ML20005D623
| ML20005D623 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/31/1987 |
| From: | Fresco A, Fullwood R, Higgins J BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20005D612 | List: |
| References | |
| CON-FIN-A-3453 NUDOCS 8911140137 | |
| Download: ML20005D623 (147) | |
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590RERAM NUCLEAR FORR STATION a PROBASILISTIC RISK ASSESSMENT-BASED-SYSTEN INSPECTION FIANS Prepared by: ) A. Fresco, J. Higgins, R. Fu11 wood, W. Gunther, R. Lewis, and J. Usher Plant Systems & Equipment Analysis Group
. Engineering Technology Division Department of Nuclear Energy Brookhaven Na'tional Laboratory p' Upton, New York.11973 May 1987 j L
Prepared for U.S. Nuclear-Regulatory Commission L. p Region I . . I h King of Prussia, PA. 19406 FIN A-3453 I i 1 e 1
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?;r CONTENTS Pate, iv LIST OF-TABLES........................................................
- LIST OF FIGURES........................................................ vii 1
- 1. INTRODUCTION..................................................
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- 2. SYSTEM PRIORITY LIST..........................................
- 3. DOMINANT ACCIDENT SEQUENCES................................... I ;
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~4. COMMON CAUSE FAILURES......................................... 4-4
- 5. SYSTEM INSPECTION TABLES......................................
- 6. ADDITIONAL REFERENCES......................................... 5-F I $
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1 TABLES l 1
-1 Table No.- Title g I 1 Table of Content s f or Inspe ction Plans . . . . . . . . . . . . . . . . . . . . . . 2 2 Systen Priority Ranking..................................... 3 ,
l Al . Reactor Protection System (RPS)............................. 6 2 Al-1 ~ Generalized-Inspection P1an................................. 6 A2 Automatic Depressurisation System ( ADS) . . . . . . . . . . . . . . . . . . . . . 10 :' A2-1 Importance Basis Failure Mode Identification. . . . . . . . . . . . . . . . 10 A2-2 I&E ~ Inspect ion Procedures f or Syst em Operation. . . . . . . . . . . . . . 13 A2-3 Mod i f ied S y s t em Wa 1kd own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 A2-3.1 Safety Relief Valve Suf fix Assignment & Associated Equipment 21 A3 Residual Heat Removal (RRR) and Low Pressure Coolant I nj e c t i on ( L PC I ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 A3-1 Importance Basis and Failure Mode Identification. . . . . . . . . . . . 24 A3-2 I&E Inspection Procedures for Systes'0peration.............. 27 A3-3 Mod i f ied Sys t em Wa1kd own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 A4 Se rvi ce Wa t e r Sys t ems ( SWS ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 A4-1 Importance Basis and Failure Mode Identification............ 31 A4-2 I & E Inspection Procedures f or System Operation. . . . . . . . . . . . 33 A4-3 Mod i f ied S y s t es Wa 1 kd own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 M5 Emergency Electric Power System (EEPS)...................... 37 A5-1 Importance Basis and Failure Identification................. 37-A5-2 I & E Inspection Procedures for Sys tem Operation. . . . . . . . . . . . 41 A5 Mod i f ied Sys t en Wa 1kd own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 A5 . Proposed Inspection Plans for Diesel Generator at l Nuclear P1 ants............................................... 44 Bl High Pressure Coolant Inj ection (HPCI) . . . . . . . . . . . . . . . . . . . . . . 47 l B1-1 Importance Basis and Failure Mode Identification. . . . . . . . . . . . 48
'B1-2 I&E Ins pection Procedures f or Sys tem Operation. . . . . . . . . . . . . . 52 B1-3 Mod i f i ed S y s t es Wa1 kd own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 B2 Reactor Core Isolat ion Cooling ( RCIC) . . . . . . . . . . . . . . . . . . . . . . . 55
- B2-1. Importance Basis and Failure Mode Identification. . . . . . . . . . . . 56 52 I&E Inspection Procedures f or System Operation. . . . . . . . . . . . . . 59 B2-3 Modified Systes Wa1kdown.................................... 60 1
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-v-TABLES Table No. Title Page B3 Core Spray (CS)............................................... 63 B3-l' Importance Basis & Failure Mode Identification................ 63 B3-3 I&E Inspection Procedures for System Operation................ 67 B3-4 Modified System Wa1kdown...................................... 68 B4 Reactor Building Closed Loop Cooling Water System (RBCLCWS)... 71 B4-1 Importance Basis and Failure Mode Identification.............. 71 B4-2 I&E Inspection Procedures for System Operation................ 74 B4-3 Modified System Wa1kdown...................................... 75 B5 Reactor Water Level........................................... 79 35-1 Importance Basis and Failure Mode Identification.............. 79 B5-2 I&E Inspection Procedures for System Operation................ 81 B5-3.1 Level Instrument Assignments for Safety System Operation...... 82 B5-3.2 Shoreham Vessel Level Trip Elevation Correlation.............. 83 B6 Feedwater, Condensate, & Power Conversion System (FW/CD/PCS).. 86 B6-1 Importance Basis and Failure Mode Identification.............. 86 B6-2 I&E Inspection Procedures for System Operation................ 90 B6-3. Modified System Wa1kdown...................................... 92 Cl St andby Liquid Cont rol ( SLC) System. . . . . . . . . . . . . . . . . . . . . . . . . . . 95 Cl-1 Importance Basis and Failure Mode Identification.............. 95 Cl-2 I&E Inspection Procedures for System Operation................ 97
,, Cl-3 Modified System Wa1kdown...................................... 98 C2 Main Steam Isolation Valves (MSIVs)........................... 101 C2-1 Importance Basis and Failure Mode Identification. . . . . . . . . . . . . . 101 C2-2 I&E Inspection Procedures for System Operation................ 102 C2-3 Modified System Walkdown...................................... 103 D1 Drywell Area Components, System Wa1kdown...................... 104 D2 Significant Human Errors, Importance Basis and Failure Mode Identification................................... 106 D3 Containment Systems, Importance Basis and Failure Mode Identification........................................... 107 D4 Reactor Building Standby Ventilation System (RBSVS) and Control Room Air Conditioning (CRAC) Chilled Water System.... . -112
. D4-1 Importance Basis and Failure Mode Identification. . . . . . . . . . . . . . 112 D4-2 I&E Inspection Procedures for System Operation................ 114 D4-3 Modified System Walkdown...................................... 115 El Plant Operations Inspection Guidance.......................... 118 E2 Periodic Surveillance & Calibration Inspection Guidance....... 128 E3 Maintenance Inspection Guidance............................... 138 i
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-vi-FIGURES Figure No. Title Page Al-1 Reactor Protection System ............................. 9 A2-1 General ADS Logic System No. 1020.201.................. 16 l A2-2 ADS Initiation Logic A System No. 1020.2............... 17 l
A2-3 ADS Initiation Logic B System No. 1020.201............. 18 l A2-4 RV Locations and Setpoints............................. 19 A2-5 Relief and Safety Valve Arrangements System No. 1020.201 20 A2-6 AD S Ai r Ac t u a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 . l A2-7 Schematic of the Nitrogen / Instrument Air Supply System Inside the Reactor Bulding and Containment...... 23 A3-1 Schematic of the RHR System Displaying the Following Modes of Operations (a) LPCI, (b) Containment Spray, and (c) RHR............................................ 30 A4-1 Reactor Building Service Water System. . . . . . . . . . . . . .. . . . 37 l AS-1 Schematic of the Electric Power Distribution System ! to the Emergency Power Buses at Shoreham from Off-Site l and On-Site Power Sources.............................. 46 i A5-2 Schematic of the 125 V DC Power Distribution........... 47 I B1-1 Shoreham High Pressure Coolant Injection System Schematic....................................... 55 1 B2-1 Reactor Core Isolation Cooling System Schematic. .. . . . . . 62 B3-1 Core Spray System Schematic (A Side)................... 70 B4-1 Simplified Schematic of the Reactor Building Closed Loop Cooling Water (RBCLCW) System.............. 78 B5-1 Leve l- Ins t rume n t Rang'e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 B6-1 Flow Diagram - Condensate System Injection to Reactor Vessel......................................... 94 B6-2 FW Pump Turbines'(Condensate and Feedwater System) , l Schematic........................ ..................... 94 Cl-1 Standby Liquid Control System in Standby Arrangement... 100 D3-1 Schematic of Containers and the Location of the Drywell Headers........................................ 109 l D3-2 Schematic of the Shoreham Containment and Reactor Building....................................... 110 , D3-3 Schematic of the Reactor Building Standby Ventilation S y s t em ( RB SV S ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111 - _D4-1 Schematic of the RBSYS and CRAC Chilled Water System... 116 l' l l-l h- , _ . .-_ _ . . . - . ,
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SHORERAM NUCLEAR POWER STATION , L PROBABILISTIC. RISK ASSESSKENT-BASED INSPECTION PLANS [.
- 1. INTRODUCTION ,
The tables and paraiFaphs in this inspection plan have been prepared to
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, provide inspection guidance based on review of the Shoreham PRA*. The guid-l ance should be used to aid in the selection of areas to: inspect and is not in-tended either to replace current I&E inspection guidance or to constitute an additional set .of inspection requirements. In using this information one ,
should realize that it is based primarily on the Shoreham PRA. Hence, recent
. system experience, f ailures, and modifications should be considered when us-ing these tables. Since plant modifications are .aormally an ongoing coatiaual process, it is recommended that relevant changes he catalogued so that these inspection plans can be periodically revised as required.
- 2. SYSTEM PRIORITY LIST Thc Shoreham core melt prevention (or front end) systems have been ranked or grouped in Table 1 according to their importance in preventing core melt.
They have been arranged into three groups of high importance ( A), medium in-portance (B), and lower importance (C). Containment or back and systems are listed in this table also (D), but are not ranked relative to the front end systems. Other plant systems not appearing in the list are generally of les-ser importance than those included here. Within each group, system importance is quite similar, hence, the systems are listed alphabetically. The systems were ranked or grouped by performing importance calculations on the systems based on their appearance in the dominant accident sequences of the event trees. Both the Birnbaum and Inspection Importance measures were calculated,
' ,, with the final grouping determined af ter a review of both lists.
- 3. DOMINANT ACCIDENT SEQUENCES The Shoreham PRA has a relatively large number of different accident sequences that contribute significantly to overall core melt risk, unlike other nuclear plant PRAs where risk is concentrated in a f ew sequences. The four types of sequences that dominate core seit risk at Shoreham are:
- Anticipated Transients Without Scram (ATWS)
- Station Blackout
- Transients with failure of FW, HPCI, RCIC, and ADS
- Flooding of reactor building elevation 8.
LOCAs outside the drywell are important to offsite risk, although not core melt.
- Science Applications, Inc. , " Final Report-Probabilistic Risk Assessment-Shoreham Nuclear Power Station-Prepared for Long Inland Lighting Company,"
SAI-372-83-PA-01, Palo Alto, CA, June 24, 1983.
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n Table 1 - Table of Contents fo'r Inspection Plans System Name Table No. Page No ']f. Reactor Protection' System (RPS)- Al , 6 , i Automatic Depressurization System ( ADS) A2 10 i Residual Heat. Removal' (RHR) and I , Low Pressure Coolant Injection (LPCI) A3 24 i -- Service Water Systems (SWS) A4 31
- Emergency Electric- Power System (EEPS) A5 - 38 High Pressure Coolant Injection (HPCI) B1 48 ;,
Reactor Core' Isolation Cooling (RCIC)' 52 56-Core- Spray (CS) . . -B3 63 Reactor Building Closed Loop Cooling Water B4 71. 1 (RBCLCW) . Reactor Water Level Measurement (RWL) B5 79-Power Conversion System (PCS)/Feedwater (FW)/ B6 86 Condensate System _ 4 Standby Liquid Control (SLC) Cl 95 Main Steam Isolation Valves (MSIVs) ' C2 101 Summary of Drywell Area Walkdown Items D1 104 Summary of'Important Human Errors D2 106 Containment Systems- . D3' 107
- Reactor Building Standby Ventilation D4 112 ,
System (RBSVS) & Control Room Air
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Conditioning (CRAC) Chilled Water System (CWS) 'E Plant Operations Inspection Guidance El 118 Periodic Surveillance and Calibration E2 128
. Inspection Guidance
. Maintenance: Inspection Guidance E3 138 9
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1 Table 2 - System Priority Ranking Front Rnd Systems Top Group: AC Power Automatic Depressurisation System (ADS) Residual Heat Removal (RHR)
" Reactor Protection System (RPS)
Reactor Building Service Water (RBSW) ,
' Middle Group: Core Spray (CS)
DC Power. High Pressure Coolant injection (HPCI) t
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Low Pressure Coolant Injection (LPCI) ' Power Conversion System (PCS)/Feedwater (FW).
- Reactor Bldg. Closed Loop Cooling Water (RBCLCW)
Reactor Core Isolation Cooling (RCIC)
- Reactor Water Level Measurement (RWL)
. Lower Group Main Steam Isolation Valves (MSIVs)
Standby Liquid Control (SLC) Turbine Building Service Water (TBSW) , AF Containment Systems Primary Containment . , Reactor Bldg. Standby Ventilation System (RBSYS) F Secondary Containment Suppression Pool l l t 4
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-4.- COMMON CAUSE FAILURES y The f ailure of multiple items from some common cause can be very signifi-cant to risk. The Shoreham PRA has identified several common cause failures that are particularly important:
- _ Loss of offsite power ;
- Failure of.all emergency diesel generators
- Common mode scras system f ailure ,
- Common mode miscalibration of reactor vessel water level instrumenta-tion
- - Internal flooding at elevation 8.
- Loss of SW causing f ailure of RBCLCW, RBSVS, CRAC, EDGs and, in turn, failure of rystems depending on these systems.
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- Clogged suppression pool strainers to ECCS.
- 5. SYSTEM INSPECTION TABLES Three tables have generally been prepared for each system to provide in-spection guidance. These tables are described below (see Table 1).
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' Table I Failure Modes Those components or licensee activities which play a dominant role in-contributing to system importance are presented, along with a brief descrip-tion of why these items are important. Inspection focus on these items will
. address approximately 95% of the risk significant areas. For experienced in-spectors, _this table is probably sufficient. This table generally uses the GE valve numbers as used in the PRA. A simplified system diagram extracted from the PRA is included for each system, which gives the valve numbers used.
Table I I&E Procedures For those who prefer additional guidance, this table $dentifies those I&E ; inspection procedures which can be used to assure the availability of the items shown in Table 2. The. inspection procedures were identified-based on the f ailure modes presented and an understanding of I&E procedures. The pto-cedures selected are'those which provide routine guidance on the principal _ , plant programmatic activities such as operations, maintenance, instrumenta-tion / control and ' eurveillance testing. There are many other inspection proce-dures which could also be used depending on the inspection criteria or the in-spector's preference. However, the procedures selected will generally provide ' adequate inspection coverage of the dominant f ailure modes. w+ .- - --. -- r - - - =-- -
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Table 1 Modified System Walkdown This table provides an abbreviated version of the licensee's system checklist, but includes only those items which are related to the dominant i failure modes. It is_ generally less than one third (sometimes only 1/10) of the normal checklist. Caution should be observed when using the checklists, since they are based on certain versions of the licensee's checklist. The re-
-vision date of the licensee's checklist that was used is indicated at the end '
of Table X-3. Where possible we have listed both the GE valve numbers used in the PRA and the LILCO valve numbers. Locations have not been provided for a significant portion of the compo-nents because they were not indicated in the LILCO operating procedures. All missing locations will be provided in the next revision. Special Tables-Some special tables are included in this plan. Table D1 is a summary of Drywell Area walkdown items. These are items which appeared in the modified system walkdown for_a given system and were located in the drywell. A summary was made of all these items and was included in one table for convenience. l Table D2 is a summary of the important human errors identified in the l PRA. f Table D3 discusses the containment systems and provides some PRA-based , insights for them. l Tables D4-1 to D4-3 pertain to the Chilled Water System (CWS) for the Reactor Building Standby Ventilation System (RBSVS) and the Control Room Air f Conditioning System (CRACS). Although the RBSVS itself is of some signifi-cance as a containment system, only the RBSYS and CRAC CWS was analyzed in the l Shoreham PRA by fault tree methodology as a separata= system. Therefore, the inspection plan for the CWS is provided as supplementary informat. ion for use at the inspector's discretion. Table A5-4 is, a separately developed Inspection Plan for the Diesel Gen-erator (DG) since a detailed fault tree was not included in the PRA for the , DGs.
- 6. ADDITIONAL REFERENCES A review of the Shoreham PRA was conducted by Brookhaven National Labora-tory during'1984. The review contains some very interesting insights into the PRA process. In particular, the inspector may find useful the information-
provided in Appendix SE on Reactor Water Level Instrument Line Pailure. The review reference is as follows: i D. Ilberg, K..Shiu, N. Hanan, E. Anavia, "A Review of the Shoreham Nuclear L Power Station Probabilistic Risk Assessment (Internal Events and Core Damsge Prequency)," NUREG/CR-4050, Brookhaven National Laboratory, June 1985. l l l
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.1 SHORERAM NUCLEAR POWER STATION EXPERIENCE-BASED INSPECTION PLAN j Reactor Protection System (RPS) f TABLE Al-1 GENERALIZED INSPECTION PLAN Discussion The Shoreham PPA uses historic industry failure data rather than plant l specific design data to establish an RPS failure probability. In Appendix B, j System Descriptions of the Shoreham PRA, the RPG system described corresponds '
I I to the version at the time of preparation of,the PRA. Since that time, impor-tant components including en analog transmitter / trip system for process sen-sors, and several circuit breakers in the electrical power supply have been installed. For completeness, this inspection plan therefore goes beyond the ' PRA to include these changes as well as inpots from the following documents: j
- Shoreham USA,R, Section 7 (January 1987)
- Shoreham, Safety Evaluation Report (SER), NUREG-0420
- Recent 1&E Information Noticos & Bulletins (January 1985 i.o June 1986)
- NUREG-0460, Anticipated Transients Without Scrams (NNS) -for Light Water Reactors
- GE Standard Technical Specifications
- I&E Manual and Inspection Procedures
- Selected Regulatory Guides & Standards
System Description
The Shoreham RPS consists of a combination of mechanical sensors, relay L logic,. circuit breakers, and motor-generator sets arranged in the " typical" ! BWR arrangement. RPS contains redundant trip subsystems, each of which '. rill j- gener. ate a reactor scram when certain parameters exceed established limits. Subsequent-to the original design upon which the FRA is based, several significant modifications have been made to the system to upgrade i;s config- !. .urations. ,0ne of the major changes was the installation of an anslog trans-I mitter/ trip system,which is described in the Shoreham FSAR, Rev. 32, 11/83. l This modification resulted in the replacement of the mechanical pressure sen-l' sors with electronic transmitters and solid state trip logic circuitry. The l_
-significant' consequence of this modification is the reduction in sensor cali-1- bration' frequency requirements from monthly to every eighteen enths. A
- second change not reflected in the PRA is associated with the solid state cir-
.cuit. breakers installed in each of the three electrical supplies to the RPS
-logic <- 1 normal, I alternate. These circuit breakers scose frequency and voltage and, when operated, interrupt power to the logic causing a trip.
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1 I A key-locked reactor mode switeb selects the necessary scram functions and bypasses for various plant operating modes. Certain scram functions such as low water level, high drywell pressure, and high reactor pressure are oper- l l able under all modes of reactor operation. During operation, logic relay contacts are closed, permitting- power f rom , the motor generator sets to be supplied through the solid state circuit l breakers and the logic relay chain to the scram solenoid operated valves , (SOVs) which supply air to the scram inlet-and outlet valves at each hydraulic i control unit., Loss of electrical power to the SOVs, either due to external reasons (loss of offsite power) or trip logic operation, results in SOV de-energization. Air pressure is removed from the scram air header through the SOVs, which opens the scram inlet and outlet valves. This allows water pres-sure to inset-the control rods. Outlet water is sent to the scran discharge l volume.- The.CRD mechanisms are latching piston-type drives. The 137 CRD mecha-
.nisms are mounted in individual housings at the bottom of the reactor vessel.
Each mechanism is coupled to its respective control rod, therefore, any piston movement within the drive mechanism results in the movement of its control rods. During a reactor scram, the high pressure water stored in the accumulator of each hydraulic control unit is directed to the bottom surface of the drive
. piston. At the same time, the volume above the piston is vented to the scram discharge volume which is at atmospheric pressure. This very large difference in pressure causes the drive pieton (and its attached control rod) to rapidly insert. This insertion time mur,t not exceed 7 seconds in accordance with the technical specifications.
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*sr As a backup to the scrom pilot air valves and the scram discharge volume solenoid valves, the entire CRD instrument air system is depressurized during a serem by the backup ceram valves, and independently by the Alternate Rod In-
- j. sertion ( ARI) scienoid valves if the ARI setpoints are exceeded. The ART function-is initiated by high reactor pressure or low reactor' water level (Level 2). In the event the reacte.r protection system failed to function, the actuated ARI function vould result in rod insertion by removing air pressure l from the scram air header through operation of solenoid operated valves.
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Inspection Areas
- 1. Review and witness RPS functica surveillance tests and preventive main-I tenance.
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- - Include witness of partial manual scram test, single rod scram, tests of individual RPS channels, and RPS circuit breaker and motor genera-tor set preventive maintenance.
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- References incl,ude: R.C. l.22, " Periodic Testing of Protection System Actuation Function," for RPS and RRCS; R.G. 1.118. " Period-ic Testing of Electric Power and Protection Systems," which en-dorses IEEE Std 338-1977, " Criteria'for Periodic Testing of Nuclear Power Generating Station Safety Systems," for RRCS only.
