CM

before October 23, 1980, were not searched

because of manhour and cost limitations.

The computerized maintenance records were

reviewed and show that many potential problems

are fixed before major system or channel failures

occur. Thus, maintenance records reflect the incipient failures to some extent.

Data from the maintenance records are shown in

Table 12. These data cover a period from May 1980

26
Table 11. Component failures in 5-year incrementsa

Number of Failures

Component

Motor-operated

valve

Cv

Manual

valves

Pump

(centrifugal)

< 500 gpm

Failure

Classification

Other

Aging

Total

Other

Aging

Total

Other

Aging

Total

Other

Aging

Total

04.9

Years

16

3

19

7

2

9

3

1

4

4

4

8

5-9.9

Years

8

7

15

4

l

5

3

4

7

7

4

11

10-14.9

Years Total

5

12

17

29

22

51

9

2

11

20

5

25

3

2

5

9

7

16

14

15

29

25

23

48

a. Data from the nuclear plant reliability data system; data for older Westinghouse plants.

to December 1986 and contain 356 records. Of

these, 171 were routine maintenance items such as

recorder paper problems, cleaning filter, etc.

Instrumentation and Control were the largest category with 74 requests. Pipe-related maintenance

items were the next largest category with

55 requests. They include flange leaks, hangers,

pipe, penetration, and snubbers. These requests

were followed by valves with 25 requests, control

circuits with 21 requests, and pumps with

18 requests.

Incident Investigation Reports. Nonroutine

events in the plant (including those that occurred

during the precommercial operations phase) are

evaluated. This evaluation may result in an incident

investigation report (IIR) that captures the important details related to the event through interviews,

analysis of logs, recorder strip charts, computer

printouts, etc. IlRs are company proprietary and

cover reportable events such as techical specification violations. The event may or may not require

NRC notification. Hence, the LERs for the station

are a subset of the IlRs. The lIRs were reviewed

through April 1985.

Some observations based on sorts of the IIR data

base are:

1. Better resolution of information in the

coded fields than LERs and NPRDS failure reports

2. Precommercial operation events captured

3. Report event frequency relatively constant

up to 1984

4. Coded reporting for LER tracking

5. Infant mortality observable in component

failure searches.

A summary of IIR causes for HPIS failures are

presented in Table 13. Procedure or personnel error

and valve failure each accounted for 28% of the

incidents, followed by I&C (17%), pumps (11076),

nozzle (6%o), and welds (5%). However, bacause

there were only 17 events covering the HPIS over

an 11-year period, the average for reportable events

was relatively low at about 1.5 per year -for the

plant studied. Estimates indicate that about half of

these events are potentially age related.

Summary of Operating

Experience

The information from the data bases indicate

that approximately 57%o of the failures lead to

27
Table 12. HPI plant data

Number of Sub breakdown

Item CM Requests of CM Requests

Routine Maintenance, Misc. 171

Pipe Maintenance 55

Flange leaks (22) Hangers (17) Pipe, penetration, orifice (12)

Snubbers (4)

Instrumentation and Control 74

Sensors, monitors (38) Sensing line leaks (15)

Valves 25

Mechanical problems (12) Packing replaced (11) Valve operator (2)

Pumps 18

Vibration (7) Pump repair (4) Mech. seals (4)

HPI pump repair (1) Boron accumulation (I)

Pump motor (X)

Other 13

Pump coolers (7) Bolts (3) Gaskets (3)

Total 356

28
Table 13. 1IR summary of causes for HPIS

failures

LER have frequency of occurrence similar to the

NPE.

A summary of problems with the HPI system

components is given next for each type of

Type of Event Percent component.

Procedure/personnel error 28

Valve failure 28

I&C failure 17

Pumps failure

Nozzle/thermal sleeve failure

Weld failure

Other

Total

11

6

5

100

Piping. Prior to 1982, there were no significant

problems with piping cracks in HPI systems for

PWRs. However, in March of 1982, cracks were

identified in the thermal sleeve and safe end where a

makeup/high pressure injection line joins the reactor coolant system at one B&W unit. 1 0 Subsequent

investigations of these lines at four other B&W

units revealed similar cracks. The apparent cause of

the cracking was thermal fatigue. The injection

lines for normal reactor coolant makeup are also

part of the high pressure injection system at B&W

plants. Because these plants do not have regenerative heat exchangers in the coolant makeup circuit,

a potential existed for makeup temperatures to be

substantially lower than the reactor coolant temperature. Temperature variations due to mixing in

the high pressure system nozzle, coupled with

hydraulic effects, were suspected to be the principle

cause of failure.

Beginning in June, 1982, cracks were also identified in the thermal sleeve to nozzle connections of

reactor coolant system branch pipes at several

Westinghouse units. The cracking was discovered in

the thermal sleeve retainer welds that attach the

sleeve to the nozzle inlet. These cracking problems

included:

1. Accumulator lines, pressurizer surge line,

and charging line nozzles at one unit

2. Accumulator line nozzle at one unit

3. Safety injection and charging line nozzles

at one unit

system degradation, but only 0.7 !70 of these failures

caused a loss of system function. The data bases

also showed that 21.3 0o of these failures were aging

related.

The data bases are in agreement on the four most

troublesome components in the HPIS. However,

the CM data ranks them differently than the NPE,

NPRDS, and 1IRs. Valves have the most failure,

followed by I&C, pumps and pipe related events.

However, CM received the most requests for I&C

problems, then pipe related problems, and finally,

valves and pumps. These requests indicate more

minor problems with I&C and pipe hanger adjustments etc., but the major reportable problems rank

valves first in frequency of failure (see Thble 14).

Table 14. Data base ranking of most troublesome HPI components

Component NPE

Valves l

I&C 2

NPRDS

1

2

3

4

IIR

1

2

3

4

CMa

3

1

Pump 3 4

Pipe, supports, nozzles 4 2

a. The CM has many requests for I&C problems and minor pipe problems which are repaired and are not reportable events.

29
4. Two accumulator lines, a safety injection

cold leg injection line, and the charging

line nozzles at one unit.

The failure mechanism was suspected to be

fatigue induced by thermal cycling.

Most of the recent stainless steel pipe crack problems at PWRs have been fatigue related rather than

corrosion related. In 1979, stress corrosion cracking was discovered in some safety system pipes containing stagnant borated water. ll No losses due to

this problem were reported in 1980-1986.

Cracks have also occurred due to vibration and

dynamic loading (water hammer). Welds and

flanges are connection stress points. Flange loosening accounted for 9%o of all pipe problems in NPE

data.

Snubbers. Snubbers perform a safety function

by restraining the motion of attached systems or

components under rapidly applied load conditions

of earthquakes, pipe breaks, or severe hydraulic

transients. LERs relating to malfunction of snubbers indicated the most frequent problem was seal

leaks in hydraulic snubbers. Mechanical snubbers

were subject to damage due to vibration. A phase I

NPAR study has been conducted on snubbers. 12

Penetrations. At least five piping penetrations

are associated with the HPIS for the plant studied.

Few problems have been found with penetrations.

One estimate for penetration problems (Issue B-26,

Reference 13) was one failure per year for all operating plants (71 at time of estimate). In that estimate, it was assumed that each plant had

40 penetrations. If five penetrations are associated

with HPIS, then about every eight years a problem

could be expected with an HPIS penetration in one

of the 71 operating plants.

Tanks. Few tank problems have occurred. Most

of the problems attributed to tanks in data bases

involved boron concentrations or fluid levels that

were out of specification. The only significant

event was to replace a boric acid injection tank at a

PWR.

Pumps. Most high pressure injection pump failures that are aging related involved seal, bearing,

and shaft problems.

Valves. The type of component most often

responsible for HPI failure was valves. This

includes all types of valves and valve operators.

Personnel error, operation, and maintenance were

the primary causes overall. When these causes were

removed, mechanical parts failure, packing leaks,

and seat or disk failure were the most often mentioned causes related to aging and service wear.

Chemistry Problems. Corrosion in pipe weld

heat zones due to contaminants and boric acid corrosion of ferritic metals are two failure causes

related to chemistry and aging. Controlling boric

acid concentrations is a safety related operational

problem. See Appendix H for problems with

borated water systems.

Microbial Influenced Corrosion. Many systems

in the majority of nuclear plants appear to be susceptible to some form of microbial influenced corrosion (MIC). This is particularly true of standby

systems where conditions exist for microbial

growths. Stainless steel 304, 316, etc., have been

affected by MIC as well as other metals. One utility

experienced MIC in the HPI and LPI pump impeller blades after initial tests followed by a period of

standby. Microorganisms have also been found in a

borated water storage tank, but no degradation was

reported.

30
HPI SAFETY ISSUES AND POTENTIAL AGING PROBLEMS

The following is a review of system and personnel

interaction safety issues13 related to HPI and aging.

Locking Out of HPIS

Power-Operated Valves

The NRC staff positions BTP EISCB 18 and BTP

RSI3 6-11 require the physical locking out of electrical

sources to specific motor-operated valves in the ECCS

including HPIS and LPIS. This method protects

against a single failure causing an undesirable component action. This assumption in the safety evaluation is

that the component is then equivalent to a similar component that is not designed for electrical operation and

can only be opened or closed by direct manual operation of the valve. Thus, no single failure (due to any

cause including aging) can both restore power to the

component and cause mechanical motion of the component. The probability of failures due to maintenance

errors, electrical faults, and mechanical failures

(7 x 1f07) was greater than the probability of the valve

mispositioning coincident with a LOCA, and as a

result of operator error (4 x 10-7). This was issue number B-8 in Reference 13, but was dropped as a safety

issue because it is an acceptable approach to meet the

single failure design criteria. However, this points out a

design shortcoming when human interaction is

required for initiating HPIS valve action.

Inadvertent Actuation of Safety

Injection in PWRs

Westinghouse and B&W plants had a high rate of

spurious or inadvertent safety injections occurring. In

the case of B&W reactors, the practice was to manually

turn on one or more HPI pumps after a reactor scram

to recover the pressurizer level. This practice contributed to the high rate of safety injections for these

plants. This practice was stopped after the accident at

the Three Mile Island plant because it was determined

the HPI pumps were not needed to maintain pressurizer level. As a result, unneeded SI in B&W plants is

now significantly lower. A possible reason for

unneeded SIs in Westinghouse plants is their design

requires more signals for initiating SI and thus more

chances for spurious signals. Actuation of the SI is

undesirable because it injects cold borated water into

the reactor, thereby subjecting injection nozzles to thermal stresses and requiring removal of boron from the

primary system before startup. Actuation was determined as not only an economic issue, but could possibly lead to a wrong response by an operator when SI is

really needed. Operator response is carried as another

issue and would include this one. This was issue 8, Reference 13. Inadvertent SI actuation due to this operating procedure was corrected by eliminating' the

procedure. However, maintenance personnel error and

other operator procedures that could inadvertently initiate SI should continue to be reduced through design,

training, and applicable updating procedures.

Switch from HPI Mode to

Recirculation Mode

The switchover from the HPI mode of operation to

the recirculation mode for accident recovery requires

realignment of a number of valves. The switchover can

be achieved by a number of manual actions, by automatic actions, or a combination of both. The three

switchover options (manual, automatic, and semiautomatic) are vulnerable with varying degrees to human

errors, hardware failure, and common cause failures.

Automatic system actuations reduce the impact of operator error in completing the switchover, but are subject

to spurious actuations. Spurious switchover of HPI to

HPR has the potential for pump damage as well as

unacceptable safety consequences. This safety issue

(No. 24) was scheduled for prioritization

(Reference 13).

High Pressure Recirculation

System Failure Due To

Containment Debris

In the recirculation mode the HPIS pumps take suction from the LPIS, which is taking suction from the

containment sump. Any debris, paint flakes or loose

material due to aging could potentially damage system

components during the HPRS mode of operation.

This is safety issue 28 (Reference 13), which has now

been scheduled for prioritization.

Failure of Demineralizer System

and the Effect on HPIS

While the demineralizer system does not directly

perform a safety function, a failure of the demineralizer system could impair the operation of the HPIS. In

the plant studied, the demineralizer system has key

31
components labeled as part of the HPIS to ensure

prompt attention if failure occurs. In Reference 13,

this problem is listed as issue 71 and is scheduled for

prioritization. This is also true of other systems and is

addressed in the next subsection on Systems

Interaction.

Systems Interaction

The HPIS interfaces with many other plant

systems. They include IE power, LPIS, RCS,

service water systems, containment spray system, monitoring, and ESFAS. In addition, part

of the HPIS is used for both normal operation

and emergency operation. There are also HPI

subsystems such as demineralizer, makeup/

letdown system, chemical control, boron injection, lubrication of HPI pumps, pump seal

cooling, and instrument air system. System

interactions for HPI may be one study recommended for the Phase 2 aging study. Overall

plant system interactions is also a Safety Issue

(A-17, Reference 13). In addition, the performance of the HPIS equipment can be affected by

the heating, ventilation, and air conditioning

system.

32
AGING ASSESSMENTS FOR THE HPIS

The third objective of the HPIS aging study was to

perform screening type assessments of the impact of

aging on operability of the HPIS. Subobjective (a),

failure modes and causes, has already been covered in

the discussion of operating experience. Based on the

review of operating experience from generic data bases,

safety issues, and plant data the remaining four subobjectives of objective 3 can be discussed. They

include: (b) identification of materials susceptible to

aging degradation, (c) stressors during operation,

(d) functional indicators that would aid in failure prediction, and (e) methods for detection and control of

aging degradation.

Preliminary Identification of

Susceptibility of Materials to

Aging

The HPIS piping, pumps, and valves are all

stainless steel which is compatible with boric acid.

The BWST is carbon steel with a liner to protect

against corrosion. Problems have occurred when

leaks in connections or valves allowed borated

water to come in contact with carbon steel components. Water evaporation results in a concentration

of boric acid causing corrosion on carbon steel

components.

The narrative descriptions in the NPE data base

were reviewed to determine if failure occurred as a

result of conditions specific to the HPIS. Several

failures were reported that resulted from the charging of cold water into a hot system and from the

handling of the water with a high boron concentration. The four most significant failures are discussed below.

High-pressure injection/makeup nozzles have

developed through wall cracks. The cracks resulted

from thermal fatigue. The thermal fatigue was

caused by turbulent mixing of hot and cold coolant

and thermal shock of the hot safe-end wall during

normal makeup. All cracks were associated with

loose thermal sleeves. Improved thermal sleeve

design and increase in minimum continuous

makeup flow to prevent thermal stratification have

been employed for failure mitigation.

Elbows in the safety injection piping between the

cold leg and the first check valve have developed

through wall cracks. The cracks occurred in the

heat-effected zone of the elbow weld, to the safety

injection piping, and in the base metal of the

elbow. The cracks resulted from high-cycle thermal

fatigue caused by cold makeup water leaking

through a closed globe valve at a pressure sufficient

to open the check valve. Mitigation methods are

being evaluated. Installation of a globe valve downstream of the check valve, rather than the existing

valve, is being considered to isolate the injection

line during normal makeup.

Motor-operated valves and check valves have

failed to operate due to boron crystallization on the

valve stems, in the valve packing, and in the valve

body. The reason for crystallization was not always

reported in the NPE, and investigations may not

have identified the cause. However, most causes

were reported as packing leaks. The valves were

usually cleaned and placed back in service. One

incident reported additional heat trace was added

to prevent future failures.

Injection boron concentration has been diluted

from leaking valves. Leaks have been reported for

both check and globe valves. Dilution of the boron

injection tanks for Westinghouse plants and safety

injection tanks for combustion engineering plants

have been reported. A few dilutions have been

reported for the borated water storage tanks of the

Babcock and Wilcox plants. Improved monitoring

of the tanks and repair of the valve seats have been

implemented as mitigation measures.

A microbe-caused-problem with stainless steel

occurred in a plant when preliminary tests were

conducted with water in the HPIS and allowed to

stand before startup. During this standby period

microorganisms caused corrosion on the pump

impeller blades.

I&C and electronic components have been susceptible to catastrophic failure. Contact wear in

switches and relays as well as corrosion on contacts

are common aging effects on electrical components. Electrical insulation ages with thermal

cycling.

Stressors for HPIS

The HPIS is subject to many stressors during

operation. They include stressors due to maintenance, operation, and testing; environmental stressors; electrical stressors on I&C; and mechanical

stressors. Various types of stressors are identified in

the following sections.

33
Maintenance, Operations, and Testing Stressors. Included in the maintenance, operations,

and testing stressors are:

1. Personnel error-resulting in inadvertent

SI actuation. If valves are not oriented

properly, or pump suction unavailable,

damage to components could result.

2. Water hammer (dynamic loading)-

incidents have been attributed to pump

startup with partially empty lines and

rapid valve motion. Most damage has been

relatively minor and involves pipe hangers

and restraints.

3. Thermal cycling-due to cold water from

the HPI system into the hot RC system.

Also reactor heatup and cooldown causes

thermal cycling.

Environmental Stressors. Included in the environmental stressors are:

1. Temperature or pressures-are environmental stressors

2. Water chemistry incorrect-impurities left

from welding and boric acid crystals are

examples of environmental stressors

3. Vibration-flow-induced vibration may

have been a contributing factor in the thermal sleeve cracking in PWRs. Pumps can

be a source of vibration if unbalanced.

Many other potential sources of vibration

exist in a nuclear plant. If the natural frequency of vibration of the connected piping is very nearly the same as the driving

frequency of the pump, then there is the

possibility of fatigue failures in the system,

particularly at the nozzles where the stress

will be highest. Vibration is detected on

major rotating equipment by instrumentation, inservice inspection, or other visual

means. Vibration problems, however, have

to be resolved on a case by case basis

4. Seismic stresses-may cause relay chatter

or fastener damage. Seismic stresses, however, are not due to recurrent conditions

due to operations.

Electrical Stressors. The electrical stressors

from switching transients and loading include:

1. Transients affecting I&C-transients can

occur from HPIS operation, external electrical faults, and lightning.

2. Low voltage affecting I&C-abnormal

voltage can occur from excessive loading,

power supply drift, and set point

adjustments.

Mechanical Stressors. Some mechanical stressors are as follows:

1. Valve misadjustment-limit switches, and

torque switches out of adjustment, and

mechanical adjustments can cause

mechanical stress

2. Pipe alignment-any misalignment of piping can stress welds, connection, and

flanges

3. Vibration-caused by motors and pumps.

Functional Indicators that would

Aid in Failure Prediction

Functional indicators include abnormal currents

and voltages for I&C equipment and indicate a

change from the normal expected values. Leaks in

pipes, flanges, and nozzles indicate problems such

as loose fasteners, cracked pipe, or corrosion. Visual indicators could include limit switch setting,

boric acid crystals, or shaft wear. Unusual vibration or noise associated with motors, valves, and

pumps could also be an indication of impending

failure.

Methods of Detection and

Control of Aging Degradation

Any of the failure detection methods also apply

to aging detection because aging is one of the mechanisms causing failure. Functional indicators

observed during normal operations or testing is one

means of detecting degradation. This detection

includes the normal control room monitoring of

gauges, charts, and computer printouts. For special problems or studies, additional condition monitoring such as vibration sensors, noise monitoring,

or current and voltage signatures may be applied.

During refueling outages, end-to-end operational

checks and functional testing results are compared

to previous baseline measurements. Any change

from the baseline should be checked out for cause

to determine if it is degradation due to aging.

Because equipment starts aging from the day it is

manufactured, control of aging begins at that time

through the management process. The management of the equipment in all phases of its life cycle

34
includes attention to shipping and conditions of

storage prior to installation. After installation, the

management of the aging process is through

inspection, surveillance, and monitoring with both

preventive and corrective maintenance.

Aging Assessment Summary

The aging assessment of the HPIS involves a

number of factors including stressors, degradation

mechanisms, and failure modes. A summary of

these various factors in the aging process along

with the inservice inspection methods is given in

Table 15.

Stressors acting on the various components contribute to the aging process. Stressors associated

with maintenance, operation, and testing include

inadvertent HPIS actuation, water hammer, and

thermal cycling. Environmental stresses include

abnormal temperatures or pressures, incorrect

water chemistry, boric acid crystals, and vibration.

Electrical stresses include external environment of

temperature, humidity and limited radiation,

abnormal voltages, and electrical transients affecting I&C. Mechanical stresses include pipe misalignment, vibration, and dynamic loading from valves

closing.

Degradation mechanisms for the HPIS passive

components (piping, thermal sleeves, and nozzles)

include fatigue, crack initiation and propagation,

and thermal embrittlement. Valves are subjected to

wear, foreign material, mechanical linkage problems, and seat or disk degradation. Motor operators have wear and loose connections as they age.

Air-operated valves main degradation mechanism

is due to contaminated air supply this contamination is moisture or oil in the air. For I&C, the degradation mechanisms are loose connections,

corrosion of terminals, and catastrophic component failures. Pumps degrade through wear, vibration, and fatigue.

The potential failure modes for the HPIS valves

and pumps are failure to operate when needed;

inadvertent operation when not called for; and during operation, a failure to operate as required. Secondary modes would include leaks, blockage, or

command faults. Piping and other passive components have failure modes of leaks, cracks, or loose

parts. For I&C, the failure modes are opens, shorts,

or failure to operate.

The inservice inspection methods for the HPIS

components include visual inspection for leaks,

volumetric inspection, and operational tests.

Materials in the HPIS susceptible to aging

include seals, and packing material in pumps and

valves. Carbon steel materials in other systems

exposed to boric acid for some period of time will

corrode. Electrical components in the I&C subsystem are subject to degradation of insulation, corrosion, and wear failures. The stainless steel piping

ages from thermal fatigue and wear. Material wear

in pumps, valves, and relay contacts is a normal

aging process.

35
Table 15. Summary of aging processes for HPIS

Major Component

Nozzles and thermal

sleeves

Stressors

System operating

transients,

thermal cycling,

vibration, water

(hammer)

Degradation

Mechanisms

Fatigue crack

initiation and

propagation

Potential

Failure Modes

Leaks through wall,

loose parts

ISI Methods

Visual inspection

volumetric inspection

Valves System operation

transients,

maintenance, and

testing

Wear, foreign

material,

mechanical linkage

faults

Leakage, fail to

operate, blockage,

command faults

Visual and operation

tests

Air-operated valves Contaminate air

supply

Parts degradation

by oil in air supply

Fail to operate Visual and

operational tests

ON I&C Electrical

transients,

temperature

maintenance,

vibration

Corrosion, loose

connections, failure

(catastrophic)

Open, shorts,

fail to operate

Testing

Pumps Systems operating

transients,

thermal cycles

Wear, vibration,

fatigue

Seal leaks, fail

to start, fail

to run

Testing, visual

inspection

Pipe supports Vibration, water

hammer

Fatigue, loosening

of connections

Breaking loose Visual inspection

Piping Vibration, water

hammer, thermal

cycles

Thermal fatigue

abrasive wear

Through the wall

leakage, or cracks

Visual inspection

volumetric inspections

Motor operators

for valves

Electrical

transients,

maintenance

Loose connections,

wear

Fail to operate Testing
REVIEW OF INSPECTION, SURVEILLANCE, AND TESTING

For the representative plant studied, the operability requirements of the HPIS are governed by the

Standard Technical Specification. These specifications require that at least two HPI pumps be operable when the reactor is critical, and that two trains

of the HPIS must be able to draw suction from the

BWST and discharge to the RCS automatically

upon ES actuation at power levels up to 60% full

power. In addition, the remaining HPI pump and

valves HP-409 and HP-410 must be operable and

valves HP-99 and HP-100 must be open when the

reactor is above 60% full power. Test or maintenance on any component is permitted provided that

operability of the redundant component in the

other train is demonstrated first, and subject to the

following conditions:

I. If reactor power is less than 6007o, the

HPIS must be restored to the appropriate

status identified above within 24-hours, or

placed in hot shutdown within an additional 12 hours. If the HPIS cannot be

restored within the following 24 hours, the

reactor must be taken to cold shutdown

within an additional 24-hour period.

2. If the power level is greater than 60%, the

inoperable component must be restored

within 72 hours, or power must be reduced

to below 60%o in another 12 hours.

The surveillance requirements in the Technical

SpecificationsI 4 and Section XI of the ASME

Boiler and Pressure Vessel Code' 5 comprise the

testing requirements for the HPIS. Technical Specifications 4.5.1.1.1 requires that during each refueling outage the HPIS be tested to demonstrate that it

responds correctly to an actuation signal. Individual components are required to be tested more frequently, as defined in Table 16.

Section XI of the ASME defines the inservice testing used by the plants. Pumps are tested quarterly

unless a relief request is granted. For these tests, vibration, differential pressure, and flow are measured.

Bearing temperatures are also measured but on a less

frequent schedule. Vibration is an excellent indicator

of pump degradation and is a good monitor for pump

aging. Periodic measurements of the electrical characteristics of the motor are not required by Section XI. A

check at one plant indicated that monitoring is not

done. Pump vibration and performance are not sensitive measurements for electrical insulation and other

motor electrical degradation. Electrical characteristic

measurements would be required to detect such aging.

