ML19354D797
| ML19354D797 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/31/1989 |
| From: | Fresco A, Shier W BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML19354D796 | List: |
| References | |
| TR-A-3875-T2C, TR-A-3875-T2C-R, TR-A-3875-T2C-R00, NUDOCS 9001220094 | |
| Download: ML19354D797 (156) | |
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Technical Report 4 TR A 3875 T2C Rev. 0
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WOLF CREEK GENERATING STATION ;
RISK BASED INSPECTION GUIDE (RIG) '
(BASED ON GENERIC INSIGHTS FROM PROBABILISTIC RISK
- ASSESSMENTS FOR WESTINGHOUSE PWRS)
December 1989 i ,
Prepared by:
- A. Fresco and W. Shier I
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Department of Nuclear Energy i Brookhaven National Laboratory l Upton, New York 11973 Prepared for: I U.S. Nuclear Regulatory Commission Washington, DC 20555
- l. FIN A 3875 !
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. -9001220094.900109 PDR ADOCK 05000482 R ___ iPCW - _ _ _ _ _ _ _ _ _ _ _ . --. .-. -_ . -
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l TABLE OF CONTENTS !
L Section Title Page
- 1. INTR O D U CTI ON . ..... .. ..... . . . .. .... . ....... . ... ..... . . ... 1
- 2. 0ENERIC DOMIN ANT ACCIDENT SEQUENCES .... .. ..... ... . ...... ... . I 2.1 LOCAs............................................................... 1 2.1.1 Small or Intermediate LOCAs (Sequence 1).... . .... ...... ..... ... ..... 1 2.1.2 Medium or Large LOCAs (Sequences 2.3) . .. .. .. ... . ..... ..... ..... 2 2.1.3 LOCA's Outside ConLainment (Sequence 4) .... .. . .. ... ........... 3 2.2 Tr an s le nt S e q u e n e e s ........ . . ... ... . ..... ..... ... . ....... . ............ ............ 5 2.2.1 Loss of All Component or Closed Cooling Water (Sequence 3)..... 5 2.2.2 Loss of DC Power (One 125V DC Bus) (Sequence 6)... ........ 5 2.2.3 Loss of Offsite Power / Station Blackout Initiators..... 6 2.2.4 Loss of Power Conversion System (PCS) or Transient Followed by Loss of PCS, with Loss of Decay Heat Removal (Sequw 10).... 8 2.3 Anticipated Transient Without Scram (ATWS) Followed by Failure of Emer.
g e nc y B oration .. ... . .. ... . . . ...... . . . . .. ..... 9 ,
! 3. CO M M ON C AUS E FAI LURE S . . ..... . ..... ... ..... . ....... . . . ...... ... ..... 9 4 IM PORT ANT HUM AN ER R O R S .. . ..... . . .. ... ...... ... . ............ .... ... 10 l
- 5. SYSTEMS INCLUDED IN GUIDE......... . .. .. . . . . . . . . . . . . . . 10 1
- 6. SYSTEM 1NSPECTION TABLES ... .. . . ... . . I1
- 7. RE FE REN C E S . ... . . ........ . . ... .. . ..... ... .... ............. . .. = . 13 7.1 Oencric Risk Based Information . . .. ...... .. . .. = ...................... 13 7.2 Other
References:
Plant Specific Risk. Based information ..... ........ ............ 14 APPEND 1X A. TABLES OF (1)IMPORTANCE B ASIS AND FAILURE MODE IDENTIFICATION.
AND (2) MODIFIED SYSTEM WALKDOWNS l A.I.l. Importance Basis and Failure Mode Identification
! Essential Service Water System (ESWS).... .... .. . . ..... A.1 A.I.2, Modified System Walkdown Essential Service Water System (ESWS)..... . . .. . ..... . .. . .......... A.5 l A.21. Importance Basis and Failure Mode Identification Safeguards (AC) Power System. .... . . . ..... . . . A.19 A.2 2. Modified System Walkdown S afegu atds ( AC) Power System .. . ... .. . . ..... ..... .. .... .. .. .. A 21 A.31. Importance Basis and Failure Mode Identification DC Power S ystem . ..... .... ....... . -. . . . . . . . . . . . . . . . . . A.25 11
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9 i TABLE OF CONTENTS (Cont'd)
. Section Title Page A.3 2. Modified System Walkdown DC Pow er S yste m . . _ .. . .. . . . . .. ....... .. ....... A 27 A.41. Importance Basis and Failure Mode Identification Component Cooling Water System. . . .. .. ... .. . . ... A.31 A.4 2. Modified System Walkdown Component Cooling Water System (CCWS). . . . ... . ... A.34 A.$.1. Importance Basis and Failure Mode Identification Reaetor Protection System (RPS) . .. . .. . .... . . . . .. . ....... A.46 A.$.2. Modified System Walkdown Reactor Protection System (RPS) ... . . . ...... .. ... .. . . ... . A.48 A.61. Importance Basis and Failure Mode Identification High Head injection / Safety injection /High Head Recirculation. A.50 A.6 2 Modified System Walkdown High Head Injection / Safety injection /High Head Recirculation...... A 54 A.71. Importance Basis and Failure Mode Identification Primary Pressure Relief S ystem . .. . ... . ..... ... .... . ... ....... ..... A.68 A.7 2. Modified System Walkdown Prim try Pressure Relie f S ystem . ... .. . ..... . . .. ...... .. . . ... ... ..... A.70 A.81. Importance Basis and Failure Mode Identification Auxiliary Feed water S ystem . ...... . ... .......... ... ...... .... .- A.74 A.8 2. Modified System Walkdown Auxillary Fecdwater S yste m . ... .... ... . . .... .. .. .... .... .. A 78 A 91. Importance Basis and Failure Mode Identification Low Head injection (LHI)/ Low Head Recirculation (LHR)..... ... ........ A.92 A.9 2, Modified System Walkdown Low Head Injection / Low Head Recirculation System: .................. A.96 A.101.lmportance Basis and Failure Mode Identification Engineered Safety Features Actuation System (ESFAS)= . . . . . . . . . . . .. A.101 A.10 2. Modified System Walkdown Engineered Safety Features Actuation System (ESFAS).... ...... ... .... ..... A.103 A.ll.l.importance Basis and Failure Mode Identification Refueling Water Storage Tank (RWST) .. .. .... . ... ..... . . .... A.104 A ll.2. Modified System Walkdown Refueling Water Storage Tank (RWST) . .. .. .. ........ .......... .-... A.105 A.121.lmportance Basis and Failure Mode Identification Power Conyersion S ystem (PCS) . ... .. .. ... ....... . ...... ......... A.108 A.12 2. Modified System Walkdown Power Conyersion System (PCS) . . . . .. - A.109 A.131.importance Basis and Failure Mode Identification Chemical and Volume Control System (CVCS) Emergency Boration. ....... A ll4 A.13 2. Modified System Walkdown Chemical and Volume Control System (CVCS) Emergency Boration... ..... A ll6 APPENDIX B. TABLES OF (1) PLANT OPERATIONS INSPECTION GUIDANCE, (2) SURVEIL.
' LANCE AND CALIBRATION INSPECTION GUIDANCE, AND (3) MAINTE.
NANCE INSPECTION GUIDANCE lii l
TABLE CF CONTENTS (Cont'd)
Section Title Page B.1 Plant Operations Inspcetion Guidance .. .... .... . . . B.1 B.2 Surveillanee and Calibration inspection Guidance..... ... . ..... . B-4 B.3 Maintenance inspection Guldanee.. . .... ... . ... . ....... B.10 APPENDIX C. CONTAINMENT AND DRYWELL WALKDOWN C.1 Con tain m e n t Walkd own . ... ... .. ..... .... ... .. . ... ............ ... C.1 i
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i t WOLF CREEIC GENERATING STATION
- 1. INTRODUCTION This inspection guide has been prepared to provide generic risk based inspection guidance for Westin house PWRs based on review of several Probabilistic Risk Assess-ments (PRAs).7 3 3 7 g3 It is not intended to apply to CE or B&W plants. The guidance should be used to aid in the selection of areas to inspect and is not intended either to replace current NRC inspection guidance or to constitute an additional set of inspection requirements. Recent system experience, failures, and modifications should be considered when reviewing these tables. Since plant modifications are normally an ongoing process,it is recommended that relevant changes be catalogued so that this inspection guidance can be periodically revised as required.
- 2. GENERIC DOMINANT ACCIDENT SEQUENCES Based upon a review of available PRAs for PWRs, eleven representative accident sequences were selected based on their contribution to core damage frequency or because of serious offsite consequences, as shown in Table 1. The details of these sequences are described below.
. 2.1 LOCA 2.1.1 Small or Intermediate LOCAs (Sequence 1)
This accident sequence is initiated by a small ($ s.2 in.) or intermediate (2 in < 0 5 6 in.) LOCA which does not depressurize the Reactor Ceolant System (RCS) below the shutoff head of the low head Emergency Core Cooling ~ System (ECCS). The Reactor Protection System (RPS) successfully scrams the reactor but the high head ECCS fails to provide adequate makeup, either in the injection or recirculation phases, resulting in core damage. The high head ECCS includes the intermediate head Safety Injection System.
Small LOCAs have actually occurred in commercial nuclear power plants and consist primarily of stuck open power operated relief valves (PORVs) and to a lesser degree reacter coolant pump (RCP) seal failures.
Failures during the injection phase are dominated by valve failures in the HH1 common discharge or spetion lines or in the mini flow lines.
l Failures during the high head recirculation (HHR) modes which can occur in the HHR
- ystem, or in any of the support systems required for long term LOCA mitigations, are the dominant contributors to these sequences. The HHR failures are themselves dominated by operator failure to correctly realign the system from the injection mode (for manual L systems) or valve failures in the common discharge or suction lines or the mini flow line i for those configutations with automatic realignment to the HHR mode, such as Wolf Creek, The Westinghouse HHR configuration takes suction from the low head recirculation
, (LHR) pump discharge. LHR malfunctions that disable HHR are the secondary contributor to HHR failures. The primary faults are LHR suction (containment sump) valve and pump malfunctions.
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t 4 MHR room cooling failures are the last major contributor. Those are attributable to electrical component failures that disable room cooler fans or service water valve failures that disables the coolers themselves. Refueling Water Storage Tank (RWST) common mode level sensors miscalibration and Service Water / Component Cooling Water malfunc-tions that disable the HHR pump coolers are less important contributors.
2.1.2 Medium or Large LOCAs (Sequences 2,3)
In accident sequences 2 and 3, an intermediate or large LOCA occurs which rapidly depressurites the reactor coolant system, followed by successful scram of the reactor.
Operation of the Low Head Injection (LHI) system is either successful but followed
- by failure of the LHR system, or the LHI system itself initially fails, either of which leads to core damage. The initiating event is an intermediate or large primary system pressure boundary failure from 6 to 29 in in diameter. No actualindustry failures of this magnitude have occurred.
2.1.2.1 Failures of the Low Head Recirculation (LHR) Mode (Sequence 2)
A major contributor to core damage for this sequence is the failure of the low head ,
ECCS in the recirculation . mode. LHR system failure is evenly divided between human errors and hardware failures. The dominant human error contributor is the failure to initiate LHR by manual realignment of the pump suction from the RWST to the containment sump.
This failure dominates those plants with non.sutomatic pump suction realignment. As noted previously, Wolf Creek has automatic realignment.
Since boron precipitation in the reactor vessel can be minimized or prevented and steam voids in the Reactor Pressure Vessel (RPV) head can be condensed by a backflush of cooling water through the core to reduce boil-off and resulting concentration of botic acid in the water remaining in the reactor vessel, a second operator error is the failure to manually switch the LHR (SI) pump discharge from cold leg to hot leg recirculation after about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident.
Hardware failures are the dominant contributors to LHR system failure for plants such as Wolf Creek with an automatic pump suction changeover feature.1mponant system valve _
malfunctions include failures of LHR containment sump valves to open or RWST suction valves to close, both including common cause failures. The failure of the low head pumps to continue to run (including common cause) is the remaining LHR hardware failure.
The common cause miscalibration of the RWST level sensors is the only major failure not directly associated with the low head ECCS. Other failures are predominantly valve failures such as rupture or failure to open of check valves, valves failing to remain open, service water to RHR heat exchanger valves failing to open, and operator failure to initiate recirculation cooling. Failures of LHR can also occur during the HHR operating mode as previously described in 2.1.1 above.
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s 2.1.2.2 Failures in the Low Head Injection (LHI) Mode (Sequence 3) i A major contributor to core damage for Sequence 3 is the failure to provide short term core injection due to accumulator or low head injection malfunctions. The accumulator failure is attributed to discharge line failures, primarily check valve failures to open or ,
5 MOV plugging. The LH1 system failure is dominated by pump failure to start or run, 1 including common cause. Human error contributors are the failure to restore the system to operable status after testing and the failure to stop the pumps if the miniflow valve fails to open. The dominant system valve failure is failure of the miniflow valve to open. Other failures include injection isolation valves failing to open or to remain open, check valves rupturing or pumps unavailable due to maintenance.
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Table 1 Representathe PWR Accident Sequences Loss of Coolant Accident Sequences F. 1. Small or medium LOCAs with failure of high head injection or recirculation.
- 2. Medium or large LOCAs with failure of low head recirculation.
_ 3. Medium or large LOCA with failure of low head injection.
- 4. LOCA outside containment.*
Transient Sequences E 5. Loss of all CCW initiator with a subsequent RCP seal LOCA.
I 6. Loss of 125V de bus initiator with failure of the auxiliary feedwater system (AFW).
- 8. Station blackout with loss of the AFW system.
Anticipated Transient Without Scram (ATWS) Sequences
- 11. Transient initiator with failure to automatically and manually scram with failure of timely emergency boration.
l "Specified because of serious offsite consequences.
"Specified based on a review of the studies that establishd precursors to potential severe core damage pccidents (NUREG/CR 2497, 3591, 4674).
r 2.1.3 LOCAs Outside Containment (Sequence ,1]
.e The commonly designated V sequence, here called Sequence 4 or LOCA outside F containment,is initiated by a failure of any one of the pairs of series interface check valves
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that isolate the high pressure reactor coolant system (RCS) from the LH1 system. The resultant flow into the LHI system is assumed to rupture the low pressure piping or components outside the containment boundary. Although core inventory makeup by the i high head injection and any available low head injection systems is initially available, the inability to switch to the recirculation mode eventually leads to core damage.
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The discharge of the Wolf Creek LHI System consists of one line from each RHR pump with a normally open MOV. Downstream of the:,e MOVs (toward the RCS) the piping is rated for primary loop conditions. The discharge lines divide to connect to each RCS cold leg. Each of these individual lines has one check valve resulting in two check valves in series. Small leakages through these valves can be accommodated without system overpressure. The failure modes of interest produce sudden, large back leakages through a pair of these interface check valves. The failure is postulated to occur in three ways:
- The dominant initiator mode is the rupture of one check valve with the previously undetected opening of the second valve. If one valve is holding pressure, the other valve can drift open and fall in the open position.
- The second initiator mode is the failure of one check valve to close upon repressurization, followed by a rupture of the second valve.
- The third initiator mode is the random rupture of the valve internals for both check valves. The gross failure of one valve could go undetected until the rupture of the second valve occurs.
The applicability of these initiator types to a specific plant is very dependent on the '
piping configuration of the LHI/RCS interface and the valve testing procedures. For example, Wolf Creek, like some other Westinghouse PWRs, has the accumulator discharge connected between the two interface check valves. This geometry imposes a specific interface valve failure order for the first and third initiator types. If the upstream (furthest from the RCS) valve fails first, the accumulator will discharge into the LHI piping, thereby i alerting the operator.
The applicability of the second initiator type is dependent on the check valve test procedure. Plants that test the interface check valves when the system is depressurized are ,
subject to this initiator type. There is no assurance that both of the valves remain closed on -
subsequent repressurizations. A test procedure that requires valve testing upon every repressurization or use can eliminate the second initiator type from consideration. (The Wolf Creek check valve test procedure should be examined carefully to ensure the potential for a test induced LOCA cutside containment is minimized.)
Based on a review of industry experience, there has been one PWR interfacing systems LOCA precursor which occurred at the Biblis A PWR in the Federal Republic of Germany on December 17,1987.' The NRC is currently in the process of evaluating the significance of this event.
, ' For BWRs, the interface valves generally consist of one normally closed MOV and one check valve. Several BWRs have experienced pressurizations of the low pressure piping, primarily due to testing errors. No gross failures have occurred, due to the lower operating pressure of the BWR RCS.
'M. Hibbs, et al., "NRC Studying implications of Unpublicized German Reactor Incident," inside NRC, 20(25),1, December 5,1988.
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A potential recovery action has been included to account for operator action to isolate
, 'the interfacing LOCA by manual closure of the LHI discharge MOV. The successful mitigation of this event is plant specific and is cependent on:
. LHI pump separation to miniinize the environmental impact of RCS blowdown on
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the second train.
- The existence of two isolatable LHI discharge headers to enable the use of the other -
! LHI loop, or the ability to use another system for RCS makeup.
. The capability of the LHI discharge MOV to isoltde the interfacing LOCA. The
- valve may not be designed to close against sura t high differential pressure, thereby failing the valve in the open position- befrw kfressurization can be effected.
2.2 Transient Sequences '
2.2.1 Loss of All Component or Closed Cooling Water (Sequence 5)
Sequence 5 represents a complete loss of the component or closed cooling water
- l. (CCW) system which precipitates a reactor coolant pump (RCP) seal LOCA and also disables the high and low head ECCS. The inability to provide high pressure makeup L results in core damage.
Lone major contributor to the . loss of CCW initiator is a pipe rupture that drains the l L system inventory before the break can be located and isolated. The second contributor is !
- the common' cause failure of all operating CCW pumps, compounded by a failure of the ;
standby pump (s) to start and run. The RCP seal LOCA and subsequent core damage is postulated to occur before CCW recovery actions can be completed.
2.2.2 Loss of DC Power (One 125V DC Bus) (Sequence 6)
This sequence n :aitiated by a non recoverable loss of a 125V DC bus. The DC power system provides control power to various systems. There hw been several partial losses ,
-of DC power at operating nuclear power plants, approximately one third of which were caused by the misalignment of breakers as part of system maintenance or surveillances. I The remainder of the precursors are due to equipment failures. A loss of one DC bus will typically disable the main feedwater system, a portion of the auxiliary feedwater system, and various DC dependent valves possibly including s PORV. This sequence postulates the failure of the remainder of the AFW system and tbe feed.and bleed mode. Failure of secondary heat removal results in core inventort losses due to PORV cycling and subsequent core damage.
The major contributor to this sequence is the failure of the remainder of the AFW system to supply sufficient flow to the steam generators. This typically involves :be failure of two additional AFW trains.* The major cause is system hardware failure including:
n some plants with two DC busses, to meet electrical separation requirements, the turbine.dsiven pump is supplied power from only one of the busses so that loss of that bus can fail both the turbine driven pump and the respective motcr-driven pump.
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t 4 pump failure to start, and discharge line faults for both the turbine and motor driven trains. .
A secondary contributor is the failure to manually start a pump which is procedurally locked out or unable to start due to a malfunction of the auto start logic.
For those plants with DC controlled PORVs, the sequence success criteria for feed and bleed is plant specific. Some PRAs assume 2 PORVs are required for success (Sequoyah, NUREG/CR-4550). The Zion risk assessment concluded that a single PORV is sufficient for success. For Sequoyah, this initiator by definition fails one valve and climinates the feed and bleed mode. Therefore, the success criteria for feed and bleed should be evaluated on a plant specific basis and the imponance of feed and bleed operation should be judged accordingly/
2.2.3 Loss of Offsite Power / Station Blackout Initiators Loss of offsite power (LOOP) accident sequences are characterized by loss of offsite ,
power followed by at least panial success of onsite emergency AC power sources. In contrast, station blackout sequences are initiated by loss of offsite power followed by total failure of onsite emergency AC power sources.
2.2.3.1 Loss of Offsite Power Initiator (LOOP) with Failure of Auxiliary Feedwater (Sequence 7)
The dominant accident sequence involving LOOP is initiated by a loss of offsite power (LOOP) with successful operation'of at least one source of emergency AC power.
Main feedwater is unavailable due to the initictor. The Auxiliary Feedwater (AFW) system fails due to common mode failures or because of random failures, in concert with the partial system unavailability due to AC power failures. The feed and bleed mode is not
- successful, generally because of system failures. Since secondary heat removal is not available, the resultant boiloff of primary coolant leads to core damage.
The LCOP initiator is one of the more common operating transients, comprising approximately 21% of all precursors to potential core damage. Although some of these initiators are weather er grid related, about 50% of the LOOP precursors are localized failures due to human error such as: maintenance errors on the main generator or switch yard breakers,. breaker misalignment dyring or post maintenance and errors related to manual breaker operation. In addition, seve-M initiators were caused by station transformer faults. The subsequent failure of one or more somces of emergency AC power, usually emergency diesel generators (EDGs) failing to start or run, is imponant because it disables a ponion of the Auxiliary Feedwater (AFW) system. The major contributor to this sequence is the failure of the AFW system to previde sufficient flow to the steam generators, panially caused by the failure of one or more (but not all) EDGs. The remainder of the system fails due to a combination of unrelated faults, such as local failures (primarily valve related) of the AFW turbine steam admission line or the AFW
'BNL. does not currently passess sufficient information to describe the Wolf Creek design.
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~ pump discharge lines and local faults of the turbine driven (TD) pump. Another contributor is TD pump unavailability due to maintenance activities.
The major human error is plant specific. If the AFW system is normally configured so that one pump is locked out, the failure to manually start this pump when needed becomes critical.' The AFW system can also be subject to several common mode failures' All are highly plant specific. One is undetected flow diversion, typically to a second unit, or else back to the Condensate Storage Tank through the mini flow line as in Wolf Creek. The second is steam binding of the pumps due to main feedwater leakage through the AFW pump discharge check valves which flashes to steam in the AFW pump. The last failure mode is the loss of the operating portion of the system due to_ a suction valve failure. This requires a system configuration with a single suction line that serves all pumps so that a single suction valve closure would disable the AFW pumps and require operator action to realign the system suction and manually start the locked out pump.
The bleed and feed mode is the option of last resort. It is highly plant specific.' In some PRAs, only one PORV is considered necessary for system success while in others, both are considered necessary, thereby significantly magnifying the importance of the PORVs themselves PORV system failures can be attributed to failure of a PORV to open on demand or prior closure of a PORV block valve, given loss of the EDG. The block valve requires AC power to reopen.