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- Detailed guidance for review of LPRM and APRM calibration is con-tained in IE Inspection Procedures 61703 and 61704. 3
- 2. Inspect senping instrument racks for correct valve configuration, "
labelling, and separation.
- 3. Ensure no abnormal RPS alarms in the control room, and verify and by-pass conditions are properly logged and justified.
- 4. Check RPS panels for-jumpers and lif ted leads. Documentation of same with the appropriate review and approval is required.
- 5. Review pos,,t work testing of RPS as'intenance tasks.
- 6. Review calibration records RPS sensors and compare results to Shore-ham technical specifications. Observe trends.
- 7. Review qualifications and training for technicians performing testing !
and/or maintenance on the system.
- 8. Review control rod drive mechanism maintenance inspection procedure and results. Insure trending of detected wear is performed.
- 9. Review surveillance and maintenance of ARI instruments.
- 10. Review preventive maintenance practices for solenoid operated valves located in the instrument air header and at-the HCU scram' inlet and outlet valves.
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SHOREHAM NUCLEAR POWER STATION L PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN l l Automatic Depressurisation System (ADS)' TABLE A2-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION , L CONDITIONS THAT CAN LEAD TO FAILURE General Guidance , i i Surveillance of the licensee's periodic. calibration, testing and/or pre- f j ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures, training and check-off lists in ac- ; cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the probability of failure for the conditions ) l listed below. The most relevant aspects are designated for each condition as I follows: l PC - Periodic. calibration activities, procedures and training. j' PT - Periodic testing activities, procedures and training. I MT - Preventive or unscheduled maintenance activities, procedures and ' L, training. OP - Normal and emergency operating procedures, check-off lists, training, etc. l- - Mission Success Criteria , i l The Automatic Depressurization System'(ADS) reduces the reactor pressure L so that flow from the LPCI and/or Core Spray Systems can enter the reactor h vessel in time to cool the core and limit fuel cladding temperatures in the l
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l event that the RCIC and RPCI systems cannot maintain the reactor water level. l There are eleven _(11) Safety Relief Valves (5RVs) installed on.the main steam l l lines.inside primary containment. Seven of these are automatically controlled t I by_the ADS. At least 3 of the 7 ADS valves are assumed by the.PRA to be re- j quired for depressurization. General Notes: 1. In 4 of the_11 leading failure modes (cutsets), operator l error by failing to initiate ADS after a component failure thereby preventing automatic actuation appears as a contribu-h.
- tor. This indicates the importance.of operator action in ini-l-
tiating ADS, and'the need to emphasize operator training. l l 2. In the ADS manual control system, the operator can use the I reset push buttons to delay or prevent the automatic opening of the relief valves if such a delay is prudent, such as in the
- case of Anticipated Transient Without Scram (ATWS). This also indicates the need to emphasize operator training.
L (__ i .,_ -- * ,- . . - -
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- 1. Failure of All ADS Valves or Valve Operators Due to an Adverse Environ-ment inside Containment Af f ecting Cable / Solenoid Coils Common mode failure of all ADS valves due to adverse environmental con-
-ditions inside containment prevents the operation of the ADS valves when re-quired (PT, OP).
- 2. Miscalibration of Reactor Water Level Channels or Transient Disables Level Instrumentation, and Operator Fails- to initiate ADS Miscalibration of-reactor water level channels, or a transient disabling the level instrumentation, combined with operator failure to initiate ADS after failure of the automatic initiation, or false indication of correct system operations prevents ADS operation when required (PC,PT,0P).
- 3. Nitrogen Supply Contaminated and Solenoids Disabled Failure of all ADS valve solenoids due to contaminated gas supplies pre-vents actuation of the ADS whe'n required (PT,0P).
- 4. Operator Error During Maintenance Disabled Valves or Solenoids Failure of all ADS valves due to common cause maintenance errors disabl-ing the valves or solenoids prevents actuation of the ADS when required (PT,MT,0F).
- 5. Common Mode Check Valve Leakage Bleeds ot f Accumulator Pressure and Failure of Backup Nitrogen Supply or Loss of Backup Supply Due to Con- ,
tainment Isolation
- #" Common cause accumulator failures combined with failure or loss of backup nitrogen-supply will prevent. ADS operation when required (PT,0P).
- 6. Common Mode Miscalibration or Transient Disables Pressure Switches on RHR !
Pumps A.B.C. and D Thereby Inhibiting ADS From Actuation Because RHR , Pumps Are Not Available Since the pressure switches on each RER pump A, B, C, and' D provides a permissive signal for ADS actuation to indicate that an RHR pump is operating, l a-Common mode miscalibration of transient disabling the pressure switches will prevent ADS initiation (PC,PT,0P).
- 7. Failure of Both Division I and'II 125V DC Electric Power Supplies Civen Common Mode Failure of All Offsite Power Sources and Failure of Diesel Generators 1-and 2 to Start and Run Failure of the Division I and Il station batteries or other de components, which provide 125V de electric power to be repaired within 2 hours of failure given loss of all ac power sources will cause failure of the ADS.
Refer to Emergency Electric Power System Table 1, Case 1, for specific inspection information. I I I l
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= )je i
. 8. ' Failure of Reactor Level Channels A and B and Operator Fails to Initiate j l
M- ADS Given Auto System Failure j Failure of both reactor level channels A and B combined with operator failure to initiate ADS given automatic system failure prevents the ADS from .) operating when required. Failure of a reactor level channel can be caused by: ] a) ' Sensor Failures (PC,PT).
.i a b) Logic Reset Switch in Reset Position Preventing Automatic Initiation !
l of an ADS Signal (PT.0P). l l
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1 l , . .i, 3 1 l l l l ! l l' [. - 1 , i- ; I l l l r ems
l i SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED ' INSPECTION PLAN Automatic Depressurisation Systems (ADS) j TABLE A2-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION i PROCEDURE FAILURE NUMBER TITLE COMPONENTS
- MODES ,
56700 Calibration Reactor low level 2,6,8
!. sensors, high dry -
well pressure sensors 61725 Surveillance and Control circuitry 2,6,8 Calibration Program ' 61726 Monthly Surveillance Observation 52051 Instrument Components and Sensors, switches 2,6,8 Systems-Procedure Review control circuitry
$2053 Instrument Components and Systems-Work Observation 52055 Instrument components and ,
I Systems-Record Review 62702 Maintenance (Refueling) ADS valves 4,7 Ar 62703 Monthly Maintenance Accumulator systems Observation Electric power systems l l 71707 Operational Safety .1,3,5,8 Verification 41700 Training- ADS operation- 1-8 41701 Requalification Training ADS operation 1-8
*Ref ers only to components identified in Tables A2-1 and A2-3.
i 1 1 4 - -
1-
' SHOREHAM NUCLEAR'54fER STAfical PROBABILISTIC CISK ASSESSNENT-BASED INSPECTION PLAN ,
AIfftMRTIC DEPIESSWttlAT90N SYSTBE t ADS) TABLE A2-3 ISODIFIED SYSTEM WALKDOIIN .~ Since most of the f olleres ossociated with ADS ere misceIIbretton errors, logic f ailures, and operator errors, verification of the items listed in this welkdown con orly Seperficially Indicate the status of the system. Observe-tion of the ADS calibration proegdores and/or f unctional test should provide the inspector. with some increased i assurance that these failure modes are averted. Desired Actuel Power Supply / Stegwired Actuel Description f.D. 900 Location Nsition Position tresker pe. Location Position Position I. Logic A Sultches S-13A 8ECR tint W ood WA " S-14A Ilot De- WA S-ISA pressed WA
- 2. Sultches for S-IA to teCR Auto WA IB21'RV-092 S*7A A.B.E,H,J,K,L
- 3. Logic B Switches S-13B HCR Isot Mmed WA S-148 ebt De-S-153 pressed ,
w WA d
- 4. Switches for $-19 to MCR Auto '
IB21'RV-092 S-7B A,B,E,H.J,K,L
- 5. SoItches for S-9,S-10 MCR CIosed WA IB21*RV-092 S-I I,S-1 D.C.F,G
- 6. SRV Dischergs IB21-TR-100 Temps IR35-P1pe Temp. IbraeI Ptol-MI/08 +
- 7. ADS Division i IHIl'Ptel- m IR4 2"PpfL-Power Supply 628 A2/10
- 8. ADS DIwIsIon iI IH11*P90L- On 1944 2'P91L-Power Supply 631 82/15
- 9. ADS Auto Inhibit S-13A 94CR 9eoreet tR2*P98L-Division I Switch BIER-10" Pret-628
/FIA&F2A
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. . Destred' Actual Power Supply / IRegelred Actuel ' ,;
LocetIon Locetion DeserIptIon 'l.D.' No. Positlon Pbst t ton Weeker9b. fbsition Positton
- 10. ADS Auto InhIblt -513B MCR Normei 1R42*PNL-82/
Division il Solteh OfRR-10. "
"PNL-631/ FIB &
F2B l l.M2 Supply P50'sev-105A SCR Iben Systen MOV P50*MLM-lO5B MCR Open - P50"809-103A ~MCR Open P50"MOV-103B MCR ' Open'
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12.M, Supply P50"02V-0696A DIl Open Minual Velves P50'02V-06968 ~ DW Open ' . P50802V-0703A 'DW Open i P50'02V-0702 DW Iben 13.Alr or M Sys em 2 80-110 Pressure psig _ 14.M2 Acc o ulators 80-110 Pressure psig fieference Documents ;,-, y-Title ' I .D. Ib. Stev. Date 1 Automatic Depressertretion SP 23.101.01 7 3/5/85 , i Systen (ADS) (LILCO Operating Procedure) i
- 2. Automatic Depressertretion .3 4/9/94 System ( ADS) - Systen 1020.201 II.lLCO System. ,
Description) for. slept i f led drawings ,
- 3. Omments by IIRC Senly 2/4/87 Resident inspector -
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SAFETY. RELIEF VALVE SUFFIX ASSIGNMENT & ASSOCIATED EQUIPMENT i 4 SAFETY REllEF VALVES - RV092 A B C* D* E- F* G H J K L ACCUNUI.ATORS A003 A '8 C D E F G H J K L CHECK VALVES F036 A -8 'C D E F G H J K L- 4'- IEMPERATURE TE101 .A B C D' Y E F G H J K L ELEMENTS VACUUM BREAKERS A B C- D E F G J H K L ADS FUNCTIONS YES YES - - YES - - YES YES YES YES
Reference:
Nuclear Botier P&lD.729E61680 (Rev. 9). Source: Shoreham PRA. Table B.2-1.
- Control provided in remote shutdown system.-
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1 i j SHOREHAM NUCLEAR P7JER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN :
- Residual East Removal System (RRIS) ,
Low Pressure Coolant injection (LPCI) System TABLE A3-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures, training and check-off lists in ' accordance with the Technical Specifications and relevant KRC bulletins and information notices should reduce the probability of failure for the condi-tions listed below. The most relevant aspects are designated for each condi-tion as follows: PC - Periodic calibration activities, procedures and training. 7 PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, training, etc. TS - Technical specifications. Mission Su_ccess Criteria The LPCI system provides vessel. inventory makeup following large pipe breaks. After reducing reactor pressure to the LPCI operating range, i.e. 409 psig, the LPCI system can also provide makeup for a small pipe break or other L demand for coolant inventory makeup. LPCI is an operating mode of the RHR system. It is divided into two loops A and B, consisting of two RHR pumps per loop, pump suction valves, LPCI injection valves, vessel level switches, I drywell pressure switches and reactor vessel pressure switches. l In the LPCI mode of RHR, the flow in loop A and loop B is from the suppression pool through either the RHR heat exchangers or the bypass around them and into the recirculation loop piping, the primary flow path for coolant injection. LPCI is automatically initiated upon: .
- Low reactor vessel water level and/or a High drywell pressure.
4
l o . LPCI (RHR) pump A is powered f rom Division I electrical power, pump B i from Division 11 and pumps C and D from Diviaion III. Only one LPCI pump is required for system success, per the FAA l l The RHR system also provides for the very important function of Decay or
- Residual Heat Removal. RHR cools the suppression pool and thereby removes i heat from containment. This is usually a long term function with sufficient timo for operator action and system realignment from the LPCI mode. Heat is transferred from the RHR system to the SW system in the RRR heat exchangers. ,
Cooling rate is determined by throttling valves P024A(B) and/or F048A(B). l l Note: The SNPS PRA provided one listing of cutsets for the LPCI and RHR systems. Failure modes 10 and 11 of the following table were derived from the listing of RHR cutsets given in the BNL review of the SNPS ; 1 PRA.I
- 1. Miscalibration of Low Resetor Pressure Permissive Channels Miscalibration of the low reactor pressure permissive channels prevents the LPCI system from being automatically initiated when required (PC, PT). !
I
- 2. Suppression Pool Water Unavailable Due to Clogged Strainers j l
Loss of the suppression pool water supply to the LPCI pumps due to j clogged strainers prevents the LPCI system from injecting into the reactor vessel (PT,MT,0P).
- 3. Both LPCI Loops Require Maintenance 1
Unavailability of both LPCI loops A and B due to maintenance prevents all F LPCI flow (MT,TS). f 4. LPCI/CS Manually Shut Off on High Level During an Accident and Level Instrumentation Fails to Automatically Initiate Af ter Having Worked at High Pressure Operator error in manually shutting off the LPCI/CS systems on Reactor Vessel High Level during an accident followed by failure of the level instrumentation to automatically initiate a second time after having worked at high pressures will prevent LPCI/CS flow when required (PT,PC,0P).
- 5. CS/LPCI Manually Shut Off on High Level During an Accident and Operators l
Are Totally Distracted Over a Period of 2-6 Hours
. Operator error in manually shutting off the LPCI/CS systems on Reactor Vessel High Level during an accident combined with operator distraction over a period of 2-6 hours results in operator failure to initiate LPCI/CS when required (OP),
1 D.11 berg, et al. , "A Review of the Shoreham Nuclear Power Station Proba-bilistic Risk Assessment (Internal Events and Core Damage Frequency)," NUREG/CR-4050, Brookhaven National Laboratory, pp. 61-62 June 1985.
i
- 6. LPCI/CS Manually Shut Off on High Level During an Accident and Operators !
Fail to Follow Procedures Af ter Recognizing the Problem ) { Operator error in manually shutting off the LPCI/CS systems on Reactor j Vessel High Level during an accident combined with operator distract, ion over a j period of 2-6 hours results in operator failure to initiate LPC1/CS when re- ) quired (OP).
- 7. Failure of All Four LPCI Pumps A.B.C.D to Start and Run '
LPCI flow is required from at least one of the four pumps for mission success (PT,MT,0P).
- 8. Failure of Flow frne Both LPCI Loops A and B to Enter the Reactor Vessel 1
Failure of flow from both LPCI loops A and B to enter the reactor vessel ' constitutes LPCI system failure. Failure of a single loop can be caused by the followingt a) Normally Closed MOV F015A or B at Primary Containment Drywell Penetration Fails Closed (PT). b) Normally Locked Open MOV F017A or B at Primary Containment Drywell Penetration Fails Closed (PT,0P). c) Air Operated Check Valve 050A Fails Stuck Closed (PT,0P). d) Normally Locked Open Manual Valve F060A or B Fails Closed (PT.0P).
- 9. LPCI Loop 1 in Maintenance and Failure of Pumps B and D or LPCI Loop 2 in Maintenance and Failure of Pumps A and C.
Given that either LPCI loop 1 or 2 is in maintenance as allowed by the Technical Specification., failure of the opposite loop pumps B and D or A and C to start and run when required constitutes LPCI system f ailure (PT,MT,TS).
- 10. Both RHR Heat Exchanger Bypass MOVs F04BA and F048B Fail Open Failure of both RHR Heat Exchanger Bypass MOVs F048A and F048B to close when required fails the suppression pool cooling mode of RHR (PT MT,0P).
- 11. Both Service Water Systes (SWS) MOVs 34A and 34B on the RRR Hest Exchangers' SW Outlet Lines Fail Closed FailuIre of the SW isolation MOVs 34A and 34B on the RHR heat exchangers .
l SW outlet lines in the closed position fails the suppression pool cooling mode of RHR (PT,MT,0P). ,
e- . -37~ SHOPEHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Residual Beat Bemoval System (RERS)/ Low Pressure Coolant lajection (LPCI) Systen TABLE A3-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE NUMBER TITLE COMPONENTS
- MODES 41700 Training All 4.5,6,10,11 52051 instrument Components and Reactor pressure 1,2,5,6 Systens-Procedure Review and level sensors 52053 Instrument Components and Systems Work Observation 52055 Instrument Components and Systems-Record Review 56700 Calibration 61725 Surveillance and Reactor pressure and 1,2,4-11 Calibration Program level sensors 3F 61726 Monthly Surveillance Suppression pool Observation strainers, pumps, MOVs Air-operated check 72700 Startup Testing-Refueling valves, manual valvas 71707 Operational Safety Verification 71710 ESF System Walkdown 62702 Maintenance (Refueling) Pumps, strainers 2,3,7,9-11 MOVs 62703 Monthly Maintenance Observation
- Refers only to components identified in Tables A3-1 and A3-3.
I
SMOREMdet MtfCLEMt POWEft STATIOut P910pm9tttSTIC RtSK ASSESS 8ENT-BASED INFECTaoss Pt#6 - Westemet fleet Stumeset (IWWFP/tm Prwesure Cootnet lejectlem (LPCtI TABLE A3-3 MODIFIED SYSTDt WALEDouN f.D. 80s. Oestred Actuel Peer Sepply/ hegelred Acteet l'escription TC~E.1/L ILEX) location Position Posltion Greester See. Locetion Pesition Position
- t. LPCI Sys 9 1980 (F015A)/ M11*PML SOf Otosed IECCi t tn/2Jn stb EL.l t F w th InJoctIon IEI 1*MD'f-036A 2.'LPCI Sys Il 18eD (FOISO)/ MIt*PML-601 Closed 15224*980C- 8tB EL.f tPE- Ce __
1njoctIon IE I 1**EDV-0369 112Y/IJu g ;
- 3. LPCI Sys i OUTIB (F017A)/
injection IEI1*MDV-037A M1t*PIEL-001 Ctcoed seCC11IN/IC1 819 EL.1I2'W On p.
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- 4. LPCI Sys Il tXffE) (F0t?9)/ Htt*PHL-601 Closed IR24*M W 819 EL.112'E On injection IEt t*98YJ-0379 It2Y/2Cl 5 e i
- 5. LPCI Sys I (F0504)/ HI t*P98L-601 Ciesed 1R24*se(X:- 119 EL.1I2'W Op '
i IsoletIon (Check) 1El1*A0V-081A 1113/83L
- 6. LPCI Sys iI (F05GB)/ s411*Put-601 Closed t#24**ECC- R8 EL.112'W (>i esototIen (Check)
- Ell'A09-0819 ii16/31L
- 7. ftfWt Sys t (F0604)/ Orywell Lectred '
injection to 8tK IEtt*Hv-7tA EL. 78' Open StenueI tsoletIon
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- 12. Sitet let 9 IP41*MDW-0349 Closed IR24'peCC- R8 EL. I12' Closad SW DIsch. t 8 25/64)
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.- . SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN ,
Service Water Systems (SWS) TABLE A4-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS TRAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic testing and/or preventive or un-scheduled maintenance activities and procedures and/or normal and emergency operating procedures, training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the conditions listed below. The most relevant aspects are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-of f lists, train-ing, etc. TS - Technical specifications. Mission Success Criteria JY The SWS is divided into two independent subsystems
. Reactor Building SWS
. Turbine Building SUS The turbine building SWS is not safety related and is not modeled in de-tail in the Shoreham PRA, hence below, SWS will refer only to the Reactor Building Service Water System (which also serves the Containment and Control Building).
- There are two SWS loops, each one consisting of two 50% capacity 8600 CPM pumps. During normal non-accident operation, the SWS is designed to provide
- cooling water to the secondary side of the Reactor Building Closed Loop Cool-ing Water (RBCLCW) heat exchangers, the Reactor Building Standby Ventilation System (RBSYS) and the Control Room Air Conditioning (CRAC) chilled water condensers, and to the main chilled water condensers. The SWS functions which may be required to be activated automatically or manually during accident conditions include the following: 9
i
. O f
RHR Heat Exchangers e ,
- Diesel Generator Engine Coolers ,
. RBCLCW Heat Exchangers e RBSYS Chilled Water Condensers
. CRAC Chilled Water Condensers .
. Emergency Makeup to Spent Fuel Pool
. Emergency Cooling Water to Ultimate Cooling Connection (direct ,
injection into the Rx Vessel via the RHR systes)
- 1. Both Service Water Loops Require Maintenance (Reactor Shutdown in 12 Hours)
Maintenance on all four SWS pumps requires shutdown of the reactor by l tech specs in 12. hours. Yet this small amount of allowed time still constitutes a significant risk. Maintenance outages of GWS pumps should be reviewed to determine compliance atid to establish any indications of trends , (MT, TS).
- 2. 11 Four SWS Pumps Fail at the Same Time Failure of all four SWS pups prevents all SWS flow. This can be caused by f ailure of logic control circuitry, bearings, lube oil, etc. (PC, PT, OP, '
TS).
- 3. Loss of Water Supply to screen Weil (Screen Well Water Level Low)
A lack of adequate water supply to the Screen Uell will prevent all SWS flow (MT, OP, TS). ,
- 4. Isolation of One Loop Required Due to Excessive Leakage and Failure to '
Isolate the Opposite Loop Excessive leakage in one SWS loop combined with failure to isolate the opposite loop from the leak prevents delivery of adequate SWS flow (OP).