Valves are tested quarterly unless a relief request

has been granted. For these tests, stroke time is

measured, usually without differential pressure.

The measurements are often made crudely using a

stop watch. Such tests would not be effective as a

monitor for aging. Periodic measurement of the

electrical characteristic for motor operators is not

required by Section XI. For resolution of IE Bulletin 85-03, most plants are using diagnostic equipment to verify torque switch setting. Although the

use of this equipment often includes electrical measurements of the operator, the tests are usually only

done once for verification and rarely repeated. For

valve testing to be a useful monitor for aging, more

accurate measurements of stroke time and periodic

measurement of electrical characteristics of the

motor operator will be needed.

Section XI also defines the inservice inspection for

welds. Welds are to be inspected volumetrically each

10 years. The cracks in the HPI nozzles and elbows

were detected by leaks, not by the ultrasonic inspection

of the inservice inspection program. The ultrasonic

techniques specified by Section XI were found to be

inadequate to detect cracks resulting from thermal

fatigue. The instrument gain had to be increased significantly and the 45-degree transducer had to be supplemented by a 60-degree shear wave transducer in order

to detect the cracks. Also, one crack developed in the

base metal of the elbow. Section XI only requires

inspection of welds. Inspection of high-stress areas of

base metal may be needed. See IE notices No. 88-01

and 88-02.l6,l7

Periodic testing standards are given in Part 3 of

Table E-5 in Appendix E. The high pressure injection system will also be inspected periodically during normal operation for leaks from pump seals,

valve packing, and flanged joints. Additional items

inspected include heat exchangers and safety valves

for leaks to atmosphere. Typical performance tests

are given in Table 17.

The following specific performance tests are performed at the representative plant studied:

1. High pressure injection valve verificationProvides verification that each valve is in its

correct position

2. High pressure injection system leakagePeriodically tests the High Pressure Injection

System outside containment for leakage

37
Table 16. Test frequency for HPIS componentsa

Component Type of Test Frequency

HPI pump 3A

HPI pump 3B

HPI pump 3C

MOV 3HP-3

MOV 3HP-4

MOV 3HP-5

AOV 3HP-16

MOV 3HP-20

AOV 3HP-21

MOV 3HP-24

MOV 3HP-25

MOV 3HP-26

Start, stop, and operating parameters

Start, stop, and operating parameters

Start, stop, and operating parameters

Stroke and leak

Stroke and leak

Stroke and leak

Stroke

Stroke and leak

Stroke and leak

Stroke

Stroke

Stroke and leak

Monthly

Monthly

Monthly

Quarterly

Quarterly

Cold SD

Quarterly

Cold SD

Cold SD

Quarterly

Quarterly

Quarterly

MOV 3HP-27

CV 3HP-101

CV 3HP-102

CV 3HP-105

CV 3HP-109

CV 3HP-113

CV 3HP-126

CV 3HP-127

CV 3HP-152

CV 3HP-153

CV 3HP-188

CV-3HP-194

MOV 3CC-7

AOV 3CC-8

CV 3CC-20

CV 3CC-24

Stroke and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Check valve function and leak

Stroke and leak

Stroke and leak

Check valve function and leak

Check valve function and leak

Quarterly

Refueling SD

Refueling SD

Refueling SD

Refueling SD

Refueling SD

Refueling SD

Refueling SD

Refueling SD

Cold SD

Refueling SD

Cold SD

Cold SD

Cold SD

Refueling SD.

Refueling SD

a. Abbreviations: AOV, air-operated valve, SD, shutdown, CV, check valve.

38
Table 17. HPIS performance testing

High pressure injection pumps

High pressure injection line

valves

High pressure injection pump

suction valves

Borated water storage tank

outlet valves

One of two pumps operates

continuously. The other pump will

be operated periodically

The remotely operated stop valves

in each line are opened partially

one at a time. The flow monitors

will indicate flow through the lines

The valves are opened and closed

individually and console lights

monitored to indicate valve position

The operational readiness of these

valves is established in completing

the pump operational test discussed

above. During this test, each valve

is tested separately

3. High pressuire injection system performance

test-Demonstrates the operability of the

HPI pumps in accordance with applicable

ASME code and identifies potential problem

areas as early as possible.

4. High pressure injection system ES TestDemonstrates the HPI System is operable

from an ES signal.

5. High pressure injection pump ventingPeriodically vents the casings of nonoperating HP pumps to prevent gas buildups

6. High pressure injection motor coolant flow

test-Periodically tests the cooling water flow

through the HP pump motors to ensure adequate upper motor bearing cooling

7. High pressure injection check valve functional test-Demonstrates the operability

of the HP System check valves.

Additional inspections for leaks, as well as functional tests may be performed periodically on

major components if required by the utility. For

example, after maintenance, functional tests may

be necessary to verify operation. Monitoring consists of comparing the performance of similar

channels and visual inspections for leaks in piping

valves and pumps. Current maintenance practices

by utilities follow recommendations by vendors for

major components, such as valves and pumps. The

Babcock & Wilcox plants that have experienced

nozzle cracking have also indicated enhanced

inspection and surveillance of the nozzle and associated piping welds.

Some utilities have used advanced I&C cable

testing during refueling outages to establish baselines and obtain trending data on electrical equipment. One such system used primarily for cable

and connection testing is the ECCAD system. 18

Another advanced surveillance system is the motor

operated valve analysis and test system

(MOVATS). 3 Since 1984, MOVATS has been used

as a diagnostic tool for motor operated valves.

39
ROLE OF MAINTENANCE IN COUNTERACTING AGING EFFECTS

Present Regulations and

Guidance

The current USNRC regulation approach to

nuclear plant maintenance is embodied in requirements for quality assurance during design, construction, and operations consistent with the safety

(10 CFR 50, Appendix B) and surveillance

requirements that ensure necessary availability, and

quality of systems and components is maintained

(10 CFR 50.36). These rules and regulations provide no clear programmatic treatment of nuclear

plant maintenance. Regulatory Guide 1.33, Revision 219 endorses ANSI N18.7-1976/ANS 3.220,

which provides no specific guidance regarding

maintenance, but covers administrative controls

and quality assurance for the operational phase of

nuclear power plants.

Current Maintenance Practices

The representative plant studied follows manufacturer's recommendations for preventive maintenance on major components and plant specific

procedures. For example, preventive maintenance

on HPI pumps and system valves includes the

following:

1. High pressure injection pump-In order to

determine internal wear of the pump, periodic efficiency tests shall be performed.

This efficiency shall be used as a comparison to initial efficiency.

2. System valves-Maintenance on remotely

operated valves shall be in accordance with

the vendor's recommendations. Manual

valve maintenance shall include checking

for packing leakage when the system is

under pressure. Safety relief valve tests

shall consist of in-place or bench testing of

setpoints as appropriate.

Corrective maintenance is also minimized by

observing good operating practices and precautions. For example, the following limits and precautions shall be followed to prevent component

damage and abnormal aging.

Normal Operation Precautions. Prior to starting the high pressure injection pumps, particular

attention must be given to the opening of all valves

in the suction line and to proper venting of the

pumps. Failure of suction could result in instant

loss of the started pump. The minimum allowable

flow of a pump is 30 gpm.

The HP pump can be started against shutoff

head, but operation of the pump in this condition

for over 30 seconds could cause the pump to overheat. The HP pump must be tripped if the motor

bearing temperature exceeds 215'F.

The HP pumps must not be started with an open

flow path to the RC pump seals. Injection seal flow

is required to all RC pumps when the RC pressure

and temperature are above 100 psig and 1900F.

The maximum flow through one letdown cooler

shall not exceed 80 gpm. The maximum seal return

temperature should not exceed 130'F to avoid damaging the demineralizer resins.

The maximum flow through one makeup filter

shall not exceed 150 gpm. A maximum pressure

drop of 30 psi should not be exceeded at any flow

rate.

High Pressure Injection Mode Precautions. The

same precautions apply during high pressure injection

in regard to suction and discharge flow of the HP

pumps as under normal operating conditions. The

possibility of pump runout due to a line break on the

discharge side must be considered. The maximum flow

of 525 gpm must not be exceeded for any length of

time because overheating of the motor may occur.

Benefit of Preventive and

Corrective Maintenance

Preventive maintenance should be performed on

the basis of need because the adoption of arbitrary

and frequent maintenance can be counter to safety.

Preventive maintenance should be supported by technical evidence and reviewed periodically. The benefits

from preventive maintenance include higher system

availability, increased life, and higher reliability.

Corrective maintenance is usually performed on

a priority basis and closely coordinated with testing. After corrective maintenance, the channel or

system is usually tested to verify that it is functionally correct. Likewise, a component or system that

fails a test will probably require adjustment or corrective maintenance. Thus, corrective maintenance

40
is a necessary part of plant operations and keeps

the HPIS functioning properly.

Improper Maintenance

Excessive preventive maintenance can have a

negative impact on safety and aging. Thirty-five

percent of nuclear plant abnormal occurrences may

be due to faulty maintenance and surveillance testing.2 1 Human error during maintenance has

involved the wrong unit or train and may increase

the potential for equipment damage. In order for a

preventive maintenance program to be effective it

must apply to both equipment with detectable degradation effects, and methods of detecting degradation before failure. Only about 25% of

equipment failures are preventable. 22 Examples of

faulty maintenance include sticking control breakers because of lack of lubrication, maintenance on

wrong train or component, and personnel errors.

Recommendations for Preferred

Maintenance Practices

For the HPIS, the preferred maintenance recommendation would be the maintenance practice recommended by the vendors for major components

such as pumps and valves. In addition, an

enhanced inspection program for leaks and cracks

in piping and nozzles would be coordinated with

corrective maintenance. A maintenance program

takes into account many factors including safety,

operations, economics, and availability. The interval for equipment maintenance frequency should

take into account potential impending failures,

their detection, and known equipment wearout

regions. Reliability centered maintenance has been

used in the aircraft industry and is being considered

in some nuclear plants. When properly applied,

reliability centered maintenance should enhance

plant safety and reduce life cycle costs.

41
CODES AND STANDARDS

Determining aging effects on equipment and life

extension is a key part of the NPAR program. One

of the outputs of the NPAR program is to recommend upgrades for old standards or recommend

development of new ones. Electrical equipment

issues relating to HPI equipment qualification and

aging are addressed by the following documents.

NUREG 058823 requires that qualification programs for electrical equipment should identify materials susceptible to aging effects and establish a schedule

for periodically replacing the equipment and material.

IEEE 323-1983,24 the industry standard upon

which the above requirements are based, includes a

number of paragraphs addressing this issue. For

example, Paragraph 6.4.2 on Operating History

states that, in order to use operating history information for establishing qualification, a record or

auditable data showing that equipment similar to

that being qualified has been exposed to levels of

environment at least as severe as those expected

from all service conditions for which the equipment

being qualified is required to function and that the

equipment satisfactorily performed the functions

required for the equipment being qualified. Those

elements of required exposure not covered by operating history may be accounted for by testing.

Regulatory Guide 1.8925 provides the guidelines for

meeting the Equipment Qualification Rule and

endorses IEEE 323-1974.

Data from naturally aged components or in-situ

measurements should be used to verify, where possible, data from artificial aging of components

used during equipment qualification.

The Standard Review Plan,2 6 Section 3.11,

"Environmental Qualification of Mechanical and

Electrical Equipment," which provides guidance

for USNRC staff in reviewing FSARs, includes

requirements for maintenance/surveillance programs for equipment located in mild environments.

Specifically, it is required that "the maintenance/

surveillance program data shall be reviewed periodically (not more than every 18 months) to ensure

that the design qualified life has not suffered thermal or cyclic degradation resulting from the accumulated stresses triggered by the abnormal

environmental conditions and the normal wear due

to its service condition. Engineering judgment shall

be used to modify the replacement program and/or

replace the equipment as deemed necessary." This

SRP guideline should be considered for possible

application in any new maintenance standard or

guide for maintenance.

The HPIS regulatory requirements and guidelines are given in Appendix E.

The Board of Nuclear Codes and Standards has

overall responsibility for codes and standards

development covering nuclear plant life extension.

The IEEE working group 3.4 is presently reviewing selected IEEE standards related to plant life

extension and plans to develop a guideline document. The mechanical components are covered by

ASME code including Sections III-C, VIII,

and XI. The process of developing recommendations for requirements similar to the electrical

equipment is an ongoing process as part of the

NPAR program.

42
HPIS AGING SYSTEM UNAVAILABILITY ASSESSMENT

The main objective of this unavailability assessment is to identify the components that contribute

to HPIS unavailability, and determine if they

change with time because of aging. Another objective is to determine the aging contribution to the

HPIS unavailability for the three emergency modes

of operation. These three operating modes are: (a)

high-pressure injection with one of three HPI

pumps [HPI(l) and HPI(2)] required, (b) highpressure injection with two of three HPI pumps

required and (c) the recirculation mode. A third

objective is to identify the type of events that contribute to HPIS unavailability.

This aging assessment is based on the linear

aging model27 and uses data from the Probabilistic

Risk Assessment (PRA)28 for the representative

plant studied and generic failure cause data on

HPIS components from a composite of nine PWR

plants that were at least 10 years old. This

approach is an approximate method that uses PRA

results (steady state models) to evaluate aging risk.

The PRA results provided the system fault trees

and baseline data for this study. The failure cause

data is used to estimate the time dependent failure

rates and the PRA is rerun at discrete times to provide the aging assessment. The software tool used

for this work was the Integrated Reliability and

Risk Analysis System (IRRAS).2 9

PRA and Basic Event Data

The PRA for the representative plant used in this

study, had been developed as part of a program to

improve the PRA capability of the electric utility

industry. This PRA was used for training personnel

from seven utilities during the course of its development. The PRA results were also used as input to

Living Schedules for plant modifications and to the

Integrated Safety Assessment Program.

The results of the PRA showed that the system

risk is dependant on support systems and events

internal to the HPIS. Only the internal events were

considered in this study.

The internal events included HPIS component

failures, human errors in leaving components

unavailable, and having a system's segment out for

maintenance. The basic event data from the PRA

for the hardware, human errors and maintenance

are given in Appendix I. The basic event data were

used to run the baseline calculations assumed to be

for a plant with random nonaging failures.

Failure Cause Data

The failure cause work30 was used to evaluate

the aging fraction (f) and the mean time to

failure (T) for each of the major HPIS components. Although, Reference 30 does not report the

specific data used in this evaluation, the files developed from the NPRDS data base as part of that

work were accessed to obtain the information.

The upper and lower bounds were developed in Reference 30 to account for the uncertainty encountered in

accurate identification of aging-related causes on the

basis of the component failure descriptions. The categorization scheme defined when a failure should be

classified as related to aging or nonaging. When insufficient information was contained in the failure

description, the aging classification was unknown.

These failures were then used to establish the upper and

lower bounds for the aging-related failure-cause fractions. The upper bounds (UP) were calculated using

the failures classified as unknown as aging-related failures, while the lower bounds (LW) are calculated using

them as nonaging-related failures. Upper and lower

bounds were developed for the data used in this report

by the same method. The upper and lower bounds for f

and T are included in the data shown in Thble 18.

In Table 18, the first three to five letters and numbers identify the system and component numbers.

The last three letters in the event name describe the

following event codes:

CVO

MVO

VxT

PPS

PPR

AVO

RVF

FIT

check valve fails to open

motor-operated valve fails to open

manual valve fails to open

pump fails to start

pump fails to run

air-operated valve fails to open

pump fails to start on low seal flow

flow transmitter fails high.

Methodology

The random nonaging failure rates from the PRA

were used for the first five-year period. Aging data

taken from the failure cause study were used to calculate a new failure rate for subsequent five-year periods.

When the constant rate contributions are incorporated, the time dependent failure rate using the linear

model is given by the equation:

43
Table 18. HPIS data for the risk assessment, aging acceleration factors, and component unavailabilities in 5-year increments for

upper and lower bounds

Aging Failure

FaIlure Fraction

Mean Time iSnion Failure

To Failure Tim Rate

(Years) (Yearn) Per Year

Aging Acceleration Factor LCwvofeft onana, an, 'Ky

(cent ane ODat Used [V £ .I TUP TLV 11T. P(51 P7I0107UP fl]LV P(151UP P(IILj Pf201UP PQ7oItV P(25hUP P(251LV pf33010U P 301lW P(351UP P(35LW P(40iUP 401 LI

