NUREG/CR-4967, 'Nuclear Plant Aging Research on High ... · liminary identification of failure...
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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
Transcript of NUREG/CR-4967, 'Nuclear Plant Aging Research on High ... · liminary identification of failure...
NUREG/CR-4967EGG-2514
Nuclear Plant Aging Research onHigh Pressure Injection Systems
Prepared by L. C. Meyer*
Idaho National Engineering LaboratoryEG&G Idaho, Inc.
Prepared forU.S. Nuclear RegulatoryCommission
I
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NUREG/CR-4967EGG-2514RM, R9
Nuclear Plant Aging Research onHigh Pressure Injection Systems
Manuscript Completed: July 1989Date Published: August 1989
Prepared byL. C. Meyer
Idaho National Engineering LaboratoryManaged by the U.S. Department of Energy
EG&G Idaho, Inc.P. 0. Box 1625Idaho Falls, ID 83415
Prepared forDivision of EngineeringOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, DC 20555NRC FIN A6389Under DOE Contract No. DE-AC07-761D01570
ABSTRACT
This report represents the results of a review of light water reactor High PressureInjection System (HPIS) operating experiences reported in the Nuclear Power Experi-ence Data Base, Licensee Event Reports (LER)s, Nuclear Plant Reliability DataSystem, and plant records. The purpose is to evaluate the potential significance ofaging as a contributor to degradation of the High Pressure Injection System. Tablesare presented that show the percentage of events for HPIS classified by cause, compo-nent, and subcomponents for PWRs. A representative Babcock and Wilcox plant wasselected for detailed study. The U.S. Nuclear Regulatory Commission's Nuclear PlantAging Research guidelines were followed in performing the detailed study that identi-fies materials susceptible to aging, stressors, environmental factors, and failuremodes for the HPIS.
In addition to the engineering evaluation, the components that contributed to sys-tem unavailability were determined and the aging contribution to HPIS unavailabilitywas evaluated. The unavailability assessment utilized an existing probabilistic riskassessment (PRA), the linear aging model, and generic failure data.
FIN No. A6389-Nuclear Plant Aging Research on High Pressure Injection Systems
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EXECUTIVE SUMMARY
Operating experience of nuclear power plants isevaluated to determine the significance of servicewear on equipment due to aging and the possibleimpact of service wear on safety. The High PressureInjection System (HPIS) and those portions ofrelated systems needed for operation of the HPISwere selected for detailed study and emphasized inthis report. This research is part of the U.S.Nuclear Regulatory Commission's (USNRC's)Nuclear Plant Aging Research (NPAR) Programand follows the NPAR guidelines.
The NPAR guidelines provided the frameworkthrough which the effect of aging on HPI was stud-ied. The products asked for in the NPAR guidelinesinclude: an identification of failure modes; a pre-liminary identification of failure causes due toaging and service wear degradation; and a review ofcurrent inspection, surveillance and monitoringmethods, including manufacturer recommendedsurveillance and maintenance practices. Perform-ance parameters or functional indicators poten-tially useful in detecting degradation are alsoidentified and preliminary recommendations aremade regarding inspection, surveillance, and moni-toring methods.
A description of the HPIS for a Babcock andWilcox PWR is presented based on informationprovided by a cooperating utility. The descriptionprovides a general understanding of a HPIS. Avariety of designs exist from the various NSSS ven-dors. However, they are all similar in that they useborated water, generally inject into the cold-leg pip-ing, utilize high head centrifugal motor drivenpumps, have motor operated valves and checkvalves, and have similar operating environments.
There is some concern that the operational dif-ferences and variations in system boundaries wouldlead to different conclusions if each type of systemwere considered separately. However, this is notconsidered likely because of the similarity of equip-ment and environment.
The HPIS for the B&W type of system is part ofthe Emergency Core Cooling System (ECCS) andhas two modes of operation under emergency con-ditions. The first is the high pressure injectionmode that is necessary to prevent uncovering of thecore for small LOCAs where high system pressure ismaintained, and to delay uncovering of the core forintermediate size LOCAs. The second provideslong-term core cooling following a LOCA using thehigh pressure recirculation mode. Part of the HPIS
is also used during normal operation to providereactor coolant pump seal cooling and to maintainthe volume of the reactor coolant system withinacceptable limits. The HPIS interfaces with manyother systems in performing its functions, whichinclude IE electrical power, service water, instru-ment air, low pressure injection system, and theEngineered Safety Feature Actuating System.
In addition to the one plant that was studied indetail, generic data bases were reviewed for HPISfailures in PWRs. This review provided larger sta-tistical bases for determining HPIS problemsrelated to aging. Data sources used include theNuclear Power Experience (NPE) data base (priorto 1986), Licensee Event Reports (LERs) (prior to1984), Nuclear Plant Reliability Data System(NPRDS) (prior to 1986), and material from anoperating nuclear plant supplied by a utility(including personnel interviews). The systemboundaries as defined by each data base were usedas reported.
A review of Nuclear Power Experience (NPE)shows that the most frequent cause of HPIS com-ponent failures is maintenance error, followed bydesign error, and mechanical disability. The fourtypes of components with the highest frequency offailure for PWRs were valves-35%o; I&C-19%;pumps-15%; and piping-7%.
Instrumentation and Control (I&C) failuresincluded the sensors, electronics, and motor con-trol centers for valves and pumps. Instrumentationaccounted for 50% of the I&C failures, valve con-trol 17%, pump control 9%, and the rest were mis-cellaneous control circuits.
HPI piping failures, after eliminating design,construction, and maintenance errors (whichaccounted for 37%), were primarily due to weldfailures-15%, corrosion-7%, and vibration-5 %.The rest of the events were spread over manycauses. The HPI pipe sizes varied from I to 14 in.,depending upon location. Failures were not domi-nated by any one particular size.
Command faults were the leading cause for valvefailures, reported in LERs. They include electricalpower or any support system that prevents the valvefrom performing its intended function. Mechanicalparts failure, seat or disk failure problems, packingfailures (leaks), and foreign material were the mostfrequent causes of the basic valve failures. Therewere 44 HPI pump failures reported in the LERdata base. Out of these, 22 events were caused by
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control, maintenance, and design error. Theremaining events found no single dominant causeof failure. Only nine failures of HPIS pumps werepotentially aging related.
The NPRDS data followed the same componentfailure pattern as NPE. The aging fraction for HPIbased 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 forthose events requiring an incident investigationreport. Plant maintenance records listed manymore events. While the top four components withthe highest frequency of failure were the same, theorder placed I&C first; pipe, supports, and nozzlessecond; then valves and pumps. These data indicatethat many minor problems associated with I&Cand pipe hangers received corrective maintenancebefore major failures occurred.
A review of the electrical standards identifiedthat operational life is based on accelerated agingtests. As naturally aged component data becomesavailable, the standards should be updated. Stand-ards for mechanical equipment are under review byASME Section XI and are supported by the NPARresearch where applicable.
The conclusions for the HPIS study are based ona review of one plant and generic information fromthe various data bases. The plant maintenancerecord contains many minor adjustments andrepairs for I&C and pipe hangers, but most of themajor components failures concerned valves andpumps. Serious piping problems concerning ther-mal sleeve and nozzle cracking were attributed tothermal fatigue. The utilities affected have takencorrective action by redesigning the thermalsleeves, using warm up lines and enhanced IS&M.Materials in seals and valve packing deteriorateswith time and results in leaks. Borated water leaks
are potentially serious because of the corrosiveaction of boric acid on carbon steel and potentialfor loss of pressure boundary.
Approximately 57% of the component failurelead to system degradation but, because of systemredundancy, only 0.7% caused loss of system func-tion. The failure modes that involve total loss ofsystem function are, failure to inject cooling waterfor emergency operations, and failure to providemakeup water or seal cooling water for normaloperations.
The specific problems related to aging were:(a) through-wall cracks occurring in the makeupnozzle and safety injection line elbow from thermalfatigue, (b) valves failing to operate due to boroncrystallization, and (c) injection boron concentra-tion diluted from leaking valves.
Inspection and surveillance review has identifiedthat electrical measurements on pump motors andvalve operators (for MOVs) could be used to detectaging. Also, that improved inservice testing ofvalves is needed to detect aging and assure operabil-ity with load. The detection of cracks caused bythermal fatigue requires enhanced ultrasonic test-ing methods. In addition, inspection of base metalin high-stress regions is needed to detect cracks inthose areas.
The HPIS unavailability assessment identifiedthat motor-operated valves contributed signifi-cantly to unavailability of the system for all threeoperating modes evaluated. The HPIS pumps weresignificant contributors to systems unavailabilityfor the two of three pumps required mode and therecirculation mode of operation. The time depen-dant unavailability assessment showed that theHPI unavailability was only moderately affected byaging with a relatively small increase over the oper-ating life.
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ACKNOWLEDGMENTS
The detailed systems studies are based on material supplied by the Duke PowerCompany, Charlotte, North Carolina.
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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 ............. ......................................... 12HPI Pump B Controls .............. ......................................... 12HPI Pump C Controls ............. ......................................... 14Motor-Operated Valves .............. ........................................ 14Pneumatic-Operated Valves .................................................. 14
HPI Pumps .................................................................... 14
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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 . ............................................................. 19System Pumps . ............................................................. 19Valves ..................................................................... 19
Nuclear Plant Reliability Data System .............................................. 19
Plant Operating Experience from Site Visit and Personnel Interviews ..... ............... 26
Nuclear Maintenance Data Base .............................................. 26Incident Investigation Reports ................................................ 27
Summary of Operating Experience ................................................. 27
Piping . ................................................................... 29Snubbers .................................................................. 30Penetrations ............................................................... 30Tanks ..................................................................... 30Pumps .................................................................... 30Valves ..................................................................... 30Chemistry Problems ............. ........................................... 30Microbial 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
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AGING ASSESSMENTS FOR THE HPIS ............................................... 33
Preliminary Identification of Susceptibility of Materials to Aging ....................... 33
Stressors for HPIS . ............................................................... 33
Maintenance, Operations, and Testing Stressors ................................. 34Environmental Stressors ............. ........................................ 34Electrical Stressors ................ .......................................... 34Mechanical 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 ............................................... 40High 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
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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, PURIFICATIONSYSTEM, 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 HPISCOMPONENTS . ............................................................... C-1
APPENDIX D-ENGINEERED SAFETY FEATURES ACTUATING SYSTEMFOR 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
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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 emergencymodes of operation . .............................................................. 47
11. HPIS unavailability vs years showing the effect of aging for the three emergencyoperating 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 andengineering 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
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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 componentunavailabilities 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
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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
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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
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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
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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
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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
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NUCLEAR PLANT AGING RESEARCHON HIGH PRESSURE INJECTION SYSTEMS
INTRODUCTION
As part of its responsibilities to protect the publichealth and safety, the USNRC is concerned withthe effect aging has on the safety of commercialnuclear power plants. To meet this responsibility,the USNRC has developed and implemented ahardware-oriented research program to investigateplant aging and the related degradation of compo-nents, systems, and structures. This program iscalled the Nuclear Plant Aging Research (NPAR)Program and is being conducted by the Electricaland Mechanical Engineering Branch of the Divi-sion of Engineering of the office of Nuclear Regu-latory Research. 1 A complementary programfocusing on pressure vessel, piping, steam genera-tor materials problems, and nondestructive exami-nation methods is being conducted by the MaterialsEngineering Branch of the Division ofEngineering.
Aging and Plant Safety
The NPAR Program is investigating how theaging of components, systems, and civil structurescan affect the safe operation of nuclear powerplants. The United States currently has approxi-mately 100 commercial pressurized and boilinglight water reactors in operation. In the context ofNPAR, aging is defined as the "cumulative degra-dation that occurs with the passage of time in acomponent, system, or structure." The main con-cern of the NPAR program is that plant safetycould be compromised if aging degradation is notdetected and corrective action taken before there isa loss of the required functional capability in acomponent, system, or structure. Consequently,aging might result in a reduction in the safety levelachieved by the defense-in-depth approach used toensure the safety of domestic reactors. Defense-in-depth requires that the public is protected from theaccidental release of fission products by a series ofmultiple barriers and engineered safety systems.
Operating plant experience provides exampleswhere age induced degradation of a key componenthas led to a reduction in the capability of a barrierto prevent the release of fission products. These
examples include degradation of valves and pipecracks.
Age degradation can also cause a loss of opera-tional readiness of engineered safety systems. Theengineered safety systems are designed to mitigatethe consequences of failure of a vital component,system, or physical barrier, such as a loss of mainfeedwater or a break in the primary system bound-ary. These systems are also designed to mitigate theeffects of events ranging from anticipated opera-tional transients such as loss of offsite power to lowprobability occurrences such as design basis seismicevents. Failures have occurred in systems such asthe auxiliary feedwater system and in the emer-gency diesel generators used to supply vital acpower to the IE Power System.a
Aging can also lead to a higher probability ofcommon mode failure. Aging can result in widescale degradation of a physical barrier or to simul-taneous degradation of redundant components.One example of this is a simultaneous degradationof the redundant valves designed to isolate the reac-tor coolant lines to a PWR. If this were to occur, afailure in the piping outside the containment couldlead to an uncontrolled release of the primary cool-ant and radioactivity outside of the containment.
NPAR Program Goals andStrategy
1. Identify and characterize aging and servicewear effects associated with electrical andmechanical components, interfaces, andsystems likely to impair plant safety.
2. Identify and recommend methods ofinspection, surveillance, and conditionmonitoring of electrical and mechanicalcomponents, and systems that will beeffective in detecting significant agingeffects before loss of safety function so
a. IE is the classification given for all the electrical power fornuclear plant safety systems.
1
that timely maintenance and repair orreplacement can be implemented.
3. Identify and recommend acceptable main-tenance practices that can be undertakento mitigate the effects of aging and todiminish the rate and extent of degrada-tion caused by aging and service wear.
The NPAR Program uses a two-phased approachto conduct aging research on the risk significantcomponents and systems in light water reactors; asillustrated in Figure 1. The first stage, Phase Imakes use of readily available information from:public and private data bases, vendor information,open literature, utility information, and expertopinion. The Phase I analysis includes a review ofthree elements:
1. The hardware design, operating environ-ment, and performance requirements
2. A survey of operating experience3. The current methods used for inspection,
surveillance, monitoring, and mainte-nance and for qualifying end-of-lifeperformance.
The results of the Phase I evaluation include anidentification of actual and potential failuremodes; a preliminary identification of failurecauses due to aging and service wear degradation;and a review of current inspection, surveillance andmonitoring practices, standards and guides. ThePhase I evaluation is used to decide if a Phase IIevaluation is warranted. If a Phase II evaluation isneeded, recommendations are developed to iden-tify the detailed engineering tests and analyses to beconducted in Phase II and which will result inimproved industry standards, guides, andpractices.
A Phase II assessment includes developing andvalidating advanced inspection, testing, monitor-ing and maintenance methods. This developmentincludes both laboratory and field testing to verifycandidate technologies. Phase II may also includeexamining and testing naturally aged componentsfrom operating power plants and developing serv-ice life prediction models.
With the completion of the Phase I and Phase IIaging assessment research, a technical basis will beavailable for use in the regulatory process. The keyend uses are shown in Figure 1. The uses envisionedfor the NPAR program results include: implement-ing improved inspection, surveillance, mainte-nance, and monitoring methods; modifying
present codes and standards; developing guidelinesfor plant life extension; and resolution of genericsafety issues.
Phase I Aging Assessment of aPWR High Pressure InjectionSystem
This report describes the results of an in-depthPhase I evaluation of the High Pressure InjectionSystems (HPIS) used in light water reactors.
The study was performed at the Idaho NationalEngineering Laboratory (INEL) and addresses thesystem aspects of the HPIS and the materials sus-ceptible to aging in components associated with theHPIS. Certain components, such as valves andpumps, have been extensively studied at othernational laboratories as part of the USNRC agingand equipment qualification programs.2-4 Operat-ing experience from the generic data bases andplant records on the HPIS are complemented bydata from these component studies where applica-ble. Specifically, Phase I NPAR component studieson valves and pumps are being pursued at the OakRidge National Laboratory (ORNL). The valvesare the HPIS component category with the mostfailures. Each motor operated valve and pump iscontrolled by a motor control center (MCC). TheMCC (and electric motors in general) are currentlybeing studied at the Brookhaven National Labora-tory (BNL) as part of the NPAR program. Datafrom these component studies are also used in sup-porting the system studies. In addition, the Engi-neered Safety Feature Actuating System (ESFAS)aging study performed at the INEL is directlyapplicable to the HPIS because it provides the actu-ating signal for initiation of the HPIS operationunder accident conditions.5
The strategy for this HPIS study follows theNPAR guidelines. Generic data bases are used toget statistical data on which HPIS componentshave experienced the most failures and the causesfor failures. Plant specific data supplied by a coop-erating utility includes design descriptions, draw-ings, maintenance records, and personnelinterviews. (The plant specific data, of course,applies to one plant considered to be typical forthose plants that use the HPI pumps for both nor-mal operation to supply makeup water and foremergency injection.) Information sources usedinclude: the Nuclear Power Experience (NPE) database, Licensee Event Reports (LERs), NuclearPlant Reliability Data System (NPRDS), IE
2
Phase I Phase II Utilization
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Figure 1. NPAR program strategy.
notices and bulletins, plant-design informationand specifications, operation and maintenance(O&M) manuals and procedures, historicalrecords, site-event records, and site interviews withmaintenance personnel.
The specific objectives of this study are:
1. Evaluate the overall operating experienceto determine if aging-related operationalproblems have developed.
2. Use specific examples from a representa-tive Babcock and Wilcox (B&W)pressurized water reactor (PWR) to illus-trate the functions of the HPIS and evalu-ate specific problems related to aging.
3. Perform screening type aging assessmentsof the impact of aging on operability. Theassessment will focus on identifying:
a. Failure modes and causesb. Materials susceptible to aging
degradationc. Stressors during operationd. Functional indicators that would aid
in failure predictione. Methods for detection and control of
aging degradation.
4. Review and provide recommendations forinspection, surveillance, and monitoring(IS&M), as well as advanced methods forIS&M.
5. Evaluate the role of maintenance in coun-teracting aging effects to include thefollowing:
a. Survey and evaluate currently usedmaintenance practices that counteractaging and service wear effects
b. Evaluate relative benefits of preven-tive and corrective maintenance
c. Identify potential mechanisms caus-ing component system degradationthrough improper maintenance
d. Provide recommendations for prefer-red maintenance practices.
6. Perform an unavailability assessment ofthe HPIS to determine which componentscontribute significantly to the HPISunavailability and how aging affectsunavailability.
Other work at the INEL related to this HPISaging study includes the reactor protection systemaging study,5 the reported failure cause study ofcomponent failures for selected systems, 6 and thedevelopment of technical criteria for use in assess-ing the residual life of the major light water reactorcomponents. 7 The reported failure cause workidentified safety systems significantly affected byaging phenomenon (of which HPIS is included).Although many component failures were identifiedin the reported failure cause work, actual HPISfailure occurred as a result of only 0.7% of thecomponent failures. This is due to control channelredundancy and priority maintenance.
The description of the HPIS for the representa-tive PWR is given first. This is followed by a reviewof the operating experience section, which providesinformation from the various data bases. HPISsafety issues and potential aging problems are dis-cussed next followed by aging assessments. Then areview of HPIS IS&M and the role of maintenanceis discussed. A section on the HPIS unavailabilityassessment identifying risk significant componentsis the last section just before the conclusions.
4
SYSTEM FUNCTIONAL DESCRIPTION
The HPIS along with the Low Pressure InjectionSystem (LPIS) and the Core Flooding System (CFS)collectively form the overall Emergency Core CoolingSystem (ECCS), which is designed to prevent coredamage from a loss-of-coolant accident (LOCA).High pressure injection is necessary to prevent uncover-ing of the core for small LOCAs, where high systempressure is maintained, and to delay uncovering of thecore for intermediate sized LOCAs. The HPIS can alsobe used to cool the core following a non-LOCA reactorshutdown (e.g., transient). This mode of HPIS opera-tion would be utilized only if normal and emergencysecondary heat removal via the steam generators can-not be achieved.
Commercial nuclear power plants have variousdesigns for HPIS in regard to boundaries, function,and terminology. A typical Westinghouse 4-loop plantuses accumulators [sometimes referred to as the UpperHead Injection System (UHIS)] as the immediateresponse system performing an ECCS function if thereactor coolant system (RCS) pressure drops. Whensystem injection is called for, the boron injection tank(BIT) subsystem is valved into the charging system tosupply borated water to the RCS. This BIT injection inindependent of the UHIS and is a HPIS function per-formed by the charging system to supply borated waterto the RCS. The system injection (SI) signal also startsthe two HPIS pumps and aligns both the HPIS andcharging systems to take suction from the refuelingwater storage tanks (RWST). The HPIS, sometimesreferred to as the High Head Injection System (HHIS),injects borated water to the RCS after the system pres-sure drops to 1500 psi. Both the HPIS and chargingsystem can be aligned to the residual heat removal sys-tem which takes suction from the containment sump.
Westinghouse 3-loop designs use an accumulatorsystem for an immediate borated water injection sys-tem and uses three pumps that perform the high pres-sure injection function including the BIT insertion.One of the pumps is also used for normal charging.The charging and high pressure injection function issimilar to the B&W system except that Westinghouseuses separate injection nozzles.
The Combustion Engineering designs have threeHPIS pumps used in the emergency mode only and aseparate charging system. The B&W system which isexampled in this report consists of three motor-drivenhigh pressure centrifugal pumps, with two primarysuction and discharge paths. One of the three pumps isalso used for supplying makeup water during normaloperation. The detailed system configuration is shown
in Figure 2. The HPIS and related systems perform thefollowing functions for the B&W type system:
1. Maintain the Reactor Coolant system (RCS)inventory during normal operation
2. Maintain proper RCS water chemistry andpurity
3. Control RCS boric acid concentration4. Provide fill and makeup for the core flood
tanks5. Provide seal injection water for the reactor
coolant pumps6. In the event of an RCS accident, provide
high pressure injection of borated waterfor emergency core cooling and plantshutdown
7. Provide long term core cooling following aLOCA using the high pressure recircula-tion system and low pressure recirculationsystem.
The first four items can be combined into onesystem called the makeup and purification systemto maintain the volume of the reactor coolant sys-tem (RCS) within acceptable limits during mostmodes of plant operation. It also recirculates reac-tor coolant for purification, addition of chemicalsfor the control of RCS corrosion, and the controlof soluble boron concentration for long term reac-tivity control.
For the purpose of this report, the HPIS confi-gured for the emergency injection mode will be ofprimary interest. However, parts of the system areshared for RC pump seal cooling, RC makeup andpurification, as well as the high pressure recircula-tion mode. Each of these configurations is brieflydiscussed, as necessary, to cover the functions ofthe HPIS shared components.
Makeup and Purification System
The makeup function is achieved primarily by aportion of the HPIS and the coolant storage andchemical addition systems. Makeup flow is sup-plied by either pump HPP-A or HPP-B (Figure 3)and is controlled automatically to balance normalleakage. Letdown flow from the RCS accommo-dates small increases in RCS volume due toinleakage from the seals of the reactor coolantpumps (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 inventorycontrol during most transient conditions other thanloss-of-coolant accidents. If the system is not capa-ble of meeting the requirements for inventory con-trol after a reactor trip, manual action can be takento start additional HPI pumps, establish additionaldischarge paths to the RCS, or align the boratedwater storage tank (BWST) for assurance of a suffi-cient suction source. Figure 3 highlights the ECCSmakeup and purification system. More detail sys-tem information is given in Appendix A.
Cooling System for RCP Seals
Seal injection flow is provided by the HPI pumpoperating to supply normal RCS makeup. Theseseals prevent the leakage of reactor coolant betweenthe shaft and the housing of the RCPs. When theRCS is at a high temperature, the seals must becooled to keep them from warping, to keep the sealfaces from becoming cracked or eroded, and to pre-vent the O-rings from extruding. The interruptionof cooling flow can result in seal damage, leadingto increased RCS leakage or small break LOCAconditions.
The ECCS, with the Seal Cooling System high-lighted, is shown in Figure 4. Details of the RCPseal cooling system are given in Appendix A.
Emergency Injection Mode of theHPIS
The HPIS provides emergency core cooling inthe event of a small break LOCA, and it also pro-vides an alternative means of core heat removal ifthe ability to cool via the steam generators is lost.Figure 5 illustrates the HPIS configuration. Formost event sequences, flow from one pump is suffi-cient; for special cases, two pumps may berequired. The HPIS is capable of supplying flow ata relatively high RCS pressure, with a shutoff headof 2900 psig (other types of systems have shutoffheads less than normal system pressure). The emer-gency mode of operation is initiated if the RCSpressure decreases to 1500 psig or if Reactor Build-ing pressure increases to 4 psig. Under these condi--tions, the following actions are automaticallyinitiated:
1. Three isolation valves in the purificationletdown line close (HP-3, HP-4, HP-5),and two isolation valves in the seal returnline 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 waterstorage tank outlet header open (HP-24and HP-25)
4. All high pressure injection pumps start.
The emergency high pressure injection flow pathis from the borated water storage tank through thehigh pressure injection pumps and into both reac-tor coolant loops. The emergency mode of opera-tion will continue until manually terminated.
High Pressure RecirculationMode
The High Pressure Recirculation System (HPRS)is one of two systems designed for long-term corecooling following a LOCA. The other system is theLow Pressure Recirculation System (LPRS). Afterexhaustion of the BWST, the LPRS and HPRS areused to recirculate water from the containmentsump to the RCS. If the LOCA is large enough, theRCS will be at a low enough pressure so that onlythe LPRS would be required. If the LOCA is small,however, the RCS will be at a pressure above theshutoff head of the LPRS pumps and the HPRSwould be required. For small break LOCAs, theHPRS and the LPRS are both required, becausethe HPRS takes its suction from the discharge ofthe LPRS. This realignment is shown in Figure 6.In this configuration, the LPIS takes its suctionfrom the reactor building sump through valvesLP-19 and LP-20. The discharge for the HPRS isthrough the HPIS nozzles into the reactor coolantsystem.
Any one pump is capable of providing enoughflow to prevent core damage for those smaller leaksizes that do not allow the RCS pressure to decreaserapidly enough to the point where only the LPRS isrequired. One high pressure line can deliver450 gpm at 1800 psig reactor vessel pressure. Oneof the three high pressure pumps is normally inoperation and a positive static head of waterensures that all pipe lines are filled with coolant.The high pressure lines contain thermal sleeves attheir connections into the reactor coolant pipe toprevent thermal stressing at the pipe juncture.
All three pumps have self-contained lubricationand mechanical seal coolant systems tied in withthe 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 highpressure recirculation mode is manually initi-ated by the following operator actions:
1. BWST supply line valves HP-24, HP-25,LP-21, and LP-22 are closed when theBWST low level alarm notifies the oper-ator. All high and low pressure ECCSpumps are also shut off at this point
2. Containment sump valves LP-19 andLP-20 are opened and the low pressurepumps LPPA and LPPB are restarted
3. Valves LP-15 and LP-16 are opened inorder to divert a portion of the LPRS
flow to the high pressure pumps that arealso restarted.
Appendix B contains more information on theLPIS as it relates to the HPIS.
After initiating the HPRS, the operator continues tocontrol the system in the recirculation mode. Tb aid theoperator, the following system conditions are moni-tored and displayed in the control room: the reactorbuilding sump level, the temperatures of water in theline from the sump to the low pressure pumps, the lowpressure pump discharge pressure, the flows in the lowpressure and high pressure supply lines to the reactorvessel, the level in the BWST, and all motor-operatedvalve positions.