2.2.3.21 Station Blackout Sequences The dominant accident sequences begin with LOOP as described in 2.2.3.1 followed by failure of all onsite power sources, resulting in a station blackout. One short term station blackout has occurred, during a loss of turbine generator and offsite power startup test. This was caused by an inadvertent isolation of the diesel generator start relays due to a failure to follow procedures.
2.2.3.2.1 Station Blackout with Failure of Decay Heat Removal (Sequence 8)
The loss of all AC power resuhs in an immediate failure of all decay heat removal systems except the turbine driven portion of the, auxiliary feedwater system. The AFW ,
system subsequently fails resulting in core damage. The major contributor to this sequence is the failure of all emergency AC power. This is dominated by the failures to start or run of all diesels or the unavailability of an EDG due to test or maintenance activities with the failure of the remaining units to start /run. The AFW system failures can occur in both the long or short term. Long term failures are attributable to station battery depletion, which results in the loss of instrumentation and control power. Short term failures are due to turbine driven pump or AFW discharge valve failures or the failure to manually open the pump discharge air operated valves.
'In some Wolf Creek procedures, upon an SI signal, the operator places the motor. driven pumps in the pull.to. lock position until power is restored to at least one safeguards bus.
'See note 3, 7
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e s 2.2.3.2.2 Station Blackout with RCP Seal LOCA (Sequence 9)
In this sequence, the loss of all AC power disables all primary system injection, as well as reactor coolant pump (RCP) seal cooling. A RCP seal LOCA occurs, accelerating the loss of the primary system inventory anri the onset of core damage.
The major contributor to this sequence is the failure of all emergency AC power. This is dominated by the failure to start /run of both emergency diesel generators (EDGs) or the unavailability of one EDG due to test or maintenance with the failure of the remaining unit.
The loss of all AC power results in a loss of cooling to the RCP seals. The RCP LOCA accelerates the loss of primary coolant and limits recovery measures to approximately one hour after the LOCA occurs. Major recovery actions are the recovery of AC power and successful restoration of the HPI component cooling.
2.2.4 - Loss of Power Conversion System (PCS) or Transient Followed by Loss of PCS, with Loss of Decay Heat Removal (Sequence 10)
The loss of the power conversion system (PCS) (or a transient followed by a loss of PCS) with the subsequent failure of the AFW system causes the primary system to overheat. The resulting system over-pressurization causes PORV cycling, a loss of system inventory and subsequent core damage. Main feed pump trips comprise over 25% of the total precursor events which have occurred. These include valid, spurious or operator induced low suction pressure trips, feed pump turbine controller failures and gradual losses of condenser vacuum or hotwell level that were not considered to be valid by the operators.
Steam dump valve closure failures, primarily due to positioner linkage problems, contrib-uted approximately 15% The remainder of the loss of PCS precursors are fairly evenly divided among condensate pump trips, feedwater recirculation, control and bypass valve malfunctions, feedwater controller failures and miscellaneous contributors including multi-ple stuck open relief valves and main turbine trips which. induced PCS isolations. The loss of the auxiliary feedwater system is the main contributor to the sequence. The majority of the system unavailability is due to operator failures to manually start either a locked out pump or a pump with a disabled auto start circuit. Hardware failures include steam admission valve and pump local faults. The unavailability of a pump or a pump discharge valve due to maintenance activities or improper position of the manual valve on the pump suction from the Condensate Storage Tank are also contributors.
Failure of a vital AC bus, primarily due.to an inverter failure, disables a steam admission valve and/or the auto start logic for a motor driven pump. .
Dependent on plant specific considerations, the feed and bleed method may be used for decay heat removal. The failures associated with this method have been described previously in 2.2.3.1.
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2.3_ Anticipated Transient Without Scram (ATWS) Followed by Failure of Emergency Boration l
This sequence is initiated by a transient from high power followed by an RPS failure j to automatically scram the reactor. The RPV has survived the initial pressure transient due to a favorable moderator temperature coefficient. The attempts to manually scram are not L successful and emergency _ boration also fails.
The initiator is a transient such as a MSIV closure, partial loss of feedwater, feedwater flow increase or a loss of reactor coolant system (RCS) flow that results in a turbine trip
- and PCS. runback. The mismatch between core power production and secondary loop removal results in RCS coolant loss through the PORVs. Core uncovery and damage occur in forty minutes or less. The Salem plant experienced a RPS automatic scram function failute that was caused by RPS breaker malfunctions.
The failure to manually scram the reactor is attributed to hardware failures of the control rods or drives that prevent insertion or operator error. The failure of emergency boration is dominated by operator failure to initiate injection. System hardware faults have a smaller contribution.
The operator actions to initiate boric acid injection is dependent on system design.
, Some plants have an in line boric acid injection tank with redundant valving that is an l integral part of the charging system /high head injection lineup. The hardware failure for this configuration is small although boron precipitation is a concern. Injection failure is primarily attributable to the failure to manually actuate the system.
At Wolf Creek, two boric acid pumps are utilized discharging through a common, normally closed high flow line to the charging pump suction header. Operator action is required to start both and open the normally closed immediate boration control valve (BG HV 8104). This configuration is more vulnerable to hardware failures related to the use of a single normal'y closed MOV and/or the system success criteria that requires both buric acid pumps to operate. However, at Wolf Creek, there is an alternate immediate boration flow path _through a normally closed manual valve (BG V-177) which must be locally operated in the Auxiliary Sullding.
- 3. COMMON CAUSE FAILURES Certain common cause failures, either hardware or human related, appear as particu-larly important to the risk of core damage from a review of the eleven representative accident sequences. They are the following:
a) Loss of offsite power (LOOP).
'b) Emergency Diesel Generators (EDGs) fail to start or continue to run, c) Component Cooling Water (CCW) pumps fail to continue running.
d) Failure of high head injection (HHI) discharge valves to open.
. e) Failute of LH1 pumps to start or continue running.
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' 4. IMPORTANT HUMAN ERRORS Similar to the previous discussion of common cause failures, certain operator errors appear as particularly important to the risk of core damage from a review of the eleven representative accident sequences. These are:
a) Failure to recover offsite power due to human error.
.b) Failure to switch from the Refueling Water Storage Tank (RWST) to the containment sump, i.e., failure to switch from the low head injection (LHI) phase to the low head recirculation (LHR) phase in response to a large or medium LOCA.
c) Failure to manually start locked out AFW pump, either turbine or motor driven.
d) Failure of manual SCRAM given ATWS or failure to initiate and successfully perform emergency boration, e) Failure to successfully isolate an interfacing LOCA condition.
- 5. SYSTEMS INCLUDED IN GUIDE Table 2 shows the systems which appeared as important based on the representative accident sequences discussed previously, as well as other generic PRA-based information.
The list is not intended to show the relative importance ranking of one system over another since the importance ranking of systems is difficult to achieve from generic insights.
> In using the list, the inspector should select systems for inspection based on both knowledge of any recent operating problems or technical specification outages, as well as on the obviously broad or important effects of support systems due to the loads served by the particular system.
Table 2 Systems Included for Wolf Creek
- 1. Essential Service Water
- 2. Safeguards (AC) Power
- 3. DC Power
- 4. Component Cooling Water (CCW) ~~
- 6. High Head injection (HHI)/ Safety Injection (SI)/High Head Recirculation (HHR)'
- 7. Primary Pressure Relief System (PPRS)
- 8. Auxiliary Feedwater (AFW)
- 9. Low Head Injection (LHI) Low Head Recirculation (LHR).*
- 10. Engineered Safety Features Actuation System
- 11. Refueling Water Storage Tank (RWST)
- 12. Power Conversion System (PCS)
- 13. Emergency Boration/ Chemical and Volume Control System (CVCS) flHR includes room coolers for HHI pumps. HHI includes the Safety injection System (SIS) at Wolf Creek.
- 1.HR includes the Residual Heat Removal System (RHRS).
10
=4 7
+1 4
_ 6. ' SYSTEM INSPECTION TABLES For each of those systems in Table 2, inspection guidance is provided in the form of a
_ failure mode table, an abbreviated walkdown checklist, and a simplified system diagram.
I.
Each of these is explained in detail below.
In using these tables, however, it is essential to remember that other systems and components are also important. If, through inattention, the failure probabilities of other systems were allowed to increase significantly, their contributions to risk might equal or exceed that of the systems in the following tables. Consequently, a balanced inspection
[: program is essential to ensuring that the licensee is minimizing plant risk. The following k tables allow an inspector to concentrate on systems and components that are most significant to risk. In so doing, however, cognizance of the status of systems performing
- other_ essential safety. functions must be maintained.
5 APPENDIX A Table AX 1 - System Failure Modes
- The introduction to this table provides a brief description of the system and the success criteria to the extent it ccold be eztracted from the system descriptions or other
. plant information. (Note that this soco ss criteria may be different from the more cc iserva-tive success criteria contained in the FSAR.)
The entries in this table are the dominant events (component failures, operator errors, etc.) contributing to system failure, provided in rank order according to their risk significance. Since most systems are designed with redundant trains,it will generally take more than one of these events to fail the entire system. No effort has been made to list all of the combinations of the events that are sufficient to produce system failure because that is usually apparent from the system description in the introduction. Where single events are sufficient to fail the entire system, that is noted in the brief discussion of the event. For certain events that are important primarily because of the circumstances of a particular accident sequence, that information-is also noted.
Because PRAs do not contain the detail necessary to attribute the listed failures to the most probable specific root causes, it is necessary for the inspector to draw from experience, plant operating history, ASME Codes, NRC Bulletins and Information Notices, INPO SOERs, vendor notices and similar sources to determine how to actually conduct inspections of the listed items. Where appropriate, codes have been included following each event description to indicate which licensee programs / activities provide inspectable aspects of the risk. These codes are as follows:
O Normal and emergency operating procedures, check off lists, technical specifications.
training. etc.
S Periodic surveillance activities, procedures and training.
M Preventive or unscheduled maintenance activities, technical specifications, procedures and training.
g f 11 gi e; -
s .
. i F-T- Periodic testing or in service inspection activities, procedures, and training.
C Periodic calibration activities, procedures, and training.
Each failure mode is correlated to the appropriate accident sequence (s) described in .
Table 1 and categorized as of high (H) medium (M) or low (L) importance. In nearly all cases, the importance categories are numerically based taking into account the event's contribution to the eleven representative accident sequences.
Table AX 2 - Modified System Walkdown This table provides an abbreviated version of the licensee's system checklist, where available, but includes only those items which are related to the dominant failure modes. It is generally much less than the normal t,hecklist. It can be used to rapidly review the line up of important system components on a routine basis. Caution should be observed when using the checklists, since they are based on certain versions of the licensee's system operating instructions. Valve numbers used are those identified in the licensee system checklists, or P&ID's.
Figure AX - Simplified System Diagram A simplified line diagram is provided for each system treated. These are intended to aid in visualizing the system configuration and the location of the components discussed in the two tables. Since they are neither complete nor controlled, they should not be used in place of up-to-date P&ID's during inspection activities.
APPENDIX B Table B1 - Plant Operations Inspection Guidance This table is a collection of all of the risk significant operator actions listed in the preceding system tables. It is provided as a cross reference for use in observing operator actions and training.
Table B2 - Surveillance and Calibration Inspection Guidance This table is a collection of all of the risk significant components listed in the preceding system tables that are considered to be significantly influenced by surveillance and calibration activities, it is provided as a cross reference to assist in selecting risk important activities for observation during inspections of the licensee's surveillance and calibration pmgrams.
Table B3 - Maintenance Inspection Guidance This table is a collection of the risk significant components listed in the preceding system tables that are considered to be significantly influenced by maintenance activities.
It is provided as a cross reference to assist the inspector in selecting risk important activities for observation during inspections of the licem,ee's maintenance program.
12
~f' t
.. 1
. Important factors include the frequency and duration of maintenance as well as errors that i degrade the component or render it inoperable when it is returned to service.
APPENDIX C ,
Table C1 - Containment and Drywell Walkdown Table L
Because they are normally inaccessible during operation, a separate walkdown check-list is provided for those components listed in the preceding system tables that are located inside the containment or drywell. This is intended for efficient inspection of those items
@1 the opponunity arises.
APPENDIX D (OPTIONAL)
System Dependency Matrix Whenever it is readily available, matrices showing the dependencies, and inter-dependencies, of front line ESF systems and support systems, and also of support systems to other suppon systems are provided to aid the inspector in determining what other systems (or trains of systems) are affected when a particular system or train fails. This can be helpfulin deciding the imponance of systems and in reviewing the adequacy of operator actions to restore the systems to service when they become inoperable.
- 7. REFERENCES 7.1. Generic Risk Based Information
- 1. R. Travis and A. Fresco, " Development of Guidance for Generic, Functionally Oriented PRA-Based Team Inspections for PWR Plants - Identification of Risk Imponant Systems Components and- Human Actions," BNL Technical 1.etter Report TLR-A-3874-Tla, October 1988 (Cover letter to Dr. J.W, Chung, USNRC, dated November 7,1988).
- 2. R. Travis, " Fin A-3874 Task Ib Inspection Matrix," BNL Technical Letter Report with cover letter to Dr. J.W. Chung,' USNRC, dated November 7,1989.
- 3. R.E. Gregg and R.E. Wright, "Appendi t Review for Dominant Generic Contributors," Idaho National Engineering Laboratory, Report No. BLB-31-88, March .1988.
13
7 7.2- Other
References:
Plant Specific Risk Based Information
- 1. M.F. Hinton and R.E. Wright, " Pilot PRA Applications Program for Inspection at Indian Point Unit 2," Idaho National Engineering Laboratory, Informal Report EGO EA-7136, Rev.1, July 1986. -
- 2. . A. Fresco, et al., " Indian Point Unit 3, Probabilistic Risk Assessment Based i System Inspection Plans," Brookhaven National Laboratory, Technical Report A-3453-87-1, Rev. O, May 1987.
4
- 3. C.L. Atwood, et al., "PRA Applications Program for Inspection at the Zion -
Nuclear Power Station Draft Report," Idaho National Engineering Laboratory, .
Informal Report EGG-EA-7304, June 1986. '
- 4. M.F. Hinton and R.E. Wright, "PRA Applications Program for Inspection at Seabrook Station Draft Report," Idaho National Engineering Laboratory, Informal Report EGG EA-7194, March 1986.
- 5. R.E. Gregg, et al., "PRA Applications Program for Inspection at the Strry Nuclear Power Station Unit 1 Draft Report," Idaho National Engineering Laboratory, Informal Report EGG-REQ 7746, July 1987. i l
- 6. R.E. Gregg and R.E. Wright, "PRA Applications Program for Inspection at '
Millstone Unit 3 Draft Report," Idaho National Engineering Laboratory, Informal i Report EGG SSRE 8016, March 1988, i 1
- 7. P. Saylor and P. Lobner (editor), " Nuclear Power Plant Sourcebook Wolf Creek :
50 482," Science Applications International Corp. Report No. SAIC 88/1996, l Revision 1, February 1989.
=e
- I 1
I I
4 14
- 4. , . - .
c( .' ( -
t p g. ' .. d t
n\ g; .
- 'y e
APPENDIX A 4
TABLES OF (1) IMPORTANCE- BASIS AND FAILURE MODE IDENTIFICATION, AND (2) MODIFIED SYSTEM WALKDOWNS e
8 i
r.: q , a iI t-j, . .
WOLF CREEK Table A.1 1. Importance Basis and Failure Mode Identification ESSENTIAL SERVICE WATER SYSTEM (ESWS) ,
Mission Success Criteria The Essential Service Water System (ESWS) consists of two 100% capacity, identical redundant cooling water trains which cool plant components for the safe shutdown of the reactor following an accident. Water is drawn from the Ultimate Heat Sink (UHS) and circulated through the components and back' to the UHS.
The ESWS also provides emergency makeup to the Spent Fuel Pool and the Compo-nent Cooling Water System (CCWS). The ESWS is the backup water supply for the Auxiliary Feed System. The site related portion of each ESWS train consists of a pump ^
with an automatic, self ' cleaning strainer, a pump prelube storage tant:, traveling water screen, supply and return piping, valves, associated instrumentation, and a discharge structure in the Ultimate Heat Sink (UHS):
The UHS is a normally submerged cooling pond, formed by providing a volume of r L
approximately 440 acre feet behind a dam built in one finger of the Main Cooling Lake.
The ESW pumps draw wMer from the UHS at a maximum temperature of 90 degrees F and a minimum temperture of 32 degrees F, and supply it for cooling or makeup to the following compor~nts or systems:
- Component Cooling Water Heat Exchanger
- Containment Air Coolers
- Diesel Generator Coolers
- Component Cooling Water Pump Room Coolers
- Centrifugal Cna er Pamp Room Coolers
- Auxiliary Feedwatu rNmp Room Conlers
- Safety Injection Pump Room Coolers L
- Residual Heat Removal Pump Room Coolers
- Containment Spray Pump Room Coolers
- Penetration Room Coolers
- Fuel Pool Cooling Pump Room Coolers A-1
lic , ,
- Control Room Air Conditioning Condensers l
- Class IE Switchgear Air Conditioning Condensers
- Indrument/ Service Air Compressors and After Coolers
- Auxiliary Feedwater System *
- Spent Fuel Pool Cooling and Cleanup System
- Component Cooling Water System Each train of the ESWS is intercoanected with the Service Water System (SWS). Two . 1 motor operated isolation valves are provided in each crosstic heater where it connects to
, the SWS, These valves are located in the pipe chase area of the 1974 foot elevation of the Control Building. In addition coullng water flow is maintained following an accident to the safeguard powered air compresers (CKA01 A and CKA01B) and associated after coolers.
The air corapressors are automatically isolated on high flow (indicative of leakage) or they l can be manually isolated.
Flow restrictive bypass lines are provided at the outlet of the Component Cooling Water (CCW) heat exchangers., .This provides a path for reduced flow during accident conditions. Motor operated bypass valves ar also provided in the outlet lines fro'm the containment air coolers, outside the contai - . t. This provides an increased amount of l flow through the coolers following a LOCA or loss of offsite power.
The ESWS normally supplies water at a higher pressure than the cooled safety related components. Therefore, if leakage occurs it will be into the system being cooled or in the I case of ESW piping and valvescinto the floor drain system. 'l Accident importance Inspectic .
Dominant Failure Modes Sequence Category Activities
- 1. Failure of any of the following EsW *alves which isolates 1.5 H s.M.T.C EsW flow to CCW heat exchangers N-- . ally closed MOV fails to Open CCW HX EEG01A CCW HX EEG01B Inlet EF.HV.51 EF HV 52 Outlet EF HV 59 EF.HV 60 '
EF V058 EF V090 Locked throttled return Bypass valve inadvertently. closed (Note: Either Train A or B MOVs are normally closed, but not both.)
- 2. EsW Pump train A or B cat for maintenance 1.5 M M. T A-2
. . - -. ,~
, , : e t
Accident Importance Inspection Dominant Failure Modes Sequence Category Activities
- 3. ESW Pump PEF 01 A or PEF 01B fails to start and run IJ M S,M,T ,
The ESW pumps are started by the diesel generator load sequencer upon ei her a Safety injection or Loss of Offsite Power Signal. Failure of the signalinputs can prevent the pumps from auto starting when required. Failure of the ESW pump pre. lube storage tanks to supply sufficient water above the pit water level to prevent the pernp line shaft bearings fromerunning dry during start up can also fail the ESWS pumps when required. The 43 gallon tank's size is based on supplying a minimum of a five. minute supply of water at six gpm, with no makeup from the pump discharge line. The tank is continuously supplied water by a connection on the ESW pump discharge, down- ;
stream of the discharge check valve and strainer, from the SWS pumps. The tank provides water to the lineshaft bearings and stuffing box continuously (provided the dis-charge line is pressurized) even during periods when the pumps are idle. When the ESW pump is running, flow in the supply line from the tank reverses and discharges through the overflow.
Locked throttled manual valves V.245 and .V 246, for .
Trains A and B. respectively, must remain Open to supply water to the tanks.
- 4. Pump discharge MOV, check valve or header isolation IJ M S.M.C valve fails to open or remain open Both ESW pump discharge lines include a vent line with a normally open motor. operated valve, EF HV 97 for Train A and EF HV 98 for Train B. The vent valves remain open for 15 seconds after pump start, then automatically close.
This is to vent the air in the pump calumn and discharge piping to prevent water hammer. Inadvertent closure of these valves when required can fail the ESWS.
Each pump also contains a check valve in the discharge line prior to the self cleaning strainer, V001 for Train A and for Train B. These check valves must open to allow 3S flow.
! Locked open manual valves on the ESW pump discharge lines must remain open:
, Train A Train B 1
EFW Inlet Upstream isolation EF V104 EF Vil3 EFW Inlet Downstream Isolation EF V107 EF V116 p
i I
s 1
l l
l l
l A3 L
I e
.s : .s Accident importance Inspection Sequence Category Activities Dominant Fallure Modes 1.5 L S.M T.C
- 5. Non essential load isolation valves fail to close Following a LOCA or loss of offsite power, the safety.
related signals isolate the SWS supply and discharge from -
the E' #J by closing the cross. tic line isolation valves.
s Two hv.ation MOVs are movided in each of the SWS supply and discharge lines to the ESWS:
Train A Train B Supply EF HV.23 EF HV.24 EF HV 25 EF HV.26 Return EF HV.39 EF HV.40 EF HV-41 EF HV.42 In addition, the normally closed MOVs, EF HV 37 for Train A and EF HV.38 for Train B must open to allow ESWS flow to discharge to the Ultimate Heat Sink, pass.
ing t4-ugh a single locked open manual valve in each
' line, V108 in Train A and Vil? in Train B.
1.5 L S,M
- 6. Pump strainer plugged Two traveling water screens DEF01A and DEF01B, are provided for each train which filter the pump suction from large debris. Spray flow from the ESW header downstream of the self cleaning strainers on the ESW pump discharge side,is provided by autematic opemng of a throttle valve and MOV (EF HV.91 Train A and EF HV.92 Train B),
- upon ESW pump start.
Small impurities and debris which may have washed through the traveling screens are filtered out by the ESW-strainers, IFEF02A and IFEF02B.'
Failure cf either the screens or strainers to function prop-erly can cause loss of de ESWS.
s A-4
,;e .e.