HPIOICVO

HP 9102CV0

HF1 I09CVO

H1713CV0

1IS2CVO

HP153CV0

HF I188CV0

LPSSCVO

LPS7CVO

HP1448WVT

LPS4VVT

*N LPS6VVT

HP24MV0

HP25KVO

HP291V0

HP4090VO

hP410hVO

HFBPPS

HPCPPS

HPAPPR

HPRAPPR

HP8PPR

HPR6FPR

HPCPPR

HPRCPPR

HPI ISAVO

HP16A8O

CSSXVO

CISGCVO

HPRC8PPS

11153RVF

W93lFlT

HpI CV

HP1I CV

1W9 CR

HPI CV

HPI CR

WlI CV

HFI CR

H11I CV

HPI CV

HP1I KV

HPI ICV

W7 IHCV

HPI HOV

W3 MCU

HPI KCV

H1I HCV

WlI KV

HP1I ItOP

HFPI MoP

HPI MOP

HPI nOP

HPI MOP

hPI MOP

WlI POV

HPI POW

HF ICy

HFPI HOP

HFI KFP

HPI FT

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.83

0.83

0.83

0.84

0.64

0.84

0.84

0.84

0.85

0.86

0.86

0.86

0.86

0.86

0.86

0.86

0.88

0.88

0.90

0.90

0.85

0.8S

0,9Z

0.63

0.63

0.63

0.63

0.63

0.63

0.U3

0.63

0.63

0.61

0.61

0.61

0.49

0.49

0.49

0.49

0.49

0.U4

0.54

0.54

0.54

0.54

0.54

0.54

0.54

0.56

0.56

0.63

0.63

0.64

0.64

0.41

6.88

6.88

6.88

6.88

6.88

6.88

6.UB

8.88

6.88

8.20

8.20

8.20

8.01

8.01

8.01

8.01

8.01

8.66

8.66

8.66

8.66

8.66

8.66

8.66

8.66

12.21

12.21

6.88

6.88

8.66

8.66

3.S6

8.58

8.58

8.U8

8.58

8. 583

8.U8

8.58

8.58

8.58

8.82

8.82

8.62

8.41

8.41

8.41

8.41

8.41

9.00

9.00

9.00

9.00

9.00

9.00

9.00

0.00

11.02

I1.12

8.58

8.58

9.00

9.00

4.00

2.7E-03

2. 7E-03

2.7E-03

2. 7E-03

2.7E-03

2.7£-03

2.7E-03

2.7E-03

2.7E-03

I. OE-00

1.OE-00

I. OE-00

2. 7E-03

2.7E-03

2.7 E-03

2.7E-03

2. 7E-03

2. 7E-03

2. 7E-03

2. 7E-03

Z. 7E-03

2.7E-03

2.7E-03

Z. 7E-03

2. 7E-03

2. 7E-03

2.7E-03

2. 7E-03

2. 7E-03

2. 7E-03

2. 7E-03

2.7E-03

I. 7E-04

1. 7 E-04

2.2E-03

2. 2E-03

2.OE-04

2.OE-04

2. OE-04

2. OE-04

2. oE-04

7.8E-04

7 .8E-04

7.BE-04

4.0E-04

4. 9E-04

4.9E-04

4.9E-04

4.9E-04

2. IE-02

2. IE-02

7.4E-OZ

7. 4E-01

7.4E-02

7.4E-01

7.4 E-02

7.4E-01

I. ZE-OZ

1. 2E-02

6. 7E-04

6.7E-04

6. SE1-03

3. 7E-03

I .IE-02

3.01-04 4.SE-05

3.OE-04 4.5E-05

3.8E-03 0.8E-04

3.8E-03 5.8E-04

3.SE-O4 5.3E-05

3.5E-04 5.3E-05

3.5E-04 5.3E-05

3.5E-04 5.3E-OS

1.5E-04 S.3E-OS

6.2E-04 0.8E-04

6.2E-04 1.8E-04

6.2E-04 1.8E-04

4.3E-02 7.5E-03

4.3E-02 7.5E-03

4.3E-02 7.SE-03

4.3E-02 7.5E-03

4.3E-02 7.SE-03

2.0E-02 5.SE-03

2.OE-02 S.SE-03

7.0E-02 1.9E-02

7.0E-00 1.9E-01

7.OE-02 1.9E-02

7.0E-01 1.0E-00

7.0E-02 1.9E-02

7.0E-01 0.9E-01

9.61-03 1.8E-03

9.6E-03 1.8E-03

I.2E-03 1.8E-04

1.2E-03 1.8E-04

81E-03 1.71-03

3.5E-03 9.7E-04

4.2E-02 2.50-03

8.7E-05

8. 71-05

8. 7E-05

8.7 E-05

9.8E-05

9.8b-05

9.8E-OS

0.8E-05

g.8E-05

3. 9E-04

3.9E-04

3. 9E-04

6.4E-03

6. 4E-03

6. 4E-03

6. 4E-03

6. 4E-03

8.4E-04

8.4E-04

Z. OE-04

2.0E-03

2. OE-04

2.0E-03

2. OE-04

2.1E-03

1. 5E-03

3. SE-03

8.7E-OS

8.7E-05

8. 4E-04

4.8E-04

3.1E-05

9.1E-OS 8.8E-OS

9.1E-OS .81E-05

1.4E-04 9.5E-05

.41E-04 9.SE-05

I.OE-04 9.9E-O0

L.OE-04 9.SE-05

1.OE-04 9.9E-05

0.OE-04 9.9E-05

l.OE-04 9.9E-05

3.5E-03 1.3E-03

3.SE-03 1.3E-03

3.5E-03 1.3E-03

7.OE-03 6.SE-03

.0OE-03 6.5E-03

7.0E-03 6.SE-03

7.OE-03 6.5E-03

7.0E-03 6.5E-03

0.IE-03 9.1E-04

.1IE-03 9.IE-04

I.IE-03 4.6E-04

011E-OZ 4.6E-03

I.IE-03 4.ff-O4

1.1E-02 4.61-03

.1IE-03 4.81-04

I.IE-02 4.61-03

1.7E-03 1.6f-03

3.6E-03 3.SE-03

I.OE-04 8.9E-05

I.OE-04 8.9E-05

9.2E-04 8.6E-04

5.3E-04 4.9E-04

6.0E-04 6.5E-05

9. 5E-05

9.5E-05

I.9E-04

I.9E-04

1. IE-04

1.IE-04

I. IE-04

I.IE-04

I.IE-04

6.6E-03

6.8E-03

6. 6E-03

7.EE-03

7. 6E-03

7.6E-03

7.6E-03

7.6E-03

1. 4E-03

0.4E-03

2.IE-03

2.1E-02

2. IE-03

2.IE-02

2. IE-03

2. IE-02

1.9E-03

3 .8E-03

1. ZE-04

1. 2E-04

1. OE-03

S. 7E-04

1. 2E-03

8.8E-05

8. 8E-05

.0OE-04

1.OE-04

9. 9E-05

9.9E-OS

9.9E-OS

S.9E-05

9.9E-05

2. 2E-03

2. ZE-03

2.2E-03

6.86-03

6.81-03

6.f6-03

6.6E-03

6.f6-03

9.9E-04

9.9E-04

7.3E-04

7.3E-03

7. 3E-04

7.3E-03

7.3E-04

7.3E-03

1. 6E-03

3. SE-03

9.2E-0S

9. ZE-05

8. 9E-04

S.IE-04

I.OE-04

S.9E-OS

9.SE-05

2.4E-04

2.4E-04

I. IE-04

I. IE-04

I.IE-04

I.IE-04

I.IE-04

9.7E-03

9. 7E-03

9. 7E-03

8. E-03

8. 1E-03

S. IE-03

8. IE-03

8. IE-03

1. SE-03

I.SE-03

3 .OE-03

3.0E-02

3.0E-03

3.OE-OZ

3.0E-03

3. OE-02

Z.OE-03

3.9E-03

I.3E-04

1.3E-04

1. IE-03

6. 2E-04

I.7E-03

8. 9E-05

8.9E-05

I IE-04

1.IE-04

I.OE-04

I.OE-04

1.OE-04

I. OE-04

I.OE-04

3. 2E-03

3. 2E-03

3. 2E-03

6.7E-03

6. 7E-03

6. 7E-03

6.7E-03

6.7E-03

I. 1E-03

I. 1E-03

9.9E-04

9.9E-03

9.9E-04

9 .9E-03

9.9E-04

9 .9E-03

I.7E-03

3. 6E-03

9.4E-05

9.4E-tS

9. IE-04

5.2E-04

1.3E-04

I.OE-04

I. OE-04

2 .9E-04

2.9E-04

1.2E-04

1. ZE-04

1. 2E-04

.2ZE-04

1.ZE-04

1.3E-02

1.3E-02

I .3E-02

8. 7E-03

8.7E-03

8. 7E-03

8. 7E-03

8.7E-03

1.9E-03

1.9E-03

4.0E-03

4.OE-02

4.0E-03

4.OE-02

4. 0E-03

4.OE-02

2.1E-03

4. OE-03

1. SE-04

1. SE-04

1.2E-03

6. 7E-04

2. 3E-03

8 .9E-0O

8. 9E-05

I. ZE-04

1. 2E-04

I.OE-04

I.OE-04

I. OE-04

.0OE-04

1.OE-04

4.1E-03

4. IE-03

4. IE-03

6.8E-Q3

6.8E-03

6 .8E-03

6.8E-03

6 .8E-03

1I. IE-03

I.IE-03

1.3E-03

1. 3E-02

1.3E-03

1. 3E-0Z

1 .3E-03

1. 3E-02

1I. 7E-03

3.6E-03

9.7E-05

9. 71-0S

9.3E-04

5.3E-04

1. 7E-04

1. IE-04

I.IE-04

3.5E-04

3. 5E-04

1.ZE-04

I.ZE-04

1.ZE-04

0.2E-04

I.2E-04

1. 6E-02

1. 6E-02

0. 6E-02

9.3E-03

9.3E-03

9 .3E-03

9. 3E-03

9 .3E-03

2.2E-03

2.2E-03

0.0E-03

4. 9E-02

4. 9E-03

4.9E-02

4.9E-03

4.9E-02

2.2E-03

*. IE-03

1.7E-04

I.7E-04

1.3E-04

7. 2E-04

2. 9E-03

9.0E-05

9.0E-05

1.3E-04

1.3E-04

I .OE-04

I.OE-04

1.OE-04

I.OE-04

1. OE-04

S.OE-03

5.0E-03

S.OE-03

5.9E-03

6.SE-03

6.9E-03

6. 9E-03

6 .9E-03

1. ZE-03

1. 2E-03

I.SE-03

1.SE-02

I.SE-03

1.SE-02

1.SE-03

1. 5E-02

1.7E-03

3.5E-03

9.9E-05

9. 9E-05

9. 6E-04

5.5S-04

2. OE-04

I.I1-04

I.IE-04

4. OE-04

4.Ot-04

1.3E-04

3.3E-04

1.31-04

1.3E-04

1.3E-04

I.9E-02

1.9E-02

1.9E102

9.9E-03

9 .9E-03

9.9E-03

9. 9E-03

9.9E-03

2.4E-03

2. 4-03

5.9E-03

5.9E-02

5.9E-03

5.9E-02

S.9E-03

5.9E-02

2.4E-03

4.3E-03

1.8£-04

1.8E-04

1. 3E-03

7. 6-04

3. 5E-03

9. IE095

9.1E095

1.3E-04

1. 3E-04

I.OE-04

I0OE-04

I.OE-04

I. 0E-04

1.OE-04

5.9E-03

5.9E-03

5. 9E-03

7. OE-03

7.OE-03

7.0E-03

7.OE-03

7.OE-03

1.3E-03

1.3E-03

I.8E-03

I.8E-02

1,8E-03

1:.8E-02

1.8E-03

1. 8E-02

1.7E-03

3.6E-03

I. OE-04

I. OE-04

9. 8E-04

5.6E-04

2.4E-04

1.2E-04

1. 2E-04

4.SE-04

4.SE-04

1.3E-04

1.3E-04

1.3E-04

1.3E-04

I.3t-04

2.2E-02

2.2E-02

2.2E-02

I. 0E-02

I. OE-OZ

I. OE-OZ

I.OE-02

I.OE-02

2. 7E-03

2.7E-03

6.8E-03

6.8E-OZ

6.8E-03

6. 8E-02

6. 8E-03

6. 8E-02

2. SE-03

4.4E-03

2.0E-04

Z. OE-04

1.4E-03

8. IE-04

4.0E-03

9. IE-05

9.1E-05

1. 4E-04

1. 4E-04

1.OE-04

1. OE-04

I .Ot-04

I. OE-04

1.0OE-04

5.8E-03

5.8E-03

S.8E-03

7. IE-03

7. IE-03

7. IE-03

7.IE-03

7.IE-03

1. 4E-03

1. 4E-03

2. OE-03

2.0E-02

2. 0E-03

2.OE-02

2. OE-03

2.OE-02

1.6E-03

3. 7E-03

I.OE-04

I. OE-04

I.0E-03

5.7E-04

2. 7E-04
Nos= X + at

where

X(t) = time-dependent failure rate

(1) The computer calculations rounded off to

7.0 x 10-2 y2 .

The calculations for probability of failure were

then performed for 5-year increments of time using

the following general equation

O = random-only failure rate

a = aging failure acceleration parameter

t = time.

Pn = Pn-I + [(AT)(aT)] (3)

where

Pn

The equation for aging acceleration parameter a

based on the moments considerations from Reference 25 is

= new probability for the new 5-year

increase

= probability for the previous 5-year

increment

Pn-I

a =[ (IfC (2) AT = five years

t = twenty-four hours (mission time) where

34 = constant from the deviation 3

f = nonrandom fraction of failures of

the component which are caused by

aging mechanisms

T = average time to failure

a acceleration parameter.

Example Calculations for P,,

Event HPAPPR

aup = 7.0 x 10-2

Pn l = P5 = 2.0 x 104

AT = 5 years

X = mean failure rate.

The time values shown in Table 18 are in years.

The data that was in terms of hours was converted

to years by using 8760 hours per year.

Sample Calculation for (2)

The event chosen for this example is the high

pressure pump A fails to run (HPAPPR). The basic

events are identified in Table I-1 of Appendix I.

The aging failure acceleration parameter is

fup = 0.86

TUP = 8.66 y

= 7.36 x 10-2 y'

aup 3 4 0.86,4.6

= 6.961 x 10-2 Y-2

T = 24 hours = 2.7 x 10-3 years

Po0 = 2.0 x 10-4 + [(5)(7.0

X 10-2)(2.7 X 10-3)]

= 1.145 x 10-3.

The computer calculations rounded off to

1.1 x 10-3.

The probabilities identified under the P heading

in Thble 18 were calculated for each component

event and for each 5-year increment. The base case

[P(5)] values were taken directly from the PRA

data. Each column of data was then used as input

to the IRRAS program for the calculations that

represent the time period for that column. Also

shown in Table 18 are mission time, failure rate,

calculated values for the aging failure acceleration

45
parameters (both upper and lower bounds), and the

component unavailabilities for each 5-year period

up to 40 years. These were the basic data and intermediate calculated results for this aging risk

evaluation.

The IRRAS software program was developed at

the INEL to run on a personal computer.

Version 2.0 was used for this analysis. The fault

trees for the three cases modeled were loaded into

the personal computer and the IRRAS program

run to determine the cut-sets for the significant

sequences. A minimal cut-set is defined as the

smallest combination of component failures,

which if they all occur, will cause the top event of

the fault tree to occur.

The outputs that were selected from the IRRAS

program were the minimum cut-set upper bound

(which is the HPIS unavailability) and the FussellVesely (F-V) importance measure. The F-V importance measure is a measure of contribution of the

event to the system unavailability. A F-V importance value of 0.01 or greater was considered

significant.

The Fussell-Vesely importance is determined by

evaluating the sequence frequency with the basic

event failure probability at its true value and again

with the basic event failure probability set to zero.

The difference between these two results is divided

by the true minimal cut-set upper bound to obtain

the Fussell-Vesely importance. In equation form

this is

always available; (b) that the probabilities for

human error events and maintenance events remain

constant for all time periods; (c) the PRA failure

rates were the random failure rates and apply for

the first 5-year interval at the representative plant;

and (d) when not given in the PRA the mean failure

rates were calculated by dividing the demand failure rate by the estimated time between demands.

In addition, the following modeling assumptions

are from the PRA (Reference 28) for the three cases

considered.

I. It was assumed that flow from the borated

water storage tank (BWST) through either

suction line (valves 3HP-24 or 3HP-25 in

Figure 10) is sufficient for all three HPI

pumps. The letdown storage tank (LDST),

which provides HPI pump suction initially

as the BWST valves open, was not assumed

to be available for emergency HPI operation since makeup to the LDST is limited

to less than 200 gpm. Furthermore, it was

assumed that adequate NPSH to the three

HPI pump exists with the LDST suction

remaining open.

2. Between the discharges of pumps 3HPP3B and 3HP-P3C there is an additional

cross-connection that is not shown in Figure 10. This line contains two normally

closed manual valves, and would serve

only as a backup to the crossover lines containing motor-operated valves 3HP-409

and 3HP-410. Therefore, its availability

was judged to have very little effect on

HPIS reliability and was not modeled.

3. Each injection loop splits into two lines

that provide flow to the cold legs downstream of the reactor-coolant pumps. For

sequences that require flow from only one

HPI pump, it was assumed that only one

of the injection-line splits was needed. For

sequences requiring the function of two

pumps, at least two of the splits were

assumed to be required.

Components that Contribute

Significantly to System

Unavailability

F-V = [F(x) - F(O)I/F(x) (4)

where

F-V = Fussell-Vesely importance

F(x) = minimal cut-set upper bound

(sequence frequency) evaluated with

the basic event failure probability at

its true value

F(O) = minimal cut-set upper bound

(sequence frequency) evaluated with

the basic event failure probability set

to zero.

Assumptions

The assumptions made in performing this analysis are: (a) that all support systems such as IE

power, service water, and low pressure injection are

The significant Fussell-Vesely importance measures calculated for the initial 5-year period and the

40-year period are summarized in Table 19 for each

of the three operating modes. The values at 40 years

46
PZR 9 i From LPI cooler

discharge

(LPR system)

Inside RB I

I 3HP.120

IHP- 3HPRCP- Orifice 1 119

3A W ~ _A

3A2 _

outlet 3HP-126

3HP-1 3HP-410

3HP110 1

seals

Orifice

3HP188

3HP-27

ES-2 opei

OrificiD I

Inside RBj

From LPI

cooler discharge

(LPR system) 3LP-57 From BWST

(LPI system)

6 10 474

Figure 10. Components that contributed significantly to HPIS unavailability for emergency modes of

operation.

47
Table 19. Events with significant Fussell-Vesely importance measures

HPI(1) HPI(2) Recirculation Mode

Event

Name

HP-24MVO

HP-25MVO

HP-26MVO

HP-CPPS

HP-148VVT

HP-CPPR

HP-APPR

HP-BPPR

5 Year

0.4503

0.4503

0.0021

0.0006

0.0003

0.0002

<0.0001

<0.0001

40-Year

Mean

0.5673

0.5673

0.0163

0.0021

0.0149

0.0046

0.0061

0.0061

5 Year

0.2592

0.2590

0.3738

0.0008

0.0039

0.0020

0.0020

0.0055

40-Year

Mean

0.2006

0.1956

0.2325

0.0290

0.2016

0.0615

5 Year

0.1705

0.1705

0.3541

0.0640

0.0297

0.0152

40-Year

Mean

0.3352

0.3352

0.4848

0.0223

0.1054

0.0318

0.0650

0.2941

HPRAPPR

HPRCPPR

0.2166 0.4638

0.1524 0.3180

are the arithmetic mean of the upper and lower

bound values at 40 years from the tables in Appendix 1.

The motor-operated valves "failure to open"

events, HP-24 MVO and HP-25 MVO, are significant for all three modes of operation for both

5 years and after 40 years aging. Event HP-26

MVO is significant for both the HPI(2) and recirculation operating modes (for both 5 and 40 years)

and becomes significant for the HPI(I) operating

mode after 40 years aging. All of the events listed

for the recirculation mode are significant for both

5- and 40-year time periods. Event HP-148 VVT

becomes significant for both HPI(I) and HPI(2)

operating modes after aging 40 years. The pump

events HPCPPS, HPCPPR, HPAPPR, and

HPBPPR become significant after 40 years for the

HPI(2) operating mode, but were not significant

events at 5 years. This is an indication of the aging

effect on the pumps.

The- conclusions' that can be drawn from the

analysis of the F-V importance measure is that

seven components involving 10 events were indentified as contributing significantly to system unavailability for at least one operating mode and time

period. The seven components are highlighted in

Figure 10, and are motor-operated valves HP-24,

HP-25, and HP-26; manual valve HP-148; and

HPIS pumps A, B, and C.

HPIS Unavailability

The cut set quantification reports were obtained

from each run of the IRRAS program. The reports

for the 5-year, 10-year, and 40-year runs are given in

Appendix I. The HPIS unavailability for each

operating mode modeled is given as the minimal

cut-set (mincut) on the reports. These HPIS

unavailabilities are summarized in Tables 20, 21,

and 22 for three emergency operating modes. These

tables present the mean, upper bound and lower

bound for 5-year increments from 5 years to

40 years.

The HPIS mean unavailability for the three

emergency operating modes are plotted in

Figure 11. The one-of-three-pumps-required operating mode had an unavailability starting at

9.59E-5 at 5 years and increased to 1,353-4 at

40 years. The two-of-three-pumps-required operating mode has the highest HPIS unavailability starting at 1.720E-4 at 5 years and increased to 4.280

E-4 at 40 years. Similarly the recirculation mode

48
Table 20. Data for HPIS unavailability

(one of three HPI pumps required)

Year

5

10

15

20

25

30

35

40

Means

9.59E-5

1.O11E-4

1.067E-4

1.120E -4

1.184E-4

1.254E-4

1.328E-4

1.353E-4

Upper Bound

1.048E-4

1.146E -4

1.235E-4

.1.348E-4

1.472E-4

1.604E-4

1.639E -4

Lower Bound

9.737E-5

9.886E-5

1.004E-4

1.019E-4

1.035E-4

1.051E-4

1.067E-4

Table 21. Data for HPIS unavailability

(two of three HPI pumps required)

Year

5

10

15

20

25

30

35

40

Means

1.720E -4

1.921E-4

2.189E-4

2.491E-4

2.889E-4

3.314E-4-

3.823E-4

4.280E-4

Upper Bound

2.053E.- 4

2.513E-4

3.028E-4

3.726E-4

4.485E -4

5.386E - 4

6.194E-4.

Lower Bound

1.788E-4

1.865E -4

1.953E-4

2.05 IE-4

2.142E -4

2.260E-4

2.365E -4

49
Table 22. Data for HPIS unavailability

(recirculation mode)

Year

5

10

15

20

25

30

35

40

Means

8.401E-7

2.OE-6

3.5E-6

5.OE-6

6.8E -6

8.7E-6

1.1E-5

1.27E-5

Upper Bound

2.716E-6

5.105E-6

7.553E -6

1.067E-5

1.394E-5

1.785E-5

2.072E-5

Lower Bound

1.339E-6

1.868E-6

2.405E-6

3.029E-6

3.485E-6

4.130E-6

4.614 E-6

10.3

U)

C',

0

.0

0

C

CO)

10Q-4

1 0-5

10-6 (

I I I I I 1IT

2 of 3 pumps

1 of 3 pumps

I Recirculation

10 7 I I I I I -

5 10 15 20 25

Years

30 35 40

8-3390

Figure 11. HPIS unavailability vs years showing the effect of aging for the three emergency operating

modes.

50
HPIS unavailability started lower at 8.401E-7 at

5 years and increased to 1.27E-5 at 40 years. The

recirculation mode had the greatest increase, but

was still an order of magnitude less than the injection modes. Normally changes in unavailability at

less than an order of magnitude are not important

in final PRA results. Based on that assumption, the

computed change for the two HPI modes would

not be expected to change the results of the full

plant PRA significantly. For the recirculation

mode, the change is more than an order of magnitude but even with the change the unavailability

remains very low. A cursory review of the representative plant PRA indicated that loss of HPI in

the recirculation mode from failure of components

does not appear in any of the important event

sequences and, therefore, a significant change

would not be expected in the full plant PRA.

Estimate for Types of Events That

Cause HPIS Unavailability

An estimate of the contribution of the types of

events to the system unavailability was obtained by

summing all the F-V importance values for a particular event type for a given time period and dividing

by the sum of all the F-V importance values for that

same time period. Because of truncation error and

the possibility that some of the events may be

included more than once in different cut-set combinations, this result will only be a rough estimate of

the contribution of each type of event to the HPIS

unavailability. The results are shown in Table 23.

As demonstrated by the values in Table 23, the

motor-operated valves failure to open is the type of

event with the highest probability of contributing

to HPIS unavailability for the two HPIS, injection

modes of operation. For these same two modes of

operation the human error events have the second

highest probability followed by the events associated with maintenance. The change in probabilities

from 5 years to 40 years may vary either up or

down depending on how the cut-sets combine. The

most noteable probability increase with age was for

the manual valves and pumps in the HPI (two of

three pumps) mode. The manual valves probability

increased from 1.4%o to 21%70 and the HPIS pumps

increased from 0.2% to 9.3%.

For the recirculation mode the maintenance

events were the most significant, 41°7o at 5 years,

but dropped to only 7.5 qo at 40 years. All the other

events showed an increased with age between

5 years and 40 years. The motor-operated valves

continued to be a contributor with 20%o at 5 years

and 39% at 40 years.

In general, all the component events had an

increase in unavailability over the 40-year period

for all three operating modes, except for motoroperated valve in the HPI (two of three pumps)

mode, which decreased, and the check valves in the

HPI (one of three pumps) mode, which had no

change.

51
Table 23. Probability for type of events causing HPIS unavailability

Percent Contribution to HPIS Unavailability

HPI

(1 of 3 pumps)

HPI

(2 of 3 pumps)

Recirculation

Mode

Event Type

Human error

Segment of system out

of service for

maintenance

Motor-operated valves

failure to open

Check valves

failure to open

Manual valves

failure to transfer

:HPIS pumps

failure to run or

failure to start

5 yr

35

5

59

0.8

0.05

0.02

40-yr

Mean

26

5

66

0.8

0.8

0.7

5 yr

39

10

49

0.4

1.4

0.2

40-yr

Mean

20

20

30

0.5

21

9.3

40-yr

5 yr Mean

25 27

41 7.5

20 39

0 0

13 22

0.9 3.9

52
CONCWSIONS

The conclusions for the HPIS aging research are

based on a detailed study of one plant and supplemented by information from generic data bases.

Experience from the generic data bases shows that

the motor-operated valves were the most troublesome components. Some failures and aging mechanisms are not being detected by current inspection

and surveillance methods. The study indicates that

aging causes only a moderate increase in system

unavailability and that motor-operated valves are

the components that contribute most often to the

system unavailability. Conclusions for the specific

study objective are presented in the following

sections.

Operating Experience

The HPIS operating experience was evaluated by

reviewing generic data bases and plant specific

data. The operating experience from the NPRDS,

covering B&W and Westinghouse plants, indicated

that approximately 57%7 of the component failures

lead to system degradation, but because of system

redunancy, only 0. 7% of HPIS component failures

caused loss of system function. The data also

showed that 21.356 of these failures were aging

related.

The NPE data covering all PWRs were reviewed

and four types of components were identified as

having the highest frequency of failure. They were

valves-35%, I&C-1956, pumps-155o, and piping7M. Both NPRDS and LERS had the same relative

ranking for these components. The corrective

maintenance ranked I&C and piping failures higher

than valve and pump failures because of many subcomponents failures that were repaired on these

particular components.

The conclusion for the data base reviews is that a

significant number of HPIS component failures

have been aging related. The term "aging related"

is used because the information given in data bases

often requires further analysis to determine extent

of aging contribution to the failure. The conclusions for specific problems related to aging are

summerized in following sections.

Specific Problems Related to

Aging

Problems with boric acid systems have occurred.

Leaks of borated water onto carbon steel caused

corrosion of the carbon steel. Such corrosion can

occur at a rate greater than I in. per year as proven

by both laboratory tests and plant experience.

Boric acid crystals have caused blockage and in one

instance precipitated out in an HPI pump causing it

to malfunction. The latter was due to an internal

valve leak.

The thermal sleeve and nozzle cracking problem

experienced in B&W plants in 1982 was attributed

to thermal fatigue. Thermal cycling occurred when

the makeup flow cycled on and off and when system injection flow occurred. The problem was corrected at the representative plant studies by

redesign of the thermal sleeve and maintaining a

continuous small makeup flow. An augmented

inservice inspection plan was also implemented.

The HPIS interacts with many of the other reactor systems. This makes it particularly vulnerable

to common mode failures when a problem in

another system can prevent HPIS from performing

its function. For example, HPI pump seal cooling is

supplied by the service water system. The HPI itself

supplies RC pump seal cooling and RC makeup

water during normal operation.

IEEE 323-1974 was the industry standard used

to address HPI electrical equipment qualification

and makes reference to the use of historical data

and the analysis of failures and trends. The design

and maintenance data should be reviewed periodically (approximately 18 months) to ensure that the

design qualified life has not suffered. The review

should include thermal or cyclic degradation resulting from accumulated stresses triggered by abnormal environmental conditions and the normal wear

due to its service condition.

Problems with the ECCS piping cracks have

occurred at Farley-2 and Tihange-1. These two

plants were Westinghouse plants utilizing the high

pressure injection pumps for both charging during

normal operation and injecting emergency core

coolant during a LOCA. The pipe cracking was

attributed to thermal fatigue that occurred because

the failure of valves in the safety injection system

allowed relatively cold water to flow back into the

primary system.

53
The B&W system that has the dual purpose

HPIS pump would potentially have the same problem. The difference in the B&W system is that it

uses the same nozzle for makeup and injection.

Aging Assessment

The aging assessment of the impact of aging on

HPIS operability involves a number of factors

including the identification of stressors, degradation mechanisms, and failure modes. The various

potential electrical, mechanical, and environmental stressors were indentified and degradation

mechanisms determined for both passive and active

components. These factors were summarized in

Table 15 along with the various failure modes and

ISI methods.

Materials in the HPIS susceptible to aging

include seals and packing material in pumps and

valves. Carbon steel materials in other systems

exposed to boric acid for some period of time will

corrode. Electrical components in the I&C subsystem are subject to degradation of insulation, corrosion, and wear failures. The stainless steel piping

ages from thermal fatigue. However, material wear

in pumps, valves, and relay contacts is a normal

aging process.

Inspection, Surveillance, and

Monitoring

The surveillance requirements in the technical

specifications for a plant and the ASME Boiler and

Pressure Code comprise the testing requirements

for the HPIS. Additional inspections for leaks may

be performed periodically along with functional

tests on major HPIS components if required by the

utility. For example, after maintenance, functional

tests may be necessary to verify operation.

Included in the surveillance methods are visual

inspections, operational testing, and calibrations

for the HPIS as a system. Individual components

may have additional tests. For example, the limitorque valve operators require periodic 18-month

inspections as part of routine maintenance for

cleaning and lubrication. Rotork valve operators

are qualified for 40-year life or 2000 open-close

cycles, therefore, no preventive maintenance program is recommended. Rotork ensures reliability by

carefully sealing out the environment during installation. Routine maintenance of inner parts is

deemed likely to cause defective sealing and, should

be avoided. The exception is that seals and lubricants subjected to radiation and heat aging must be

replaced based on the qualified life.

Some advanced methods for IS&M on mitigating

IGSCC in piping developed by EPRI have included

stress improvement methods, alternate materials,

and improved water chemistry. Another advanced

monitoring method developed by the INEL for

cable monitoring is the ECCAD system. This technology has been transferred to industry and at least

two operating utilities have used the system.

Another advanced surveillance system is the

motor-operated valve analysis and test system

(MOVATS). Since 1984, the MOVATS system has

been used as a diagnostic tool for motor-operated

valves.

Maintenance

Maintenance practices for major HPIS components follow vendor's recommendations closely, in

addition to plant specific maintenance procedures.

Corrective maintenance is minimized by observing

good operating practices and precautions. The benefits of preventive maintenance include higher system availability, increased life, and higher

reliability.

In general, up to 35% of nuclear plant abnormal

occurrences may be due to faulty maintenance and

surveillance testing. In order for preventive maintenance to be effective, it must be applied to equipment that has degradation effects that are

detectable. Also, methods of detecting degradation

must be available.

For the HPIS, the recommendation is to continue the maintenance recommended by the vendors for major components such as pumps and

valves. In addition, an enhanced inspection program for detecting pipe leaks and cracks (including

nozzles) using advanced monitoring methods

should be implemented.

HPIS Unavailability Assessment

This HPIS aging system unavailability assessment demonstrates the linear aging model technique using failure cause data can be used to update

a representative plant PRA with time to calculate

the change of unavailability as the system ages.

This approach, as presented, is suitable for exploratory investigations and identification of components that contribute significantly to system

unavailability. The results of this assessment apply

to the plant studied and only to other plants of the

54
same design configuration. More advanced aging

reliability models with better methods of verifying

that failure data exhibits aging and for calculating

the time dependent failure rates are being

developed.

Seven components involving 10 events were identified as risk significant for the HPIS by the Fussell

Vesely measure. These events were motor-operated

valves HP-24, HP-25, and HP-26; failure to operate, HPIS pumps A, B, and C; failure to start or

run, and manual valve HP-148; failing to transfer.