9
INTERFACES WITH SUPPORTING SYSTEMS
The supporting systems are those systemsrequired for the HPIS to perform its function. Fail-ure in a support system can affect the operation ofHPIS.
ice water system. The LPSW system also suppliescooling flow to the heat exchangers of the compo-nent cooling system.
Instrument AirElectrical
Electric power is supplied to the three HPISpumps by three independent 4160 volt buses. Themotor-operated valves (MOVs) also require acmotive power. Emergency power is available to crit-ical components in the event that the normal powersource is lost. Control power for the HPI pumps isprovided from the dc power system. Control powerfor all other electrical components is derived fromthe same source as the motive power. The HPIScomponents and their power supplies are listed inAppendix C.
Following a loss-of-coolant accident, assuming asimultaneous loss of normal power sources, the emer-gency power source and both the LPIS and HPIS willbe in full operation within 25 seconds after actuation.All calculations for the representative plant studiedhave assumed a 25 second delay from receipt of theactuation signal to start of flow for both the HPI andLPI systems. Upon loss of normal power sourcesincluding the startup source and initiation of an engi-neered safeguards signal, the 4160 volt engineered safe-guards power line is connected to the emergency powersource. The emergency unit will start up and accelerateto full speed in 23 seconds or less. An analysis hasshown that by energizing the HPI and LPI valves(which have opening times of 14 seconds and 15 sec-onds respectively at normal bus voltage) and pumps atless than 100% voltage and frequency, the design injec-tion flow rate (HPI - 450 gpm, LPI - 3000 gpm) willbe obtained within 25 seconds.
Service Water
Cooling water for the HPI pump motors is pro-vided by the LPSW system. Backup cooling flowcan be made available by local manual action fromthe elevated storage tank of the high pressure serv-
The instrument air system supplies motive powerto the number of valves required for HPIS to func-tion in the normal makeup and RCP seal coolingmodes. Upon loss of instrument air, all the pneu-matic valves transfer to or remain in the closedposition with the exception of control valve(3HP-31), which opens fully for seal injection flowto the RCPs.
ECCS Pump Room Coolers
New plants (0-5 years old) for all four U.S. NSSSvendors have pump room coolers. Older plants(5-15 years old) may or may not have pump roomcoolers. Plants older than 15 years do not havethem.
Engineered Safety FeaturesActuating System
The HPIS is one of the Engineered Safeguards(ES) Systems that is automatically actuated by theESFAS. The aging study on ESFAS is covered inReference 5.
For normal operation, automatic control signalsare supplied for two flow control valves. They arethe normal makeup flow control valve, HP-120,and valve HP-31 for RC pump seal flow control (seeFigure 2).
During emergency conditions ES signals are pro-vided to components in the HPIS that must changestate. The components required for function of theHPIS receive signals from the ESFAS when theRCS pressure is low (1500 psig), or when the reac-tor building pressure increases to 4 psig. TheES-actuated HPIS components as well as the ESchannel doing the actuation are listed in Table 1.Appendix D describes the ESFAS system for theHPIS 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 ActuationChannel
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 aremade of stainless steel which is resistant to corro-sion from boric acid used in the borated water. Themajor components are discussed in this section.More detailed aging related design information onthese components is given in Appendix E.
Valves
For the purposes of this report, a valve is definedas the valve body and all its internal parts, the valveoperator (motor, solenoid, hand wheel, etc.), andany limit and torque switches mounted on the valvebody or operator needed to make the valve func-tion. The HPIS uses many types of valves and valveoperators including motor-operated valves (MOV),pneumatic-operated valves, solenoid-operatedvalves, manual-operated valves, check valves, andsafety relief valves. The component NPAR studieshave extensively covered aging of valves (see Refer-ences 2 and 3).
Valve failures and how they affect system opera-tions is an important consideration in this research.The failure modes that can affect system operationinclude failure to open, failure to close, internalleakage, external leakage, and plugged. Generalfailure mechanisms that exist independent of valvetype include normal wear, excessive wear, corro-sion, foreign material contamination, and excessivevibration. Control valves such as the HPI flow con-trol valves can also have a failure mode in whichthey fail to operate as required. They are designedto constantly change position during operation. Asystem failure occurs when a valve has lost the abil-ity to control system parameters.
Types of valves are discussed further inAppendix E and valve failures are covered in a latersection on Operating Experience.
Instrumentation and Control
In general, there are three variations of the engi-neered safeguards control stations as shown in Fig-ure 7. The first station is a controller and monitorfor a motor-operated valve automatically operatedby an ESFAS signal. Feedback information on thevalve position is provided by a limit switch. Thesecond station is a pump motor control station thathas inputs from either the ESFAS or main controlroom. Instrumentation for the HPI pumps
includes local indication of pump discharge pres-sure and suction pressure. Discharge flow throughthe pump crossover lines, discharge pressure, andlow pressure are also provided to the control room.The third type of control station is one that controlsan air operated pilot valve to control the pneumaticvalve. Two stations are shown on this drawingmeaning that the valve can be controlled from twodifferent locations. The "C" in the triangles onthese drawings indicates computer monitoring forcontrol room display.
The operator can override the automatic flowcontrol on HPI by adjustment of the flow control-ler so the flow rate may be reduced to match the lossof coolant from the vessel when small line breaksoccur.
HPI Pump A Controls. The HPI pump A is nor-mally controlled from the control room with amanual four position (start-run-off-auto) switch.In the start position, power is supplied to the pumpcircuit breaker closing coil and pump starts. In autoposition, the pump automatically starts on low sealinjection flow, or loss of voltage on the main feederbus as sensed by the main feeder Bus Monitor.When in the off position, the control switch ener-gizes the circuit breaker trip coil and the pumpmotor is de-energized.
In the event of a reactor accident, the ESFASchannel I will automatically start HP injectionpump A regardless of its manual control switchlocation. If the control switch was in the off posi-tion when the start occurred, the pump will con-tinue to run after the ESFAS signal is reset until themanual pump control switch is cycled out of the offposition and back.
After an ESFAS signal is initiated, the automaticcontrol can be overridden by pressing the manualpushbutton in the control room. This override canbe done only when a safeguards trip signal, or testsignal is present. When in the manual mode, thepump control then functions as if no ESFAS signalis present. Control is returned to the automaticmode for safeguards when the auto pushbutton isoperated in the control room or if the ESFAS tripsignal is cleared at the safeguards cabinet.
HPI Pump B Controls. The controls for pump Bare essentially the same as those provided for pump Aexcept that in the event of an accident it receives anautomatic start signal from both ESFAS channels 1
12
ES command
!Engineered 4Safeguards n CControl Station Auto &
It ~~~~~~~~~~~~~~Contro ( l
l ~~~~~control ndica ton-
Us~~~~~Nra Control)l ~~~~~controlIndication 4
_Safeguards A, cond _4 sol|Control Station Auto coto im
l Indication I Indicallon IIl Indlcatlon I4 Inialo
I_____
7*1904
Figure 7. Representative ESE actuated systems.
13
and 2. In order to take manual control of pump Bwhen an ESFAS signal is present, both channels Iand 2 must be put into the manual mode.
HPI Pump C Controls. The controls for HPIpump C are essentially the same as for pump Aexcept that pump C is not used for normal opera-tion. In the event of a reactor accident, HPIpump C receives an automatic start signal fromESFAS channel 2.
Motor-Operated Valves. The ESFAS providesautomatic operation for the valves listed in Table 1.After an ESFAS signal is initiated, the automatic con-trol signal can be overridden by manual pushbutton.While in the manual mode the valve is controlled as ifno ESFAS signal is present. Control is returned to theautomatic mode for safeguards when the auto push-button is operated or when the ESFAS trip signal iscleared at the safeguards cabinet. Manual controls areprovided in the control room and safeguards cabinet.
Pneumatic-Operated Valves (HP-5 and HP-21). The pneumatic-operated valves are controlled bya 125 Vac solenoid operated valve that controls the airline operating the pneumatic valve. Manual control isprovided in the control room and automatic operationis provided by ESFAS channel 2. A manual overridesimilar to the MOV and HPI pump controls is pro-vided. Interlocks in the control circuit will close HP-5in the event of an excessively high letdown temperatureor low instrument air pressure. Similarly, valve HP-21is interlocked to close on low instrument air or if allfour of the individual RC pump seal return valves areclosed.
HPI Pumps
The HPI pumps are vertical multiple-stage-centrifugal pumps with mechanical seals. The wettedparts of the pumps are stainless steel. Figure 8 is a pic-ture of a HPI pump. Operation of the HPI pumpsrequires lubricating oil cooling and pump seal cooling.The HPI pump cooling is accomplished via the LPSWsystems where heat generated in the pump lubricatingoil and seals is removed via heat exchangers. The HPIpump data is given in Appendix E.
than those encountered during emergency opera-tion. Pipe sizes range from I to 14 in. Design pres-sures and temperatures for piping are given inAppendix E.
Nozzles and Thermal Sleeves
The high pressure injection/makeup (HPI/MU)nozzles with thermal sleeves are located on all four coldlegs of the reactor coolant piping. They provide emer-gency core cooling and normal makeup flow to theprimary coolant system. In general, one or two of thelines are used for both HPI and MU, while the remain-ing nozzles are used for HPI alone. The thermal sleevesare incorporated into the nozzle assembly to provide athermal barrier between the cold HPI/MU fluid andthe hot HPIS nozzle. This prevents thermal shock andfatigue of the nozzle. See Appendix F for a discussionof nozzle cracking.
Piping Penetrations
The reactor building penetrations for HPI linesassociated with the representative plant studiedinclude the following: the letdown from the reac-tor coolant system to the demineralizers is a 2.5 in.line with remotely controlled valves on the insideand outside of containment for isolation control;two (1 in.) nozzle warming lines with check valvesfor isolation control; a normal makeup inlet line(4 in.) having a check valve on the inside of con-tainment; two seal injection lines that supply RCpump seal cooling (2.5 in.) with a check valveinside of containment; and the emergency injectionline (4 in.) with a check valve on inside of contain-ment. The penetrations must be able to maintainreactor building pressure seal. Reactor coolant iso-lation under accident conditions is maintained byvalves on the inside and outside of thepenetrations.
Pipe Hangers
Pipe hangers are rigid carbon steel supports forthe various HPI piping. Their purpose is to resistthe dead weight loads of the piping and water.
SnubbersHPI Piping
The HPI piping is stainless steel and designed fornormal operation. The normal operating systemtemperature and pressure requirements are greater
Snubbers move freely at low acceleration andlock up at higher acceleration, to provide supportfor seismic and other dynamic loads. They may bemechanical or hydraulic.
14
TanksThe borated water storage tank is the source of water
for the emergency HPIS. It is a 350,000 gallon coatedcarbon steel tank. For the representative plant studied
it is considered part of the low pressure injection sys-tem because it also supplies borated water for LPISand containment spray systems.
Figure 8. HPI pump.
15
OPERATING EXPERIENCE
The HPIS operating experience is based oninformation from generic data bases, plant recordsfrom the representative PWR studied, and othersources such as USNRC information notices andbulletins for specific problems. Information wastaken from the NPE, LER, and NPRDS genericdata bases. These data bases were adequate fordrawing general conclusions about the ECCS withemphasis on the HPIS. The system boundariesused were those indentified by these data bases.The discussion of each data base identifies thecomponents or the subsystem covered by that database. Although the components and subsystemsthat support the HPIS function are generallyincluded as part of the HPIS, actual high pressureexists only in that part of the system from the HPIpumps to the injection nozzles. All the componentsthat are included in the HPIS are important to itsservice and therefore, are included in this study.
Nuclear Power Experience Data
The NPE automated retrieval system was devel-oped and introduced by the S. M. Stoller Corpora-tion at Boulder, Colorado. This system containsinformation on nuclear plant components availablefrom the public domain. The index and key wordsare computerized, allowing a rapid search of thesystem for specific articles with titles and referencenumbers to hard copy volumes. The system isupdated quarterly and is a convenient source forobtaining generic information on problem areas.The NPE data is summarized in this section. Moredetailed data from NPE on the various HPI andassociated ECCS components are given in Appen-dix G.- The NPE data base includes the following ECCS
subsystems: safety injection, high head injection,upper head injection, boron injection, recircula-tion phase, containment spray, and accumulatortank. All the above systems are part of the HPISexcept containment spray. The containment spraysystem could not be conveniently removed, thus, toindicate the data summaries from the NPE alsoinclude containment spray, they are called ECCSsummaries. The components in these subsystemsare similar in that they have similar operating envi-ronments and must be compatible with boratedwater.
For U.S. Nuclear Power Plants, from startupthrough 1986, there were 1552 articles on the ECCS
for PWRs. Component failure events listed in orderof frequency of occurrence for these PWRs is givenin Table 2. Valve failures are listed most frequently(35°%), followed by I&C (19%), pumps (15°%), andpipes (7%). Pipes in this case include nozzles andpenetrations.
The ECCS failure causes are listed in Table 3.The top three causes in order of frequency of occur-rence were maintenance error, design error, andmechanical disability. This was followed with. 4°0%each for local I&C, set point drift, chemistry out ofspec, and subcomponent sticking. The failurecauses that are considered potentially aging relatedand identified in Table 3 account for 28% of thefailures. The cause is identified only as potentiallyaging related because the root cause can not alwaysdetermined from the reported events.
After eliminating human and procedure errors, abreakdown of subcomponents causing ECCS fail-ure is given in Table 4 for PWRs. Moving internalparts for valves, pumps, and motors accounted for15°10 of the problems. Just a little over half of thesewere caused by valve stem or disc seating problems.Instrumentation accounted for 12% of failures andcontrol 11°0% of failures. The drive sources for mostof the valves and pumps are electric motors whichaccounted for 66% of drive/actuator failures.
Table 2. ECCS component failure ranking(NPE)
PWR(%0)Component
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)
PWRCause .(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 seatingBlade/impellerCoupling shaftBearing/bushingPiston/diaphragm
Instrumentation
Percent (a) Subcomponent Percent (a)
15 Drive/Actuator 8
(52)(9)
(16)(14)( 9)
12
Pneumatic airHydraulicMotorOther
Chemistry
Relay/breakersWire cableConnectorsSeal/gasketFitting/flange
(10)(3)
(66)(21)
BistablesTransmittersSense linesIndicator
(33)(25)(25)(17)
7
65
554
Control
Limit SWSolenoid valve
Torque SWOther
11 External Support
(14)(1 1)(23)(52)
Mounting/fastenerBody/casingOther misc.
4
4
212
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 50.72for occurrences before 1984 and 10 CFR 50.73 forevents after January 1, 1984) require nuclearpower plants to report significant events to theUSNRC. The pre-1984 LER data base has beenused as a source of reliability data. Events reportedto the LER system after January 1, 1984 are onlythose that are, or lead to, safety-significant events.If a component fails and can be replaced within thetime constraint of the limiting condition for opera-tion (LCO), no LER is required. This limitedreporting would not be expected to provide an ade-quate representation of failure experience; there-fore, only the LERs prior to 1984 were used in thisaging study. Data from licensee event reports onHPI pumps and valves from LERs are presented inthis section. The LER data covers HPI and CVCSvalves and pumps for B&W, W, and CE plants.
Pump Failures. A total of 44 events associatedwith HPI pumps were reported in the LERs for allPWRs from January 1, 1972 to September 30,1980.8 For the HPI pumps, the leading failurecause was control malfunction with 14 failureevents out of a total of 44. The next most frequentcauses were maintenance and design error with fourevents each. This is followed by three events eachfor operation error, failed internals, and unknown.These HPI pump failure causes are summarized inTable 5.
The failure mode experienced most often wasdoes not start for 24 of the 44 events. This was fol-lowed by does not continue to run (8 events) andloss of function (7 events). Most of the problems(25) were discovered by performance testing.TWelve were discovered by normal operations (SeeTable 6).
System Pumps. The system pump category cov-ers all HPIS/CVCS pumps apart from the mainHPI pumps, (i.e., charging pumps, makeuppumps, etc.). There were 130 events reported wherethe leading cause was seal or packing failure (29),followed by control malfunction (13), loss of pres-sure boundary (12), and drive train failure (9).Table 7 ranks these causes by number of events.
Valves. A summary of HPIS valve failure fromLER data for 1976 through 1980 is given inTable 8.9 About 42% of the valve failures wereclassified as potentially aging related. In this cate-gory, mechanical controls (parts failed or out ofadjustment) were the most frequent with 8%, fol-lowed by seat or disk failure 6.9%, and packingfailure 5%. It is interesting to note that commandfaults (faults due to power source, controls or sup-porting systems) caused 32% of HPIS valve fail-ures. See Appendix G for additional LER data onvalves.
Nuclear Plant Reliability DataSystem
The NPRDS was developed by the EquipmentAvailability Task Force of the Edison Electric Insti-tute (EEI) in the early 1970s under the direction ofthe American National Standards Institute(ANSI). The NPRDS was maintained by the South-west Research Institute under contract to the EEIthrough 1981. Since January 1982, the NPRDShas been under the direction of the Institute ofNuclear Power Operation. The components cov-ered in the NPRDS for HPIS are announciators,circuit breakers, safety function instruments,motors, pumps, valves, and valve operators. ForWestinghouse plants upper head injection, theBoron Injection Tank along with associated valvesis also included. For B&W plants, filters areincluded as well as letdown, purification, and alsoCVCS components common to the CVCS andHPIS systems.
The NPRDS data for all Westinghouse andBabcock and Wilcox plants were compiled for theHPIS and the aging fraction determined. Combus-tion Engineering data was unavailable fromNPRDS at the time these data were compiled.Aging fraction is the ratio of aging-related failuresto total number of failures. The B&W systemsincluded in this sort were the letdown purificationand makeup systems and high pressure injectionsystem. The Westinghouse systems included werethe High Pressure Injection System and UpperHead Injection Subsystem. The results of this sortare shown in Table 9. The overall aging fractionsfor 1036 failures is 0.213. This means that 21.3%of the failures were aging related. The HPIS com-ponents ordered by aging fractions for categories ofdesign, aging, testing, and human is shown inTable 10. Valves caused the most failures, followedby valve operators and instrumentation. These
19
Table 5. HPI pump failure causes (LER)
Numberof
EventsItem
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 77qoof HPI component failures in HPIS.
The NPRDS system effect code identifies theeffect on the system caused by the componentfailure. The codes were taken directly from theNPRDS failure records. The NPRDS has five sys-tem effect categories and are defined as follows:
* Loss of System Function-A componentfailure that, singularly results in the systembeing unable to perform its intendedfunction (i.e., all trains, channels, etc.,inoperable).
* Degraded System Operation-The systemis capable of fulfilling its intended func-tion, but some feature of the system isimpaired.
* Loss of Redundancy-Loss of one systemfunctional path.
* Loss of Subsystem/Channel-A partialloss of system functional path.
* System Function Unaffected-Failure didnot affect the operation of the system.
The fractions for each system effect are shown inTable 9. Approximately 57% of the failures lead tosystem degradation, but only 0.7% caused a loss ofsystem functions.
The motor-operated valve data from the nuclearplant reliability data system for Westinghouseplants had enough events (56) to show an agingtrend (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 maintenanceB Does not start N During normal operationsC Loss of function R During records reviewD Does not continue to run T During testingU Unknown U Unknown
(b) Type of Event
Code . (d) Event Classification
Failure Command Description Code Description
S Nonrecurring, not common cause D DemandR T Recurring, not common cause T Time (continuous operation)C U Nonlethal common cause U Unknown
V Recurring nonlethal common causeN Lethal common cause
W
Table 7. Cause for all system pump failures other than main HPI pumps (LERs)
Numberof
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 122Aging failures 221Test and maintenance failures 83Human related failures 27Other failures 583
Total 1036
Failure Category Fractions
Design fraction 0.118Aging fraction 0.213Test and maintenance fraction 0.080Human related fraction 0.026Other fraction 0.563
System Effect Totals
Loss of system function 7Degraded system operation 197Loss of redundancy 138Loss of subsystem/channel 251System function unaffected 443
Total 1036
System Effect Fractions
Loss of system function fraction 0.007Degraded system operation fraction 0.190Loss of redundancy fraction 0.133Loss of subsystem/channel fraction 0.242System 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 - - 0.722recorder
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 0.702transmitter
Heater 36 0.111 0.111 0.111 0.167 0.500
Instrumentation: 19 0.158 0.105 - - 0.737controller
Instrumentation: 153 0.124 0.039 0.007 - 0.830switch
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 - - 0.833computation module
a. Components ordered by aging fractions.
25
'u
161
co,
a)
Ez3
12 i
a Total number of failures-. Aging caused failures
8
4
i I - - - - - - - - - - - - - - - - - - --
5 10 15 Years in service9-6104
Figure 9. Failure data for HPIS motor-operated valves in 5-year increments (data from nuclear plantreliability data system).
in five-year increments. This also appears inTable 11 along with data for check valves, manualvalves, and HPIS pumps. The check valves andmanual valves showed no trend on aging failures.The HPIS pumps had a significant increase in thenumber of failures in the 10- to 14.9-year periodfor both aging and other causes.
Plant Operating Experience fromSite Visit and PersonnelInterviews
An operating B&W plant was visited and site per-sonnel interviewed. Included in the interviews werethe HPI maintenance supervisor, I&C supervisors,and an electrical specialist. Detailed I&C drawingson HPIS were reviewed with plant personnel, aswell as computer printouts of corrective mainte-nance (CM) requests, incident investigation reports(IIR), and test procedures for HPIS. All mainte-nance work is initiated with a maintenance workrequest and when the work is finished, a descrip-tion of the problem and corrective action is written
on the request sheet. These CM requests are thenfiled for record purposes.
Nuclear Maintenance Data Base. CorrectiveMaintenance summaries were taken from theNuclear Maintenance Data Base. This plant datasystem summarizes all CM reports as a one-linesummary, work required reference number, anddate. Any channel or component found deficientduring implementation of calibration and testingprocedures would have a CM request written to cor-rect the problem.
The CM request records were computerized bythe utility starting October 23, 1980 and these werethe ones reviewed. Microfilm records for CMbefore October 23, 1980, were not searchedbecause of manhour and cost limitations.
The computerized maintenance records werereviewed and show that many potential problemsare fixed before major system or channel failuresoccur. Thus, maintenance records reflect the incipi-ent failures to some extent.
Data from the maintenance records are shown inTable 12. These data cover a period from May 1980
26
Table 11. Component failures in 5-year incrementsa
Number of Failures
Component
Motor-operatedvalve
Cv
Manualvalves
Pump(centrifugal)< 500 gpm
FailureClassification
OtherAgingTotal
OtherAgingTotal
OtherAgingTotal
OtherAgingTotal
04.9Years
163
19
729
314
448
5-9.9Years
87
15
4l5
347
74
11
10-14.9Years Total
51217
292251
92
11
205
25
325
97
16
1415
29
252348
a. Data from the nuclear plant reliability data system; data for older Westinghouse plants.
to December 1986 and contain 356 records. Ofthese, 171 were routine maintenance items such asrecorder paper problems, cleaning filter, etc.Instrumentation and Control were the largest cate-gory with 74 requests. Pipe-related maintenanceitems were the next largest category with55 requests. They include flange leaks, hangers,pipe, penetration, and snubbers. These requestswere followed by valves with 25 requests, controlcircuits with 21 requests, and pumps with18 requests.
Incident Investigation Reports. Nonroutineevents in the plant (including those that occurredduring the precommercial operations phase) areevaluated. This evaluation may result in an incidentinvestigation report (IIR) that captures the impor-tant details related to the event through interviews,analysis of logs, recorder strip charts, computerprintouts, etc. IlRs are company proprietary andcover reportable events such as techical specifica-tion violations. The event may or may not requireNRC notification. Hence, the LERs for the stationare a subset of the IlRs. The lIRs were reviewedthrough April 1985.
Some observations based on sorts of the IIR database are:
1. Better resolution of information in thecoded fields than LERs and NPRDS fail-ure reports
2. Precommercial operation events captured3. Report event frequency relatively constant
up to 19844. Coded reporting for LER tracking5. Infant mortality observable in component
failure searches.
A summary of IIR causes for HPIS failures arepresented in Table 13. Procedure or personnel errorand valve failure each accounted for 28% of theincidents, followed by I&C (17%), pumps (11076),nozzle (6%o), and welds (5%). However, bacausethere were only 17 events covering the HPIS overan 11-year period, the average for reportable eventswas relatively low at about 1.5 per year -for theplant studied. Estimates indicate that about half ofthese events are potentially age related.
Summary of OperatingExperience
The information from the data bases indicatethat approximately 57%o of the failures lead to
27
Table 12. HPI plant data
Number of Sub breakdownItem 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 HPISfailures
LER have frequency of occurrence similar to theNPE.
A summary of problems with the HPI systemcomponents is given next for each type ofcomponent.Type of Event Percent
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 significantproblems with piping cracks in HPI systems forPWRs. However, in March of 1982, cracks wereidentified in the thermal sleeve and safe end where amakeup/high pressure injection line joins the reac-tor coolant system at one B&W unit. 1 0 Subsequentinvestigations of these lines at four other B&Wunits revealed similar cracks. The apparent cause ofthe cracking was thermal fatigue. The injectionlines for normal reactor coolant makeup are alsopart of the high pressure injection system at B&Wplants. Because these plants do not have regenera-tive heat exchangers in the coolant makeup circuit,a potential existed for makeup temperatures to besubstantially lower than the reactor coolant tem-perature. Temperature variations due to mixing inthe high pressure system nozzle, coupled withhydraulic effects, were suspected to be the principlecause of failure.
Beginning in June, 1982, cracks were also identi-fied in the thermal sleeve to nozzle connections ofreactor coolant system branch pipes at severalWestinghouse units. The cracking was discovered inthe thermal sleeve retainer welds that attach thesleeve to the nozzle inlet. These cracking problemsincluded:
1. Accumulator lines, pressurizer surge line,and charging line nozzles at one unit
2. Accumulator line nozzle at one unit3. Safety injection and charging line nozzles
at one unit
system degradation, but only 0.7 !70 of these failurescaused a loss of system function. The data basesalso showed that 21.3 0o of these failures were agingrelated.
The data bases are in agreement on the four mosttroublesome 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&Cproblems, then pipe related problems, and finally,valves and pumps. These requests indicate moreminor problems with I&C and pipe hanger adjust-ments etc., but the major reportable problems rankvalves 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 injectioncold leg injection line, and the chargingline nozzles at one unit.
The failure mechanism was suspected to befatigue induced by thermal cycling.
Most of the recent stainless steel pipe crack prob-lems at PWRs have been fatigue related rather thancorrosion related. In 1979, stress corrosion crack-ing was discovered in some safety system pipes con-taining stagnant borated water. ll No losses due tothis problem were reported in 1980-1986.
Cracks have also occurred due to vibration anddynamic loading (water hammer). Welds andflanges are connection stress points. Flange loosen-ing accounted for 9%o of all pipe problems in NPEdata.
Snubbers. Snubbers perform a safety functionby restraining the motion of attached systems orcomponents under rapidly applied load conditionsof earthquakes, pipe breaks, or severe hydraulictransients. LERs relating to malfunction of snub-bers indicated the most frequent problem was sealleaks in hydraulic snubbers. Mechanical snubberswere subject to damage due to vibration. A phase INPAR study has been conducted on snubbers. 12
Penetrations. At least five piping penetrationsare 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 oper-ating plants (71 at time of estimate). In that esti-mate, it was assumed that each plant had40 penetrations. If five penetrations are associatedwith HPIS, then about every eight years a problemcould be expected with an HPIS penetration in oneof the 71 operating plants.
Tanks. Few tank problems have occurred. Mostof the problems attributed to tanks in data basesinvolved boron concentrations or fluid levels thatwere out of specification. The only significantevent was to replace a boric acid injection tank at aPWR.