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Essential Service Water System (ESWS)
TABLE A.I.2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow. Sup. Required Actual Descripdon ID No. location Position Position Breaker # locadon Position Posidon Train A EF.V104 Locked ;
E$W A Inlet Open -
Upstream isola.
tion ESW A Inlet EF.V107 Locked Downstrearn Open Isolation ESW A to Uld. EF.V108 Locked mate Heat St.s Open Header Isolation Traveling Water EF.V003 Locked '
Screen IA Open Wash Isoladon ESW Pump A . EF.V002 Locked Discharge Isola. Open L'
tion ESW Pump A EF.V245 Locked Pre. Lube Stor. Throttled age Tank 1A Fill ESW Pump A EF.V162 Locked Pre. Lube Stor. Closed age Tank Drain ESW Traveling EF.V262 Closed
- Water Screen Open 1A Warm Water Header Up.
l , stream Isolation
- In accordance with STF GP.001, EF.V262 and EF.V264 rnay need to be open to comply with seasonal requirements.
A-5
e .
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Essential Service Water System (ESWS)
TABLE A.12 MODIFIED SYSTEM WALKDOWN (Cent'd)
Desired Actual Pow. Sup. Required Actual ID No. Loca6cn Position Position Breaker # Location Position Position Description ESY! Tssveling EF.V264 Closed
- Water scru.n Open c 1A Warm Water
- Header Down-
-; stream Isolation ESW A to Air EF.V045 Locked compressor / Open Aftercooler A isolation Air Compressor / EF.V143 Locked Aftercooler A Throttled ESW A Retam
' Isolation Air Compressor EF.V346 Im ked A Aftercooler Open Check Valve EF V046 Down-stream Isolation Train B EF.V113 Locked M Inlet Open Upstream Isola.
tion .
ESW B Inlet EF.V116 Locked Downstream Open -
Isolation
- In accordance with STN GP 001. EF V262 and EF.V264 may need to be open to comply with seasonal requirements.
A-6 i
~
, y.. .
]
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Essential Service Water System (ESWS) -
TABLE A.12 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Locadon Posidon Position Breaker # location Position Posidon ESW B to Ulti- EF Vil7 Locked' mate Heat Sink Open Header Isolation Traveling Water EF V006 lecked Screen IB Wash Open
, Isolation ESW Pump B EF V005 lecked Discharge Isola. Ope,n tion ESW Pump L EF V246 Locked ,_.
Pre Lube Stor- Throttled . . ._ . !
age Fill l
ESW Pump B EF-V163 Locked I Pro Lube Stor. Closed I age Tank B Iso-lation ESW Traveling EF V263 Closed
- Water Screen Open IB Warm Water Header Down-stream Isolation
- In' accordance with STN GP-001. EF V263 and EF-V263 may need to be open to comply with seasonal requirements. ,
A-7
_ _ . _ . ___ -__ -_._m___.-__ ____._
, s l
WOLF CREEK GENERATING STNIION RISK. BASED INSPECTION GUIDE Essential Service Water System (ESWS) l 1
TABLE A.I.2 MODIFIED SYSTEM WALKDOWN (Cont'd) -
Desired Actual Pow. Sup. Required Actual Descriptica ID No.~ 1.ocadon Position Position Breaker # location Posidon Posidon ESW B to Air EF.V075 Locked Comprenor/ Open )
. Aftercooler B Isolation Air Compressor / EF.Vl44 Locked Aftercooler B Throttled ESW B Return Isolation l Air Compressor EF.V345 Locked B Aftercooler Open I Check Valve ,
EF.V076 Down.
strearn Isolation l
EF HV.23 NG Control Build. ON l ESW to SWS O!AGF1 ing Train A )
Isoladon l EF HV.25 NG Same ON ESW to SWS 02AHF1 Isolation l l
EF HV-41 NG Same ON l ESW to SWS OlAFR4 l Isolation EF MV.37 NG Same ON 4 ESW u UHS OlAERI I Isolation l EF HV-39 NG Same ON ESW to SWS 02AFR1 Isolation Traveling Water NG ESW ON Screen Motor 05EBF211 Pumphouse DFEF01A Space Train A Hester 1
A8
f- -O. ,
!?
- WOLF CREEK GENERATING STATION ,
RISK. BASED INSPECTION GUIDE i Essential Service Water System (ESWS)
TABLE A.12 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. location Position Position Breaker # Location Position Posioon EF HV 91: NG Same ON Screen Wash ' 05EDF2 Water Valve EF HV.97 ESW NG Same ON Pump Discharge 05EDF3 Line Air Dis- 1 charge Valve DFEF01 A ESW NG05EDF4 ESW ON Traveling Water Pumphouse Screen Train A l
EFPDVl9 NG05EEF3 Same ON ESW Strainer Backwash Trash Valve DFEF02A NG05EFF3 Same ON ESW Self.
Cleaning Strainer ESWA/SW EF HIS. Main Open Cross Connect 25 Control Valve (RLO19) Room Switch
( Lineup Train A ESWA/SW EF HIS. Main Open Cross Connect 23 Control l Valve (RL019) Room l, Switch Lineup Train A ESWA to SWS EF HIS. Main Open/
Isolation
- 39 Control Closed (RL019) Room Switch Lineup Train A
A9
. . 1 l
WOLF CREEK GENERATING STATION i RISK. BASED INSPECTION GUIDE Essential Service Water System (ESWS)
TABLE A.12 MODIFIED SYSTEM WALKDOWN (Cont'd) -)
Desired Actual Pow. Sup. Required Actual !
Description ~ ID No. Location Position Position Breaker # 1.ecation Position Position ,
l ESWA to SWS EF HIS. Main Open/
Isolation' 41 Control Closed (RLD19) Room Switch Lineup Train A ESWA to Ulti. EF HIS- Main Closed mate Heat Sink 37 Control I (RLh19) Room Switch Lineup .
Train A ESW Pump A EF Same Normal HIS.
55A ESW A Dis. EF Same Closed charge Isolation HIS 85 ESW Pump A EF Same Open Discharge Air . HIS-97 Release Valve ESW Screen EF Same' Fast Wash Speed Se. HIS 3 lector Switch -
ESW Traveling EF Same Auto Water Screen A HIS 3 i ESW Pump A EF Same Normal HIS.55B
A 10
i 7 ..:
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Essential Service Water System (ESWS) e TAltLE A.12 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual - Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # location Position Posidon ESW Self EF - Same Normal Cleaning HIS 19 Strainer EF HV 24 NG Control Build. ON ESW to SW 01AGF2 ing Train B Isolation i EF HV 42 NG Same ON ESW to SW OlAFR3 Isolation EF HV 26 NG Same ON ESW to SW 02AHF2 Isolation EF HV 38 NG Same ON ESW to UHS 02AHF3 Isolation Valve EF HV 40 NG Same ON ESW to SW 02AER4 Isolation Traveling Water NG ESW ON Screen Motor 06EBP211 Pumphouse DFEF01B Space Train B Heater EF HV 92, NG Same ON Screen Wash 06EDF2 Water Valve EF HV 98, NG Same ON ESW Pump Dis. 06EDF3 charge Line Air Discharge Valve A ll
WOLF CREEK GENERATING STATION t RISK. BASED INSPECTION GUIDE Essential Service Water System (ESY/S)
TABLE A.12 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position DFEF01B ESW NG06EDF4 ESW ON Traveling Water Pumphouse Screen Train B EFPDV20 ESW NG06EEF3 Same ON Strainer Back.
wash Trash Valve DFEF02B, ESW NG06EFF3 Same ON Self Cleaning Strainer
- ESWB/SW EF Main Open Cross Connect HIS 26 Control Valve (RL019) Room Switch Lineup Train B Control Room EF Same Normal Isolate Switch HS 26A for EF HV.26 (NG02A.
HF2)
ESWB/SW EF Same Open Cross Connect HIS 24 Valve (RLD19)
ESWB to SWS EF Same Open/
Isolation
- HIS-42 Closed ESWB to SWS EF Same Open/
Isolation
- HIS-40 Closed
A-12
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Essential Service Water System (ESWS)-
TABLE A.I.2 MODIFIED SYSTEM WALKDOWN (Cont'd) i Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position ESWB to Ulti. EF Same Closed mate Heat Sink HIS 38 Control Room EF HS. Main Normal -
Isolate Switch 38A Control for EF HV 38 (NG02A. Room HF3) Switch Lineup Train B ESW B Dis. EF Same Closed charge Isolation HIS 86 ESW Pump B EF Same Normal HIS 56A ESW B Screen EF ESW Auto Wash HIS-4 Pumphou.
se' Switch Lineup Train B ESW Screen B EF Same. Auto Spray Valve HIS 92 .
ESW Pump B EF Same Normal HIS 56B ESW Self EF Same Normal Cleaning HIS 20 Strainer B s
l A-13 l
l l
J
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Essential Service Water System (ESWS)
TABLE A.12 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Descripdon ID No. Location Position Position Breaker # Location Position Position ESW Pump B EF Same Open .
Discharge Air HIS 98 Release Valve ESW B Screen EF Same Fast Wash Speed Se- HS 4 lecdon Switch A-14
=ugigi - -
- @ q)
,k & Tk IMAGE EVALUATION ,[gj Mi*
% gf/ TEST TARGET (MT-3) If g#4 k///7[i/
NA'$, *%[j/[@ ;f l.0 if IM Ell y 'y El i,1
[5 lMe 1.8 1.25 1.4 1.6
- 150mm >
- 6" >
4
- ~%z,, /<$4%%
+$f*$ffff; o# L sm .
s +Q*k+
. + .p l <d*4j t +-
- ig$'
IMAGE EVALUATION TEST TARGET (MT-31
///%j /"4
- 4 k/// %*
'W 'k , '
l.0 !!82 B14 E lf HM u !? m I.8
=
1.25 l.4 1.6 4 - 150mm >
4 6" >
t / '4 4)#y*/, ;;;g , yy ,,,
, , ,, ////g%
54.:, ,- . ..
.. :- t .
,,.,y,..
- & 4 l 0 4'IS.h
- + IMAGE EVALUATION ///'p/ gb4 gf/ #,
k//f7%Y "
q; TEST TARGET (MT-3)
</,p y,, /4,,, f+4
+ %
l.0 E EE M uts g=n t eu 11 5M m
Ubb 1.25 I.4 1.6 i
4 150mm >
4 6" >
&*)g'
% ,7, / ;*y)A////h sp
'/b p y7////;
o i 44*
re 4 CA - __. ;..'
WOLF CREEK GENERATING STATION (ESWS)
TABLE A.12 (Cont'd)
REFERENCE DOCUMENTS
. TITLE 1.D. NO. REV DATE Licensed Operator Initial Training bsson:
- 1. " Essential Service Water System" LO 1408900 000 02/29/88
- 2. " Service Water System" LO 1407600 001 02/25/87 Drawings
- 1. ShTCPS *P&lD - Essential Service Water System" M.12EF01 0 12/09/86
- 2. SNUD."~, "P&lD - Essential Service Water System" M 12EF02 0 12/09/86
- 3. SNUPPS *P&lD - Service Water System" M 12EA01 0 08/07/84
- 4. ShTPPS "P&lD - Service Water System" M 12EA02 0 08/14!85 Procedures
- 1.
- Essential Service Water Valve Breaker and Switch Lineup" CKL.EF 120 12 02/10/89 l
l A-15
=-
- 'l r m r 3 COOLNG t PUMP LAKE l
- 888 -
--* COOLERS t ; t j r , r ,
RHR HEAT i CCWS e EXCHANGERS v J t J ,
a L
- .i i r y I r , r , ;
L LA.TIMATE --* ESWS - TO RNM .l m '
EXCHANGER ' HEAT SINK HEAT SINK L ; t ; ;
r , *
, DIESEL ,TO SWS .
GENERATORS
' DISCHARGE L J t
i r ' '
OTHER '
SAFEGUARD ccws - componentcesshewesersyneem '
ESWS - Essensaf Ser*e wseer System HEAT LOADS ;
( l sws - serenwaist syneem l Figure A.1-1. " Wolf Creek Cooling Water Systems Functional Disgram" i (Source ' SAIC 88/1996, Figure 3.1) .
e
- = - %-, .s- w. .- .. % --==
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. . -- c WOLF CREEK GENERATING STATION.
Table A.21, Importance Basis and Failure Mode Identification SAFEGUARDS (AC) POWER SYSTEM Misdon Success Criteria 1 The Safeguards (AC) Power System consists of two 4160 VAC buses, four 480 VAC l buses, four 120 VAC vital instrumentation- buses, four 125 VDC buses, two dedicated diesel generators, and their associated motor control centers, breakers, transformers; i chargers, inverters, and batteries. l I
Each 4160 VAC bus is normally powered from offsite power sources through.either
' the No. 7 or startup transformer. On loss of offsite power the breakers open and the diesel generators start and their associated breakers close to load the diesels on the emergency j buses. The 4160 VAC buses provide power to the large pumps such as the centrifugal charging, the safety injection, CCW and residual heat removal pumps.
Each 4160 VAC bus feeds two 480 VAC buses through transformers. The 480 VAC -
buses are primarily used to power a multitude of MOVs and small pumps such as the charging pump oil and diesel cooling water pumps. They also provide power to four battery chargers.
Accident Importance laspection Doeninant Failure Modes Sequence Category Activities
- 1. Failure of EDGs DGNE01, DGNE02 to start & run following 7,8.9 H 0.S.M.C loss of offsite power.
- 2. EDGs unavailable due to test or maintenance, With only two 7.8.9 H M diesels, this unavailability should be relatively important.
- 3. Failure to restore AC power after station blackout with con- 8.9 H O current RCP seal LOCA. (Refer to Emergency Procedure EMG CS 02 %ss of All AC Power with S1 Required.")
- 4. less of vital AC bus. (Refer to Off. Normal Procedure OFN 10 H M.s.C 00-021, %ss of Vital 120 VAC Instrument Bus", or OFN 00 027, %ss of Vital 480 VAC Bus NG01. NG02. NG03 and NG04.")*
- 5. Improper EDG post maintenance valve or breaker lineup. 7.8.9 M O.M ESW valves V052. V053. V079. V080
' Vital AC is critical support to operation of turbine driven AFW pump for auto actuation and staam control valve. See Table A.10,.1.
A-19
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Importance Inspection :
~
Accident 0; Dominant Failure Modes Sequence - Category Activities i- 6.' -
Cooling water valves for EDG fail' to open. (BNL does not 7,8.9 - L S.M,C )
- currently possess adequate information on EDO cooling water flow control.) ,
L
- 7. Failum 'of EDG output breaker to close. (BNL does not -7.8.9 3.M.C currently possess sufficient EDG information to identify components.)
~
- 8. Failure to transfer to reserve source of AC power and failure 7.8.9 L S.M -)
of EDO start signal. _
l
- 9. Failun of Invener* 6.7,10 L- M.S.C Inverter No.11. Breaker No.111 -
Invertet No.13. Breaker No. 311 inverar No.12. Breaker No. 211 ,
Inveter No.14. Breaker No. 411 j
' Vital AC is critien) support to operation of turbine driven AFW pump for auto actuation and steam control valve. See Table A.101.
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A-20
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-. WOLF CREEK GENERATING STATION
- RISK BASED INSPECTION GUIDE Safeguards-(AC) Power System
, TABLE A.2 2 MODIFIED SYSTEM WALKDOWN I
1 .
Desired Actual - Pow. Sup. Required Actual ;
l- ~
Description ID No. Location Position Position Breaker # location Position Position ESW A to EF. locked Diesel Gener. V272 Open -
ator IA Cool-(: en and IE Swgr. Con-denser SA Isola-tion ESW.to Diesel EF- lecked Generator IA V052 Open Coolers !sola.
tion Diesel Gener. EF. . Locked stor 1A Cool. V053' Throttled ers ESW A 40%
Return isola. Open tion Diesel Gener. EF- Locked stor IA Cool. V273 Open ers and IE Swgr. A/C Condenser SA ESW Peturn Isolation ESW B to EF. lacked Diesel Gener. V274 Open stor IB Cool-ers and IE Swgr. A/C ,
Condenser SB Isolation ESW B to EF. Locked Diesel Gener. V079 Open stor B Coolers isolation Diesel Gener. EF. Locked stor B Coolers V080 Throttled ESW Return 30%
! solation Open A-21
T
, WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Safeguards (AC) Power System TABLE A.2 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
~
Desired Actual Pow. Sup. Required ' Actual Description ID No, location Position Position Breaker # Location Position Position -
Diesel Geer. EF. Locked stor IB Cool. V275 Open ers and IE Swgr. Con.
denser SB ESW B Return Isolation ESW A to EF. Locked _ _ _
Peneestion V041 Open Room Coolm 15A isolation Penetration EF. Locked Room Cooler V042 Open 15A ESW A Return isola.
tion ESW B to EF. Locked Peneustion V083 open Room Cooler ISB Isolation Penetration EF. Locked Room Cooler V084 Open ,
15B ESW. B Return hole.
tion NCTTE: Electrical lineups for the Safeguards Power Sysism loads are shown on the respective Modified System Walkdown Tables.
A-22
i 1
WOLF CREEK GENERATING STATION
[ Safeguards (AC) Power System]
TABLE A,2 2-(Cont'd)
REFERENCE DOCUMENTS TITLE I.D. NO. REV DATE WCGS " Licensed Operator Training Document
- 1. WCGS
- Essential Service Water Valve. Breaker and CKL EF 120 12 02/10/89 Switch Line.:p"
- 2. WCGS " Diesel Generator NE01 and NE02 Valve Checklist" CKL KJ.121 6 08/23/87 c
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t Figure A.2-1. " Wolf Creek 4160 VAC and 480 VAC Electric Power Distribution ^
System Showing Component Locations" j
(Source SAIC 88/1996 Figure 3.6-2) t e Y
_ - - . _ _ - _ _ - . . _ _ - . -_ ____-_.____.___.__.._-___.____.-_-____t - _ - - - . .
\
WOLF. CREEK GENERATING STATION i
LTable A.31. Importance Basis and Failure Mode Identification
- n l DC POWER SYSTEM ;
Mission Success Criteria Loss of 125V DC power is an accident sequence initiator which contributes to loss of decay heat removal capability through loss of PORV's and AFW.125V DC power is supplied for the plant protection system, control and instrumentation and other loads for start up, operation and shutdown modes of plant operation. The four 125V DC buses are supplied by four station batteries and also from the 480V AC buses through battery chargers. The 120V AC vital buses are fed from the 125V DC buses through four station inverters (uninterruptible power supplies).
Accident Importance laspection Dominant Fallure Modes Sequence Category Activities
- 1. Loss of 125V DC bus 6 H 0,S,M T.C Train A BUS NK01 BUS NK03 Switches 89NK0101 89NK0301 89NK0102 89NK0302 Train B BUS NK02 BUS NK04 Switches 89NK0201 89NK0401 89NK0202 89NK0402
- 2. Failure of on line charger and failure of spare to energir.e on 6 M S.M,T.C demand Train A Charger NK21 Charger NK23 Breakers 52NG0103 52NG0303 Switches 89NK0102 89NK0302 Train B l Charger NK22 Charger NK24 Breakers 52NG0203 52NG0403 l.
Switches 89NK0102 89NK0302 l
A-25
e e 1
Accident Importsace Inspection - l Dominant Failure Modes Sequence Category Activities
- 3. - Operational test or maintenance error resulting in 6 L O.M l a) deenergizing or cascading of DC power supplies i b) failure to properly restore batteries or l charger after maintenance !
Refer to Operating Procedure SYS NK 131 *Energitation of 125V DC (Class IE) System (NK01, NK02, NK03, NK04)". -
]
Train A Bauery NK11 Bauery NK13 Chrager NK21 Charger NK23 l I
Train B l
attery NK12 Bauery NK14 .
- t. arger NK22 Charger NK24 1 I
4 Failui of Batteries 6 L M.S.T Battery NKil, NK13 NK12, NK14 4 l
Battery failure typically oc:urs during extended station black-out scenarios where AC power is lost for several hours. Refer i to Emergency Procedure EMG C.0
- Loss of All AC Power" for load shedding instructions
- 5. Loss of battery room ventilation. 6 L M.S.T As in battery failure, loss of battery room ventilation is
. typically a long term failure occuring durirg extendert stadon J
blackout scenarios. It can cause two types of faults:
- a. Hydrogen building with risk n explo.
sion, and
- b. Temperature increase with risk of bat.
. tery charge output failure.
A 26 1
-.. , .y i
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE DC Power System TABLE A,3 2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow, Sup. Required Actual Description ID No. Location Position Position Breaker # tocation Position Position Bus NK01 Bat. 89NK Closed tery NKil Out. 0101 put Switch Bus NK01 Bat. 89h% Closed tery Charger 0102 NK21 Output Switch Bus NK03 Bat. 89NK Closed tery NK13 Out. 0301 put Swit;lt Bus NK03 Bat- 89NK Closed tery Charger 0302 NK23 Output Switch Bus NK02 Bat. 89NK Closed -
, tery NK12 Out. 0201 L put Switch
(
Bus NK02 Bat- 89h% Closed tery Charger 0202 NK28 Output Breaker l
l l
l A-27
L . ( )
." ] l i
r WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE DC Power System TABLE A.3 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow.' Sup. Required Actual Description ID No. Locatica Position Position Breaker # location Position Position Bus NK04 Bat. 89NK Closed 1 tery NK14 Out. 0401 1 put Switch Bus NK04 Bat. 89NK Closed tery Charger NO2 1 NK24 Output Switch 480V Bus $2NG Closed NG01 Input 0103 Breaker to Bat.
tery Charger .