The HPIS unavailability increased over the aging

period modeled from 5 to 40 years for all three

emergency operating modes. Even after the

increases, the unavailabilities of the system are not

much higher for the high pressure injection modes

and the full plant PRA would not be expected to

change significantly. For the recirculation mode,

the unavailability increased significantly, but even

after the increase, the value remained very low. A

cursory review of the representative plant PRA

indicated this change would not be expected to

change the results of the full plant PRA.

The equipment failures that have the highest contribution to HPIS unavailability are the motoroperated valves failure to open.

55
REFERENCES

1. NuclearPlantAging Research (NPAR) Program Plan, NUREG-1 144, U.S. Nuclear Regulatory Commission, 1985.

2. W. L. Greenstreet, G. A. Murphy, D. M. Eissenberg, Aging and Service Wear of Electric MotorOperated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants, NUREG/CR4234, Vol. 1 (ORNL-6170/Vl), Martin Marietta Energy Systems, Inc., Oak Ridge National

Laboratory, June 1985.

3. J. L. Crowley and D. M. Eissenberg, Evaluation of the Motor-Operated Valve Analysis and Test System (MOVATS) to Detect Degradation, Incorrect Adjustments, and Other Abnormalities in MotorOperated Valves, NUREG/CR-4380, ORNL-6226 Oak Ridge National Laboratory, January 1986.

4. Data Summaries of Licensee Event Reports of Valves at U.S. Commercial Nuclear Power Plants,

NUREG/CR-1363, Rev. 1, October 1982.

5. L. C. Meyer, NuclearPlantAgingResearch on ReactorProtection Systems, NUREG/CR-4740, EGG2467, Idaho National Engineering Laboratory, January 1988.

6. B. M. Meale and D. G. Satterwhite, An Aging Failure Survey of Light Water Reactor Safety Systems

and Components, NUREG/CR-4747, EGG-2473, Vol. 1, Idaho National Engineering Laboratory,

July 1987.

7. V. N. Shah, Residual Life Assessment of Major Light Water Reactor Components-Overview,

NUREG/CR-4731, EGG-2469, Vol 1, Idaho National Engineering Laboratory, June 1987.

8. C. L. Atwood, Common Cause Fault Rates for Pumps, NUREG/CR-2098, EGG-EA-5289,

February 1983.

9. W. H. Hubble and C. Miller, Data Summaries of Licensee Event Reports of Valves at U.S. Commercial Nuclear Power Plants, NUREG/CR-1363, EGG-EA-5816, Rev. 1, October 1982.

10. Cracking in Piping of Makeup Coolant Lines at B& WPlants, U.S. NRC Information Notice 82-09,

March 31, 1982.

11. Pipe Cracks in Stagnant Borated Water System at PWR Plants, U.S. NRC Information Notice I E

79-19.

12. S. H. Bush, P. G. Nessler, R. E. Dodge,Aging and Service Wear ofHydraulic and Mechanical Snubbers Used in Safety-Related Piping and Components of Nuclear Power Plants, NUREG/CR-4279,

PNL,5479, Vol. 1, February 1986.

13. A Prioritization of Generic Safety Issues, NUREG-0933, December 1983.

14. Standard Technical Specificationsfor Babcock and Vilcox Pressurized Water Reactors, NUREG-01 03,

Revision 4, Fall 1980.

15. ASME Boiler and Pressure Vessel Code. An American Standard, ANSI/ASME BPV-l 1-1, American

Society Mechanical Engineers, New York, 1983.

16. Safety Injection Pipe Failure, USNRC Information Notice 88-01, Janaury 27, 1988.

56
17. Thermal Stresses in Piping Connected to Reactor Coolant Systems, USNRC Information Notice 88-08

Supplement No. 1. June 24, 1988.

18. M. R. Dinsel, M. R. Donaldson, F. T. Soberano, In Situ Testing of the Shippingport Atomic Power

Station Electrical Circuits, NUREG CR-3956, Idaho National Engineering Laboratory, April 1987.

19. QualityAssuranceProgram Requirements (Operation), Regulatory Guide 1.33, U.S. Nuclear Regulatory Commission, February 1978.

20. ANS-3.2 Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power

Plants N18.7, 1976.

21. Maintenance and Surveillance Program Plan (Turkey Point Unit 3), NRC Docket Number 50-250,

U.S. Nuclear Regulatory Commission, October 10, 1984, (Available from NRC Public Document

Room).

22. D. J. Sheliga, "Calculation of Optimum Preventive Maintenance Intervals for Electrical Equipment," 1980 IEEEIndustrial Applications Society Meeting, Cincinnati, Ohio.

23. Interim Staff Position on Environmental Qualfflcation of Safety-Related Electrical Equipment,

NUREG-0588, Rev. 1, U.S. Nuclear Regulatory Commission, July 1981.

24. Qualifying Class IE Equipment for Nuclear Power Generating Stations, IEEE 323-1974, Institute of

Electrical and Electronics Engineers.

25. Environmental Qualijcation of ElectricEquipmentforNuclear Power Plants, Regulatory Guide 1.89,

U.S. Nuclear Regulatory Commission.

26. Standard Review Plan for Review of Safety AnalysisReportsforNuclearPower Plants, NUREG-0800,

July 1981.

27. W. E. Vesely, Risk Evaluation of Aging Phenomena: The Linear Aging Reliability Model and Its

Extensions, NUREG/CR-4769, May 1987.

28. Oconee PRA, A Probabilistic Risk Assessment of Oconee 3, The Nuclear Safety Analysis Center,

Electric Power Research Institute Palo Alto, California and Duke Power Company, Charlotte, North

Carolina, NSAC 60, Vol. 1, Vol. 2, Vol. 3, and Vol. 4, June 1984.

29. K. D. Russell and M. B. Sattison, Integrated Reliability and Risk Analysis Systems (IRRAS) User's

Guide-Version 2.0 draft NUREG/CR-51 11, EGG-2535, March 1988.

30. B. M. Meale and D. G. Satterwhite, An Aging Failure Survey ofLight Water Reactor Safety Systems

and Components, NUREG/CR4747, EGG-2473, Vol. 2, July 1988.

57
APPENDIX A

REACTOR COOLANT MAKEUP SYSTEM, PURIFICATION SYSTEM,

AND COOLANT INJECTION AND RETURN FOR RC PUMP SEALS

A-1
APPENDIX A

REACTOR COOLANT MAKEUP SYSTEM, PURIFICATION SYSTEM,

AND COOLANT INJECTION AND RETURN FOR RC PUMP SEALS

The part of the HPIS used during both normal

and emergency operation is shown in Figure A-1.

During normal operation, the High Pressure Injection system receives reactor coolant as letdown

water for purification and chemistry control. The

letdown water is cooled in one or both of the parallel connected letdown coolers in order to prevent

damage to the purification system ion exchange

resins. Cooling water for the letdown coolers is provided by the Component Cooling system. The

cooled letdown is routed through the N16 Decay

Tank, which provides sufficient flow delay to allow

for decay of radioactive nitrogen-16. The RC

makeup system is shown in Figure A-2. (The N-16

decay tank is not shown on Figure A-2.)

Flow from the decay tank to the purification system is controlled by a fixed block orifice. In the

event additional flow is necessary, an orifice bypass

valve is provided.

In the purification system, the letdown stream

enters the purification demineralizer through the

purification demineralizer supply header, and

leaves through the purification demineralizer discharge header. A valve is provided between the supply and discharge headers to allow the

demineralizer to be bypassed if necessary. A threeway valve in the discharge header allows the demineralizer discharge to be directed to either the

letdown filters or the debarking demineralizer supply header. From the debarking demineralizer supply header, the letdown can be sent to the Coolant

Treatment system or through the debarking demineralizer. The debarking demineralizer effluent is

sent via the debarking demineralizer discharge

header to the letdown filters.

Feed and bleed controls allow batch addition of

demineralized water and/or concentrated boric

acid for control of RC system boric acid concentration and restores the letdown storage tank level as

necessary. The makeup batch enters the system

upstream of the letdown filters.

Letdown from the purification system and/or

makeup from the Coolant Treatment system passes

through one or both of the two parallel connected

letdown filters to the letdown storage tank.

Downstream from the letdown filters, return

water from the reactor coolant pump seals is supplied to the letdown storage tank. The seal return

water is taken from each of the four RC pumps and

is cooled by one or both of the parallel connect RC

seal return coolers before entering the letdown storage tank. The seal return and injection system is

shown in Figure A-3.

The letdown storage tank serves as the normal

suction source for the HP injection pumps; however, during safety injection, the borated water

storage tank is automatically connected as an additional source. These sources provide injection

water until RC system pressure is low enough for

low pressure injection.

The three high pressure injection pumps are connected in parallel with cross connecting suction and

discharge headers thus enabling any pump to supply any discharge line. During normal operation,

only one pump is required for normal makeup and

reactor coolant pump seal injection. A second

pump serves as a backup for normal operation. All

three HPI pumps are started automatically by a signal from the engineered safeguards system in the

event of a low RC system pressure or a high reactor

building pressure condition.

Three principal flow paths are provided from the

HPI pump discharge header: an injection line to

reactor coolant Loop A, a reactor coolant pump

seal injection line, and an injection line to reactor

coolant loop B. Normally HPI pump A or B is

operated to supply makeup flow to RC loop A and

RC pump seal injection flow; pump C is used only

for injection to loop B in the event of an accident.

An additional emergency cross connect line is provided from the HPI pump discharge header to both

RC loops as a backup for safety injection.

The water supplied for reactor coolant pump seal

injection is routed through the two parallel connected seal filters before being distributed to the

four reactor coolant pumps.

Description of Components

Letdown Coolers. The letdown coolers are of

the shell and spiral tube type with a shell material

of carbon steel and a tube material of stainless

steel. The two letdown coolers are installed in

A-3
188 ES2 open 148 114 113 \__.} 5

Orifice Orifice X ~~~~~~~HP-P3C ES-2 1 on

From LPI 3LP-5#

cooler discharge _ I (LPR system) 3LP-57

Inside RB]

Figure A-1. Part of the HPIS used during normal and emergency operations.

From BWST

(LPI system)

a io 474

A-4
6 10 473

Figure A-2. Reactor coolant makeup and purification system.

A-5
RCP.3A1

r 3HP-21

R~CP.3A2 -LES 2)

To seat retufn

RCP-381- SHP-20 coolers an~d LOST

RCP.38l (ES1).l)

RS

AtCP soa "Ir

SHP.1 32

RCP seal

I rot 3A

3HP. 22

I to DUl

Figure A-3. RC pump seal return and injection system.

parallel with one normally in use and the second as

a spare. For maximum letdown flow, both coolers

are required.

Seal Return Coolers. The seal return coolers are

of the shell and tube type with carbon steel as the

shell material and stainless steel as the tube material. Of the two coolers installed in parallel, one is

normally in operation.

Makeup Filters. The material of all wetted parts

of the filter vessel is stainless steel. The filter elements themselves are of depth type design. The

materials are wound cotton around a stainless steel

core. Two makeup filters are installed in parallel

with one normally in use.

Reactor Coolant Pump Seal Filters. The filter

assembly consists of the filter vessel and one element. The filter vessel is of the in-line type with

flanged ends. The element is pleated stainless

steel. One of the two filters is required during

normal operation. A bypass line is installed

around the filters.

Letdown Storage Tank. The total volume of the

stainless steel letdown storage tank is 600 ft3. The

water content under steady state conditions is about

350 ft3 while the rest is gas space. The tank is

equipped with a manway, 4-in. inlet and outlet connections, a 1-in. relief valve connection, a 1-in. vent

connection, two 3/4-in. sampling connections (one in

the gas space, the other in the water space), a 1-in.

inlet connection for H2 and N2 and connections for

level, pressure, and temperature indication.

The inlet connects to a nozzle that sprays the

incoming water into the gas space. For hydrogen

and nitrogen addition, a ring header in the water

space is provided.

Purification Demineralizers. The purification

demineralizers function is to purify the portion of

the reactor coolant let down through the letdown

coolers. Each demineralizer is sized for a flow rate

of 70 gpm.

The inlet flow is distributed in an inlet header.

The outlet is equipped with a similar header having

a 110 mesh stainless steel screen to contain the resin

beads. The vessel contains a manway, a resin inlet,

resin outlet, and a vent connection to the high

A-6
activity waste tank in addition to the water inlet and

outlet. The material of all parts wetted by the process fluid is 304 stainless steel.

At the inlet line of each demineralizer, a small

capacity relief valve is installed in order to prevent

pressure buildup during ambient temperature

changes in the event that the demineralizer is isolated. The reactor coolant quality specifications are

given in Table A-I.

Reactor Coolant Pump Seal Return Cooler

Type Shell and tube

Heat Transferred, Btu/hr 2.2 x 106

Seal Return Flow, lb/hr 1.25 x 105

Seal Return Temperature 145 x 127

Change, 'F

Material, shell/tube CS/SS

Design Pressure, psig 150

Design Temperature,0 F 200

Recirculated Cooling Water 1.25 x 105

Flow (ea.), lb/hr

Code ASME Sec. III-C & VIII

Deborating Demineralizers. The debarking

demineralizers are used to reduce the boron concentration in the reactor coolant during the latter

part of the core lifetime when the boron concentration must be lower. Each demineralizer is sized for a

70 gpm flow rate, the same as the purification

demineralizers.

Letdown Storage Tank

Volume, ft3

Design Pressure, psig

Design Temperature,°F

Material

Code

600

100

200

SS

ASME Sec. III-C

Purification Demineralizer

Type

Component Data

Letdown Cooler

Type

Heat Transferred, Btu/hr

Letdown Flow, lb/hr

Letdown Cooler Inlet/Outlet

Temperature, F

Material, shell/tube

Design Pressure, psig

Design Temperature,°F

Component Cooling

Water Flow (ea.), lb/hr

Code

Material

Volume, ft3

Flow, gal/min

Vessel Design Pressure, psig

Vessel Design Temperature,0 F

Code

Mixed bed, boric acid saturated

Ss

85

70

150

200

ASME Sec. III-C

Shell and spiral tube

16.0 x 106

3.5 x 104

555/120

CS/SS

2,500

600

2 x 105

ASME Sec. IlI-C & VIII

Letdown Filter

Design Flow Rate, gal/min

Material

Design Temperature

Design Pressure

Code

80

Ss

200

150

ASME Sec. III-C

Table A-1. Reactor coolant quality

Total solids, including dissolved and undissolved material, but excluding 7LiOH

and H3BO3 (max), ppm

Boron (max), ppm

Lithium as 7Li, ppm (when required for pH adjustment)

pH at 770F

Dissolved oxygen as 02

Chlorides as Cl- (max), ppm

Hydrogen as H2, std cc/L H20

Fluorides as Fl- (max), ppm

a. Equivalent range as 7LiOH is 1.455 to 5.82 ppm.

b. Equivalent pH at 600°F is 6.8 to 7.8.

c. With proper H2 0 specification at critical condition, dissolved 02 is assumed not to be present.

1.0

2,270

_c

0.15

15-40

0.15

A-7
APPENDIX B

LOW PRESSURE INJECTION SYSTEM INTERFACE WITH HPI

B-I
APPENDIX B

LOW PRESSURE INJECTION SYSTEM INTERFACE WITH HPI

The LPI system is connected to the HPIP system in

two respects: (a) Under accident conditions they both

take their suction from the borated water storage tank

(the reactor building spray system also takes suction

from the BWST), and (b) for high pressure recirculation

mode of operation, the HPIS takes suction from the

LPIS output.

The HPRS mode is initiated when the BWST borated

water is depleted and conditions require high pressure

recirculation. Both HPI and LPI pumps are shut down

if operating and the LPI valves are reconfigured by

opening the two valves of the reactor building emergency sump (valves LP-19 and LP-20), starting at least

one LPI pump, closing valves LP-21 and LP-22 from

the BWSI, and opening valves LP-15 and LP-16 to the

suction of the HPI pumps.

B-3
APPENDIX C

ELECTRICAL POWER REQUIREMENTS FOR HPIS COMPONENTS

C-1
APPENDIX C

ELECTRICAL POWER REQUIREMENTS FOR HPIS COMPONENTS

The voltage requirements for the various valves

and HPI pumps are given in Table C-i. The IE

safety power system supplies the power for the HPI

components. Grid voltage is regulated to + 10°70.

Each plant must decide what voltage degradation

represents an acceptable level.

C-3
Table C-1. HPIS pumps and valve electrical power requirements

Voltage

Description Requirement

HPI Pump Aa 4 kV

HPI Pump Ba 4 kV

HPI Pump Ca 4 kV

HP-I, Letdown cooler A inlet isolation valve 208 V

HP-2, Letdown cooler B inlet isolation valve 208 V

HP-3, Letdown cooler A outlet valve 208 V

HP-4, Letdown cooler B outlet valve 208 V

HP-5, Letdown isolation valve 125 Vdc

HP-6, Letdown orifice inlet valve 208/120 Vac

HP-8, Purification demineralizer inlet valve 208/120 Vac

HP-9, Spare purification demineralizer inlet valve 208/120 Vac

HP-I l, Spare purification demineralizer outlet valve 208/120 Vac

HP-13, Purification demineralizer bypass valve 208/120 Vac

HP-14, Bleed control valve 208 V

HP-16, Makeup isolation valve 208 V

HP-17, Makeup filter isolation valve 208/120 Vac

HP-18, Makeup filter isolation valve 208/120 Vac

HP-19, Makeup filter bypass valve 208/120 ac

HP-20, RC pump seal return valve 208 V

HP-21, RC pump seal return isolation valve 125 Vdc

HP-22, Letdown storage tank drain valve 208/120 Vac

HP-23, HPI pump suction valve 208 V

HP-24, Borated water to HPI pump A valve 208 V

HP-25, Borated water to HPI pump C valve 208 V

HP-26, HPI to loop A reactor inlet valve 208 V

HP-27, HPI to loop B reactor inlet valve 208 V

HP-98, HPI pumps suction crossover valve 208 V

HP-1 15, HPI pumps discharge crossover valve 208 V

HP-226, RC pump A2 seal return isolation valve 208 V

HP-228, RC pump Al seal return isolation valve 208 V

HP-230, RC pump B2 seal return isolation valve 208 V

HP-232, RC pump BI seal return isolation valve 208 V

HP-409, HPI isolation from HPI pump crossover valve 600 V

HP-410, HPI isolation from HPI pump crossover valve 600V

GWD-19, Letdown storage tank vent valve 208/120 Vac

a. In the event of a loss of 4 kV power, emergency power can be provided to one HPI pump from the auxiliary service water

switchgear.

C-4
APPENDIX D

ENGINEERED SAFETY FEATURES ACTUATING SYSTEMS FOR HPIS

D-1
APPENDIX D

ENGINEERED SAFETY FEATURES ACTUATING SYSTEMS FOR HPIS

The HPIS is one of the Engineered Safeguards

Systems that is automatically actuated by the Engineered Safety Feature Actuating System (ESFAS).

The other safeguards systems actuated by ESFAS

include the low-pressure injection system, reactor

building isolation, reactor building cooling system,

and the reactor building spray system.

The generic ESFAS diagram for a representative

B&W plant is shown in Figure D-1. Figure D-1 is

presented to illustrate the interconnections between

major ESFAS subsystems. The three blocks on the

left side of Figure D-1 are identical analog subsystems that receive pressure transducer inputs. The

output lines from the analog subsystems go to the

two identical center logic subsystems where the two

out of three logic decides whether an ESF system is

actuated. On the right side of the figure are the five

ESF actuated systems. Although only the HPIS

will be addressed in detail in this report, the other

systems are covered only to show system interactions. The simplified one-line diagram for initiation of the High Pressure Injection (HPI) systems

(ESFAS channels 1 and 2) is shown in Figure D-2,

along with the related aging data for the various

components. A detailed discussion of ESFAS is

covered in Reference D-1.

In case of a LOCA, the HPIS will be initiated

when the RC pressure drops to 1550 psig. If the

high pressure injection channels fail to maintain

RC pressure and it continues to decrease, then at

600 psig the core flooding system will automatically dump water into the core. This will happen

automatically, with no means for manual control.

If the RC pressure continues to drop, at 550 psig

the low pressure injection channels (3 and 4) will be

actuated in the same manner as the high pressure

channels. Anytime there is a large RC leak or rupture, the coolant will flash to steam as it escapes

from the system. This will cause the building pressure to increase. When the building pressure

increases to 3.0 psig, the building cooling and isolation channels will be actuated as well as high and

low pressure injection channels if they have not

already been actuated.

The ESFAS Unit Control Module provides the

output contacts to drive the final actuating device

control circuitry. Interfacing capability is also provided to allow manual control of the final device

after safety action has been initiated. If the

remotely mounted manual control equipment is

activated before safety action is initiated, the system safety function will not be inhibited.

The engineered safeguards for actuated devices

on ESFAS channels 1 and 2 are shown in

Table D-1, they include HPI pumps, valves, and IE

power breakers and busses.

Electrical power for operating the HPIS control

relays is taken from the power source for the associated device. Loss of power, a relay, or a device does

not impair system functions because there is a second redundant device for each required function.

D-3
Wide-range

RCS pressure

tranSmitter

Narrow-range

RB pressure

transminer

RB pressure

Switches

Wide-range

RCS pressure

transmitter

o Narrow-range

- RMB pressure I

transminer

RB pressure

Swi:Ches

Wide-range

PCS pressure

transmitter

Narrow-range

RB pressure (

traismitler

RB pressure

Switches

Pressure signals Trip signals Actuation signals

Figure D-1. Engineered safeguard system.
Reactor _ I ESPS cabinet in

building I control room

Manual

bypass

Inhibit

High

pressure

Injection

system See Note 1 activation

Reactor building pressure I

ESPS analog subsystem 2

ESPS Channel 1

See Note 2

Component Reactor

coolantalibratmBuffelOgi 2og3 tegt conitro pressure power lCaibrate sBfler Bistable Bistable logic 23 test conitrol Comments

transmitter supply tetaplfe

Environment

Temperature 120F average 50 to 80 F

Radiation 3 x 104 RAD INIA

p interfaces

6" EO

Pressure TAP 110V power and module Interlock

10 year tile 1 40 years

Testing Monthly Funtionat test, response time test at 18 months

Catibration 18 months 18 months

Maintenance or relueling 18 months

Signal 0 to 2000 PSI 4.20 ma 0 o to 10 Vdc

Physical

data

Stressors See Note I

Indicators Drilt. Drift, failure. contact resistance change, binding or bent parts

of moisture

degradation intrusion.

wearout,

failure,

signal

variance from

similar channels

1S9 Note e: The transmitter, cable, reactor building penetration, terminal strip and connectors are Identical to that

of the RPS RC presuure channel

Note 2; The low pressure Injection Is Identical.

Figure D-2. ESFAS channel for initiating the high pressure injection with related aging and engineering

data.

....
Table D-1. Engineered safeguards actuated devices for Channels 1 and 2

Devices Channel I Channel 2 Channels 1 and 2

HPI pumps and

valves

HP-PA

HP-24

HP-26

HP-3

HP-4

HP-20

HP-PC

HP-25

HP-27

HP-5

HP-21

Emergency power

source start

(Channel B)

HP-PB

Electrical power

sources and

Equipment

Emergency power

source start

(Channel A)

LOAD SHED &

STBY. BKR. I

Standby BUS

FEED BKR. I

Emergency power

source start

(Channel B)

LOAD SHED &

STBY. BKR. 2

Standby BUS

FEED BKR. 2

D-6
REFERENCES

D-1. L. C. Meyer, Nuclear Plant Aging Research on Reactor Protection Systems, NUREG/CR-4740,

EGG-2467, Idaho National Engineering Laboratory (January 1988).

D-7
APPENDIX E

COMPONENT DESIGN INFORMATION

E-1
APPENDIX E

COMPONENT DESIGN INFORMATION

Piping

The high pressure injection and low pressure injection lines are designed for normal operating conditions. The normal operating system temperature and

pressure requirements are greater than those encountered during emergency operation. The low pressure

injection system piping and valves are subjected to

more severe conditions during decay heat removal

operation than during emergency operation and, therefore, operate well within the design conditions.

Table E-1 gives the design pressure and temperatures of

these systems. To ensure system integrity, major piping

has welded connections except where flanges are dictated for maintenance reasons. All piping for the HPI

system is stainless steel. Pipe sizes include 1, 2 1/2, 4,

6, 8, and 14 in.

Generic guidance is needed for predicting and mitigating and the effects of piping damage due to flowassisted erosion, corrosion, or cavitation. Results from

the Surry-2 rupture on feedwater pipes show that erosion and corrosion can occur in single-phase flow.

What lessons from that incident can be applied to HPI

piping is not yet known. Some factors that can effect

piping damage include pipe design, fluid dynamics,

piping material, and water chemistry.

HPI Pumps

Each HPI pump can deliver 450 gpm at

1700 psig reactor vessel pressure. Water is drawn

through a single suction header from the BWST

and pumped through injection lines that enter the

reactor building on opposite sides. Each injection

line splits into two lines inside the reactor building,

but outside the secondary missile shield, to provide

four injection paths to the RCS cold legs. The four

Table E-1. Engineered safeguards piping design conditions

Temperature

(OF)

Pressure

(psig)

High Pressure Injection System

From the pump discharge to upstream of the

stop check valves inside the secondary

shielding.