Pumps. Most high pressure injection pump fail-ures that are aging related involved seal, bearing,and shaft problems.
Valves. The type of component most oftenresponsible for HPI failure was valves. Thisincludes all types of valves and valve operators.Personnel error, operation, and maintenance werethe primary causes overall. When these causes wereremoved, mechanical parts failure, packing leaks,and seat or disk failure were the most often men-tioned causes related to aging and service wear.
Chemistry Problems. Corrosion in pipe weldheat zones due to contaminants and boric acid cor-rosion of ferritic metals are two failure causesrelated to chemistry and aging. Controlling boricacid concentrations is a safety related operationalproblem. See Appendix H for problems withborated water systems.
Microbial Influenced Corrosion. Many systemsin the majority of nuclear plants appear to be sus-ceptible to some form of microbial influenced cor-rosion (MIC). This is particularly true of standbysystems where conditions exist for microbialgrowths. Stainless steel 304, 316, etc., have beenaffected by MIC as well as other metals. One utilityexperienced MIC in the HPI and LPI pump impel-ler blades after initial tests followed by a period ofstandby. Microorganisms have also been found in aborated water storage tank, but no degradation wasreported.
30
HPI SAFETY ISSUES AND POTENTIAL AGING PROBLEMS
The following is a review of system and personnelinteraction safety issues13 related to HPI and aging.
Locking Out of HPISPower-Operated Valves
The NRC staff positions BTP EISCB 18 and BTPRSI3 6-11 require the physical locking out of electricalsources to specific motor-operated valves in the ECCSincluding HPIS and LPIS. This method protectsagainst a single failure causing an undesirable compo-nent action. This assumption in the safety evaluation isthat the component is then equivalent to a similar com-ponent that is not designed for electrical operation andcan only be opened or closed by direct manual opera-tion of the valve. Thus, no single failure (due to anycause including aging) can both restore power to thecomponent and cause mechanical motion of the com-ponent. The probability of failures due to maintenanceerrors, electrical faults, and mechanical failures(7 x 1f07) was greater than the probability of the valvemispositioning coincident with a LOCA, and as aresult of operator error (4 x 10-7). This was issue num-ber B-8 in Reference 13, but was dropped as a safetyissue because it is an acceptable approach to meet thesingle failure design criteria. However, this points out adesign shortcoming when human interaction isrequired for initiating HPIS valve action.
Inadvertent Actuation of SafetyInjection in PWRs
Westinghouse and B&W plants had a high rate ofspurious or inadvertent safety injections occurring. Inthe case of B&W reactors, the practice was to manuallyturn on one or more HPI pumps after a reactor scramto recover the pressurizer level. This practice contrib-uted to the high rate of safety injections for theseplants. This practice was stopped after the accident atthe Three Mile Island plant because it was determinedthe HPI pumps were not needed to maintain pressur-izer level. As a result, unneeded SI in B&W plants isnow significantly lower. A possible reason forunneeded SIs in Westinghouse plants is their designrequires more signals for initiating SI and thus morechances for spurious signals. Actuation of the SI isundesirable because it injects cold borated water intothe reactor, thereby subjecting injection nozzles to ther-mal stresses and requiring removal of boron from theprimary system before startup. Actuation was deter-
mined as not only an economic issue, but could possi-bly lead to a wrong response by an operator when SI isreally needed. Operator response is carried as anotherissue and would include this one. This was issue 8, Ref-erence 13. Inadvertent SI actuation due to this operat-ing procedure was corrected by eliminating' theprocedure. However, maintenance personnel error andother operator procedures that could inadvertently ini-tiate SI should continue to be reduced through design,training, and applicable updating procedures.
Switch from HPI Mode toRecirculation Mode
The switchover from the HPI mode of operation tothe recirculation mode for accident recovery requiresrealignment of a number of valves. The switchover canbe achieved by a number of manual actions, by auto-matic actions, or a combination of both. The threeswitchover options (manual, automatic, and semiauto-matic) are vulnerable with varying degrees to humanerrors, hardware failure, and common cause failures.Automatic system actuations reduce the impact of oper-ator error in completing the switchover, but are subjectto spurious actuations. Spurious switchover of HPI toHPR has the potential for pump damage as well asunacceptable safety consequences. This safety issue(No. 24) was scheduled for prioritization(Reference 13).
High Pressure RecirculationSystem Failure Due ToContainment Debris
In the recirculation mode the HPIS pumps take suc-tion from the LPIS, which is taking suction from thecontainment sump. Any debris, paint flakes or loosematerial due to aging could potentially damage systemcomponents during the HPRS mode of operation.This is safety issue 28 (Reference 13), which has nowbeen scheduled for prioritization.
Failure of Demineralizer Systemand the Effect on HPIS
While the demineralizer system does not directlyperform a safety function, a failure of the demineral-izer system could impair the operation of the HPIS. Inthe plant studied, the demineralizer system has key
31
components labeled as part of the HPIS to ensureprompt attention if failure occurs. In Reference 13,this problem is listed as issue 71 and is scheduled forprioritization. This is also true of other systems and isaddressed in the next subsection on SystemsInteraction.
Systems Interaction
The HPIS interfaces with many other plantsystems. They include IE power, LPIS, RCS,service water systems, containment spray sys-
tem, monitoring, and ESFAS. In addition, partof the HPIS is used for both normal operationand emergency operation. There are also HPIsubsystems such as demineralizer, makeup/letdown system, chemical control, boron injec-tion, lubrication of HPI pumps, pump sealcooling, and instrument air system. Systeminteractions for HPI may be one study recom-mended for the Phase 2 aging study. Overallplant system interactions is also a Safety Issue(A-17, Reference 13). In addition, the perform-ance of the HPIS equipment can be affected bythe heating, ventilation, and air conditioningsystem.
32
AGING ASSESSMENTS FOR THE HPIS
The third objective of the HPIS aging study was toperform screening type assessments of the impact ofaging on operability of the HPIS. Subobjective (a),failure modes and causes, has already been covered inthe discussion of operating experience. Based on thereview of operating experience from generic data bases,safety issues, and plant data the remaining four subob-jectives of objective 3 can be discussed. Theyinclude: (b) identification of materials susceptible toaging degradation, (c) stressors during operation,(d) functional indicators that would aid in failure pre-diction, and (e) methods for detection and control ofaging degradation.
Preliminary Identification ofSusceptibility of Materials toAging
The HPIS piping, pumps, and valves are allstainless steel which is compatible with boric acid.The BWST is carbon steel with a liner to protectagainst corrosion. Problems have occurred whenleaks in connections or valves allowed boratedwater to come in contact with carbon steel compo-nents. Water evaporation results in a concentrationof boric acid causing corrosion on carbon steelcomponents.
The narrative descriptions in the NPE data basewere reviewed to determine if failure occurred as aresult of conditions specific to the HPIS. Severalfailures were reported that resulted from the charg-ing of cold water into a hot system and from thehandling of the water with a high boron concentra-tion. The four most significant failures are dis-cussed below.
High-pressure injection/makeup nozzles havedeveloped through wall cracks. The cracks resultedfrom thermal fatigue. The thermal fatigue wascaused by turbulent mixing of hot and cold coolantand thermal shock of the hot safe-end wall duringnormal makeup. All cracks were associated withloose thermal sleeves. Improved thermal sleevedesign and increase in minimum continuousmakeup flow to prevent thermal stratification havebeen employed for failure mitigation.
Elbows in the safety injection piping between thecold leg and the first check valve have developedthrough wall cracks. The cracks occurred in theheat-effected zone of the elbow weld, to the safetyinjection piping, and in the base metal of the
elbow. The cracks resulted from high-cycle thermalfatigue caused by cold makeup water leakingthrough a closed globe valve at a pressure sufficientto open the check valve. Mitigation methods arebeing evaluated. Installation of a globe valve down-stream of the check valve, rather than the existingvalve, is being considered to isolate the injectionline during normal makeup.
Motor-operated valves and check valves havefailed to operate due to boron crystallization on thevalve stems, in the valve packing, and in the valvebody. The reason for crystallization was not alwaysreported in the NPE, and investigations may nothave identified the cause. However, most causeswere reported as packing leaks. The valves wereusually cleaned and placed back in service. Oneincident reported additional heat trace was addedto prevent future failures.
Injection boron concentration has been dilutedfrom leaking valves. Leaks have been reported forboth check and globe valves. Dilution of the boroninjection tanks for Westinghouse plants and safetyinjection tanks for combustion engineering plantshave been reported. A few dilutions have beenreported for the borated water storage tanks of theBabcock and Wilcox plants. Improved monitoringof the tanks and repair of the valve seats have beenimplemented as mitigation measures.
A microbe-caused-problem with stainless steeloccurred in a plant when preliminary tests wereconducted with water in the HPIS and allowed tostand before startup. During this standby periodmicroorganisms caused corrosion on the pumpimpeller blades.
I&C and electronic components have been sus-ceptible to catastrophic failure. Contact wear inswitches and relays as well as corrosion on contactsare common aging effects on electrical compo-nents. Electrical insulation ages with thermalcycling.
Stressors for HPIS
The HPIS is subject to many stressors duringoperation. They include stressors due to mainte-nance, operation, and testing; environmental stres-sors; electrical stressors on I&C; and mechanicalstressors. Various types of stressors are identified inthe following sections.
33
Maintenance, Operations, and Testing Stres-sors. Included in the maintenance, operations,and testing stressors are:
1. Personnel error-resulting in inadvertentSI actuation. If valves are not orientedproperly, or pump suction unavailable,damage to components could result.
2. Water hammer (dynamic loading)-incidents have been attributed to pumpstartup with partially empty lines andrapid valve motion. Most damage has beenrelatively minor and involves pipe hangersand restraints.
3. Thermal cycling-due to cold water fromthe HPI system into the hot RC system.Also reactor heatup and cooldown causesthermal cycling.
Environmental Stressors. Included in the envi-ronmental stressors are:
1. Temperature or pressures-are environ-mental stressors
2. Water chemistry incorrect-impurities leftfrom welding and boric acid crystals areexamples of environmental stressors
3. Vibration-flow-induced vibration mayhave been a contributing factor in the ther-mal sleeve cracking in PWRs. Pumps canbe a source of vibration if unbalanced.Many other potential sources of vibrationexist in a nuclear plant. If the natural fre-quency of vibration of the connected pip-ing is very nearly the same as the drivingfrequency of the pump, then there is thepossibility of fatigue failures in the system,particularly at the nozzles where the stresswill be highest. Vibration is detected onmajor rotating equipment by instrumenta-tion, inservice inspection, or other visualmeans. Vibration problems, however, haveto be resolved on a case by case basis
4. Seismic stresses-may cause relay chatteror fastener damage. Seismic stresses, how-ever, are not due to recurrent conditionsdue to operations.
Electrical Stressors. The electrical stressorsfrom switching transients and loading include:
1. Transients affecting I&C-transients canoccur from HPIS operation, external elec-trical faults, and lightning.
2. Low voltage affecting I&C-abnormalvoltage can occur from excessive loading,power supply drift, and set pointadjustments.
Mechanical Stressors. Some mechanical stres-sors are as follows:
1. Valve misadjustment-limit switches, andtorque switches out of adjustment, andmechanical adjustments can causemechanical stress
2. Pipe alignment-any misalignment of pip-ing can stress welds, connection, andflanges
3. Vibration-caused by motors and pumps.
Functional Indicators that wouldAid in Failure Prediction
Functional indicators include abnormal currentsand voltages for I&C equipment and indicate achange from the normal expected values. Leaks inpipes, flanges, and nozzles indicate problems suchas loose fasteners, cracked pipe, or corrosion. Vis-ual indicators could include limit switch setting,boric acid crystals, or shaft wear. Unusual vibra-tion or noise associated with motors, valves, andpumps could also be an indication of impendingfailure.
Methods of Detection andControl of Aging Degradation
Any of the failure detection methods also applyto aging detection because aging is one of the mech-anisms causing failure. Functional indicatorsobserved during normal operations or testing is onemeans of detecting degradation. This detectionincludes the normal control room monitoring ofgauges, charts, and computer printouts. For spe-cial problems or studies, additional condition mon-itoring such as vibration sensors, noise monitoring,or current and voltage signatures may be applied.During refueling outages, end-to-end operationalchecks and functional testing results are comparedto previous baseline measurements. Any changefrom the baseline should be checked out for causeto determine if it is degradation due to aging.
Because equipment starts aging from the day it ismanufactured, control of aging begins at that timethrough the management process. The manage-ment of the equipment in all phases of its life cycle
34
includes attention to shipping and conditions ofstorage prior to installation. After installation, themanagement of the aging process is throughinspection, surveillance, and monitoring with bothpreventive and corrective maintenance.
Aging Assessment Summary
The aging assessment of the HPIS involves anumber of factors including stressors, degradationmechanisms, and failure modes. A summary ofthese various factors in the aging process alongwith the inservice inspection methods is given inTable 15.
Stressors acting on the various components con-tribute to the aging process. Stressors associatedwith maintenance, operation, and testing includeinadvertent HPIS actuation, water hammer, andthermal cycling. Environmental stresses includeabnormal temperatures or pressures, incorrectwater chemistry, boric acid crystals, and vibration.Electrical stresses include external environment oftemperature, humidity and limited radiation,abnormal voltages, and electrical transients affect-ing I&C. Mechanical stresses include pipe misalign-ment, vibration, and dynamic loading from valvesclosing.
Degradation mechanisms for the HPIS passivecomponents (piping, thermal sleeves, and nozzles)include fatigue, crack initiation and propagation,and thermal embrittlement. Valves are subjected to
wear, foreign material, mechanical linkage prob-lems, and seat or disk degradation. Motor opera-tors have wear and loose connections as they age.Air-operated valves main degradation mechanismis due to contaminated air supply this contamina-tion is moisture or oil in the air. For I&C, the degra-dation mechanisms are loose connections,corrosion of terminals, and catastrophic compo-nent failures. Pumps degrade through wear, vibra-tion, and fatigue.
The potential failure modes for the HPIS valvesand pumps are failure to operate when needed;inadvertent operation when not called for; and dur-ing operation, a failure to operate as required. Sec-ondary modes would include leaks, blockage, orcommand faults. Piping and other passive compo-nents have failure modes of leaks, cracks, or looseparts. For I&C, the failure modes are opens, shorts,or failure to operate.
The inservice inspection methods for the HPIScomponents include visual inspection for leaks,volumetric inspection, and operational tests.
Materials in the HPIS susceptible to aginginclude seals, and packing material in pumps andvalves. Carbon steel materials in other systemsexposed to boric acid for some period of time willcorrode. Electrical components in the I&C subsys-tem are subject to degradation of insulation, corro-sion, and wear failures. The stainless steel pipingages from thermal fatigue and wear. Material wearin pumps, valves, and relay contacts is a normalaging process.
35
Table 15. Summary of aging processes for HPIS
Major Component
Nozzles and thermalsleeves
Stressors
System operatingtransients,thermal cycling,vibration, water(hammer)
DegradationMechanisms
Fatigue crackinitiation andpropagation
PotentialFailure Modes
Leaks through wall,loose parts
ISI Methods
Visual inspectionvolumetric inspection
Valves System operationtransients,maintenance, andtesting
Wear, foreignmaterial,mechanical linkagefaults
Leakage, fail tooperate, blockage,command faults
Visual and operationtests
Air-operated valves Contaminate airsupply
Parts degradationby oil in air supply
Fail to operate Visual andoperational tests
ON I&C Electricaltransients,temperaturemaintenance,vibration
Corrosion, looseconnections, failure(catastrophic)
Open, shorts,fail to operate
Testing
Pumps Systems operatingtransients,thermal cycles
Wear, vibration,fatigue
Seal leaks, failto start, failto run
Testing, visualinspection
Pipe supports Vibration, waterhammer
Fatigue, looseningof connections
Breaking loose Visual inspection
Piping Vibration, waterhammer, thermalcycles
Thermal fatigueabrasive wear
Through the wallleakage, or cracks
Visual inspectionvolumetric inspections
Motor operatorsfor valves
Electricaltransients,maintenance
Loose connections,wear
Fail to operate Testing
REVIEW OF INSPECTION, SURVEILLANCE, AND TESTING
For the representative plant studied, the opera-bility requirements of the HPIS are governed by theStandard Technical Specification. These specifica-tions require that at least two HPI pumps be opera-ble when the reactor is critical, and that two trainsof the HPIS must be able to draw suction from theBWST and discharge to the RCS automaticallyupon ES actuation at power levels up to 60% fullpower. In addition, the remaining HPI pump andvalves HP-409 and HP-410 must be operable andvalves HP-99 and HP-100 must be open when thereactor is above 60% full power. Test or mainte-nance on any component is permitted provided thatoperability of the redundant component in theother train is demonstrated first, and subject to thefollowing conditions:
I. If reactor power is less than 6007o, theHPIS must be restored to the appropriatestatus identified above within 24-hours, orplaced in hot shutdown within an addi-tional 12 hours. If the HPIS cannot berestored within the following 24 hours, thereactor must be taken to cold shutdownwithin an additional 24-hour period.
2. If the power level is greater than 60%, theinoperable component must be restoredwithin 72 hours, or power must be reducedto below 60%o in another 12 hours.
The surveillance requirements in the TechnicalSpecificationsI 4 and Section XI of the ASMEBoiler and Pressure Vessel Code' 5 comprise thetesting requirements for the HPIS. Technical Speci-fications 4.5.1.1.1 requires that during each refuel-ing outage the HPIS be tested to demonstrate that itresponds correctly to an actuation signal. Individ-ual components are required to be tested more fre-quently, as defined in Table 16.
Section XI of the ASME defines the inservice test-ing used by the plants. Pumps are tested quarterlyunless a relief request is granted. For these tests, vibra-tion, differential pressure, and flow are measured.Bearing temperatures are also measured but on a lessfrequent schedule. Vibration is an excellent indicatorof pump degradation and is a good monitor for pumpaging. Periodic measurements of the electrical charac-teristics of the motor are not required by Section XI. Acheck at one plant indicated that monitoring is notdone. Pump vibration and performance are not sensi-tive measurements for electrical insulation and other
motor electrical degradation. Electrical characteristicmeasurements would be required to detect such aging.
Valves are tested quarterly unless a relief requesthas been granted. For these tests, stroke time ismeasured, usually without differential pressure.The measurements are often made crudely using astop watch. Such tests would not be effective as amonitor for aging. Periodic measurement of theelectrical characteristic for motor operators is notrequired by Section XI. For resolution of IE Bulle-tin 85-03, most plants are using diagnostic equip-ment to verify torque switch setting. Although theuse of this equipment often includes electrical mea-surements of the operator, the tests are usually onlydone once for verification and rarely repeated. Forvalve testing to be a useful monitor for aging, moreaccurate measurements of stroke time and periodicmeasurement of electrical characteristics of themotor operator will be needed.
Section XI also defines the inservice inspection forwelds. Welds are to be inspected volumetrically each10 years. The cracks in the HPI nozzles and elbowswere detected by leaks, not by the ultrasonic inspectionof the inservice inspection program. The ultrasonictechniques specified by Section XI were found to beinadequate to detect cracks resulting from thermalfatigue. The instrument gain had to be increased signif-icantly and the 45-degree transducer had to be supple-mented by a 60-degree shear wave transducer in orderto detect the cracks. Also, one crack developed in thebase metal of the elbow. Section XI only requiresinspection of welds. Inspection of high-stress areas ofbase metal may be needed. See IE notices No. 88-01and 88-02.l6,l7
Periodic testing standards are given in Part 3 ofTable E-5 in Appendix E. The high pressure injec-tion system will also be inspected periodically dur-ing normal operation for leaks from pump seals,valve packing, and flanged joints. Additional itemsinspected include heat exchangers and safety valvesfor leaks to atmosphere. Typical performance testsare given in Table 17.
The following specific performance tests are per-formed at the representative plant studied:
1. High pressure injection valve verification-Provides verification that each valve is in itscorrect position
2. High pressure injection system leakage-Periodically tests the High Pressure InjectionSystem outside containment for leakage
37
Table 16. Test frequency for HPIS componentsa
Component Type of Test Frequency
HPI pump 3AHPI pump 3BHPI pump 3C
MOV 3HP-3MOV 3HP-4MOV 3HP-5
AOV 3HP-16MOV 3HP-20AOV 3HP-21
MOV 3HP-24MOV 3HP-25MOV 3HP-26
Start, stop, and operating parametersStart, stop, and operating parametersStart, stop, and operating parameters
Stroke and leakStroke and leakStroke and leak
StrokeStroke and leakStroke and leak
StrokeStrokeStroke and leak
MonthlyMonthlyMonthly
QuarterlyQuarterlyCold SD
QuarterlyCold SDCold SD
QuarterlyQuarterlyQuarterly
MOV 3HP-27CV 3HP-101CV 3HP-102
CV 3HP-105CV 3HP-109CV 3HP-113
CV 3HP-126CV 3HP-127CV 3HP-152
CV 3HP-153CV 3HP-188CV-3HP-194
MOV 3CC-7AOV 3CC-8CV 3CC-20CV 3CC-24
Stroke and leakCheck valve function and leakCheck valve function and leak
Check valve function and leakCheck valve function and leakCheck valve function and leak
Check valve function and leakCheck valve function and leakCheck valve function and leak
Check valve function and leakCheck valve function and leakCheck valve function and leak
Stroke and leakStroke and leakCheck valve function and leakCheck valve function and leak
QuarterlyRefueling SDRefueling SD
Refueling SDRefueling SDRefueling SD
Refueling SDRefueling SDRefueling SD
Cold SDRefueling SDCold SD
Cold SDCold SDRefueling 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 linevalves
High pressure injection pumpsuction valves
Borated water storage tankoutlet valves
One of two pumps operatescontinuously. The other pump willbe operated periodically
The remotely operated stop valvesin each line are opened partiallyone at a time. The flow monitorswill indicate flow through the lines
The valves are opened and closedindividually and console lightsmonitored to indicate valve position
The operational readiness of thesevalves is established in completingthe pump operational test discussedabove. During this test, each valveis tested separately
3. High pressuire injection system performancetest-Demonstrates the operability of theHPI pumps in accordance with applicableASME code and identifies potential problemareas as early as possible.
4. High pressure injection system ES Test-Demonstrates the HPI System is operablefrom an ES signal.
5. High pressure injection pump venting-Periodically vents the casings of nonoper-ating HP pumps to prevent gas buildups
6. High pressure injection motor coolant flowtest-Periodically tests the cooling water flowthrough the HP pump motors to ensure ade-quate upper motor bearing cooling
7. High pressure injection check valve func-tional test-Demonstrates the operabilityof the HP System check valves.
Additional inspections for leaks, as well as func-tional tests may be performed periodically on
major components if required by the utility. Forexample, after maintenance, functional tests maybe necessary to verify operation. Monitoring con-sists of comparing the performance of similarchannels and visual inspections for leaks in pipingvalves and pumps. Current maintenance practicesby utilities follow recommendations by vendors formajor components, such as valves and pumps. TheBabcock & Wilcox plants that have experiencednozzle cracking have also indicated enhancedinspection and surveillance of the nozzle and asso-ciated piping welds.
Some utilities have used advanced I&C cabletesting during refueling outages to establish base-lines and obtain trending data on electrical equip-ment. One such system used primarily for cableand connection testing is the ECCAD system. 18
Another advanced surveillance system is the motoroperated valve analysis and test system(MOVATS). 3 Since 1984, MOVATS has been usedas a diagnostic tool for motor operated valves.
39
ROLE OF MAINTENANCE IN COUNTERACTING AGING EFFECTS
Present Regulations andGuidance
The current USNRC regulation approach tonuclear plant maintenance is embodied in require-ments for quality assurance during design, con-struction, and operations consistent with the safety(10 CFR 50, Appendix B) and surveillancerequirements that ensure necessary availability, andquality of systems and components is maintained(10 CFR 50.36). These rules and regulations pro-vide no clear programmatic treatment of nuclearplant maintenance. Regulatory Guide 1.33, Revi-sion 219 endorses ANSI N18.7-1976/ANS 3.220,which provides no specific guidance regardingmaintenance, but covers administrative controlsand quality assurance for the operational phase ofnuclear power plants.
Current Maintenance Practices
The representative plant studied follows manu-facturer's recommendations for preventive mainte-nance on major components and plant specificprocedures. For example, preventive maintenanceon HPI pumps and system valves includes thefollowing:
1. High pressure injection pump-In order todetermine internal wear of the pump, peri-odic efficiency tests shall be performed.This efficiency shall be used as a compari-son to initial efficiency.
2. System valves-Maintenance on remotelyoperated valves shall be in accordance withthe vendor's recommendations. Manualvalve maintenance shall include checkingfor packing leakage when the system isunder pressure. Safety relief valve testsshall consist of in-place or bench testing ofsetpoints as appropriate.
Corrective maintenance is also minimized byobserving good operating practices and precau-tions. For example, the following limits and precau-tions shall be followed to prevent componentdamage and abnormal aging.
Normal Operation Precautions. Prior to start-ing the high pressure injection pumps, particular
attention must be given to the opening of all valvesin the suction line and to proper venting of thepumps. Failure of suction could result in instantloss of the started pump. The minimum allowableflow of a pump is 30 gpm.
The HP pump can be started against shutoffhead, but operation of the pump in this conditionfor over 30 seconds could cause the pump to over-heat. The HP pump must be tripped if the motorbearing temperature exceeds 215'F.
The HP pumps must not be started with an openflow path to the RC pump seals. Injection seal flowis required to all RC pumps when the RC pressureand temperature are above 100 psig and 1900F.
The maximum flow through one letdown coolershall not exceed 80 gpm. The maximum seal returntemperature should not exceed 130'F to avoid dam-aging the demineralizer resins.
The maximum flow through one makeup filtershall not exceed 150 gpm. A maximum pressuredrop of 30 psi should not be exceeded at any flowrate.
High Pressure Injection Mode Precautions. Thesame precautions apply during high pressure injectionin regard to suction and discharge flow of the HPpumps as under normal operating conditions. Thepossibility of pump runout due to a line break on thedischarge side must be considered. The maximum flowof 525 gpm must not be exceeded for any length oftime because overheating of the motor may occur.
Benefit of Preventive andCorrective Maintenance
Preventive maintenance should be performed onthe basis of need because the adoption of arbitraryand frequent maintenance can be counter to safety.Preventive maintenance should be supported by tech-nical evidence and reviewed periodically. The benefitsfrom preventive maintenance include higher systemavailability, increased life, and higher reliability.
Corrective maintenance is usually performed ona priority basis and closely coordinated with test-ing. After corrective maintenance, the channel orsystem is usually tested to verify that it is function-ally correct. Likewise, a component or system thatfails a test will probably require adjustment or cor-rective maintenance. Thus, corrective maintenance
40
is a necessary part of plant operations and keepsthe HPIS functioning properly.
Improper Maintenance
Excessive preventive maintenance can have anegative impact on safety and aging. Thirty-fivepercent of nuclear plant abnormal occurrences maybe due to faulty maintenance and surveillance test-ing.2 1 Human error during maintenance hasinvolved the wrong unit or train and may increasethe potential for equipment damage. In order for apreventive maintenance program to be effective itmust apply to both equipment with detectable deg-radation effects, and methods of detecting degra-dation before failure. Only about 25% ofequipment failures are preventable. 22 Examples offaulty maintenance include sticking control break-ers because of lack of lubrication, maintenance onwrong train or component, and personnel errors.