NK21 480V Bus 52NG Closed NG03 Input 0303 Breaker to Bat-tery Charger NK23 480V Bus 52NG Closed NG02 Input 0203 Breaker to Bat.
tery Charger NK22 480V Bus $2NG Closed NGG4 Input NO3 Breaker to Bat, tery Charger NK24 i
A-28
O *
- r. ,
t WOLF CREEK GENERATING STATION DC Power System TABLE A.3 2 (Cont'd)
REFERENCE DOCUMENTS TITLE I.D. NO. REV DATE WCGS Licensed Operator Training Document
- 1. *DC ud Instrument Power Systems" LO 1506300 000 02/03/88 Drawings i 1. " Class IE 125V DC System Meter & Relay Diagram" E 0 INK 01(Q) 15 12/02/86
- 2. " Class IE 125V DC System Meter & Relay Diagram" E 0 INK 02(Q) 17
- 3. " Low Voltage System Class IE 480V Single Line E llNG01(Q) 1 07/14/87 Meter & Relay Diagram" j' 4. " Low Voltage System Class IE 480V Single Line E 11NG02(Q) 1 07/14/87 j Meter & Relay Diagram" i
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System Showing Component Locations" -
l (Source SAIC 88/1996, Figure 3.6-4) ,
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l WOLF CREEK GENERATING STATION Table A.41. Importance Basis and Failure Mode Identification ,
COMPONENT COOLING WATER SYSTEM Mission Success Criteria The compor'ent cooling water (CCW) system is a non radioactive, closed loop cooling water system that removes heat generated by various plant components and transfers this heat to the Essential Service Water (ESW) system. The CCW system consists of two 100%
capacity trains that provide cooling of engineered safety features and an additional loop that cools non essential equipment. This non essential loop is common to both trains and can be isolated during accident conditions. Mission success is provided by one of the two 100% capacity pumps and the one heat exchanger in one of the two essential trains.
The essential trains of the CCW system provide cooling for the following components:
SI containment spray pumps Centrifugal charging pumps L RHR pumps and heat exchangers Fuel pool cooling heat exchangers Heat loads generated by the following components are removed by the non.cssential l CCW loop:
1 Letdown heat exchanger Excess letdown heat exchanger Positive displacement charging pump Seal water heat exchanger Reactor coolant pumps -
Automatic control of certain components in CCW system components is provided. The automatic actions and initiation signals include the following:
- Automatic startup of the standby CCW pump in a train based on low discharge pressure on the operating pump.
- Automatic startup of a CCW pump based on startup of a centrifugal charging pump in the same train.
l - CCW pump startup and isolation of the non essential loop based on a safety injection signal.
- Automatic makeup to the CCW surge tanks from the demineralized water storage and transfer system.
A 31
s .
-- Non-essential loop isolation valve closure based on a high containment isolation
'M Signal.
- Isolation of the surge tank vent based on high radiation in any CCW system loop.
Remote manual operation of the CCW pumps, inlet isolation Valves for the RHR heat exchangers, and cooling to the reactor coolant pumps and excess letdown heat exchanger is provided from the control room. l Accident importance Inspection I Dominant Failure Modes Sequence Cattfory Activities
- 1. Pumps fall to start & run 1.5 H S.M.T.C Pumps can fall to start after automatic initiate signals or i manus) initiation from the control room. l Train A Train B PEG 01A PEG 01B 1 PEG 01C - PEG 01D ]
- 2. Local fault of heat exchanger valves that isolate or severely restrict CCW flow Train A Train B Heat Exchanger EEG01A . EEG01B luked open manual valves CCW Inlet V019 V044 Outlet V035 V060
- 3. . CCW pumps out for maintenance (PEG 01A and PEG 01C)
- PEG 01B and PEG 01D 4 Local fault of CCW pump suction and discharge valves re-stricting CCW flow Train A Train B Pump suction valves V132 (Pump PEG 01A) V138 (Pump PEG 01B)
(te.cked open manual valves) V135 (Purnp PEG 01C) V141 (Pump PEG 01D)
(Normahy open MOVS) HVI5 (Common) HV16 (Common) 1 Pump aischarge valves V004 (Pump PEG 01A) V013 (Pump PEG 01B)
V008 (Pump PEG 01C) V017 (Pump PEG 010)
Suction check valves VISO (Common) V131 (Common)
Discharge check valves V005 (Pump PEG 01A) V012 (Pump PEG 01B)
V007 (Pump PEG 01C) V016 (Pump PEG 01D) e A-32
'e j ,
,. s .. .
. . Accident - ' importance - Inspection
., Dominant Failure Modes Sequence Category Acthities
- 5. Local fault of manual valves restricting flow to ECCS pump coolers Train A Train B 1 locked open, manual valves:
- 1 Safety injection pump oil cooler: V040 V065 RHR pump seal cooler: V042 V067 4 Centrifugal charging pump oil cooler: V039 V064 1 Inlet header valve for all of the above components V043 V063 l
l j
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A-33
3- ,
l 7 .
1:
WOLF CREEK GENERATING STATION 1- RISK. BASED INSPECTION GUIDE L
Component Cooling Water System (CCWS)
TRAIN A TABLE A.4 2 MODIFIED SYSTEM WALKDOWN 1-l- Desired Actual Pow Sup. Required Actual Descripdon ID No. Location Position Position Breaker # !acation Position Posiden CCW Pump A EG CCWA Locked Suction isolation V132 Open Valve CCW Pump A EG CCWA Locked Discharge Isola. V004 Open tion Valve
, CCW Pump C EG CCWA- Locked Suction Isolation V135 Open Valve l
l l
CCW Pump C EG CCWA Locked Discharge Isola. V008 Open tion Valve 1
CCW Heat Ex. EG CCWA Locked changer A Inlet V019 Open Isolation Valve CCW Heat Ex- EG CCWA Locked changer A Out. V035 Open let Isolation l Valve CCW Heat Ex. EG CCWA Open
- changer A V205 l Temp Bypass l Upstream Isola.
l tion Valve A 34 O
F: y ,
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE J i
Component Cooling Water System (CCWS) l l
TRAIN A l
TABLE A.4 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup, Required Actual Descripdon ID No, location Position Posidon Breaker # Location Position Position CCW Heat Ex. EG CCWA Open changer A V?')6 Temp Bypass Downstream isolation Valve CCW Heat Ex. EF CCWA Locked /
changer A ESW V058 Throttled Return B) pass CCW to Fuel EG FPHXA Locked /
Pool Cooling V200 Throttled /
Isolation HX Open A 625 mms e
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, a 1 i
WOLF CREEK GENERATING STATION j RISK. BASED INSPECTION GUIDE Component Cooling Water System (CCWS)
]
TRAIN B I TABLE A.4 2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow. Sup. Required Actual Descripdon ID No. Location ' Position Position Breaker # Location Position Position 1 l
CCW Purnp B EG CCWB Locked Suction Isolation V138 Open Valve CCW Purnp B EG CCWB Locked Discharge Isola. V013 Open j tion Valve !
CCW Pump D EG CCWB Locked i Suction isolation V141 Open I Valve CCW Pump D EG CCWB Locked Discharge Isola. V017 Open tion Valve CCW Heat Ex. EG CCWB Locked changer B Inlet V044 Open Isolation Valve CCW Heat Ex. EG CCWB Locked changer B Out. V060 Open let Isolation Valve CCW Heat Ex. EG CCWB Open changer B Temp V207 Bypass Up-stream Isolation Valve t
A-36
i u
I
' WOLF CREEK GENERATING STATION l RISK. BASED INSPECTION GUIDE- )
1 Component Cooling Water System (CCWS) l TRAIN B i
TABLE A.4 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. location Position Position Breaker # location Position Position CCW Heat Ex. EG CCWB Open changer B V208 Temp Bypass Dowitstream Isoittion Valve CCW Heat Ex. EF CCWB Open ,
changer B V060 ESW Return Bypass CCW to Fuel EG FPHXB Locked /
Pool Cooling V201 Throttled HX B Open Isolaten 6.25 turns A 37
1,. '
WOLF CREEK GENERATING STATION l
- RISK BASED INSPECTION GUIDE Component Cooling Water System (CCWS) j l
1 l 1
i l - TABLE A,4 2 MODIFIED SYSTEM WALKDOWN (Cont'd) l l- Daired Actual Pow. Sup. Required Actual I
Description ID No. l.ocation ' Position Position Breaker # 1.ecation Position Position l
l CCW Pumps EF Aux. Locked A&C Room V056 Bldg. Open Cooler Ventila.
SGLilA ESW tion ESW Inlet Isolation Train A l l- CCW Pumps EF Locked A&C Room V057 Open -
! Cooler l SGLilA ESW l Outlet Isola. l I tion -
l~ *-
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Room Cooler V013 Throttled l l
t- IIA ESW Re.
l turn isolation 1
I ESW Locked CCW HX1 A EF ESW A Return V058 Train A Throttled HV 59 Bypass l Isolation l
ESW to CCW EG Locked Pumps Train A V182 Throttled Isolation l
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l A-38 i
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f?
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Component Cooling Water System (CCWS)
TABLE A,4 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position CCW Pumps EF ESW - Locked B&D Room V088 Train B Open Cooler SOL 11B ESW Inlet Iso!a'. ion CCW Pumps EF ESW Locked B&D Room V089 Train B Open Cooler SGLilB ESW Outlet Isolation CCW Pump GL ESW Locked Room Cooler V021 Train B Throttled 11B ESW Re.
tura Isolation CCW HX1B EF ESW Locked i ESW B Return V090 Train B Throttled HV 60 Bypass i Isolation ESW to CCW EG ESW Locked l- Pumps Train V18S Train B Throttled B Isolation 1
I l
s A 39
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Component Cooling Water System (CCWS)
?
j l
L TABLE A.4 2 MODIFIED SYSTEM WALKDOWN (Cont'd) l Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position !
l ESW A to EF HIS. Main Open/
CCW HX A' 51 Control Closed l Room l Switch I Line up I Train A ESW A Return EF HIS. Same Open/
A*
I EF HV.$1 NG03CMF1 Auxiliary ON l ESW Supply Building to CCW HX A Train A Isolation EF HV 59 NG03CHF2 Same ON ESW from CCW HX A Isolation
- For NORMAL conditions only one CCW HX is in operation with the other Train HX on reduced flow througn the Return Bypass. Either EF HIS 59 or EF HIS 60 should be OPEN.
l I-l l
l A-40
e ..
WOLF CREEK GENERATING STATION RISK.DASED INSPECTION GUIDE-Component Cooling Water System (CCWS)
TABLE A.4 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Locedon Position Position Breaker # Location Posidon Position ESW B to EF Main Closed /
CCW HX B* HIS $2 Control = Open .
Room Switch Line-up Train B ESW B Return EF Same Closed / l fron, CCW HX HIS 60 Open B*
EF HV 52 NG Auxiliary ON ESW Supply 04CNF3 Building to CCW HX B Train B Isolation EF HV 60 NG Same ON ESW from StCHF2 CCW HX B isolation -
- For NORMAL conditions only one CCW HX is in operation with the other Train HX on reduced flow through the Return Egass. Either EF HIS 59 or EF HIS-60 should be OPEN.
A-41
o .- 1 WOLF CREER GENERATING STATION i CCWS I TABLE A.4 2 (Cont'd)
REFERENCE DOCUMENTS 1 i
L TITLE 1.D. NO. REV DATE I t
. 1
(~. Licensed Operator Initia! Training Lesson )
1.' ' Component Cooling Water System" LO 1400800 000 03/01/88 Drswings
- 1. SNUPPS *P&lD . Component Cooling Water System
- M 12EG01 . 12/09/86
- 2. SNUPPS *P&lD Component Cooling Water System" M 12EG02 1 12/09/86 i
'1
- 3. SNUPPS *P&lD . Component Cooling Water System
- M 02EG03 17 02/28/85 i Procedures
- 1. " Component Cooling Water System Valve, Switch and Breaker CKL EG 120 9 11/09/87 Lineup" l
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j WOLF CREEK GENERATING STATION Table A.51. Importance Basis and Failure Mode Identification )i l
REACTOR PROTECTION SYSTEM (RPS)
Mission Success Criteria The Reactor Protection System (RPS) automatically keeps the reactor operating within a safe region by shutting down the reactor whenever the limits of the region are exceeded (or reached). The system acts to limit the consequences of Condition 11 events (faults of ,
moderate frequency, such as a loss of feedwater) by, at most, a shutdown of the reactor and j turbine, with the plant capable of returning to operation after corrective action. Whenever a direct process or calculated variable reaches a setpoint the reactor will be shutdown in j order to protect against either gross damage to fuel cladding or loss of system integrity ;
which could lead to release of radioactive fission products into the Containment. !
The following systems typically make up the Reactor Protection System.
- a. Process Instrumentation and Control System
- b. Nuclear Instrumentation System .
- c. Solid State Logic Protection System
- d. Reactor Trip Switchgear
- c. Manual Actuation Circuit The RPS contains sensors which, when connected with analog circuitry consisting of two to four redundant channels, monitor various plant parameters. The RPS also contains ,
digital circuitry, consisting of two redundant logic trains, which receive inputs from the analog protection channels to complete the logic necessary to automatically open the reactor trip breakers.
Each of the two trains, A and B, is capable of opening a separate and independent reactor trip breaker, RTA and RTB, respectively and a bypass breaker, BYB and BYA, respectively. The two trip breakers in series connect three phase AC power from the rod drive motor generator sets to the rod drive power cabinets. During plant power operation, a DC undervoltage coil on each reactor trip breaker holds a trip plunger out against its spring, allowing the power to be available at the rod control power supply cabinets. For reactor trip, a loss of DC voltage to the undervoltage coil releases the trip plunger and trips open the breaker. When either of the trip breakers opens, power is interrupted to the rod drive power supply, and the control rods fall, by gravity, into the core. The rods cannot be withdrawn until the trip breakers are manually reset. The trip breakers cannot be reset until the abnormal condition which initiated the trip is corrected. Bypass breakers BYA and BYB are provided to permit testing of the trip breakers.
A 46
p 1
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1 The RPS provides for manual initiation of a reactor trip by the operator, and l automatically ini:iates a reactor trip for the following: i (1) Whenever necessary to prevent fuel damage for an anticipated malfunction l (Condition II). l (2) To limit core damage for infrequent faults (Condition III).
(3) So that the energy generated in the core is compatible with the design provisions to protect the reactor coolant pressure boundary (RCPB) for limiting faults )
(Condition IV).
l The reactor trip system also initates a turbine trip signal whenever a reactor trip is initiated. This is to prevent reactivity insertion that would otherwise result from excessive reactor system cooldown and to avoid unnecessary actuation of the Engineered Safety Features Actuation Systeem (ESFAS).
Accident Importance Inspection Dominant Failure Modes Sequence Category Acthities
- 1. Instrument failure due to calibration / maintenance error, or it 11 S.M.C ,
random failure which inhibits initiation of reactor trip signal
- 2. Reactor trip breaker. 52/RTA or $2/RTB. or trip b> Tass 11 M S.htC breaker. 52/BYA or $2/BYB. fails to open
- 3. Operator failure to manually scram scactor following ATWS 11 L O The WCGS UFSAR states that, pending implementation of equipment from sensor output to the fmal actuation device that is diverse from the reactor trip system which will auto-matically initiate the Aualliary Feedwater System and a tur.
bine trip under conditions indicative of an ATWS:
- a. Emergeney procedures have been developed to train operators to recognise ATWS events, including consideration of scram indicators, rod position indicators, flua monitors, pressurizer level and pressure indicators, pressurizer relief valve and safety indicators, and any other alarms annunciated in the control room, with emphasis on alarms not processed through the electrical portion of the reactor scram system,
- b. Operators have been trained to take actions in the event of an ATWS. includin; consideration of manually scramming the reactot by using the manual scram button. prompt actuation of the suailiary feedwater system to ensure delivery to the full capacity of this system, and initiation of turbine trip.The operators have also been trained to initiate boration by actuating safety injection systems to bring the facility to a safe. shutdown condition.
A 47
1 o e WOLF CREEK GENERATING STATION ,
RISK. BASED INSPECTION GUIDE Reactor Protection System (RPS)
TABLE A.5 2 MODIFIED SYSTEM WALKDOWN The Reactor Protection System is a normally energized system whose operability must be assured by extensive surveillance testing. Observation of the conduct of this testing will provide the inspector with direct input regarding the safety function capability of the system. System walkdown during normal power operation will reveal little regarding the safety function status. However, the following may be checked:
COMPONENT REQUIRED STATUS ACTUAL STATt'S t, Reactor Trip Breakers RTA Closed ,
RTB Closed
- 2. Reactor Trip Bypass Breakers BYA Open j BYB Open l
- 3. Annuneistor Panel - RPS No windows !!!uminated
- 4. RPS Trip Status Panet No bypass lights itiuminated, P 7, P 8, P 10, intermediate range hl flux, low power range hl flux pennissive lights illuminated .
- 5. RPS Permissive and Bypass Status Panel No lights illuminated
- 6. Process Instrument Bistables Mode Switches No RPS channel in test lights illu.
minated A-48
WOLF CREEK GENERATING STATION
, Reactor Protection System (RPS)
TABLE A.5 2 REFERENCE DOCUMENTS TITLE 1.D. NO. REV DATE
- 1. WCCS Updated Safety Analysis Report Section issue 89 05 1989
. 7.2 "Resero Trip System", Section 15.8 ' Anticipated Transients Without Scram"
- 2. WCCS Licensed Operator Training Document LO 1301200 000 03/14/88
- Resetor Protection Systein" i
A 49
i i
WOLF CREEK GENERATING STATION i
Table A.61. Importance Basis and Failure Mode Identification HIGli HEAD INJECTION / SAFETY INJECTION /HIGli IIEAD RECIRCULATION
. Mission Success Criteria l The high head and safety injection systems (HHI/SI) provide core cooling and l negative reactivity addition to the primary systems following small loss of coolant I accidents (LOCAs) where reactor coolant system (RCS) pressure does not reduce suffi- l ciently to permit flow from the low pressure injection system. The high head certrifugal i changing pumps and the intermediate SI pumps provide injection flow up to the shutoff I head of the charging pumps (2500 psig). These charging and SI pumps take suction from i the refueling water storage tank (RWST); the normal suction path from the volume control ,
tank to the charging pumps is automatically isolated based on a high head injection initiation signal.
Major valves in the HHJ flow path include locked open manual valves, check valves, and motor operated valves that are normally closed and open automatically based on a -
safety injection signal. In the safety injection flow path, normally open, motor operated valves are located in the RWST suction line.
During the recirculation mode of operation, the residual heat removal (RHR) pumps supply provide cooled, recirculated water to the SI pumps for injection to the RCS. An automatic switchover to the containment sump for suction to the RHR pumps occurs as the water level is the RWST is reduced to approximately 36%. This level signal,- combined ,
with an active SI signal, causes the two, normally closed, containment sump to RHR pump suction MOVs to open. When these MOVs reach the full open position, a limit switch signals the MOVs in the RWST to RHR pump suction line to close. Thus, the containment sump provides the long term source of water to the RHR pumps. Manual actions required during this sequence include initiation of CCW flow to the RHR heat exchangers to provide cooling for the recirculated flow.
Mission success for the HHI and SI systems is provided by the operation of one of two centrigugal charging, safety injection and residual heat removal pumps during both the injection and recirculation phases of operation.
A 50
O O' !
. c -
i Accident importance laspection ,
Dominant Failure Modes Sequence Category Aethities
- 1. Failure to switch from RWST to the containment sump for the 1 H- O low pressure recirculation system Automatic switchover is .
provided base 4 on low RWST level. MOVs EJ HV $811 NB open and BN HV 8812 A/B close. Operator netion is required j for initiation of CCW flow to the RHR heat eachangers. )
. MOVs EG HV 101/102 must be opened from the Control i Room i
- 2. Failure of HHI discharge valves to open, including common 1 H S.M.T.C cause failures (includes check valves)
D1 HV 8803B (BIT Inlet)
Et HV 8837A Bt HV 8837B s Dt HV 8801A (BIT Outlet) q Bt HV 8801D (BIT O stlet)
Valves EM HV 8837 A/B are modulating solenoid valves that can be used as an alternate charging path Check Valves: BB V 8948A BB V 8948B BB V 894BC BB V 8948D EM HV 8481A EM HV 84BlB ,
- 3. Failure of HPR suction valves to open, including common 1 M S.M,T,C
' cause failure MOVs: El HV 8811A EJ HV 8811B (Normally closed, containment sump to RHR pump suction valves)
- 4. Failure of pump rerum line (miniflow) valve to open fails 1 M S.M.T.C operating pump Centrifugal Charging Purnps: HV 8810 HV 8811 Safety injection Pumps: HV 8814A i HV 8814B l
- 5. Electrical failures (pewer cable / breaker) disable HHR pump 1 M S.M room coolers Circuit Breaker Train A Train B Si Pump Room Cooler $2NG01ABF3 52NG02ACF3 s
A 51 1
e ,
j j
Accident limportance inspection l Dominnet railure Modes Sequente Category Arthitles j
- 6. Failure of Service Water System vahe to open or remain open 1 M S.M.T i disables HHR pump room cooling .
Manual Yahes - Locked Open j TRAIN A EF.V032 51 Pump Room Cooler EF.V033 S1 Pump Room Cooler Return
. EF.V037 RHR Pump Room Cooler i EF.V038 RHR Pump Room Cooler Re:um EF.YO29 Centrifugal Charging Pump ;
Room Cooler !
EF.V030 Centrifugal Charging Pump Room Cooler Return EF.V056 CCW Pump Room Cooler EF.V057 CCW Pump Room Cooler Return TRAIN B EF.V065 $1 Pump Room Cooler EF.V066 St Pump Room Cooler Isolation )
EF.V061 RHR Pump Room Cooler
)
EF.V062 RHR Pump Room Cooler Isolation ]
EF.V068 Centrifugal Charging Pump ,
Room Cooler j EF.V069 Centrifugal Charging Pump Room ]
Cooler Isolation 1
. EF.V088 CCW Pump Room Cooler EF.V089 CCW Pump Room Cooler Return
- 7. Local fault of purnps/ pumps fail to start or run 1 M 5,M,T.C
. Safety injection Pumps: PEM01A PEM01B RHR Pumps: PD01A PD01B Centrifugal Charging Pumps: PBG05A PBG05B
- 8. Failure of valve to open in the common portion of the HHI 1 M S.M,TC suction line from the RWST MOVs BN 14V 112 D/E arr the RWST to centrifugal charg.
ing pump suction supply valves and open based on an SI signal.
- 9. Plugging of manual valve in the HRI and SI suction line (or in 1 M 5,M the containment sump strainers)
MOVs E,M HV.8924, EM HV 8807A and EM HV 8807B provide recirculation flow for both HHI and 51.
A 52
-c .