- 5. One Loop in Maintenance with Two Pump Failures in the Opposite Loop One SWS loop in maintenance combined with f ailures of both pumps in the opposite loop will, prevent all SWS flow (PC, PT, NT, OP). ,
- 6. Failure of Operator to Cross-Connect TBSW to Supply Reactor Building SWS i i The TBSW System can provide a backup supply to the Reactor Building SWS.
If an accident signal exists it must be overridden by the operator to open the cross connect valves (OP).
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- 33 i
l SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN j l Service Unter Systes (SUS) TABLE A4-2 I&E INSPECTION PROCEDURES POR SYSTEM OPERATION . ; 1 i PROCEDURE FAILURE NUMBER TITLE COMPONENTS
- HODES 1
56700 Calibration SW Pumps, Valves 2,5 l 61725 Surveillance Testing & l Calibration Program 61726 Monthly Surveillance Observation 62702 Maintenance SW pumps, valves, 1,3,5, screens t 62703 Honthly Maintenance Observation 71707 Operational Safety Verification SW pum9s, valves 2,3,4,5 screens 71710 ESF System Walkdown JF' 41700 Training TBSW cross connect 6 to RB SW 41701 Requalification Training
- Refers only to components identified in Tables A4-1 and A4-3.
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- 5. DEDV-03tA. Set P-003A 1 Pet *04V- Scrooneell Closed l
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I General Walkdown Items
- 1. Visually Inspect SW pumps, motors, and supporting components for unusual operation, leakage, etc. l
- 2. Visually inspect the SW piping and components for leakage. ,
i
. Reference Documents Title I.D. No. Rev. Date .
i Service Water SP-23.122.01 16 3/20/86 l (LILCO Operating Procedure) ; e i 4 4 M'
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l SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED ' INSPECTION PLAN 1 tsorgency Electric Power System (EEPS) TABLE A5-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS TRAT CAN LEAD TO FAILURE General Guidance j Surveillance of the licensee's periodic testing and/or preventive or un-scheduled maintenance activities and procedures and/or normal and emergency i operating procedures, training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the conditions listed below. The I most relevant aspects are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. j PT - Periodic testing activities, procedures and training. j NT - Preventive or unscheduled asintenance activities, procedures and 1 training. l OP - Normal and energency operating procedures, check-off lists, train- 1 l ing, etc. TS - Technical specifications. ] Mission Success Criteria De Energency Electric Power System provides emergency power, when re-quired, to the plant's safety systems and components. There are two indepen-dent offsite power sources: The Normal Station Service (NSS) transformer, connected to the transmission systen side of the generator 138 kV circuit breakers, and the Reserve Station Service (RSS) transformer connected to the ' 69 kV transmission system and to an onsite 55 MW gas turbine generator. The onsite EEPS is divided into three independent divisions designated Divisions I, II, and III, each with its separate diesel generator. The dis-tribution system of each division consists of a 4kV emergency bus, a 480 V i load center, several Motor Control Centers (MCCs), and several low-voltage distribution panels. Instrumentation and control equipment and wiring are also segregated into three separate divisions, designated Divisions I, II, and III. . There are three independent 125V DC power systems corresponding to the three standby AC power divisions. Each DC division is energized by its own j battery and charger. The battery chargers are supplied from separate 480V MCCs, each of which is connected to'an independent 480 v AC bus. Each 125V DC battery bank has sufficient capacity without its charger to independently supply the required loads for two at least (2) hours. n- _ l
. . . - - - . - - . - -. - - ~ . _ . _ _ _ - . . ..- -__
i i 39 , Each of the three. independent 4 kV emergency buses has connections to two independent offsite power supplies and to a single onsite diesel generator. i in the event of loss of the preferred offsite source, a switch is automati-cally ande to the alternate offsite source. The diesels start automatically upon: total Loss of Offsite Power (LOOP), low reactor water level, or high ! drywell pressure. The diesels can also be started locally or remote manually . from the Control Room. l Note: Since the time when the Shorehaa PRA was originally prepared, several significant changes to the EEPS either have been made or will be made. Specifically, no credit was taken for the original gas turbine installation because it did not have " black start" capability. Sub- t sequently, a second gas turbine with such capability has been in-i stalled offsite. Secondly, prior to receiving the 5% low power operating license, four non-saf ety GM-END portable diesels were installed. l Finally, three safety grade Colt diesels have been installed which can act either as the primary or backup diesels. However, these diesels are only electrically connected to the onsite power system in , a temporary manner. Final configuration and operability will not take place until a later date. The inspector should be aware that these changes have not been incor-potated into the PRA, and therefore, have not been entered into the inspection plan. However, they should not be overlooked in the in- * . spection process. ,
, The diesel generators included in the PRA are the three safety-grade-Transamerica DeLaval Inc. (TDI) units.
The PRA determined the f ailure combinations (cutsets) required to cause the following failures CASE 1 - Loss of all three 125 V DC Divisions CASE 2 - Loss of any one of the three 480 V AC Divisione For CASE 1, the sole cutset causing loss of all three 125 V DC divisions ist
- 1. All Diesel Generators Require Maintenance (Shutdown in 12 Hours) and Common Mode Failure of All Of f site Power Sources (LOOP) and Direct Power Source (125 VDC 'Systes) is Not Repaired Within 2 Hours of Failure All three diesel generators requiring maintenance followed by a loss of all offsite power sources and failure to repair any of the batteries or other f' 125 V DC system components within 2 hours will prevent electrical power from being supplied to the three 125 V DC divisions (PT MT,0P).
-, , ., v- . ,m-4-.,. - ,-,-,--- ,.-
. o i Note: A, proposed method to inspect diesel generators is provided herein as ;
Table A5-4. This inspection plan is taken f rom NUREG/CR-4440 "A Review of Emergency Diesel Generator Performance at Nuclear Power ' Plants" by J.C. Riggins and P Subudhi, SNL, November 1985. i For CASE 2, the cutsets resulting in unavailability of any one of three 480 V AC divisions are:
- 1. Voltage Transf ormer Failure (PT.MT,0P). *
- 2. Normally Closed 4 kV to 480 V tus Transformer treaker Fails Open
( PT . MT ,0P ) .
- 3. 480 V Bus or Switchgear Unavailable (PT.MT,0P.TS). t
?
- 4. 4 kV tus or Switchgear Unavailable (PT.MT,0P.TS).
- 5. Common Mode Loss of Offsite Power and One Diesel Generator Fails to Start ,
and Run. Similar to CASE 1. Item 1. 9 b e
- e e
e
-L- _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
s 41 SHORERAM NUCLEAR POWER STATION PROSABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Energemey Electrical Power System (EEPS) TABLE A5-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE
- NUMBER TITLE COMPONENTS
- MODES 41700 Training Offsite power sources, CASE 1:1 diesel-generators, switchgear, transformers, battery sets, chargers, breakers 61725 Surveillance Testing and Offsite power sources, CASE 1:1 Calibration Program diesel-generators , CASE 23 switchgear, transf ormers, 1-$
battery sets, chargers, breakers 61726 Monthly Surveillance Observation 71707 ' Operational Safety Verification J 9 71710 ESP System Walkdown 62702 Maintenance (Refueling) Diesel-generat ors , CASE lit switchgsar, battery CASE 23 sets, chargers 3,4 62703 Monthly Maintenance Observation
*Ref ers only to components identified in Tables A5-1 and A5-3. ,
1- . l l I l e
SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED ! INSPECTION PLAN i tsorgeacy Electric Power System (EEPS) . i TABLE A5-3 MODIFIED SYSTEM WALKDOWN i Control Room. l Verify that power is available to or.from the following: From the two off-site power sources j 1 ~ 138 kV Normal Station Service (NSS) Transformer 69 kV Reserve Station Service (RSS) Transformer !
- 2. To the three 4 kV ac busses .
101 ~
~102 103 I
- 3. To the three 480 V ac busses .
111 , 112 113 4., To the three 125 V de busses : r A B ! C ,
- 5. Verify that all disel-generator (DG) alaras are cleared. :
Diesel Generator Rooms , L Verify the following
- 1. DG local / remote selector switch is in remote position.
- 2. Ventilation f an switches are in Auto or Standby.
p 3. Governor oil' level is satisf actory.
- 4. - Air receiver tank pressure is 205 psig.
L _5. Fuel oil' day tank level is satisfactory 275 gallons.
- 6. Electric innersion heater is operating correctly to maintain the engine
- jacket cooling water at 150*F when the engine is not running.
- 7. AC motor-driven water circulation pump for moving the water through the i jacket cooling water system when the engine is not running, is operating. ,
9 e
~ T*,
- i i
Verify the followings f
- 1. Power is available through battery chargers to:
Division I Division II Division III t 125 V de Bus A 125 V de Bus B 125 V de Bus C Battery Charger AC-Al Battery Charger BC-B1 Battery Charger BC-C1 , MCC-0A1 MCC-081 PNL-C1 MCC 0A2 MCC-052 PNL-C4 r PNL-Al PNL-B1 Inverter , PNL-A2 PNL-B2 i
'Inve rter
- 2. Check the station batteries, battery chargers, and inverters for abnormalities.
- 3. Check the station transformers 4 kV-480 V T-101-4. T-102-4, and T-103-5 for abnormalities.
9 6 9
I
~44-i SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-SASED INSPECTION PLAN tmargency Electrical Power System (EEPS) ,
i TABLE A5-4 PROPOSED INSPECTION PLAN FOR DIESEL GENERATORS AT NUCLEAR PLANTS A. Obj ective To review and evaluate Diesel Generator design operation, and mainte-nonce at NPPs to ensure that the DGs will be available when needed to power
- safety systems.
B. Details
- 1. The inspection of the following items should focus on DG auxiliary systems as follows: Fuel Injection System. Turbocharger, Starting System Speed / Load Control, Jacket Water, Cooling Water, Lube Oil, Puel 011. Control and Monitoring Systems, and Generator. 6
- 2. Using the LER, 50.55e, and Part 21 systems computer printout, select 3 recent failures (within 2 years) for followup at the NPP. When at the plant select an additional 2 failures from the internal systems.
Evaluate the licensee's response to these failures for proper failure analysis, corrective action, notification of vendor, Part 21 evaluation and documentation.
- 3. Maintenance: Refer to IE I.P.s 62700 and 62702, as they apply to DG -
maintenance. Additionally, does the NPP have, and have they imple-mented the DG vendors' saintenance recommendations (especially those , recommendations unique to nuclear service DGs such as Colt's de-scribed in NSAC-79)? Are maintenance personnel specially trained on DGs? Is f ailure information fed back into maintenance progras? has the NPP implemented recommendations of various studies referenced in Section 4 above.
- 4. Design Change Control Select two DG modifications and verify prop-er implementation. Utilizing information from DG vendor inspection on modifications recommended, verify that NPP is receiving all pertinent information in this area from the vendor. (Reference IE I.P. 37700).
- 5. Spare Parts. and Procurement: Review how spare parts and services are
- purchased and parts stored, both from DG vendor and direct f rom sub-vendor. Verify adequate Part 21 and QA, particularly when vendors are only supplying commercial grade parts and services (e.g., Wood-ward Governor and Stewart and Stevenson). Verify ASME code specified where appropriate. Tour spare parts storage area. (Reference IE I.P. 387015).
,x
. . 6. Training: Ensure appropriate DG specific training given to mainte- 3 nance, operations, QA, and management personnel. Are there adequate i documents to describe DG operation onsite (both asin engine and aux- :
iliary systes)? (Reference IE 1.P. 41700). t
- 7. Observe DGs in operation. Ensure they run smoothly and are operated per procedure. Look for abnormal vibration and leaks (air, fuel oil, or lube oil). Checs that readings are within specified limits. Are ,
limits per DG vendor recommendations? Are recommendations clearly l specified? Is air quality in DG room satisfactory without excessive i dust? Are control cabinets properly gasketed? Are instruments cali- . brated? Is trending of operating data performed to detuct degrada- ' tion early?
- 8. Is NPP receiving all appropriate service information from vendort design, maintenance, operational, ete? This is especially important for General Motors DG owners (verify they receive " Power Pointers" '
from GM).
- 9. Review site practices to limit DG cold f ast starts. ,
- 10. Reliability records and calculations: Check logs, procedures, and calculations versus Reg. Guide 1.108 criteria.
- 11. Enrure that pertinent studies on DG performance have been reviewed and recommendations implemented as appropriate (e.g., NUREG/CR-0660 and NSAC-79).
- 12. Torquingt onsure plant has adequate specifications for all torquing.
Ensure it is documented and done with calibrated equipment. Observe ; re-torquing if in progress. Source J.C. Higgins and M. Subudhi, "A Review of Energency Diesel Generator Perf ormance at Nuclear Power Plants," NUREG/CR-4440, Brookhaven National Laboratory, November 1985. References
- 1. NSAC-79 "A Limited Performance Review of Fairbanks Morse and General -
Motors Diesel Generators at Nuclear Plants," Nuclear Safety Analysis Center. Electric Power Research Institute, April 1984.
- 2. G. Boner and H. Hanners, " Enhancement of Onsite Energency Diesel Generator Reliability," NUREG/CR-0660, University of Dayton, February 1979.
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Sources Shoreham PRA, Figure B.5-2. Figure A5-2. Schematic of the 125 Y DC Power Distribution t h - - . _ _ _ _ _ _ _ - _ _ _ - _ _ _ . a . _ , . . , _ _ _ , . _ . , , , , _ _ _ . _ . , , , _ _ _ , . . , , , , , , , _ , ., ., _ , . .
( . 48-SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSKENT-BASED INSPECTION PLAN Eigh Pressere Coolant Injection (EPCI) System TABLE B1-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS TRAT CAN LEAD TO FAILURE General Guidance . . Surveillance of the licensee's periodic testing, calibration and/or pre-ventive or unscheduled saintenance activities and procedures and/or normal and ) emergency operating procedures, training and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of f ailure for the conditions listed be- I low. The most relevant aspects are designated for each condition as follows: l PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, training, etc. TS - Technical Specifications. Mission Success Criteria The HPCI system is designed to inject water into the reactor vessel through the feedvater line over a range of pressure from 150 to 1100 psi. Initially, domineralised water from the Condensate Storage Tank is used as the water supply with automatic switchover to the suppression pool as the backup supply. The HPCI turbine is driven by steam extracted from main steamline "B" upstress of the main steen11ne isolation valves. The inboard isolation valve in the steam line to the HPCI turbine is normally open and the outboard valve is normally closed with the exhaust line discharging to the suppression pool below the waterline. ' The HPCI system requires only DC power from the st& tion battery and steam extracted from the nuclear systes. The RPCI controls automatically start the system and bring it to design flow rate within 25 seconds after receipt of a reactor pressure vessel low water level (Level 2) signal or a primary contain-ment drywell high pressure signal. Failure Condition I
- 1. MPCI System in Maintenance Maintenance disables the HPCI system. The Technical Specifications should be checked for allowed outage time and compared to previous maintenance p_ _ ._- __ ~
. . -49 !
l outages to determine compliance and to estat,lish any indications of trends (MT,TS). l
- 2. Failure of HPCI Pusp or looster Pump or Turbine to Start and Run Failure of either the HPCI pump or its booster pump or the turbine to start and run prevents all MPCI flow. This can be caused by failure of logic circuitry, bearings, lube oil, etc. (PT,MT).
- 3. Normally Closed Steam Supply Isolation MOV F001 or F003 Fails Closed 1 1
I Failure of either MOV F001 or F003 to open prevents steam flow to the ' l HPCI turbine, thereby preventing HPCI flew (PT,MT).
- 4. Normally Closed HPCI Turbine Stop Valve F051 or control valve F052 Fails Closea ,
J Failure of either Stop Valve F051 or Control Valve F052 to open when re-
~
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quired prevents steam flow to the HPCI turbine, thereby preventing HPCI flow (PT.MT). , 1
- 5. Normally Closed HPCI Pump Discharge Isolation MOV F006 Fails Closed j
Failure of HPCI pump discharge isolation MOV F006 to open when required prevents HPCI flow (PT,MT).
- 6. HPCI Lube 011 Cooler Shutoff MOV F059 Fails Closed ,
Failure of the lube oil cooler shutof f MOV F059 to open will cause over- ! p heating and eventual failure of the HPCI pump thereby preventing HPCI flow (FT,MT).
- 7. Auxiliary 011 Pump Fails to Start and Run Start up of the turbine auxiliary oil' pump and proper functioning of the ;
hydraulic control systes are required for operation of the turbine stop and I control valves (PT). l
- 8. False signal Indicates Righ Area Temperature ,
l A false signal from either of the area temperature sensors causes the steam supply to the HPCI turbine to be isolated, thereby preventing HPCI flow (PC PT). ,
- 9. Operator Inadve'rtently Isolates Shutoff MOV F042 From the Suppression l Pool l
Inadvertent closure of MOV F042 prevents flow f rom the Suppression Pool from entering the HPCI pump suction, af ter the Condensate Storage Tank inven-tory has already been depleted (OP). l 1
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- 10. Miscalibration' of High Level Sensors in Suppression Pool or Low Level I Sensors in Condensate Storage Tank (CST) ]
l The HPCI suppression pool suction MOV F042 automatically opens, provided l that MPCI steam supply pressure is not low or auto-isolation is not present, .i upon. generation of either a low water level signal in the CST or a high level : signal in the suppression pool, If either set of sensors is miscalibrated '! either too low or too high, proper operation of the HPCI system when required '
'l may be jeopardised (PC.PT).
- 11. Common Mode Miscalibration of Turbic a Exhaust Diaphragm Pressure Sensors i or Steam Supply Line Pressure Sensors Miscalibration of either the turbine exhaust diaphragm pressure sensors or the steam supply line pressure sensors can cause inadvertent isolation of ;
the RPCI turbine, thereby preventing HPCI flow (PC,PT),
- 12. Miscalibration of Turbine Trip Pressure or Level Sensors A false signal caused by miscalibration of the turbine high exhaust pressure or reactor pressure vessel water level sensors will cause inadvertent tripping of the HPCI turbine (PC,PT).
- 13. HPCI Turbine Fails to Start (During Subsequent Re-Starts)
Failure of either the HPCI pump or its booster pump or the turbine to start and run prevents all HPCI flow. This can be caused by failure of logic circuitry, bearings, lube oil, etc. (PT,MT).
- 14. Normally Closed HPCI Discharge Isolation MOV F006 or Suppression Pool Isolation MOV F042 Fa11s Closed (During Subsequent Re-Starts)
~
i Failure of the HPCI pump discharge isolation MOV P006, or the pump suction isolation MOV F042 from the suppression pool, to open or remain.open when required prevents HPCI flow (PT). j 15. jbygg.jy Closed Turbine Stop Valve or Control Valve Fails Closed (During Suk;13 tent Re-Starts) l Failure of e'ither Stop Valve F051 or Control Valve F052 to open when L required prevents steam flow to the HPCI turbine, thereby preventing HPCI flow (PT,WT).
- 16. Failure of HPCI Pump or Booster Pump to Start and Run (During Subsequent '
l Re-Starts) l Failure of either the HPCI pump or its booster pump or the turbine to l-start'and run prevents all HPCI flow. This can be caused by failure of logic I- circuitry, bearings, lube oil, etc. (PT,MT). l l b l-l L
- h. . _ . _ _ _ _ .
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- 17. False Signal Indicates High Steam Line AP, sensors N004 or N005 A false signal indicating high steamline AP and therefore a possible leak or rupture.of the HPCI steam supply line causes isolation of the HPCI system uith operator action required to recover (PC,PT).
'18. False Signal Indicates High Turbine Exhaust Pressure Sensor N017A or N017B A false signal indicating a high turbine exhaust pressure from sensor N017A or N017B will trip the HPCI turbine (PC.PT).
- 19. False Signs 1 Indicates Low Pressure on Pump Suction (No. 10)
~
A f alse signal indicating pump suction low pressure will inadvertently trip the HPCI turbine (PC.PT).
- 20. Failure of Reisy Logic for Suppression Pool or Condensate Storage Tank Level Indication (During Subsequent Operation)
Failure of the relay logic for the suppression pool or CST will prevent automatic switchover from the CST to the pool, thereby interrupting HPCI flow when required (PC,PT).
- 21. Instability in Steam Exhaust Line Causes High Exhaust Pressure During Initial Start-up or Subsequent Start-up Unstable, flow conditions in the HPCI turbine exhaust line will cause the turbine to trip on high exhaust pressure (OP.PT).
p
- 22. HPCI' Auto Reset Not Roset Failure to reset the HPCI system to the automatic mode will prevent auto-matic actuation of the system when required (OP.TS).