Recommendations for PreferredMaintenance Practices
For the HPIS, the preferred maintenance recom-mendation would be the maintenance practice rec-ommended by the vendors for major componentssuch as pumps and valves. In addition, anenhanced inspection program for leaks and cracksin piping and nozzles would be coordinated withcorrective maintenance. A maintenance programtakes into account many factors including safety,operations, economics, and availability. The inter-val for equipment maintenance frequency shouldtake into account potential impending failures,their detection, and known equipment wearoutregions. Reliability centered maintenance has beenused in the aircraft industry and is being consideredin some nuclear plants. When properly applied,reliability centered maintenance should enhanceplant safety and reduce life cycle costs.
41
CODES AND STANDARDS
Determining aging effects on equipment and lifeextension is a key part of the NPAR program. Oneof the outputs of the NPAR program is to recom-mend upgrades for old standards or recommenddevelopment of new ones. Electrical equipmentissues relating to HPI equipment qualification andaging are addressed by the following documents.
NUREG 058823 requires that qualification pro-grams for electrical equipment should identify materi-als susceptible to aging effects and establish a schedulefor periodically replacing the equipment and material.
IEEE 323-1983,24 the industry standard uponwhich the above requirements are based, includes anumber of paragraphs addressing this issue. Forexample, Paragraph 6.4.2 on Operating Historystates that, in order to use operating history infor-mation for establishing qualification, a record orauditable data showing that equipment similar tothat being qualified has been exposed to levels ofenvironment at least as severe as those expectedfrom all service conditions for which the equipmentbeing qualified is required to function and that theequipment satisfactorily performed the functionsrequired for the equipment being qualified. Thoseelements of required exposure not covered by oper-ating history may be accounted for by testing.
Regulatory Guide 1.8925 provides the guidelines formeeting the Equipment Qualification Rule andendorses IEEE 323-1974.
Data from naturally aged components or in-situmeasurements should be used to verify, where pos-sible, data from artificial aging of componentsused during equipment qualification.
The Standard Review Plan,2 6 Section 3.11,"Environmental Qualification of Mechanical andElectrical Equipment," which provides guidancefor USNRC staff in reviewing FSARs, includesrequirements for maintenance/surveillance pro-grams for equipment located in mild environments.Specifically, it is required that "the maintenance/surveillance program data shall be reviewed period-ically (not more than every 18 months) to ensurethat the design qualified life has not suffered ther-mal or cyclic degradation resulting from the accu-mulated stresses triggered by the abnormalenvironmental conditions and the normal wear dueto its service condition. Engineering judgment shallbe used to modify the replacement program and/orreplace the equipment as deemed necessary." ThisSRP guideline should be considered for possibleapplication in any new maintenance standard orguide for maintenance.
The HPIS regulatory requirements and guide-lines are given in Appendix E.
The Board of Nuclear Codes and Standards hasoverall responsibility for codes and standardsdevelopment covering nuclear plant life extension.The IEEE working group 3.4 is presently review-ing selected IEEE standards related to plant lifeextension and plans to develop a guideline docu-ment. The mechanical components are covered byASME code including Sections III-C, VIII,and XI. The process of developing recommenda-tions for requirements similar to the electricalequipment is an ongoing process as part of theNPAR program.
42
HPIS AGING SYSTEM UNAVAILABILITY ASSESSMENT
The main objective of this unavailability assess-ment is to identify the components that contributeto HPIS unavailability, and determine if theychange with time because of aging. Another objec-tive is to determine the aging contribution to theHPIS unavailability for the three emergency modesof operation. These three operating modes are: (a)high-pressure injection with one of three HPIpumps [HPI(l) and HPI(2)] required, (b) high-pressure injection with two of three HPI pumpsrequired and (c) the recirculation mode. A thirdobjective is to identify the type of events that con-tribute to HPIS unavailability.
This aging assessment is based on the linearaging model27 and uses data from the ProbabilisticRisk Assessment (PRA)28 for the representativeplant studied and generic failure cause data onHPIS components from a composite of nine PWRplants that were at least 10 years old. Thisapproach is an approximate method that uses PRAresults (steady state models) to evaluate aging risk.The PRA results provided the system fault treesand baseline data for this study. The failure causedata is used to estimate the time dependent failurerates and the PRA is rerun at discrete times to pro-vide the aging assessment. The software tool usedfor this work was the Integrated Reliability andRisk Analysis System (IRRAS).2 9
PRA and Basic Event Data
The PRA for the representative plant used in thisstudy, had been developed as part of a program toimprove the PRA capability of the electric utilityindustry. This PRA was used for training personnelfrom seven utilities during the course of its develop-ment. The PRA results were also used as input toLiving Schedules for plant modifications and to theIntegrated Safety Assessment Program.
The results of the PRA showed that the systemrisk is dependant on support systems and eventsinternal to the HPIS. Only the internal events wereconsidered in this study.
The internal events included HPIS componentfailures, human errors in leaving componentsunavailable, and having a system's segment out formaintenance. The basic event data from the PRAfor the hardware, human errors and maintenanceare given in Appendix I. The basic event data wereused to run the baseline calculations assumed to befor a plant with random nonaging failures.
Failure Cause Data
The failure cause work30 was used to evaluatethe aging fraction (f) and the mean time tofailure (T) for each of the major HPIS compo-nents. Although, Reference 30 does not report thespecific data used in this evaluation, the files devel-oped from the NPRDS data base as part of thatwork were accessed to obtain the information.
The upper and lower bounds were developed in Ref-erence 30 to account for the uncertainty encountered inaccurate identification of aging-related causes on thebasis of the component failure descriptions. The cate-gorization scheme defined when a failure should beclassified as related to aging or nonaging. When insuf-ficient information was contained in the failuredescription, the aging classification was unknown.These failures were then used to establish the upper andlower bounds for the aging-related failure-cause frac-tions. The upper bounds (UP) were calculated usingthe failures classified as unknown as aging-related fail-ures, while the lower bounds (LW) are calculated usingthem as nonaging-related failures. Upper and lowerbounds were developed for the data used in this reportby the same method. The upper and lower bounds for fand T are included in the data shown in Thble 18.
In Table 18, the first three to five letters and num-bers identify the system and component numbers.The last three letters in the event name describe thefollowing event codes:
CVOMVOVxTPPSPPRAVORVFFIT
check valve fails to openmotor-operated valve fails to openmanual valve fails to openpump fails to startpump fails to runair-operated valve fails to openpump fails to start on low seal flowflow transmitter fails high.
Methodology
The random nonaging failure rates from the PRAwere used for the first five-year period. Aging datataken from the failure cause study were used to calcu-late a new failure rate for subsequent five-year periods.When the constant rate contributions are incorpo-rated, the time dependent failure rate using the linearmodel is given by the equation:
43
Table 18. HPIS data for the risk assessment, aging acceleration factors, and component unavailabilities in 5-year increments forupper and lower bounds
Aging FailureFaIlure Fraction
Mean Time iSnion Failure
To Failure Tim Rate(Years) (Yearn) Per Year
Aging Accele-ration Factor LCwvofeft onana, an, 'Ky
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(cent ane ODat Used [V £ .I TUP TLV 11T.
HPIOICVOHP 9102CV0
HF1 I09CVOH1713CV01IS2CVO
HP153CV0HF I188CV0LPSSCVOLPS7CVOHP1448WVTLPS4VVT
*N LPS6VVTHP24MV0HP25KVOHP291V0HP4090VOhP410hVOHFBPPSHPCPPSHPAPPRHPRAPPRHP8PPRHPR6FPRHPCPPRHPRCPPRHPI ISAVOHP16A8OCSSXVOCISGCVOHPRC8PPS11153RVFW93lFlT
HpI CVHP1I CV1W9 CRHPI CVHPI CRWlI CVHFI CRH11I CVHPI CVHP1I KVHPI ICVW7 IHCV
HPI HOVW3 MCUHPI KCVH1I HCVWlI KVHP1I ItOPHFPI MoPHPI MOPHPI nOPHPI MOP
hPI MOP
WlI POV
HPI POW
HF I CyHFPI HOPHFI KFPHPI FT
0.900.900.900.900.900.900.900.900.900.830.830.830.840.640.840.840.840.850.860.860.860.860.860.860.860.88
0.880.900.900.850.8S0,9Z
0.630.630.630.630.630.630.U30.630.630.610.610.610.490.490.490.490.490.U40.540.540.540.540.540.540.540.560.560.630.630.640.640.41
6.886.886.886.886.886.886.UB8.886.888.208.208.208.018.018.018.018.018.668.668.668.668.668.668.668.6612.2112.216.886.888.668.663.S6
8.588.588.U88.588. 5838.U88.588.588.588.828.828.628.418.418.418.418.419.009.009.009.009.009.009.000.0011.02I1.128.588.589.009.004.00
2.7E-032. 7E-032.7E-032. 7E-032.7E-032.7£-032.7E-032.7E-032.7E-03I. OE-001.OE-00I. OE-002. 7E-032.7E-032.7 E-032.7E-032. 7E-032. 7E-032. 7E-032. 7E-03Z. 7E-032.7E-032.7E-03Z. 7E-032. 7E-032. 7E-032.7E-032. 7E-032. 7E-032. 7E-032. 7E-032.7E-03
I. 7E-041. 7 E-042.2E-032. 2E-032.OE-042.OE-042. OE-042. OE-042. oE-047.8E-047 .8E-047.BE-044.0E-044. 9E-04
4.9E-044.9E-044.9E-042. IE-022. IE-027.4E-OZ7. 4E-017.4E-027.4E-017.4 E-027.4E-01I. ZE-OZ1. 2E-026. 7E-046.7E-046. SE1-033. 7E-03I .IE-02
3.01-04 4.SE-053.OE-04 4.5E-053.8E-03 0.8E-043.8E-03 5.8E-043.SE-O4 5.3E-053.5E-04 5.3E-053.5E-04 5.3E-053.5E-04 5.3E-OS1.5E-04 S.3E-OS6.2E-04 0.8E-046.2E-04 1.8E-046.2E-04 1.8E-044.3E-02 7.5E-034.3E-02 7.5E-034.3E-02 7.SE-034.3E-02 7.5E-034.3E-02 7.SE-032.0E-02 5.SE-032.OE-02 S.SE-037.0E-02 1.9E-027.0E-00 1.9E-017.OE-02 1.9E-027.0E-01 1.0E-007.0E-02 1.9E-027.0E-01 0.9E-019.61-03 1.8E-039.6E-03 1.8E-03I.2E-03 1.8E-041.2E-03 1.8E-0481E-03 1.71-033.5E-03 9.7E-044.2E-02 2.50-03
8.7E-058. 71-058. 7E-058.7 E-059.8E-059.8b-059.8E-OS0.8E-05
g.8E-053. 9E-043.9E-043. 9E-046.4E-036. 4E-036. 4E-036. 4E-036. 4E-038.4E-048.4E-04Z. OE-042.0E-032. OE-042.0E-032. OE-042.1E-031. 5E-033. SE-038.7E-OS8.7E-058. 4E-044.8E-043.1E-05
9.1E-OS 8.8E-OS9.1E-OS .81E-051.4E-04 9.5E-05.41E-04 9.SE-05
I.OE-04 9.9E-O0L.OE-04 9.SE-051.OE-04 9.9E-050.OE-04 9.9E-05l.OE-04 9.9E-053.5E-03 1.3E-033.SE-03 1.3E-033.5E-03 1.3E-03
7.OE-03 6.SE-03.0OE-03 6.5E-03
7.0E-03 6.SE-037.OE-03 6.5E-037.0E-03 6.5E-030.IE-03 9.1E-04.1IE-03 9.IE-04
I.IE-03 4.6E-04011E-OZ 4.6E-03I.IE-03 4.ff-O41.1E-02 4.61-03.1IE-03 4.81-04
I.IE-02 4.61-031.7E-03 1.6f-033.6E-03 3.SE-03I.OE-04 8.9E-05I.OE-04 8.9E-059.2E-04 8.6E-045.3E-04 4.9E-046.0E-04 6.5E-05
9. 5E-059.5E-05I.9E-04I.9E-04
1. IE-041.IE-04I. IE-04I.IE-04I.IE-046.6E-036.8E-036. 6E-037.EE-03
7. 6E-037.6E-037.6E-037.6E-031. 4E-030.4E-032.IE-032.1E-022. IE-032.IE-022. IE-032. IE-021.9E-033 .8E-031. ZE-041. 2E-041. OE-03S. 7E-041. 2E-03
8.8E-058. 8E-05
.0OE-041.OE-04
9. 9E-059.9E-OS9.9E-OSS.9E-059.9E-052. 2E-032. ZE-032.2E-036.86-036.81-036.f6-036.6E-036.f6-039.9E-049.9E-047.3E-047.3E-037. 3E-047.3E-037.3E-047.3E-031. 6E-033. SE-039.2E-0S9. ZE-058. 9E-04S.IE-04I.OE-04
S.9E-OS9.SE-052.4E-042.4E-04I. IE-04I. IE-04I.IE-04I.IE-04I.IE-04
9.7E-039. 7E-039. 7E-038. E-038. 1E-03S. IE-038. IE-038. IE-031. SE-03I.SE-033 .OE-033.0E-023.0E-033.OE-OZ3.0E-033. OE-02Z.OE-033.9E-03I.3E-041.3E-041. IE-036. 2E-04I.7E-03
8. 9E-058.9E-05I IE-041.IE-04I.OE-04I.OE-041.OE-04I. OE-04I.OE-043. 2E-033. 2E-033. 2E-036.7E-036. 7E-036. 7E-036.7E-036.7E-03I. 1E-03I. 1E-03
9.9E-049.9E-039.9E-049 .9E-039.9E-049 .9E-03I.7E-033. 6E-039.4E-059.4E-tS9. IE-045.2E-041.3E-04
I.OE-04I. OE-042 .9E-042.9E-041.2E-041. ZE-041. 2E-04.2ZE-04
1.ZE-041.3E-02
1.3E-02I .3E-028. 7E-03
8.7E-038. 7E-038. 7E-038.7E-031.9E-031.9E-034.0E-034.OE-024.0E-034.OE-024. 0E-034.OE-022.1E-034. OE-031. SE-041. SE-041.2E-03
6. 7E-042. 3E-03
8 .9E-0O8. 9E-05I. ZE-041. 2E-04I.OE-04
I.OE-04
I. OE-04.0OE-041.OE-04
4.1E-034. IE-034. IE-036.8E-Q36.8E-036 .8E-036.8E-036 .8E-031I. IE-03I.IE-031.3E-031. 3E-021.3E-031. 3E-0Z1 .3E-031. 3E-021I. 7E-033.6E-039.7E-059. 71-0S9.3E-045.3E-041. 7E-04
1. IE-04I.IE-04
3.5E-043. 5E-041.ZE-04I.ZE-041.ZE-040.2E-04I.2E-04
1. 6E-021. 6E-020. 6E-029.3E-039.3E-039 .3E-039. 3E-039 .3E-032.2E-032.2E-030.0E-034. 9E-024. 9E-034.9E-024.9E-034.9E-022.2E-03*. IE-031.7E-04I.7E-041.3E-047. 2E-042. 9E-03
9.0E-059.0E-051.3E-041.3E-04I .OE-04I.OE-041.OE-04
I.OE-04
1. OE-04S.OE-035.0E-03S.OE-035.9E-036.SE-036.9E-036. 9E-036 .9E-031. ZE-031. 2E-03I.SE-03
1.SE-02I.SE-03
1.SE-02
1.SE-03
1. 5E-021.7E-033.5E-039.9E-059. 9E-059. 6E-045.5S-042. OE-04
I.I1-04I.IE-04
4. OE-044.Ot-041.3E-043.3E-041.31-041.3E-041.3E-04I.9E-021.9E-021.9E1029.9E-039 .9E-039.9E-039. 9E-039.9E-032.4E-032. 4-035.9E-035.9E-025.9E-035.9E-02S.9E-035.9E-022.4E-034.3E-031.8£-041.8E-041. 3E-037. 6-043. 5E-03
9. IE0959.1E0951.3E-041. 3E-04I.OE-04I0OE-04I.OE-04I. 0E-041.OE-04
5.9E-035.9E-035. 9E-037. OE-037.OE-037.0E-037.OE-037.OE-03
1.3E-031.3E-03I.8E-03I.8E-021,8E-031:.8E-02
1.8E-031. 8E-021.7E-033.6E-03I. OE-04I. OE-049. 8E-045.6E-042.4E-04
1.2E-041. 2E-044.SE-044.SE-041.3E-041.3E-041.3E-041.3E-04I.3t-042.2E-022.2E-022.2E-02I. 0E-02I. OE-OZI. OE-OZI.OE-02I.OE-022. 7E-032.7E-036.8E-036.8E-OZ6.8E-036. 8E-026. 8E-036. 8E-022. SE-034.4E-032.0E-04Z. OE-041.4E-038. IE-044.0E-03
9. IE-059.1E-051. 4E-041. 4E-041.OE-041. OE-04I .Ot-04I. OE-041.0OE-045.8E-035.8E-03S.8E-037. IE-037. IE-037. IE-037.IE-037.IE-031. 4E-031. 4E-032. OE-032.0E-022. 0E-032.OE-022. OE-032.OE-021.6E-033. 7E-03I.OE-04I. OE-04I.0E-035.7E-042. 7E-04
Nos= X + at
where
X(t) = time-dependent failure rate
(1) The computer calculations rounded off to7.0 x 10-2 y2 .
The calculations for probability of failure werethen performed for 5-year increments of time usingthe 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 abased on the moments considerations from Refer-ence 25 is
= new probability for the new 5-yearincrease
= probability for the previous 5-yearincrement
Pn-I
a =[ (IfC (2) AT = five years
t = twenty-four hours (mission time)where
34 = constant from the deviation3
f = nonrandom fraction of failures ofthe component which are caused byaging 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 convertedto years by using 8760 hours per year.
Sample Calculation for (2)
The event chosen for this example is the highpressure pump A fails to run (HPAPPR). The basicevents 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 to1.1 x 10-3.
The probabilities identified under the P headingin Thble 18 were calculated for each componentevent and for each 5-year increment. The base case[P(5)] values were taken directly from the PRAdata. Each column of data was then used as inputto the IRRAS program for the calculations thatrepresent the time period for that column. Alsoshown in Table 18 are mission time, failure rate,calculated values for the aging failure acceleration
45
parameters (both upper and lower bounds), and thecomponent unavailabilities for each 5-year periodup to 40 years. These were the basic data and inter-mediate calculated results for this aging riskevaluation.
The IRRAS software program was developed atthe INEL to run on a personal computer.Version 2.0 was used for this analysis. The faulttrees for the three cases modeled were loaded intothe personal computer and the IRRAS programrun to determine the cut-sets for the significantsequences. A minimal cut-set is defined as thesmallest combination of component failures,which if they all occur, will cause the top event ofthe fault tree to occur.
The outputs that were selected from the IRRASprogram were the minimum cut-set upper bound(which is the HPIS unavailability) and the Fussell-Vesely (F-V) importance measure. The F-V impor-tance measure is a measure of contribution of theevent to the system unavailability. A F-V impor-tance value of 0.01 or greater was consideredsignificant.
The Fussell-Vesely importance is determined byevaluating the sequence frequency with the basicevent failure probability at its true value and againwith the basic event failure probability set to zero.The difference between these two results is dividedby the true minimal cut-set upper bound to obtainthe Fussell-Vesely importance. In equation formthis is
always available; (b) that the probabilities forhuman error events and maintenance events remainconstant for all time periods; (c) the PRA failurerates were the random failure rates and apply forthe first 5-year interval at the representative plant;and (d) when not given in the PRA the mean failurerates were calculated by dividing the demand fail-ure rate by the estimated time between demands.
In addition, the following modeling assumptionsare from the PRA (Reference 28) for the three casesconsidered.
I. It was assumed that flow from the boratedwater storage tank (BWST) through eithersuction line (valves 3HP-24 or 3HP-25 inFigure 10) is sufficient for all three HPIpumps. The letdown storage tank (LDST),which provides HPI pump suction initiallyas the BWST valves open, was not assumedto be available for emergency HPI opera-tion since makeup to the LDST is limitedto less than 200 gpm. Furthermore, it wasassumed that adequate NPSH to the threeHPI pump exists with the LDST suctionremaining open.
2. Between the discharges of pumps 3HP-P3B and 3HP-P3C there is an additionalcross-connection that is not shown in Fig-ure 10. This line contains two normallyclosed manual valves, and would serveonly as a backup to the crossover lines con-taining motor-operated valves 3HP-409and 3HP-410. Therefore, its availabilitywas judged to have very little effect onHPIS reliability and was not modeled.
3. Each injection loop splits into two linesthat provide flow to the cold legs down-stream of the reactor-coolant pumps. Forsequences that require flow from only oneHPI pump, it was assumed that only oneof the injection-line splits was needed. Forsequences requiring the function of twopumps, at least two of the splits wereassumed to be required.
Components that ContributeSignificantly to SystemUnavailability
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 withthe basic event failure probability atits true value
F(O) = minimal cut-set upper bound(sequence frequency) evaluated withthe basic event failure probability setto zero.
Assumptions
The assumptions made in performing this analy-sis are: (a) that all support systems such as IEpower, service water, and low pressure injection are
The significant Fussell-Vesely importance mea-sures calculated for the initial 5-year period and the40-year period are summarized in Table 19 for eachof the three operating modes. The values at 40 years
46
PZR 9 i From LPI coolerdischarge
(LPR system)Inside RB I
I 3HP.120
IHP- 3HP-RCP- Orifice 1 1193A W ~ _A
3A2 _outlet 3HP-126
3HP-13HP-410
3HP-110 1
seals
Orifice
3HP-188
3HP-27ES-2 opei
OrificiD I
Inside RBj
From LPIcooler discharge(LPR system) 3LP-57 From BWST
(LPI system)
6 10 474
Figure 10. Components that contributed significantly to HPIS unavailability for emergency modes ofoperation.
47
Table 19. Events with significant Fussell-Vesely importance measures
HPI(1) HPI(2) Recirculation Mode
EventName
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-YearMean
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-YearMean
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-YearMean
0.3352
0.3352
0.4848
0.0223
0.1054
0.0318
0.0650
0.2941
HPRAPPR
HPRCPPR
0.2166 0.4638
0.31800.1524
are the arithmetic mean of the upper and lowerbound values at 40 years from the tables in Appen-dix 1.
The motor-operated valves "failure to open"events, HP-24 MVO and HP-25 MVO, are signifi-cant for all three modes of operation for both5 years and after 40 years aging. Event HP-26MVO is significant for both the HPI(2) and recircu-lation operating modes (for both 5 and 40 years)and becomes significant for the HPI(I) operatingmode after 40 years aging. All of the events listedfor the recirculation mode are significant for both5- and 40-year time periods. Event HP-148 VVTbecomes significant for both HPI(I) and HPI(2)operating modes after aging 40 years. The pumpevents HPCPPS, HPCPPR, HPAPPR, andHPBPPR become significant after 40 years for theHPI(2) operating mode, but were not significantevents at 5 years. This is an indication of the agingeffect on the pumps.
The- conclusions' that can be drawn from theanalysis of the F-V importance measure is thatseven components involving 10 events were indenti-fied as contributing significantly to system unavail-ability for at least one operating mode and timeperiod. The seven components are highlighted in
Figure 10, and are motor-operated valves HP-24,HP-25, and HP-26; manual valve HP-148; andHPIS pumps A, B, and C.
HPIS Unavailability
The cut set quantification reports were obtainedfrom each run of the IRRAS program. The reportsfor the 5-year, 10-year, and 40-year runs are given inAppendix I. The HPIS unavailability for eachoperating mode modeled is given as the minimalcut-set (mincut) on the reports. These HPISunavailabilities are summarized in Tables 20, 21,and 22 for three emergency operating modes. Thesetables present the mean, upper bound and lowerbound for 5-year increments from 5 years to40 years.
The HPIS mean unavailability for the threeemergency operating modes are plotted inFigure 11. The one-of-three-pumps-required oper-ating mode had an unavailability starting at9.59E-5 at 5 years and increased to 1,353-4 at40 years. The two-of-three-pumps-required operat-ing mode has the highest HPIS unavailability start-ing at 1.720E-4 at 5 years and increased to 4.280E-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
I I I I I -10 7
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 operatingmodes.
50
HPIS unavailability started lower at 8.401E-7 at5 years and increased to 1.27E-5 at 40 years. Therecirculation mode had the greatest increase, butwas still an order of magnitude less than the injec-tion modes. Normally changes in unavailability atless than an order of magnitude are not importantin final PRA results. Based on that assumption, thecomputed change for the two HPI modes wouldnot be expected to change the results of the fullplant PRA significantly. For the recirculationmode, the change is more than an order of magni-tude but even with the change the unavailabilityremains very low. A cursory review of the repre-sentative plant PRA indicated that loss of HPI inthe recirculation mode from failure of componentsdoes not appear in any of the important eventsequences and, therefore, a significant changewould not be expected in the full plant PRA.
Estimate for Types of Events ThatCause HPIS Unavailability
An estimate of the contribution of the types ofevents to the system unavailability was obtained bysumming all the F-V importance values for a partic-ular event type for a given time period and dividingby the sum of all the F-V importance values for thatsame time period. Because of truncation error andthe possibility that some of the events may beincluded more than once in different cut-set combi-
nations, this result will only be a rough estimate ofthe contribution of each type of event to the HPISunavailability. The results are shown in Table 23.
As demonstrated by the values in Table 23, themotor-operated valves failure to open is the type ofevent with the highest probability of contributingto HPIS unavailability for the two HPIS, injectionmodes of operation. For these same two modes ofoperation the human error events have the secondhighest probability followed by the events associ-ated with maintenance. The change in probabilitiesfrom 5 years to 40 years may vary either up ordown depending on how the cut-sets combine. Themost noteable probability increase with age was forthe manual valves and pumps in the HPI (two ofthree pumps) mode. The manual valves probabilityincreased from 1.4%o to 21%70 and the HPIS pumpsincreased from 0.2% to 9.3%.
For the recirculation mode the maintenanceevents were the most significant, 41°7o at 5 years,but dropped to only 7.5 qo at 40 years. All the otherevents showed an increased with age between5 years and 40 years. The motor-operated valvescontinued to be a contributor with 20%o at 5 yearsand 39% at 40 years.
In general, all the component events had anincrease in unavailability over the 40-year periodfor all three operating modes, except for motor-operated valve in the HPI (two of three pumps)mode, which decreased, and the check valves in theHPI (one of three pumps) mode, which had nochange.
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)
RecirculationMode
Event Type
Human error
Segment of system outof service formaintenance
Motor-operated valvesfailure to open
Check valvesfailure to open
Manual valvesfailure to transfer
:HPIS pumpsfailure to run orfailure to start
5 yr
35
5
59
0.8
0.05
0.02
40-yrMean
26
5
66
0.8
0.8
0.7
5 yr
39
10
49
0.4
1.4
0.2
40-yrMean
20
20
30
0.5
21
9.3
40-yrMean5 yr
25 27
41 7.5
20 39
0 0
13 22
0.9 3.9
52
CONCWSIONS
The conclusions for the HPIS aging research arebased on a detailed study of one plant and supple-mented by information from generic data bases.Experience from the generic data bases shows thatthe motor-operated valves were the most trouble-some components. Some failures and aging mecha-nisms are not being detected by current inspectionand surveillance methods. The study indicates thataging causes only a moderate increase in systemunavailability and that motor-operated valves arethe components that contribute most often to thesystem unavailability. Conclusions for the specificstudy objective are presented in the followingsections.
Operating Experience
The HPIS operating experience was evaluated byreviewing generic data bases and plant specificdata. The operating experience from the NPRDS,covering B&W and Westinghouse plants, indicatedthat approximately 57%7 of the component failureslead to system degradation, but because of systemredunancy, only 0. 7% of HPIS component failurescaused loss of system function. The data alsoshowed that 21.356 of these failures were agingrelated.