Accident importance Inspection Dominant Failure Modes Sequence Category Activities
- 10. HRI and Si pump return line (miniflow) valve fails to close: 1 L S.M.T.C l interlock prevents HPR suction valves from opening
]
$1 Pump: Normally open MOVs operated l from the Main Control Room l EM HV 8814A I' EM HV 8814B BN HV 8813 Centrifugal Charging Pump: MOVs that cycle open and closed for flows between 174 )
gpm and 259 gpm. i
- 11. Local pump failures ! L S.M j
- failure of control cable to MCC '
- failure of pump breaker to close Power Source ESFRM1 Centrifugal Charging Pump 1A ;
ESFRM2 Centrifugal Charging Pump 1B ;
ESFRM) Safety injection Pump 1A ,
ESFRM2 Safety injection Pump 1B ESFRMI RHR Pump 1A ESFRM2 RHR Pump 1B
- 12. Pump in maintenance 1 L M Two CCP. $1, or RHR pumps should not be simultaneously out for maintenance l
A 53 l
o .
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE j High Head Injection / Safety Injection /High Head Recirculation
)
1 TABLE A.6 2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow, Sup. Required Actual Description ID No. Location Position Position Breaker 8 Location Position Position
$1 Pump B Dis. EM. Locked charge Isolation 8921B Open 1
I
$1 Pump A Dis. EM. Locked !
charge Isolation 8921A Open l
$1 Pumps to EM. Locked Accum Injection VD98 Throttled Cold Leg 4 Throttle ,
$! Pumps to EM. Locked Accum Injection V095 Throttled Cold Leg 1 Throttle St Pump B to EM. 14cked RCS Hot Leg 4 V092 Throttled Throttle <
$1 Pump B to EM. Locked RCS Hot bg 1 V091 Throttled Throttle
)
A 54
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE :
High Head Injection / Safety Injection /High Head Recirculation TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker a Location Position Position Boron Inj. to EM. Locked l RCS Cold Leg V110 Throtded 4 Throttle Boron inj. to EM. Locked RCS Cold leg V107 Throttled 1 Throttle
$1 Pumps to EM. Locked '
Accum injection V097 Throttled Cold Leg 3 Throttle
$1 Pumps to EM. Locked Accum injection V096 Throttled Cold Leg 2 Throttle
$1 Pump A to EM. Locked RCS Hot Leg 3 V090 Throttsed Throttle 51 Pump A to EM. locked RCS Hot Les 2 V069 Throttled Throttle I
l l
l l
A 55
..c -.r -
O *
. . l WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE High Head Injection / Safety Injection /High Head Recirculation l
TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd) ,
i Desired Actual Pow. Sup. Required Actual I Descripdon ID No. location Position Posiden Breaker # 1ecation Position Position I i
RCS Cold leg V109 Throttled l 3 hrottle l
Boron inj. to EM. lacked RCS Cold leg V108 Throttled I 2 Throttle 1
RWST Outlet BN. Locked l
Isolation V011 Open i Valve
)
A-56
. o !
. o WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE High Head Injection / Safety Injection /High Head Recirculation j TABl E A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd) 4 Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # location Position Position CCW Trat A EG V414 CCWA Open/ ;
Supply isolation Locked ,
Valve to Post Closed' Accident Sam-pling Coolers
~
CCW Train A EG V416 CCWA Open/
Return isolation Locked Valve to Post Closed' Accident Sam-pling Coolers RHR HX A EJ V033 CCWA Locked CCW Outlet Throttled Isolation Valve CCW A to SIP / EG V038 2026AB locked RHR Pump /CCP Open Coolers isolation Valve !
CCW to $1 EG.V040 SIA locked Pump 1 A Oil- Open Cooler Isolation Valve St Pump 1A EM- $1A locked Oil. Cooler Out. V099 Throttled l let Isolation 0.4 turns Valve open CCW to RHR EG VD42 RHRA locked Pump Seal Open Cooler IA Iso-lation Valve
' Idle train must be locked closed l
l A 57
.-. ,r
O ' ?
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE High Head Injection / Safety Injection /High Head Recirculation TABLE A 6 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position RHR Pump A EJ.V070 RHRA Locked CCW Return Throttled Isolation Valve CCW to CCP EG.V039 CCPA Locked Oil Cooler $A Open isolation Valve CCP A CCW BG. CCPA Locked Return Isolation Y259 Throttled Valve 1.4 turns open SIP /RHR/CCP EG V043 2026AB Locked Coolers Return Open isolation Valve CCW Pumps EF VO$6 CCWA Locked A&C Rwm Opn Cooler SGL11A ESW Inlet Iso.
lation CCW Pumps EF.V057 CCWA Locked A&C Room Open ,
Cooler SGL11A ESW Outlet Isolation CCW Train B EG V413 CCWB Operv Supply isolation Locked Valve to Post Closed
- Accident Sam, pling Coohrs CCW Train B EG V415 CCWB Operd Return Isolation Locked Valve to Post Closed
- Accident Sam-pling Coolers
' Idle train must be locked closed A 58
c .
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE High Head Injection / Safety Injection /High Head Recirculation TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Ar.tual Pow. Sup. Required Actual Description ID No. Location Position Position Breake # tocation Position Position
- RHR HX B EJ.YO38 CCWB LocLed CCW Outlet Throttled Isolation Valve Train B to $1P/ EG.YO63 2026AB Locked RHR Pump /CCP Open Coolen Isolation Valve
$1 Pump 1B EG YO65 SIB Locked Oil-Cooler Isola. Open tion Valve St Pump IB EG V103 SIB Locked Oil Cooler Out. Throttled let isolation 0.45 tums Valve open CCW to RHR EG.YO67 RHRB tocked Pump Seal Open +
Cooler IB !so.
lation Valve 1
l l
l l
A 59 l
I WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE High Head Injection / Safety injection /High Head Recirculation TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actud Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position RHR Pwnp B EJ.V071 RHRB Locked CCW Return Throttled .
Isolation Valva 0.2 twns open CCW to CCP EG.V0H CCPB tocked Oil Coole $B Open .
Isolation Valve CCP A CCW BO. CCPB tocked Return Isolation V268 *ihtottled Valve SIP /RHR/CCP EG.V068 2026AB Locked Coolers CCWB Open Return isolation Valve A 60
l 1 . .
r WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE High Head Injection / Safety Injection /High Head Recirculation TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Descripdon ID No. Locadon Position Position Breaker # Location Position Position E$W A to EF V029 CCPA Locked Cent. Charging Open Pump Room Cooler 12A !so.
lation Cent. Charging EF.YO30 CCPA Locked Pump Room Open Cooler 12A ESW Rerum Isolation ESW A to $1 EF V032 $1A Locked P. imp Room Open Cooler 9A Iso-lation
$1 Pump Room EF.V033 51A locked Cooler 9A ESW Open A Return Isola-tion -
ESW A to RHR EF.V037 RHRA Locked Pump Room Open Cooler 10A iso-lation RHR Pump EF V038 RHRA lacked Room Cooler Open 10A ESW A Return Isolation t
A 61
O 8 WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE ,
1 High Head Injection / Safety Injection /High Head Recirculation i TABl.E A 6 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position j l
RHR Pump GL V010 Availiary Locked Room Cooler Building Throtded .
10A ESW Re- Ventilation ;
tum Isolation ESW Train A ;
S1 Pump Room GL V009 Same Locked Cooler 9A ESW Throtded l Retum Isolation Cent. Charging GL.V008 Same Lacked Pump Room Throtded Cooler 12A ESW Rerum Isolation i
Cent. Charging GL V019 Availiary Locked i Pump Room Building Throttled
- Cooler 12B Ventilation ,
! solation Train B St Pump Room GL V018 Same Locked Cooler 9B ESW Throttled Return Isolation RHR Pump GL.V017 Same Locked Room Cooler Throttled 10B ESW Re.
tum isolation t
A 62 e
- t
. o WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE !
l High Head Injection / Safety Injection /High Head Recirculation ,
TABLE A.6 2 MODIFIED SYSTEM WALXDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. LocaGon Position Position Breaker # Location Position Position :
ESW B to RHR EF YO61 RHRB Locked :
Pump Room '
Open Cooler 10B 1so-lation RHR Pump EF V062 RHRB Locked Room Cooler Open 10B ESW B Return isolation ESW B to EF.V068 CCPB Locked Cent. Charging Open
^
Pump Room Cooler 12B !so-lation Cent. Charging EF.V069 CCPB Locked Pump Room Throttled Cooler 12B ESW B Return Isolation ESW B to SI EF V06S SIB lacked Pump Room Open Cooler 9B !so-lation 51 Pump Room EF.V066 SIB Locked '
Cooler 9B ESW Open B Rerum Isola-tion A 63
~ --
-- - ' w r*+-
. . l
{
l WOLF CREEK GENERATING STATION (ESWS) e TABLE A.6 2 (Cont'd)
REFERENCE DOCUMENTS l TITLE I.D. NO. REV DATE
. Licensed Operator Initial Training lessons:
- 1. " Essential Service Water System
- LO 1408900 000 02/29/88
- 2.
- Service Water System
- LO 1407601 001 02/2$/87
- 3.
- Emergency Core Cooling System and Safety injection
- LO 1300600 000 01/28/88 !
- 4.
- Residual Heat Removal System
- LO 1300$00 001 08/10/88 Drawings
- 1. SNUPPS *P&!D - Essential Service Water System
- M 12EF01 0 12/09/86
- 2. SNUPPS *P&lD - Essential Service Water System
- M 12EF02 0 12/09/86
- 3. $NUPPS *P&lD - Service Water System
- M.12EA01 0 08/07/84
- 4. SNUPPS *P&lD - Service Water System
- M 12EA02 0 S. SNUPPS *P&!D Chemical and Volume Control System
- M 12BG03 3 8/10/87
- 6. SNUPPS *PilD Chemical and Volume Control System
- M.12BGOS 1 7/14/87
- 7. SNUPPS *P&lD . Residual Heat Removal System
- M.12EJ01 1 7/14/87 Procedures
- 1.
- Essential Service Water Valve Breaker and Switch Lineup" CKL.EF 120 12 02/10/89
- 2.
- Component Cooling Water System Valve, Breaker and Switch CKL.EG.120 9 11/09/87 Lineup
- 3.
- Chemical and Volume Control System Normal Valve Lineup
- CKL.BG 120 12 01/08/89
- 4.
- Chemical and Volume Control System Switch and Breaker CKL.BG.130 8 02/10/89 Lineup"
$.
- Safety injection System Lineup Checklists
- CKL.EM.120 7 04/20/88
- 6. *RHR Normal System Lineup" CKL.EJ 120 9 08/13/88 l
A 64
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A-66
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i A 67
WOLF CREEK GENERATING STATION Table A 7-1. Importance Basis and Failure Mode Identification !
1 l
PRIMARY PRESSURE RELIEF SYSTEM l
Mission Success Criteria i
The primary pressure relief system (PPRS) provides protection from i overpressurization of the primary system to ensure that primary integrity is maintained.
The PPRS also provides the means to reduce the RCS pressure if necessary. The PPRS is composed of three code safety relief valves (SRV) and two power operated relief valves (PORVs). The code safety valves are important only for ATWS scenarios. The PORVs provide RCS pressure relief at a set point below the SRVs. The PORVs discharge to the pressurizer relief tank. Each PORV is provided with a motor operated block valve. The i PORVs automatically open on high RCS pressure or are manually opened at the discretion of the operator. The block valves are normally open unless a PORY is leaking.
The PPRS is dependent on the AC power buses for motive and control power to the PORV block valves, vital AC power for control power to the PORVs, and the containment air system for motive power to the PORVs. However, the PORVs are provided with [ air / ,
nitrogen) bottles sized to provide approximately [ ] openings of each valve. '
The success criteria for the PPRS vary depending on the application. The success criterion for the PPRS following a transient event demanding PORV opening is that the PORVs successfully reclose. The success criterion for the PPRS following a transient and failure of the AFWS is that both PORVs successfully open on demand. The success criterion fot the PPRS following a small LOCA with failure of the AFWS and for the '
support system function provided to HHI in the emergency boration mode is that [one or more) PORVs successfully open on demand. The success criterion for ATWS is that [3 SRVs or 2 SRVs and 2 PORVs open).3 Accident Importance Inspection Dominent Failure Modes Sequence Category Activities
$RVs 8010 A.B.C
- 3. PORY block valve closed 7 M O.M HV-8000A. HV.8000B 3
BNL does not currently possess the appropriate data sources to complete this information.
A 68
o .o l
Accident importance Inspection i Dominant Failure Modes Sequence Category Aethllles ;
i 4 Operator error in bleed a feed activities comes lack of RCS 6 M O l cooling j (No specific emergency procedure for bleed and feed appears to exist for WCGS) i I
i i
+
F A 69
a WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE !
l Primary Pressure Relief System l l
TABLE A.7 2 MODIFIED SYSTEM WALKDOWN )
Desired Actual Pow. Sup. Required Actual {
Description ID No. Locanon Position Position Breaker # Location Position Position J
Pressurizer PCV- Contain. Closed Closed _
PORY 455A ment El l
Pressurizer PCV. Closed Closed ,
PORY 456A l
Pressurizer SRV. Not Cagged Safety Relief 8010A Valve l
Pressuriter SRV. Not Cagged I Safety Relief 8010B l Valve )
l Pressurizer SRY. Not Gagged Safety Relief 8010C Valve Pressurizer HV. Open PORV Block 8000A Valve (PCV.
455A)
Pressurizer HV. Open PORY Block 8000B Valve (PCV.
456A)
A 70 -
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Primary Pressure Relief System TABLE A.7 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual - Pow. Sup- Required Actual -
Description . 19 No. Location Position Posidon Breaker # location Position Position PZR PORY BB BB HIS. Main Closed / Auto PCV 455A 455A Control Board RLO21 PZR PORY BB BB HIS- Same Closed / Auto PCV-456A 456A BB HV 8000A BB HIS. Same Open PZR Power Re. 8000A lief PCV-455A Inlet BB HV 8000B BB HIS. Same Open l; PZR Power Re. 8000B lief PCV 456A Inlet' PZR Relief Iso. BB HIS- Same Arm l lation Valves 8000A ll (BB HV 8000A I'
and BB PCV-455A) l PZR Relief Iso. BB HIS. Same Arm l !ation Valves 8000B l (BB HV 8000B and BB PCV.
456B)
F2R Relief Iso. BB HIS- Main Normal lation Valve 8000C Control (BB HV 8000A) Board
- l. NGO) i; PZR Relief Iso- BB HIS. Main Normal l
!ation Valve 8000D Control l (BB HV 8000B) Board NG02 A-71
WOLF CREEK GENERATING STATION WCGS
-PPRS TABLE A.7 2 (Cont'd)
REFERENCE DOCUMENTS TITLE I.D. NO, REV DATE Documents <
- 1. P. Saylor and P. Lobner (ed.). " Nuclear Power Plant S/JC 88/1996 1 February 1989 System Sourcebook . Wolf Creek 50-482." Science Applications International Corp.
Procedures j
- 1. Reactor Coolant System Lineup CKL.BB.110 l 8 02/10/89 l
1 j
4 I
A-72
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WOLF CREEK GENERATING STATION I
Table A.81. Importance Basis and Failure Mode Identification AUXILIARY FEEDWATER SYSTEM Mission Success Criteria The Auxiliary Feedwater (AFW) system provides feedwater to the steam generators (SGs) to allow continued heat removal from the primary system when main feedwater is unavailable. In this capacity the AFW system serves as one of the means to perform the
, safety function of early core heat removal following a transient or small LOCA.
The AFW system is a three train system which consists of two motor driven pumps (MDPs) and one turbine driven pump (TDP). Each MDP discharges to two of the four SGs.
The TDP is twice the capacity of the MD pumps and discharges to all of the SGs. Each pump takes suction through a common header from the Condensate Storage Tank (CST) or from an Essential Service Water header. The CST has a capacity of approximately 466,200 gallons. The Technical Specifications require the CST to be operational with at least 281, 000 gallons of water. Each flow path from an AFW pump discharge to a SG has two check valves in series and a normally open air operated valve.
The two MDPs start automatically on receipt of an AFW actuation signal (AFAS).
This signal is generated in response to any of the following conditions: SG water level low low, presence of the ESF signal, station blackout, or trip of main feedwater pumps.
The same signal causes the TDP throttle / trip valve to open automatically starting the TDP.
.In the event that low AFW pump suction pressure is sensed, indicating faults in the condensate storage tank suction lines, suction is automatically switched to the ESW headers' by opening the ESW header isolation valves.
' The AFW system depends on AC power for motive power to MDP motors and for control power to AONs, DC power for control power to MDP's, TD pump and the associated- air operated discharge valves, and AFAS for automatic actuation.
In addition to the dependencies listed above, the AFW system also interfaces with the instrument air system, and HVAC, and SWS HVAC provides room cooling for the TDP and instrument air is provided to the TDP discharge valves.
A-74
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- 4 Accident importance Inspection Dominant Failure Modes Sequence Category Activities
- 1. Failure to manually start locked out standby pump 7,10,6 H O Accordmg to WCGS training document LO 1406100 in some emergency procedures, upon a SI signal, the operator is instructed to p' ace the motor driven AFW pumps in the pull-to lock position until power is restored to at least one safe.
guards bus.
. 2. Local fault of valve in turbine driven pump dischars,e to 6,8,7 H S.M T steam generators.
Inadvertent closure of locked open manual valve V055 pre-vents AFW flow to all four steam generators from the TDP,
- 3. Failure to manually start pump given auto. start failure 10,6,7 H O MDP PALO1A [MDP PALOlB] can be manually staned by the Contrul Roorn handswitch AL HIS 23A (RLOOS), [AL HIS 22A (RLOO5)), the handswitch at the Auxiliary Shutdown Panel (ASP) AL HIS 231, [AL HIS 22B (RP118)], and locally '
at the NB01 bus [ ].
TDP PALO2 is manually started by depressing the Actuate push button on Main Control Board Panet RL018 wluch in turn opens the three steam supply valves ABHV 5, ALHV-6, r
and FCHV 312.
- 4. Turbine driven pump PALO2 fails to start or run 10,8,7,6 H 5,M,T.C Several possible failure mechanisms can lead to failure to start or run, e.g. hardware failure of pump or turbine
- 5. Motor driven pump PALOI A or PALOlB fails to start or run 6,10,7 H S.M T.C Similarly, as in 3 above, failure to start or run can be caused by pump or motor hardware faults, etc.
- 6. Local fault of valve in motor driven pump discharge to steam 6,7 H S.M.T generator inadvertent closure of locked open manual valve V045 pre-vents flow to Steam Generators B and C from MDP PALOl A.
Similarly, inadvertent closure of locked open manual valve V031 prevents flow to Steam Generators A and D from MDP PALOlB.
- 7. Turbine driven pump PALO2in maintenance 10,7,6 H M The Wolf Creek Tech. Specs, limit the allowed outage time of one AFW pump to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 8. Steam supply valve or throttle / trip valve faib to open (or 10.7,6 H S.M T.C other valve faults in steam admission line) for turbine driven pump Steam for the TDP is supplied by normally closed air operated ABHV $ from SG B and ABHV 6 from SG C, and through the normally closed trip and throttle valve FCHV 312. Failure of ABHV 5 and ABHV 6 to open or of FCHV 312 to open prevents operation of the TDP.
A-75
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Accident . Importance Inspection - 1 Dominant Failure Modes Sequence Category Activities -
9.' Local fault of suction valve from the condensate storage tank 10,7 M- O.S.M -
(CST) . ,
Inadvenent closure of locked open manual valve V15 blocks all now from the CST to the AFW pumps, forcing reliance !
' upon the ESWS as a suction source
- 10. AFW Gow control valve in maintenance fails delivery from 7,6,10 M M TD pump i The TDP. discharges to the SGS through four air operated discharge valves:
I 9 . Valve No. SG ALHV8 A ALHV10 ! B :
ALHV12 C ALHV6 D Maintenance on any of the above prevents flow to the respec.
tive SG.
' 11. Undetected flow diversion 7 M O
. Inadvertent closure oflocked open manual valves V031 V045 l or V055 described in 2 and 6 above diverts all flow back to I
- the CST through the mini-flow line.
- 12. . Undetected FW leakage back through pump discharge vaJves 7 .M O causes steam binding
.WCGS training document LO1406100 alerts the operator to i this condition and indicates that it can be desacted by routine l
. temperature monitoring via the plant computer and manual local testing. :The condition can be cleared by venting and runnmg the.affected pump.
- 13. Local fault of motor. driven pump power breaker 10,7 M S,M .
(See AC power system)
- 14. Turbine driven pump in test - 10,6 L S
~ 3ach AFW pump must be tested every 31 days according to ,
the WCGS Tech Specs.
/
1 A-76 3
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7 , .. , i
.j Accident Importance Inspection l Dominant Fallure Modes Sequence Category Aethitles 1; i
- 15. Local fauk of AFW actuation signal logic fails to actuate MD 10 L S,C 1 pump and/or TD pump steam valves 1 1
The MDPs are automatically actuated upon any one of the ;
following signals:
a) 2 of 4 low. low water levels in any one steam generator.
b) 51 signal. l
. c) Less of offsite power and station normal auxiliary power.
d) Loss of both MFW pumps.
The TDP automatically starts upon either of the following '
I signals- l l
a) 2 of 4 low. low water levels in any 2 of 4 steam generators.
l l
b) Loss of offsite power and station normal auxihary power, l
l- 16. Failure to restore TD pump from testing 6 L O i l l The TDP is subject to penodic testing under procedure I 17, Failure to restore TD pump discharge valve V055 after test 6 L O l
Locked open manual valve V055 does not have control room I I
position indication. i l
L- 18. Failure to manually open TD pump discharge AOVs ALFIV 6, L 8 O l l' 8,10,12. '
1 Each valve has a 25 cu. ft. N2accumulator to backup the air l supply should it be lost. '
- 19. MD pump PAL Ol A or OlB in maintenance. 10 L M t-1 1 As in 10 above, each pump may be inoperable up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, i
sccording to the Tech Specs.
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A-77 l l
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WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE i Auxiliary Feedwater System TABLE A.3 2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position CST Supply to AP.V015 Aux. Locked AFW Pumps Bldg. Open Suction Isolation (AB) 014 Valve AFW Pumps AP.V001 A.B. 024 Locked Miniflow to Open CST Isolation Valve TD AFW Pump AL.V011 A.B. 125 Locked Suction from Open ESW Train A Isolation TD AFW Pump AL.V014 A.B. 125 Locked Suction from Open ESW Train B Isolation l
l- MD AFW Pump Al V008 A.B.125 Locked !