200 3,050

High pressure injection pump.

From upstream of the stop check valves to

the reactor inlet line.

Low Pressure Injection System

(Portion used with HPI)

From the borated water storage tank to

upstream of the borated water storage tank

outlet valves.

From upstream of the borated water storage

tank outlet valve to upstream of the electric

motor operated valves in the borated water

feed lines.

200/150 2,800/3,050

650 2,500

150 Static

200 100

E-3
connections are located between the reactor coolant pump discharge and the reactor inlet nozzles.

Operation of the HPIS pumps requires lubrication oil cooling and pump seal cooling. Charging

pump cooling is accomplished via the Low Pressure

Service Water (LPSW) System where heat generated in the pump lubricating oil and seals is

removed via heat exchangers and transferred to

ultimate heat sink. Electrical power is supplied by

three independent 4160 V buses to the three HPI

pumps. After receiving an actuation signal, the

HPI pumps can reach full speed within six seconds. The HPI pump data are given in Table E-2

and the HPI pump characteristics curves of total

dynamic head and net position suction head are

shown in Figure E-1. The HPI pumps are vertical

stage-centrifugal pumps with mechanical seals.

The wetted parts of the pumps are stainless steel.

Valves

All remotely operated valves in the Emergency Core

Cooling Systems are manufactured and inspected in

accordance with the intent of the ASME Nuclear

Power Piping Code B31.7. Liquid penetrant, radiography, ultrasonic, and hydrotesting are performed as the

code classification requires.

The seats and discs of these valves are manufactured from materials free from galling and seizing.

All valve material is certified to be in accordance

with ASTM specifications. All remotely operated

valves in these systems are of the backseating type

and equipped with stem leak-off provisions.

The valves and their design conditions are listed

in Table E-3. Actual system operating conditions

are significantly less severe than design conditions

shown in Table E-3.

1. Tank Problems - While there have been a

number of events at PWRs involving tanks

with boron concentrations or fluid levels that

were out of specification, most of these events

did not result in an outage or derating. The

only significant event was an outage to replace

a boric acid injection tank at a PWR. The

borated water storage tank (BWST) is the primary water source for the HPI during the

HPI emergency operation mode. A single

header connects the borated water supply to

the charging pump. The pumps take suction

from the 350,000 gallon BWST (".2000 ppm

boron). Specifications for the BWST are

given in Table E4.

Thble E-5 gives the regulatory requirements and

guidelines for the design equipment qualification and

testing of the HPIS.

Table E-2. High pressure injection pump data

]Type Vertical, multistage, centrifugal

mechanical seal

Capacity, gal/min

Head, ft H2O (at sp. gr. = 1)

Motor horsepower, nameplate hp

Pump material

(See Figure E-l)

(See Figure E-l)

600

SS wetted parts

Design pressure, psig

Design temperature,'F

2,800/3,050

200/150

E-4
(D

0.

U1)

I-5

7000

6000

5000

4000

3000

2000

1000

_ 40

_ 35

- 30 ^

(n

25 cD

20 .

_ 15 D6

_z

600

7-815se

300

Capacity (gpm)

Figure E-1. HPI pump characteristics.

E-5
Table E-3. Active-HPIS and LPIS reactor coolant pressure boundary valves

Valve

Number System Service

Size

(in.)

System

Design

Rating

System

Condition

During

Operation

Motor

Operator

Type Type

Valve

Movement

HP-3 High pressure

injection

HP-4 High pressure

injection

HP-5 High pressure

injection

HP-26 High pressure

injection

HP-27 High pressure

injection

LP-17 Low pressure

LP-18 injection

Letdown

cooler

outlet

Letdown

cooler

outlet

Letdown

line RB

isolation

HP inj

RB

isolation

HP inj

RB

isolation

LP inj

RB

isolation

2-1/2

2-1/2

2-1/2

4

4

10

2500 psig 2170 psig

650xF I35xF

40-100 gpm

2500 psig 2170 psig

650xF I35xF

40-100 gpm

2500 psig 1270 psig

200xF 2950 psig

40-140 gpm

3050 psig 2200-

200xF 2950 psig

120-245xF

450 gpm

3050 psig 2200-

200xF 2950 psig

450 gpm

1 20-245xF

2500 psig 255 psig

300xF 28OxF

3000 gpm

Globe Limitorque Full open

SMB-00-15 to

Full close

Globe Limitorque Full open

SMB-00-15 to

Full close

Globe Sheffer Full open

Piston to

Full close

Globe Limitorque Full close

SMB-1-25 to

Full open

Globe Limitorque Full close

SMB-1-25 to

Full open

Gate Limitorque Full close

SMB-4-100 to

Full open

Table E-4. Borated water storage tank

Capacity, gal 388,000

Material Carbon steel/coated inside

Design pressure, psig Atmospheric

Design temperature,°F 150

Code AWWA D-100

E-6
Table E-5. HPIS regulatory requirements and guidelines

Criteria Title

Applicability

HPIS Remarks

Part I Design

ASME boiler and pressure vessel

Code, Section Ill-C, VIII, and XI

1. - Life extension under review

in ASME Section XI

2. General Design Criteria (GDC), Appendix A

to 10 CFR Part 50

b. GDC 2

c. GDC 4

d. GDC 13

e. GDC 19

j. GDC 24

-!a

Design basis for protection against

natural phenomena

Environmental and missile design

basis

Instrumentation and control

Control room

Separation of protection and control

systems

R

R

R

R

R

3. Regulatory Guides (RG)

c. RG 1.53

d. RG 1.62

e. RGI.75

f. RG 1.89

Application of the single-failure

criterion to nuclear power plant

protection systems

Manual initiation of protection actions

Physical independence of electric

systems

Environmental qualification of electrical

equipment for nuclear power plants

Gb

G

G

G Positions on end of life and

maintenance
Table E-5. (continued)

Applicability

Criteria Title HPIS Remarks

3. Regulatory Guides (RG) (continued)

g. RG 1.105 Instrument spands and setpoints G

h. RG 1.118 Periodic testing of electric power and

protection systems

G

Aging should be taken into

account

May require updating to take

into account studies on

increasing surveillance

intervals

6. IEEE Standards

a. 279-1971 Criteria for protection system for

nuclear power generating stations

60

R

b. 379-1977 Application of the single failure R

criteria to NPGS Class IE systems

Part 2 Electrical equipment qualification

1. 10 CFR 50.49 R All replacement equipment

purchased after 2/22/83

2. IEEE-323-1974 General guide for qualifying Class IE

electrical equipment for nuclear power

generating stations (1971)

R

3. IEEE 344-1975 Recommended practices for seismic R

qualification of Class IE electrical

equipment for nuclear power generating

stations

All replacement equipment

purchased after 2/22/83

Seismic qualification

Verify with naturally aged

components

4. Regulatory Guide 1.89 Environmental qualification of

electric equipment for nuclear power

plants

G
Table E-5. (continued)

Criteria Title

Applicability

HPIS Remarks

5. NUREG-0588 Interim staff position on

Environmental qualification of

safety-related electrical equipment

G Update using naturally aged

component data

Part 3 Testing requirements

1. Standard Technical Specifications

Section 3/4.5 and 4.5.11.1 M~

2. IEEE-Std 338-1977

3. Regulatory Guide 1.68

Criteria for periodic testing of nuclear

power generating station safety systems

Initial test programs for water-cooled

nuclear power plants

R

R

R

a. R = required.

b. C = guideline.
APPENDIX F

MAKEUP/HPI NOZZLE CRACKING

F-I
APPENDIX F

MAKEUPIHPI NOZZLE CRACKING

Makeup Nozzles

The 2.5-in. high pressure injection connections

with thermal sleeves are located on all four cold legs

of the reactor coolant piping. The four HPI/MU

nozzles (one per cold leg) are used to: (a) provide a

coolant source for emergency core cooling, and

(b) supply normal makeup (purification flow) to

the primary system. In general, one or two of the

nozzles are used for both HPI and MU, while the

remaining nozzles are used for HPI alone.

The incorporation of a thermal sleeve into a nozzle assembly is a common practice in the nuclear

industry to provide a thermal barrier between the

cold HPI/MU fluid and the hot high pressure

injection nozzle. This helps prevent thermal shock

and fatigue of the nozzle. The purpose of the safeend is to make the field weld easier (pipe to safeend) by allowing similar metals to be welded. The

dissimilar metal weld between the safe-end and the

nozzle can then be made under controlled conditions in the vendor's shop. The use of the safe-end

also eliminates the need to do any post-weld heat

treating in the field. Failures in these HPI/MU nozzles may preclude the proper functioning of the

ECCS and/or the normal fluid makeup of the primary system.

Make-Up Nozzle Cracking

Problem

Cracks were found in the normal makeup high

pressure injection (MU/HPI) nozzles of several

B&W plantsFl following an inspection of the eight

B&W plants licensed to operate. These cracks

appeared to be directly related to loose or missing

thermal sleeves. As a result, a B&W Owners'

Group Task Force was established to identify the

cause of the failures and recommend modifications

to eliminate future failures.

The B&W Task Force completed a generic investigation of the MU/HPI nozzle component cracking problem and has a report on the findings of that

investigation (Reference F-2). The report presents

relevant facts and probable failure scenarios, as

well as recommended modifications to thermal

sleeve designs, makeup system operating conditions, and ISI plans. Failure analysis indicated that

the cracks were initiated on the inside diameter and

were propagated by thermal fatigue. Inspections at

Midland have also shown that gaps may be present

between the thermal sleeve and safe-end in the contact expansion joint. These findings along with

stress analysis and testing have implicated insufficient contact expansion of the thermal sleeves as

the most probable root cause of the failures.

Possible Solutions

As a result of their investigation, B&W made the

following recommendationsFl for solving the

problem:

1. Reroll the upstream end of the thermal

sleeve when inspections indicate that a gap

exists, or repair and/or replace damaged

components

2. Maintain a continuous MU flow greater

than 1.5 gpm

3. Implement an augmented ISI plan

4. Perform a detailed stress analysis of a nozzle with a modified thermal sleeve design

to justify long term operation.

The NRC staff reviewed and evaluated the recommendations of the Task Force and supported all

four recommendations. The makeup HPI nozzle is

shown in Figure F-1 with both the old and new

design.

F-3
New thermal sleeve design

2.000" dia. ref. F r 1.500" dia.

Thermal,

sleeve

-I " j Safe end

| - , Contact

expanded

Old thermal sleeve design .010" to .015"

diametrical clearance

between thermal

sleeve & nozzle l.D. 7.81S5

Figure F-1. Makeup HPI nozzle (new and old design).

F-4
REFERENCES

F-i. "Make-Up Nozzle Cracking in Babcock and Wilcox (B&W) Plants," A Prioritization of Generic

Safety Issues, NUREG 0933, Issue 69, December 1983.

F-2. 177FuelAssembly Owners' Group Safe End Task Force Report on Generic Investigation of HPI/MU

Nozzle Component Cracking, B&W Document No. 77-1140611-00, Babcock and Wilcox

Company, 1983.

F-5
APPENDIX G

OPERATING EXPERIENCE DATA

G-1
APPENDIX G

OPERATING EXPERIENCE DATA

Nuclear Power Experience Data

for HPIS

Table G-2. I&C failures for HPI (NPE)

PWR Instrumentation

The NPE data includes the following ESF sys- (ho) (%0)

tems under the heading Systems Injections Recirculation and Containment Spray. Valves, pipes, and

many of the major components will be similar for

these systems. The NPE data base could not be

sorted on subsystems to eliminate containment

spray events. Because of the similarity of components that handle borated water, this should not significantly affect the conclusions.

PWRs:

safety injection (SI)

high head injection (HHI)

(a.k.a. upper head inj - UHI)

boron injection

recirculation phase

containment spray

accumulator tank system

Component

Control

Valves

Pumps

Electrical and

Miscellaneous

Instrumentation

Level

Pressure

Flow

Heat Tracing

Temperature

Other

Misc. and other

39

50

11

100

(17)

( 9)

(13)

(16)

(32)

(18)

( 1)

( 8)

(26)

NOTE: Numbers in parentheses are a percentage breakdown

for instrumentation channels for all instrumentation failures.

For all U.S. Nuclear Power Plants, from startup

through 1986, there were 1552 articles in NPE on

HPIS for PWRs. The ranking for frequency of

HPI valve failures by function is given in

Table G-1. The majority of valve failures (60%)

were for stop (on/off) valves. Table G-2 gives a

similar ranking for type of I&C channel failures.

About half are for instrumentation. A subbreakdown is shown in parentheses for various

types of measurements.

Based on data from NPE, the causes for pipe

failures are given in Table G-3. Design error, construction error, and maintenance error are the leading causes for pipe failures. Pipe support (including

snubbers) failure causes are broken down in

Table G-4. Maintenance error, design error, water

Table G-3. Causes for pipe failures (NPE)

Table G-1. Valve failure ranking by

function (NPE)

Cause Percent

PWR

Valve Function (%)

Stop (on/off) 60

Check 12

Control 8

Containment isolation 8

Relief 4

Other 8

No. of events 537

Design construction error

Weld failure

Maintenance error

Corrosion

Vibration

Blockage

Mechanical disability

Water hammer

Foreign material

Environmental effects

Other

23

15

14

7

5

4

4

3

3

2

20

1 00-

G-3
hammer, and mechanical disability are the leading

causes for pipe support failures.

In the NPE data base, the heading tanks include

I&C as well as valves associated with the tanks. Percentage of problems associated with tank subcomponents are shown in Table G-5 based on 194

articles.

LER Data on Valves

The LER data for valves for the HPI and make

up system is shown in Table G-6.G-1 Command

faults are those faults occurring in the control,

power, or other support system that prevents the

valve from performing its intended function. This

data covered years 1976 through 1980.

Table G-4. Causes for pipe support

failures (NPE)

Item

Maintenance error

Design construction error

Water hammer

Mechanical disability

Vibration

Wearout

Weld failure

Corrosion

Broken part/damage

Environment

Fatigue

Other

Percent

23

20

20

13

6

4

3

2

2

2

1

4

100

Table G-5. Subcomponents involved in

tank problems (NPE)

Item Percent

Water chemistry 34

I&C 25

Valve moving internals 17

Seals, gaskets, packing 3

Valve operators 2

Fittings, flanges 2

Other 17

100

Table G-6. Summary of valve failures and command faults for HPIS AND CVCS

(LERs)

PWR

Command

Failures Faults Total

System No. Percent No. Percent No. Percent

High pressure coolant

injection

Chemical volume

control (makeup)

48

92

5 26

10 11

8

3

74 6

103 8

G-4
REFERENCES

G-1. W. H. HubbleandC. MillerDataSummariesofLicenseeEventReportsof Valvesat U.S. Commercial

Nuclear Power Plants, EGG-EA-5125 NUREG/CR-1363, October 1982.

G-5
APPENDIX H

PROBLEMS WITH BORATED WATER SYSTEMS

H-I
APPENDIX H

PROBLEMS WITH BORATED WATER SYSTEMS

A number of cracking incidents were experienced

in safety-related stainless steel piping systems and

portions of systems containing oxygenated, stagnant, or essentially stagnant borated water during

the period from November 1974 to February 1977.

Metallurgical investigations revealed these cracks

occurred in the weld heat affected zone of type 304

stainless steel material. The cracks initiated on the

ID surface and propagated in a mode typical of

stress corrosion cracking. Analysis indicated the

probable corrosives to be chloride and oxygen contamination in the affected systems. Cracks were

also found as a result of local boric acid buildup

and confirmed by liquid penetrant tests. Conclusions were drawn that cracking incidents were due

to IGSCC originating in the pipe ID. The cracking

was localized to the heat affected zone where

type 304 stainless steel is sensitized to during welding. More information can be found in the IE

Information Notice 79-19 on this problem. No

problems were reported in 1980-82.

For the PWR plant used in the detailed HPI study,

the system operation procedures stated a concern that

following initial high pressure injection and HPI recirculation boron may plate out in the core and block

cooling channels. Therefore, at a predetermined time,

HPI flow is realigned from cold leg injection to hot leg

injection. This reverses flow direction through the core

and dissolves any crystallized boron.

On April 20, 1987, the NRC Information Notice

No. 86-109 alerted recipients of possible degradation

of the reactor coolant system pressure boundary resulting from boric acid corrosion. The boric acid corrosion

of ferritic steel components in the RCS was first

noticed at Arkansas One when inspecting HPI nozzles

and safe ends as a follow-up to the thermal stress cracking problem in 1982. Boric acid will rapidly corrode

ferritic (carbon) steel components if a small leak occurs

near a hot surface. The boric acid solution will boil and

concentrate, becoming more acidic and more corrosive. In addition, the evaporation of the water will

cause the boric acid crystals to accumulate at that

point. Other plants have experienced boric acid corrosion of ferritic steel due to leakage of borated water.

One involved threaded fasteners and is discussed in lE

Notice 82-02.

The boric acid corrosion is most active where the

metal surface is cool enough to remain wet. If the surface were hot enough to dry out, the loss of electrolyte

would slow the corrosion rate. Laboratory tests and

plant experience have shown corrosion rates of > 1-in.

depth per year in ferritic steel (Reference

Article 522 VII Safety System, A ECCS NPE).

Many PWRs have no positive means of detecting

boron dilution during cold shutdown. There have been

25 reported instances (Issue 22, NUREG-0933) of

inadvertent boron dilution during maintenance and

refueling. Although none have occurred, the possibility

of an inadvertent criticality is a paramount safety concern. Although undesirable, the consequences of an

unmitigated boron dilution event is not severe enough

to warrant backfit of additional protective features at

operating plants. Such an event at an operating plant

would represent a breakdown in a licensee's ability to

control its plant.

If there is no clear indication of boric acid leakage to

bolts, boric acid corrosion of bolts could go undetected

until failure of the bolts. Present ISI procedures do not

require mandatory visual inspection of bolts and UT

inspections are not required on pressure-retaining bolts

of less than 2 in. in diameter. This problem was listed

as safety issue 29, NUREG-0933 and given a high priority ranking for resolution.

The most recent alert to boric acid problems is from

NRC information notices 86-108 supplement I, "Degradation of Reactor Coolant System Pressure Boundary Resulting from Boric Acid Corrosion." In this

instance, about 500 lb of boric acid crystals were found

on the RV head as a result of a small leak in an instrument tube seal.

The loss of safety injection capability has occurred

at two plantsH-1 as a result of common mode failure

of SI pumps from crystallization of boric acid. In one

instance, leaky valves in the discharge line of the boron

injection tank (BIT) enabled concentrated boric acid to

flow through the low pressure discharge line (SI pump

suction) and to precipitate in pumps not heat traced. In

the other case, boric acid crystallization blockage

occurred between the boric acid storage tank (BAST)

and the charging pump. Solidified boric acid has also

blocked mixing tank pumpsH- 2 and level indicator

failures have occurred due to blocked sensing lines.

H-3
REFERENCES

H-I. Nuclear Power Experience Article, PWR 7A489.

H-2. Nuclear Power Experience Article, PWR7A497.

H-4
APPENDIX I

HPIS RISK ASSESSMENT DATA SUMMARIES

I-1
APPENDIX I

HPIS RISK ASSESSMENT DATA SUMMARIES

This appendix contains the baseline data from

the PRA'-I in Tables I-l, I-2, and 1-3. The description of each event is given in these tables.

Fault tree importance measures for the three

cases studied are given in Thbles I-4 through I-18.

These are direct printouts from the IRRAS Program for the 5-year, 10-year upper and lower

bounds, and 40-year upper and lower bounds.

The fault tree cut set quantification reports for the

three cases are given in Tibles 1-19 through 1-33 for the

same time periods as the importance measures. The

top of this display gives the family and fault tree names

along with the minimal cut set upper bound value as

calculated by the point estimate quantification. The

lower portion showed the quantified minimal cut sets.

The first column gave the ranking of the cut sets

according to the probability. The second column indicated an approximate percentage contribution of the

cut set to the minimal cut set upper bound. The third

column gave the frequency of each individual cut set

and the last column gave a listing of the basic events

making up each cut set.

The code for the fault tree heading on Tables 1-4

through 1-33 identifies the mode of operation that

applies for the table.

HPI is for the injection that requires one of three

pumps.

HP201 is for the injection mode that requires

two of three pumps.

HPRI is for the recirculation mode.

1-3
Table 1-1. Basic events for the high-pressure injectionlrecirculation system: hardware

Event

Name Descriptionab

Mean

UnavailabilityC

Mean

Failure

Rate

(yr-')

HPI0ICVO

HP102CVO

HP109CVO

Tilting-disk check valve

3HP-101 fails to open on

demand

Tilting-disk check valve

3HP-102 fails to open on

demand

Tilting-disk check valve

3HP-109 fails to open on

demand

8.7 -5 1.7-4

8.7-5 1.7 -4

8.7 -5 2.2-3

HP113CVO

HP152CVO

HP153CVO

HP 188CVO

LP55CVO

LP57CVO

HP 148VVT

Tilting-disk check valve

3HP-1 13 fails to open on

demand

Stop check valve 3HP-152

fails to open on demand

Stop check valve 3HP-153

fails to open on demand

Swing check valve 3HP-188

fails to open on demand

Swing check valve 3LP-55

fails to open on demand

Swing check valve 3LP-57

fails to open on demand

Manual valve 3HP-148

transfers closed

8.7 -5 2.2-3

9.8 -5 2.0 -4

9.8-5 2.0-4

9.8 -5 2.0-4

9.8 -5 2.0-4

9.8 -5 2.0-4

3.9-4 7.8-4

LP54VVT

LP56VVT

HP24MVO

HP25MVO

Manual valve 3LP-54

transfers closed

Manual valve 3LP-56

transfers closed

MOV 3HP-24 fails to open

on demand

MOV 3HP-25 fails to open

on demand

3.9-4 7.8-4

3.9-4 7.8-4

6.4-3 4.9-2

6.4-3 4.9-2

1-4
Table l-1. (continued)

Event

Name

HP26MVO

HP409MVO

HP410MVO

HPBPPS

HPCPPS

HPAPPR

HPRAPPR

HPBPPR

HPRBPPR

HPCPPR

HPRCPPR

HP15AVO

HP16AVO

HP12OUCF

HP120EPF

Descriptionab

MOV 3HP-26 fails to open

on demand

MOV 3HP-409 fails to open

on demand

MOV 3HP-410 fails to open

on demand

Pump HP-P3B fails to start

on demandd

Pump HP-P3C fails to start

on demand

Pump HP-P3A fails to rune

Pump HP-P3A fails to run

during recirculatione

Pump HP-P3B fails to run

Pump HP-P3B fails to run

during recirculatione

Pump HP-P3C fails to run

Pump HP-P3C fails to run

during recirculatione

AOV valve 3HP-15 fails to

open on demandf

Piston-operated valve 3HP-6

fails to open on demandd

AOV 3HP-120 control signal

fails lowg

AOV 3HP-120 E/P module

output fails lowh

Mean

Unavailabilityc

6.4-3

6.4 -3

6.4-3

8.4-4

8.4 -4

2.0-4

2.0-3

2.0 -4

2.0-3

2.0-4

2.0-3

1.6-3

3.5 -3

1.7 -4

4.6-5

Mean

Failure

Rate

(yr-')

4.9 -2

4.9-2

4.9 -2

2.1 -2

2.1 -2

7.4-2

7.4+1

7.4-2

7.4-1

7.4 -2

7.4-1

1.2-2

1.2-2

6.2-2

1.7-2

I-5
I

Table l-1. (continued)

Event

Name Descriptionab

Mean

Unavailabilityc

Mean

Failure

Rate

(yr')

HP120VPF

CS55CVO

CS86CVO

HPRCBPPS

AOV 3HP-120 pneumatic

positioner output fails lowh

Tilting-disk check valve

3CS-55 fails to open on

demand'

Tilting-disk check valve

3CS-36 fails to open on

demandf

RC bleed-transfer pump 3B

fails to start on

demand)

4.6-5 1.7 -2

8.7-5 6.7 -4

8.7-5 6.7 -4

8.4 -4 6.5 -3

HPM53RVF

HP31FTT

HP3 1UCT

HP3 1VPT

HP3 lEPT

CC24CVO

CC14CVO

CC20CVO

CCBPPS

CCBLSF

Pump HP-P3B fails to start

on low seal flowk

Flow transmitter controlling

AOV 3HP-31 output fails

highk

AOV 3HP-31 control signal to

E/P module fails highl

Valve positioner output for

AOV 3HP-31 fails highk

Output fails high from E/P

module for AOV-HP-31 k

Swing check valve 3CC-24

fails to open on demand

Tilting-disk check valve

3CC-14 fails to open on

demand

Swing check valve 3CC-20

fails to open on demand

Pump CC-P3B fails to start

on demandm

Pump CC-P3B fails to actuate

on low CC flowk

4.8 -4 3.7 -3

3.1 -5 1.1 -2

6.0-4 2.2-I

4.6-5 1.7 -2

4.6-5 1.7 -2

9.8 -5 2.0 -4

8.7 -5 1.7 -4

9.8 -5 2.0-4

8.4 -4 6.5 -3

2.4 -4 1.8 -3

I-6
Table 1-1. (continued)

Mean

Failure

Event Mean Rate

Name Descriptionasb UnavailabilityC (yr-t )

CCAPPR Pump CC-P3A fails to run 2.0 - 4 1.8 - 1

CCABPSF Limit switch for valve 2.3 - 4 8.2-2

3CC-7 or 3CC-8 opens,

causing interlock to failn

a. Events used in the RISK Assessment

b. Abbreviations and acronyms: MOV, motor-operated valve; AOV, air-operated valve; RC, reactor coolant.

c. Unavailability" as used here includes contributions from all failure modes of interest. For some events there is more than one

failure mode of interest, and the mean unavailability is the sum of the contributions from each failure mode, using an appropriate

mean failure rate and duration.

d. Failure rate multiplied by 10 due to pumping sump water.

e. Pump is run for normal makeup; I month is maximum.

f. Cycled during operation.

g. Controls continuously; level transmitter failure rate x 10.

h. Controls continuously (generic data).

i. Also cycled during operation.

j. Also operated during operation; used data for HPI pumps.

k. OPRA generic data.