The NPE data covering all PWRs were reviewedand four types of components were identified ashaving the highest frequency of failure. They werevalves-35%, I&C-1956, pumps-155o, and piping-7M. Both NPRDS and LERS had the same relativeranking for these components. The correctivemaintenance ranked I&C and piping failures higherthan valve and pump failures because of many sub-components failures that were repaired on theseparticular components.
The conclusion for the data base reviews is that asignificant number of HPIS component failureshave been aging related. The term "aging related"is used because the information given in data basesoften requires further analysis to determine extentof aging contribution to the failure. The conclu-sions for specific problems related to aging aresummerized in following sections.
Specific Problems Related toAging
Problems with boric acid systems have occurred.Leaks of borated water onto carbon steel causedcorrosion of the carbon steel. Such corrosion canoccur at a rate greater than I in. per year as provenby both laboratory tests and plant experience.Boric acid crystals have caused blockage and in oneinstance precipitated out in an HPI pump causing itto malfunction. The latter was due to an internalvalve leak.
The thermal sleeve and nozzle cracking problemexperienced in B&W plants in 1982 was attributedto thermal fatigue. Thermal cycling occurred whenthe makeup flow cycled on and off and when sys-tem injection flow occurred. The problem was cor-rected at the representative plant studies byredesign of the thermal sleeve and maintaining acontinuous small makeup flow. An augmentedinservice inspection plan was also implemented.
The HPIS interacts with many of the other reac-tor systems. This makes it particularly vulnerableto common mode failures when a problem inanother system can prevent HPIS from performingits function. For example, HPI pump seal cooling issupplied by the service water system. The HPI itselfsupplies RC pump seal cooling and RC makeupwater during normal operation.
IEEE 323-1974 was the industry standard usedto address HPI electrical equipment qualificationand makes reference to the use of historical dataand the analysis of failures and trends. The designand maintenance data should be reviewed periodi-cally (approximately 18 months) to ensure that thedesign qualified life has not suffered. The reviewshould include thermal or cyclic degradation result-ing from accumulated stresses triggered by abnor-mal environmental conditions and the normal weardue to its service condition.
Problems with the ECCS piping cracks haveoccurred at Farley-2 and Tihange-1. These twoplants were Westinghouse plants utilizing the highpressure injection pumps for both charging duringnormal operation and injecting emergency corecoolant during a LOCA. The pipe cracking wasattributed to thermal fatigue that occurred becausethe failure of valves in the safety injection systemallowed relatively cold water to flow back into theprimary system.
53
The B&W system that has the dual purposeHPIS pump would potentially have the same prob-lem. The difference in the B&W system is that ituses the same nozzle for makeup and injection.
Aging Assessment
The aging assessment of the impact of aging onHPIS operability involves a number of factorsincluding the identification of stressors, degrada-tion mechanisms, and failure modes. The variouspotential electrical, mechanical, and environmen-tal stressors were indentified and degradationmechanisms determined for both passive and activecomponents. These factors were summarized inTable 15 along with the various failure modes andISI methods.
Materials in the HPIS susceptible to aginginclude seals and packing material in pumps andvalves. Carbon steel materials in other systemsexposed to boric acid for some period of time willcorrode. Electrical components in the I&C subsys-tem are subject to degradation of insulation, corro-sion, and wear failures. The stainless steel pipingages from thermal fatigue. However, material wearin pumps, valves, and relay contacts is a normalaging process.
Inspection, Surveillance, andMonitoring
The surveillance requirements in the technicalspecifications for a plant and the ASME Boiler andPressure Code comprise the testing requirementsfor the HPIS. Additional inspections for leaks maybe performed periodically along with functionaltests on major HPIS components if required by theutility. For example, after maintenance, functionaltests may be necessary to verify operation.
Included in the surveillance methods are visualinspections, operational testing, and calibrationsfor the HPIS as a system. Individual componentsmay have additional tests. For example, the limitor-que valve operators require periodic 18-monthinspections as part of routine maintenance forcleaning and lubrication. Rotork valve operatorsare qualified for 40-year life or 2000 open-closecycles, therefore, no preventive maintenance pro-gram is recommended. Rotork ensures reliability bycarefully sealing out the environment during instal-lation. Routine maintenance of inner parts isdeemed likely to cause defective sealing and, shouldbe avoided. The exception is that seals and lubri-
cants subjected to radiation and heat aging must bereplaced based on the qualified life.
Some advanced methods for IS&M on mitigatingIGSCC in piping developed by EPRI have includedstress improvement methods, alternate materials,and improved water chemistry. Another advancedmonitoring method developed by the INEL forcable monitoring is the ECCAD system. This tech-nology has been transferred to industry and at leasttwo operating utilities have used the system.Another advanced surveillance system is themotor-operated valve analysis and test system(MOVATS). Since 1984, the MOVATS system hasbeen used as a diagnostic tool for motor-operatedvalves.
Maintenance
Maintenance practices for major HPIS compo-nents follow vendor's recommendations closely, inaddition to plant specific maintenance procedures.Corrective maintenance is minimized by observinggood operating practices and precautions. The ben-efits of preventive maintenance include higher sys-tem availability, increased life, and higherreliability.
In general, up to 35% of nuclear plant abnormaloccurrences may be due to faulty maintenance andsurveillance testing. In order for preventive mainte-nance to be effective, it must be applied to equip-ment that has degradation effects that aredetectable. Also, methods of detecting degradationmust be available.
For the HPIS, the recommendation is to con-tinue the maintenance recommended by the ven-dors for major components such as pumps andvalves. In addition, an enhanced inspection pro-gram for detecting pipe leaks and cracks (includingnozzles) using advanced monitoring methodsshould be implemented.
HPIS Unavailability Assessment
This HPIS aging system unavailability assess-ment demonstrates the linear aging model tech-nique using failure cause data can be used to updatea representative plant PRA with time to calculatethe change of unavailability as the system ages.This approach, as presented, is suitable for explor-atory investigations and identification of compo-nents that contribute significantly to systemunavailability. The results of this assessment applyto the plant studied and only to other plants of the
54
same design configuration. More advanced agingreliability models with better methods of verifyingthat failure data exhibits aging and for calculatingthe time dependent failure rates are beingdeveloped.
Seven components involving 10 events were iden-tified as risk significant for the HPIS by the FussellVesely measure. These events were motor-operatedvalves HP-24, HP-25, and HP-26; failure to oper-ate, HPIS pumps A, B, and C; failure to start orrun, and manual valve HP-148; failing to transfer.
The HPIS unavailability increased over the agingperiod modeled from 5 to 40 years for all three
emergency operating modes. Even after theincreases, the unavailabilities of the system are notmuch higher for the high pressure injection modesand the full plant PRA would not be expected tochange significantly. For the recirculation mode,the unavailability increased significantly, but evenafter the increase, the value remained very low. Acursory review of the representative plant PRAindicated this change would not be expected tochange the results of the full plant PRA.
The equipment failures that have the highest con-tribution to HPIS unavailability are the motor-operated valves failure to open.
55
REFERENCES
1. NuclearPlantAging Research (NPAR) Program Plan, NUREG-1 144, U.S. Nuclear Regulatory Com-mission, 1985.
2. W. L. Greenstreet, G. A. Murphy, D. M. Eissenberg, Aging and Service Wear of Electric Motor-Operated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants, NUREG/CR-4234, Vol. 1 (ORNL-6170/Vl), Martin Marietta Energy Systems, Inc., Oak Ridge NationalLaboratory, June 1985.
3. J. L. Crowley and D. M. Eissenberg, Evaluation of the Motor-Operated Valve Analysis and Test Sys-tem (MOVATS) to Detect Degradation, Incorrect Adjustments, and Other Abnormalities in Motor-Operated 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, EGG-2467, Idaho National Engineering Laboratory, January 1988.
6. B. M. Meale and D. G. Satterwhite, An Aging Failure Survey of Light Water Reactor Safety Systemsand 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. Commer-cial 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 E79-19.
12. S. H. Bush, P. G. Nessler, R. E. Dodge,Aging and Service Wear ofHydraulic and Mechanical Snub-bers 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, AmericanSociety 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 88-08Supplement No. 1. June 24, 1988.
18. M. R. Dinsel, M. R. Donaldson, F. T. Soberano, In Situ Testing of the Shippingport Atomic PowerStation Electrical Circuits, NUREG CR-3956, Idaho National Engineering Laboratory, April 1987.
19. QualityAssuranceProgram Requirements (Operation), Regulatory Guide 1.33, U.S. Nuclear Regula-tory Commission, February 1978.
20. ANS-3.2 Administrative Controls and Quality Assurance for the Operational Phase of Nuclear PowerPlants 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 DocumentRoom).
22. D. J. Sheliga, "Calculation of Optimum Preventive Maintenance Intervals for Electrical Equip-ment," 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 ofElectrical 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 ItsExtensions, 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, NorthCarolina, 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'sGuide-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 Systemsand 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 normaland emergency operation is shown in Figure A-1.During normal operation, the High Pressure Injec-tion system receives reactor coolant as letdownwater for purification and chemistry control. Theletdown water is cooled in one or both of the paral-lel connected letdown coolers in order to preventdamage to the purification system ion exchangeresins. Cooling water for the letdown coolers is pro-vided by the Component Cooling system. Thecooled letdown is routed through the N16 DecayTank, which provides sufficient flow delay to allowfor decay of radioactive nitrogen-16. The RCmakeup system is shown in Figure A-2. (The N-16decay tank is not shown on Figure A-2.)
Flow from the decay tank to the purification sys-tem is controlled by a fixed block orifice. In theevent additional flow is necessary, an orifice bypassvalve is provided.
In the purification system, the letdown streamenters the purification demineralizer through thepurification demineralizer supply header, andleaves through the purification demineralizer dis-charge header. A valve is provided between the sup-ply and discharge headers to allow thedemineralizer to be bypassed if necessary. A three-way valve in the discharge header allows the demin-eralizer discharge to be directed to either theletdown filters or the debarking demineralizer sup-ply header. From the debarking demineralizer sup-ply header, the letdown can be sent to the CoolantTreatment system or through the debarking demin-eralizer. The debarking demineralizer effluent issent via the debarking demineralizer dischargeheader to the letdown filters.
Feed and bleed controls allow batch addition ofdemineralized water and/or concentrated boricacid for control of RC system boric acid concentra-tion and restores the letdown storage tank level asnecessary. The makeup batch enters the systemupstream of the letdown filters.
Letdown from the purification system and/ormakeup from the Coolant Treatment system passesthrough one or both of the two parallel connectedletdown filters to the letdown storage tank.
Downstream from the letdown filters, returnwater from the reactor coolant pump seals is sup-
plied to the letdown storage tank. The seal returnwater is taken from each of the four RC pumps andis cooled by one or both of the parallel connect RCseal return coolers before entering the letdown stor-age tank. The seal return and injection system isshown in Figure A-3.
The letdown storage tank serves as the normalsuction source for the HP injection pumps; how-ever, during safety injection, the borated waterstorage tank is automatically connected as an addi-tional source. These sources provide injectionwater until RC system pressure is low enough forlow pressure injection.
The three high pressure injection pumps are con-nected in parallel with cross connecting suction anddischarge headers thus enabling any pump to sup-ply any discharge line. During normal operation,only one pump is required for normal makeup andreactor coolant pump seal injection. A secondpump serves as a backup for normal operation. Allthree HPI pumps are started automatically by a sig-nal from the engineered safeguards system in theevent of a low RC system pressure or a high reactorbuilding pressure condition.
Three principal flow paths are provided from theHPI pump discharge header: an injection line toreactor coolant Loop A, a reactor coolant pumpseal injection line, and an injection line to reactorcoolant loop B. Normally HPI pump A or B isoperated to supply makeup flow to RC loop A andRC pump seal injection flow; pump C is used onlyfor injection to loop B in the event of an accident.An additional emergency cross connect line is pro-vided from the HPI pump discharge header to bothRC loops as a backup for safety injection.
The water supplied for reactor coolant pump sealinjection is routed through the two parallel con-nected seal filters before being distributed to thefour reactor coolant pumps.
Description of Components
Letdown Coolers. The letdown coolers are ofthe shell and spiral tube type with a shell materialof carbon steel and a tube material of stainlesssteel. The two letdown coolers are installed in
A-3
188 ES2 open 148 114 113 \__.} 5Orifice X ~~~~~~~HP-P3C 1Orifice ES-2 on
From LPI 3LP-5#cooler discharge _
I (LPR system) 3LP-57Inside 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-21R~CP.3A2 -LES 2)
To seat retufn
RCP-381- SHP-20 coolers an~d LOSTRCP.38l (ES1).l)
RS
AtCP soa "Ir
SHP.1 32
RCP sealI 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 asa spare. For maximum letdown flow, both coolersare required.
Seal Return Coolers. The seal return coolers areof the shell and tube type with carbon steel as theshell material and stainless steel as the tube mate-rial. Of the two coolers installed in parallel, one isnormally in operation.
Makeup Filters. The material of all wetted partsof the filter vessel is stainless steel. The filter ele-ments themselves are of depth type design. Thematerials are wound cotton around a stainless steelcore. Two makeup filters are installed in parallelwith one normally in use.
Reactor Coolant Pump Seal Filters. The filterassembly consists of the filter vessel and one ele-ment. The filter vessel is of the in-line type withflanged ends. The element is pleated stainlesssteel. One of the two filters is required duringnormal operation. A bypass line is installedaround the filters.
Letdown Storage Tank. The total volume of thestainless steel letdown storage tank is 600 ft3. Thewater content under steady state conditions is about350 ft3 while the rest is gas space. The tank isequipped with a manway, 4-in. inlet and outlet con-nections, a 1-in. relief valve connection, a 1-in. ventconnection, two 3/4-in. sampling connections (one inthe gas space, the other in the water space), a 1-in.inlet connection for H2 and N2 and connections forlevel, pressure, and temperature indication.
The inlet connects to a nozzle that sprays theincoming water into the gas space. For hydrogenand nitrogen addition, a ring header in the waterspace is provided.
Purification Demineralizers. The purificationdemineralizers function is to purify the portion ofthe reactor coolant let down through the letdowncoolers. Each demineralizer is sized for a flow rateof 70 gpm.
The inlet flow is distributed in an inlet header.The outlet is equipped with a similar header havinga 110 mesh stainless steel screen to contain the resinbeads. 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 andoutlet. The material of all parts wetted by the proc-ess fluid is 304 stainless steel.
At the inlet line of each demineralizer, a smallcapacity relief valve is installed in order to preventpressure buildup during ambient temperaturechanges in the event that the demineralizer is iso-lated. The reactor coolant quality specifications aregiven in Table A-I.
Reactor Coolant Pump Seal Return Cooler
Type Shell and tubeHeat Transferred, Btu/hr 2.2 x 106Seal Return Flow, lb/hr 1.25 x 105Seal Return Temperature 145 x 127Change, 'FMaterial, shell/tube CS/SSDesign Pressure, psig 150Design Temperature,0 F 200Recirculated Cooling Water 1.25 x 105Flow (ea.), lb/hrCode ASME Sec. III-C & VIII
Deborating Demineralizers. The debarkingdemineralizers are used to reduce the boron con-centration in the reactor coolant during the latterpart of the core lifetime when the boron concentra-tion must be lower. Each demineralizer is sized for a70 gpm flow rate, the same as the purificationdemineralizers.
Letdown Storage Tank
Volume, ft3
Design Pressure, psigDesign Temperature,°FMaterialCode
600100200SSASME Sec. III-C
Purification Demineralizer
Type
Component Data
Letdown CoolerTypeHeat Transferred, Btu/hrLetdown Flow, lb/hrLetdown Cooler Inlet/OutletTemperature, F
Material, shell/tubeDesign Pressure, psigDesign Temperature,°FComponent CoolingWater Flow (ea.), lb/hrCode
MaterialVolume, ft3
Flow, gal/minVessel Design Pressure, psigVessel Design Temperature,0 FCode
Mixed bed, boric acid satu-ratedSs8570150200ASME Sec. III-C
Shell and spiral tube16.0 x 1063.5 x 104555/120
CS/SS2,5006002 x 105
ASME Sec. IlI-C & VIII
Letdown Filter
Design Flow Rate, gal/minMaterialDesign TemperatureDesign PressureCode
80Ss200150ASME Sec. III-C
Table A-1. Reactor coolant quality
Total solids, including dissolved and undissolved material, but excluding 7LiOHand 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 7 LiOH 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 intwo respects: (a) Under accident conditions they bothtake their suction from the borated water storage tank(the reactor building spray system also takes suctionfrom the BWST), and (b) for high pressure recirculationmode of operation, the HPIS takes suction from theLPIS output.
The HPRS mode is initiated when the BWST borated
water is depleted and conditions require high pressurerecirculation. Both HPI and LPI pumps are shut downif operating and the LPI valves are reconfigured byopening the two valves of the reactor building emer-gency sump (valves LP-19 and LP-20), starting at leastone LPI pump, closing valves LP-21 and LP-22 fromthe BWSI, and opening valves LP-15 and LP-16 to thesuction 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 valvesand HPI pumps are given in Table C-i. The IEsafety power system supplies the power for the HPI
components. Grid voltage is regulated to + 10°70.Each plant must decide what voltage degradationrepresents an acceptable level.
C-3
Table C-1. HPIS pumps and valve electrical power requirements
VoltageDescription Requirement
HPI Pump Aa 4 kVHPI Pump Ba 4 kVHPI Pump Ca 4 kV
HP-I, Letdown cooler A inlet isolation valve 208 VHP-2, Letdown cooler B inlet isolation valve 208 VHP-3, Letdown cooler A outlet valve 208 V
HP-4, Letdown cooler B outlet valve 208 VHP-5, Letdown isolation valve 125 VdcHP-6, Letdown orifice inlet valve 208/120 Vac
HP-8, Purification demineralizer inlet valve 208/120 VacHP-9, Spare purification demineralizer inlet valve 208/120 VacHP-I l, Spare purification demineralizer outlet valve 208/120 Vac
HP-13, Purification demineralizer bypass valve 208/120 VacHP-14, Bleed control valve 208 VHP-16, Makeup isolation valve 208 V
HP-17, Makeup filter isolation valve 208/120 VacHP-18, Makeup filter isolation valve 208/120 VacHP-19, Makeup filter bypass valve 208/120 ac
HP-20, RC pump seal return valve 208 VHP-21, RC pump seal return isolation valve 125 VdcHP-22, Letdown storage tank drain valve 208/120 Vac
HP-23, HPI pump suction valve 208 VHP-24, Borated water to HPI pump A valve 208 VHP-25, Borated water to HPI pump C valve 208 V
HP-26, HPI to loop A reactor inlet valve 208 VHP-27, HPI to loop B reactor inlet valve 208 VHP-98, HPI pumps suction crossover valve 208 V
HP-1 15, HPI pumps discharge crossover valve 208 VHP-226, RC pump A2 seal return isolation valve 208 VHP-228, RC pump Al seal return isolation valve 208 V
HP-230, RC pump B2 seal return isolation valve 208 VHP-232, RC pump BI seal return isolation valve 208 VHP-409, HPI isolation from HPI pump crossover valve 600 V
HP-410, HPI isolation from HPI pump crossover valve 600VGWD-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 waterswitchgear.
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 SafeguardsSystems that is automatically actuated by the Engi-neered Safety Feature Actuating System (ESFAS).The other safeguards systems actuated by ESFASinclude the low-pressure injection system, reactorbuilding isolation, reactor building cooling system,and the reactor building spray system.
The generic ESFAS diagram for a representativeB&W plant is shown in Figure D-1. Figure D-1 ispresented to illustrate the interconnections betweenmajor ESFAS subsystems. The three blocks on theleft side of Figure D-1 are identical analog subsys-tems that receive pressure transducer inputs. Theoutput lines from the analog subsystems go to thetwo identical center logic subsystems where the twoout of three logic decides whether an ESF system isactuated. On the right side of the figure are the fiveESF actuated systems. Although only the HPISwill be addressed in detail in this report, the othersystems are covered only to show system interac-tions. The simplified one-line diagram for initia-tion 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 variouscomponents. A detailed discussion of ESFAS iscovered in Reference D-1.
In case of a LOCA, the HPIS will be initiatedwhen the RC pressure drops to 1550 psig. If thehigh pressure injection channels fail to maintainRC pressure and it continues to decrease, then at
600 psig the core flooding system will automati-cally dump water into the core. This will happenautomatically, with no means for manual control.If the RC pressure continues to drop, at 550 psigthe low pressure injection channels (3 and 4) will beactuated in the same manner as the high pressurechannels. Anytime there is a large RC leak or rup-ture, the coolant will flash to steam as it escapesfrom the system. This will cause the building pres-sure to increase. When the building pressureincreases to 3.0 psig, the building cooling and iso-lation channels will be actuated as well as high andlow pressure injection channels if they have notalready been actuated.
The ESFAS Unit Control Module provides theoutput contacts to drive the final actuating devicecontrol circuitry. Interfacing capability is also pro-vided to allow manual control of the final deviceafter safety action has been initiated. If theremotely mounted manual control equipment isactivated before safety action is initiated, the sys-tem safety function will not be inhibited.
The engineered safeguards for actuated deviceson ESFAS channels 1 and 2 are shown inTable D-1, they include HPI pumps, valves, and IEpower breakers and busses.
Electrical power for operating the HPIS controlrelays is taken from the power source for the associ-ated device. Loss of power, a relay, or a device doesnot impair system functions because there is a sec-ond redundant device for each required function.
D-3
Wide-rangeRCS pressure
tranSmitter
Narrow-rangeRB pressuretransminer
RB pressureSwitches
Wide-rangeRCS pressure
transmitter
o Narrow-range- RMB pressure I
transminer
RB pressureSwi:Ches
Wide-rangePCS pressure
transmitter
Narrow-rangeRB pressure (traismitler
RB pressureSwitches
Pressure signals Trip signals Actuation signals
Figure D-1. Engineered safeguard system.
Reactor _ I ESPS cabinet inbuilding I control room
Manualbypass
InhibitHighpressureInjectionsystemactivationSee Note 1
Reactor building pressure IESPS analog subsystem 2
ESPS Channel 1See Note 2
Component ReactorcoolantalibratmBuffelOgi 2og3 tegt conitropressure power lCaibrate sBfler Bistable Bistable logic 23 test conitrol Comments
transmitter supply tetaplfeEnvironment
Temperature 120F average 50 to 80 F
Radiation 3 x 104 RAD INIA
p interfaces6" EO
Pressure TAP 110V power and module Interlock
10 year tile 1 40 yearsTesting 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
Physicaldata
Stressors See Note I
Indicators Drilt. Drift, failure. contact resistance change, binding or bent partsof moisturedegradation intrusion.
wearout,failure,signalvariance fromsimilar channels
1S9Note e: The transmitter, cable, reactor building penetration, terminal strip and connectors are Identical to thatof 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 engineeringdata.
....
Table D-1. Engineered safeguards actuated devices for Channels 1 and 2
Devices Channel I Channel 2 Channels 1 and 2
HPI pumps andvalves
HP-PAHP-24HP-26HP-3HP-4HP-20
HP-PCHP-25HP-27HP-5HP-21Emergency powersource start(Channel B)
HP-PB
Electrical powersources andEquipment
Emergency powersource start(Channel A)LOAD SHED &STBY. BKR. IStandby BUSFEED BKR. I
Emergency powersource start(Channel B)LOAD SHED &STBY. BKR. 2Standby BUSFEED 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 injec-tion lines are designed for normal operating condi-tions. The normal operating system temperature andpressure requirements are greater than those encoun-tered during emergency operation. The low pressureinjection system piping and valves are subjected tomore severe conditions during decay heat removaloperation than during emergency operation and, there-fore, operate well within the design conditions.Table E-1 gives the design pressure and temperatures ofthese systems. To ensure system integrity, major pipinghas welded connections except where flanges are dic-tated for maintenance reasons. All piping for the HPIsystem is stainless steel. Pipe sizes include 1, 2 1/2, 4,6, 8, and 14 in.
Generic guidance is needed for predicting and miti-gating and the effects of piping damage due to flow-assisted erosion, corrosion, or cavitation. Results from
the Surry-2 rupture on feedwater pipes show that ero-sion and corrosion can occur in single-phase flow.What lessons from that incident can be applied to HPIpiping is not yet known. Some factors that can effectpiping damage include pipe design, fluid dynamics,piping material, and water chemistry.
HPI Pumps
Each HPI pump can deliver 450 gpm at1700 psig reactor vessel pressure. Water is drawnthrough a single suction header from the BWSTand pumped through injection lines that enter thereactor building on opposite sides. Each injectionline splits into two lines inside the reactor building,but outside the secondary missile shield, to providefour 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 thestop check valves inside the secondaryshielding.
200 3,050
High pressure injection pump.
From upstream of the stop check valves tothe reactor inlet line.
Low Pressure Injection System(Portion used with HPI)
From the borated water storage tank toupstream of the borated water storage tankoutlet valves.
From upstream of the borated water storagetank outlet valve to upstream of the electricmotor operated valves in the borated waterfeed lines.
200/150 2,800/3,050
650 2,500
150 Static
200 100
E-3
connections are located between the reactor cool-ant pump discharge and the reactor inlet nozzles.
Operation of the HPIS pumps requires lubrica-tion oil cooling and pump seal cooling. Chargingpump cooling is accomplished via the Low PressureService Water (LPSW) System where heat gener-ated in the pump lubricating oil and seals isremoved via heat exchangers and transferred toultimate heat sink. Electrical power is supplied bythree independent 4160 V buses to the three HPIpumps. After receiving an actuation signal, theHPI pumps can reach full speed within six sec-onds. The HPI pump data are given in Table E-2and the HPI pump characteristics curves of totaldynamic head and net position suction head areshown in Figure E-1. The HPI pumps are verticalstage-centrifugal pumps with mechanical seals.The wetted parts of the pumps are stainless steel.
Valves
All remotely operated valves in the Emergency CoreCooling Systems are manufactured and inspected inaccordance with the intent of the ASME NuclearPower Piping Code B31.7. Liquid penetrant, radiogra-phy, ultrasonic, and hydrotesting are performed as thecode classification requires.
The seats and discs of these valves are manufac-tured from materials free from galling and seizing.
All valve material is certified to be in accordancewith ASTM specifications. All remotely operatedvalves in these systems are of the backseating typeand equipped with stem leak-off provisions.
The valves and their design conditions are listedin Table E-3. Actual system operating conditionsare significantly less severe than design conditionsshown in Table E-3.
1. Tank Problems - While there have been anumber of events at PWRs involving tankswith boron concentrations or fluid levels thatwere out of specification, most of these eventsdid not result in an outage or derating. Theonly significant event was an outage to replacea boric acid injection tank at a PWR. Theborated water storage tank (BWST) is the pri-mary water source for the HPI during theHPI emergency operation mode. A singleheader connects the borated water supply tothe charging pump. The pumps take suctionfrom the 350,000 gallon BWST (".2000 ppmboron). Specifications for the BWST aregiven in Table E4.
Thble E-5 gives the regulatory requirements andguidelines for the design equipment qualification andtesting of the HPIS.
Table E-2. High pressure injection pump data
]Type Vertical, multistage, centrifugalmechanical 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
300Capacity (gpm)
Figure E-1. HPI pump characteristics.
E-5
Table E-3. Active-HPIS and LPIS reactor coolant pressure boundary valves
ValveNumber System Service
Size(in.)