SHOREHAM NUCLEAR POWER STATION PkOBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Eigh Pressure Coolant Injection (EPCI) System TABLE B1-2 I&F. IWSPECTION PROCEDURES FOR SYSTEM OPERATION FAILURE PROCEDURE NUMBER TITLE COMPONENTS MODES 61725 Surveillance Testing and HPCI pumps, turbine 2-22 Calibration Program MOVs F001, F003, turbine stop valve ; 61726 Monthly Surveillance F051, turbine control _j Observation valve F052, MOV F006, 71707 Operational Safety lube oil cooler MOV Verification F059, Aux. Oil Pump, 71710 ESF System Walkdown MOV F042, level, , pressure, temperature l sensors 62702 Maintenance HPCI Pumps, turbine, 1-6 MOVs F001, F003, 62703- Monthly Maintenance turbine stop valve Observation F051, turbine control i F052, MOV F006, lube j l' oil cooler MOV F059 53051 Instrument components Level, pressure 8,10-12, and. Systems-Procedure temperature sensors 17-20 Review
$3053- Instrument Components and
.Systees-Work Observation-53055 Instrument components and
- Systems-Record Review l-56700 Calibration l 41700 Training MOV F042, HPCI turbine 9,21,22 L
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TMILE Bi-3 IODIFIED SYSTEM IfALKDOWN (Cont'd) 1 . 0 . 80 0 Desired Actuel Power Sopply/ Required Actuel Description _ G.E. Litf0 Location Position Position Breaker too. Locetton Position Position 13.W'Cl Systen lHit' Control Re. IR4 2*PNL- Reley fte. Closed Torbine PNL-620 82/7 125 Y de j e Relay logic Olst. Penet (Blue) i 14.te'Cl Torbine IHil' Control Rs. IR4 2'PNL- Same Closed . & Turbine PfeL-601 B2/5 l Volvo (A0W-082, LCV-091
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r SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Core Isolation Cooling (RCIC) System TABLE B2-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS TRAT CAN LEAD TO FAILURE _ General Guidance Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities and procedures and/or normal and emergency operating procedures, training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the conditions listed below. The most relevant aspects,are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, training, etc. TS - Technical specifications. Mission Success Criteria The RCIC system consists of one 100% capacity steam turbine-driven pump which injects cooling water flow from the Condensate Storage Tank to the - reactor vessel over a range of pressure from 100 to 1100 psi through Feedwater Loop A in order to assure that sufficient water inventory is maintained in the reactor vessel to permit adequate core cooling during the following conditions: a)- Transients that include loss of normal feedwater. b) LOCAs with break sizes that do not depressurize the reactor. c) Hot Shutdown conditions. The suppression pool provides the backup water supply. !- 1. RCIC in Maintenance Given that HPCI is Available and Not in Maintenance Maintenance disables the RCIC system. The tech specs should be checked for allowed outage time and compared to previous maintenance outages to , determine compliance and establish any indications of trends (MT,TS). (
37
- 2. Failure of RCIC Pump or Turbine to Start or Run Failure of_the pump or turbine'to start or run prevents all RCIC flow.
This can be caused by failure of logic control circuitry, bearings, lube oil, etc. (PT).
- 3. RCIC Pump Discharge Isolation MOV F013 Normally Closed Fails Closed Failure of the RCIC pump discharge isolation valve to open when required prevents all RCIC flow (PT, NT).
- 4. Motor-Operated Gate Valve F008 or F045 in Steam Line to RCIC Turbine Normally Closed Fails Closed Failure of either normally closed DC-powered valve to open when required prevents steam flow to the RCIC turbine, thereby preventing pump operation (PT, MT).
- 5. Equipment Room ( Area) High Temperature Sensor 600A, 600B, 601 A, 601B, 602B, 603A, 603B, 604A, or 604B Gives False Signal A false signal from any one of the equipment area high temperature sensors will generate a system isolation signal to automatically trip the RCIC turbine', thereby preventing RCIC flow (PC, PT).
- 6. Miscalibration of RCIC Turbine Trip Level and Pressure Sensors Causes False Turbine Trip A false signal from the RCIC pump suction low pressure and/or turbine JF high exhaust pressure and/or reactor vessel high water level sensor caused by i common mode failure initiates a RCIC turbine trip, thereby preventing RCIC '
flow (PC, PT).
- 7. RCIC Pump or Turbine Fails to Start on Subsequent Starts Failure of the RCIC pump or turbine to start subsequent to successful operation prevents RCIC flow (PT, MT).
- 8. Sensors N006 for Low RCIC Pump Suction Pressure Gives False Signal for Turbine Trip A f alse signal f rom the low pump suction pressure sensor N006 will cause .
the RCIC tur,bine to trip, thereby preventing RCIC flow (PC, PT).
- 9. Sensor N017 or'N018 for RCIC Turbine Steamline High 6P Gives False Signal or Instrument Line Break Causes Turbine Trip A false signal from the RCIC turbine steam line high AP sensors N017 or N018, or an instrument line break, will cause the RCIC turbine to trip, thereby preventing RCIC flow (PC, PT).
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'10. RCIC Turbine High Exhaust = Pressure Sensors N009A or N009B Cive False
-Signal for Turbine Trip .
A; false signal indicating high RCIC turbine exhaust pressure will trip the RCIC turbine thereby preventing RCIC flow (PC, PT).
- 11. RCIC Turbine Lubrication System Fails The lubrication system for the RCIC system can fail due to valves failing
. to open,' pump f ailure, etc. , thereby preventing RCIC turbine operation and RCIC flow (PT, MT).
- 12. Instability in Steam Exhaust Line Causes High RCIC Turbine Exhaust Pressure Steam exhaust line instability can induce a high exhaust'line' pressure, thereby = tripping the RCIC turbine and preventing RCIC flow (OP, PT).
- 13. Automatic HPCI Shutoff Fails and Operator Fails to Shut Off HPCI Failure to shutoff MPCI when RCIO is, running may cause the'RCIC pump to trip upon high reactor water level (PT, OP).
l- . L L . L M _
..' .. 59-t:
SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Beactor Core Isolation Cooling (RCIC) System TABLE B2-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION FAILURE PROCEDURE TITLE COMPONENTS
- MODES-NUMBER .
41700 Training RCIC pump & turbine 5,6,8-12 9 temperature, pressure level, & speed sensors ; area temperature sensors, lube oil - cooler isolation valve. 53051 Instrument components & Temperature, pressure, 5-10 Systems Procedure Review level, speed sensors. 4 53053 Instrument components & Area temperature Systems Work Observation sensors i 53055 Instrument components & Systems - Record Review 56700 Calibration 61725. Surveillance & Testing & RCIC pump & turbine 1-4, 1 Calibration Program temperature, pressure 11-13 er- 61726 Monthly Surveillance level & speed sensors, Observation area temperature sensors 71707 Operational Safety Verification RCIC discharge valve 71710 ESP Systen Walkdown F013, lube oil cooling 62702 Maintenance water valve F046, lube 62703 Monthly Maintenance Observation oil cooling water PCV, turbine stop valve 044, turbine control valve, steaaline isolation , valves F008, F0045 i
- Refers only to components identified in Tables B2-1 and 82-3.
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. TAHLE 82-3 NODIFIED SYSTEN WALMDOWN (Omt'di 3.0 lap. . Desired Actuel Power Sepply/ . Regelree Acteet ! Description G.E. LILCO Locetion~ Position Posttion ' Breaker No. Locetion Posttien. Posttion 11.nCIC Process' 1Hil' See On R2*PNL-A2/ . On l ' lastroent Cebinet PNL-613 04 l l 12.e. ItCtC fieley fH118 fieler 5tn. M42*PNL-A2/ On taple - P98L-62 8 125V de 07 Dist.Panet El. 44'
- b. RCIC fleI oy . IHtl' Some M42*P91L-92/ On Logic P98L-618 09 13.c. Ste m Leak IH1 l' - Saoe m42*PNL-A2/ On Detector P90L-614 .
005 M42*PNL-82/ I b. Steen Leak Some Sene On l Detectoe 012 I M42*P98L-82/. On 14.RCIC.9ysten tHile Same ' 10 Volve Opatrol P90L-602 I ( A0W-0821083) l steference Doements Title f.D. 800 Rev. Date ! (Litc0 Opersting Procedurel [' Reactor Core isoletion Coollag Syste (RCIC) SP 23.119.01 14 3/24/96 i l Velves not Identiflod in Dorehen ItCDC flow diagre 1 RCic turbine stop volves - 044 i l
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1 l SHOREHAM NUCLEAR STATION PROBABILISTIC RISK ASSESSKENT-BASED INSPECTION PLAN Core Spray (CS) System . TABLE B3-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION l CONDITIONS THAT CAN LEAD TO FAILURE
.c General Guidance .
Surveillance of the licensee's periodic testing and/or preventive or j unscheduled maintenance activities and procedures and/or normal and emergency- j operating procedures, training, and check-off lists in accordance with the j Technic'al Specifications and relevant hdC bulletins and information notices - l
- should reduce the probability of failure for the conditions listed below. The )
most relevant aspects are designated for each condition as follows: l l PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. l HT - Preventive or unscheduled maintenance activities, procedures and ] training. , OP - Normal and emergency operating procedures, check-of f lists, l training, etc. l TS - Technical specifications. Mission Success Criteria l # The Core Spray (CS) System provides inventory makeup and cooling during large LOCA breaks or other conditions requiring low pressure makeup, including such cases as ADS depressurization following a small break. It consists of two independent spray loops, each with a 100% capacity centrifugal pump, and a , separate spray sparger in the reactor vessel above the core for each loop, I l which is supplied from the suppression pool, h The 'aystem starts automatically upon:
- 1. Low reactor vessel water level and/or
- 2. High pressure in the drywell, and if
- 3. Reactor vessel pressure is low enough.
- 1. Failure of Differential Pressure Sensor 005 Due to Miscalibration or Common Mode Failure of Channel Miscalibration or common mode failure of the channel for differential pressure sensor 005 across the CS pump discharge valves prevents automatic opening of the valves (PC.PT).
- 2. , - Common Mode Failure of Both Core Spray Pumps Failure of both CS pumps to operate causes insufficient flow from the CS system when required (PT). .
- 3. Suppression Pool Water Unavailable Due to Clogged Strainer Clogging of the single strainer in the common CS pump suction line from
- the Suppression Pool will prevent all CS flow (PT,MT,0P).
- 4. Both Core Spray Pumps Require Maintenance - Reactor Shutdown Within 12 Hours .
Maintenance on both CS pumps prevents all CS flow when required. The Technical Specifications should be checked for allowed outage time and
. compared to previous maintenance outages to determine compliance and to establish any indications of trends (MT,TS).
- 5. Failure of Both Core Spray Loop A and Loop B .
Failure of both CS loops to start and run when required prevents all CS flow. Failure of a loop can include the followings a) Normally Closed Discharge Isolation Valve F005A or F005B Fails Closed l (PT) b) Pump A or Pump B Fails to Start and Run (PT) c) Failure of Differential Pressure Sensor to open Pump Discharge Valve ( PC PT) d) Failure of Differential Pressure Relay Logic (PC,PT) e) Loss of Pump Lubrication (PT,MT,0P) f) Failure of Strainer Blocks Flow (PT,0P) g) Flow Diverted to Suppression Pool Via Normally Closed Test Line MOV-i F015A or F015B Failing Open (PT,0P) h) Normally Open Suction Line MOV F001 A or F001B Fails Closed (PT,0P) i) Discharge Line Check Valve F006A or F006B Fails Stuck Closed (Inside Drywell)(PT) 1 j) Discharge Line Check Valve F003A or F003B Fails Stuck Closed (PT) k) Locked Open Discharge Line Manual Isolation valve F007A or F007B Fails Closed (Inside Drywell)(PT)
I 1 65- I l 6.. Conditional' Probability That Room Cooling is Required and Failure of
~
Reactor Building Standby ventilation System Civen that area cooling is required for the Core Spray pumps, failure of J the~ Reactor Building Standby Ventilation System (RBSYS) will cause failure of I the Core Spray System. Refer to Tables 03 and D4 for RBSYS failures. l 1
- 7. Reactor Vessel Level Instrumentatioa Fails Af ter Raving Worked at High l Pressures l l
And . a) LPCI/CS Manually Shut Of f on High Level During an Accident or b) Operators Totally Distracted Over a Period of 1-6 Hours and Fail to j Restart LPCI/CS When Required i or c) Operators Fail to Follow Procedures Af ter Recognizing the Problem and Fail to Restart LPCI/CS When Required Failure of the reactor vessel level instrumentation and various operator errors in failing to activate the LPCI or CS systems when required can cause plant damage.. Attention should be paid particularly to operator training (PC PT,0f).
;jr 8. Common Mode Failure of Both Loops Caused by Manual Valves F007A and F007B Being Lef t in the Closed Position Both locked open manual valves F007A and F007B inadvertently being lef t in the closed. position will cause loss of both CS loops (PT,MT,0P).
- 9. Common Mode Failure of All Offsite Power Sources and DC1 and DC2 Both Fail to Start and Run l
L Loss of all offsite power sources combined with failure of the Division I L and Division II diesel generators to start and run will prevent actuation of l the CS system. Refer to the Emergency Electric Power System inspection plan, i Table A5, for further details.
- 10. Failure 'of Loop A (Loop B) and Loop B (Loop A) Battery Charger
Unavailable, or Power Cable f rom 480V MCC to Loop B (Loop A) Charger Fails, and Repair Not Achieved Within 2 Hours i This combination of f ailures will prevent both loops of the CS system f rom being actuated when required. Reference should be made to Item 5 of this table and also to the Emergency Electric Power System (EEPS) inspection plan, Table AS, for further details. y, - - - - , - ,
- 11. Failure of Loop A (Loop B) and Loop B (Loop A) Voltage Transformer or Bus --
Transf ormer Breaker, or Unavailability of Loop B (Loop A) Bus or Switchgear- l This combination of failures will-prevent both loops of the CS system from , being actuated when required. Reference should-be made to Item 5 of this ( table and also to the EEPS inspection plan, Table AS, for further details. ' l
- 12. Suppression Pool Water Unavailable Due to Rupture l
Loss of the Suppression Pool water inventory prevents the reactor vessel from being cooled'b/ the CS system. To minimize this occurrence, the results of the. containment integrated leak rate teste can be reviewed to determine if ^ any leaks in the Suppression Pool had been identified. Also, plant records could be reviewed to detect any abnormal nitrogen usage which might indicate leakage of the pool in the inerted containment and/or a visual inspection of the pool could be conducted. 4 9 h ~ ~ ~ ~
._ E________________._______
, . SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN ,
Core Spray (CS) Systes ;
. TABLE B3-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION :
l', PROCEDURE - FAILURE NUMBER. TITLE COMPONENTS
- MODES 52051 Instrument Components and Resctor Low Level 1,5,7 Systems-Procedure Review Sensors, Drywell Righ Pressure 52053 Instrument Components and Sensors, Reactor Systems - Work Observation Vessel Pressure Sensors
$2055 Instrument components and '
Systems ' Record Review 56700 Calibration l 61725- Surveillance and Calibration Sensors, Pumps, 1,2,3,5 Control Program Strainers, Check Valves 7,8,12 Manual Valves, Discharge MOVs Suppression Pool l
~ .nr 61726 Monthly Surveillance i Observation 72700 Startup Testing-Refueling l 62702 Maintenance (Refueling) Pumps, Discharge 3,4,5,8 MOVs, Strainers 62703 Monthly Maintenance -
j Observation l 1 1 71707 Operational Safety Verification Pumps. Strainers, MOVs 3.5,8- j
~* '
Suppression Pool, 12 ! Electric Power System j I 71710 ESF System Walkdown {,
~
*Ref ers only to components identified in Tables B3-1 and B3-3.
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i SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN i Reactor Building closed Loop Cooling Water System (RBCLCWS) TABLE B4-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION l CONDITIONS THAT CAN LEAD TO FAILURE General Guidance l Surveillance of the licensee's periodic testing and/or preventive or unscheduled maintenance activities and procedures and/or normal and emergency operating procedures, training, and check-off lists in accordance with the Technical Specifications and relevant NRC bulletine and information notices I should reduce the probability of failure for the conditions listed below. The most relevant aspects are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and
~
trainings
.OP - Normal anc emergency operating procedures, check-off lists, training, etc.
- TS - Technical specifications.
L ! Mission Success Criteria f-'# The RBCLCW consists of three 50% motor driven pumps designed to provide 1600 l GPM each of cooling water, two 100% capacity RBCLCW heat exchangers and I associated piping, valves, and instrumentation. Cooling water must always be available to: I a) At least one of the two reactor recirculation pump seal coolers, motor bearing coolers, and motor winding coolers. b) At least two of the four RHR pump seal coolers. c) At least one of the two spent fuel pool cooling water heat exchangers.
-Following an accident or a head tank low-level signal, motor-operated isolation valves installed in the discharge header isolate each redundant loop
- from the other and also from non-safety related portions of the system.
During normal operation, two pumps are in operation and the third pump in
, automatic standby. Each pump can supply either redundant loop.
i 4
+
- 1. Failure of the Service Water System to Provide Flow to the RBCLCW System
. Loops Loss.of SWS flow to both'ioops of the RBCLCW system disables-the RBCLCW system.- The SWS inspection plan-(A4) should be referenced.
2 . Both RBCLCW Loops Require Maintenance - Reactor Shutdown Within 12 Hours
. Maintenance on both loops of RBCLCW requires shutdown of the reactor. The '
tech specs should be checked for allowed outage time and compared to previous maintenance outages. to determine compliance and to establish any indications of trends (HT,TS).
- 3. Failure of One Loop and Kaintenance on the Other Loop Failure of one loop together with maintenance on the other loop prevents all RBCLCW flow. Failure of one loop can be caused by the followings a) False Indication of Suction Pressure Stops Pump (PC PT) b) Pump Fails-to Continue Running (PT,0P) c) Pump Filter Clogged (PT,0P) d) RX Inlet (MOV-0421,B) or Outlet valve (MOV-052A,B) Fails Closed
( PT MT,0P) e) Temperature Control Valve (TCV-001W/X, 001Y/Z) Fails Closed (PT,MT,OP) f) Normally Open- SWS Motor Operated Valve Fails Closed (PT,MT,0P) g) Normally Open- Manual Valve (16V-009 A,B) in SWS Injection Line Fails , Closed (PT MT,0P) h) RX Ruptures' or Excessive Leakage (PT,0P) i) Head Tank Fails to Fulfill its Function to ' Provide NPSH (OP) j) Pump Discharge Check Valve (005A,B,C) Fails Closed (PT,0P) k) Pipe Rupture or Excessive Leakage in Pump Suction Lines, Pump Connections or Discharge Lines (PT,0P)
-1) -Faulty Motor 0verload Breaker Stops Pump (PT,0P)
- 4. Failure of Both RBCLCW Loops Failure of both loops can be caused by a combination of the events mentioned in 3 above.
-- . - - -. -.~ . . - . . . . - . . . - ,
I
. . 73 i
$.' Maintenance of One SWS toop and' Failure of the Opposite RBCLCW Loop
- I Maintenance of the SWS loops is described in:the SWS inspection plan,-
. Table A4, while. failure of.an RBCLCW loop is described in 3 above.
6.- Maintenance of One SWS Loop and Maintenance of the Opposite RBCLCW Loop 1 Maintenance of the SWS loops is described in the SWS inspection plan.
*- -Table A4, while maintenance on an RBCLCW loop is described in 2 above.
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l SHOREHAM NUCLEAR POWER STATION
-PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Building Closed Loop Cooling Water (RBCLCW) System TABLE B4-2 16E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE NUMBER TITLE COMPONENTS
- MODES L61725 Surveillance Testing and Pumps, heat exchangers 1,3,4,5
. Calibration Program & Inlet'& Outlet Valves 61726. Monthly Surveillance SWS NOVs, Check Valves, Observation Head Tanks, Filters
^
71707 Operational Safety Verification ' 71710 ESF System Walkdown 62702 Maintenance Pumps, Heat Exchangers -2,3,5,6 62703 Monthly Maintenance MOVs, SWS MOVs, Filters Observation
- Refers only to components identifie3 in Tables B4-l'and 34-3.
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l l l l - . TABLE B4-3 IWDIFIED SYSTE! % ANet (Cont'd) Desired Actuel Pe w Seeply/ Itagelred Actimt Description 1 . 0 . 90s . Location Position Position Gr *skv Pb. Locetton Nsition msttion 13.flBCLCW W SIf . IP4 t*fEDV-037B etnin O tl. Closed . 19t24'9800-1928/ fer. Bldg. Closed Outlet'IsoletIon Ihm.(IItB-01) 6EH El. I12' Pop A Ofscharge 005A %en m ock velvo l PWp B Discherge 00 2 Open e D eck Velve ! Pop C Discharge OOSC Open D ock Vstve 14.feCLCW Sk E-11ASW IP41*t6V-00094 fte. 9tdg. locked .- Omntelnment inlet El. 8' S Open y e 15.ftBCtCW W E-tIBSW IPe t* 16V-0009B Some Locked (bateinment latet @en i A e 2 e .' e e _,
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TABLE B4-3 MODIFIED SYSTEM WALKDOWN (Cont'd) Desired Actual J Position /L Position /- j Description I.D. No. Location Condition Condition 'l l Tank Level Check (Rupture and Leakage) J
..- RBCLCW Head Tank A Level Condition
,- RBCLCW Head Tank j B Level Condition Pump Filter' Check for Possible Clogging Pump P005A Suction Line Filter Pump P005B Suction -
Line Filter ; Pump P005C Suetion.
"' Line Filter.
Rupture and Leakage Checka Heat Exchanger E-001A 3 Heat Exchanger E-001B ,
'# - Pump Suction Line P005A <
Pump-Discharge Line P005A- ,i Pump. Connections P005A' Pump Suction Line P005B -1 Pump Discharge Line P005B : Pump Connections P005B Pump Suction Line P005C Pump Discharge Line P005C , Pump.Conneetions P005C .j I Reference Documents ~ Title I.D. No. Rev. Date (LILCO Operating Procedures): 1.-Reactor Building Closed Loop
. SP 23.118.01 12 3/24/86 Cooling Water (RBCLCW) System 2.- Service Water SP 23.122.01 18 3/20/86 G
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PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Water Level ) 1 TABLE B5-1 IMPORTANCE BASIS AND PAILURE MODE IDEKTIFICATION l CONDIT10hS TRAT CAN LEAD TO PAILURE J General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre- i ventive or unscheduled maintenance activities, procedures and training and/or ; normal and emergency operating procedures, training and check-off lists in ac- l cordance with the Technical Specifications and relevant NRC bulletina and in-formation notices should reduce the probab131ty of f ailure for the conditions listed below. The most relevant aspects are designated for each condition as followst . PC - Periodic calibrStion activities, procedures and training. PT - Periodic testing activities, procedurec and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency orarating procedures, check-off liste, > training, etc. Mission Success Criteria
## The Reactor Water Level (RWL) syarna at Ehorehan (and most all BWRs) is
.important in that, besides providing indication, it provides automatic trips and system initiations to accomplish many safety functions. There are mul-tiple RWL instruments used for this purpose as shown in Table B5-1-1 which was extracted from the Shorehan PRA. The various trip setpoints are shown in Table B5-1-2, also taken from the Shoreham PRA. The PRA includes effects of failures of the RWL system in two main ways (1) failure of the RWL systen to I initiate safety systems or trips, and (2) a break in a RWL instrument line outside the drive 11 leading to a LOCA outside containment.