1. OPRA generic data x 5 for five modules in series.

m. Used data for HPI pumps.

n. OPRA generic data x 2 for two limit switches.

1-7
Table 1-2. Basic events for the high-pressure injection/recirculation system: human errors

Event

Name

HP2425MVH

HP26MVCH

HP27MVH

HP409MVH

HP410MVH

HPSEGOH

Descriptiona Event lxrpeb

U

Mean

Unavailability C

MOVs 3HP-24 and 3HP-25 left 5.0-5

unavailable

MOV 3HP-26 left unavailable

MOV 3HP-27 left unavailable

MOV 3HP-409 left unavailable

MOV 3HP-410 left unavailable

Manual valves 3HP-33 or

3HP-100 left closed

inadvertently

U

U

U

U

U

1.5 -4

1.5-4

2.9 -4

2.9-4

2.2-4

HP418VVH

HPBCPPH

HP24MVH

HP25MVH

HP26MVH

HPCROSSH

HPBPPH

HPCPPH

HPLDSTH

HPRCPH

Manual valve 3HP-148 left

closed inadvertently

Pumps HP-P3B and HP-P3C

left unavailable

Operator fails to open MOV 3HP-24

Operator fails to open MOV 3HP-25

Operator fails to open MOV 3HP-26

Operator fails to open HPI

crossover valves, 3HP-409,

3HP-H10

Operator fails to start

pump HP-P3B

Operator fails to start

pump HP-P3C

Operator fails to initiate

letdown storage tank

Operator fails to trip

RCPs on loss of cooling

flow

U

U

OF(Recovery)

OF(Recovery)

OF(Recovery)

OF(Recovery)

OF(Recovery)

OF(Recovery)

OF

OF

6.1-4

1.5-4

1.0-2

1.0

1.0 -2

1.0

1.0-2

1.0

1.0-2

1.0-2

1.0

1.0-2

1.0

1.0-2

1.0-2

a. Abbreviations and acronyms: MOV, motor-operated valve; RCP, reactor coolant pump.

b. Definition of event types; U, unavailability error; OF, "operator fails to" error.

c. Mean unavailability consists of all human-error contributions for a particular event.

I-8
Table 1-3. Basic events for the high-pressure injection/recirculation system: maintenance

Event

Name

HP26MVM

HP409MVM

HP410MVM

HPSEGKM

Mean

Unavailability

2.7-4 -

2.7-4 A

2.7-4

- 1.5-3

Components in Maintenance Block

HPSEGMM 1.5-3

MOV 3HP-26

MOV 3HP-409

MOV 3HP-410

Pump HP-P3a; manual valves 3HP-106 and

3HP-103; check valve 3HP-105

Pump HP-P3B; manual valves 3HP-107 and

3HP-1 10; check valve 3HP-109

Pump HP-P3C; MOV 3HP-27; manual valves

3HP-111, 3HP-14 and 3HP-148; check valve

3HP-1 13

MOV 3HP-24; check valve 3HP-101

MOV 3HP-25; check valve 3HP-102

HPSEGHM 6.1-4

HPSEGPM

HPSEGQM

2.6-4

2.6-4

1-9
Table 1-4. Fault tree HPI importance measures report

(5 year)

Family: HPI

Fault Tree: HPI

(Sorted by Fussell-Vesely)

Event Name

HP2425MVH

HP24MVO

HP25MVO

HPSEGQM

HPSEGHM

HPSEGPM

HP101CVO

HP102CVO

HPSEGKM

HPSEGMM

HPBCPPH

HPCROSSH

HP26MVO

HPCPPS

HP148VVH

HP148VVT

HPCPPR

HP188CVO

HP27MVH

HP26MVM

HP113CVO

HP410MVO

HP26MVCH

HPSEGOH

HPBPPS

HPAPPR

HP409MVO

HPBPPR

HP109CVO

HP410MVM

HP409MVH

HP409MVM

HP152CVO

HP153CVO

Probability

of

Failure

5.OOOE -005

6.400E-003

6.400E-003

2.600E-004

6.1OOE-004

2.600E-004

8.700E-005

8.700E-005

1.500E-003

1.500E-003

1.500E -004

l.OOOE-002

6.400E-003

8.400E-004

6.100E -004

3.900E - 004

2.OOOE-004

9.800E-005

1.500E-004

2.700E-004

8.700E-005

6.400E - 003

1.500E-004

2.200E-004

8.400E-004

2.000E-004

6.400E-003

2.OOOE-004

8.700E-005

2.700E-004

2.900E-004

2.700E-004

9.800E-005

9.800E -005

FussellVesely

Importance

5.213E -001

4.503E -001

4.502E-001

3.523E -002

2.046E-002

1.757E -002

6.121E-003

6.120E -003

2.816E-003

2.815E-003

2.451E -003

2.212E -003

2.145E-003

6.308E -004

4.581E -004

2.929E -004

1.502E -004

1.162E -004

1.126E-004

7.322E -005

6.533E -005

5.414E-005

5.027E - 005

4.772E -005

2.990E -005

9.328E - 006

9.167E -006

8.384E-006

3.097E -006

2.284E -006

4.154E- 007

3.867E -007

1.137E-008

1.137E -008

Risk

Reduction

Ratio

2.089E + 000

1.819E + 000

1.819E + 000

1.037E + 000

1.021E + 000

1.018E+000

1.006E + 000

1.006E + 000

1.003E + 000

L.003E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.OO1E+000

l .OOOE + 000

I.OOOE + 000

.OOOE + 000

.OOOE + 000

1 .OOOE + 000

l.OOOE + 000

1.000E + 000

1.OOOE + 000

.OOOE + 000

I .OOOE + 000

1.000E + 000

1.OOOE + 000

1 .000E + 000

1.OOOE + 000

l.OOOE + 000

1.OOOE + 000

1.000E + 000

1.OOOE + 000

.OOOE + 000

1.OOOE + 000

Risk

Achievement

Ratio

1.043E + 004

7.088E + 001

7.087E+001

6.861E + 001

1.699E + 000

6.856E + 001

7.132E+001

7.131E+001

2.585E + 000

1.005E + 000

1.734E+001

1.219E + 000

1.333E +000

1.750E + 000

1.750E + 000

1.751E+000

1.751E+000

2.186E+000

1.751E+000

1.271E + 000

1.751E+000

1.008E + 000

1.335E + 000

1.217E+000

1.036E + 000

1.047E + 000

1.001E+000

I.042E + 000

1.036E + 000

1.008E + 000

1.001E+000

l.OOE + 000

1.OOOE + 000

1.OOOE + 000

1-10
Table 1-5. Fault tree HPI importance measures report

(10 year upper bound)

Family: HPI

Fault Tree: HPI

(Sorted by Fussell-Vesely)

Event Name

HP24MVO

HP25MVO

HP2425MVH

HPSEGQM

HPSEGHM

HPSEGPM

HPIOCVO

HP102CVO

HPCROSSH

HP26MVO

HPSEGKM

HPSEGMM

HP 148VVT

HPBCPPH

HPCPPR

HPCPPS

HP 148VVH

HPAPPR

HPBPPR

HP26MVM

HP188CVO

HP27MVH

HPSEGOH

HP1 13CVO

HP26MVCH

HPBPPS

HP41 OMVO

HP409MVO

HP109CVO

HP410MVM

HP409MVH

HP409MVM

HP153CVO

HP152CVO

Probability

of

Failure

7.000E-003

7.OOOE-003

5.OOOE-005

2.600E-004

6.100E-004

2.600E-004

9.100E-005

9.100E-005

I.OOOE-002

7.OOOE-003

1.500E-003

1.500E-003

3.500E-003

1.500E -004

1. lOOE-003

1.1 OOE - 003

6.1 OOE - 004

1. lOOE-003

1. lOOE-003

2.700E-004

1.OOOE-004

1.500E-004

2.200E-004

1.400E-004

1.500E-004

1.1 OOE -003

7.OOOE-003

7.000E-003

1.400E-004

2.700E-004

2.900E-004

2.700E -004

1.OOOE-004

I.OOOE-004

FussellVesely

Importance

4.91 1E -001

4.909E -001

4.771E-001

3.535E -002

2.066E-002

1.758E-002

6.383E -003

6.382E -003

5.265E -003

5.058E -003

2.905E -003

2.902E -003

2.758E -003

2.258E -003

8.667E - 004

8.667E -004

4.806E-004

1.936E -004

1.887E-004

1.792E-004

1.224E -004

1.182E -004

1.145E -004

1.1 03E -004

1.084E -004

1.038E-004

7.704E-005

2.649E-005

1.322E -005

2.972E -006

1.097E -006

1.022E -006

1.221E -008

1.221E -008

Risk

Reduction

Ratio

1.965E + 000

1.964E + 000

1.912E + 000

1.037E + 000

1.021 E + 000

1.018E +000

1.006E + 000

1.006E + 000

1.005E + 000

1.005E + 000

L.003E + 000

1.003E + 000

1.003E + 000

1.002E + 000

l.001E+000

l.OOlE+ 000

1.000E + 000

1.OOOE + 000

1.000E + 000

1.000E + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1. OOOE + 000

1.OOOE + 000

1.OOOE + 000

.OOOE + 000

1.OOOE + 000

1.OE + 000

1.000E + 000

1.OOOE + 000

L.OOOE + 000

1.000E + 000

1.OOOE + 000

Risk

Achievement

Ratio

7.064E + 001

7.062E + 001

9.542E + 003

6.863E + 001

1.710E+000

6.858E + 001

7.112E + 001

7.11OE+001

1.521E + 000

1.717E+000

2.574E + 000

1.071E +000

1.785E + 000

1.605E + 001

1.787E + 000

1.787E + 000

1.787E + 000

1.176E + 000

1.171E+000

1.663E + 000

2.224E + 000

1.788E + 000

1.520E + 000

1.788E + 000

1.722E + 000

1.094E + 000

1.011E+000

1.004E + 000

1.094E + 000

1.O11E+000

1.004E + 000

1.004E + 000

1.OOOE + 000

L.OOOE + 000

I-ll
Table 1-6. Fault tree HPI importance measures report

(10 year lower bound)

Family: HPI

Fault RTee: HPI

(Sorted by Fussell-Vesely)

Event Name

HP2425MVH

HP24MVO

HP25MVO

HPSEGQM

HPSEGHM

HPSEGPM

HPIOICVO

HP102CVO

HPCROSSH

HP26MVO

HPSEGKM

HPSEGMM

HPBCPPH

HP148VVT

HPCPPS

HP148VVH

HPCPPR

HP188CVO

HP27MVH

HP26MVM

HP113CVO

HP26MVCH

HPSEGOH

HP410MVO

HPBPPS

HPAPPR

HPBPPR

HP409MVO

HP109CVO

HP410MVM

HP409MVH

HP409MVM

HP152CVO

HP153CVO

Probability

of

Failure

5.000E-005

6.500E-003

6.500E-003

2.600E-004

6.100E-004

2.600E-004

8.800E-005

8.800E-005

1.OOOE-002

6.500E-003

1.500E-003

1.500E-003

1.500E-004

1.300E-003

9.100E-004

6.1OOE-004

4.600E -004

9.900E-005

1.500E-004

2.700E-004

9.500E-005

1.500E-004

2.200E-004

6.500E- 003

9.100E-004

4.600E-004

4.600E -004

6.500E-003

9.500E -005

2.700E-004

2.900E -004

2.700E-004

9.900E -005

9.900E-005

FussellVesely

Importance

5.135E-001

4.572E -001

4.571E -001

3.527E -002

2.052E -002

1.758E-002

6.189E -003

6.188E -003

3.098E-003

2.984E -003

2.840E-003

2.839E -003

2.417E-003

9.896E -004

6.927E - 004

4.643E - 004

3.502E -004

1.181E - 004

1.142E -004

1.069E-004

7.231E -005

6.887E -005

6.705E - 005

5.962E -005

4.937E -005

3.641E -005

3.426E -005

1.335E -005

5.154E -006

2.477E-006

5.958E -007

5.547E -007

1.167E -008

1.167E - 008

Risk

Reduction

Ratio

2.055E + 000

1.842E + 000

1.842E + 000

1.037E + 000

1.021E + 000

1.018E+000

1.006E + 000

1.006E + 000

1.003E + 000

1.003E + 000

1.003E + 000

1.003E + 000

1.002E + 000

l .001E + 000

1.001E +000

.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1.000E + 000

1. OOOE + 000

1.OOOE + 000

1.000E + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

l.OOOE + 000

1.OOOE + 000

1.000E + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

L.OOOE + 000

Risk

Achievement

Ratio

1.027E + 004

7.085E + 001

7.084E+ 001

6.863E + 001

1.701E+000

6.858E + 001

7.130E+001

7.129E+ 001

1.307E+ 000

1.456E + 000

2.588E + 000

1.017E+000

1.711E +001

1.760E + 000

1.761E+000

1.761E+000

1.761E+ 000

2.192E + 000

1.761E+000

1.396E + 000

1.761E+000

1.459E + 000

1.305E + 000

1.009E + 000

1.054E + 000

1.079E + 000

1.074E + 000

1.002E + 000

1.054E + 000

1.009E + 000

1.002E + 000

1.002E + 000

1 .OOOE + 000

.OOOE + 000

1-12
Table 1-7. Fault tree HPI importance measures report

(40 year upper bound)

Family: HPI

Fault Tree: HPI

(Sorted by Fussell-Vesely)

Event Name

HP24MVO

HP25MVO

HP2425MVH

HPSEGQM

HP148VVT

HPSEGHM

HPCROSSH

HP26MVO

HPSEGPM

HPAPPR

HPBPPR

HP1OICVO

HP102CVO

HPCPPR

HPSEGKM

HPSEGMM

HPCPPS

HPBCPPH

HPBPPS

HP148VVH

HP26MVM

HP113CVO

HPSEGOH

HP26MVCH

HP410MVO

HP409MVO

HP188CVO

HP27MVH

HP109CVO

HP410MVM

HP409MVH

HP409MVM

HP152CVO

HP153CVO

Probability

of

Failure

1.OOOE -002

I.OOOE-002

5.000E-005

2.600E-004

2.200E-002

6.100E -004

l.OOOE -002

1.OOOE -002

2.600E-004

6.800E-003

6.800E-003

1.200E - 004

1.200E -004

6.800E-003

1.500E -003

1.500E-003

2.700E-003

1.500E-004

2.700E-003

6.100E -004

2.700E - 004

4.500E-004

2.200E-004

1.500E-004

1.OOOE-002

1.OOOE-002

1.300E-004

1.500E-004

4.500E-004

2.700E -004

2.900E-004

2.700E-004

1.300E-004

1.300E-004

FussellVesely

Importance

6.338E -001

6.333E -001

3.050E -001

3.265E -002

2.402E -002

2.354E-002

2.135E -002

2.080E -002

1.605E -002

1.150E-002

1.147E - 002

7.604E -003

7.599E-003

7.424E -003

6.959E -003

6.943E -003

2.948E -003

1.529E-003

8.121E -004

6.660E-004

5.510E-004

4.913E -004

4.685E - 004

3.120E -004

3.072E -004

2.190E-004

1.687E -004

1.638E-004

1.353E -004

8.295E - 006

6.351E -006

5.913E -006

2.176E -008

2.176E-008

Risk

Reduction

Ratio

2.730E + 000

2.727E + 000

1.439E + 000

1.034E + 000

1.025E + 000

1.024E + 000

1.022E + 000

1.021E+ 000

1.016E+000

1.012E+000

1.012E+ 000

1.008E + 000

1.008E + 000

1.007E + 000

1.007E + 000

1.007E + 000

1.003E + 000

1.002E + 000

1.Oil + 000

1.001l+000

l.OOE + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

l.OOOE + 000

.OOOE + 000

1.OOOE + 000

1.OOOE + 000

1.OOOE + 000

I .OOOE + 000

1.OOOE + 000

1.OOOE + 000

Risk

Achievement

Ratio

6.372E + 001

6.367E + 001

6.102E + 003

6.272E + 001

2.068E + 000

1.918E + 000

3.113E + 000

3.059E + 000

6.270E + 001

2.679E + 000

2.676E + 000

6.434E + 001

6.429E + 001

2.084E + 000

3.900E + 000

2.358E + 000

2.089E + 000

1.119E+001

1 .300E + 000

2.09 1E + 000

3.040E + 000

2.09 1E + 000

3.129E + 000

3.080E + 000

1.030E + 000

1.022E + 000

2.297E + 000

2.092E + 000

1.301E+ 000

1.031E+O000

1.022E + 000

1.022E + 000

1.OOOE + 000

1.OOOE + 000

I-13
Table I-S. Fault tree HPI importance measures report

(40 year lower bound)

Family: HPI

Fault Tree: HPI

(Sorted by Fussell-Vesely)

Event Name

HP24MVO

HP25MVO

HP2425MVH

HPSEGQM

HPSEGHM

HPSEGPM

HPCROSSH

HP26MVO

HPlOICVO

HP102CVO

HP148VVT

HPSEGKM

HPSEGMM

HPBCPPH

HPCPPR

HPCPPS

HPAPPR

HPBPPR

HP148VVH

HP26MVM

HPBPPS

HPSEGOH

HP26MVCH

HP27MVH

HP188CVO

HP113CVO

HP410MVO

HP409MVO

HP109CVO

HP410MVM

HP409MVH

HP409MVM

HP152CVO

HP153CVO

Probability

of

Failure

7.100E-003

7.1003E-003

5.000E-005

2.600E1-004

6.100E-004

2.600E -004

1.OOOE-002

7.100E-003

9.100E-005

9.1003E-005

6.800E1-003

1.500E1-003

1.500E1-003

1.500E - 004

2.000E-003

1.400E1-003

2.000E-003

2.0003E-003

6. 100E - 004

2.700E1-004

1.400E-003

2.200E1-004

1.500E -004

1.500E1-004

I.OOOE-004

1.400E-004

7.100E-003

7.100E-003

1.400E -004

2.700E1-004

2.900E1-004

2.700E-004

I.OOOE-004

1.OOOE-004

FussellVesely

Importance

4.958E -001

4.956E1-001

4.684E -001

3.5311E-002

2.121E1-002

1.751E-002

8.4103E-003

8.048E1-003

6.354E -003

6.352E-003

5.705E1-003

3.422E -003

3.417E -003

2.226E -003

1.678E1-003

1.175E-003

7.648E1-004

7.558E1-004

5.118E-004

2.9031E -004

2.184E1- 004

1.831E-004

1.700E1-004

1.258E-004

1.226E -004

1.175E-004

9.537E -005

4.373E -005

2.184E1-005

3.627E1-006

1.786E -006

1.663E -006

1.223E1-008

1.223E -008

Risk

Reduction

Ratio

1.983E+000

1.982E + 000

1.881E +000

1.037E + 000

1.022E + 000

1.018E+000

1.008E + 000

1.008E + 000

1.006E + 000

1.006E + 000

1.006E + 000

1.003E + 000

1.003E + 000

1.002E + 000

1.002E + 000

1.001E+000

1.001E+000

1.001E + 000

1.001E± +000

.OOOE + 000

1.OOOE + 000

.OOOE + 000

1.OOOE + 000

.OOOE + 000

1.OOOE + 000

.OOOE + 000

.OOOE + 000

1.OOOE + 000

1.0OOOE + 000

.OOOE + 000

.OOOE + 000

1.OOOE + 000

I.OOOE + 000

.OOOE + 000

Risk

Achievement

Ratio

7.031E+001

7.028E + 001

9.369E + 003

6.834E + 001

1.733E1+000

6.8301+001

1.833E+±000

2.125E+3000

7.079E + 001

7.077E + 001

1.833E1+000

2.770E + 000

1.210E + 000

1.584E + 001

1.837E + 000

1.838E+000

1.382E+±000

1.377E + 000

1.838E+000

2.075E + 000

1.156E+±000

1.832E1+000

2.133E+000

1.839E1+000

2.226E + 000

1.839E+±000

1.013E + 000

1.006E + 000

1.156E+±000

1.013E + 000

1.006E + 000

1.006E + 000

.OOOE + 000

1.0OOOE + 000

1-14
Table 1-9. Fault tree HP201 importance measures report

(5 year)

Family: HPI

Fault Tree: HP201

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HP2425MVH

HP24MVO

HP25MVO

HPSEGHM

HPSEGQM

HPSEGKM

HPSEGMM

HP26MVM

HPSEGPM

HP26MVCH

HPSEGOH

HPCPPS

HPBPPS

HP 148VVH

HPBPPR

HP148VVT

HPlOlCVO

HP102CVO

HPBCPPH

HPAPPR

HPCPPR

HP409MVO

HP410MVO

HP27MVH

HP113CVO

HP109CVO

HP409MVH

HP409MVM

HP410MVM

HP188CVO

HP153CVO

HP152CVO

Probability

of

Failure

1.OOOE-002

6.400E-003

5.000E-005

6.400E-003

6.400E-003

6.1 OOE - 004

2.600E-004

1.500E-003

1.500E-003

2.700E-004

2.600E-004

1.500E-004

2.200E -004

8.400E-004

8.400E -004

6. 100E -004

2.000E-004

3.900E -004

8.700E -005

8.700E-005

1.500E-004

2.000E-004

2.000E-004

6.400E -003

6.400E -003

1.500E-004

8.700E -005

8.700E -005

2.900E-004

2.700E-004

2.700E -004

9.800E -005

9.800E-005

9.800E-005

FussellVesely

Importance

3.967E -001

3.738E -001

2.906E-001

2.592E-001

2.510E -001

5.964E-002

3.568E -002

3.509E -002

3.376E -002

1.575E -002

1.013E -002

8.760E-003

8.642E-003

8.432E -003

7.326E -003

6.123E -003

5.505E -003

3.915E -003

3.523E -003

3.412E-003

2.790E -003

2.036E -003

2.008E-003

1.854E-003

1.798E -003

1.506E -003

8.733E -004

7.588E -004

8.401E- 005

7.821E -005

7.584E -005

6.516E -005

6.363E -009

6.363E -009

Risk

Reduction

Ratio

1.657E + 000

1.597E + 000

1.410E + 000

1.350E + 000

1.335E + 000

1.063E + 000

1.037E + 000

1.036E + 000

1.035E + 000

1.016E+ 000

1 .010E + 000

1.009E + 000

1.009E + 000

1.009E + 000

1.007E + 000

1.006E + 000

1.006E + 000

1.004E + 000

1.004E + 000

1.003E + 000

1.003E + 000

1.002E + 000

L.002E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.00E+ 000

1.001E + 000

.OOOE + 000

1 .OOOE + 000

1 .OOOE + 000

I.OOOE + 000

.OOOE + 000

.OOOE + 000

Risk

Achievement

Ratio

4.025E + 001

5.902E + 001

5.813E + 003

4.122E + 001

3.995E + 001

2.178E+000

3.869E + 001

8.642E + 000

1.622E+001

5.933E +001

3.993E + 001

5.939E + 001

4.026E + 001

1.103E+001

9.714E + 000

1.103E+001

2.847E + 001

1.103E + 001

4.147E +001

4.020E + 001

1.958E + 001

1.118E+001

1.103E +001

1.288E + 000

1.279E + 000

1.103E+001

1. 104E +001

9.721E + 000

1.290E + 000

1.290E + 000

1.281E+000

1.665E + 000

I .OOOE + 000

1 .OOOE + 000

1-15
Table 1-10. Fault tree HP201 importance measures report

(10 year upper bound)