(' A Suction from Open - l
'. ESWS MD AFW Pump Al V005 A.B.125 Locked B Suction from Open ESWS i
MD AFW Pump AL.V040 A.B. 135 Locked l A Miniflow to Open l CST Isolation l
{ l
[ I l MD AFW Pump Al V028 A.B. 135 Locked-l B Miniflow to Open CST Isolation 1
i 1
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- j. A-78 L
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WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE j Auxiliary FeedWater System I
TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker s Location Position Position MD AFW Pump AL V02) Aux. Open L A Suction Pres. Bldg.(A.
sure Transmitter B.) 135 AL IT 25 Isola-tion MD AFW Pump AL V018 A.B.135 Open B Suction Pres-sure Transmitter AL PT 24 Isola-tion ,
TD AFW Pump AL V024 A.B. 135 Open Suction Pressure Transmitter AL PT 26 Isolation TD AFW Pump AL V055 Aux. Locked Disch Isolation , Bldg. Open (A.B.)
135 TD AFW Pump AL V052 A.B.135 Locked Miniflow Open Recire, to CST
! solation MD AFW Pump AL VN3 A.B. 135 Locked A Discharge Open Isolation MD AFW Pump AL V047 A.B.135 Locked A Discharge to Open SGB HV 9 Inlet Isolation MD AFW Pump AL V049 A.B.135 Locked A Discharge to Open SGB Header Isolation MD AFW Pump ALVM4 A B.135 Locked A Discharge to Open SGC HV Il in.
let Isolation A-79
.e .
l WOLF CREEK GENERATING STATION :
RISK BASED INSPECTION GUIDE '1 1
Auxiliary Feedwater System l l
l TABI.E A.8 2 ' MODIFIED SYSTEM WALKDOWN (Cont'd) )
Desired Actual Pow. Sup. Required Actual 1 Description ID No. location Position Position Breaker # Location Position Position MD AFW Pump &V046 A.B.135 Locked l A Discharge to Open SGC Header Isolation MD AFW Pump LV031 Aux. Locked B Discharge Bldg. Open isolation - (A.B.)
135 !
l MD AFW Pump AL.V035 A.B. 135 Locked i B Discharge to Open SGD HV 5 Inlet Isolation MD AFW Pump AL V037 A.B.135 Locked B Discharge to Open SGD Header Isolation MD AFW Pump & V032 A.B. 135 Locked B Discharge to Open SGA HV 7 Inlet isolation MD AFW Pump & V034 A.B. 135 Locked B Discharge to Open SGA Header Isolation TD AFW Pump AL-V056 A.B. 135 Locked B Discharge to Open SGA HV 8 Inlet Isolation TD AFW Pump AL V058 Aux. Locked Discharge to Bldg. Open SGA Header (A.B.)
Isolation 135 TD AFW Pump &V061 A.B. 135 Locked Discharge to Open SGD HV 6 Inlet Isolation l '.
A-80
WOLF CREEK GENERATING STATION RISK. BASED' INSPECTION GUIDE m Auxiliary Feedwater System TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # Location Position Position TD AFW Pump AL.V063 A.B. 135 Locked Discharge to Open SGD Header Isolation TD AFW Pump AL.V066 A.B. 135 Locked Discharge to Open _
SGB HV.10 In.
let Isolation TD AFW Pump AL.V068 A.B. 135 Locked Discharge to Open SGB Header Isolation TD AFW Pump AL.V071 A.B.135 Locked Discharge to Open SGC HV.12 In.
let Isolation TD AFW Pump AL.V073 Aux. Locked Discharge to Bldg. Open SG.C Header (A.B.)
Isolation 135 Main Steam AB. A.B. 145 Locked Loop 2 to V085 Open AFWP Turbine HV 5 Inlet Iso.
lation Main Steam AB. A.B.145 Locked loop 3 to V087 Open AFWP Turbine HV.6 Inlet Iso.
lation TD AFW Pump AL AFW D Locked Discharge to HV.6 Valve Neutral SGD isolation Room TD AFW Pump AL AFW A Locked Discharge H V.8 Valve Neutral Header to SGA Room Isolation A-81
e .
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Auxiliary Feedwater System TABLE A.8 2. MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup, Required Actual Description ID No. Location Position Posidon Breaker # Location Position Position TD AFW Pump AL AFW B Locked Discharge HV 10 Valve Neutral Header to SGB Room Isolation TD AFW Pump AL AFW C locked Discharge HV 12 Valve Neutral Header to SGC Room isolation Main Steam AB Main Locked loop 2 to HV5 Stream Neutral AFWP Turbine Tunnel isolation BC' Main Steam AB Same Locked loop 3 to HV-6 Neu*:al AFWP Turbine Isolation ESW A to Aux EF V047 MD AFW Locked FW Pump A Pump Open Room Cooler Room 2A Isolation Aux FW Pump EF V048 Same Locked Room Cooler Open 2A ESW Return isolation ESW B to Aux EF V077 MD AFW Locked FW Pump B Pump Open Room Cooler Room 2B Isolation Aux FW Pump EF V078 Same Locked Room Cooler Open 2B ESW B Re.
turn Isolation A-82
WOLF CREEK GENERATING STATION .
RISK. BASED INSPECTION GUIDE Auxiliary Feedwater System TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. locadon Position Posiden Breaker # location Position Position AL HV 30 52NG Closed ESW to MD NCCF3 AFW Purnp B AL HV 33 TD 52NG Closed AFWP ESW 04CCF4 Train B Suc6on isolation AL HV 34 Con- $2NG Closed densate Storage 04CNF1 to MD AFWP B
AL HV 31 52NG Closed ESW to MD 03CCF3 AFWP A
,- AL HV 32 TD 52NG Closed AFWP ESW 03CCF4 Train A Suction Isolation AL HV 35 Con- $2NG Closed densate Storage 03CEF3 to MD AFWP A
AL HV 35 Con- 52NG Closed densate Storage 03CEF4 to TD AFWP 1
AFWP B Dis- CLFt15 chuge Header l Lo SG D isola-
tion -
A-83
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I WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Auxiliary Feedwater System TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow Sup. Required Actual Description ID No. Location Position Position Breaker # location Position Position AL HV.9 MD NG Closed AFWP A Dis. OlBAR114 charge Header to SG B Isola.
tion -
AL HV.11 MD NG Closed AFWP A Dis. OlBAR115 charge Header to SG C isola.
tion Panel RP053CC PN0823 Closed Panel RP053DB NNO307 Closed Panel RP053EB PN0819 Closed Panel RP053EA PN0716 Closed Panel RP053CD PN0723 - Closed Panet RP053BC NN0418 Closed Panet RP053BC NN0416 Closed t
A-84
t' ,a WOLF CREEK GENERATING STATION
, RISK. BASED INSPECTION GUIDE Auxillary Feedwater System TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. kcation Position Position Breaker # Location P9sition Position Panel RP053AC NN0120 Closed Panel RP053AC NN0ll6 Closed Panel RP053DA NN0208 Closed Aux Relay Rack NK4419 Closed RP335 AFW Pump 152NB0105 Racked DPALOIA Up Breaker AFW Pump 152NB0205 Racked DPALOlB Up Breaker MD AFWP - NK4101 On DPALOIA Con.
trol Power (Via.
SWGR NB01)
MD AFWP NK4401 On DPALOlB Con.
trol Power (Via. .
SWGR hT02)
A-85
l WOLF CREEK GENERATING STATION
, RISK BASED INSPECTION GUIDE Auxiliary Feedwater System ;
l~
l: TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. location Position Position Breaker # location Position Position 1x .
AL SG D Control Open MDAFWP B HK5A Room I Control Valve RLOO6
SG D TDAFWP AL C.R. Open Control Valve HK 6A RLOO6 AL HV 6 l
SG A ~ AL C.R. Open MDAFWP B HK 7A RLOO6 l Control Valve AL HV 7 1
^
Control Valve HK 8A RLOO6 AL HV 8 1 l SG B AL C.R. Open l l MDAFWP A HK.9A RLOO6 Control Valve AL HV 9 l-SG B TDAFWP AL C.R. Open Control Valve K 10A RLOO6 AL HV 10 l-SG C AL C.R. Open
)
MDAFWP A HKllA RLOO6 l Control Valve AL HV ll' l
, Control Valve HK 12A RLOO6 l AL HV 12 ESW to MD AL C.R. Closed AFWP B Valve HIS-30A RLOO5 L AL HV.30 l
A-86 l
l
WOLF CREEK GENERATING STATION '
RISK BASED INSPECTION GUIDE .
Auxiliary Feedwater System TABLE A.8 2 MODIFIED SYSTEM WALNDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # tocation Position Position ESW to MD AL C.R. Closed AFWP A Valve HIS.31 A RLOO5 AL HV.31 ESW to TD AL C.R. Closed ,
AFWP Valve HIS.32A RLOO5 AL HV.32 ESW to TD AL C.R. Closed AFWP Valve HIS.33A RL005 -
AFWP B Valve ES.34A RLOO5 AL HV.34 CST to MD AL C.R. Open t, AFWP A Valve HIS.35A RLOO5 l- AL HV.35 CST to TD AL C.R. Open AFWP Valve HIS 36A RLOO5 AL HV.36 l
1 I
l A 87
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a WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE le _-
Auxiliary Feedwater System TABLE A,8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. location Position Position Breaker # Locatian Position Position Loop 3 Sicam AB Control Closed to AFWP Tur- HIS 6A Room bine Valve AB R1h05 HV.6 Loop 2 Steam AB C.R. Closed to AFWP Tur. HIS.5A RLOO5 bine Valve AB HV 5 loop 2 Warmup AB C.R. Closed Steam to AFWP HIS48 RLOO5 Turbine AB HV4B Loop 3 Wannup AB C.R. Ckned Steam to AFWP HIS49 RLOO5 Turbine AB HV49 AFWP Turbine FC C.R. Closed Bypass Trap to HIS.10 RLOO5 Cond. FC LV.
10 AFWP Trap FC C.R. Open Isol. Valve FC HIS 310 RLOO5 FV.310 AFWP Turbine FC C.R. Closed . .
Mech. Trip / HIS. RLOO5 Throttle Valve 312A FC HV 312 AFWP Turbine FC C.R. 3850 RPM Speed Govenor HIK. RLOOS Control 313A AFWP PALOlA l AL iC.R. Normal Motor Control His.23A RLDOS , Aftet Switch ' Stop _
A-88
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- WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Auxiliary Feedwater System TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd) r Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Posidon Breaker # beation Position Position AFWP PALOlB AL Control Normal Motor Control }US.22A Room After Switch Rt.D05 Stop Control Room RP Remote Normal Isolate Switch HIS.1 Shutdown Panel Control Room RP Remote Normal s isolate Switch HIS 2 Shutdowm Panel A-89 I
, WOLF CREEK GENERATING STATION
-AFWS TABLE A.8 2 (Cont'd)
REFERENCE DOCUMENTS TITLE 1.D. NO. REV DATE Documents
- 1. Wolf Creek Generating Station Technical Specifications as in effect February 1988. " Condensate Storage Tank," Par. 3.7.1.4 Licensed Operator Initial Training Lessons:
- 1. WCCS " Licensed Operator Training Document - Auxiliary LO1406100 001 12/29/87 Feedwater" Procedures
- 1. WCGS " Auxiliary Feedwater Normal Lineup" CKL.AL-120 11 01/08/89 Drawings P&lD - Auxiliary Feedwater System M.12AL01(Q) 0 04/27/85 -
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WOLF CREEK GENERATING STATION :
Table A 91, Importance Basis and Failure Mode Identification LOW HEAD INJECTION (LHI)/ LOW HEAD RECIRCULATION (LHR)
Mission Success Criteria The LHl/LHR system injects borated water from the Refueling Water Storage Tank (RWST)into the RCS to provide core cooling water during the injection phase of a large l
break LOCA. '
Four accumulators are available to flood the core with borated water immediately l following a large break LOCA. They are designed to minimize core damage until the safety injection pumps can provide adequate water for core cooling. Each tank is pressurized with nitrogen at 650 psig and contains a minimum water volume of 850 ft' with a minimum l boron concentration of 2000 ppm.
The accumulators are self-contained, self actuating, and passive in nature. Each tank l
is connected to the RCS at one of the. reactor inlets (cold legs). Two check valves, held closed by RCS pressure, provide isolation during normal operation. The' tanks can be isolated by motor operated valves during plant shutdown and depressurization. The accu- 1
_l mulators are not dependent on any support systems. Three of the four tanks provide sufficient water to cover the core _ following a Design Basis Accident (DBA), assuming the contents of one of the four tanks spilled through the break.
The LHI/LHR system can be aligned to take suction from the containment sump and maintain a borated water cover over the reactor core for extended periods of time in the recirculation phase. Manual startup of the CCW system is required to provide cooling to the RHR heat exchangers thereby cooling the recirculation flow.
The LHI/LHR system consists of two pumps taking suction from separate RWST discharge headers which discharge into cross-connected headers. Upon receipt of a SIAS, the two pumps will start and the injection line motor-operated valves will open. When RCS pressure drops below 600 psig, the LHI/LHR will begin to deliver flow to the cold legs.
Mission success is accomplished by operation of one LHI pump following a DBA.
t l-A 92
y ,
[1 )*' .s-L tec
(:
" importance Inspection .
N . Accident Dominant Failure Modes Sequence Category Activities
- 1. Accumulator failure including common inode check valve 3 H- S.M T c_ failure or plugging of MOVs
~
Normally open (wlth breakers racked out) MOVs in the accumulator discharge lines EP HV 8808A,B C and D b Check valves in the accumulator discharge lines Cold bg 1 '8956A - 8945A Cold leg 2 8956B 8945B Cold bg 3 8956C 8945C Cold Leg 4 . 8956D 8945D
- 2. Operator failure to isolate interfacing LOCA 4 H O
, RHR Train A HV 8701A PV.8702A RHR Train B HV.8701B-PV 8702B These MOVs are controlled from the main control room and !
are interlocked such that they cannot be opened if RCS 1 pressure exceeds 360 psig and automatically close if RCS l pressure exceeds 682 psig.
- 3. Operator failure to successfully switch from LHI to LHR 2 H O -l including valve alignment errors -[
The valve lineup for recirculation is automatic, Operator l action is required to align the CCW syst.em to provide cooling . ,
of the RHR heat exchangers during the recirculation phase. 1 4 - LHI pump (s) fail to start or run including common cause 1,3 H !
S.M T.C failure ~
i Pumps PEJ01A j PEJ01B These pumps start automatically based on a safety injection signal, q
-' I
' 5. Failure of LHR suction (containment sump) valves to open 1,2 H S.M,T,C -i
' These valves are located outside containment but are con. i L> tained by a pressure tight vessel rated at 60 psig, 1
- 6. Failure of LH1 suction valve from RWST to close 1,2
~
These valves are normally open and have position indication on the ESFAS panel. .
A-93 a
g# -
,s-Accident Importance Inspection Dominant Fallure Modes = Sequence . Category Activities J
- 7. Failure to realign system after testing - 3 M O
- 8. Cold leg holation valve fails to close for switch to hot leg 2 M S.M,T.C
, recirculation. Operstor action is required to close these valves.
EJ HV 8809 A and B
- 9.- Pump discharge crossover valve falh to close 2 M S.M,T,C Remotely operated MOVs: HV 8716 A and B
- 10. Failure to switch from cold les to hot leg recirculation 2 M S.M,T Switch to hot leg recirculation is accomplished by:
Closing valves EJ HV 8809 A and B.
Opening valves EJ HV B716 A and B, and
, Opening valve EJ HV 8840.
- 11, LHI pump return line (miniflow) valve fails to open or remain 1,3 M S.M.T C open, including common cause and operator faih to stop pump Flow control valves: FCV 610 FCV 611
- 12. Containment sump plug 1 L S,M 13, LH hot les recirculation discharge valve falh to open 2 L S.M,T Motor operated valves HV 8802 A and B
- 14. Heat eachanger cooling water valves fall to open (CCW 2 L S.M T.C system failure)
Motor operated valves: HV 101 (RHR HE A)
- 15. Injection isolation valves fail to remain open 1,2,3 L S,M,T C Cold les injection header valves: HV 8809 A and B SI pump suction (from RWST) HV-8923 A and B Hot leg injection header holation valves HV-8802 A and B
- 16. Recirculation suction velves rupture / fail to remain closed 2 L S.M.T Isolation valves that must be closed during recirculation in.
clude:
Motor operated valves: 8813,8814 A and B (SI pump mini flow lines)
Motor. operated valves: LCV 112 D and'E (RWST to charging pump suction)
Motor. operated valves: 8806 A and B (RWST to SI pump suction)
A 94
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q l'
Accident Importance Inspection Dominent Failure Modes Sequence Category Activities -
i
- 17. Injection MOVs rupture / fall to remain closed (interfacing 4 L S M,T j 1
LOCA)
Motor operated valves:
Normally open HV.8809 A and B (RHR HX A and B Dis.
charge) j Normally closed HV.8701A and PV.8702A (RHR A to RCS l Loop !) Interlocked with HV 8704A St Pump A Suetion -l Normally closed HV 8701B and.PV.8702B (RHR B to RCS l Loop 4) interlocked with HV.8704B S1 Pump B Suction
- 18. Injection check valves: failure modes include rupture (inter. 4 L, MT facing LOCA) failure to open, and failure to remain open.
Cold leg injection check valves:
Loop 1: 8948A and 8818A Loop 2: 8948B and 8818B Loop 3: 8948C and 8818C
- 1. cop 4: 8948D and 8818D Hot leg injection check valves: l a
Loop 2: 8841A and 8949B i Loop 3: 8841B and 8949C
- 19. Pumps unavailable due to maintenance 1.2.3 L M One of two RHR pumps must be available (PEJ01 A or PEJ02B)
- 20. Operator failure to initiate recirculation cooling 2 L O Operator action is required to initiate CCW flow to the RHR heat exchangers. Inadequate CCW flow is annunciated until l flow exceeds 7000 gpm.
l
- 25. Lifting of system relief valve below set point 3 L S.T Relief valves: PSV 8856 A/B and PSV 8842. l (Serpoint: 600 psig) l l
1 l
1 1
. i A-95 l l
1 1
. ~ . . ,-
l WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Low Head Injection / Low Head Recirculation System TABLE A.9 2 MODIFIED SYSTEM WALKDOWN Desired - Actual Pow. Sup. Required Actual Description ID. No. Location Position Position Breaker # Location Position Position RHR Pump B EJ.V071 RHR B Locked CCW Return Throttled '
lsoladon RHR Pump B EJ. RHR B Locked Discharge Isola- 8724B Open tion RHR Pump A EJ V070 RHR A Locked CCW Return Throttled Isolation RHR Pump A EJ. RHR A Locked Discharge Isola- 8724A Open >-
tion ESW to RHR EJ V061 RHR B Locked ;
Pump Room Open Cooler 10B Iso- .i i
lation RHR Pump GL V017 RHR B Locked i Room Cooler - Throttled 10B ESW Re-turn Isolation RHR Pump. EJ V062 RHR B Locked Room Cooler Open 10B ESW Re.
turn isoladon i.
I' i -.
l L
i-
{
A-96 l
L
- 3. ,
WOLF CREEK GENERATING STATION '
RISK BASED INSPECTION GUIDE
' Low Head Injection / Low Head Recirculation System-TABLE A.9 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual- Pow Sup. Required Actual Description ID No. Location Position Position Breaker # 1.ocation Position Position RHR Pump EF V038 RHR A Locked _
Room Cooler Open -
10A ESW Re.'
turn Isolation RHR Pump . GL V010 RHR A Locked Room Cooler Thrott!cd 10A ESW Re-turn Isolation ESW to RHR EF V037 RHR A Locked Pump Room Open Cooler 10A Iso-lation RHR Train B to EJ.V002 RHR Closed CVCS letdown HXB Isolation 1
RHR Train A EJ V001 RHR Closed j to CVCS let. HXA j down Isolatin I l
l j
I 4
A-97 o
p ,
e .
WOLF CREEK GENERATING STATION
, RISK.B ASED INSPECTION GUIDE Low Head Injection / Low Head Recirculation System ,
TABLE A.9 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location - Position Position Breaker # Location Position Position RHR HtB U V038 CCW B Locked '
CCW Outlet Throttled Isolation Valve RHR HtA U V033 CCW.A Locked CCW Outlet .. Throttled Isolation Valve -
RHR Train A U.FCV. RHRHtA Closed 52NO ON Mini Flow 610 01ACF6 Valve RHR Train B U FCV. RHRHtB Closed 52NO ON l
Mini Flow 611 02AGR3 Valve RHR Train A U HV. Contain. Closed 52NG ON ;
RCS RHR Isola. 8701A ment El. ~ 01BEF2 l tion Valve )
l RHR Train B U HV. Contain. Closed 52NO ON j RCS RHR isola. 8701B ment El.~ OlBDF3 I tion Valve 1
l l'
l l
l l
l l
A-98 1'
i w
,- , - + , ,
m e
o-t WOLF CREEK GENERATING STATION i RISK. BASED INSPECTION GUIDE i
Low Head Injection / Low Head Recirculation System TABLE A.9 2 MODIFIED SYSTEM WAL.KDOWN (Cont-d)
Desired Actual Pow. Sup. Required Actual Det.cripdon ID No. Location Posithn Posidon Breater # kcation Position Position RHR Train A U HY. RHRHXA Open 52NO ON Cross Tie Isols- 8716A OlBERI tion Yalve RHR Train B D HV. RHRlD3 Open $2NO ON Cross Tie Iscla. 8716B 02BDR3 tion Valve RHR Train A U HV- Contain. Closed 52NO NELEC ON Sump to RllR 8811 A ment E1.~ OlBFF3 PRM Pump A RHR Train B U HV- Contain. Closed - 52NO SELEC ON Sump to RHR 8811B ment El. ~ 02BEF2 PRM Pump B RHR Train A U.HV. NPENRM Open $2NG ON RHR to Cold BBD9A OlBCRI 14g Isoladon ,
Valve
]
RHR Train B D HV. SPENRM Open $2NG ON RHR to Cold 8809B 02BBR2 leg Isolation Valve ItHR to Hot les D HV. SPENRM Closed 52NG ON isoladon Valve 8840 02BBR3 1
?
I t
i A.99
)
i
o WOLF CREEK GENERATING STATION LHl/LilR i 1
TABLE A.9 2 (Cont'd) ;
i REFERENCE DOCUMENTS i I
TITLE 1.D. NO. REY DATE 1
WCCS Licensed Operator init ie\ Training Documents
- 1.
- Emergency Core Coolir.g System and Safety iniection" LOl300600 000 01/28/88 ;
- 2.