SystemDesignRating
SystemCondition
DuringOperation
MotorOperator
Type TypeValve
Movement
HP-3 High pressureinjection
HP-4 High pressureinjection
HP-5 High pressureinjection
HP-26 High pressureinjection
HP-27 High pressureinjection
LP-17 Low pressureLP-18 injection
Letdowncooleroutlet
Letdowncooleroutlet
Letdownline RBisolation
HP injRBisolation
HP injRBisolation
LP injRBisolation
2-1/2
2-1/2
2-1/2
4
4
10
2500 psig 2170 psig650xF I35xF
40-100 gpm
2500 psig 2170 psig650xF I35xF
40-100 gpm
2500 psig 1270 psig200xF 2950 psig
40-140 gpm
3050 psig 2200-200xF 2950 psig
120-245xF450 gpm
3050 psig 2200-200xF 2950 psig
450 gpm1 20-245xF
2500 psig 255 psig300xF 28OxF
3000 gpm
Globe Limitorque Full openSMB-00-15 to
Full close
Globe Limitorque Full openSMB-00-15 to
Full close
Globe Sheffer Full openPiston to
Full close
Globe Limitorque Full closeSMB-1-25 to
Full open
Globe Limitorque Full closeSMB-1-25 to
Full open
Gate Limitorque Full closeSMB-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 TitleApplicability
HPIS Remarks
Part I Design
ASME boiler and pressure vesselCode, Section Ill-C, VIII, and XI
1. - Life extension under reviewin ASME Section XI
2. General Design Criteria (GDC), Appendix Ato 10 CFR Part 50
b. GDC 2
c. GDC 4
d. GDC 13
e. GDC 19
j. GDC 24
-!a
Design basis for protection againstnatural phenomena
Environmental and missile designbasis
Instrumentation and control
Control room
Separation of protection and controlsystems
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-failurecriterion to nuclear power plantprotection systems
Manual initiation of protection actions
Physical independence of electricsystems
Environmental qualification of electricalequipment for nuclear power plants
Gb
G
G
G Positions on end of life andmaintenance
Table E-5. (continued)
ApplicabilityHPISCriteria Title Remarks
3. Regulatory Guides (RG) (continued)
g. RG 1.105 Instrument spands and setpoints G
h. RG 1.118 Periodic testing of electric power andprotection systems
G
Aging should be taken intoaccount
May require updating to takeinto account studies onincreasing surveillanceintervals
6. IEEE Standards
a. 279-1971 Criteria for protection system fornuclear power generating stations
60
R
Rb. 379-1977 Application of the single failurecriteria to NPGS Class IE systems
Part 2 Electrical equipment qualification
1. 10 CFR 50.49 R All replacement equipmentpurchased after 2/22/83
2. IEEE-323-1974 General guide for qualifying Class IEelectrical equipment for nuclear powergenerating stations (1971)
R
R3. IEEE 344-1975 Recommended practices for seismicqualification of Class IE electricalequipment for nuclear power generatingstations
All replacement equipmentpurchased after 2/22/83
Seismic qualification
Verify with naturally agedcomponents
4. Regulatory Guide 1.89 Environmental qualification ofelectric equipment for nuclear powerplants
G
Table E-5. (continued)
Criteria TitleApplicability
HPIS Remarks
5. NUREG-0588 Interim staff position onEnvironmental qualification ofsafety-related electrical equipment
G Update using naturally agedcomponent data
Part 3 Testing requirements
1. Standard Technical SpecificationsSection 3/4.5 and 4.5.11.1
M~2. IEEE-Std 338-1977
3. Regulatory Guide 1.68
Criteria for periodic testing of nuclearpower generating station safety systems
Initial test programs for water-coolednuclear 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 connectionswith thermal sleeves are located on all four cold legsof the reactor coolant piping. The four HPI/MUnozzles (one per cold leg) are used to: (a) provide acoolant source for emergency core cooling, and(b) supply normal makeup (purification flow) tothe primary system. In general, one or two of thenozzles are used for both HPI and MU, while theremaining nozzles are used for HPI alone.
The incorporation of a thermal sleeve into a noz-zle assembly is a common practice in the nuclearindustry to provide a thermal barrier between thecold HPI/MU fluid and the hot high pressureinjection nozzle. This helps prevent thermal shockand fatigue of the nozzle. The purpose of the safe-end is to make the field weld easier (pipe to safe-end) by allowing similar metals to be welded. Thedissimilar metal weld between the safe-end and thenozzle can then be made under controlled condi-tions in the vendor's shop. The use of the safe-endalso eliminates the need to do any post-weld heattreating in the field. Failures in these HPI/MU noz-zles may preclude the proper functioning of theECCS and/or the normal fluid makeup of the pri-mary system.
Make-Up Nozzle CrackingProblem
Cracks were found in the normal makeup highpressure injection (MU/HPI) nozzles of severalB&W plantsFl following an inspection of the eightB&W plants licensed to operate. These cracksappeared to be directly related to loose or missingthermal sleeves. As a result, a B&W Owners'Group Task Force was established to identify thecause of the failures and recommend modificationsto eliminate future failures.
The B&W Task Force completed a generic inves-tigation of the MU/HPI nozzle component crack-ing problem and has a report on the findings of thatinvestigation (Reference F-2). The report presentsrelevant facts and probable failure scenarios, aswell as recommended modifications to thermalsleeve designs, makeup system operating condi-tions, and ISI plans. Failure analysis indicated thatthe cracks were initiated on the inside diameter andwere propagated by thermal fatigue. Inspections atMidland have also shown that gaps may be presentbetween the thermal sleeve and safe-end in the con-tact expansion joint. These findings along withstress analysis and testing have implicated insuffi-cient contact expansion of the thermal sleeves asthe most probable root cause of the failures.
Possible Solutions
As a result of their investigation, B&W made thefollowing recommendationsFl for solving theproblem:
1. Reroll the upstream end of the thermalsleeve when inspections indicate that a gapexists, or repair and/or replace damagedcomponents
2. Maintain a continuous MU flow greaterthan 1.5 gpm
3. Implement an augmented ISI plan4. Perform a detailed stress analysis of a noz-
zle with a modified thermal sleeve designto justify long term operation.
The NRC staff reviewed and evaluated the rec-ommendations of the Task Force and supported allfour recommendations. The makeup HPI nozzle isshown in Figure F-1 with both the old and newdesign.
F-3
New thermal sleeve design
2.000" dia. ref.
F r 1.500" dia.
Thermal,sleeve
-I " j Safe end
| - , Contactexpanded
Old thermal sleeve design.010" to .015"diametrical clearancebetween thermalsleeve & 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 GenericSafety Issues, NUREG 0933, Issue 69, December 1983.
F-2. 177FuelAssembly Owners' Group Safe End Task Force Report on Generic Investigation of HPI/MUNozzle Component Cracking, B&W Document No. 77-1140611-00, Babcock and WilcoxCompany, 1983.
F-5
APPENDIX G
OPERATING EXPERIENCE DATA
G-1
APPENDIX G
OPERATING EXPERIENCE DATA
Nuclear Power Experience Datafor HPIS
Table G-2. I&C failures for HPI (NPE)
PWR Instrumentation(ho) (%0)The NPE data includes the following ESF sys-
tems under the heading Systems Injections Recircu-lation and Containment Spray. Valves, pipes, andmany of the major components will be similar forthese systems. The NPE data base could not besorted on subsystems to eliminate containmentspray events. Because of the similarity of compo-nents that handle borated water, this should not sig-nificantly affect the conclusions.
PWRs:
safety injection (SI)high head injection (HHI)(a.k.a. upper head inj - UHI)boron injectionrecirculation phasecontainment sprayaccumulator tank system
Component
ControlValvesPumpsElectrical andMiscellaneous
InstrumentationLevelPressureFlowHeat TracingTemperatureOther
Misc. and other
39
50
11
100
(17)( 9)(13)
(16)(32)(18)( 1)( 8)(26)
NOTE: Numbers in parentheses are a percentage breakdownfor instrumentation channels for all instrumentation failures.
For all U.S. Nuclear Power Plants, from startupthrough 1986, there were 1552 articles in NPE onHPIS for PWRs. The ranking for frequency ofHPI valve failures by function is given inTable G-1. The majority of valve failures (60%)were for stop (on/off) valves. Table G-2 gives asimilar ranking for type of I&C channel failures.About half are for instrumentation. A sub-breakdown is shown in parentheses for varioustypes of measurements.
Based on data from NPE, the causes for pipefailures are given in Table G-3. Design error, con-struction error, and maintenance error are the lead-ing causes for pipe failures. Pipe support (includingsnubbers) failure causes are broken down inTable G-4. Maintenance error, design error, water
Table G-3. Causes for pipe failures (NPE)
Table G-1. Valve failure ranking byfunction (NPE)
Cause Percent
PWRValve Function (%)
Stop (on/off) 60Check 12Control 8Containment isolation 8Relief 4Other 8No. of events 537
Design construction errorWeld failureMaintenance errorCorrosionVibrationBlockageMechanical disabilityWater hammerForeign materialEnvironmental effectsOther
2315147544332
201 00-
G-3
hammer, and mechanical disability are the leadingcauses for pipe support failures.
In the NPE data base, the heading tanks includeI&C as well as valves associated with the tanks. Per-centage of problems associated with tank subcom-ponents are shown in Table G-5 based on 194articles.
LER Data on Valves
The LER data for valves for the HPI and makeup system is shown in Table G-6.G-1 Commandfaults are those faults occurring in the control,power, or other support system that prevents thevalve from performing its intended function. Thisdata covered years 1976 through 1980.
Table G-4. Causes for pipe supportfailures (NPE)
Item
Maintenance errorDesign construction errorWater hammerMechanical disabilityVibrationWearoutWeld failureCorrosionBroken part/damageEnvironmentFatigueOther
Percent
2320201364322214
100
Table G-5. Subcomponents involved intank problems (NPE)
Item Percent
Water chemistry 34I&C 25Valve moving internals 17Seals, gaskets, packing 3Valve operators 2Fittings, flanges 2Other 17
100
Table G-6. Summary of valve failures and command faults for HPIS AND CVCS(LERs)
PWR
CommandFaultsFailures Total
System No. Percent No. Percent No. Percent
High pressure coolantinjection
Chemical volumecontrol (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. CommercialNuclear 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 experiencedin safety-related stainless steel piping systems andportions of systems containing oxygenated, stag-nant, or essentially stagnant borated water duringthe period from November 1974 to February 1977.Metallurgical investigations revealed these cracksoccurred in the weld heat affected zone of type 304stainless steel material. The cracks initiated on theID surface and propagated in a mode typical ofstress corrosion cracking. Analysis indicated theprobable corrosives to be chloride and oxygen con-tamination in the affected systems. Cracks werealso found as a result of local boric acid buildupand confirmed by liquid penetrant tests. Conclu-sions were drawn that cracking incidents were dueto IGSCC originating in the pipe ID. The crackingwas localized to the heat affected zone wheretype 304 stainless steel is sensitized to during weld-ing. More information can be found in the IEInformation Notice 79-19 on this problem. Noproblems were reported in 1980-82.
For the PWR plant used in the detailed HPI study,the system operation procedures stated a concern thatfollowing initial high pressure injection and HPI recir-culation boron may plate out in the core and blockcooling channels. Therefore, at a predetermined time,HPI flow is realigned from cold leg injection to hot leginjection. This reverses flow direction through the coreand dissolves any crystallized boron.
On April 20, 1987, the NRC Information NoticeNo. 86-109 alerted recipients of possible degradationof the reactor coolant system pressure boundary result-ing from boric acid corrosion. The boric acid corrosionof ferritic steel components in the RCS was firstnoticed at Arkansas One when inspecting HPI nozzlesand safe ends as a follow-up to the thermal stress crack-ing problem in 1982. Boric acid will rapidly corrodeferritic (carbon) steel components if a small leak occursnear a hot surface. The boric acid solution will boil andconcentrate, becoming more acidic and more corro-sive. In addition, the evaporation of the water willcause the boric acid crystals to accumulate at thatpoint. Other plants have experienced boric acid corro-sion of ferritic steel due to leakage of borated water.One involved threaded fasteners and is discussed in lENotice 82-02.
The boric acid corrosion is most active where themetal surface is cool enough to remain wet. If the sur-face were hot enough to dry out, the loss of electrolytewould slow the corrosion rate. Laboratory tests andplant experience have shown corrosion rates of > 1-in.depth per year in ferritic steel (ReferenceArticle 522 VII Safety System, A ECCS NPE).
Many PWRs have no positive means of detectingboron dilution during cold shutdown. There have been25 reported instances (Issue 22, NUREG-0933) ofinadvertent boron dilution during maintenance andrefueling. Although none have occurred, the possibilityof an inadvertent criticality is a paramount safety con-cern. Although undesirable, the consequences of anunmitigated boron dilution event is not severe enoughto warrant backfit of additional protective features atoperating plants. Such an event at an operating plantwould represent a breakdown in a licensee's ability tocontrol its plant.
If there is no clear indication of boric acid leakage tobolts, boric acid corrosion of bolts could go undetecteduntil failure of the bolts. Present ISI procedures do notrequire mandatory visual inspection of bolts and UTinspections are not required on pressure-retaining boltsof less than 2 in. in diameter. This problem was listedas safety issue 29, NUREG-0933 and given a high pri-ority ranking for resolution.
The most recent alert to boric acid problems is fromNRC information notices 86-108 supplement I, "Deg-radation of Reactor Coolant System Pressure Bound-ary Resulting from Boric Acid Corrosion." In thisinstance, about 500 lb of boric acid crystals were foundon the RV head as a result of a small leak in an instru-ment tube seal.
The loss of safety injection capability has occurredat two plantsH-1 as a result of common mode failureof SI pumps from crystallization of boric acid. In oneinstance, leaky valves in the discharge line of the boroninjection tank (BIT) enabled concentrated boric acid toflow through the low pressure discharge line (SI pumpsuction) and to precipitate in pumps not heat traced. Inthe other case, boric acid crystallization blockageoccurred between the boric acid storage tank (BAST)and the charging pump. Solidified boric acid has alsoblocked mixing tank pumpsH- 2 and level indicatorfailures 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 fromthe PRA'-I in Tables I-l, I-2, and 1-3. The descrip-tion of each event is given in these tables.
Fault tree importance measures for the threecases studied are given in Thbles I-4 through I-18.These are direct printouts from the IRRAS Pro-gram for the 5-year, 10-year upper and lowerbounds, and 40-year upper and lower bounds.
The fault tree cut set quantification reports for thethree cases are given in Tibles 1-19 through 1-33 for thesame time periods as the importance measures. Thetop of this display gives the family and fault tree namesalong with the minimal cut set upper bound value ascalculated by the point estimate quantification. Thelower portion showed the quantified minimal cut sets.
The first column gave the ranking of the cut setsaccording to the probability. The second column indi-cated an approximate percentage contribution of thecut set to the minimal cut set upper bound. The thirdcolumn gave the frequency of each individual cut setand the last column gave a listing of the basic eventsmaking up each cut set.
The code for the fault tree heading on Tables 1-4through 1-33 identifies the mode of operation thatapplies for the table.
HPI is for the injection that requires one of threepumps.
HP201 is for the injection mode that requirestwo of three pumps.
HPRI is for the recirculation mode.
1-3
Table 1-1. Basic events for the high-pressure injectionlrecirculation system: hardware
EventName Descriptionab
MeanUnavailabilityC
MeanFailureRate(yr-')
HPI0ICVO
HP102CVO
HP109CVO
Tilting-disk check valve3HP-101 fails to open ondemand
Tilting-disk check valve3HP-102 fails to open ondemand
Tilting-disk check valve3HP-109 fails to open ondemand
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 valve3HP-1 13 fails to open ondemand
Stop check valve 3HP-152fails to open on demand
Stop check valve 3HP-153fails to open on demand
Swing check valve 3HP-188fails to open on demand
Swing check valve 3LP-55fails to open on demand
Swing check valve 3LP-57fails to open on demand
Manual valve 3HP-148transfers 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-54transfers closed
Manual valve 3LP-56transfers closed
MOV 3HP-24 fails to openon demand
MOV 3HP-25 fails to openon 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)
EventName
HP26MVO
HP409MVO
HP410MVO
HPBPPS
HPCPPS
HPAPPR
HPRAPPR
HPBPPR
HPRBPPR
HPCPPR
HPRCPPR
HP15AVO
HP16AVO
HP12OUCF
HP120EPF
Descriptionab
MOV 3HP-26 fails to openon demand
MOV 3HP-409 fails to openon demand
MOV 3HP-410 fails to openon demand
Pump HP-P3B fails to starton demandd
Pump HP-P3C fails to starton demand
Pump HP-P3A fails to rune
Pump HP-P3A fails to runduring recirculatione
Pump HP-P3B fails to run
Pump HP-P3B fails to runduring recirculatione
Pump HP-P3C fails to run
Pump HP-P3C fails to runduring recirculatione
AOV valve 3HP-15 fails toopen on demandf
Piston-operated valve 3HP-6fails to open on demandd
AOV 3HP-120 control signalfails lowg
AOV 3HP-120 E/P moduleoutput fails lowh
MeanUnavailabilityc
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
MeanFailure
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)
EventName Descriptionab
MeanUnavailabilityc
MeanFailureRate(yr')
HP120VPF
CS55CVO
CS86CVO
HPRCBPPS
AOV 3HP-120 pneumaticpositioner output fails lowh
Tilting-disk check valve3CS-55 fails to open ondemand'
Tilting-disk check valve3CS-36 fails to open ondemandf
RC bleed-transfer pump 3Bfails to start ondemand)
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 starton low seal flowk
Flow transmitter controllingAOV 3HP-31 output failshighk
AOV 3HP-31 control signal toE/P module fails highl
Valve positioner output forAOV 3HP-31 fails highk
Output fails high from E/Pmodule for AOV-HP-31 k
Swing check valve 3CC-24fails to open on demand
Tilting-disk check valve3CC-14 fails to open ondemand
Swing check valve 3CC-20fails to open on demand
Pump CC-P3B fails to starton demandm
Pump CC-P3B fails to actuateon 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)
MeanFailure
Event Mean RateName 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 8.2-23CC-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 onefailure mode of interest, and the mean unavailability is the sum of the contributions from each failure mode, using an appropriatemean 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
EventName
HP2425MVH
HP26MVCH
HP27MVH
HP409MVH
HP410MVH
HPSEGOH
Descriptiona Event lxrpeb
U
MeanUnavailability C
5.0-5MOVs 3HP-24 and 3HP-25 leftunavailable
MOV 3HP-26 left unavailable
MOV 3HP-27 left unavailable
MOV 3HP-409 left unavailable
MOV 3HP-410 left unavailable
Manual valves 3HP-33 or3HP-100 left closedinadvertently
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 leftclosed inadvertently
Pumps HP-P3B and HP-P3Cleft unavailable
Operator fails to openMOV 3HP-24
Operator fails to openMOV 3HP-25
Operator fails to openMOV 3HP-26
Operator fails to open HPIcrossover valves, 3HP-409,3HP-H10
Operator fails to startpump HP-P3B
Operator fails to startpump HP-P3C
Operator fails to initiateletdown storage tank
Operator fails to tripRCPs on loss of coolingflow
U
U
OF(Recovery)
OF(Recovery)
OF(Recovery)
OF(Recovery)
OF(Recovery)
OF(Recovery)
OF
OF
6.1-4
1.5-4
1.0-21.0
1.0 -21.0
1.0-21.0
1.0-2
1.0-21.0
1.0-21.0
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
EventName
HP26MVM
HP409MVM
HP410MVM
HPSEGKM
MeanUnavailability
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 and3HP-103; check valve 3HP-105
Pump HP-P3B; manual valves 3HP-107 and3HP-1 10; check valve 3HP-109
Pump HP-P3C; MOV 3HP-27; manual valves3HP-111, 3HP-14 and 3HP-148; check valve3HP-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: HPIFault Tree: HPI
(Sorted by Fussell-Vesely)
Event Name
HP2425MVHHP24MVOHP25MVOHPSEGQMHPSEGHM
HPSEGPMHP101CVOHP102CVOHPSEGKMHPSEGMM
HPBCPPHHPCROSSHHP26MVOHPCPPSHP148VVH
HP148VVTHPCPPRHP188CVOHP27MVHHP26MVM
HP113CVOHP410MVOHP26MVCHHPSEGOHHPBPPS
HPAPPRHP409MVOHPBPPRHP109CVOHP410MVM
HP409MVHHP409MVMHP152CVOHP153CVO
Probabilityof
Failure
5.OOOE -0056.400E-0036.400E-0032.600E-0046.1OOE-004
2.600E-0048.700E-0058.700E-0051.500E-0031.500E-003
1.500E -004l.OOOE-0026.400E-0038.400E-0046.100E -004
3.900E - 0042.OOOE-0049.800E-0051.500E-0042.700E-004
8.700E-0056.400E - 0031.500E-0042.200E-0048.400E-004
2.000E-0046.400E-0032.OOOE-0048.700E-0052.700E-004
2.900E-0042.700E-0049.800E-0059.800E -005
Fussell-Vesely
Importance
5.213E -0014.503E -0014.502E-0013.523E -0022.046E-002
1.757E -0026.121E-0036.120E -0032.816E-0032.815E-003