Family: HPI

Fault Tree: HP201

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HP24MVO

HP25MVO

HP2425MVH

HPSEGHM

HPSEGKM

HPSEGMM

HPBPPR

HP148VVT

HPSEGQM

HPAPPR

HPCPPR

HPCPPS

HP26MVM

HPSEGPM

HPBPPS

HPSEGOH

HP148VVH

HP26MVCH

HPIOICVO

HP102CVO

HPBCPPH

HP409MVO

HP410MVO

HP27MVH

HP113CVO

HP IO9CVO

HP409MVH

HP409MVM

HP188CVO

HP410MVM

HP153CVO

HP152CVO

Probability

of

Failure

l.000OE-002

7.OOOE-003

7.000E-003

7.000E-003

5.000E-005

6.100E-004

1.500E-003

1.500E-003

1.I OOE - 003

3.500E -003

2.600E-004

1.I OOE - 003

1. IOOE - 003

1. IOOE - 003

2.700E-004

2.600E-004

1. IOOE - 003

2.200E-004

6.100E-004

1.500E-004

9.100E-005

9.100E-005

1.500E-004

7.000E-003

7.OOOE-003

1.500E-004

1.400E -004

1.400E-004

2.900E-004

2.700E-004

l.OOOE - 004

2.700E-004

L.OOOE-004

I.OOOE-004

FussellVesely

Importance

3.622E -001

3.428E-001

2.581E - 001

2.506E-001

2.435E -001

1.009E-001

7.965E -002

7.848E -002

5.335E -002

4.507E -002

3.144E-002

1.462E -002

1.417E -002

1.417E -002

1.321E - 002

9.249E -003

8.080E -003

7.896E -003

7.855E -003

7.345E -003

3.355E -003

3.258E -003

3.007E-003

2.469E -003

1.951E- 003

1.932E -003

1.803E-003

1.028E -003

1.023E -004

9.524E -005

7.990E -005

7.524E-005

7.908E -009

7.908E - 009

Risk

Reduction

Ratio

1.568E + 000

1.522E+000

1.348E + 000

1.334E + 000

1.322E + 000

1.11 2E + 000

1.087E + 000

1.085E + 000

1.056E + 000

1.047E + 000

1.032E + 000

1.015E+000

1.014E + 000

1.014E + 000

1.013E+000

1.009E + 000

1.008E +000

1.008E + 000

1.008E + 000

1.007E + 000

1.003E + 000

1.003E + 000

1.003E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.001E + 000

.OOOE + 000

1.000E + 000

I.OOOE + 000

.OOOE + 000

1.OOOE + 000

.OOOE + 000

Risk

Achievement

Ratio

3.685E + 001

4.963E + 001

3.759E + 001

3.654E+ 001

4.871 E + 003

6.431E+000

1.369E + 001

3.900E + 001

4.925E + 001

1.382E + 001

3.552E + 001

1.427E + 001

1.385E + 001

1.385E + 001

4.991E + 001

3.655E + 001

8.337E + 000

3.687E + 001

1.386E+001

4.996E + 001

3.785E + 001

3.678E+001

2.101E +001

1.350E+000

1.277E + 000

1.387E + 001

1.387E + 001

8.344E + 000

1.353E + 000

1.353E+ 000

1.799E + 000

1.279E + 000

I .OOOE + 000

1 .OOOE + 000

1-16
Table 1-11. Fault tree HP201 importance measures report

(10 year lower bound)

Family: HPI

Fault Tree: HP201

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HP2425MVH

HP24MVO

HP25MVO

HPSEGHM

HPSEGKM

HPSEGMM

HPSEGQM

HPBPPR

HP26MVM

HP148VVT

HPCPPS

HPSEGPM

HPSEGOH

HP26MVCH

HPBPPS

HP148VVH

HPAPPR

HPCPPR

HPlOICVO

HP102CVO

HPBCPPH

HP409MVO

HP410MVO

HP27MVH

HP113CVO

HP109CVO

HP409MVH

HP409MVM

HP410MVM

HP188CVO

HP153CVO

HP152CVO

Probability

of

Failure

l.OOOE-002

6.500E - 003

5.OOOE-005

6.500E-003

6.500E-003

6.100E-004

1.500E-003

1.500E-003

2.600E-004

4.600E-004

2.700E-004

1.300E-003

9.1OOE-004

2.600E-004

2.200E-004

I .500E -004

9.100E-004

6.100E-004

4.600E-004

4.600E -004

8.800E-005

8.800E-005

1.500E-004

6.500E-003

6.500E-003

1.500E-004

9.500E-005

9.500E-005

2.900E-004

2.700E-004

2.700E-004

9.900E-005

9.900E-005

9.900E-005

FussellVesely

Importance

3.874E -001

3.653E-001

2.796E -001

2.569E-001

2.489E -001

7.407E -002

5.031E- 002

4.902E-002

3.463E -002

1.606E-002

1.516E-002

1.448E-002

1.014E-002

9.890E - 003

8.442E-003

8.429E -003

7.648E-003

6.795E -003

5.231E - 003

5.124E -003

3.478E -003

3.370E -003

2.906E-003

1.981E - 003

1.810OE-003

1.671E -003

1.058E -003

7.984E - 004

8.836E -005

8.227E -005

7.517E-005

7.026E -005

6.912E -009

6.912E-009

Risk

Reduction

Ratio

1.632E + 000

1.576E + 000

1.388E+000

1.346E + 000

1.331E+000

1.080E + 000

1.053E + 000

1.052E + 000

1.036E + 000

1.016E+000

1.015E+000

1.015E+000

1.OIOE+000

1.O1OE+000

1.009E + 000

1.009E + 000

L.008E + 000

1.007E + 000

1.005E + 000

1.005E + 000

1.003E + 000

1.003E + 000

1.003E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.001E+ 000

1.001E+ 000

1.OOOE + 000

I .OOOE + 000

I.OOOE + 000

.OOOE + 000

1 .OOOE + 000

1.OOOE + 000

Risk

Achievement

Ratio

3.934E + 001

5.683E + 001

5.593E + 003

4.025E + 001

3.904E + 001

3.607E + 000

1.036E + 001

2.402E + 001

3.783E+ 001

3.581E + 001

5.713E + 001

1.212E + 001

1.212E + 001

3.902E + 001

3.935E + 001

5.718E + 001

9.397E + 000

1.213E+ 001

1.236E + 001

1.213E + 001

4.050E + 001

3.928E + 001

2.035E + 001

1.303E + 000

1.277E + 000

1.213E + 001

1.213E + 001

9.403E + 000

1.305E+000

1.305E + 000

1.278E + 000

1.710E+000

.OOOE + 000

I .OOOE + 000

1-17
Table 1-12. Fault tree HP201 importance measures report

(40 year upper bound)

Family: HPI

Fault Tree: HP201

(Sorted by Fussell-Vesely)

Event Name

HPBPPR

HP 148VVT

HPCROSSH

HP24MVO

HP25MVO

HP26MVO

HPSEGHM

HPSEGKM

HPSEGMM

HPAPPR

HPCPPR

HP2425MVH

HPCPPS

HPSEGQM

HP148VVH

HPBPPS

HP409MVO

HP113CVO

HP26MVM

HPSEGPM

HPSEGOH

HP26MVCH

HPBCPPH

HPIOICVO

HP27MVH

HP102CVO

HP410MVO

HP109CVO

HP409MVH

HP409MVM

HP188CVO

HP410MVM

HP152CVO

HP153CVO

Probability

of

Failure

6.800E - 003

2.200E -002

1.OOOE3- 002

1.OOE - 002

1.OOOE3- 002

I. OOOE -002

6. 1 OE -004

1.5SOOE- 003

I. SOOE -003

6.800E - 003

6.800E -003

5.000OE- 005

2.700E -003

2.600E -004

6. 100E -004

2.700E -003

1.00E - 002

4.50E - 004

2.700E -004

2.600E -004

2.200E -004

1.50E - 004

1 .500E - 004

1.200E -004

1.50E - 004

1.200E - 004

I.000OE -002

4.50E - 004

2.900E -004

2.700E -004

1.300E -004

2.700E -004

1.300 - 004

1.300E -004

FussellVesely

Importance

4.584E - 001

3.008E-001

1.726E-001

1.710E -001

1.67SE-001

1.632E - 001

1.370E - 001

1.289E - 001

1.283E - 001

9.870E -002

9.296E-002

8.068E-002

3.691E -002

1.298E -002

8.339E -003

6.779E -003

6.224E -003

6.152E -003

4.401E - 003

4.334E-003

3.716E-003

2.447E-003

2.405E-003

2.052E -003

2.051E -003

2.010E -003

1.861E-003

1.130E-003

1.805E -004

1.680E -004

1.1 19E -004

5.024E-005

1.428E-008

1.428E- 008

Risk

Reduction

Ratio

1.846E + 000

1.430E + 000

1.209E +000

1.206E + 000

1.201E+ 000

1.195E + 000

1.159E + 000

1.148E+000

1.147E + 000

1.llOE+000

1.102E + 000

1.088E+ 000

1.038E + 000

1.013E+000

1.008E + 000

1.007E + 000

1.006E + 000

1.006E + 000

1.004E + 000

1.004E + 000

1.004E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.001E + 000

1.000E + 000

1.000E + 000

1.OE + 000

1.OOOE + 000

1.000E + 000

1.OOOE + 000

Risk

Achievement

Ratio

6.703E + 001

1.435E +-001

1.808E-+ 001

1.792E + 001

1.758E1+001

1.715E+001

1.216E + 001

1.848E1+001

6.398E + 001

1.539E+ 001

1.456E+001

1.614E + 003

1.462E + 001

1.732E± 001

1.464E + 001

3.504E + 000

1.616E+ 000

1.465E+ 001

1.729E + 001

1.766E + 001

1.788E+001

1.731E+001

1.699E + 001

1.809E + 001

1.465E+001

1.774E+001

1.184E+000

3.509E + 000

1.622E + 000

1.622E + 000

1.860E+ 000

1.186E+000

1 .OOOE + 000

1.OOOE + 000

I-18
Table 1-13. Fault tree HP201 importance measures report

(40 year lower bound)

Family: HPI

Fault Tree: HP201

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HP24MVO

HP25MVO

HP2425MVH

HPSEGHM

HPBPPR

HPSEGKM

HPSEGMM

HP148VVT

HPAPPR

HPCPPR

HPSEGQM

HPCPPS

HP26MVM

HP148VVH

HPBPPS

HPSEGPM

HPSEGOH

HP26MVCH

HPBCPPH

HP101CVO

HP409MVO

HP102CVO

HP27MVH

HP113CVO

HP410MVO

HP lO9CVO

HP409MVH

HP409MVM

HP188CVO

HP410MVM

HP153CVO

HP152CVO

Probability

of

Failure

1.OOOE - 002

7.100E-003

7.100E-003

7.100E-003

5.000E -005

6.100E-004

2.000E-003

1.500E-003

1.500E-003

6.800E-003

2.000E-003

2.OOOE-003

2.600E-004

1.400E-003

2.700E-004

6.100E-004

1.400E-003

2.600E-004

2.200E-004

1.500E-004

1.500E-004

9.100E-005

7.100E -003

9.100E-005

1.S00E-004

1.400E-004

7.100E-003

1.400E-004

2.900E-004

2.700E-004

1.000E-004

2.700E-004

1.OOOE-004

I.000E-004

FussellVesely

Importance

3.196E -001

3.018E-001

2.302E-001

2.236E -001

2.113E-001

1.323E -001

1.298E -001

1.138E-001

1.127E -001

1.024E-001

3.137E -002

3.010E -002

2.751E -002

2.107E -002

1.147E-002

9.182E -003

8.974E -003

8.137E-003

6.953E-003

6.377E -003

3.191E -003

2.951E -003

2.895E-003

2.866E -003

2.258E -003

2.107E -003

1.771E- 003

8.974E -004

1.182E -004

1.I01E -004

8.655E -005

6.733E -005

8.529E -009

8.529E -009

Risk

Reduction

Ratio

1.470E + 000

1.432E + 000

1.299E + 000

1.288E + 000

1.268E + 000

1.153E+ 000

1.149E+000

1.128E+000

1.127E+000

1.1 14E+ 000

1.032E + 000

1.031E+000

1.028E + 000

1.022E + 000

1.012E+000

1.009E + 000

1.009E + 000

1.008E + 000

L.007E + 000

1.006E + 000

L.003E + 000

L.003E + 000

1.003E + 000

1.003E + 000

1.002E + 000

1.002E + 000

1.002E + 000

1.001E+ 000

I.000E + 000

1.OOOE + 000

1.OOOE + 000

1.000E + 000

1.OOOE + 000

1.000E + 000

Risk

Achievement

Ratio

3.263E + 001

4.321E + 001

3.318E+001

3.226E + 001

4.228E + 003

9.572E + 000

6.542E + 001

1.744E + 001

5.657E+ 001

1.594E+001

1.664E + 001

1.601E+001

3.139E+001

1.602E + 001

4.346E + 001

1.603E + 001

7.401E + 000

3.228E + 001

3.259E + 001

4.350E + 001

2.223E + 001

3.340E+001

1.405E + 000

3.248E + 001

1.604E + 001

1.604E+ 001

1.248E + 000

7.409E + 000

1.408E + 000

1.408E + 000

1.865E1+000

1.249E + 000

1.OOOE + 000

1.OOOE + 000

1-19
Table 1-14. Fault tree HPRI importance measures report

(5 year)

Family: HPI

Fault 'ftee: HPRI

(Sorted by Fussell-Vesely)

Event Name

HPSEGMM

HPSEGKM

HPSEGHM

HP26MVO

HPCROSSH

HPBCPPH

HPRAPPR

HP2425MVH

HP25MVO

HP24MVO

HPRCPPR

HPCPPS

HP148VVH

HP148VVT

HPCPPR

Probability

of

Failure

1.500E-003

1.500E-003

6.100E-004

6.400E-003

1.000E-002

1.500E-004

2.000E-003

5.OOOE-005

6.400E-003

6.400E-003

2.OOOE -003

8.400E-004

6.IOOE-004

3.900E-004

2.OOOE-004

FussellVesely

Importance

4.757E - 001

4.757E -001

4.757E -001

3.541E-001

3.541E-001

2.673E - 001

2.166E - 001

2.081E-001

1.705E -001

1.705E -001

I .524E -001

6.399E-002

4.647E -002

2.971E-002

1.524E -002

Risk

Reduction

Ratio

1.907E + 000

1.907E + 000

1.907E + 000

1.548E + 000

1.548E+000

1.365E + 000

1.276E+000

1.263E + 000

1.206E + 000

1.206E + 000

1.180E + 000

1.068E + 000

1.049E + 000

1.031E+000

1.015E+000

Risk

Achievement

Ratio

1l.OOIE + 000

2.868E + 002

7.691E + 001

5.597E + 001

3.606E + 001

1.783E + 003

1.091E + 002

4.160E + 003

2.747E + 001

2.747E + 001

7.703E + 001

7.712E + 001

7.714E + 001

7.715E + 001

7.717E + 001

1-20
Table 1-15. Fault tree HPRI importance measures report

(10 year upper bound)

Family: HPI

Fault Tree: HPRI

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HPRAPPR

HPRCPPR

HP2425MVH

HP25MVO

HP24MVO

HPSEGKM

HPSEGMM

HPSEGHM

HP148VVT

HPBCPPH

HPCPPS

HPCPPR

HP148VVH

Probability

of

Failure

1.OOOE-002

7.OOOE-003

1.I OOE - 002

1. lOOE - 002

5.OOOE-005

7.OOOE-003

7.000E-003

1.500E-003

1.500E-003

6.100E-004

3.500E-003

1.500E -004

1.I OOE -003

1.100E-003

*6.100E - 004

FussellVesely

Importance

4.618E -001

4.618E -001

,4.010E-001

2.835E - 001

2.301E -001

2.255E -001

2.255E-001

1.529E-001

1.529E - 001

1.529E-001

9.021E-002

8.267E -002

2.835E -002

2.835E-002

1.572E-002

Risk

Reduction

Ratio

1.858E + 000

1.858E + 000

1.669E + 000

1 .396E + 000

1.299E + 000

1.291E + 000

1.291E+000

1.181E+000

1.181E+000

1.181E + 000

1.099E + 000

1.090E + 000

1.029E + 000

1.029E + 000

1.016E+000

Risk

Achievement

* Ratio

4.672E + 001

6.651E+001

3.705E + 001

2.649E + 001

4.596E + 003

3.298E + 001

3.298E + 001

9.234E+001

1.OOOE + 000

2.668E + 001

2.668E + 001

5.520E + 002

2.674E + 001

2.674E + 001

2.676E + 001

I-21
Table 1-16. Fault tree HPRI importance measures report

(10 year lower bound)

Family: HPI

Fault Tree: HPRI

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HPRAPPR

HPSEGHM

HPSEGKM

HPSEGMM

HP2425MVH

HPRCPPR

HP25MVO

HP24MVO

HPBCPPH

HP148VVT

HPCPPS

HP148VVH

HPCPPR

Probability

of

Failure

l.OOOE-002

6.500E-003

4.600E-003

6.100E-004

1. 500E -003

I.500E-003

5.OOOE-005

4.600E-003

6.500E-003

6.500E-003

1.500E-004

1.300E-003

9.1OOE-004

6.100E-004

4.600E-004

FussellVesely

Importance

4.121E -001

4.121E -001

3.170E -001

3.004E-001

3.004E - 001

3.004E-001

2.277E -001

2.234E -001

1.924E-001

1.924E-001

1.677E -001

6.312E - 002

4.418E -002

2.962E-002

2.234E-002

Risk

Reduction

Ratio

1.701E+000

1.701E + 000

1.464E + 000

1.429E + 000

1.429E + 000

1.429E + 000

1.295E + 000

1.288E + 000

1.238E+000

1.238E+000

1.202E + 000

1.067E +000

1.046E + 000

1.031E+000

1.023E + 000

Risk

Achievement

Ratio

4.180E+001

6.399E + 001

6.959E + 001

4.938E + 001

1.813E +002

1.OOOE + 000

4.550E + 003

4.933E + 001

3.041E + 001

3.041E + 001

1.119E+003

4.949E + 001

4.951E+001

4.953E + 001

4.953E+001

I-22
Table 1-17. Fault tree HPRI importance measures report

(40 year upper bound)

Family: HPI

Fault Tree: HPRI

(Sorted by Fussell-Vesely)

Event Name

HPRAPPR

HP26MVO

HPCROSSH

HP25MVO

HP24MVO

HPRCPPR

HP2425MVH

HP148VVT

HPCPPR

HPSEGMM

HPSEGKM

HPSEGHM

HPCPPS

HPBCPPH

HP 148VVH

Probability

of

Failure

6.800E-002

Il.OOOE-002

1.OOOE-002

1.OOOE-002

I.OOOE-002

6.800E-002

5.OOOE-005

2.200E-002

6.800E-003

1.500E-003

1.500E-003

6.1 00E - 004

2.700E - 003

1.500E-004

6.1 00E - 004

FussellVesely

Importance

4.923E - 001

4.861E -001

4.861E -001

3.354E-001

3.354E -001

3.282E -001

1.677E-001

1.062E-001

3.282E -002

2.461E -002

2.461E- 002

2.461E -002

1.303E -002

1.084E -002

2.944E -003

Risk

Reduction

Ratio

1.969E + 000

1.946E + 000

1.946E + 000

1.505E + 000

1.505E + 000

1.488E + 000

1.201E + 000

1.1 19E + 000

1.034E + 000

1.025E + 000

1.025E + 000

1.025E + 000

1.013E+000

1.OlIE+000

1.003E + 000

Risk

Achievement

Ratio

7.747E + 000

4.91 1E + 001

4.911E +001

3.420E + 001

3.420E + 001

5.498E + 000

3.350E+003

5.720E + 000

5.793E + 000

1.OOOE + 000

1.542E + 001

5.809E + 000

5.813E + 000

7.323E + 001

5.823E + 000

1-23
Table 1-18. Fault tree HPRI importance measures report

(40 year lower bound)

Family: HPI

Fault Tree: HPRI

(Sorted by Fussell-Vesely)

Event Name

HPCROSSH

HP26MVO

HPRAPPR

HPRCPPR

HP25MVO

HP24MVO

HP2425MVH

HP148VVT

HPSEGMM

HPSEGKM

HPSEGHM

HPBCPPH

HPCPPR

HPCPPS

HP 148VVH

Probability

of

Failure

l.OOOE-002

7.100E -003

2.OOOE-002

2.OOOE-002

7.100E- 003

7.1OOE-003

5.OOOE-005

6.800E-003

1.500E - 003

1.500E-003

6.100E-004

1.500E-004

2.000E - 003

1.400E-003

6.100E-004

FussellVesely

Importance

4.835E -001

4.835E -001

4.353E -001

3.078E-001

2.349E -001

2.349E -001

2.330E-001

1.046E-001

9.060E-002

9.060E-002

9.060E -002

4.866E -002

3.078E -002

2.154E -002

9.387E -003

Risk

Reduction

Ratio

1.936E+000

1.936E + 000

1.771E+000

1.445E + 000

1.307E+ 000

1.307E+ 000

1.304E + 000

1.11 7E + 000

1. IlOOE + 000

1. lOOE + 000

1.1 OOE + 000

1.051E+000

1.032E + 000

1.022E + 000

1.009E + 000

Risk

Achievement

Ratio

4.886E + 001

6.861E+001

2.233E + 001

1.608E + 001

3.385E+001

3.385E + 001

4.654E + 003

1.628E+ 001

1.000E + 000

5.508E + 001

1.633E+ 001

3.254E + 002

1.636E + 001

1.637E + 001

1.638E + 001

1-24
Table 1-19. Fault tree HPI cut sets quantification report

(5 year)

Family: HPI

Fault Tree: HPI

Mincut Upper Bound 9.590E - 005

Cut % 1o Cut

No. Total Set Freq. Cut Sets

1 52.1 52.1 5.OE-005 HP2425MVH

2 94.8 42.7 4.1E-005 HP24MVO, HP25MVO

3 96.6 1.7 1.7E-006 HP24MVO, HPSEGQM

4 98.3 1.7 1.7E -006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

5 98.9 .6 5.6E-007 HP102CVO, HP24MVO

6 99.5 .6 5.6E-007 HPlOICVO, HP25MVO

7 99.7 .2 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

8 99.8 .1 5.4E -008 HP26MVO, HPCPPS, HPCROSSH

9 99.8 .0 3.9E -008 HP148VVH, HP26MVO, HPCROSSH

10 99.9 .0 3.9E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

11 99.9 .0 2.5E-008 HP148VVT, HP26MVO, HPCROSSH

12 99.9 .0 2.3E-008 HPIOlCVO, HPSEGQM

13 99.9 .0 2.3E -008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM

14 99.9 .0 1.3E -008 HP26MVO, HPCPPR, HPCROSSH

Table 1-20. Fault tree HPI cut sets quantification report

(10 year upper bound)

Family: HPI

Fault 'lee: HPI

Mincut Upper Bound 1.048E -004

Cut % Oo Cut

No. Total Set Freq. Cut Sets

1 47.7 47.7 5.OE -005 HP2425MVH

2 94.5 46.8 4.9E-005 HP24MVO, HP25MVO

3 96.2 1.7 1.8E-006 HP24MVO, HPSEGQM

4 97.9 1.7 1.8E -006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

5 98.5 .6 6.4E -007 HP102CVO, HP24MVO

6 99.2 .6 6.4E -007 HPIOICVO, HP25MVO

7 99.4 .2 2.4E -007 HP148VVT, HP26MVO, HPCROSSH

8 99.6 .2 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

9 99.7 .1 7.7E-008 HP26MVO, HPCPPR, HPCROSSH

10 99.7 .1 7.7E-008 HP26MVO, HPCPPS, HPCROSSH

11 99.8 .0 4.3E1-008 HP148WH, HP26MVO, HPCROSSH

12 99.8 .0 4.3E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

13 99.9 .0 2.4E-008 HPlOlCVO, HPSEGQM

14 99.9 .0 2.4E -008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM

1-25
-7

Table 1-21. Fault tree HPI cut sets quantification report

(10 year lower bound)

Family: HPI

Fault ltee: HPI

Mincut Upper Bound 9.737E - 005

Cut % o Cut

No. Total Set Freq. Cut Sets

1 51.4 51.4 5.OE-005 HP2425MVH

2 94.7 43.4 4.2E - 005 HP24MVO, HP25MVO

3 96.5 1.7 1.7E-006 HP24MVO, HPSEGQM

4 98.2 1.7 1.7E - 006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

5 98.8 .6 5.7E-007 HPIOICVO, HP25MVO

6 99.4 .6 5.7E-007 HP102CVO, HP24MVO

7 99.6 .2 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

8 99.7 .1 8.5E -008 HP148VVT, HP26MVO, HPCROSSH

9 99.8 .1 5.9E-008 HP26MVO, HPCPPS, HPCROSSH

10 99.8 .0 4.OE-008 HP148VVH, HP26MVO, HPCROSSH

l1 99.8 .0 4.OE -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

12 99.9 .0 3.E -008 HP26MVO, HPCPPR, HPCROSSH

13 99.9 .0 2.3E-008 HPlOlCVO, HPSEGQM

14 99.9 .0 2.3E -008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM

1-26
Table 1-22. Fault tree HPI cut sets quantification report

(40 year upper bound)