- Residual Heat Removal System" LOl300$00 001 08/10/88 Procedures ____
- l. WCCS *RHR Normal System Lincug CKL EJ 120 9 08/13/88 Drawings
- 1. 'PA ID - Rc4idual Heat Removal System" M 12EJ01(Q) 1 07/14/87
- 2. *P&lD - Reactor Coolant System" 2 07/14/87 MJ2BB01(Q)
I t
A-100
0 o I
e WOLF CREEK GENERATING STATION Table A.101. Importance Basis and Failure Mode Identification ENGINEERED SAFETY FEATURES ACTUATION SYSTEh! (ESFAS)
. _ ~ .
Mission Success Criteria The Engineered Safety Features Actuation System (ESFAS) is designed to sense selected plant parameters, determine whether or not predetermined safety limits are being exceeded, and,if they are, to form logic combinations based on exceedence of the selected parameter limits. Once the required 10gic combination has been fonned, the ESFAS sends actuation signals to those ESF components that respond to the particular condition that exists.
The ESFAS consists of two portions of circuitry: Analog circuitry provides redundant channels that generate actuation signals concerned with the auxiliary feedwater and ventilation systems; the digital circuitry provides two redundant logic trains that receive inputs from the analog protection channels and provide the necessary logic to activate ,
reqv. ired ESF systems concerned with reactor safety and containment integrity. Each digital train is capable of actuating the required ESF equipment. The ESPAS depends on the electric power system to provide 120V AC for instrumentation and 125V DC for instru.
mentation and logic circuits.
l The specific automatic actuation signals provided by the ESFAS include:
l
- 1. Safety Injection Signal (SIS) ,
- 2. Containment Isolation Signal Phase A (CISA) ,
- 3. Containment Isolation Signal Phase B (CISB) l, 4. Containment Purge Isolation Signal (CPIS) l 5. Containment Spray Actuation Signal (CSAS) l 6. Fuel Building Isolation Signal (FBIS)
- 7. Control Room Ventilation Isolation Signal (CRVIS)
- 8. Main Steam Isolation Signal (MSLIS)
- 9. Feedwater Isolation Signal (FWIS)
- 10. Auxiliary Feedwater Actuation Signal for Motor and Turbine Driven Pumps (AFAS M, AFAS T)
- 11. Auxiliary Feedwater Low Suction Pressure Switchover (LSP)
- 12. Steam Generator Blowdown and Sample Isolation Signal (SOBSIS)
Several plant parameters are monitored by the ESFAS to generate actuation signals for safety systems listed above. These plant parameters are given below with an indication of the coincidence required for mission success.
A 101
O '
f
)
Low steamline pressure (2/3 coincidence for 1/4 steam generators) l High steamline pressure rate of decrease (2/3 coincidence for 1/4 steam generators) :
Low pressurizer pressure (2/4 coincidence) l High containment pressure (2/3 coincidence)
Containment pressure high 3 (2/4 coincidence)
Containment pressure high 2 (2/3 coincidence) i NB Bus unden'oltage condition (2/4 coincidence) )
Steam generator level lo lo (2/4 coincidence for 1/4 steam generators) l Steam generator level hi hi (2/4 coincidence for 1/4 steam generators)
Low RCS T,,, )
High containment atmosphere radiation level l High containment purge system radiation level l High fuel building ventilation system radiation level (1/2 coincidence) 2 Accident 1.nportance Ir;spection Dominant ranure Modes Sequence Category Activities
- 1. Failure of automatic trutistion logic (most critical for Aakil- 6.10 i l in y Fcedwater (AFW)irdtiadon) through following scenarios
- s) Latrurwnt failae through catrurauon or mabtenance error M O.S.M,T C .
NOTE: Motor driven AFW pumps are initiated enjo lo steam generator level (2/4 coincidence on one steam generator) or ,
l on a trip of both main feed pumps The turbine d'iven r ATW l
pumps are initiated by a lo lo steam genetator level (2/4 l coincidence on 2/4 steam generators on an undervoltage con-dition on bus NB0ll/NB02 (2/4 coincidence). ,
l b) Logic relays fail to close M T.M l c) Failure of 120V vital AC L M.S.C (see Table A.21) '
l
)
A 102
WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Engineered Safety Features Actuation System (ESFAS)
TABLE A.10 2 h10DIFIED SYSTEh! WALKDOWN !
(
The ESFAS is a normally energized system which must de energize to actuate (close) the relay contacts (with the exception of the Containment High High Pressure network -
which must energize to actuate). Operability must be assured by extensive surveillance -
testing, the observation of which will provide the inspector with direct input regarding the safety function capability of the system. System walkdown during normal power operation will only reveal whether certain circuits are properly aligned.
Such alignment checks could include:
- a. Observing that channels are not bypassed or in test.
- b. Ensuring that instrument root valves are open, particularly those instruments which initiate a reactor trip, safety injection, or start of auxiliary feedwater. -
REFERENCE DOCUMENTS TITLE I.D. NO. REV DATE WCGs. Licensed Operator Initial Training Docurr.ent
- t.
- Engineered Safety Features Actustion System" LOl301?01 000 10D0/s?
l A 103
^
. i O O WOLF CREEK GENERATING STATION l
Table A.ll 1. Importance Basis and Failure Mode Identification j l
REFUELING WATER STORAGE TANK (RWST) I i
l Mission Success Criteria 1 i
The Refueling Water Storage Tank (RWST), although a passive component, is the i source of water supply during three safety significant modes of operation: high pressure I injection, containment spray, and low pressure injection. It is also critical during the switchover, from the injection phase to high or low pressure recirculation from the containment sump upon receipt of a RWST low leve) signal, i
During the injection phase, if primary system pressure remains above the LHl/RHR l pump shutoff head, the pumps discharge to the RWST through the minimum now recirculation lines until the RCS pressure is sufficiently reduced to allow inflow.
If needed for ECCS injection, th RWST supplies the neiessary amount of borated water to provide the requirca net positive suction head (NPSH) to the RHR pumps, prior to auto switchover of the RHR pumps suctions to the containment sumps. This could occur as quickly as 14 minutes after r>ctuation. This auto switchover occurs at approximately 36% -
RWST level, on 2 out of the 4 Lo Lo.1 Level bistables. The operators initiate CCW Gow to the RHR heat exchangers prior to reaching this setpoint. The operators also get a MCB annunciator at this level (1 out of 4 Lo Lo 1 level bistables) to alert them to verify that the auto switchover is occurring. The auto switchover works on a 2 out of 4 level logic concurrent with an SIS. The auto switchover annunciator annunciates when the first level detector reaches 36(7c; therefore the auto switchover may occur sometime after the alarm is received.
Accident Importance Inspection Dominent Failure Modes Sequence Category Acthities
- 1. Common cause miscalibration of RwsT level sensors which 1.2 M O.s.C fails manual realignment of high and low pressure ECCS:
LT.930. LT 931. LT 932. LT 933 A 104
o . i WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Refueling Water Storage Tank (RWST)
TABLE A.112 MODirIED SYSTEM WALKDOWN Desired Actual Pow. Sup. Required Actual Description ID No, location Positica Position Breaker a Location Position Position RWST LT 930 BN. 713 Open isolation V007 RWST LT 931 BN. 713 Open isolation V008 g RWST LT 932 BN. /13 Open ,
Isolaten V009 RWST LT 9E B N. 713 Open "
Isc16h V010 P
r i
i l
l A 105 l
WOLF CREEK GENERATING STATION RWST TABLE A.112 (Cont'd)
REFERENCE DOCUMENTS '
i TTTLE 1.D. NO. REV DATE WCGS Licensed Operator Initial Training Document
- 1.
- Emergency Core Cooling System and Safety Injection" LOl300600 000 01/28/88 l
l l A.106 l
s A ING "' '
g-T ' STEA SIS J
t _ CENTRIFifGAL
, ' CHARGING PUMPS REFtXt.fMG WATER II STORAC,E TANK 2 06 A ,JC _ SAFETY INJECTION
- pgup A 5 HV
- t 80'2A - 2 RHR PUMP A
' CSAS W WASTE HOLDUP 3 _ CONTAINMENT ggg
- SPRAY PUMP 8 FUEL POOL - HCV HCV ' EDUCTOR
~
g CLEANUP PUMPS' FUEL POOt. 6800A e3006 LCV SI3 CLEANUF FUMPS 112D _
SI SYSTEM _ CENTRtFUGAL p -
' CHARGING PUMPS TEST LINE :
l HV i
88068 , _ SAFETY INJECTION
! CVCS BLENOING: W -
- PUMP 8 HV 88128 ,
CON TAINME NT M 7 p g pygp g SPRAf TEST LINE : .' - - p gen 8787 CSAS HV _ mm
' ' HV l
8813 RETURv4 - 4 . . . CONTAINMENT S1 PUMPS r --
' SPRAY PUMP A M en- CDUCTOR
- _ . . .. --- -- ... .... ... -. . - ~
~~
Figure A.11-1. "Wo f Cr:ek Refueling Water Storage Tank' (Source WCGS-Lesson Tc.st I.O !300600, Rev-OfX) Figure 10)
JKA m3 % - '
WOLF CREEK GENERATING STATION I
Table A.121. Importance Basis and Failure Mode Identification '
i POWER CONVERSION SYSTEM (PCS)
Mission Success Criteria !
l The power conversion system (PCS) can be used to provide feedwater to the steam j generators following a transient. The PCS consists of two 67% capacity turbine driven i main feedwater pumps, a motor driven startup feedwater pump, three 50% capacity motor- !
driven condensate pumps, and the hotwell inventory. The inventory of the hotwell (with l the CST as a backup supply) is assumed sufficient for all mission times of interest. The ;
feedwater regulating valves will close after a reactor scram, due to plant control logic. The j feedwater pumps remain on, and the miniflow valves will open. Feedwater can then be i provided to the SGs, through the feedwater regulating valve bypass valve. The PCS is ]
dependent on non class 1E DC power and instrument air. The success criterion for the PCS '
is restoration of flow from one or more main feedwater pumps to one or more steam gerseTatorb.
Attleent Importchee Ir.sperWn Dominant Fallere Modea Sequence Category Acibities
- 1. Loss of Peort c:nvers!on System is ar. important transiera 10 11
. mnt when toupled with loss of ATW. Failure modes for the PCS we:
a) FW lirie locak wim fallare of operance to isalare O.T b) Failure of main FW or condensate pumps to continue S.M.T running. Dere arc numerous trip irutiators for the FW and condensate pumps. ;
c) Failure of main FW and condensate pumps to start and run O following loss of DC bus (see Table A.31)
The MFW pump turbines are each supported by separate emergency DC motor. driven oil pumps which start when the oil pressure in the line drops 25 psig below normal.
Emergency pump A is powered from PJ.0106 and Pump B from PJ.0107.
A-108
. o j WOLF CREEK GENERATING STATION !
RISK BASED INSPECTION GUIDE Power Conversion System (PCS)
TABLE A.12 2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow. Sup. Required Actual Description ID No. Location Petition Position Breaker # Location Position Position Condensate AD. 422 Open Purnp 1A V092 Recire. Valve FV.78 Inlet Iso.
lation !
Condensate AD. 422 Open Pump 1A V093 Recire. Valve FV.78 Outlet isolation Condensate AD. 422 Open -
Pump IB V090 Recire. Valve .
- FV.15B IMet l I'"II'" !
._ = .
Conderuste AD. 422 Open ._. ,
Pump IB VM1 Recire, Yalve -
FV 13B Oudet isd.ation ,
Condensate AD. 422 Open -
Pump ic V088 Recire. Valve FV.22B trJet isolation Condensate AD. 422 Open Pump IC V089 Recire. Valve FV.22B Outlet isolation A 109
i . o WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Power Conversion System (PCS)
TABLE A.12 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Location . Position Position Breaker W Location Position Position Steam Seal & AD. 4 25 Lecksi 9di Stage Ea. V280 Open' haust Drain to LP Condenser Isolation Condenser Vac- AD HIS. Control Closed vum Breakers 113 Board Valves (AD. RLO23 HV113 A.B.C.
D)
Condewate AD FIK- Same Auto l Pump A Recire. 7B RLO23 Yelve (AD.
FV?D)
?
Cen6ensate ! AD FIK. Same Auto .
l Pump B Re:src. 15D RLO23 Valve (AD-
- FVi$P) !
Cendensam AD FIK- Sarne Auto _. 1 P;tmp C I;ev,r:. 22B RLX3 Valve (AD- ;
FV21ti) l AD HVil3B $2 PG13 On
-l Vacuum Breaker R JF2 Valve for IP Condenser
- Note: Valve must be locked open at all times except when line FC-005 BC.1 is depressurized (AFW Pump Turbine Condensate Drain Header).
5 A 110
o o WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE Power Conversion System (PCS)
TABLE A.12 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
Desired Actual Pow. Sup. Required Actual Description ID No. Locadon Position Position Breaker # Location Position Position AD.HV113D 52PG13 On Vacuum Breaker R CR2 Valve for LP Condenser AD.HV113A $2PG14 On Vacuum Breaker R EF4 Valve for LP Condenser
~
g AD.HVil3C 52PG14 On Vacuum Bretker R EF5 Valve for HP Condenser Motor Driven ,
AE.V342 ] 431 Open l SGFW Pump I l Mini Flow l>olas j den i
$GFW Pump B AE.V011 432 Lxked -
Suc6cn PT4 Open - - .
! solation 4 I
! I hP.7 Pumn 3 AE.VMS i 434 Open
'1xtun 1scl. tion I
i SGFW Pump A AE.V009 434 Open Suction isolation i
i .
l SGFW Pump A AE.V013 442 Locked l Suction PT 6 Open -
l Isolation l $GFW Pump A AE.V026 442 Locked l
Recire. FV.2B Open inlet isolation l
l A 111
. . I i
WOLF CREEK GENERATING STATION i RISK BASED INSPECTION GUIDE i
Power Conversion System (PCS) !
i TABLE A.12 2 MODIFIED SYSTEM WALKDOWN (Cont'd) ;
Desired Actual Pow. Sup. Required Aen:a1 ,
Description ID No. Location Position Position Breaker # Location Position Posinon SGFW Pump B AE V024 442 Locked Recire. FV 1B Open Inlet Isolation f 1
Emergency Lube 69 Open Oil Bearing PJ0106 Pump DPFOC3A Emergency Lube 69 Open Oil Bearing PJ010'l Pump DPFOC3B P
F 9
A-ll2
p- ,
1 *
- i r
WOLF CREEK GENERATING STATION l PCS :
TABLE A.12 2 (Cont'd)
REFERENCE DOCUMENTS TITLE !.D. NO. REV DATE WCGS Licensed Oretator Initial Training Documents:
- 1.
- Main Teedwater System" LO1505900 000 03/17/88
- 2.
- Main Condensate System" -
LO1505600 009 01/03/89 i
Procedures
.- 1. WCGS
- Condensate System Valve and Breaker Lineup" CKL AD 120 8 04/15/87
- 2. WCCS
- Main Feedwater System Valve and Electrical Lineup" CKL AE.120 10 03/31/88 41 We N - -
enm ulasus suest . . _ sur - . er # -- e ,
gener --
Gens
- %M- W WW e sur es.e P ee tsunse emunes a --+ mason WB eh-emaguen apur y smD
?
A ll3
O O l
- . j I
WOLF CREEK GENERATING STATION
. l Table A.131. Importance Basis and Failure Mode Identification )
l CHEMICAL AND VOLUME CONTROL SYSTEh! (CYCS) Eh1ERGENCY )
BORATION
. Mission Success Criteria !
l The chemical and volume control system (CVCS) provides several major functions ]
during startup, normal operation, emergency operation, and shutdown of the reactor. The j RCS boron concentration is normally controlled by the makeup portion of the CVCS.
However, there are occasions when it is necessary to borate at a rate that exceeds the normal, maximum capability of the makeup system. In these situations, the CVCS is initiated either by a SIAS or manually to rapidly inject concentrated boric acid into the i '
RCS. Of concern are the situations where the CVCS can be initiatedonly manually (i.e.,
following ATWS). )
Immediate boration flow comes directly from the two boric acid tanks using both boric acid transfer pumps. It is sent directly to the suction of the charging pumps through immediate boration valve BG HV 8104, which in turn inject into the RCS cold legs. To initiate immediate boration, the Control Room operator must perform the following:
- Open the immediate borat on i control valve BG HV 8104.
- Start both boric acid transfer pumps.
- Observe the immediate boration flow meter on main control board panet P. LOO 2 for proper indication of flow.
Since the boration flow bypasses the normal Reactor Makeup System and its indica.
tion, there is no record of the total amount of boric acid that has been added. If it is required to add a specific amount of boric acid, the control room operator must manually calculate the boric acid addition by observing the boric acid flow rate and the time d6 ration '
of the immediate boration.
Alternate immedia'e boration (manual immediate boration) is used if the normal immediate boration path is inoperable due to blockage or the immediate boration valve fails to function. This alte*nate immediate boration path is from the Boric Acid Transfer pumps through boric acid flow control valve BG FCV 110A (which is normally closed and fails open) to a manual valve. BG V-177, operated locally in the Auxiliary Building from the Safety Injection Pump roon "A". When the alternate immediate boration valve BG V.
177 is opened and the boric acio transfer pumps are started, boric acid will flow directly to the suction of the charging pumps. Indication of the boric acid flow will be available to the control room operator on the normal boric acid flow recorder. Also, the flow will be totalized on the boric acid counter.
A ll4
)
o C
. = i D-Accident importance Inspection Dominant Failure Modes Sequence Categor) Acthities 1 Failure to initism and perform emergency boration. Initiation 11 H O of emerging boration is a manual operation by the Control Room operator.
Refer to OflLNormd Procedure OFN 00 009 "Immediate Boration")
- 2. Single valve failure, to open preventing boric acid flow. 11 M S.M.T +
Pnneipal failure modes are power or cr.tical circuit fault MOV HV Blod is the immediate boration valve, and is con-trolled by the Control Room operator.
Check va!ve %174 must also aDow passage of the boric acid.
Normally closed boric neid flow control valve. BC FCV 110A must successfully fail open, or be manually opened from the Control Room. Non, ally closed manual vcse BG V.177 must be loca])y opened.
- 3. Failurs of bcr e cek pumps to provide sufficient flow 11 M S.M.T Boric Acid Transfee Pumps:
PBG02A PB002B
- 4. Charging pumps unavailable due to maintenance or failura to 11 M S.M.T run ceratifugd Charging Pumps: PBG05A and P3003fl positive Displacement Purrp: PBGM J
s A-115 s.c---
. . . j WOLF CREEK GENERATING STATION RISK. BASED INSPECTION GUIDE )
Chemical and Volume Control System (CVCS) Emergency Boration TABLE A.13 2 MODIFIED SYSTEM WALKDOWN I
Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker # tocation Position
)
Position :
1 BAT A Outlet 8461A Locked Isolation Valve Open BAT B Outlet $461B Locked Isolation Valve Open Boric Acid $463 Locked Transfer Pump Open _
A Suction isola.
tion Valve Doric Acid V348 Locked Transfe: Panp Open .
A Discharge Isoisuon Valve Boric Acid $475 Locked .
Transfer Pump Open B Suction Isola.
tion Valve Boric Acid V166 1sked Trr,asfer Purnp Open B Discharge isolation Valve A 116
e .
WOLF CREEK GENERATING STATION I
CYCS/I.B.
TABLE A.13 2 (Cont'd) !
REFERENCE DOCUMENTS TITLE 1.D. NO. REV DATE WCGS Licensed Operator Initial Training Documents:
- 1.
- Chemical & Volume Control System
- LOl30M40 000 12/29/69 ,
Procedures
- 1. WCGS Off. Normal Procedure Immediate Boration* OFN00-009 2 01/03/69
- 2. WCGS
- Chemical and Volume Control System CKL BG 120 12 01/08/89 Normal Valve Lineup
- Drawings ;
- 1. *PalD Chemical & Volume Control System
- M 12BG01(Q) 0 N/27/85
- 2. ' PAID Same" M.12BG02(Q) 2 07/14/87
- 3. *P&!D Sama" M.12BG03(Q) 3 0 0i'0/87 ,
- 4. *P&ID Same" M 12BGM(Q) 0 10,'22/24 '
- 5. "P&lD Same* _
M.12BG0$(Q) 1 07/14/67 l
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APPENDIX B f
G TABLES OF
. 1 (I) PLANT OPERATIONS INSPECTION GUIDANCE +
(2) SURVEILLANCE AND CALIBRATION INSPECTION GUIDANCE (3) MAINTENANCE INSPECTION GUIDANCE 4 i
i
, +
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Table B.I Plant Operations Inspection Guidance Recognizing that the normal system lineup is important for any given standby safety system, the following human errors are identified as important to risk.