2.451E -0032.212E -0032.145E-0036.308E -0044.581E -004
2.929E -0041.502E -0041.162E -0041.126E-0047.322E -005
6.533E -0055.414E-0055.027E - 0054.772E -0052.990E -005
9.328E - 0069.167E -0068.384E-0063.097E -0062.284E -006
4.154E- 0073.867E -0071.137E-0081.137E -008
RiskReduction
Ratio
2.089E + 0001.819E + 0001.819E + 0001.037E + 0001.021E + 000
1.018E+0001.006E + 0001.006E + 0001.003E + 000L.003E + 000
1.002E + 0001.002E + 0001.002E + 0001.OO1E+000l .OOOE + 000
I.OOOE + 000.OOOE + 000.OOOE + 000
1 .OOOE + 000l.OOOE + 000
1.000E + 0001.OOOE + 000
.OOOE + 000I .OOOE + 0001.000E + 000
1.OOOE + 0001 .000E + 0001.OOOE + 000l.OOOE + 0001.OOOE + 000
1.000E + 0001.OOOE + 000
.OOOE + 0001.OOOE + 000
RiskAchievement
Ratio
1.043E + 0047.088E + 0017.087E+0016.861E + 0011.699E + 000
6.856E + 0017.132E+0017.131E+0012.585E + 0001.005E + 000
1.734E+0011.219E + 0001.333E +0001.750E + 0001.750E + 000
1.751E+0001.751E+0002.186E+0001.751E+0001.271E + 000
1.751E+0001.008E + 0001.335E + 0001.217E+0001.036E + 000
1.047E + 0001.001E+000I.042E + 0001.036E + 0001.008E + 000
1.001E+000l.OOE + 0001.OOOE + 0001.OOOE + 000
1-10
Table 1-5. Fault tree HPI importance measures report(10 year upper bound)
Family: HPIFault Tree: HPI
(Sorted by Fussell-Vesely)
Event Name
HP24MVOHP25MVOHP2425MVHHPSEGQMHPSEGHM
HPSEGPMHPIOCVOHP102CVOHPCROSSHHP26MVO
HPSEGKMHPSEGMMHP 148VVTHPBCPPHHPCPPR
HPCPPSHP 148VVHHPAPPRHPBPPRHP26MVM
HP188CVOHP27MVHHPSEGOHHP1 13CVOHP26MVCH
HPBPPSHP41 OMVOHP409MVOHP109CVOHP410MVM
HP409MVHHP409MVMHP153CVOHP152CVO
Probabilityof
Failure
7.000E-0037.OOOE-0035.OOOE-0052.600E-0046.100E-004
2.600E-0049.100E-0059.100E-005I.OOOE-0027.OOOE-003
1.500E-0031.500E-0033.500E-0031.500E -0041. lOOE-003
1.1 OOE - 0036.1 OOE - 0041. lOOE-0031. lOOE-0032.700E-004
1.OOOE-0041.500E-0042.200E-0041.400E-0041.500E-004
1.1 OOE -0037.OOOE-0037.000E-0031.400E-0042.700E-004
2.900E-0042.700E -0041.OOOE-004I.OOOE-004
Fussell-Vesely
Importance
4.91 1E -0014.909E -0014.771E-0013.535E -0022.066E-002
1.758E-0026.383E -0036.382E -0035.265E -0035.058E -003
2.905E -0032.902E -0032.758E -0032.258E -0038.667E - 004
8.667E -0044.806E-0041.936E -0041.887E-0041.792E-004
1.224E -0041.182E -0041.145E -0041.1 03E -0041.084E -004
1.038E-0047.704E-0052.649E-0051.322E -0052.972E -006
1.097E -0061.022E -0061.221E -0081.221E -008
RiskReduction
Ratio
1.965E + 0001.964E + 0001.912E + 0001.037E + 0001.021 E + 000
1.018E +0001.006E + 0001.006E + 0001.005E + 0001.005E + 000
L.003E + 0001.003E + 0001.003E + 0001.002E + 000l.001E+000
l.OOlE+ 0001.000E + 0001.OOOE + 0001.000E + 0001.000E + 000
1.OOOE + 0001.OOOE + 0001.OOOE + 0001. OOOE + 0001.OOOE + 000
1.OOOE + 000.OOOE + 000
1.OOOE + 0001.OE + 0001.000E + 000
1.OOOE + 000L.OOOE + 0001.000E + 0001.OOOE + 000
RiskAchievement
Ratio
7.064E + 0017.062E + 0019.542E + 0036.863E + 0011.710E+000
6.858E + 0017.112E + 0017.11OE+0011.521E + 0001.717E+000
2.574E + 0001.071E +0001.785E + 0001.605E + 0011.787E + 000
1.787E + 0001.787E + 0001.176E + 0001.171E+0001.663E + 000
2.224E + 0001.788E + 0001.520E + 0001.788E + 0001.722E + 000
1.094E + 0001.011E+0001.004E + 0001.094E + 0001.O11E+000
1.004E + 0001.004E + 0001.OOOE + 000L.OOOE + 000
I-ll
Table 1-6. Fault tree HPI importance measures report(10 year lower bound)
Family: HPIFault RTee: HPI
(Sorted by Fussell-Vesely)
Event Name
HP2425MVHHP24MVOHP25MVOHPSEGQMHPSEGHM
HPSEGPMHPIOICVOHP102CVOHPCROSSHHP26MVO
HPSEGKMHPSEGMMHPBCPPHHP148VVTHPCPPS
HP148VVHHPCPPRHP188CVOHP27MVHHP26MVM
HP113CVOHP26MVCHHPSEGOHHP410MVOHPBPPS
HPAPPRHPBPPRHP409MVOHP109CVOHP410MVM
HP409MVHHP409MVMHP152CVOHP153CVO
Probabilityof
Failure
5.000E-0056.500E-0036.500E-0032.600E-0046.100E-004
2.600E-0048.800E-0058.800E-0051.OOOE-0026.500E-003
1.500E-0031.500E-0031.500E-0041.300E-0039.100E-004
6.1OOE-0044.600E -0049.900E-0051.500E-0042.700E-004
9.500E-0051.500E-0042.200E-0046.500E- 0039.100E-004
4.600E-0044.600E -0046.500E-0039.500E -0052.700E-004
2.900E -0042.700E-0049.900E -0059.900E-005
Fussell-Vesely
Importance
5.135E-0014.572E -0014.571E -0013.527E -0022.052E -002
1.758E-0026.189E -0036.188E -0033.098E-0032.984E -003
2.840E-0032.839E -0032.417E-0039.896E -0046.927E - 004
4.643E - 0043.502E -0041.181E - 0041.142E -0041.069E-004
7.231E -0056.887E -0056.705E - 0055.962E -0054.937E -005
3.641E -0053.426E -0051.335E -0055.154E -0062.477E-006
5.958E -0075.547E -0071.167E -0081.167E - 008
RiskReduction
Ratio
2.055E + 0001.842E + 0001.842E + 0001.037E + 0001.021E + 000
1.018E+0001.006E + 0001.006E + 0001.003E + 0001.003E + 000
1.003E + 0001.003E + 0001.002E + 000l .001E + 0001.001E +000
.OOOE + 0001.OOOE + 0001.OOOE + 0001.OOOE + 0001.OOOE + 000
1.000E + 0001. OOOE + 0001.OOOE + 0001.000E + 0001.OOOE + 000
1.OOOE + 0001.OOOE + 000l.OOOE + 0001.OOOE + 0001.000E + 000
1.OOOE + 0001.OOOE + 0001.OOOE + 000L.OOOE + 000
RiskAchievement
Ratio
1.027E + 0047.085E + 0017.084E+ 0016.863E + 0011.701E+000
6.858E + 0017.130E+0017.129E+ 0011.307E+ 0001.456E + 000
2.588E + 0001.017E+0001.711E +0011.760E + 0001.761E+000
1.761E+0001.761E+ 0002.192E + 0001.761E+0001.396E + 000
1.761E+0001.459E + 0001.305E + 0001.009E + 0001.054E + 000
1.079E + 0001.074E + 0001.002E + 0001.054E + 0001.009E + 000
1.002E + 0001.002E + 0001 .OOOE + 000
.OOOE + 000
1-12
Table 1-7. Fault tree HPI importance measures report(40 year upper bound)
Family: HPIFault Tree: HPI
(Sorted by Fussell-Vesely)
Event Name
HP24MVOHP25MVOHP2425MVHHPSEGQMHP148VVT
HPSEGHMHPCROSSHHP26MVOHPSEGPMHPAPPR
HPBPPRHP1OICVOHP102CVOHPCPPRHPSEGKM
HPSEGMMHPCPPSHPBCPPHHPBPPSHP148VVH
HP26MVMHP113CVOHPSEGOHHP26MVCHHP410MVO
HP409MVOHP188CVOHP27MVHHP109CVOHP410MVM
HP409MVHHP409MVMHP152CVOHP153CVO
Probabilityof
Failure
1.OOOE -002I.OOOE-0025.000E-0052.600E-0042.200E-002
6.100E -004l.OOOE -0021.OOOE -0022.600E-0046.800E-003
6.800E-0031.200E - 0041.200E -0046.800E-0031.500E -003
1.500E-0032.700E-0031.500E-0042.700E-0036.100E -004
2.700E - 0044.500E-0042.200E-0041.500E-0041.OOOE-002
1.OOOE-0021.300E-0041.500E-0044.500E-0042.700E -004
2.900E-0042.700E-0041.300E-0041.300E-004
Fussell-Vesely
Importance
6.338E -0016.333E -0013.050E -0013.265E -0022.402E -002
2.354E-0022.135E -0022.080E -0021.605E -0021.150E-002
1.147E - 0027.604E -0037.599E-0037.424E -0036.959E -003
6.943E -0032.948E -0031.529E-0038.121E -0046.660E-004
5.510E-0044.913E -0044.685E - 0043.120E -0043.072E -004
2.190E-0041.687E -0041.638E-0041.353E -0048.295E - 006
6.351E -0065.913E -0062.176E -0082.176E-008
RiskReduction
Ratio
2.730E + 0002.727E + 0001.439E + 0001.034E + 0001.025E + 000
1.024E + 0001.022E + 0001.021E+ 0001.016E+0001.012E+000
1.012E+ 0001.008E + 0001.008E + 0001.007E + 0001.007E + 000
1.007E + 0001.003E + 0001.002E + 0001.Oil + 0001.001l+000
l.OOE + 0001.OOOE + 0001.OOOE + 0001.OOOE + 0001.OOOE + 000
1.OOOE + 000l.OOOE + 000
.OOOE + 0001.OOOE + 0001.OOOE + 000
1.OOOE + 000I .OOOE + 0001.OOOE + 0001.OOOE + 000
RiskAchievement
Ratio
6.372E + 0016.367E + 0016.102E + 0036.272E + 0012.068E + 000
1.918E + 0003.113E + 0003.059E + 0006.270E + 0012.679E + 000
2.676E + 0006.434E + 0016.429E + 0012.084E + 0003.900E + 000
2.358E + 0002.089E + 0001.119E+0011 .300E + 0002.09 1E + 000
3.040E + 0002.09 1E + 0003.129E + 0003.080E + 0001.030E + 000
1.022E + 0002.297E + 0002.092E + 0001.301E+ 0001.031E+O000
1.022E + 0001.022E + 0001.OOOE + 0001.OOOE + 000
I-13
Table I-S. Fault tree HPI importance measures report(40 year lower bound)
Family: HPIFault Tree: HPI
(Sorted by Fussell-Vesely)
Event Name
HP24MVOHP25MVOHP2425MVHHPSEGQMHPSEGHM
HPSEGPMHPCROSSHHP26MVOHPlOICVOHP102CVO
HP148VVTHPSEGKMHPSEGMMHPBCPPHHPCPPR
HPCPPSHPAPPRHPBPPRHP148VVHHP26MVM
HPBPPSHPSEGOHHP26MVCHHP27MVHHP188CVO
HP113CVOHP410MVOHP409MVOHP109CVOHP410MVM
HP409MVHHP409MVMHP152CVOHP153CVO
Probabilityof
Failure
7.100E-0037.1003E-0035.000E-0052.600E1-0046.100E-004
2.600E -0041.OOOE-0027.100E-0039.100E-0059.1003E-005
6.800E1-0031.500E1-0031.500E1-0031.500E - 0042.000E-003
1.400E1-0032.000E-0032.0003E-0036. 100E - 0042.700E1-004
1.400E-0032.200E1-0041.500E -0041.500E1-004I.OOOE-004
1.400E-0047.100E-0037.100E-0031.400E -0042.700E1-004
2.900E1-0042.700E-004I.OOOE-0041.OOOE-004
Fussell-Vesely
Importance
4.958E -0014.956E1-0014.684E -0013.5311E-0022.121E1-002
1.751E-0028.4103E-0038.048E1-0036.354E -0036.352E-003
5.705E1-0033.422E -0033.417E -0032.226E -0031.678E1-003
1.175E-0037.648E1-0047.558E1-0045.118E-0042.9031E -004
2.184E1- 0041.831E-0041.700E1-0041.258E-0041.226E -004
1.175E-0049.537E -0054.373E -0052.184E1-0053.627E1-006
1.786E -0061.663E -0061.223E1-0081.223E -008
RiskReduction
Ratio
1.983E+0001.982E + 0001.881E +0001.037E + 0001.022E + 000
1.018E+0001.008E + 0001.008E + 0001.006E + 0001.006E + 000
1.006E + 0001.003E + 0001.003E + 0001.002E + 0001.002E + 000
1.001E+0001.001E+0001.001E + 0001.001E± +000
.OOOE + 000
1.OOOE + 000.OOOE + 000
1.OOOE + 000.OOOE + 000
1.OOOE + 000
.OOOE + 000.OOOE + 000
1.OOOE + 0001.0OOOE + 000
.OOOE + 000
.OOOE + 0001.OOOE + 000I.OOOE + 000
.OOOE + 000
RiskAchievement
Ratio
7.031E+0017.028E + 0019.369E + 0036.834E + 0011.733E1+000
6.8301+0011.833E+±0002.125E+30007.079E + 0017.077E + 001
1.833E1+0002.770E + 0001.210E + 0001.584E + 0011.837E + 000
1.838E+0001.382E+±0001.377E + 0001.838E+0002.075E + 000
1.156E+±0001.832E1+0002.133E+0001.839E1+0002.226E + 000
1.839E+±0001.013E + 0001.006E + 0001.156E+±0001.013E + 000
1.006E + 0001.006E + 000
.OOOE + 0001.0OOOE + 000
1-14
Table 1-9. Fault tree HP201 importance measures report(5 year)
Family: HPIFault Tree: HP201
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHP2425MVHHP24MVOHP25MVO
HPSEGHMHPSEGQMHPSEGKMHPSEGMMHP26MVM
HPSEGPMHP26MVCHHPSEGOHHPCPPSHPBPPS
HP 148VVHHPBPPRHP148VVTHPlOlCVOHP102CVO
HPBCPPHHPAPPRHPCPPRHP409MVOHP410MVO
HP27MVHHP113CVOHP109CVOHP409MVHHP409MVM
HP410MVMHP188CVOHP153CVOHP152CVO
Probabilityof
Failure
1.OOOE-0026.400E-0035.000E-0056.400E-0036.400E-003
6.1 OOE - 0042.600E-0041.500E-0031.500E-0032.700E-004
2.600E-0041.500E-0042.200E -0048.400E-0048.400E -004
6. 100E -0042.000E-0043.900E -0048.700E -0058.700E-005
1.500E-0042.000E-0042.000E-0046.400E -0036.400E -003
1.500E-0048.700E -0058.700E -0052.900E-0042.700E-004
2.700E -0049.800E -0059.800E-0059.800E-005
Fussell-Vesely
Importance
3.967E -0013.738E -0012.906E-0012.592E-0012.510E -001
5.964E-0023.568E -0023.509E -0023.376E -0021.575E -002
1.013E -0028.760E-0038.642E-0038.432E -0037.326E -003
6.123E -0035.505E -0033.915E -0033.523E -0033.412E-003
2.790E -0032.036E -0032.008E-0031.854E-0031.798E -003
1.506E -0038.733E -0047.588E -0048.401E- 0057.821E -005
7.584E -0056.516E -0056.363E -0096.363E -009
RiskReduction
Ratio
1.657E + 0001.597E + 0001.410E + 0001.350E + 0001.335E + 000
1.063E + 0001.037E + 0001.036E + 0001.035E + 0001.016E+ 000
1 .010E + 0001.009E + 0001.009E + 0001.009E + 0001.007E + 000
1.006E + 0001.006E + 0001.004E + 0001.004E + 0001.003E + 000
1.003E + 0001.002E + 000L.002E + 0001.002E + 0001.002E + 000
1.002E + 0001.00E+ 0001.001E + 000
.OOOE + 0001 .OOOE + 000
1 .OOOE + 000I.OOOE + 000
.OOOE + 000
.OOOE + 000
RiskAchievement
Ratio
4.025E + 0015.902E + 0015.813E + 0034.122E + 0013.995E + 001
2.178E+0003.869E + 0018.642E + 0001.622E+0015.933E +001
3.993E + 0015.939E + 0014.026E + 0011.103E+0019.714E + 000
1.103E+0012.847E + 0011.103E + 0014.147E +0014.020E + 001
1.958E + 0011.118E+0011.103E +0011.288E + 0001.279E + 000
1.103E+0011. 104E +0019.721E + 0001.290E + 0001.290E + 000
1.281E+0001.665E + 000I .OOOE + 0001 .OOOE + 000
1-15
Table 1-10. Fault tree HP201 importance measures report(10 year upper bound)
Family: HPIFault Tree: HP201
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHP24MVOHP25MVOHP2425MVH
HPSEGHMHPSEGKMHPSEGMMHPBPPRHP148VVT
HPSEGQMHPAPPRHPCPPRHPCPPSHP26MVM
HPSEGPMHPBPPSHPSEGOHHP148VVHHP26MVCH
HPIOICVOHP102CVOHPBCPPHHP409MVOHP410MVO
HP27MVHHP113CVOHP IO9CVOHP409MVHHP409MVM
HP188CVOHP410MVMHP153CVOHP152CVO
Probabilityof
Failure
l.000OE-0027.OOOE-0037.000E-0037.000E-0035.000E-005
6.100E-0041.500E-0031.500E-0031.I OOE - 0033.500E -003
2.600E-0041.I OOE - 0031. IOOE - 0031. IOOE - 0032.700E-004
2.600E-0041. IOOE - 0032.200E-0046.100E-0041.500E-004
9.100E-0059.100E-0051.500E-0047.000E-0037.OOOE-003
1.500E-0041.400E -0041.400E-0042.900E-0042.700E-004
l.OOOE - 0042.700E-004L.OOOE-004I.OOOE-004
Fussell-Vesely
Importance
3.622E -0013.428E-0012.581E - 0012.506E-0012.435E -001
1.009E-0017.965E -0027.848E -0025.335E -0024.507E -002
3.144E-0021.462E -0021.417E -0021.417E -0021.321E - 002
9.249E -0038.080E -0037.896E -0037.855E -0037.345E -003
3.355E -0033.258E -0033.007E-0032.469E -0031.951E- 003
1.932E -0031.803E-0031.028E -0031.023E -0049.524E -005
7.990E -0057.524E-0057.908E -0097.908E - 009
RiskReduction
Ratio
1.568E + 0001.522E+0001.348E + 0001.334E + 0001.322E + 000
1.11 2E + 0001.087E + 0001.085E + 0001.056E + 0001.047E + 000
1.032E + 0001.015E+0001.014E + 0001.014E + 0001.013E+000
1.009E + 0001.008E +0001.008E + 0001.008E + 0001.007E + 000
1.003E + 0001.003E + 0001.003E + 0001.002E + 0001.002E + 000
1.002E + 0001.002E + 0001.001E + 000
.OOOE + 0001.000E + 000
I.OOOE + 000.OOOE + 000
1.OOOE + 000.OOOE + 000
RiskAchievement
Ratio
3.685E + 0014.963E + 0013.759E + 0013.654E+ 0014.871 E + 003
6.431E+0001.369E + 0013.900E + 0014.925E + 0011.382E + 001
3.552E + 0011.427E + 0011.385E + 0011.385E + 0014.991E + 001
3.655E + 0018.337E + 0003.687E + 0011.386E+0014.996E + 001
3.785E + 0013.678E+0012.101E +0011.350E+0001.277E + 000
1.387E + 0011.387E + 0018.344E + 0001.353E + 0001.353E+ 000
1.799E + 0001.279E + 000I .OOOE + 0001 .OOOE + 000
1-16
Table 1-11. Fault tree HP201 importance measures report(10 year lower bound)
Family: HPIFault Tree: HP201
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHP2425MVHHP24MVOHP25MVO
HPSEGHMHPSEGKMHPSEGMMHPSEGQMHPBPPR
HP26MVMHP148VVTHPCPPSHPSEGPMHPSEGOH
HP26MVCHHPBPPSHP148VVHHPAPPRHPCPPR
HPlOICVOHP102CVOHPBCPPHHP409MVOHP410MVO
HP27MVHHP113CVOHP109CVOHP409MVHHP409MVM
HP410MVMHP188CVOHP153CVOHP152CVO
Probabilityof
Failure
l.OOOE-0026.500E - 0035.OOOE-0056.500E-0036.500E-003
6.100E-0041.500E-0031.500E-0032.600E-0044.600E-004
2.700E-0041.300E-0039.1OOE-0042.600E-0042.200E-004
I .500E -0049.100E-0046.100E-0044.600E-0044.600E -004
8.800E-0058.800E-0051.500E-0046.500E-0036.500E-003
1.500E-0049.500E-0059.500E-0052.900E-0042.700E-004
2.700E-0049.900E-0059.900E-0059.900E-005
Fussell-Vesely
Importance
3.874E -0013.653E-0012.796E -0012.569E-0012.489E -001
7.407E -0025.031E- 0024.902E-0023.463E -0021.606E-002
1.516E-0021.448E-0021.014E-0029.890E - 0038.442E-003
8.429E -0037.648E-0036.795E -0035.231E - 0035.124E -003
3.478E -0033.370E -0032.906E-0031.981E - 0031.810OE-003
1.671E -0031.058E -0037.984E - 0048.836E -0058.227E -005
7.517E-0057.026E -0056.912E -0096.912E-009
RiskReduction
Ratio
1.632E + 0001.576E + 0001.388E+0001.346E + 0001.331E+000
1.080E + 0001.053E + 0001.052E + 0001.036E + 0001.016E+000
1.015E+0001.015E+0001.OIOE+0001.O1OE+0001.009E + 000
1.009E + 000L.008E + 0001.007E + 0001.005E + 0001.005E + 000
1.003E + 0001.003E + 0001.003E + 0001.002E + 0001.002E + 000
1.002E + 0001.001E+ 0001.001E+ 0001.OOOE + 000I .OOOE + 000
I.OOOE + 000.OOOE + 000
1 .OOOE + 0001.OOOE + 000
RiskAchievement
Ratio
3.934E + 0015.683E + 0015.593E + 0034.025E + 0013.904E + 001
3.607E + 0001.036E + 0012.402E + 0013.783E+ 0013.581E + 001
5.713E + 0011.212E + 0011.212E + 0013.902E + 0013.935E + 001
5.718E + 0019.397E + 0001.213E+ 0011.236E + 0011.213E + 001
4.050E + 0013.928E + 0012.035E + 0011.303E + 0001.277E + 000
1.213E + 0011.213E + 0019.403E + 0001.305E+0001.305E + 000
1.278E + 0001.710E+000
.OOOE + 000I .OOOE + 000
1-17
Table 1-12. Fault tree HP201 importance measures report(40 year upper bound)
Family: HPIFault Tree: HP201
(Sorted by Fussell-Vesely)
Event Name
HPBPPRHP 148VVTHPCROSSHHP24MVOHP25MVO
HP26MVOHPSEGHMHPSEGKMHPSEGMMHPAPPR
HPCPPRHP2425MVHHPCPPSHPSEGQMHP148VVH
HPBPPSHP409MVOHP113CVOHP26MVMHPSEGPM
HPSEGOHHP26MVCHHPBCPPHHPIOICVOHP27MVH
HP102CVOHP410MVOHP109CVOHP409MVHHP409MVM
HP188CVOHP410MVMHP152CVOHP153CVO
Probabilityof
Failure
6.800E - 0032.200E -0021.OOOE3- 0021.OOE - 0021.OOOE3- 002
I. OOOE -0026. 1 OE -0041.5SOOE- 003I. SOOE -0036.800E - 003
6.800E -0035.000OE- 0052.700E -0032.600E -0046. 100E -004
2.700E -0031.00E - 0024.50E - 0042.700E -0042.600E -004
2.200E -0041.50E - 0041 .500E - 0041.200E -0041.50E - 004
1.200E - 004I.000OE -0024.50E - 0042.900E -0042.700E -004
1.300E -0042.700E -0041.300 - 0041.300E -004
Fussell-Vesely
Importance
4.584E - 0013.008E-0011.726E-0011.710E -0011.67SE-001
1.632E - 0011.370E - 0011.289E - 0011.283E - 0019.870E -002
9.296E-0028.068E-0023.691E -0021.298E -0028.339E -003
6.779E -0036.224E -0036.152E -0034.401E - 0034.334E-003
3.716E-0032.447E-0032.405E-0032.052E -0032.051E -003
2.010E -0031.861E-0031.130E-0031.805E -0041.680E -004
1.1 19E -0045.024E-0051.428E-0081.428E- 008
RiskReduction
Ratio
1.846E + 0001.430E + 0001.209E +0001.206E + 0001.201E+ 000
1.195E + 0001.159E + 0001.148E+0001.147E + 0001.llOE+000
1.102E + 0001.088E+ 0001.038E + 0001.013E+0001.008E + 000
1.007E + 0001.006E + 0001.006E + 0001.004E + 0001.004E + 000
1.004E + 0001.002E + 0001.002E + 0001.002E + 0001.002E + 000
1.002E + 0001.002E + 0001.001E + 0001.000E + 0001.000E + 000
1.OE + 0001.OOOE + 0001.000E + 0001.OOOE + 000
RiskAchievement
Ratio
6.703E + 0011.435E +-0011.808E-+ 0011.792E + 0011.758E1+001
1.715E+0011.216E + 0011.848E1+0016.398E + 0011.539E+ 001
1.456E+0011.614E + 0031.462E + 0011.732E± 0011.464E + 001
3.504E + 0001.616E+ 0001.465E+ 0011.729E + 0011.766E + 001
1.788E+0011.731E+0011.699E + 0011.809E + 0011.465E+001
1.774E+0011.184E+0003.509E + 0001.622E + 0001.622E + 000
1.860E+ 0001.186E+0001 .OOOE + 0001.OOOE + 000
I-18
Table 1-13. Fault tree HP201 importance measures report(40 year lower bound)
Family: HPIFault Tree: HP201
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHP24MVOHP25MVOHP2425MVH
HPSEGHMHPBPPRHPSEGKMHPSEGMMHP148VVT
HPAPPRHPCPPRHPSEGQMHPCPPSHP26MVM
HP148VVHHPBPPSHPSEGPMHPSEGOHHP26MVCH
HPBCPPHHP101CVOHP409MVOHP102CVOHP27MVH
HP113CVOHP410MVOHP lO9CVOHP409MVHHP409MVM
HP188CVOHP410MVMHP153CVOHP152CVO
Probabilityof
Failure
1.OOOE - 0027.100E-0037.100E-0037.100E-0035.000E -005
6.100E-0042.000E-0031.500E-0031.500E-0036.800E-003
2.000E-0032.OOOE-0032.600E-0041.400E-0032.700E-004
6.100E-0041.400E-0032.600E-0042.200E-0041.500E-004
1.500E-0049.100E-0057.100E -0039.100E-0051.S00E-004
1.400E-0047.100E-0031.400E-0042.900E-0042.700E-004
1.000E-0042.700E-0041.OOOE-004I.000E-004
Fussell-Vesely
Importance
3.196E -0013.018E-0012.302E-0012.236E -0012.113E-001
1.323E -0011.298E -0011.138E-0011.127E -0011.024E-001
3.137E -0023.010E -0022.751E -0022.107E -0021.147E-002
9.182E -0038.974E -0038.137E-0036.953E-0036.377E -003
3.191E -0032.951E -0032.895E-0032.866E -0032.258E -003
2.107E -0031.771E- 0038.974E -0041.182E -0041.I01E -004
8.655E -0056.733E -0058.529E -0098.529E -009
RiskReduction
Ratio
1.470E + 0001.432E + 0001.299E + 0001.288E + 0001.268E + 000
1.153E+ 0001.149E+0001.128E+0001.127E+0001.1 14E+ 000
1.032E + 0001.031E+0001.028E + 0001.022E + 0001.012E+000
1.009E + 0001.009E + 0001.008E + 000L.007E + 0001.006E + 000
L.003E + 000L.003E + 0001.003E + 0001.003E + 0001.002E + 000
1.002E + 0001.002E + 0001.001E+ 000I.000E + 0001.OOOE + 000
1.OOOE + 0001.000E + 0001.OOOE + 0001.000E + 000
RiskAchievement
Ratio
3.263E + 0014.321E + 0013.318E+0013.226E + 0014.228E + 003
9.572E + 0006.542E + 0011.744E + 0015.657E+ 0011.594E+001
1.664E + 0011.601E+0013.139E+0011.602E + 0014.346E + 001
1.603E + 0017.401E + 0003.228E + 0013.259E + 0014.350E + 001
2.223E + 0013.340E+0011.405E + 0003.248E + 0011.604E + 001
1.604E+ 0011.248E + 0007.409E + 0001.408E + 0001.408E + 000
1.865E1+0001.249E + 0001.OOOE + 0001.OOOE + 000
1-19
Table 1-14. Fault tree HPRI importance measures report(5 year)
Family: HPIFault 'ftee: HPRI
(Sorted by Fussell-Vesely)
Event Name
HPSEGMMHPSEGKMHPSEGHMHP26MVOHPCROSSH
HPBCPPHHPRAPPRHP2425MVHHP25MVOHP24MVO
HPRCPPRHPCPPSHP148VVHHP148VVTHPCPPR
Probabilityof
Failure
1.500E-0031.500E-0036.100E-0046.400E-0031.000E-002
1.500E-0042.000E-0035.OOOE-0056.400E-0036.400E-003
2.OOOE -0038.400E-0046.IOOE-0043.900E-0042.OOOE-004
Fussell-Vesely
Importance
4.757E - 0014.757E -0014.757E -0013.541E-0013.541E-001
2.673E - 0012.166E - 0012.081E-0011.705E -0011.705E -001
I .524E -0016.399E-0024.647E -0022.971E-0021.524E -002
RiskReduction
Ratio
1.907E + 0001.907E + 0001.907E + 0001.548E + 0001.548E+000
1.365E + 0001.276E+0001.263E + 0001.206E + 0001.206E + 000
1.180E + 0001.068E + 0001.049E + 0001.031E+0001.015E+000
RiskAchievement
Ratio
1l.OOIE + 0002.868E + 0027.691E + 0015.597E + 0013.606E + 001
1.783E + 0031.091E + 0024.160E + 0032.747E + 0012.747E + 001
7.703E + 0017.712E + 0017.714E + 0017.715E + 0017.717E + 001
1-20
Table 1-15. Fault tree HPRI importance measures report(10 year upper bound)
Family: HPIFault Tree: HPRI
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHPRAPPRHPRCPPRHP2425MVH
HP25MVOHP24MVOHPSEGKMHPSEGMMHPSEGHM
HP148VVTHPBCPPHHPCPPSHPCPPRHP148VVH
Probabilityof
Failure
1.OOOE-0027.OOOE-0031.I OOE - 0021. lOOE - 0025.OOOE-005
7.OOOE-0037.000E-0031.500E-0031.500E-0036.100E-004
3.500E-0031.500E -0041.I OOE -0031.100E-003
*6.100E - 004
Fussell-Vesely
Importance
4.618E -0014.618E -001,4.010E-0012.835E - 0012.301E -001
2.255E -0012.255E-0011.529E-0011.529E - 0011.529E-001
9.021E-0028.267E -0022.835E -0022.835E-0021.572E-002
RiskReduction
Ratio
1.858E + 0001.858E + 0001.669E + 0001 .396E + 0001.299E + 000
1.291E + 0001.291E+0001.181E+0001.181E+0001.181E + 000
1.099E + 0001.090E + 0001.029E + 0001.029E + 0001.016E+000
RiskAchievement
* Ratio
4.672E + 0016.651E+0013.705E + 0012.649E + 0014.596E + 003
3.298E + 0013.298E + 0019.234E+0011.OOOE + 0002.668E + 001