Family: HPI

Fault Tree: HPI

Mincut Upper Bound 1.639E - 004

Cut QO %0 Cut

No. Total Set Freq. Cut Sets

1 61.0 61.0 1.OE-004 HP24MVO, HP25MVO

2 91.5 30.5 5.OE-005 HP2425MVH

3 93.1 1.6 2.6E-006 HP24MVO, HPSEGQM

4 94.7 1.6 2.6E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

5 96.0 1.3 2.2E-006 HP148VVT, HP26MVO, HPCROSSH

6 96.8 .7 1.2E-006 HPlOlCVO, HP25MVO

7 97.5 .7 1.2E-006 HP102CVO, HP24MVO

8 98.1 .6 L.OE-006 HP148VVT, HPAPPR, HPBPPR

9 98.5 .4 6.8E -007 HP26MVO, HPCPPR, HPCROSSH

10 98.7 .2 3.IE-007 HPAPPR, HPBPPR, HPCPPR

11 98.9 .2 2.7E-007 HP26MVO, HPCPPS, HPCROSSH

12 99.0 .1 2.2E - 007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

13 99.2 .1 2.2E-007 HP148VVT, HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

14 99.3 .1 2.2E -007 HP148VVT, HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

15 99.4 .1 1.2E-007 HPAPPR, HPBPPR, HPCPPS

16 99.4 .1 8.9E -008 HP148VVT, HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

17 99.5 .0 6.9E -008 HPAPPR, HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

18 99.5 .0 6.9E -008 HPBPPR, HPCPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

19 99.6 .0 6.1E-008 HP148VVH, HP26MVO, HPCROSSH

20 99.6 .0 6.1E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

21 99.6 .0 5.9E -008 HP148VVT, HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM

22 99.7 .0 4.8E-008 HP148VVT, HP24MVO, HPSEGOH

23 99.7 .0 4.5E-008 HP113CVO, HP26MVO, HPCROSSH

24 99.7 .0 3.3E-008 HP148VVT, HP26MVCH, HPCROSSH

25 99.7 .0 3.1E-008 HPIOICVO, HPSEGQM

26 99.7 .0 3.E -008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM

27 99.8 .0 2.8E-008 HP148VVH, HPAPPR, HPBPPR

28 99.8 .0 2.8E -008 HPAPPR, HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM

29 99.8 .0 2.7E -008 HPAPPR, HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM

30 99.8 .0 2.7E - 008 HPBPPS, HPCPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

31 99.8 .0 2.7E-008 HPBPPR, HPCPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

32 99.8 .0 2.2E-008 HP148VVT, HP26MVO, HP409MVO, HP410MVO

33 99.9 .0 2.1E-008 HP113CVO, HPAPPR, HPBPPR

34 99.9 .0 1.8E -008 HP26MVM, HPCPPR, HPCROSSH, /HPSEGHM, /HPSEGQM

35 99.9 .0 1.5E-008 HP26MVO, HPBCPPH, HPCROSSH

36 99.9 .0 1.5E-008 HP26MVO, HP27MVH, HPCROSSH

37 99.9 .0 1.5E -008 HP24MVO, HPCPPR, HPSEGOH

1-27
Table 1-23. Fault tree HPI cut sets quantification report

(40 year lower bound)

Family: HPI

Fault Tree: HPI

Mincut Upper Bound 1.067E - 004

Cut % 50 Cut

No. Total Set Freq. Cut Sets

1 47.2 47.2 5.OE - 005 HP24MVO, HP25MVO

2 94.1 46.9 5.OE-005 HP2425MVH

3 95.8 1.7 1.8E-006 HP24MVO, HPSEGQM

4 97.6 1.7 1.8E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

5 98.2 .6 6.5E-007 HPIOICVO, HP25MVO

6 98.8 .6 6.5E-007 HP102CVO, HP24MVO

7 99.2 .5 4.8E-007 HP148VVT, HP26MVO, HPCROSSH

8 99.4 .2 2.2E - 007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

9 99.6 .1 1.4E-007 HP26MVO, HPCPPR, HPCROSSH

10 99.7 .1 L.OE-007 HP26MVO, HPCPPS, HPCROSSH

11 99.7 .0 4.3E-008 HP148VVH, HP26MVO, HPCROSSH

12 99.7 .0 4.3E - 008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

13 99.8 .0 2.7E-008 HP148VVT, HPAPPR, HPBPPR

14 99.8 .0 2.4E-008 HPIOlCVO, HPSEGQM

Table 1-24. Fault tree HP201 cut sets quantification report

(5 year)

Family: HPI

Fault flee: HP201

Mincut Upper Bound 1.720E - 004

Cut % % Cut

No. Total Set Freq. Cut Sets

1 37.2 37.2 6.4E-005 HP26MVO, HPCROSSH

2 66.3 29.1 5.OE-005 HP2425MVH

3 90.1 23.8 4.IE-005 HP24MVO, HP25MVO

4 91.7 1.6 2.7E - 006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM

5 92.6 1.0 1.7E-006 HP24MVO, HPSEGQM

6 93.6 1.0 1.7E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

7 94.5 .9 1.5E -006 HP26MVCH, HPCROSSH

8 95.3 .8 1.4E-006 HP24MVO, HPSEGOH

9 96.0 .7 1.3E-006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM

10 96.7 .7 1.3E-006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

11 97.3 .5 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM

12 97.6 .3 5.8E-007 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM

13 97.9 .3 5.6E-007 HP102CVO, HP24MVO

14 98.3 .3 5.6E-007 HPlOICVO, HP25MVO

1-28
Table 1-25. Fault tree HP201 cut sets quantification report

(10 year upper bound)

Family: HPI

Fault Tree: HP201

Mincut Upper Bound 2.053E - 004

Cut % !o Cut

No. Total Set Freq. Cut Sets

1 34.1 34.1 7.OE-005 HP26MVO, HPCROSSH

2 58.5 24.4 5.OE-005 HP2425MVH

3 82.3 23.9 4.9E-005 HP24MVO, HP25MVO

4 84.9 2.6 5.2E-006 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM

5 86.7 1.9 3.8E-006 HP148VVT, HPBPPR

6 88.1 1.3 2.7E-006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM

7 88.9 .9 1.8E - 006 HP24MVO, HPSEGQM

8 89.8 .9 1.8E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

9 90.6 .8 1.6E-006 HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

10 91.4 .8 1.6E - 006 HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

11 92.2 .8 1.6E-006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM

12 93.0 .8 1.6E - 006 HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

13 93.8 .8 1.6E-006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

14 94.6 .8 1.5E - 006 HP24MVO, HPSEGOH

15 95.3 .7 1.5E-006 HP26MVCH, HPCROSSH

16 95.9 .6 1.2E-006 HPAPPR, HPBPPR

17 96.5 .6 1.2E-006 HPBPPR, HPCPPR

18 97.1 .6 1.2E-006 HPBPPR, HPCPPS

19 97.5 .4 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM

20 97.9 .3 6.7E -007 HP148VVH, HPBPPR

21 98.2 .3 6.7E-007 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM

22 98.5 .3 6.4E-007 HP102CVO, HP24MVO

23 98.8 .3 6.4E-007 HPlOlCVO, HP25MVO

24 99.0 .2 3.4E-007 HP26MVO, HP409MVO, HP410MVO

25 99.1 .1 2.3E-007 HPBCPPH, HPSEGKM

26 99.2 .1 2.2E -007 HPBCPPH, /HPSEGHM, /HPSEGKM, HPSEGMM

27 99.3 .1 2.2E -007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM

28 99.4 .1 2.1E -007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM

29 99.5 .1 2.IE-007 HPI13CVO, /HPSEGHM, /HPSEGKM, HPSEGMM

30 99.6 .1 1.7E-007 HP27MVH, HPBPPR

31 99.7 .1 1.7E -007 HPBCPPH, HPBPPR

32 99.7 .1 1.5E-007 HP113CVO, HPBPPR

33 99.8 .0 5.7E -008 /HPSEGHM, HPSEGOH, HPSEGPM, /HPSEGQM

34 99.8 .0 5.2E -008 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM

35 99.8 .0 3.9E-008 HP148VVT, HPAPPR, HPCROSSH

36 99.8 .0 3.7E - 008 HP148VVT, HP409MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

37 99.8 .0 2.7E-008 HP148VVT, HP409MVO, HPAPPR

38 99.9 .0 2.4E-008 HPlOlCVO, HPSEGQM

1-29
Table 1-26. Fault tree HP201 cut sets quantification report

(10 year lower bound)

Family: HPI

Fault Tree: HP201

Mincut Upper Bound 1.788E - 004

Cut 0o qo Cut

No. Total Set Freq. Cut Sets

1 36.4 36.4 6.5E - 005 HP26MVO, HPCROSSH

2 64.3 28.0 5.OE - 005 HP2425MVH

3 87.9 23.6 4.2E-005 HP24MVO, HP25MVO

4 89.5 1.5 2.7E -006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM

5 90.5 1.1 1.9E -006 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM

6 91.5 .9 1.7E-006 HP24MVO, HPSEGQM

7 92.4 .9 1.7E - 006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

8 93.3 .8 1.5E-006 HP26MVCH, HPCROSSH

9 94.1 .8 1.4E-006 HP24MVO, HPSEGOH

10 94.8 .8 1.4E-006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

11 95.6 .8 1.4E-006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM

12 96.1 .5 9.1E-007 HP 148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM

13 96.5 .4 6.9E-007 HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

14 96.9 .4 6.9E -007 HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

15 97.3 .4 6.9E -007 HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

16 97.6 .3 6.OE-007 HP148VVT, HPBPPR

17 97.9 .3 5.7E-007 HPIOICVO, HP25MVO

18 98.2 .3 5.7E-007 HP102CVO, HP24MVO

19 98.5 .2 4.2E -007 HPBPPR, HPCPPS

20 98.6 .2 2.8E-007 HP148VVH, HPBPPR

21 98.8 .2 2.8E -007 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM

22 98.9 .2 2.7E -007 HP26MVO, HP409MVO, HP410MVO

23 99.1 .1 2.3E-007 HPBCPPH, HPSEGKM

24 99.2 .1 2.2E -007 HPBCPPH, /HPSEGHM, /HPSEGKM, HPSEGMM

25 99.3 .1 2.2E -007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM

26 99.4 .1 2.IE-007 HPAPPR, HPBPPR

27 99.6 .1 2.1E-007 HPBPPR, HPCPPR

28 99.6 .1 1.4E-007 HPI13CVO, /HPSEGHM, /HPSEGKM, HPSEGMM

29 99.7 .1 1.4E -007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM

30 99.7 .0 6.9E -008 HP27MVH, HPBPPR

31 99.8 .0 6.9E-008 HPBCPPH, HPBPPR

32 99.8 .0 5.7E -008 /HPSEGHM, HPSEGOH, HPSEGPM, /HPSEGQM

33 99.8 .0 4.4E-008 HP113CVO, HPBPPR

34 99.9 .0 2.3E-008 HPlOlCVO, HPSEGQM

35 99.9 .0 2.3E-008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM

36 99.9 .0 1.9E -008 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM

37 99.9 .0 1.9E-008 HPlOlCVO, HPSEGOH

38 99.9 .0 1.4E -008 HPCPPS, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM

1-30
Table 1-27. Fault tree HP201 cut sets quantification report

(40 year upper bound)

Family: HPI

Fault Tree: HP201

Mincut Upper Bound 6.194E - 004

Cut % %1⁄4 Cut

No. Total Set Freq. Cut Sets

1 24.2 24.2 1.5E-004 HP148VVT, HPBPPR

2 40.3 16.1 1.OE-004 HP24MVO, HP25MVO

3 56.4 16.1 1.OE-004 HP26MVO, HPCROSSH

4 64.5 8.1 5.OE-005 HP2425MVH

5 72.0 7.5 4.6E - 005 HPBPPR, HPCPPR

6 79.4 7.5 4.6E -005 HPAPPR, HPBPPR

7 84.8 5.3 3.3E-005 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM

8 87.7 3.0 1.8E-005 HPBPPR, HPCPPS

9 89.4 1.6 1.OE -005 HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

10 91.0 1.6 1.0E-005 HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

11 92.7 1.6 1.OE -005 HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

12 93.3 .7 4.1E-006 HP148VVH, HPBPPR

13 94.0 .7 4.1E -006 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM

14 94.6 .7 4.OE -006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM

15 95.3 .7 4.OE -006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

16 95.8 .5 3.1E-006 HP113CVO, HPBPPR

17 96.2 .4 2.7E -006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM

18 96.6 .4 2.6E-006 HP24MVO, HPSEGQM

19 97.1 .4 2.6E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

20 97.4 .4 2.2E - 006 HP24MVO, HPSEGOH

21 97.7 .2 1.5E-006 HP26MVCH, HPCROSSH

22 97.9 .2 1.5E-006 HP148VVT, HP409MVO, HPAPPR

23 98.1 .2 1.5E-006 HP148VVT, HPAPPR, HPCROSSH

24 98.3 .2 1.2E-006 HP102CVO, HP24MVO

25 98.5 .2 1.2E-006 HPIOICVO, HP25MVO

26 98.7 .2 1.OE-006 HPBCPPH, HPBPPR

27 98.9 .2 1.OE-006 HP27MVH, HPBPPR

28 99.0 .2 1.OE-006 HP26MVO, HP409MVO, HP410MVO

29 99.2 .1 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM

30 99.3 .1 6.7E-007 HP113CVO, /HPSEGHM, /HPSEGKM, HPSEGMM

31 99.4 .1 6.7E-007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM

32 99.5 .1 4.6E-007 HP409MVO, HPAPPR, HPCPPR

33 99.5 .1 4.6E-007 HPAPPR, HPCPPR, HPCROSSH

34 99.6 .1 3.3E-007 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM

35 99.6 .1 3.3E-007 HP148VVT, HP409MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

36 99.7 .0 2.3E -007 HPBCPPH, HPSEGKM

37 99.7 .0 2.2E-007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM

1-31
Table 1-28. Fault tree HP201 cut sets quantification report

(40 year lower bound)

Family: HPI

Fault Tree: HP201

Mincut Upper Bound 2.365E - 004

Cut %O %/o Cut

No. Total Set Freq. Cut Sets

1 30.0 30.0 7.1E-005 HP26MVO, HPCROSSH

2 51.3 21.3 5.0E-005 HP24MVO, HP25MVO

3 72.5 21.1 5.0E-005 HP2425MVH

4 78.2 5.8 1.4E-005 HP148VVT, HPBPPR

5 82.5 4.3 1.OE-005 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM

6 84.2 1.7 4.OE-006 HPBPPR, HPCPPR

7 85.9 1.7 4.0E-006 HPAPPR, HPBPPR

8 87.2 1.3 3.E -006 HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM

9 88.4 1.3 3.0E-006 HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

10 89.7 1.3 3.E -006 HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM

11 90.9 1.2 2.8E-006 HPBPPR, HPCPPS

12 92.0 1.1 2.7E -006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM

13 92.9 .9 2.IE-006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM

14 93.8 .9 2.IE-006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM

15 94.6 .8 1.8E-006 HP24MVO, HPSEGQM

16 95.4 .8 1.8E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM

17 96.0 .7 1.6E-006 HP24MVO, HPSEGOH

18 96.7 .6 1.5E-006 HP26MVCH, HPCROSSH

19 97.2 .5 1.2E-006 HP148VVH, HPBPPR

20 97.7 .5 1.2E -006 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM

21 98.1 .4 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM

22 98.4 .3 6.5E-007 HPlOICVO, HP25MVO

23 98.6 .3 6.5E-007 HP102CVO, HP24MVO

24 98.8 .2 3.6E-007 HP26MVO, HP409MVO, HP410MVO

25 98.9 .1 3.OE-007 HP27MVH, HPBPPR

26 99.0 .1 3.0E-007 HPBCPPH, HPBPPR

27 99.2 .1 2.8E-007 HP113CVO, HPBPPR

28 99.2 .1 2.3E-Q007 HPBCPPH, HPSEGKM

29 99.3 .1 2.2E-007 HPBCPPH, /HPSEGHM, /HPSEGKM, HPSEGMM

30 99.4 .1 2.2E - 007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM

31 99.5 .1 2.1E-007 HP113CVO, /HPSEGHM, /HPSEGKM, HPSEGMM

32 99.6 .1 2.1E-007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM

33 99.7 .1 1.4E-007 HP148VVT, HPAPPR, HPCROSSH

34 99.7 .0 1.OE-007 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM

35 99.8 .0 I.OE-007 HP148VVT, HP409MVO, HPAPPR

36 99.8 .0 7.2E-008 HP148VVT, HP4Q9MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

37 99.8 .0 5.7E -008 /HPSEGHM, HPSEGOH, HPSEGPM, /HPSEGQM

38 99.8 .0 4.OE -008 HPAPPR, HPCPPR, HPCROSSH

1-32
Table 1-29. Fault tree HPRI cut sets quantification report

(5 year)

Family: HPI

Fault Tree: HPRI

Mincut Upper Bound 8.401E - 005

Cut % qo Cut

No. Total Set Freq. Cut Sets

1 26.7 26.7 2.2E - 007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

2 42.0 15.2 1.3E-007 HP26MVO, HPCROSSH, HPRCPPR

3 53.9 11.9 1.OE-007 HP2425MVH, HPRAPPR

4 63.6 9.8 8.2E-008 HP24MVO, HP25MVO, HPRAPPR

5 72.5 8.9 7.5E -008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM

6 79.8 7.3 6.1E -008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

7 86.2 6.4 5.4E -008 HP26MVO, HPCPPS, HPCROSSH

8 90.9 4.6 3.9E -008 HP148VVH, HP26MVO, HPCROSSH

9 95.5 4.6 3.9E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

10 98.5 3.0 2.5E-008 HP148VVT, HP26MVO, HPCROSSH

11 100.0 1.5 1.3E -008 HP26MVO, HPCPPR, HPCROSSH

Table 1-30. Fault tree HPRI cut sets quantification report

(10 year upper bound)

Family: HPI

Fault Tree: HPRI

Mincut Upper Bound 2.716E -006

Cut 010 %o Cut

No. Total Set Freq. Cut Sets

1 28.4 28.4 7.7E -007 HP26MVO, HPCROSSH, HPRCPPR

2 48.6 20.3 5.5E -007 HP2425MVH, HPRAPPR

3 68.4 19.8 5.4E -007 HP24MVO, HP25MVO, HPRAPPR

4 77.5 9.0 2.4E -007 HP148VVT, HP26MVO, HPCROSSH

5 85.7 8.3 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

6 88.6 2.8 7.7E - 008 HP26MVO, HPCPPR, HPCROSSH

7 91.4 2.8 7.7E - 008 HP26MVO, HPCPPS, HPCROSSH

8 94.2 2.8 7.5E - 008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM

9 96.9 2.7 7.3E - 008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

10 98.4 1.6 4.3E-008 HP148VVH, HP26MVO, HPCROSSH

11 100.0 1.6 4.3E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

I-33
Table 1-31. Fault tree HPRI cut sets quantification report

(10 year lower bound)

Family: HPI

Fault Tree: HPRI

Mincut Upper Bound 1.339E - 006

Cut qO %o Cut

No. Total Set Freq. Cut Sets

1 22.3 22.3 3.0E-007 HP26MVO, HPCROSSH, HPRCPPR

2 39.5 17.2 2.3E - 007 HP2425MVH, HPRAPPR

3 56.3 16.8 2.2E-007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

4 70.8 14.5 1.9E-007 HP24MVO, HP25MVO, HPRAPPR

5 77.1 6.3 8.5E-008 HP148VVT, HP26MVO, HPCROSSH

6 82.7 5.6 7.5E - 008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM

7 87.4 4.7 6.3E - 008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

8 91.8 4.4 5.9E-008 HP26MVO, HPCPPS, HPCROSSH

9 94.8 3.0 4.OE-008 HP148VVH, HP26MVO, HPCROSSH

10 97.7 3.0 4.OE -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

11 100.0 2.2 3.ME - 008 HP26MVO, HPCPPR, HPCROSSH

Table 1-32. Fault tree HPRI cut sets quantification report

(40 year upper bound)

Family: HPI

Fault Tree: HPRI

Mincut Upper Bound 2.072E - 005

Cut % % Cut

No. Total Set Freq. Cut Sets

1 32.8 32.8 6.8E-006 HP26MVO, HPCROSSH, HPRCPPR

2 65.6 32.8 6.8E-006 HP24MVO, HP25MVO, HPRAPPR

3 82.0 16.4 3.4E-006 HP2425MVH, HPRAPPR

4 92.7 10.6 2.2E-006 HP148VVT, HP26MVO, HPCROSSH

5 95.9 3.3 6.8E-007 HP26MVO, HPCPPR, HPCROSSH

6 97.2 1.3 2.7E-007 HP26MVO, HPCPPS, HPCROSSH

7 98.3 1.1 2.2E-007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

8 99.1 .7 1.5E -007 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

9 99.4 .4 7.5E -008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM

10 99.7 .3 6.1E-008 HP148VVH, HP26MVO, HPCROSSH

11 100.0 .3 6.1E-008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

1-34
Table 1-33. Fault tree HPRI cut sets quantification report

(40 year lower bound)

Family: HPI

Fault Tree: HPRI

Mincut Upper Bound 4.614E - 006

Cut %O 0 Cut

No. Total Set Freq. Cut Sets

1 30.8 30.8 1.4E-006 HP26MVO, HPCROSSH, HPRCPPR

2 52.6 21.9 1.OE-006 HP24MVO, HP25MVO, HPRAPPR

3 74.3 21.7 1 .0E - 006 HP2425MVH, HPRAPPR

4 84.8 10.5 4.8E -007 HP148VVT, HP26MVO, HPCROSSH

5 89.6 4.9 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM

6 92.7 3.1 1.4E-007 HP26MVO, HPCPPR, HPCROSSH

7 94.9 2.2 1.OE-007 HP26MVO, HPCPPS, HPCROSSH

8 96.5 1.6 7.5E -008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM

9 98.1 1.6 7.5E -008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM

10 99.1 .9 4.3E-008 HP148VVH, HP26MVO, HPCROSSH

11 100.0 .9 4.3E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM

1-35
REFERENCES

I-l. Oconee PAA, A Probabilistic Risk Assessment of Oconee 3, The Nuclear Safety Analysis Center, Electric Power Research Institute Palo Alto, California and Duke Power Company, Charlotte, North

Carolina, NSAC 60, Vol. 1, Vol. 2, Vol. 3, and Vol. 4, June 1984.

1-36
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 12-89) (Assigned by NRC. Add Vol.. Supp., Rev..

320C,3202. BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/CR-4967

2. TITLE AND SUBTITLE EGG-2514

3. DATE REPORT PUBLISHED

Nuclear Plant Aging Research . . MONTH YEAR

ton High Pressure Injection Systems August 1989

4. FIN Oh GRANT NUMBER

A6389

5. AUTHOR(S) 6. TYPE OF REPORT

Research

Leroy C. Meyer 7. PERIOD COVERED {Inclusive Dares)

B. P ER FORMING ORGAN IZATION - NAME AND ADDR ESS {if NRC, provide Division, Office or Region, U.& Nuclear Regularory Commission, and mailing address: if contractor. provide name and mailing addressj

EG&G Idaho, Inc.

P.O. Box 1625

Idaho Falls, ID 83415

9. SPONSOR ING ORGAN IZATI ON -NAME AND ADDR ESS 1f NRC. type `s as above- if contracor. provide NRC Division. Office or Region, US. Nuclear Regulatory Commission. and mailing address j

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

This report represents the results of a review of light water reactor High Pressure

Injection System (HPIS) operating experiences reported in the Nuclear Power Experience Data Base, Licensee Event Reports (LER)s, Nuclear Plant Reliability Data

System, and plant records. The purpose is to evaluate the potential significance of

aging as a contributor to degradation of the High Pressure Injection System. Tables

are presented that show the percentage of events for HPIS classified by cause, component, and subcomponents for PWRs. A representative Babcock and Wilcox plant was

selected for detailed study. The U.S. Nuclear Regulatory Commission's Nuclear Plant

Aging Research guidelines were followed in performing the detailed study that identifies materials susceptible to aging, stressors, environmental factors, and failure

modes for the HPIS.

In addition to the engineering evaluation, the components that contributed to system unavailability were determined and the aging contribution to HPIS unavailability

was evaluated. The unavailability assessment utilized an existing probabilistic risk

assessment (PRA), the linear aging model, and generic failure data.

12. KEY WORDS/DESCRI PTO RS (List evords or phrases that will assist researchers in locating the report.) 13. AVAILABSI LIT Y STATEMENT

High Pressure Injection System (HPIS) 14 SECURITYeCLASSIFICATIO

Aging 9(rhis Page)

Unclassified

(This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-891 "U.S. GOVERNKNT PRINTING OrFICEt`1989- 241-590100230
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

SPECIAL FOURTH-CLASS RATE POSTAGE e FEES PAID

USNRC

PERMIT No. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300