Failure Conditions
- 1. Common Mode Miscalibration,
. Miscalibration.of multiple RWL channels / instruments can result in failure of many f ront line systems to trip or automatically actuate. (PC,VT).
4
- 2. Instrument Line Break Ta11ure of a RWL instrument reference line outside the drywell, leacing to a small break LOCA that bypasses containment and causes that channel to fail high (PC.PT,0P,MT).
7 ,
- 3. Operator Error ,
Given the LOCA in #2 above the operator, in the process of attempting to isolate the leak, causes a leak in the oppcsite channel of RWL instrumentation (OF).
- 4. Electronics Failure ,
Tailure of electronic circuitry of RWL instrumentation one channel . l ( PC. PT) .
$. Flashing of RWL Reference Leg During a station blackout drywell cooling is lost and the drywell temper-ature increases. As a result the RWL instrument reference les can flash caus-ing errohsously high readings. Operators must be properly trained to recog-nize this eventuality (0P).
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SHORENAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED r INSPECTION PLAN taaetor unter Level - TABLE B5-2 16E INSPECTION PROCEDURES FOR SYSTEM OPERATION PAILURE l PROCEDURE NUMBER TITLE COMPONENTS MODES 61725 Sorveillance Testing & RwL Instruments 1,4 Calibration Program 56700 calibration RwL Instruments 1,4 41700 Training Operator Error 3,5 3 71707 Operational Safety Insttumeut 2 Verification Line Broak 71710 ESF System Walkdown t JW 4 e 4
1 33 o .
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Table 35-3.1 LEVEL INSTRUMENT ASSIGNMENTS I FOR SAFETY SYSTEM CPERATION 1 SIDE A SIDE B FUNCTION INSTRUMENT POWER INSTRUMENT POWER _ t Seren & m -N000A(0 ) PS A LT E1200C(0) 95 A RHR i L: -483(O) PS I LT E1.NOICD(0) 95 8
- 150 l MPCI frio LT s21-N091C(LB) DC.A LT 821-et*0(LI) DC 4 1
WC1 LT 821 NC91A(L2) DC.A (f 8214915(L2) (Xl.4 l Initiste LT 821491C(L2) DC.A LT E1.NC910(L2) DC.8 RCIC frip LT E1.N0910(L3) DC.g LT E 1 491A(L8) DC.A RCC LT E1.NC91 A(c) DC.A LT 821 4 91B(L2) DC-8 Initiate LT Mi-4*1C(L2) OC.A LT B21.@91C L L21 DC.4 ! M51V LT-821 481 A(L2) RPS A Lf.821-N001C(L2 ) APS A l i LT-821 4815(L2 ) F5' 8 LT-4C1-N081D(L21 95 6 l l- . Afwl '. T 821-@91A(L2) 00 4 L1 321.N091B N*) DC4 t WT LT E1-N0'iC(L2) DC.A LT 321. 409t0(u) DC-8 [ i J Afwl LT 521-NC91 A(L2 ) DC.A (f $21.N0918(t2) CC.8 AAI LT 821-NC91C(L2) DC.A LT 821-NC 10(L2) DC-8 l I l LPCI LT 821-N091 A(L1) DC.A LT E1-40914(L1) CC-4
- LPCs LT E1-N091C(L1) DC.A LT Ei NC910(L1) DC.8 l LT Mi-495A(O OC.A LT E14095B(0) DC-8 i ADI LT 821-N091A(L1 DC.A LT E1 4918(L1) DC-8 LT E1-4910(L1)
LT E1.N091C(L1) DC-A DC 8 t l . f Food ed Lt C2-40AA(LI) Vital AC Lt C2-@040(L8) INST S Main TT LT O2-N004C(L8) IN57 A . ! Narrow ' LT C 2-404A(12) Vital AC LT 02-N0048(12) INST S - Range LT C2 204C(IND) INST A Disoley' LT 02-N00AA(RCC). Vital AC LT C2-40044(RCC). INST 8 , WR 'LT 821-@81A(EC) PS-A LT 821.N001C(RCC) RPSB . l Disoley' LT E1-N0810( LW) RP58 i I f Shutsom LT E1-427(12) INST A Upeet LT C 2-N017( RCC) INST A Fuel Zone LT E143f A( ACC) 1N37 A kt C -4278(DC) INST S S6urce: Shoreham PRA,
- EC e necorsert 1.2 : Indiretor. * * * ~ *
. Recorner suitenes bet,een sensors.
l m -- -
- -- -x -_z
.2 I
Table s5-3.2 SHOROMM VE5SEL LEVEL TRIP ELEVATION CORRELATION INCE S ABOVE: INSTR. VESSEL TAF ZERO (3) ZERO
- (1)
REFERENCE fl[SCRIPTION 69.5 586.25 227.69 Tap "a"~ nozzle ~5 team tap for condensing chambers. 576.75 218.19 60 Marrow and wide range upscale. Level 8 RCIC, HPCI Turistne Steam Inlet valve closure. Close main turbine ~ 212.69/216.94 54.5/58.75 571.25/75.5(4) stop valves. Trip feed pumps. 42.75 559.5 200.94 Level 7 Feedseter control high level alare. . 33.4 550.25 191.69 Level 4 Feeduster control low level alarm. 4, Scram and close RpRt shutdown cooling 12.5 529.25 i' Level 3 isolation valves. ADS level peralssive. 170.69 For wide, narrow, shutdown / upset range Instr. zero instr. Narrow range and shutdown range 158.19 0 516.75 downscale.
-7.75 509.0 150.44 Tap *b" nozzle Narrow range tap (variable leg). 483.5 124.94 -33.25 feeduater sparger.
Initfate RCIC and HPCI. 'Close primary tevel 2 system isolation valves (except Rim shut- ^ down isolation valves). Trip recirculation -38 478.75 120.19 puses. Close M51Vs. 408.56 50 -108.19 . fuel zones upscale. t Notes: (1) Top of active fuel, approximate. (3) Level instrument zero. (2) Vessel zero, cold (approximate). (4) m/y, x = L8 for HPCi/RCIC y = L8 for feeduater and sein turbine trip.
+o %
1 Table B5-3.2 (continued) SHORDMH VESSEL LEVEI TRIP ELEVATION CORREtATION i Im.m.S A00VE: TAF INSTR. VESSEL REFERENCE flESCRIPTION (1) ZERO (3) ZERO a tevel 1 Initiate LPCS med Rett. Start Div. 1 Div. 2,and Div. 3 standby diesels. Contribute to ADS. 25.69 -132.5 384.25 Nfde range downscale 8'.19 -150 366.75 TAF Top of active fuel. Fuel zone inst. Zero. 0 -158.19 358.56 Tap "c" Wide range tap (variable leg). -0.56 -158.15 358 ., Jet pump section. -50.56 -208.75 308 BAF Bottom of active fuel. Fuel zone down- ! scale. - 150 -308.19 208.56 l D 1 Tap "d" Fuel zone var 1able leg. - 226.56 -384.75 132 Source: Shoreham PRA. Table B.5-6. i
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O D(NSATE CHAMBtR WPSCT V r3! 1 (+) 180 in. R
- RfCCRD (TYPICAL)
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TYP t NARA0W RANGt + 50 TAP 5 TRANSMITTER a g UP$tT/$MUTDOWN I I TRANSMITTtt [ j r _ !. _m::.s _ t_se_ _ _ _ _ u_tt_( _w_) _ _ _ _ _ _ _ _ _ _ . o _ru_tt zoat p TTP 4 INST ltRD TAPS
,# vint RAN u ttvet g TRAN5NITTER S t 2 a a
. - - _ -. 'f'"_'!_^E85_N_'._ _. __ _ ._ _ _ _ _. E . E J..L !50 i a-1 TYPt Of 2 TAPS
. a 8 Furl ZONE TaAN5MITTER Source.: Shoreham PRA Figure B.5-4. !
Figure B5-1. Level Instrument Ranges i 1 m
( 1 i I SHORERAM NUCLEAR POWER . STATION PROBABILISTIC RISK ASSESSKENT-BASED INSPECTION PLAN Feeduster, Coodensate, and Power Conversion Systems (FW/CD/PCS) , TABLE B6-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION i CONDITIONS TRAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures, training and check-off lists in ae-cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the probability of failure for the conditions listed below. The most relevant aspects are designated for each condition as follows: - PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedores, check-off lists, trainir.g, etc. Mission Success Criteria The Power Convetsion System (PCS) consists of the Main Steam System, the Main Turbine Generator, the Turbine Bypass Valves (4 valves, 25% capacity), the Main Condenser, and the Circulating System. It functions to convert reac-tor power to electricity, or to merely reject heat to Long Island Sound via the turbine bypass valves, the condenser, and the circulating water system. The feedwater and condensate systems transfer the condensate from the main condenser hotwells to the reactor feedwater pumps, pre-heat the feedwater and return it to the reactor pressure vessel to be converted to steam. These systems are not initiated automatically but are normally operating, non-safety systems which automatically control reactor level within a predetermined range. Manual / automatic startup, operation and shutdown take place from the con-trol room. Loss of offsite power automatically trips, the main turbine, the . f eedwater, and the condensate systems. Automatic shutdown of the f eedwater system also occurs upon isolation of thq MSIVs or a FW turbine trip.
- There are two steam turbine-driven FW pumps, two condensate booster pumps, and two condensate pumps, all divided into two partially separated flowpaths A and B. At least one flow path containing one FW pump, one conden-sate booster pump, and one condensate pump is required to maintain reactor
..Mmw*...,-- . .- 2
i
. e l i
water level following a trip. The CD system can be used without the FW system to maintain reactor level if actions are taken to depressurite the reactor to below approximately'540 psia. A. Feedwater System
- 1. Failure of Long Tern Operator Actions to Control Feedwater During Cool-down (e.g. Failure to Provide Long-Tern Makeup to the Condenser)
Operator actions are an essential aspect of long-ters feedwater control during cooldown. Loss of feedwater can occur by such actions as failure to provide askeup to the condenser to maintain proper condenser level (OP). I
- 2. Miscalibration of Sensors Causes False FW Systes Trip
)
Miscalibration of sensors such as high reactor water level can cause FW i isolation due to a falso high water level signal or failure of the FW pump j turbines due to gross moisture carryover (PC.P.T). j Miscalibration of the lue reactor pressure sensors can cause FW pump trip due to a low turbine steam supply pressure, although condensate flow would continue (PC.PT). i Miscalibration of or f alse signals from the low condenser vacuum sensors, paths A and B. can also cause an automatic trip of the FW system on low con-denser vacuum, although condensate flow would continue in this case as well ( PC .PT) .
- 3. Common Mode Failure of All Of f site Power Sources Common mode failure of all offsite power sources causes an automatic trip of both the feedwater and condensate systems. Refer to emergency electric power system table 1, case 2, for specific inspection information.
- 4. FW Systes Trips Due to Failure in FW Control Logic, Failure of FW Turbine Control Causing Loss of H.P. Steam. Spurious MSIV Closure, or Genuine Isolation Signal Causing MSIV Closure A FW system trip causes loss of the system. Such trips can be caused by several instrumentation and controis-related problems such as failure in the FW control or FW turbine control logic components or a spurious closure of the MSIVs. A genuine containment isolation signal also causes MSIV closure lead-ing to FW trip. The former f ailures can be related to periodic calibration and testing activities while the latter is part of the normal system response ,
$ to a containment isolation and, in fact, the inspection focus should be on en-suring this response (PC PT).
- 5. Transient condition Leads to FW Trip y
Sudden, sharp etriations in operating conditions such as reactor water level and other process variables can cause FW trip. Such transients can be caused by operator errcre (OP). O
i l
- 6. Stean Jet Air Ejector Legs Unavailable and Mechanical Vacuus Pump f 8
Unavailable Cause Loss of Condenser vacuum L The loss of the capability of the steam jet air ejectors to maintain I condenser vacuum combined with unavailability of the mechanical vacuum pump will lead to a loss of condenser vacuum which in turn leads to loss of FW. a) Stesa Jet Air Ejectors (SJAE) Legs Unavailable This can be caused by: ; i) Loss of Main Air Ejectors (OP.MT) l
- 11) Test or Maintenance of SJAE (PT.MT) iii) 0f f Gas Systen Failure / Trips (OP.MT) ,
b) Mechanical Vacuus Pump Unavailable : l This can be -caused byt \ i i) Operator Fails to Start Pump (OP)_ t
- 11) Pump Les in Test or Maintenance (PT.MT) 111) Pump Fails to Start and Run (OP,PT) iv) Co,ndenser Vacuus Limitation (OP) v) 09erational Limit Prohibits Pump Operation Above 5% Power (OP)
B. Condensate System 1.. Failure of Long Ters Operator Actions to Control Condensate During Cooldown (e.g. , Failure to Provide Long-Term Make-up to the Condenser) Operator actions are an essential aspect of long-term condensate control l during cooldown. Loss of condensate can occur by such actions as failure to provide make-up to the condenser to maintain proper condenser level (OP).
- 2. Common Mode Failure of All Offsite Power Sources Common mode f ailure of all offsite power sources causes an automatic trip
- of both the feedwater and condensate systems. Refer to Energency Electric Power System. Table A5-1, for specific inspection information.
- 3. Flow Control Instruments Fail to Supply Sigr.a1 or Supply False Signal to Trains A and B Failure of the flow control instruments to supply signals or by supplying false signals to the FW/CD flow control devices can cause a CD system trip (PC.PT).
-= . .
i i
* * .g9 i
- 4. Rupture of Piping or Heat Exchangers ;
Rupture of piping or of the various heat exchangers within the CD system , can cause unavailability of the CD system (IS).
- 5. Condensate or condensate Rooster Pumps A and B Fail to Continue Running Tailure of either the condensate pump or condensate booster pump in Train '
A combined with f ailure of either pump in Train B will cause loss of the CD system (OP.MT).
- 6. Condensate Domineraliser System (CDS) Failure '
Since the CDS processes all CD flow from both condensate pumps, any , f ailure in the CDS can cause total loss of the CD system (OP.MT).
- 7. Rupture of Condenser Botwell A rupture of the condenser hotwell tubes would allow contaminated circulating water to enter the CD system or cause loss of the hotwell inventory, thereby preventing CD system flow (IS,0P).
- 8. Failure of Drain Coolers in FW Hester Legs A and B Failure of both FW heater lege A and B drain coolers will cause CD system unavailability (IS,0P,MT).
C. Power Conversion System (PCS) f
- 1. Failure of Turbine Bypass Valves Mechanically or Electronica11y The 4 Turbine Bypass Valves are automatically and sequentially '
controlled. Post-accident they must be used since the main turbine is not available.
- 2. Loss of vacuus Vacuum can be lost by a leak in the condenser or by loss of the circulating water system.
6
1 l l SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED r INSPECTION PLAN ! Feedwater, Condensate, and Feuer Conversion System (FW/CD/PCS) TABLE B6-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE NUMBER TITLE COMPONENTS MODES l 41700 Training FW control system A11, 5,6 i I SJAE, Mech. vacuu's pumps, offgas system l condensate & booster pumps. CD domineraliser system, condenser, j drain coolers I i
-52051 Instrument Components and Reactor High level, A 2,4 l Systems-Procedure Review Reactor low pressure, Bt3 l Condenser low vacuum Ct1 j sensors, FW control 2 logic 52053 Ir.strument Components and l Systems-Work Observation i 1
52055 Instrument Components and Systems-Record Review 56700 Calibration 61725 Surveillance and FW control system At2,4,6 ) Calibration Program FW turbine control, Ct1 i Isolation signal SJAE, Mech. Vacuum l pumps CD flow control system, heat exchangers, , Piping, Condenser Hotwell, Drain Coolers 61726 Monthly Surveillance Observation 72700 'Startup Testing-Refueling ,
,w,- , .,-c g
9}. j TABLE B6-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION (Cont'd) PROCEDURE FAILURE' NUMBER TITLE COMPONENTS MODES 11707 Operational Safety SJAE, Mech. Vacuus A6 i Verification Pump, CD flow control instruments, B:5,6,8 condensate & booster C:1,2 pump' s ,CD Domineraliser 71710 ESF System Walkdown System Drain Coolers 62702 Maintenance (Refueling) > 62703 Monthly Maintenance 1 Observation 6 4 t
e . ! SHOREHAM NUCLEAR POWER STATION PROB ABILISTIC RISK ASSESSKENT-BASED INSPECTION PLAN . Feedwater, Condensate, and Power Conversion Systems (FW/CD/PCS) TABLE B6-3 MODITIED SYSTEM WALKDOWN LIST
. i b
A. Feedwater System Since the large majority of system f ailure modes involves operator errors, false instrument signals, control logic failures or miscalibration of sensors, a system walkdown will reveal little tangible information. However, , the probability of failureo leading to loss of condenser vacuum through fail-ures of the steaa jet air ejectors and the mechanical vacuum pump can be reduced by system walkdown for those components described in the next section for the Condensate System. (Loss of condenser vacuum causes loss of the Teedwater System but not the Condensate System.) B. Condensate System The failure modes for the Condensate System consist, for the most part, of operator errore, failure of the flow control instruments, loss of off-site power, failure of the Condensate and Booster Pumps to continue running, etc. Again, a system walkdown will yield very little tangible inf ormable inf orma-tion with the stesption of the following! C. Power Conversion System (PCS) The main failure mode of the PCS is failure of the turbine bypass valves either mechanically or electrically. MSIVs are covered in Table C2.
- 1. For FW System Failure Caused by Loss of Condenser Vacuum Desired Actual Description I.D. No. Location Position Position
- a. Air Ejector, Condenser IN21-16V- Turbine Bldg. Closed A Bypass 00llA El. 37' Down-stream RT
- b. Air Ejector, Condenser IN21-20V- Same Open A Inlet 0012A
- c. Air Ejector, Condenser IN21-20V- Same Open e
A Outlet 0013A
'd. Air Ejector, Condenser IN21-16V- Same Closed B Bypass 0011B
- e. Air Ejector, Condenser IN21-20V- Same Open B Inlet 0012B Same Open
- f. Air Ejector, Condenser IN21-20V- Same Open B Outlet 0013B
I l
. e !
t- ; l J
- 1. Condensate (Cont'd)
Desired Actual i Description I.D. No. Location Condition Condition
- g. Mechanical Vacuum Pump Turbine Bldg. Status (Should be off during El. 15' power operation.)
- h. Offgas System- Radweste Bldg. Status
- 2. Condensate Domineraliser System Desired Actual Description I.D. No. Location Condition Condition 3a. Drain Cooler A Turbine Bldg. Status El. 15' 3b. Drain Cooler B Same Status
- 4. Walkdown of High Rupture Risk Components Desired Actual Description I.D. No. Location Conditi on Oonditiot,
- s. Piping __,
Status
- b. Heat Exchangers Status
- c. Condensor Hot Well -
Status Referenco Document _s Title I.D. No. Rev. Date (LILCO Operating Procedures)
- 1. Feedwater and Feedwater Control SP 23 109.01 10 3/21/86 Systems
- 2. Condensate SP 23.103.01 7 4/21/86
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- 93 SHORERAM NUCLEAR POWER STATION l PROBABILISTIC RISK ASSESSKENT-BASED i INSPECTION PLAN Standby Liquid Control (SLC) System ,
TABLE Cl-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION l t f CONDITIONS TRAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or j normal and emergency operating procedures, training and check-off lists in ac-cordance with the Technical Specifications and relevant NRC bulletins and in- , formation notices should reduce the probability.of failure for the conditions listed below. The most relevant aspects are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and , training. OP , Normal and emergency operating procedures, check-off lists, ; training, etc. Mission Success Cetteria y The purpose of the Standby Liquid Control (SLC) system is to achieve sub-criticality in the reactor by insertion of negative rcuctivity into the core independently of the Control Rod Drive system. The SLC system is manually ini-tinted from the main control room if the otierator determines that sufficient # negative reactivity cannot be achieved or osintained with the control rods. - Only one SLC pump is allowed to operete at one time. Once th,e manual initta-tion signal is received, both explosive valves art accuated, one of the two pumps is started, and the neutron absorber solution is pumped into the RPV.
- 1. Operator Fail's to Initiate the SLC Failure of the operator to initiate the SLC when required is a high probability failure mode. Operator training is a crucial factor (0P).
- 2. Inadequate 'SLC Poison Mixing in the Reactor Vessel Failure of the SLC poison, sodium pentaborate, to adequately six within the reactor vessel will prevent achieving sufficient negative reactivity.
,- , - , - . - , --- , - , + - - -
t i
?
- 3. Insufficient Boron Concentration in the SLC Storage Tank The SLC Storage Tank is required to maintain 4290 lbs of sodium ,
pentaborate at all times. Insufficient concentration will prevent adequate negative reactivity when required (PT,0P.TS).
- 4. Suction Line From SLC Storage Tank Plugged Plugging of the SLC pumps suction line from the storage tank by -
precipitation of the sodium pentaborate solution will prevent insertion of negative reactivity when required (PT.0P).