2.668E + 0015.520E + 0022.674E + 0012.674E + 0012.676E + 001
I-21
Table 1-16. Fault tree HPRI importance measures report(10 year lower bound)
Family: HPIFault Tree: HPRI
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHPRAPPRHPSEGHMHPSEGKM
HPSEGMMHP2425MVHHPRCPPRHP25MVOHP24MVO
HPBCPPHHP148VVTHPCPPSHP148VVHHPCPPR
Probabilityof
Failure
l.OOOE-0026.500E-0034.600E-0036.100E-0041. 500E -003
I.500E-0035.OOOE-0054.600E-0036.500E-0036.500E-003
1.500E-0041.300E-0039.1OOE-0046.100E-0044.600E-004
Fussell-Vesely
Importance
4.121E -0014.121E -0013.170E -0013.004E-0013.004E - 001
3.004E-0012.277E -0012.234E -0011.924E-0011.924E-001
1.677E -0016.312E - 0024.418E -0022.962E-0022.234E-002
RiskReduction
Ratio
1.701E+0001.701E + 0001.464E + 0001.429E + 0001.429E + 000
1.429E + 0001.295E + 0001.288E + 0001.238E+0001.238E+000
1.202E + 0001.067E +0001.046E + 0001.031E+0001.023E + 000
RiskAchievement
Ratio
4.180E+0016.399E + 0016.959E + 0014.938E + 0011.813E +002
1.OOOE + 0004.550E + 0034.933E + 0013.041E + 0013.041E + 001
1.119E+0034.949E + 0014.951E+0014.953E + 0014.953E+001
I-22
Table 1-17. Fault tree HPRI importance measures report(40 year upper bound)
Family: HPIFault Tree: HPRI
(Sorted by Fussell-Vesely)
Event Name
HPRAPPRHP26MVOHPCROSSHHP25MVOHP24MVO
HPRCPPRHP2425MVHHP148VVTHPCPPRHPSEGMM
HPSEGKMHPSEGHMHPCPPSHPBCPPHHP 148VVH
Probabilityof
Failure
6.800E-002Il.OOOE-0021.OOOE-0021.OOOE-002I.OOOE-002
6.800E-0025.OOOE-0052.200E-0026.800E-0031.500E-003
1.500E-0036.1 00E - 0042.700E - 0031.500E-0046.1 00E - 004
Fussell-Vesely
Importance
4.923E - 0014.861E -0014.861E -0013.354E-0013.354E -001
3.282E -0011.677E-0011.062E-0013.282E -0022.461E -002
2.461E- 0022.461E -0021.303E -0021.084E -0022.944E -003
RiskReduction
Ratio
1.969E + 0001.946E + 0001.946E + 0001.505E + 0001.505E + 000
1.488E + 0001.201E + 0001.1 19E + 0001.034E + 0001.025E + 000
1.025E + 0001.025E + 0001.013E+0001.OlIE+0001.003E + 000
RiskAchievement
Ratio
7.747E + 0004.91 1E + 0014.911E +0013.420E + 0013.420E + 001
5.498E + 0003.350E+0035.720E + 0005.793E + 0001.OOOE + 000
1.542E + 0015.809E + 0005.813E + 0007.323E + 0015.823E + 000
1-23
Table 1-18. Fault tree HPRI importance measures report(40 year lower bound)
Family: HPIFault Tree: HPRI
(Sorted by Fussell-Vesely)
Event Name
HPCROSSHHP26MVOHPRAPPRHPRCPPRHP25MVO
HP24MVOHP2425MVHHP148VVTHPSEGMMHPSEGKM
HPSEGHMHPBCPPHHPCPPRHPCPPSHP 148VVH
Probabilityof
Failure
l.OOOE-0027.100E -0032.OOOE-0022.OOOE-0027.100E- 003
7.1OOE-0035.OOOE-0056.800E-0031.500E - 0031.500E-003
6.100E-0041.500E-0042.000E - 0031.400E-0036.100E-004
Fussell-Vesely
Importance
4.835E -0014.835E -0014.353E -0013.078E-0012.349E -001
2.349E -0012.330E-0011.046E-0019.060E-0029.060E-002
9.060E -0024.866E -0023.078E -0022.154E -0029.387E -003
RiskReduction
Ratio
1.936E+0001.936E + 0001.771E+0001.445E + 0001.307E+ 000
1.307E+ 0001.304E + 0001.11 7E + 0001. IlOOE + 0001. lOOE + 000
1.1 OOE + 0001.051E+0001.032E + 0001.022E + 0001.009E + 000
RiskAchievement
Ratio
4.886E + 0016.861E+0012.233E + 0011.608E + 0013.385E+001
3.385E + 0014.654E + 0031.628E+ 0011.000E + 0005.508E + 001
1.633E+ 0013.254E + 0021.636E + 0011.637E + 0011.638E + 001
1-24
Table 1-19. Fault tree HPI cut sets quantification report(5 year)
Family: HPIFault Tree: HPI
Mincut Upper Bound 9.590E - 005
Cut % 1o CutNo. Total Set Freq. Cut Sets
1 52.1 52.1 5.OE-005 HP2425MVH2 94.8 42.7 4.1E-005 HP24MVO, HP25MVO3 96.6 1.7 1.7E-006 HP24MVO, HPSEGQM4 98.3 1.7 1.7E -006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM5 98.9 .6 5.6E-007 HP102CVO, HP24MVO
6 99.5 .6 5.6E-007 HPlOICVO, HP25MVO7 99.7 .2 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM8 99.8 .1 5.4E -008 HP26MVO, HPCPPS, HPCROSSH9 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, HPCROSSH12 99.9 .0 2.3E-008 HPIOlCVO, HPSEGQM13 99.9 .0 2.3E -008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM14 99.9 .0 1.3E -008 HP26MVO, HPCPPR, HPCROSSH
Table 1-20. Fault tree HPI cut sets quantification report(10 year upper bound)
Family: HPIFault 'lee: HPI
Mincut Upper Bound 1.048E -004
Cut % Oo CutNo. Total Set Freq. Cut Sets
1 47.7 47.7 5.OE -005 HP2425MVH2 94.5 46.8 4.9E-005 HP24MVO, HP25MVO3 96.2 1.7 1.8E-006 HP24MVO, HPSEGQM4 97.9 1.7 1.8E -006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM5 98.5 .6 6.4E -007 HP102CVO, HP24MVO
6 99.2 .6 6.4E -007 HPIOICVO, HP25MVO7 99.4 .2 2.4E -007 HP148VVT, HP26MVO, HPCROSSH8 99.6 .2 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM9 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, HPCROSSH12 99.8 .0 4.3E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM13 99.9 .0 2.4E-008 HPlOlCVO, HPSEGQM14 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: HPIFault ltee: HPI
Mincut Upper Bound 9.737E - 005
Cut % o CutNo. Total Set Freq. Cut Sets
1 51.4 51.4 5.OE-005 HP2425MVH2 94.7 43.4 4.2E - 005 HP24MVO, HP25MVO3 96.5 1.7 1.7E-006 HP24MVO, HPSEGQM4 98.2 1.7 1.7E - 006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM5 98.8 .6 5.7E-007 HPIOICVO, HP25MVO
6 99.4 .6 5.7E-007 HP102CVO, HP24MVO7 99.6 .2 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM8 99.7 .1 8.5E -008 HP148VVT, HP26MVO, HPCROSSH9 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, /HPSEGMM12 99.9 .0 3.E -008 HP26MVO, HPCPPR, HPCROSSH13 99.9 .0 2.3E-008 HPlOlCVO, HPSEGQM14 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: HPIFault Tree: HPI
Mincut Upper Bound 1.639E - 004
Cut QO %0 CutNo. Total Set Freq. Cut Sets
1 61.0 61.0 1.OE-004 HP24MVO, HP25MVO2 91.5 30.5 5.OE-005 HP2425MVH3 93.1 1.6 2.6E-006 HP24MVO, HPSEGQM4 94.7 1.6 2.6E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM5 96.0 1.3 2.2E-006 HP148VVT, HP26MVO, HPCROSSH
6 96.8 .7 1.2E-006 HPlOlCVO, HP25MVO7 97.5 .7 1.2E-006 HP102CVO, HP24MVO8 98.1 .6 L.OE-006 HP148VVT, HPAPPR, HPBPPR9 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, HPCROSSH12 99.0 .1 2.2E - 007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM13 99.2 .1 2.2E-007 HP148VVT, HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM14 99.3 .1 2.2E -007 HP148VVT, HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM15 99.4 .1 1.2E-007 HPAPPR, HPBPPR, HPCPPS
16 99.4 .1 8.9E -008 HP148VVT, HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM17 99.5 .0 6.9E -008 HPAPPR, HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM18 99.5 .0 6.9E -008 HPBPPR, HPCPPR, /HPSEGHM, HPSEGKM, /HPSEGMM19 99.6 .0 6.1E-008 HP148VVH, HP26MVO, HPCROSSH20 99.6 .0 6.1E -008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM
21 99.6 .0 5.9E -008 HP148VVT, HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM22 99.7 .0 4.8E-008 HP148VVT, HP24MVO, HPSEGOH23 99.7 .0 4.5E-008 HP113CVO, HP26MVO, HPCROSSH24 99.7 .0 3.3E-008 HP148VVT, HP26MVCH, HPCROSSH25 99.7 .0 3.1E-008 HPIOICVO, HPSEGQM
26 99.7 .0 3.E -008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM27 99.8 .0 2.8E-008 HP148VVH, HPAPPR, HPBPPR28 99.8 .0 2.8E -008 HPAPPR, HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM29 99.8 .0 2.7E -008 HPAPPR, HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM30 99.8 .0 2.7E - 008 HPBPPS, HPCPPR, /HPSEGHM, HPSEGKM, /HPSEGMM
31 99.8 .0 2.7E-008 HPBPPR, HPCPPS, /HPSEGHM, HPSEGKM, /HPSEGMM32 99.8 .0 2.2E-008 HP148VVT, HP26MVO, HP409MVO, HP410MVO33 99.9 .0 2.1E-008 HP113CVO, HPAPPR, HPBPPR34 99.9 .0 1.8E -008 HP26MVM, HPCPPR, HPCROSSH, /HPSEGHM, /HPSEGQM35 99.9 .0 1.5E-008 HP26MVO, HPBCPPH, HPCROSSH
36 99.9 .0 1.5E-008 HP26MVO, HP27MVH, HPCROSSH37 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: HPIFault Tree: HPI
Mincut Upper Bound 1.067E - 004
Cut % 50 CutNo. Total Set Freq. Cut Sets
1 47.2 47.2 5.OE - 005 HP24MVO, HP25MVO2 94.1 46.9 5.OE-005 HP2425MVH3 95.8 1.7 1.8E-006 HP24MVO, HPSEGQM4 97.6 1.7 1.8E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM5 98.2 .6 6.5E-007 HPIOICVO, HP25MVO
6 98.8 .6 6.5E-007 HP102CVO, HP24MVO7 99.2 .5 4.8E-007 HP148VVT, HP26MVO, HPCROSSH8 99.4 .2 2.2E - 007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM9 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, HPCROSSH12 99.7 .0 4.3E - 008 HP26MVO, HPCROSSH, HPSEGHM, /HPSEGKM, /HPSEGMM13 99.8 .0 2.7E-008 HP148VVT, HPAPPR, HPBPPR14 99.8 .0 2.4E-008 HPIOlCVO, HPSEGQM
Table 1-24. Fault tree HP201 cut sets quantification report(5 year)
Family: HPIFault flee: HP201
Mincut Upper Bound 1.720E - 004
Cut % % CutNo. Total Set Freq. Cut Sets
1 37.2 37.2 6.4E-005 HP26MVO, HPCROSSH2 66.3 29.1 5.OE-005 HP2425MVH3 90.1 23.8 4.IE-005 HP24MVO, HP25MVO4 91.7 1.6 2.7E - 006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM5 92.6 1.0 1.7E-006 HP24MVO, HPSEGQM
6 93.6 1.0 1.7E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM7 94.5 .9 1.5E -006 HP26MVCH, HPCROSSH8 95.3 .8 1.4E-006 HP24MVO, HPSEGOH9 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, HPSEGMM12 97.6 .3 5.8E-007 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM13 97.9 .3 5.6E-007 HP102CVO, HP24MVO14 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: HPIFault Tree: HP201
Mincut Upper Bound 2.053E - 004
Cut % !o CutNo. Total Set Freq. Cut Sets
1 34.1 34.1 7.OE-005 HP26MVO, HPCROSSH2 58.5 24.4 5.OE-005 HP2425MVH3 82.3 23.9 4.9E-005 HP24MVO, HP25MVO4 84.9 2.6 5.2E-006 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM5 86.7 1.9 3.8E-006 HP148VVT, HPBPPR
6 88.1 1.3 2.7E-006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM7 88.9 .9 1.8E - 006 HP24MVO, HPSEGQM8 89.8 .9 1.8E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM9 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, HPSEGMM12 93.0 .8 1.6E - 006 HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM13 93.8 .8 1.6E-006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM14 94.6 .8 1.5E - 006 HP24MVO, HPSEGOH15 95.3 .7 1.5E-006 HP26MVCH, HPCROSSH
16 95.9 .6 1.2E-006 HPAPPR, HPBPPR17 96.5 .6 1.2E-006 HPBPPR, HPCPPR18 97.1 .6 1.2E-006 HPBPPR, HPCPPS19 97.5 .4 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM20 97.9 .3 6.7E -007 HP148VVH, HPBPPR
21 98.2 .3 6.7E-007 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM22 98.5 .3 6.4E-007 HP102CVO, HP24MVO23 98.8 .3 6.4E-007 HPlOlCVO, HP25MVO24 99.0 .2 3.4E-007 HP26MVO, HP409MVO, HP410MVO25 99.1 .1 2.3E-007 HPBCPPH, HPSEGKM
26 99.2 .1 2.2E -007 HPBCPPH, /HPSEGHM, /HPSEGKM, HPSEGMM27 99.3 .1 2.2E -007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM28 99.4 .1 2.1E -007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM29 99.5 .1 2.IE-007 HPI13CVO, /HPSEGHM, /HPSEGKM, HPSEGMM30 99.6 .1 1.7E-007 HP27MVH, HPBPPR
31 99.7 .1 1.7E -007 HPBCPPH, HPBPPR32 99.7 .1 1.5E-007 HP113CVO, HPBPPR33 99.8 .0 5.7E -008 /HPSEGHM, HPSEGOH, HPSEGPM, /HPSEGQM34 99.8 .0 5.2E -008 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM35 99.8 .0 3.9E-008 HP148VVT, HPAPPR, HPCROSSH
36 99.8 .0 3.7E - 008 HP148VVT, HP409MVO, /HPSEGHM, HPSEGKM, /HPSEGMM37 99.8 .0 2.7E-008 HP148VVT, HP409MVO, HPAPPR38 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: HPIFault Tree: HP201
Mincut Upper Bound 1.788E - 004
Cut 0o qo CutNo. Total Set Freq. Cut Sets
1 36.4 36.4 6.5E - 005 HP26MVO, HPCROSSH2 64.3 28.0 5.OE - 005 HP2425MVH3 87.9 23.6 4.2E-005 HP24MVO, HP25MVO4 89.5 1.5 2.7E -006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM5 90.5 1.1 1.9E -006 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM
6 91.5 .9 1.7E-006 HP24MVO, HPSEGQM7 92.4 .9 1.7E - 006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM8 93.3 .8 1.5E-006 HP26MVCH, HPCROSSH9 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, HPSEGMM12 96.1 .5 9.1E-007 HP 148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM13 96.5 .4 6.9E-007 HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM14 96.9 .4 6.9E -007 HPCPPR, /HPSEGHM, /HPSEGKM, HPSEGMM15 97.3 .4 6.9E -007 HPAPPR, /HPSEGHM, /HPSEGKM, HPSEGMM
16 97.6 .3 6.OE-007 HP148VVT, HPBPPR17 97.9 .3 5.7E-007 HPIOICVO, HP25MVO18 98.2 .3 5.7E-007 HP102CVO, HP24MVO19 98.5 .2 4.2E -007 HPBPPR, HPCPPS20 98.6 .2 2.8E-007 HP148VVH, HPBPPR
21 98.8 .2 2.8E -007 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM22 98.9 .2 2.7E -007 HP26MVO, HP409MVO, HP410MVO23 99.1 .1 2.3E-007 HPBCPPH, HPSEGKM24 99.2 .1 2.2E -007 HPBCPPH, /HPSEGHM, /HPSEGKM, HPSEGMM25 99.3 .1 2.2E -007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM
26 99.4 .1 2.IE-007 HPAPPR, HPBPPR27 99.6 .1 2.1E-007 HPBPPR, HPCPPR28 99.6 .1 1.4E-007 HPI13CVO, /HPSEGHM, /HPSEGKM, HPSEGMM29 99.7 .1 1.4E -007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM30 99.7 .0 6.9E -008 HP27MVH, HPBPPR
31 99.8 .0 6.9E-008 HPBCPPH, HPBPPR32 99.8 .0 5.7E -008 /HPSEGHM, HPSEGOH, HPSEGPM, /HPSEGQM33 99.8 .0 4.4E-008 HP113CVO, HPBPPR34 99.9 .0 2.3E-008 HPlOlCVO, HPSEGQM35 99.9 .0 2.3E-008 HP102CVO, /HPSEGHM, HPSEGPM, /HPSEGQM
36 99.9 .0 1.9E -008 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM37 99.9 .0 1.9E-008 HPlOlCVO, HPSEGOH38 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: HPIFault Tree: HP201
Mincut Upper Bound 6.194E - 004
Cut % %¼ CutNo. Total Set Freq. Cut Sets
1 24.2 24.2 1.5E-004 HP148VVT, HPBPPR2 40.3 16.1 1.OE-004 HP24MVO, HP25MVO3 56.4 16.1 1.OE-004 HP26MVO, HPCROSSH4 64.5 8.1 5.OE-005 HP2425MVH5 72.0 7.5 4.6E - 005 HPBPPR, HPCPPR
6 79.4 7.5 4.6E -005 HPAPPR, HPBPPR7 84.8 5.3 3.3E-005 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM8 87.7 3.0 1.8E-005 HPBPPR, HPCPPS9 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, HPSEGMM12 93.3 .7 4.1E-006 HP148VVH, HPBPPR13 94.0 .7 4.1E -006 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM14 94.6 .7 4.OE -006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM15 95.3 .7 4.OE -006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM
16 95.8 .5 3.1E-006 HP113CVO, HPBPPR17 96.2 .4 2.7E -006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM18 96.6 .4 2.6E-006 HP24MVO, HPSEGQM19 97.1 .4 2.6E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM20 97.4 .4 2.2E - 006 HP24MVO, HPSEGOH
21 97.7 .2 1.5E-006 HP26MVCH, HPCROSSH22 97.9 .2 1.5E-006 HP148VVT, HP409MVO, HPAPPR23 98.1 .2 1.5E-006 HP148VVT, HPAPPR, HPCROSSH24 98.3 .2 1.2E-006 HP102CVO, HP24MVO25 98.5 .2 1.2E-006 HPIOICVO, HP25MVO
26 98.7 .2 1.OE-006 HPBCPPH, HPBPPR27 98.9 .2 1.OE-006 HP27MVH, HPBPPR28 99.0 .2 1.OE-006 HP26MVO, HP409MVO, HP410MVO29 99.2 .1 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM30 99.3 .1 6.7E-007 HP113CVO, /HPSEGHM, /HPSEGKM, HPSEGMM
31 99.4 .1 6.7E-007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM32 99.5 .1 4.6E-007 HP409MVO, HPAPPR, HPCPPR33 99.5 .1 4.6E-007 HPAPPR, HPCPPR, HPCROSSH34 99.6 .1 3.3E-007 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM35 99.6 .1 3.3E-007 HP148VVT, HP409MVO, /HPSEGHM, HPSEGKM, /HPSEGMM
36 99.7 .0 2.3E -007 HPBCPPH, HPSEGKM37 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: HPIFault Tree: HP201
Mincut Upper Bound 2.365E - 004
Cut %O %/o CutNo. Total Set Freq. Cut Sets
1 30.0 30.0 7.1E-005 HP26MVO, HPCROSSH2 51.3 21.3 5.0E-005 HP24MVO, HP25MVO3 72.5 21.1 5.0E-005 HP2425MVH4 78.2 5.8 1.4E-005 HP148VVT, HPBPPR5 82.5 4.3 1.OE-005 HP148VVT, /HPSEGHM, /HPSEGKM, HPSEGMM
6 84.2 1.7 4.OE-006 HPBPPR, HPCPPR7 85.9 1.7 4.0E-006 HPAPPR, HPBPPR8 87.2 1.3 3.E -006 HPBPPR, /HPSEGHM, HPSEGKM, /HPSEGMM9 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, HPCPPS12 92.0 1.1 2.7E -006 HP26MVM, HPCROSSH, /HPSEGHM, /HPSEGQM13 92.9 .9 2.IE-006 HPBPPS, /HPSEGHM, HPSEGKM, /HPSEGMM14 93.8 .9 2.IE-006 HPCPPS, /HPSEGHM, /HPSEGKM, HPSEGMM15 94.6 .8 1.8E-006 HP24MVO, HPSEGQM
16 95.4 .8 1.8E-006 HP25MVO, /HPSEGHM, HPSEGPM, /HPSEGQM17 96.0 .7 1.6E-006 HP24MVO, HPSEGOH18 96.7 .6 1.5E-006 HP26MVCH, HPCROSSH19 97.2 .5 1.2E-006 HP148VVH, HPBPPR20 97.7 .5 1.2E -006 HPBPPR, HPSEGHM, /HPSEGKM, /HPSEGMM
21 98.1 .4 9.1E-007 HP148VVH, /HPSEGHM, /HPSEGKM, HPSEGMM22 98.4 .3 6.5E-007 HPlOICVO, HP25MVO23 98.6 .3 6.5E-007 HP102CVO, HP24MVO24 98.8 .2 3.6E-007 HP26MVO, HP409MVO, HP410MVO25 98.9 .1 3.OE-007 HP27MVH, HPBPPR
26 99.0 .1 3.0E-007 HPBCPPH, HPBPPR27 99.2 .1 2.8E-007 HP113CVO, HPBPPR28 99.2 .1 2.3E-Q007 HPBCPPH, HPSEGKM29 99.3 .1 2.2E-007 HPBCPPH, /HPSEGHM, /HPSEGKM, HPSEGMM30 99.4 .1 2.2E - 007 HP27MVH, /HPSEGHM, /HPSEGKM, HPSEGMM
31 99.5 .1 2.1E-007 HP113CVO, /HPSEGHM, /HPSEGKM, HPSEGMM32 99.6 .1 2.1E-007 HP109CVO, /HPSEGHM, HPSEGKM, /HPSEGMM33 99.7 .1 1.4E-007 HP148VVT, HPAPPR, HPCROSSH34 99.7 .0 1.OE-007 HP148VVT, HPCROSSH, /HPSEGHM, HPSEGKM, /HPSEGMM35 99.8 .0 I.OE-007 HP148VVT, HP409MVO, HPAPPR
36 99.8 .0 7.2E-008 HP148VVT, HP4Q9MVO, /HPSEGHM, HPSEGKM, /HPSEGMM37 99.8 .0 5.7E -008 /HPSEGHM, HPSEGOH, HPSEGPM, /HPSEGQM38 99.8 .0 4.OE -008 HPAPPR, HPCPPR, HPCROSSH
1-32
Table 1-29. Fault tree HPRI cut sets quantification report(5 year)
Family: HPIFault Tree: HPRI
Mincut Upper Bound 8.401E - 005
Cut % qo CutNo. Total Set Freq. Cut Sets
1 26.7 26.7 2.2E - 007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM2 42.0 15.2 1.3E-007 HP26MVO, HPCROSSH, HPRCPPR3 53.9 11.9 1.OE-007 HP2425MVH, HPRAPPR4 63.6 9.8 8.2E-008 HP24MVO, HP25MVO, HPRAPPR5 72.5 8.9 7.5E -008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM
6 79.8 7.3 6.1E -008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM7 86.2 6.4 5.4E -008 HP26MVO, HPCPPS, HPCROSSH8 90.9 4.6 3.9E -008 HP148VVH, HP26MVO, HPCROSSH9 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: HPIFault Tree: HPRI
Mincut Upper Bound 2.716E -006
Cut 010 %o CutNo. Total Set Freq. Cut Sets
1 28.4 28.4 7.7E -007 HP26MVO, HPCROSSH, HPRCPPR2 48.6 20.3 5.5E -007 HP2425MVH, HPRAPPR3 68.4 19.8 5.4E -007 HP24MVO, HP25MVO, HPRAPPR4 77.5 9.0 2.4E -007 HP148VVT, HP26MVO, HPCROSSH5 85.7 8.3 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM
6 88.6 2.8 7.7E - 008 HP26MVO, HPCPPR, HPCROSSH7 91.4 2.8 7.7E - 008 HP26MVO, HPCPPS, HPCROSSH8 94.2 2.8 7.5E - 008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM9 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: HPIFault Tree: HPRI
Mincut Upper Bound 1.339E - 006
Cut qO %o CutNo. Total Set Freq. Cut Sets
1 22.3 22.3 3.0E-007 HP26MVO, HPCROSSH, HPRCPPR2 39.5 17.2 2.3E - 007 HP2425MVH, HPRAPPR3 56.3 16.8 2.2E-007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM4 70.8 14.5 1.9E-007 HP24MVO, HP25MVO, HPRAPPR5 77.1 6.3 8.5E-008 HP148VVT, HP26MVO, HPCROSSH
6 82.7 5.6 7.5E - 008 HP2425MVH, /HPSEGHM, HPSEGKM, /HPSEGMM7 87.4 4.7 6.3E - 008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM8 91.8 4.4 5.9E-008 HP26MVO, HPCPPS, HPCROSSH9 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: HPIFault Tree: HPRI
Mincut Upper Bound 2.072E - 005
Cut % % CutNo. Total Set Freq. Cut Sets
1 32.8 32.8 6.8E-006 HP26MVO, HPCROSSH, HPRCPPR2 65.6 32.8 6.8E-006 HP24MVO, HP25MVO, HPRAPPR3 82.0 16.4 3.4E-006 HP2425MVH, HPRAPPR4 92.7 10.6 2.2E-006 HP148VVT, HP26MVO, HPCROSSH5 95.9 3.3 6.8E-007 HP26MVO, HPCPPR, HPCROSSH
6 97.2 1.3 2.7E-007 HP26MVO, HPCPPS, HPCROSSH7 98.3 1.1 2.2E-007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM8 99.1 .7 1.5E -007 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM9 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: HPIFault Tree: HPRI
Mincut Upper Bound 4.614E - 006
Cut %O 0 CutNo. Total Set Freq. Cut Sets
1 30.8 30.8 1.4E-006 HP26MVO, HPCROSSH, HPRCPPR2 52.6 21.9 1.OE-006 HP24MVO, HP25MVO, HPRAPPR3 74.3 21.7 1 .0E - 006 HP2425MVH, HPRAPPR4 84.8 10.5 4.8E -007 HP148VVT, HP26MVO, HPCROSSH5 89.6 4.9 2.2E -007 HPBCPPH, /HPSEGHM, HPSEGKM, /HPSEGMM
6 92.7 3.1 1.4E-007 HP26MVO, HPCPPR, HPCROSSH7 94.9 2.2 1.OE-007 HP26MVO, HPCPPS, HPCROSSH8 96.5 1.6 7.5E -008 HP24MVO, HP25MVO, /HPSEGHM, HPSEGKM, /HPSEGMM9 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, NorthCarolina, 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 NUMBER12-89) (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 PUBLISHEDNuclear Plant Aging Research . . MONTH YEAR
ton High Pressure Injection Systems August 19894. FIN Oh GRANT NUMBER
A63895. AUTHOR(S) 6. TYPE OF REPORT
ResearchLeroy 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. providename and mailing addressj
EG&G Idaho, Inc.P.O. Box 1625Idaho 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 EngineeringOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, D 20555
10. SUPPLEMENTARY NOTES
11. ABSTRACT (200 words or less)
This report represents the results of a review of light water reactor High PressureInjection System (HPIS) operating experiences reported in the Nuclear Power Experi-ence Data Base, Licensee Event Reports (LER)s, Nuclear Plant Reliability DataSystem, and plant records. The purpose is to evaluate the potential significance ofaging as a contributor to degradation of the High Pressure Injection System. Tablesare presented that show the percentage of events for HPIS classified by cause, compo-nent, and subcomponents for PWRs. A representative Babcock and Wilcox plant wasselected for detailed study. The U.S. Nuclear Regulatory Commission's Nuclear PlantAging Research guidelines were followed in performing the detailed study that identi-fies materials susceptible to aging, stressors, environmental factors, and failuremodes for the HPIS.
In addition to the engineering evaluation, the components that contributed to sys-tem unavailability were determined and the aging contribution to HPIS unavailabilitywas evaluated. The unavailability assessment utilized an existing probabilistic riskassessment (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 SECURITYeCLASSIFICATIOAging 9(rhis Page)
Unclassified(This Report)
Unclassified15. NUMBER OF PAGES
16. PRICE
NRC FORM 335 (2-891 "U.S. GOVERNKNT PRINTING OrFICEt`1989- 241-590100230
UNITED STATESNUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555
SPECIAL FOURTH-CLASS RATEPOSTAGE e FEES PAID
USNRC
PERMIT No. G-67
OFFICIAL BUSINESSPENALTY FOR PRIVATE USE, $300