System Failure Discussion Normal & Emergency Failure of Emergency Diesel Gen- Table A.21, item 1 AC Power crators (EDGs) to start or run
[DGhT01,DONE02]
Failure to restore AC power after Table A.21, item 3 station blackout w/ concurrent RCP seal LOCA
[See Table A.21, item 3]
% 1mproper EDG post meintenance Table A.21, item 5 valve or breaker lineup
[ESW V052,V053,V079,V080)
DC Power Loss of 125V DC bus Table A.31, item 1
[See Table A.31, item 1]
Operational test or maintenance er- Table A.31, item 3 ror rt.sulting in a) de energizing or cascading of DC power supplies b) failure to properly restore bat-teries or charger after maintenance
[See Table A.31, item 3)
Reactor Protection Sys- Operator failure to manually scram Table A.51, item 3 tem (RPS) reactor following ATWS High Head injection / Failure to switch from RWST to Table A.61, item 1 Recirculation the containment sump via the Low Head Rectreulation system
[See Table A 61, item 1)
Primary Pressure Relief PORY block valve closed Table 'A.71, Item 3 System [HVS000A.HV8000B]
Operator error in bleed and feed Table A.71, item 4 activities causes lack of RCS cool-ing Auxiliary Feedwater Failure to manually start locked Table A.81, Item 1 out standby pump (MDPPALOI A.B1 B1 l
, . I
^
. , 1 Table B.1 Plant Operations Inspection Guidance (Cont'd) I System Failure Discussion j Auxiliary Feedwater Failure to manually start pump Table A.81, item 3 l (Cont'd) given auto start failure :
[MDP PALO1 A,B/TDP PALO2)
Local fault of valve in motor. Tat,le A.81. Item 6 I driven pump discharge to steam l generator i
[V045 V031] j Local fault of suction valve from Table A.81. Item 9 )
the condensate storage tank (CST)
[V055)
Undetected flow diversion Table A.81, Item 11 )
[V031,V045,V0$$) i Undetected FW leakage back Table A.81, item 12 through pump discharge valves caures steam binding Failure to restore TD pump from Table A.81, item 16 l testing i
[PALO2)
Failure to restore TD pump dis- Table A.31, item 17 i charge valve after test j
[V055) ]'
Failure to manually open TD Table A.81, Jtem 18 pump discharge AOVs l [ ALHV6,8,10,12) ,
Low Head Injection / Operator failure to isolate interfac- Table A.91, Item 2 Recirculation ing LOCA
[HV 8701 A,B HV 8702A.B) )
Operator failure to successfully Table A.91, llem 3 l switch from LHI to LHR including 1 valve alignment errors j Failure to realign system after Table A.91, item 7 '
testing l l
Operator failure to stop pumps if Table A.91, item 12 pump return line (miniflow) valve falls to open or remain open
[FCV 610,FCV 611]
Operator failure to initiate recircu- Table A 91, Item 23 lation cooling (See Table A.91. Item 231 B2
p, Table B.1 Plant Operations Inspection Guidance (Cont'd)
System Failure Discussion Engineered Safety Fea- Failure of automatic initiation Table A.101. Item 1 tures Actuation logie by instrument failure through calibration or maintenance error Refueling Water Stor- Common cause miscalibration of Table A ll 1, item 1 age Tank (RWST) RWST level sensors which falls manual realignment of high and low pressure ECCS (LT-930,LT 931,LT-932,LT 933)
Power Conversion Loss of PCS (and AFWS) by a) Table A,121, item 1 FW line break with operator fail. (a) & (c) ure to isolate break c) Failure of MFW and condensate pumps to start or run following loss of DC bus
[ Buses PJ 0106,PJ 0107)
Emergency Boration Operator failure to initiate and Table A.131. Item 1 perform emergency boration__
B3
If WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Table B.2 Surveillance and Calibration Inspection GUIDANCE The listed components are the risk significant components for which surveillance and/or calibration should minimize failure.
System Failure Discussion Essential Service Water Failure of valves which isolate Table A.1 1, item 1 SW flow to CCW heat exchangers
[See Table A.11, item 1)
Pumps fall to start or run Table A.1 1, item 3
[ PEF 01A, PEF 01B)
Pump discharge MOV, check valve Table A.1 1, item 4 or header isolation valve falls to open or remain open
[See Table A.1 1, item 4]
Non essential load isolation valves Table A.1 1, item 5 fall to close
[See Tarle A.1 1, item 5)
Pump rtrainers plugged Table A.1 1, item 6
[See Table A.1 1, item 6)
Normal and Emergency Smergency diesel generators Table A.21, Item 1 AC Power (EDGs) fall to start or run
[DGbT01,DONE02)
Loss of vital AC bus Table A.21, item 4
[See Table A.21, Item 4)
Cooling water valves for EDG fall Table A.21, item 6 to open
[See Table A.21, item 6) -
Failure of EDG output breakers to Table A.21, item 7 close
[See Table A.21, Item 7)
Failure to transfer to reserve Table A.21, item 8 source of AC power and failure of EDG start signal Failure of inverter of MG set Table A.21, item 9
[See Table A.21, item 9)
DC Power Loss of 125V DC bus Table A.31, item 1 ISee Table A.31. Item 11 B4
o' o ,
Tabic B.2 Surveillance and Calibration Inspection Guidance (Cont'd) .
System Failure. Discussion ,
i DC Power (Cont'd) Failure of on line charger and fall. Table A.31, !!cm 2 !
ure of space to energize on de- ;
mand !
[See Table A.31. Item 2]
Failure of batteries Table A.31, item 4
{NKil,NK12 NK13,NK14)
Loss of battery room ventilation Table A.31, item 5
[See Table A.31, item 5) ,
Component Cooling Pumps fail to start or run Table A.41, item 1 :
Water System [ PEG 01 A,1B,1C,1D)
Local fault of heat exchanger Table A.41, item 2 valves which isolate or severely !
reduce CCW flow
[See Table A.41. Item 2)
Pump discharge or suction valves Table A.41, !!cm 4 fall to open or remain open -
[See Table A.41, !!em 4]
Failure to open or remain open of Table A.41, item 5 any local valve that disables all ECCS pump coolers
[See Table A.41, !!cm 5)
Reactor Protection Instrument failure due to Table A.51, !!cm 1 calibration / maintenance error or random failure which inhibits initi-ation of reactor trip signal Reactor trip breaker or trip bypass Table A.51, Jtem 2 ,
breaker fails to open
[52RTA,52RTB,52BYA,52BYB)
High Head Injection / Failure of HPI discharge valves to Table A.61, Item 2 Recirculation open including common cause fail-ure (includes check valves)
[See Table A.61, Jtem 2)
Failure of HHR suction valves to Table A.61, !!em 3 open including common cause fall-ure (includes check valves)
[See Table A.61, item 3)
Failure of pump return line Table A.61, Item 4 (mininow) valve to open falls op- ,
erating pump
[HV 8810,HF 8811 HV-8814A.HV-8814B1 B5
e .
Table B.2. Surveillance and Calibration Inspection Guidance (Cont'd)
System Failure Discussion HHl/HHR (Cont'd) Electrical failures (power cable / Table A.61. Item 5 breaker) disable HHR pump room cooling (See Table A.61. Item 5)
Failure of service water system Table A.61. Item 6 valve to open or remain open dis-ables HHR pump room cooling
[See Table A.61, Item 6)
Local fault of pumps / pumps fall to Table A.61. Item 7 start or run
[PEM01 A,B/PEJ01 A,B/PB005 A,B)
Failure of valve to open in the Table A.61. Item 8 common portion of the HHI suc-tion line from the RWST
[BN LCV-ll2 D/E) -
Plugging of manual valve in the Table A.61, item 9 HHI and $1 suction line (or in the containment sump strainers)
HHI and SI pump return line Table A.61, item 10 (mininow) valve falls to close; in-terlock. prevents HHR suction valves from opening
[See Table A.61, item 10)
Local pump failures: Table A.61, item 11
- failure of control cable to MCC
- failure of pump breaker to close (See Table A.61. Item 11]
Primary Pressure Relief PORV falls to open when required Table A.71, item 1 System for feed and bleed mode
-[PCV 455A.PCV-456)
Auxiliary Feedwater Local fault of valve in turbine ' Table A.81, item 2 driven pump discharge to steam generator
[V055)
Turbine driven pump falls to start Table A.81, !!em 4 or mn ,
IPALO21 B6
n l c o L Table B.2 Surveillance and Calibration inspection Guidance (Cont'd)
System Failure Discussion Auxiliary Feedwater Motor driven pump falls to start Table A.81, item 5 (Cont'd) or run .
[PALOl A.PALOlB)
Local fault of valve in motor Tab!c A.81, item 6 driven pump discharge to steam generator
[V045,V031)
Steam supply or throttle / trip valve Table A.81. Item 8 falls to open (or other valve faults in steam admission line) for tur-bine driven pump
[ABHV 5,ABHV 6,FCHV 312)
Local fault of suction valve from Table A.81, item 9 the CST s [V15)
Local fault of motor driven pump Table A.81, item 13 power breaker Turbine driven pump in test Table A.81, item 14
[PALO2)
Local fault of AFW actuation sig. Table A 81, item 15 nal logic falls to actuate MD pump and/or TD pump steam valves
[See Table A.81, Item 15)
Low Head injectior./ Accumulator failure, including Table A.91, item 1 Recirculation check valve failure or plugging of MOVs
[See Table A.91, item 1)
LH1 pumps fall to start or run in. Table A.91, Item 4 cluding common cause failure
[PEJ 01 A.PEJ OlB)
Failure of LHR suction (contain. Table A.91, item 5 ment sump) valves to open (EJ HV 8811A,HV 8811B]
Failure of LHI suction valve from Table A.91, !!cm 6 RWST to close Cold leg isolation valve falls to Table A.91, item 8 close (EJ HV-8809A.HV 8809B1 B7
Table B.2 Surveillance and Calibration Inspection' Guidance (Cont'd)'
- System Failure Discussion -
LHl/LHR (Cont'd) Pump discharge crossover valve Table A.91. Item 9
. falls to close
Table A.91, Item 10 Failure.to switch from cold leg to hot leg recirculation
[See Table A.91, item 10]
LH1 pump return line (miniflow) Table A.91, item 11 valve falls to open or remain g-open, including common cause (EJ FCV 610.FCV 611]
Table A.91, item 12 Containment sump plugs LH hot leg recirculation discharge Table A 91, item 13 valve fails to open n
[EJ HV-8802A.HV 8802B]
Heat exchanger cooling water Table A.91, Item 14 valves fall to open (CCW failure)
[HV 101(RHR HXA),HV 102(RHR HXB)]
Injection isolation valves fail to Table A.91, item 15
=
remain open
[See Table A.91, Item 15]
=
Recirculation suction valves Table A.91, item 16 rupture / fall to remain closed E [See Table A.9.1, item 16]-
{ Injection MOVs rupture / fall to re- Table A.91, item 17
[
f main closed (interfacing LOCA)
[Sec Table A.9-1. Item 17]
Injection check valves rupture (in. Table A.91, item 18 terfacing LOCA), failure to open, failure to remain open
[See Table A.91, item 18]
Lifting of system relief valve be- Table A.91, item 21 low setpoint.
Engineered Safety Fea- Failure of automatic initiation Table A.10-1, item 1 a
tures Actuation logic by:
a) instrument failure through cali-
- bration or maintenance error b) logic relays falling to close c) failure of 120V vital AC B-8 k ---
'N- W 4- .
Table B.2- Surveillance and Calibration Inspection Guidance (Cont'd)
- System Failure Discussion
, . Refueling Water Common cause miscalibration of Table A.11 1, item 1 Storage Tank (RWST) RWST level sensors which fails manual realignment of high and low head ECCS
[LT 930,LT 931,LT 932 LT 933]
Power Conversion Loss of PCS (& AFWS) by failure Table A.121, item 1(b) of MFW or condensate pumps to continue running
[See Table A.121, item 1(b))
Emergency Boration Single valve failure to open pre. Table A.131, item 2 (CVCS) venting boric acid flow due to power or control circuit fault (HV8104,FCV 110A]
Failure of boric acid pumps to Table A.13-1 Item 3 provide sufficient flow
[PBG02A,PBG02B]
Charging pumps unavailable due Table A.131, item 4 to maintenance or failure to run IPBG04,PBG05A,PBG05B1 B9 i
a . .
WOLF CREEK GENERATING STATION RISK BASED INSPECTION GUIDE Table B.3 Maintenance Inspection Guidance The components listed here are significant to risk because of unavailability for mainte.
. nance. The dominant contributors are usually frequency and duration of maintenance, with some contribution due to improperly performed maintenance.
System Failure Discussion Essential Service Water Failure of valves which isolate Table A.1 1 Item 1 SW flow to CCW heat exchangers
[See Table A.11 Item Il Pump train A or B out for main- Table A.1 1, item 2 tenance Pumps fail to start or run Table A.1 1, item 3
[ PEF 01A and PEF 01B]
Pump discharge MOV, check valve Table A.1 1, item 4 or header isolation valve falls to open or remttin open
[See Table A.11 Item 4)
Non essential load isolation valves Table A.1-1, item 5 fail to close
[See Table A.11, Item 5)
Pump strainers plugged Table A.1 1, item 6
[See Table A.1-1, Item 6]
Normal and Emergency Emergency diesel generators Table A.21, item 1 AC Power (EDGs) fall to start or run
[DGNE01,DGNE02]
. EDGs unavailable due to mainte- Table A.21, item 2 nance
[DGNE01 DGNE02]
Loss of a vital AC bus Table A.21, item 4
[See Table A.21 Item 4)
Improper EDG post maintenance Table A.2-1, item 5 valve or breaker lineup'
[ESW V052,V053,V079,V080)
Cooling water valves for EDG fall Table A.21. Item 6 to open
[See Table A.21, Item 61 B 10
u/'
M W' Table B.3. Maintenance Inspection Guidance (Cont'd)
System Failure Discussion L Normal and Emergency Failure of EDG output breakers to Table .~A.21, item 7 -
AC Power (Cont'd) close
[See Table A.21, item 7] ;
Failure to transfer to reserve Table A.21, Item 8 source of AC power and failure of EDG start signal ,
Failure of inverter or MG set Table A.21, !!cm 9
[See Table A.21 Item 9] ,
DC Power Loss of 125V DC bus Table A.31, Item 1 l
[See Table A.31, Item 1) !
Failure of on line charger and fall- Table A.31, Item 2-ure of spare to energize on de-mand
[See Table A.31. Item 2} !
Operational test or' maintenance er- Table A.31. Item 3 ror resulting in _;
a) de energizing or cascading of DC power supplies 1 b) failure to properly restore bat-teries or charger after maintenance
[See Tab!c A.31 Item 3]
Failure of batteries Table A.31, Item 4
[NK11,NK12,NK13,NK14]
L Loss of battery room-ventilatio Table A.31, Item 5
[See Table A.31 Item 5)
Component Cooling- Pumps fall to start or run Table A 41, item 1 Water System [ PEG 01 A,lB,1C,1D) l Local fault of heat exchanger Table A.41, item 2 valves which isolate or severely reduce CCW flow or SW coolant l flow
[See Table A 4-1. Item 2]
l Pumps out for maintenance Table A 41 Item 3
[ PEG 01 A,1B,1 Cold) :,
- Pump discharge or suction valves Table A.41, item 4 L _ fall to open or remain open L [See Table A 41, item 4]
1 i Failure to open or remain open of Table A.41, item 5 any local valve that disables all l ECCS pump coolers i ISee Table A.41, Item 51 1
B 11
%W
- Table B.3 Maintenance inspection Guidance (Cont'd)
System Failure Discussion l
v Reactor Protection Instrument failure due to Table A.51, item 1 calibration / maintenance error, or random failure which inhibits initi-ation of reactor trip signal Reactor trip breaker or trip bypass Table A.51, item 2 breaker falls to open 152RTA,52RTB,52BYA,52BYB)
High Head injection / Failure of HH1 discharge valves to Table A.61, item 2 Recirculation open including common cause fall-ure (includes check valves)
[See Table A.61, item 2)
Failure of HHR suction valves to Table A.61, Item 3 open including common cause fall-ure (includes check valves)
[See Table A.61, item 3)
Failure of pump return line Table A.61, item 4 (miniflow) valve to open falls op-erating pump
[HV8810,HV8811,HV 8814A.HV-8814B)
Electrical failures (power cable / Table A.61, item 5 breaker) disable HHR pump room cooling
[See Table A.61 Item 5)
Failure of service water system Table A.61, item 6 valve to open or remain open dis-ables HHR pump room cooling
[See Table A.6 l, item 6)
Local fault of pumps / pumps fall to Table A.61, item 7 start or run
[PEM01 A,B/PEJ01 A,B/PBG05 A,B)
Failure of valve to open in the Table A.6 2, Item 8 common portion of the Hill suc-tion line from the RWST
[BN LCV-112 D/E]
Plugging of manual valve in the Table A.61, item 9 common HHI suction line (or in the containment sump strainers)
B-12
.g , y.
.,: t. .,
Table B.3 Maintenance Inspection Guidance (Cont'd)
System - Failure Discussion
-]
High Head Injection / HHI and SI pump return line Table A.61, Jtem 10 -
Recirculation (Cont'd) (miniflow) valve fails to close; in- i terlock prevents HHR suction I valves from opening ,
[See Table A.61, item 10]
Local pump failures: Table A.61, Item 11
- failure of control cable to MCC
- failure of pump breaker to close
[See Table A.61. Item 11]
Pump in maintenance Table A.61, Item 13 -
[PEM01 A,B/PEJ01 A,B/PBG05 A,B]
Primary Pressure Relief. PORV falls to open when required Table A.71 Item 1 System - for feed and bleed mode
[PCV 455A.PCV-456)
Failure of PORV/SRV to resent Table A.71, Item 2 causing small LOCA
[PCV-455A PCV-456/SRV-8010A, '
B C) ;
PORV block valve closed Table A.71, Jtem 3
[HV 8000A, HV 8000B]
Auxiliary Feedwater Local fault of valve in turbine- Table A.81, Jtem '2 driven pump discharge to steam generator
[V055)
Turbine driven pump falls to start Table A.81, item 4 or run
[PALO2)
Motor driven pump falls to start Table A.8-1 Item 5 ~
or run
[PALOI A,PALOlB]
Local fault of valve in motor Table A.81, Jtem 6 driven pump discharge to steam generator
[V045,V031)
Turbine driven pump in mainte- Table A.81, Jtem 7 nance IPALO21 B-13
n ;
Table B.3 Maintenance Inspection Guidance (Cont'd) l l System - Failure Discussion Auxiliary Feedwater Steam supply valve or throttle / trip Table A.81, item 8 i:
(Cont'd) valve falls to open (or other valve faults. In steam admission line) for turbine driven pump-
[ABHV 5,ABHV-6,FCHV 312]
4 Local fault of suction valve from' Table A.81, item 9 the CST
[V15)
AFW flow control valve in main- Table A.81, item 10 tenance falls delivery from TD j pumps ;
I-
[ALHV6/ALHV8/ALHV10/
ALHV12]
l_
Local fault of motor driven pump - Table A.81, item 13 l
power breaker i Motor driven pump in maintenance Table A.81, Item 19
[ PAL 01A. PAL 01B)
. Low Head injection / Accumulator failure including Table A.91, Item 1 Recirculation check valve failure or plugging of MOVs
[See Table A.91, item 1)
LH1 pumps fall to start or run in.
Table A.91, item 4 cluding common cause failure (PEJ Ol A,PEJ 01B]
Failure of LHR suction-(contain. Table A.91, item 5 ment sump) valves to open
[EJ HV 8811 A,HV-8811B) l: Failure of LHI suction valve from Table A.91, item 6 l RWST to close i
[EJ HV-8812A HV 8812B)
Cold leg isolation valve falls to Table A 91, item 8 -
close l [EJ HV 8809A,HV 8809B)
Pump discharge crossover valve Table A.91, item 9 falls to close
[EJ HV 8716A.HV-8716B)
[ Failure to switch from cold leg to Table A.91, item 10 hot leg recirculation ISee Tn.ble A.91, item 101 B-14 L
i l' ,
, 1
- + 1 J
L . . .
/"0 **
' Table B.3 Maintenance Inspection Guidance (Cont'd) 1 g System Failure Discussion Low Head Injection /J LHI pump return line (miniflow) Table A.91. Item 11 Recirculation (Cont'd) . valve falls to open or remain open >
[EJ FCV 610,FCV 611]- l Containment sump plugs Table A.91, Jtem 12-LH hot leg recirculation discharge Table A.91, item 13 '
valve falls to open
Heat exchanger cooling water Table A.91 Item 14 valves fall to open (CCW failure)
[HV 101(RHR HXA),HV 102(RHR HXB))
. Injection isolation valves fall to . Table A.91, item 15 remain open .
[See Table A.91, Item 15] ;
Recirculation suction 'valves Table' A.91, item 16 rupture / fall to remain closed
[See Table A,91. Item 16) -
Injection MOVs rupture / fall to re- Table A.91. Item 17 main closed (interfacing LOCA)
[See Table A.91. Item 17]
Injection check valves rupture (in. Table A.9-1, Item 18 terfacing LOCA), failure to open, failure to remain open
[See Table A 91. Item -18]
Pumps unavailable due to mainte- Table A.91, Item 19 nance
[PEJ01 A PEJ01B)
Lifting of system relief valve be- Table A.91, Item 21 .
Iow setpoint l'
[PSV 8856A/B,PSV 8842]
Engineered Safety Fea- Failure of automatic initiation Table A.10-1, Item 1 l
tures Actuation logic by:
a) instrument failure through cali-bration or maintenance error b) logic relays failing to close c) failure of 120V vital AC L Refueling Water Stor- Failure to realign system after re- Table A.11 1. Item 2 age Tank (RWST) fueling outage.
[ Refueling proceduresl B-15 l
u_
l
- e .
Table B.3 ' Maintenance Inspection Guidance (Cont'd)
Q.
System Failure Discussion Power Conversion' - Loss of PCS (& AFWS) by Table A.121, item 1(b) b) failure of MFW or condensate pumps to continue running
[See Table A.121. Item 1(b)]
Emergency Boration Single valve tallure to open pre- Table A.131, item 2 (CVCS) ' venting boric acid flow due to power or control circuit fault
[HV 8104,FCV 110A]
Failure of boric acid pumps to Table A.131. Item 3 provide sufficient flow
[PBG02A PBG02B]
Charging pumps unavailable due Table A,131, item 4 to maintenance or failure to run iPBG04,PB005 A PBGOSB1 n
B-16 ,
, sa.--p e + = 1 O i .
', ? '
- s t
APPENDIX C 4
CONTAINMENT AND DRYWELL WALKDOWN 1
a v
- i k
e
.a v L, _+ WOLF CREEK GENERATING STATION i
RISKoBASED INSPECTION GUIDE Table C.1 ' Containment Walkdown Discussion Since the containment is generally inaccessible during normal plant operation, those components listed in the preceding tables which are located within the containment are listed below:
Desired Actual Description- ID No. Location Position - Position RHR Train A EJ HV- Containment Closed RCS RHR Isolation Valve 8701A El, RHR Train B EJ HV- Containment Closed RCS RHR lsolation Valve 8701B El.
RHR Train A EJ HV- Containment Closed Sump to RHR Pump A 8811 A El.
RHR Train B EJ HV- Containment Closed Sump to RHR Pump B 8811B El.
Pressurizer PORY PCV-455A Containment Closed El.
Pressurizer PORV PCV 456A Closed Pressurizer Safety Relief Valve SRV 8010A Not Gagged Pressurizer Safety Relief Valve SRV 8010B Not Gagged Pressurizer Safety Relief Valve SRV 8010C Not Gagged C-1
e.
TABLE C.1 CONTAINMENT WALKDOWN (Cont'd) r Desired Actual Description ID No. Location Position Position
' Pressurizer PORY Block Valve HV 8000A Open (PCV 455A)
. Pressurizer PORV Block Valve HV 8000B Open (PCV 456 A) t k
C-2
..