- 5. Check Valve 006 or 007 Fails - Stuck Closed Failure of either check valve. 006 or 007, prevents SLC pump flow discharge f rom entering the reactor vessel (PT). l
- 6. Normally Locked open Manual valve 008 in Common Pump Discharge Line to the RPV Fails Closed
- Inadvertent closure of manual valve 008 by maintenance or test error prevents SLC pump flow discharge from entering the reactor vessel (PT MT,0P).
- 7. Normally Locked Open Manual valve 001 in the Supply Line From the SLC -
. Storage Tank Fails closed -
Inadvertent closure of annual valvo 001 by maint6 nance or test error > preve.nts flow from the SLC storage ttnk te the SLC puepe suct ton (FT,MT,0P).
- 8. Normally Locked Closed gnual Valve 014 or 015,in t33 SLC Pug Suttion ',1na Feils Open l
Inadvertent opening of manus 1 valve 014 will allow desiperalized cater to bypass the GLC s?.orage trok while inadvertent c.pening of manual valve 015 will divert flow from the SLC storage tank away f rom the suction of the putas and ; to the drain. tank. la both cases, edequate SLC flow into the roector vessc3 ' will be prevented (PT.NT,0P).
- 9. SLC Tank Fails During ATWS Demand Failure of the SLC Tank during an ATWS demand prevents adequate SLC flow into the reactor vessel when required (PT).
4 A - . . _ , - _ _ _ _ _ . _ _ . - . , - . _ _ _ _ _ _ - - _ - _ _ _ . . _ - - - _ .
SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Standby Liquid Control (SLC) System TABLE Cl-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE NUMBER TITLE COMPONENTS MODES 41700 Training Manual initiation 1 avitches 61725 Surveillance and Storage tank, 3-9 Calibration Control Pumps, Manual Valves Program 001, 008, 014. 015; Check velves 006, 007 61726 Monthly Surveillance Observation 71707 Operational Safety Verification 71710 ESF System Walkdown 62702 Maintenance Manual valves 6-8 001, 008, 014, 62703 Morithly Maintenance 015, Pumps y , _ Observation __
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k TMILE C1-? MOelF?!D SYSTEM 18AtM00mm 10 set'of l .D. Sep. Ossited .%."Lat Pouer $appply/ IIngetred Actuel Descriptton G.E. LI LCO Locatten Positton %sition fresher sb. Location % sition Posittee 9.SLC 5%mp A 1C4l' WA l' ~ ~ ~ ~ ~ 8'CC-1113/9CG en P-024A 3 10.SLC Pump B IC4 t' Some 8800-1123/4AE Os 1 P-024 l '- t1.SLC Tonk 1C4 l' 8tr. 6-1dg. etC-1120/1EF On Moeter A WOO 94 E', t;2' I2.SLC Tonk 1C4 t' Sopr3 . 8E00-11#C/Isee On Meeter B +0098 J I3.SeeIb VeIve l'-004A IC4t' 8ER ._ 8ECC-1II3/9CG Oe (Explosive EV-010A ' veIvet Pwr, A 14.Sgelb Velve F0048 fC41* Some 9ECC-t 123/4AE On (Ewplosive EV-010B l Velve) Pwr. B . ! 15.teorsei Hoet 141- 800-t12A/ Os , Tracing 017FSH 2Mt(Rt ClreuIts ime1-essL-31Bfett Os j , IM 1-PIEL-2B9tK2 Om ' e ! 16.nndendent tag 1 , setE-1I1A/ Off i i Huet Trac- O t9FSH 2L8t(Rt Ing ClrceIts 18481-Ptol-22814C1 On 1481-P90L-2AGtR2 Oe t#US-Pfel-se29 De ! 17.SBLC Tonk #C41' Temp.elttiln Temp. Element TE-001 toch spec IstR a i theits , I i I t i floference Document ; i - 3 Tltto str. Strv. Date i (LILCO @ereting Proceduret SP-23.123.01 to 8/6/85 Standby Ligold Omtrol i i e _ ~ . . _ - _ - _ _ _ _ _ _ - . - _ _ . - . _ _> _ ,
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101 SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Main Steam Isolation Valves (MgIVs) TABLE C2-1 IMPORTANCE BASIS AND PAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activities, procedures and training and/or normal and emergency operating procedures, training and check-of f lists in accordance with the Technical Specifications and relevant NRC bulletins and information notices should reduce the probability of failure for the condi-tions listed below. The most relevant aspects are designated for each condi-tion as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled maintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, training, etc. Mission Success Cri'geria Av There are two redundant MSIVs for sach of the fove main stoaa 11aeb. They are cit 4td by springs and air /lg and opened by cir/@2 Their closure time is ' 3 to 10 seconds. Post-accident they must be open to utilize the PCS for host removal and must be closed for containsent isolation. They are leak tested each refueltog cycle (or 2 years) per Technical Specifications.
- 1. 113IVs Fail to Reasin Open or Operetor Fails to Pe-Ocen Valves During a transient the MSIVs may be spuriously closed or they any close automatically on low reactor pressure. The MSIVs must be open for the Power Conversion Systen to be used for heat removal from the reactor and the con-tainment. If the MSIVs have closed and they are needed for a heat removal path, the operator aust reopen then remote manually from the control room. N 2 isolation va'1ves (P50*MOV-105A and B) must be realigned before the MSIVs can be reopened. The ac'cident override switch may need to be operated to open the N2 1 solation valves (PC, PT, NT, OP).
- 2. MSIVs Fail to Close or Leak Excessively .
If containment isolation is required, then the MSIVs must close and pro-vide a leak tight seal. (PT MT). l L
r
-102- . .
- - SHORERAM NUCLEAR POWER STATION PROSABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Main Steam Isolation Valves (MBifs)
TABLE C2-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION PROCEDURE FAILURE NUMBER TITLE COMPONT.NTS MODES 41700 Training MSIVs 1 61725 Surveillance Testing & MSIVs 2 Calibration Program 61726 Monthly Surveillance Observation 61720 Containment Local Leak Rate Testing 62702 Maintenance MSIVs 1,2 62703 Monthly Maintenance Observation . L l I ri 1 l~ L. 0 .- e f- ,. .
103-SHOREHAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Main Staas Isolation Talves (MSIVs) TABLE C2-3 MODIPIED SYSTEM WALKDOWN COMPONENT DESIRED PANEL DESCRIPTION ID NO. LOCATION READING ACTUAL
- 1. hrbine Main Steam MCB-01 MCR Approx.
Console Line Flowmeters Equal 2 1Hi1*MPX N2 1 solation P50*MOV- MCR Open Valves to In- 103A board MSIV F50*MOV- MCR Open 103B P50* MOV- MCR Open 105A P50*MOV- MCR Open 105B
- 3. lH11*MPX N2 1 solation P50-01V- RB Open Valves to out- 0872A Board MSIV F50-01V- RB Open 0872B P50-01V- RB Open
# Ob92A P50-01V- RB D)>en ~
06928 1* 9 l l
~104- . . ;
SHOREHAM NUCLEAR POWER STATION UNIT 1 PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Drywell Area Components TABLE D1 SYSTEM WALKDOWN Discussion ,
'!he drywell is a high radiation area that is inerted with nitrogen during normal plant operation and is therefore not accessible. To facilitate the in-spection of components located within the drywell during periods when access ;
is permissible, those components are summarized below: GE LILCO DESIRED ACTUAL SYSTEM COMPONENT ID NO. LOCATION POSITION POSITION
- 1. SLC Standby Liqyid Control F008 C41* El. 100' Locked ,
Manual Isolation Valve HV-023 open to Rx Vessel
- 2. RHR RRR System I F0608 1 Ell
- El. 78' Locked Injection to Rx Vessel HV-71B Open ,
- 3. RHR RHR Systes II F0605 1 Ell
- El. 78' Locked Injection to Rx Vessel HV-71B Open __
- 4. HPCI Steam Suply Inboard F002 IE41* 11. ? 01osed _ , , _
1 isol. Valet to RPCI MOV-04! Turbiu
- 5. kCIC St has Suppl 3 frboard F007 IE51* EL. t Ci v,id. __
Itcl. Val n to 2.CIC MOV-041 Turbino ! PSD*02V (1. t Opat. i 6. ADF/ y furply to ADS
. Tj SRVs ( so !!SIVs 0696A P50*02V E1. 7 Open 0696B
- 7. RHR RHR. Pump Strainer FL-078 Supp. Pooit Clear
- 8. RRR RHR Pump Strainer FL-079 Supp. Poolt clear I
- 9. CS Core Spray Loop A' F007A 1E21* . El. 125' Locked Injection Manual HV-071A ,
Isolation Valve Open 10.CS Core Spray Loop B F007B lE21* El. 125' locked ) Injection Manual HV-071B j Isolation Valve Open tAetually in Wetwell. Inspect only when accessible. l N E t;'"
~ -
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Reference Documents Title No. Rev Date
- 1. Automatic Depressurization SP.23.101 01 7 3/5/85 System ( ADS)
- 2. Core Spray System (CSS) SP.23 203 01 19 4/10/86
- 3. High Pressure Coolant Injection SP.23 202 01 12 4/10/86 (HPCI) Systen .
- 4. Reactor Core Isolation Cooling (RCIC) System SP.23.119.01 14 3/24/86
- 5. Residual Heat Renoval (RRR) SP.23.121 01 21 2/19/86
- 6. Standby Liquid Control (SLC) SP.23 123.01 10 8/8/85 6
1 1 4 0 l. t-u
106- .. , a . li. ' SHORERAM NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN b Significant Bussan Errors TABLE D2 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Since these items are significant human or operator errors identified by the PRA, the primary inspection areas to address them would be operator training and procedure reviews. These areas are not listed in order of importance.
- 1. Failure to initiate Standby Liquid Control (SLC) when required (e.g.
during Anticipated Transient Without Scram-ATWS)
- 2. Miscalibrations of - Reactor Vessel Level
- Pump Discharge Valve AP Instruments for Core Spray and Low Pressure Coolant Injection (LPCI)
- 3. Failure to manually initiate the ADS, if required and automatic initiation fails.
'WS scenario.
- 4. Failure to inhibit ~ ADS duri-
- 5. Human errors during normai ition leading to feedwater system trips.
- 6. Failure to recover offsit. tr.
- 7. Failure to properly r .omponents after test or maintenance.
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-107-1 SHOREHAM NUCLEAR POWER STATION PROBARILISTIC RISK ASSESSHENT-BASED
- INSPECTION PLAN Containment Systems TABLE D3 IHPORTANCE BASIS AND FAILURE MODE IDENTIFICATION CONDITIONS THAT CAN LEAD TO FAILURE General Guidance Surveillance of the licensee's periodic calibration, testing and/or pre-ventive or unscheduled maintenance activitica, procedures and training and/or normal and emergency operating procedures, training and check-off lists in ac- ,
cordance with the Technical Specifications and relevant NRC bulletins and in-formation notices should reduce the probability of f ailure for the conditions listed below. The most relevant aspects are designated for each condition as follows: PC - Periodic calibration activities, procedures and training. PT - Periodic testing activities, procedures and training. MT - Preventive or unscheduled naintenance activities, procedures and training. OP - Normal and emergency operating procedures, check-off lists, 1 training, etc. Mission Success Criteria
' 9 The main containment systems considered here are the Primary Containment, the Suppression Pool, the Secondary Containment and the Reactor Building Standby Ventilation System (RBSVS). Tne Primary Containment serves to contain fission products post-accident and consists of the drywell and suppression pool walls and the containment isolation system. From a risk standpoint, small leaks are not a major concern, but large leaks, on the order of the tech spec limit. (La) or larger can be very significant to risk. The Suppression Pool serves to limit containment pressure, scrub fission products, and provide -
a source of water for the ECCS. The Secondary Containment and the RSSVS work together to help ensure any -leakage f rom the Primary Containment is either captured or released through an elevated, filtered vent path. The RBSVS also has area coolers to limit temperatures in the various ECCS equipment areas. Since there.are no detailed fault trees or minimal cutsets in the PRA for these containment systems, this inspectinn plan does not have the detail of the front end systems. Important Aspects of Containment Systems
- 1. Emergency procedures and training established to vent containment wetwell (Suppression Pool) during containment overpressure scenarios.
' ~ ~ ~
7 j$$( . 108 :
. 7 Maintaining functionability of SP due to its multiple uses of condensing si , 2.
steam, scrubbing fission products, and providing the ECCS with water.
+
- 3. Sequences that bypass containment or result in early containment failure
'are important to offsite risk. Important areas associated with bypass of containment are containment isolation failure due to valve failures or grois leakage through containment. One important area associated with
.early containment overpressure is SP bypass.. (The SP bypass leak test ' -
addresses this.) t 0 9 6 l-L l (1 , i u . L i . l l 1 1 .
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-110- ' '
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-118- . .
4 P I 1 SHOREHAM NUCLEAR POWER STATION l PROBABILISTIC RISK ASSESSMENT-BASED L INSPECTION PLAN 1 Rasetor Building Standby Ventilation System (RBSYS) and I Control Roon Air Conditioning (CRAC) Chilled Water System ) l TABLE D4-1 IMPORTANCE BASIS AND FAILURE MODE IDENTIFICATION . , i h CONDITIONS THAT CAN LEAD TO FAILURE L General Guidance j ! Surveillance of the licensee's periodic testing and/or preventive or un- l scheduled maintenance activities and procedures and/or normal and emergency operating procedures and check-off lists in accordance with the Technical (. Specifications and relevant NRC bulletins and information notices should re-l doce the probability of f ailure for the conditions listed below. The most i l relevan*. aspects are' designated for each condition as follows: l PC - Periodic calibration activities, procedures and training. ! PT - Periodic testing activities, procedures and training.
- MT - Preventive or unscheduled maintenance activities, procedures and t raining. ;
L OP - t oraal and emergency operating procedures, check-of f lists, train-ing, etc. Mission Success Criteria The RBSVS and CRAC chilled water system (CWS) are designed to remove heat gains generated in the control room; the relay, emergency switchgear, and com-puter rooms; the Motor Generator (MG) rooms; the Motor Control Center (MCC) rooms; and the reactor building secondary containment during both normal and ; accident conditions. Heat removed from these areas is rejected through the Service Water System (SWS) to Long Island Sound. Both systems consist of two identical redundant subsystems, designated A and B. Each subsystem is comprised of two centrifugal water chillers, a con-densing water and chilled water pump for each chiller, unit coolers, AC units,
- a. surge tank, piping and necessary contr01s for proper operation. Either of
- the redundant subsystems is adequate to cool all the saf ety-related equipment in all.three divisions.
e
- 1. Failure of Both' Trains of Service Water Since the SWS is required for RBSVS and CRAC CWS operation failure of the h SWS causes failure of the RBSVS and CRAC CWS as well. Refer to Table A4 f or the SWS inspection plan.
4
113-
- 2. ' Failure of Both Trains of RVSYS and CRAC CWS Failure of both trains of RBSVS and CRAC CWS constitutes system failure.
Failure of a single train can be caused by
- a. Operator Shuts off the Operating Train Without Recognizing that the Alternate System is Ineffective (0P)
- b. Unit Coolers 2A and 3A (or 2B and 3B) Fail Because Normally Closed PCV019A (PCV019B) Fails Open so that the Header Crossover Bypass Line is Inadvertently Open (PT, OP)
- c. Chilled Water Supply Header Blocked Because Normally Open Manual Valve 028A (0288) Fails Closed (PT, OP)
- d. Chilled Water Return Header Blocked Because Normally Open Manual j valve 027A (0278) Fails Closed (PT, OP)
- e. System A (System B) Pipe Failure (PT, OP)
- 3. SWS Train A (or Train B) in Maintenance and Train B (Train A) of RBSVS and CRAC CWS Fails Maintenance on one train of SWS combined with failure of the opposite
- train of RBSVS and CRAC CWS constitutes system failure.- Refer to Table A4
.for the SWS inspection plan and to Item 2 above for failure of a train of RBSVS and CRAC CWS.
- 4. Miscalibration or Common Made Failure of Reactor Water Level Sensors and er Kiscalibration of Drywell High Pressure Sensors During an accident condition, the_ accident signal starts all four water -;
chillers, their associated chilled water pumps, and condensing water pumps. The System A and B crossover and crossover bypass valves close, providing two full capacity independent chilled water systema. All coolers in the RBSVS be-come operational. Miscalibration or common mode failure of reactor water level sensors and of the drywell' high pressure sensors prevents the accident ! signal from being generated when required. This failure is common to the LPC1 and CS systems (PC, PT).
- 5. Common Mode Failure of All Offsite Power Sources and Diesel Generators 1 and 2 Fail and the Direct Power Sources (125 VDC System) is not Repaired Within 2 Hours of Failure Loss of all offsite power sources combined with failure of the Divisions 1 and II diesel generators to start and run and failure to repair the 125 VDC power sources within 2 hours of failure will prevent actuation and operation of the RBSVS and CRAC CWS when required. Refer to Case 2. No. 5 of the Emer-gency Electric Power System inspection plan, Table A5-1, for further details.
i
-114- , ,
SHOREHAM NUCLEAR POWER STATION , l PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLAN Reactor Building Standby Ventilation System (RBSYS) sad ' Csatrol Room Air Conditioning (CRAC) Chilled Water System (CWS) i
- l. TABLE D4-2 I&E INSPECTION PROCEDURES FOR SYSTEM OPERATION .
V PAILURE PROCEDURE MODES NUMBER TITLE COMPONENTS L 56700 Calibration Reactor Low Level 4 Sensors High -Drywell Pressure Sensors Control Circuitry 4 l 61725 Surveillance & Calibration i Control Program
=$2051 Instrument components and Sensors , Switches, 4 ,
f Systems-Procedure Review Control Circuitry 1 ,, 52053 Instrument Components and Sensors, Switches , 4 Systems-Work Observation Control Circuits 52055 Instrument Components and Sensors, Switches, 4 Systems-Record Review Control Circuits Maintenance (Refueling) Service Water System 1,3,5 62702 Monthly Maintenance 0bservation Electric Power Systems 62703 71707 Operational Safety Verification Valves PCV019A/B 2 Manual Valves 027A/B 71710 ESF System Walkdown 028A/B Training RPSVS & CRAC CWS 2 41700 41701 Requalification Training W , _ ._-v, -
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-118--
SHORERAM NUCLEAR POWER STATION j- , PROBABILISTIC RISK ASSESSMENT-BASED ' INSPECTION PLANS i TABLE El - PLANT OPERATIONS INSPECTION GUIDANCE i Recognizing that the normal system lineup is important for any given ' standby safety system, the following human cerors are identified'in the PRA as important to. risk.. SYSTEM FAILURE DISCUSSION i Reactor Inspect sensing instrument racks for Table Al-1, Item 2 Protection correct valve configuration, labelling
' System (RPS) and-separation.
Ensure no abnormal RPS alarms in the Table.Al-1, Item 3 L cont-rol room,.and verify and bypass conditions are properly logged and justified. . Check RPS panels for jumpers and lif ted Table Al-1, Item 4 leads. Documentation of same with the appropriate review and approval is required. Automatic Failure of All ADS Valves or Valve Table A2-1. Item 1 Depressuriz- Operators Due to an Adverse Environment ation System' Inside, Containment Affecting (ADS) . Cable / Solenoid Coils Miscalibration of Reactor Water Level Table A2-1, Item 2 Channels or Transient Disables Level Instrumentation, and Operator Fails to Initiate ADS Nitrogen Supply Contaminated and Table A2-1. Item 3 Solenoids Disabled Operator Error During Maintenance Table A2-1, Item 4 Disabled Valves or Solenoids Common Mode Check Valve Leakage Bleeds Table A2-1, Item 5 of f ' Accumulator Pressure and Failure of Backup Nitrogen Supply or Loss of Backup Supply Due to Containment Isolation Common Mode Miscalibration or Transient Table A2-1, Item 6 Disables Pressure Switches on RHR Pumps A, B, C, and D thereby Inhibiting ADS from Actuation Because RHR Pumps Are Not Available
.- . 119 Table El (Cont'd)
SYSTEM FAILURE- DISCUSSION Automatic Failure of Reactor Level Channels A and Table A2-1, Item 8b Depressurie- 8 and Operator Fails to Initiate ADS ation System Given Auto System Failure: Logic Roset ( ADS) (Cont'd) Switch in Reset Position Preventing Automatic Initiation of an ADS Signal
. Residual Heat. Suppression Pool Water Unavailable Due Table A3-1, Item 2 Removal System to Clogged Strainers i LPCI/CS Manually Shut Of f on High Level Table A3-1, Item 4' and During an Accident and Level Instrumentation Fails to Automatically Low Pressure Initiate After Having Worked at High Coolant Injec- Pressure tion (LPCI)
System CS/LPCI Manually Shut Of f on High Level Table A3-1, Item 5 During an Accident and Operators Are Totally Distracted Over a Period of 2-6 Hours LPCI/CS Manually Shut Of f on High Table A3-1, Item 6 , Level During an Accident and Operators Fail to Follos Procedures Af ter j, Recognizing the Problem Failure of All Four LPCI Pumps A,B,C,D Table A3-1, Item 7 to Start and Run Failure of Flow f rom Both LPCI Loops A Table A3-1 Item 8b and B to Enter the Reactor Vessel: Normally Locked Open MOV F017A or B at Primary Containment Drywell Penetration Fails Closed Failure of Flow from Both LPCI Loops A Table A3-1, Item 8e and B to Enter the Reactor Vessel: Air Operated Check Valve 050A Fails Stuck Closed Failure of Flow from Both LPCI Loops A Table A3-1, Item 8d and B to Enter the Reactor Vessel: Normally Locked Open Manual Valve F060A or B Fails Closed Both RRR Heat Exchanger Bypass HOVs Table A3-1, Item 10 F048A and F048B Fail Open Both Service Water System (SWS) MOVs Table A3-1, Item 11 34A and 34B on the RHR Heat Exchangers' SW Outlet Lines Fail Closed 9
b +'9- -120- . . Table E-l' (Cont'd) SYSTEM FAILURE DISCUSSION Service. Water All Four SWS Pumps Fail at the Same Time Table A4-1,-Item 2 Systems-(SWS) Loss of Water Supply to Screen Well Table A4-1, Item 3 (Screen Well Water Level Low) B Isolation of One Loop Required Due to Table A4-1, Item 4 Excessive Leakage and Failure to Isolate the Opposite Loop , q One Loop in Maintenance with Two Pump Table A4-1. Item 5 Failures in the Opposite Loop Failure of Operator to Cross-Connect Table A4-1, Item 6 TBSW to Supply Reactor Building SWS Emergency All" Diesel Generators Require Table A5-1, Case 1 l Electrical Maintenance (Shutdown in 12 Hours) and Item 1 l Power System Common Mode Failure of All Offsite , l (EEPS) Power Sources (LOOP) and Direct Power Source (125 VDC System) is Not Repaired l Within 2 Hours of Failure Voltage Transformer Failure ' Table AS-1, Case 2, Item 1 Normally Closed 4 kV to 480 V Bus Table A5-1, Case-2, l Transformer Breaker Fails Open Item 2 l Table AS-1, Case 2, 480 V Bus or Switchgear Unavailable Item 3 j I 4 kV Bus or Switchgear Unavailable- Table AS-1, Case 2, Item 4 ; Common Mode Loss of Offsite Power and Table A5-1, Case 2,- ) One Diesel Generator Fails to Start Item 5 ' and Run. High Pressure Failure of HPCI Pump or Booster Pump Table B1-1, Item 2 Coolant or Turbine to Start and Run
- i Injection ,
(HPCI) System Auxiliary 011 Pump Fails to Start Table B1-1 Item 7 and Run p False Signal Indicates High Area Table B1-1, Item 8 -i Temperature Operator Inadvertently Isolates Shutoff Table B1-1, Item 9 MOV F042 From the Suppression Pool e (
.+.
__ .=w-
-- ~. . - . . -. . . -
-121-
~
7.E * .e Table E-1 (Cont'd) .1 SYSTEM FA1 LURE DISCUSSION j l High Pressure- HPCI Turbine Fails to Start (During Table B1-1, Item 13 1 Coolant Subsequent Re-Starts)- ~
' Inj ection , .
-(HPCI) System Failure of HPCI Pump or Booster Pump to Table El-1, Item 16 1
-(Cont'd) Start and Run (During Subsequent Re-Starts)
. False Signal Indicates High Steam Line. Table B1-1, Item 17 A P, Sensors N004 or N005 q False Signal Indicates High Turbine Table B1-1, Item 18 Exhaust Pressure Sensor N017A or N017B ;
3 False Signal; Indicates Low Pressure on Table B1-1, Item 19 Pump Suction (No. 10) Instability in Steam Exhaust Line Table B1-1, Item 21 Causes High Exhaust Pressure During ' Initial Start-up or Subsequent Start-up HPCI Auto Reset Not Reset Table B1-1, Item 22 .; Reactor Core Failure of RCIC Pump or Turbine to Table B2-1, Item 2 Isolation Start or Run Cooling (RCIC)
,,, System Equipment Room ( Area) High Temperature Table B2-1, Item 5 Sensor 600A, 600B, 601A, 601B, 602B, 603A, 603B, 604A, or 604B Gives False Signal RCIC Pump or Turbine Fails to Start on Table B2-1, Item 7 Subsequent Starts Sensor N006 for Low RCIC Pump Suction Table B2-1, It em 8 Pressure Gives False Signal.for Turbine Trip Sensor N017 or N018 for RCIC Turbine Table-B2-1, Item 9 Steamline High P Cives False Signal or Instrument Line Break Causes-Turbine Trip RCIC Turbine High Exhaust Pressure Table B2-1. Item 10 Sensors N009A or N009B Cive False Signal for Turbine Trip
{( RCIC Turbine Lubrication System Falls Table B2-1, Item 11 Instability in Steam Exhaust Line Table B2-1, Item 12 Causes High RCIC Turbine Exhaust Pressure 1
p ,
-188- . ,
e Table E-1-(Cont'd) SYSTEM FAILURE DISCUSSION RCIC (Cont'd) Automatic HPCI. Shutoff Fails and , Table B2-1, Itam 13-Operator Fails to Shut Off HPCI Core Spray (CS); Common Mode Failure of Both Core Spray Table B3-1. Item 2 .i System Pumps Suppression Pool Water Unavailable Due Table B3-1, Item 3 to Clogged Strainer Failure of Both Core Spray Loop A and_ Table B3-1. Item Sb ! Loop B: Pump A or' Pump B Fails to Start ! 2 and Run Failure of Both Core Spray Loop A and Table B3-1, Item Se Loop B:. Loss of Pump Lubrication i Failure of Both Core Spray Loop A and Table B3-1 Item 5f Loop B: Failure of Strainer Blocks Flow i Failure of Both Core Spray Loop A and Table B3-1. Item 5g Loop B Flow Diverted to Suppression Pool Via Normally Closed Test Line
. MOV F015A or F015.B Failing Open Failure of Both Core Spray Loop A and Table B3-1, Item Sh Loop B: Normally Open Suction Line '
MOV F001A or F001B Fails Closed Reactor Vessel Level Instrumentation Table B3-1, Item 7a , Fails Af ter Having Worked at High-Pres 1ures, and LPCI/CS Manually Shut Off on High Level During an Accident Reactor Vessel Level Instrumentation Table B3-1,-Item 7b Fails After Having Worked at High Pressures, and Operaters Totally Distracted Over a Period of 1-6 Hours-and Fail to Rastart LPCI/CS When Required Reactor Vessel Level Instrumentation Table B3-1, Item 7c . Falls After Having Worked at High
. Pressures, and Operators Fail to Follow Procedures After Recognizing the Problem' and Fail to Restart LPCI/CS When Required Common Mode Failure of Both Loops Table B3-1, Item 8-Caused by Manual Valves F007A and F007B Being Lef t in the Closed Position l
1 l _" "' '~ " ~ " ~ " ' * *
[ t ni is *- 123-jst Table E-1 (Cont'd)
-SYSTEM FAILURE DISCUSSION
" Reactor Failure of One Loop and Maintenance on Table B4-1,-Item 3a ,
Building' the Other Loop, Failure Caused by False I Closed Loop _ fndication of Suction Pressure Stops Pump i Cooling Water System Failure of One Loop and Maintenance on- Table B4-1, Item 3b (RBCLCWS) the Other Loop, Failure Caused by Pump Failure to Continue Running Failure of one Loop and Maintenance on .Teble B4-1, Item 3e l the Other Loop, Failure Caused by < Clogged Pump Filter
?
Failure of One Loop and Maintenance on- Table B4-1, Item 3d
-the Other Loop, Failure Ceosed by HK "
Inlet (MOV-0421,B) or Outlet Valve. (MOV-052A,B) Failing Closed Failure of One Loop and Maintenance on Table B4-1, Item 3e the Other Loop, Failure Caused.by Temperature Control Valva (TCV-00lW/X, 00lY/Z) Failing Closed u ' Feilure of One Loop and Maintenance on Table B4-1, Item 3f-the Other Loop, Failure Caused by Normally Open SW5 Motor Operated Valve
,,, Failing Closed Failure of One Loop and Maintenance on Table B4-1, Item 3g the Other Loop, Failure Caused by Normally Open. Manual valve (16V-009A,B) in SWS Injervion Line Failing Closed Failure of One Loop and Maintenance on Table B4-1, Item 3h the Other Loop, Failure Cauend by HX Ruptures or Excessive Leakage Failure of One Loop and Maintenance on Table B4-1, Item 31 y the Other Loop, Failure Caused by Head Tank Failure to Fulfill its Function to \
Provide NPSH Failure of One Loop rad Maintenance on Table B4-1, Item 3j the Other Loop, Failure Caused by' Pump [ Discharge Check Valve (005A,B C) Failing
. Closed j 1
l l l l l l l l
- .-. . .- - . - ~ ~ . - . - -
.5
-184- . .
i Table - E ( Cont 'd ) SYSTEM FAILURE DISCUSSION e Reactor Failure of One Loop and Maintenance on Table B4-1 Item 3k Building' the Other Loop, Failure Caused by Pipe l Closed Loop Rupture or Excessive Leakage in Pump L Cooling Water Suction Lines, Pump Connections or System Discharge Lines (RBCLCWS)
-(Cont'd) Failure of One Loop and Maintenance on Table B4-1, Item 31 the Other Loop, Failure Caused by Faulty Motor Overload Breaker Stopping Pump Failure of Both RBCLCW Loops Table B4-1, Item 4 Maintenance of One SWS Loop and Failure Table B4-1 Item 5 of the Opposite RBCLCY Loop Reactor Instrument Line Break Table B5-1, Item 2
- Water Level Operator Error Table B5-1, Item 3 i
Flashing of RWL Reference Leg Table B5-1, Item 5 ,
'Feedwater, Failure of Long Tara Operator Actions Table B6-1, Condensate, to Control Feedwater During Cooldown Item A.I.
and Power (e.g., Failure to Provide Long-Term Conversion Makeup to the Condenser) Systems (FW/CD/PCS)' FW System Trips Due to Failure in FW Table B6-1,
. Control Logic, Failure of FW Turbine Item A.4 Control Cauaing Loss of H.P. Steam, Spurious MSIV Closure, or Genuine Isolation Signal Causing MSIV Closure Transient Condition Leads to FW Trip Table B6-1, Item A.5.
Steam Jet Air Ejectors (SJAE) Legs Table B6-1, Unavailable: Loss of Main Air Ejectors Item A.6.a.i. s Steam Jet Air Ejectors (SJAE) Legs Table B6-1, Unavailable Off Gas System Item A.6.a.iii Failure / Trips
- Mechanical Vacuum Pump Unavailable: Table B6-1, Operator Fails to Start Pump Item A.6.b.i Mechanical Vacuum Pump Unevailable: Table B6-1, Pump Fails to Start and Run Item A.6.b.iii Mechanical Vacuum Pump Unavailable: Table B6-1, Condenser Vacuum Limitation Item A.6.b.iv w .-
L
-125-Table E-1 (Cont'd).
SYSTEM FAILURE DISCUSSION Feedwater, Mechanical Vacuum Pump Unavailable: Table B6-1,.
-Condensate, Operational Limit Prohibits Pump- Item A.6.b.v and Power Operation Above 5% Power (OP)
Conversion Systems Failure of Long Term Operator Actions Table B6-1, (FW/CD/PCS) to Control Condensate During Cooldown' Item B.1 (Cont'd) (e.g. , Failure to Provide Long-Term
. -Make-up to the Condenser) .
Condensate System: Flow Control Table B6-1, Instruments Fail to Supply Signal or Item B.3-g, - Supply False, Signal to Trains A and B Condensate or Condensate Booster Pumps- Table B6-1, j A and B Fail to Continue Running Item B.S. Condensate Domineralizer System (CDS) Table B6-1, Failure Item B.6. Rupture of Condenser Hotwell Table B6-1, Item B.7. , Failure of-Drain Coolers in FW Heater Table B6-1, j Legs A and B Item B.8.
- JP' Power Conversion System: Feilure of Table B6-1 Turbine Bypass Valves Mechanically or- Item C.I.
Electronically Power Conversion Systes: Loss of Vacuum Table B6-1, Item C.2. t Standby Operator Fails to Initiate the SLC Table Cl-1, Item 1 f Liquid Control (SLC) System Insufficient Boron Concentration in Table Cl-1, Item 3 the SLC Storage Tank Suction Line From SLC Storage Tank Table Cl-1, Item 4 Plugged Normally Locked Open Manual Valve 008 Table Cl-1, item 6-in Common Pump Discharge Line to the EFV Fails Closed Normally Locked Open Manual Valve 001 Table Cl-1, Item 7 in the Supply Line From the SLC Storage Tank Fails Closed I e
-126-- _. ,
j Table E-1 (Cont'd) ,
. SYSTEM FAILURE DISCUSSION s,
Normally. Locked Closed Manual Valve 014 Table Cl-1, Item 8 SLC (Cont'd) or 015 in the SLC Pump Suction Line - Fails Open Main Steam MSIVs Fail to Remain Open or Operator Table C2-1. Item 1 - Isolation Fails to Re-Open Valves. Valves (MSIVs) Significant' Failure to initiate Standby Liquid Table D2, Item 1 Control (SLC) when required (e.g. 1 Human Errors during Anticipated Transient Without- , i
' Scram - ATWS)
Failure to manually initiate the ADS Table D2,- Item 3 if required and automatic initiation f ail's Failure to inhibit ADS during ATWS Table D2, Item 4 1 scenario Human errors during normal operation Table D2, Item 5 ; leading to feedwater system trips
- Failure to recover of fsite power Table D2, Item 6 L
Reactor . Failure of Both' Trains of RVSVS and Table D4-1, Item 2a Building ' CRAC CWS: Operator Shuts off the , Standby Operating Train Without Racognizing that Ventilation the t.1 ternate System is Inef fective System (RBSYS)
- Failure'of Both Trains of RVSVS and Table D4-1, Item 2b and CRAC CWS: Unit Coolers 2A and 3A (or 2B and 3B) Fail Because Normally LControl Room. Closed PCV019A (PCV019B) Fails Open so ,
Air, Condition- that the Header Crossover Bypass'Line is
' ing - ( CRAC) . Inadvertently Open Chilled Water 4
-System Failure-of.Both Trains of RVSVS and Table D4-1, Item 2c CRAC CWS: Chilled Water Supply Header Blocked Because Normally Open Manual Valve 028A (0288) Fails Closed Failure of Both Trains of RVSVS and Table D4-1, Item 2d CRAC CWS Chilled Water Return Header Blocked Because Normally Open Manual Valve 027A (0275) Fails Closed Failure of Both Trains of RVSVS and Table D4-1, Item 2e CRAC CWS: System A (System B) Pipe Failure i'
ma _:__-__________ .. -
{ .
'-127-t -,
Table'E-1'(Cont'd): SYSTEM- FAILURE DISCUSSION' li Control Room: SWS Train-A (or Train B)-in Maintenance Table.D4-1, Item 3
.- Air Condition- and Train B (Train A) of RBSVS and
-ing (CRAC). CRAC CWS Fails-H. Chilled Water- '
Systoo (Cont'd)
.7 1
1: E l ( ,. ** I g '. i 4
-s,-
l
-128- , ,
. SHORERAM NUCLEAR POWER STATION l PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLANS TABLE- E2 PERIODIC SURVEILLANCE AND CALIBRATION INSPECTION GUIDANCE ]
l The listed components are the risk significant components for which proper surveillance and/or calibration should minimize failure. -l l SYSTEM FAILURE DISCUSSION Reactor Review and witness RPS function Table Al-1, Item 1 Protection- surveillance tests and preventive System (RPS) . maintenance; include witness of partial-manual scram test, single rod scram, ' tests of individual RPS channels, and RPS circuit breaker and motor generator set preventive maintenance. Review calibration records RPS sensors Table Al-1, Item 6. and compare results to Shoreham- ; technical specifications. Observe trends. Review qualifications and training for Table Al-1, Item 7 technicians performing testing and/or maintenance on the system. Review surveillance and maintenance of Table Al-1, Item 9 ARI instruments. - Review preventive maintenance practices Table Al-1, Item 10 i for solenoid operated valves located in o the instrument air- header and at the HCU scram inlet and outlet valves. , Automatic Failure of All ADS Valves or Valve Table A2-1, Item 1 ll Depressuriza- Operators Due to an Adverse Environment tion System Inside Containment Af fecting
.(ADS) Cable /Solanoid Coils Miscalibration of Reactor Water Level Table A2-1, Item 2 Channels or Transient Disables Level Instrumentation, and Operator Fails to ..
Initiate ADS Nitrogen Supply Contaminated and Table A2-1 Item 3 , Solenoids Disabled Operator Error During Maintenance Table A2-1, Item 4 Disabled Valves or Solenoids I l~ Common Mode Check Valve Leakage Bleeds Table A2-1, Item 5 off Accumulator Pressure and Failure of Backup Nitrogen Supply or Loss of Backup Supply Due to Containment Isolation l 1e i . e-~ _ _ _ , _ -= __ - :-
-129-Table E-2 (Cont'd) y SYSTEM FAILURE- DISCUSSION Automatic ; Common Mode Miscalibration or Transient Table A2-1. Item- 6 Depressuriza- Disables. Pressure Switches on RHR Pumps ,
tion System A, B, C, and D thereby Inhibiting ADS (ADS) from Actuation Because RHR Pumps Are (Cont'd) Not Available-Failure of Reactor Level Channels A and Table A2-1, Item 8a B and Operator Fails to Initiate ADS Given Auto System Failure Sensor Failures Failure' of Reactor Level Channels A and Table A2-1, Item 8b B and Operator Fails to Initiate ADS Given Auto System Failures Logie Roset Switch in Reset Position Preventing Automatic Initiation of an ADS Signal Residual Heat Miscalibration of Low Reactor Pressure Table A3-1, Item 1 Removal System Permissive Channels (RHRS) Suppression Pool Water Unavailable Due Table A3-1, Item 2 And to Clogged Strainers Low Pressure LPCI/CS Manually Shut Of f on High Level Table A3-1. Item 4 Coolant During an Accident and Level Inj ection Instrumentation Fails to Automatically
'JF (LPCI) System Initiate Af ter Having Worked at High Pressure Failure of All Four LPCI Pumps A,B,C,D Table A3-1, Item 7 to Start and Run Failure of Flow from Both LPCI Loops A Table A3-1 Item 8a and B to Enter the Reactor Vessel
- Normally Closed MOV F015A or B at Primary Containment Drywell Penetration Fails Closed Failurelof Flow from Both LPCI Loops A Table A3-1, Item 8b and B to Enter the Reactor Vessel:
Normally Locked Open NOV F017A or B at Primary Containment Drywell Penetration Fails Closed Failure of Flow from Both LPCI Loops A Table A3-1. Item 8e and B to Enter the Reactor Vessel: Air Operated Check Valve 050A Fails Stuck Closed
.c
-130- .-
l 1 Table E-2 (Cont'd) SYSTEM FAILURE DISCUSSION ! Low Pressure Failure of Flow from Both LPCI Loops A Table A3-1, Item 8d j Coolant- and B to Enter the Reactor Vessels Inj ection Normally Locked Open Manual Valve F060A * (LPC1) System or'B Fails Closed - (Cont'd) LPCI Loop 1 in _ Maintenance and Failure Table A3-1, Item 9 of Pumps B and D or LPCI Loop 2 in Maintenance and Failure of Pumps A and C , Both RHR Heat Exchanger Bypass MOVs Table A3-1, Item 10 F048A and F048B Fail Open-Both Service Water System (SWS) MOVs Table A3-1,. Item 11 34A and 34B on the RRR Heat Exchangers' - SW Outlet Lines Fail Closed Service Water All Four SWS Pumps Tail at the Same Table A4-1, Item 2 Systems (SWS) Time i One Loop in Maintenance with Two Pump Table A4-1 Item 5 Failures in the Opposite Loop Emergency All Diesel Generators Require Table AS-1, Case 1 Electric Power Maintenance (Shutdown in 12 Hours) and Item 1 System (EEPS) Common Mode Failure of All Of fsite Power Sources (LOOP) and Direct Power Source (125 VDC System) is Not Repaired Within 2 Hours of Failure Voltage Transformer Failure Table A5-1,: Case 2 Item 1
- - Normally Closed 4 kV to 480 V Bus Table A5-1, Case 2, Transformer Breaker Fails Open Item 2 480 V Bus or Switchgear Unavailable Table A5-1, Case 2, Item 3 4 kV Bus or Switchgear Unavailable Table A5-1, Case 2, Item 4 4
Common Mode Loss of Of fsite Power and Table AS-1, Case 2, One Diesel Generator Fails to Start Item 5 and'Run.
, 'l**%- - -_-_--------------_-.--------m- - - - - _ - - - _ _ _ - - - - - - -
4
,. ; :.- . 131 Table E-2-(Cont'd)
SYSTEM FAILURE DISCUSSION t High Pressure Failure of HPCI Pump or Booster Pump Table B1-1, Item 2
. Coolant- or Turbine to Start and Run Inj ection l (RPCI) System Normally Closed Steam Supply' Isolation Table 31-1, Item 3 MOV F001 or F003 Fails Closed Normally Closed'HPCI Turbine Stop Valve Table B1-1, Item 4 F051 or Control Valve F052-Fails Closed i
Normally Closed HPCI Pump Discharge Table B1-1. Item 5 Isolation MOV F006 Fails Closed HPCI Lube Oil Cooler Shutoff MOV F059 Table B1-1, Item 6 i Fails Closed Auxiliary 011 Pump Fails to Start Table B1-1, Item 7 and Run False Signal Indicates High Area Table B1-1, Item 8 'I Temperature i
. 1 Miscalibration of High Level Sensors in Table B1-1, Item 10 -l Suppression Pool or Low Level Sensors in Condensate Storage Tank (CST)
Common Mode Miscalibration of Turbine Table B1-1, Item 11
'# Exhaust Diaphraga Pressure Sensors or Steam Supply Line Pressure Sensors Miscalibration of Turbine Trip Pressure Table B1-1, Item 12 or Level Sensors
, HPCI Turbine Fails to Start (During Table B1-1, Item 13 Subsequent Re-Starts)
. Normally Closed HPCI Discharge Table B1-1, Item 14 Isolation MOV F006 or Suppression Pool Isolation MOV F042 Fails Closed (During Subsequent Re-Starts)
.. Normally Closed Turbine Stop Valve or Table B1-1, Item 15 Control Valve Fails Closed (During Subsequent Re-Starts)
Failure of HPCI Pump or Booster Pump to Table B1-1, Item 16 Start and Run (During Subsequent Re-Starts) Palse Signal Indicates Righ Steam Line Table B1-1, Item 17 AP,' Sensors N004 or N005 11 i
-132- , ... j i
I Table E-2 (Cont'd). SYSTEM FAILURE DISCUSSION High Pressure False Signal Indicates High Turbine Table B1-1, Item 18 i Coolant Exhaust Pressure Sensor N017A or N017B j
. Inj e ction (HPCI) System False Signal' Indicates Low Pressure on Table B1-1 Item 19 (Cont'd) Pump Suction (No. 10) -
1 Failure of Relay Logic for Suppression Table B1-1. Item 20 ; Pool or Condensate Storage Tank Level
- Indication (During Subsequent Operation)
Instability in Steam Exhaust Line Causes Table B1-1, Item 22 i High Exhaust Pressure During Initial Start-up or Subsequent Start-up Reactor Core Failure of RCIC Pump or Turbine to Table B2-1 Item 2 , Isolation Start or Run Cooling (RCIC) System RCIC Pump Discharge Isolation MOV F013 Table B2-1 Item 3 Normally Closed Fails Closed ,, Motor-Operated Gate Valve F008 or F045 ' Table B2-1, Item 4 l in Steam Line to RCIC Turbine Normally Closed Fails Closed Equipment Room ( Area) Righ Temperature Table B2-1. Item 5 Sensor 600A, 600B, 601A, 601B, 602B, ' 603A, 603B, 604A, or 604B Gives False
-Signal Miscalibration of RCIC Turbine Trip. Table B2-1, Item 6 Level and Pressure Sensors Causes False Turbine Trip RCIC Pump or Turbine Fails to Start on Table B2-1, Item 7-Subsequent Starts Sensors N006 for Low RCIC Pump Suction Table B2-1, Item 8 Pressure Gives False Signal for Turbine Trip ;
Sensor N017 or N018 for RCIC Turbine Table B2-1 Item 9 . Steamline High P Gives False Signal or Instrument Line Break Causes Turbine Trip RCIC Turbine High Exhaust Pressure Table B2-1 Item 10 Sensors N009A or N009B Give False Signal for. Turbine Trip RCIC Turbine Lubrication System Fails Table B2-1, Item 11
-133-Table E-2 (Cont'd)
SYSTEM FAILURE- DISCUSSION Reactor Core. Instability in Steam Exhaust Line Table B2 l, Item 12 Isolation Causes High RCIC Turbine Exhaust Pressure Cocy ng (RCIC) i syd ca- Automatic HPCI Shutoff Fails and Table B2-1, Item 13 l (Cont'd) Operator Fails to Shut Off HPCI ! Core Spray (CS) Failure of Differential Fressure Sensor Table B3-1, Item 1 System 005 Due to Miscalibration or Common Mode Failure of Channel
'I1 Common Mode Failure of Both Core Spray Table B3-1, Item 2 Pumpw i
Suppression Pool Water Unavailable Due Tabis B3-1. Item 3 to Clogged Strainer Failure of Both Core Spray Loop A and Table B3-1, Item 5a Loop B: Normally Closed Discharge ' Isolation Valve F005A or F005B Fails ; Closed l l Failure of Both Core Spray Loop A and Table B3-1, Item 5b Loop B Pump A or Pump B Fails to Start and Run g p Failure of Both Core Spray Loop A and Table 53-1, Item Se Loop B: Failure of Differential Pressure-Sensor to Open Pump Discharge Valve. .; Failure of Both Core Spray Loop A and Table B3-1 Item 5d Loop B: Failure of Differential Pressure Relay Logic Failure of Both Core Spray Loop A and Table B3-1. Item Se- 1 Loop B: Loss of Pump Lubrication Failure of Both Core Spray Loop A and Table B3-1, Item 5f Loop B: Failure of Strainer Blocks Flow Failure of Both Core Spray Loop A and Table B3-1, Item Sg
- Loop B: Flow Diverted to Suppression Pool Via Normally Closed Test Line MOV F015A or F015B Failing Open Failure of Both Core Spray Loop A and Table B3-1, Item Sh Loop B: Normally Open Suction Line MOV F00lA or F001B Fails Closed 1
-134- . .
Table E-2 (Cont'd) SYSTEM FAILURE DISCUS $10N Core Spray (CS) Failure of Both Core Spray Loop A and Table B3-1, item 51 System Loop B Discharge Line Check Valve (Cont'd) F006A or F0065 Fails Stuck Closed (Inside Drywell) . Failure of Both Core Spray Loop A and Table B3-1, item 5j Loop B Discharge Line Check Valve F003A or F003B Fails Stuck Closed , Failure of Both Core Spray Loop A and Table B3-1, item $k Loop B: Locked Open Discharge Line Manual Isolation Valve F007A or F007B Fails Closed (Inside Drywell) Reactor Vessel Level Instrumentation Table B3-1, item 7a Fails Af ter Having Worked at High Pressures, and LPC1/CS Manually Shut Off on High Level During an Accident Reactor Vessel Level Instrumentation Table B3-1. Item 7b Fails Af ter having Worked at High Pressures, and Operators Totally Distracted Over a Period of 1-6 Hours and fail to Restart LPC1/CS When Required Reactor Vessel Level Instrumentation Table B3-1, item 7e Fails Af ter Having L'orked at High Pressures, and Operators Fail to Follow Procedures After Recognizing the Problem and Fail to Restart LPC1/CS When Required Cos=on Mode Failure of Both Loops Table B3-1, item 8 C&aded by Manual Valves 7007A and 7007B Being Left in the Closed Position Reactor Failure of one Loop and Maintenance on Table B4-1, item 3a Building the Other Loop; Failure Caused by False Closed Loop Indication of Suction Pressure Stoping Cooling Water Pump System (RBCLCWS) Failure of One Loop and Maintenance on Table B4-1. Item 3b , the Other Loop; Failure Caused by Pump Failure to Oontinue Running Failure of one Loop and Maintenance on Table B4-1, item 3c the Other Loop Failure Caused by Clogged Pump Filter l
= **
' i 135 l Table E-2 (Cont'd)
SYSTEM FAILURE DISCUSSION i Reactor Failure of One Loop and Maintenance on Table B4-1 Item 3d Building the Other Loopt failure Caused by HX Closed Loop Inlet (MOV-042A,5) or Outlet Valve l Cooling Water (MOV-052A,5) Failing Closed l System ; (RBCLCWS) Failure of One Loop and Maintenance on Table B4-1, Item 3e (Cont'd) the Other Loop; Failure Caused by
. Temperature Control Valve (TCV-00lW/ X, ,
00lY/Z) Failing Closed ' Failure of One Loop and Maintenance on Table B4-1, Item 3f the Other Loop: Failure Caused by Normally Open SWS Motor Operated Valve - Failing Closed ; Failure of One Loop and Maintenance on Table B4-1, item 3g the Other Loop: Failure Caused by . Normally Open Manual Valve (16V-009A.2) in SWS Injection Line Failing Closed Failure of One Loop and Maintenance on Table B4-1, Item 3h : the other Loop; Failure caused by HX Ruptures or Excessive Leakage
- Failure of One Loop and Maintenance on Table B4-1. Item 3j p the Other Loop Failure Caused by Pump Discharge Check valve (00$A,5,C) Failing ;
Closed , Failure of One Loop and Maintenance on Table B4-1, Item 3k the Other Loop; Failure Caused by Fipe Rupture or Excessive Leakage in Fump Suction Lines, Pump Connections or Discharge Lines Failure of One Loop and Maintenance on Table B4-1. Item 31 the Other Loop: Failure Caused by Faulty Motor Overload Breaker Stoping Pump I Failure of Both RBCLCW Loops Table B4-1. Item 4 l Maintenance of one SWS Loop and Failure Table B4-1. Item 5 of the Opposite RBCLCW Loop l Reactor Common Mode Miscalibration Table B5-1, Item 1 Water Level Instrument Line Break Table 35-1. Item 2 l Electronics Failure Table B5-1. Item 4 l
- i. ,
i -136- e - l Table E-2 (Cont'd) SYSTEM FAILURE DISCUSSION
' Feedwater, Miscalibration of Sensors Causes Talse Table B6-1, Condensate. TW Systes Trip Item A.2 and Power Conversion F9 Systes Trips Due to Failure in TW Table B6-1, Systems Control Logic, Failure of TW Turbine Ites A.4 (FW/CD/PCS) Control Causing Loss of R.P. Steam, Spurious MSIV Closure, or Genuine Isolation Signal Causing MSIV Closure Steen Jet Air Ejectors (SJ AE) Legs Table B6-1, Unavailable: Test or Maintenance of SJAE Ites A.6.a.iii.
Mechanical vacuun Pump Unavailable: Table B6-1 Pump Les in Test or Maintenance Iten A.6.b.ii Mech'anical vacuum Pump Unavailable: Table B6-1, Pump Fails to Start and Run ,Ites A.6.b.iii Condensate System: Flow Control Table B6-1, Instruments Fail to Supply Signal or Item B.3 Supply False Signal to Trains A and B Condensate System: Rupture of Piping or Table B6-1 Meat Exchangers item B.A. Condensate Systems Rupture of Condenser Table B6-1 Hotwell Ites B.7.
-Condensate System: Failure of Drain Table B6-1, Coolers in FW Heater Legs A and B Item B.B.
Power Conversion Systems Loss of vacuus Table B6-1 Ites C.2. Standby Liquid Insufficient Boron Concentration in Table Cl-1, item 3 Control (SLC) the SLC Storage Tank System Suction Line From SLC Storage Tank Table Cl-1. Item 4 Plugged Check Valve 006 or 007 Fails - Stuck Table Cl-1. Item 5 ' Closed Normally Locked Open Manual Valve 008 Table Cl-1, Item 6
) in Common Pump Discharge Line to the RPV Fails Closed Normally Locked Open Manual valve 001 Table Cl-1. Item 7 in the Supply Line From the SLC Storage Tank Fails Closed 4
eqpe- - +w g--
.~. __ . _ _ -_
e '
. -137-Table E-2 (Cont'd) !
SYSTEM FAILURE DISCUSSION Standby Liquid Normally Locked Closed Manual valve 014 Table Cl-1, Item 8 Control (SLC) or 015 in the SLC Pump Suction Line System Fails Open 1 (Cont'd) SLC Tank Fails During ATWS Demand Table Cl-1. Item 9 Main Steam MSIVs Tail to Remain Open or Operator Table C2-1, Item 1 Isolation Fails to Re-Open Valves Valves (MSIVs) > MSIVs Tati to Close or Leak Excessively Table C2-1 Item 2 Significant Miscalibrations of t Reactor Vessel Level, Table D2, Item 2 Human Errors Pump Discharge Valve P Instruments for Core Spray and Low Pressure Coolant Injection (LPCI) Failure to properly restore components Table D2, Item 7 after test or saintenance. Reactor Tailure of Both Trains of RVSVS and Table D4-1 Building CRAC CWSt Unit Coolers 2A and 3A (or Item 2.b Standby 2B and 38) Fail Because Normally Closed Ventilation PCV019A (PCV0198) Fails Open so that System (RBSYS) the Header Crossover Bypass Line is Inadvertently Open . And JF Failure of Both Trains of RVSVS and Table D4-1, Control Room CRAC CWS t Chilled Water Supply Header Item 2.c Air Condition- Blocked Because Normally Open Manual ing (CRAC) Valve 028A (0288) Fails Closed Chilled Water System Failure of loth Trains of RVSVS and Table D4-1, CRAC CWSt Chilled Water Return Header Ites 2.d Blocked Because Normally Open Manual valve 027A (0278) Fails Closed Failure of Both Trains of RYSYS and Table D4-1, CRAC CWS System A (System B) Fipe Item 2.e Failure 4 SWS Train A (or Train B) in Maintenance Table D4-1, Item 3 and Train B (Train A) of RBSYS and CRAC CWS Fails ! Miscalibration or Common Mode Failure Table D4-1. Item 4 of Reactor Water Leval Sensors and Miscalibration of Drywell High Pressure Sen' sors l .
136- , , , SHOREMAN NUCLEAR POWER STATION PROBABILISTIC RISK ASSESSMENT-BASED INSPECTION PLANS TABLE E3 - MAINTENANCE INSPECTION CUIDANCE The components listed here are significant to risk because of unavail-ability for maintenance. The dominant contributors are usually frequency and duration of maintenance, with some contribution due to improperly performed , maintenance. SYSTEM FAILURE DISCUSSION Reactor Review and witness RPS function Table Al-1, Iten 1 Protection surveillance tests and preventive System (RPS) maintenance; include witness of partial manual scras test, single rod scras, tests of individual RPS channels, and
~ RPS circuit breaker and motor generator set traventive maintenance.
Review post work testing of RPS Table Al-1, Item 5 maintenance tasks. Review calibration records RPS sensors Table Al-1, Item 6 and compare results to Shorehan technical specifications. ' Observe trends. Review qualifications and training for Table Al-1, Item 7 technicians performing testing and/or maintenance on the system. Review control rod drive mechanism Table Al-1, Item 8 maintenance inspection procedure and results. Insure trending of detected wear is performed. Review surveillance and maintenance of Table Al-1. Iten 9 ARI instruments. Review preventive maintenance practices Table Al-1, Iten 10 for solenoid operated valves located in the instrument air header and at the HCU scram inlet and outlet valves. e Automatic Operator Error During Maintenance Table A2-1, Item 4 Depressuriza- Disabled Valves or Solenoids tion System h ( ADS) 4 d
. n.-..
* * * .g39 j l
Table E-3 (Cont'd) SYSTEM FAILURE DISCUSSION I 1 Residual Heat Suppression Pool Water Unavailable Due Table A3-1. Item 2 l Removal System to Clogged Strainers ) Both LPCI Loops Require Maintenance Table A3-1. Item 3 And Failure of All Four LPCI Pumps A,3 C.D Table A3-1. Item 7 Low Pressure to Start and Run Coolant ' Inj ection LPCI Loop 1 in Maintenance and Fai1Gre Table A3-1. Item 9 (LPCI) System of Pumps B and D or LPCI Loop 2 in ! Maintenance and Failure of Pumps A and C. j Both RHR Heat Exchanger Bypass NOVs Table A3-1, Item 10 F048A and F048B Fail Open noth Service Water System (SWS) MOVs Table A3-1 Item 11 34A and 34B on the RHR Reat Exchangers' ' SW Outlet Lines Fail Closed Service Water Both Service Water Loops Require Table A4-1. Item 1 Systema (SWS) Maintenance (Reactor Shutdown in 12 Hours) Loss of Water Supply to Screen Well Table A4-1. Item 3 (Screen Well Water Level Low) , jy- One Loop in Maintenance with Two Pump Table A4-1, Item 5 Failures in the Opposite Loop l Energency All Diesel Generators Raquire Table A5-1 Case 1 Electric Maintenance (Shutdown in 12 Bours) and Ites 1 l Power System Common Mode Failure of All Of fsite Power (EEPS) Sources (LOOP) and Direct Power Source (125 VDC System) is Not Repaired Within 2 Hours of Failure r I Voltage Transformer Failure Table A5-1. Case 2 Item 1 Normally Closed 4 kV to 480 V Bus Table AS-1, Case 2, Transformer Breaker Fails Open Item 2 6 480 V Bus or Switchgear Unavailable Table A5-1, Case 2
, Item 3 4 kV Bus or Switchgear Unavailable Table AS-1, Case 2 Ites 4 Common Mode Loss of Offsite Power and Table A5-1, Case 2, One Diesel Generator Fails to Start and Item 5 Run.
I f _ _ .. - -.
-140- , . ,
Table E-3 (Cont'd) SYSTEM FAILURE DISCUSSION j 1 High Pressure HPCI System in Maintenance Table B1-1, Item 1 , ; Coolant i Inj ection Failure of HPCI Pump or Booster Pump Table B1-1, Item 2 l (HPCI) System or Turbine to Start and Run i t 1 Normally Closed Steam Supply Isolation Table B1-1, Item 3 i MOV F001 or F003 Fails Closed
~
Normally Closed HPCI Turbine Stop Valve Table B1-1, Item 4 F051 or Control Valve F052 Fails Closed j Normally Closed HPCI Pump Discharge Table B1-1. Item 5 I Isolation MOV F006 Fails Closed ' l HPCI Lube 011 Cooler Shutoff MOV F059 Table B1-1, Item 6 Fails Closed HPCI Turbine Fails to Start (During Table B1-1, Item 13 Subsequent Re-Starts) 1 l Normally Closed Turbine Stop Valve or Table B1-1, Item 15 Control Valve Fails Closed (During Subsequent Re-Starts) Failure of HPCI Pump or Booster Pump Table B1-1, Item 16 to Start and Run (During Subsequent i Re-Starts) i Reactor Core RCIC in Maintenance Given that MPCI is Table B2-1. Item 1 Isolation Available and Not in Maintenance 3 Cooling (RCIC) System RCIC Pump Discharge Isolation MOV F013 Table B2-1 Item 3 Normally Closed Fails Closed Motor-operated Gate Valve F008 or F045 Table B2-1. Item 4 < in Steam Line to RCIC Turbine Normally Closed Fails Closed RCIC Turbine Lubrication Systen Fails Table B2-1, Item 11 l Core Spray Suppression Pool Water Unavailable Due Table B3-1, Item 3 < (CS) System to Clogged Strainer Both Core Spray Pumps ' Require Table B3-1, Item 4 Maintenance - Reactor Shutdown Within 12 Hours FaiiureofBothCoreSprayLoopAand Table 23-1, Item Se Loop Bt Loss of Pump Lubrication
, _ . _ . _. . _ . . _ _ . _ , . , _ -, __ _ _ _ _ . _ . _ _ _ _ . _ _ .- _ _m__ _ _- ___ -
c '. s ' o
-141- i Table E-3 (Cont'd) i SYSTEM FAILURE DISCUSSION Core Spray Common Mode Failure of Both Loops Table B3-1, Item 8 (CS) System Caused by Manual Valves F007A and F007B (Cont'd) Being Left in the Closed Position Reactor Both RBCLCW Loops Require Maintenance - Table B4-1, Item 2 Building Reactor Shutdown Within 12 Hours r Closed Loop . Cooling Water Failure of One Loop and Maintenance on Table B4-1 Item 3d System the Other Loopt Failure Caused by HX (RBCLCWS) Inlet (MOV-042A,B) or Outlet Valve (MOV-052A,B) Failing Closed ;
Failure of one Loop and Maintenance on Table B4-1, Item 3e the Other Loopt Failure Caused by Temperature Control Valve (TCV-00lW/ X, 00lY/Z) Failing Closed Failure of One Loop and Maintenance on Table B4-1, Item 3f ' the other Loopt Failure Caused by Normally Open SWS Motor Operated Valve Failing Closed , Failure of One Loop and Maintenance on Table B4-1, item 3g l the Other Loop Failure Caused by Normally Open Manual Valve (16V-009A,B) l JP' in SWS Injection Line Failing Closed Failure of Both RBCLCW Loops Table B4-1, Item 4 Maintenance of One SWS Loop and Table B4-1, Item 5 Failure of the Opposite RBCLCW Loop j Maintenance of One SWS Loop and Table B4-1, Item 6 Maintenance of the Opposite RBCLCW Loop Reactor Instrument Line Break Table B5-1 Item 2 Water Level Feedwater, , Steam Jet Air Ejectors (SJAE) Legs Table B6-1, Condensate, Unavailable Loss of Main Air Ejectors Item A.6.a.1 ,
- and Power Conversion Steam Jet Air Ejectors (SJ AE) Legs Table B6-1, Systems Unevailablet Test or Maintenance of Iten A.6.a.ii (FW/CD/ PCS) SJAE Steam Jet Air Ejectors (SJAE) Legs Table B6-1, Unavailablet Off Gas System Item A.6.a.iii ,
Failure / Trips ,
i
-142- i Table E-3 (Cont'd) ;
SYSTEM FAILURE DISCUSSION Feedwater, Mechanical Vacuum Pump Unavailable: Table 86-1, ; Condensate, Pump Leg in Test or Maintenance Item A.6.b.it and Fower Conversion Condensate or Condensate Booster Pumps Table B6-1, . Systems A and B Fail to Continue Running Item B.5. . ! (FW/CD/PCS) (Cont'd) Condensate Domineraliser System (CDS) Table B6-1, ! Failure Item B.6. j Failure of Drain Coolers in FW Heater Table B6-1, Legs A and B Item B.8. , Standby Liquid Normally Locked open Manual Valve 008 Table Cl-1, Item 6 f Control (SLC) in Common Pump Discharge Line to the RFV i System Fails Closed l i Normally Locked Op3n Manual Valve 001 Table Cl-1, Item 7 ' in the Supply Line From the SLC Storage j i Tank Fails Closed : f . Normally Locked Closed Manual Valve 014 Table Cl-1. Item 8 or 015 in the SLC Pump Suction Line , Fails Open Main Steam MSIVs Fail to Remain Open or Operator Table C2-1. Item 1 [' Isolation Fails to Re-Open Valves Valves.(MSIVs) ! MSIVs Tail to Close or Leak Table C2-1, Item 2 , Excessively : Significant Failure to properly restore components Table D2, Item 7 , Human Errors after test or maintenance ' O e 9 e 4 e _. . . , . _ , .w -- ,}}