ML20043H926
| ML20043H926 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/31/1990 |
| From: | Shier W BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20043H913 | List: |
| References | |
| CON-FIN-A-3875 TR-A-3875-T2D, TR-A-3875-T2D-R-DR, TR-A-3875-T2D-R00-DR, NUDOCS 9006270088 | |
| Download: ML20043H926 (107) | |
Text
{{#Wiki_filter:. > 4 1
- Technical Report A 3875 T2D Rev. 0 - .
Tf t URaxi ; YOGTLE ELECTRIC GENERATING PLANT l RISK. BASED INSPECTION GUIDE (RIG) (BASED ON GENERIC INSIGHTS FROM PROBABILISTIC RISK 6 ASSESSMENTS FOR WESTINGHOUSE PWRS) . i 1 March 1990 j i Prepared by: W. Shier ; Department of Nuclear Energy Brookhaven National Laboratory l, Upton, New York 11973 Prepared for: U.S. Nuclear Regulatory Commission Washington, DC 20555 l , i FIN A-3875 L 0 111 I 90062700GG 900424 DR ADOCK0500f.{4 l
-- - - - . _ _ . _ . _.____.___________________________l
I Abstreet . This report provides risk based inspection guidance for the Vogtle Electric Ocnerating Station based on prior work sponsored by the Office of Nuclear Reactor Regulation. U.S. Nuclear Regulatory Commission. The inspection guidance developed for Vogtle is based on generic insights for Probabilistic Risk Assessments (PRAs) completed for several Westinghouse PWRs. Thirteen Vogtle systems have been analyzed for potential failure modes based on the results of previous analyses for Westinghouse plants that have completed detailed PRAs. The Vogtle plant does not have a PRA: thus, the Vogtle results are based on experience, engineering judgements, and the generic risk. based inspection guide previously developed by BNL. The results presented should be useful for the selection plant areas to be inspected and should complement the existing NRC inspection guides and requirements. t ill
. . 4
-- TABLE OF CONTENTS Dn ABSTRACT ........................................... ill
- 1. INTR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I
- 2. OHNERIC DOMINANT ACCIDENT SEQUENCES . . . . . . . . . . . . . . . 1 i
2.1 LOCAs...................................... ! 2.1.1 Small or Intermediate LOCAs (Sequence 1) . . . . . . . . . . . . . . . . I 2.1.2 Medium or Large LOCAs (Sequences 2.3). . . . . . . . . . . . . . . . . - 2 2.1.3 LOCA's Outside Containment (Sequence 4) . . . . . . . . . . . . . . . . 3 2.2 Transient Sequence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2.1 Loss of All Component Cooling Water (Sequence 5) . . . . . . . . . . 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 (Sequence s 7.8 and 9) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.2.4 Loss of Feedwater System or Transient Followed by Loss of l Feedwater with Loss of Decay Heat Removal (Sequence 10) . . . . . 8 2.3 Anticipated Transient Without Scram (ATWS) Followed by Failure of Emergency Boration . . . . . . . . . . . . . . . . . . . . . . . . 9
- 3. COMMON CAUSE FAILURES . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
- 4. IMPORTANT HUMAN ERRORS . . . . . . . . . . . . . . . . . . . . . . . . . . 10
- 5. SYSTEMS INCLUDED IN OUIDE . . . . . . . . . . . . . . . . . . . . . . . . . 10
- 6. SYSTEM INSPECTION TABLES . . . . . . . . . . . . . . . . . . . . . . . . . 11
- 7. R E FE R EN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 7.1 Generic Risk Based Information . . . . . . . . . . . . . . . . . . . . . . . 13 7.2 Other
References:
Plant Specific Risk Based Information . . . . . . . . 14 l APPENDIX A - Tables of (1) Imrortance Basis and Failure Mode Identification.' and (2) Modified System Walkdowns . . . . . . . . . . . . . . . . . . . . . A1 APPENDIX B . Tables of (1) Plant' Operations Inspection Guidance. (2) Surveillance and Calibration Inspection Guidance, and (3) Maintenance Inspection Guidance. . . . . . . . . . . . . . . . . . . B1 APPENDIX C Containment and Drywell Walkdown. ........ ....... C1 l Y
. e . .
l 1 I LIST OF TABLES
.P,,ag 1 Representative PWR Accident Sequences. . . . . . . . . . . . . . . . . . . . . 3 2 Systems Included in Generic RIGS for PWRs. . . . . . . . . . . . . . . . . . I1 3 APPENDIX A Importance Basis and Failure Mode Identification A2 '
I ! A.1 1 Nuclear Service Cooling Water System (SWS) . . . . . . . . . . . . . . . . . A.2 1 Normal / Emergency AC Power System . . . . . . . . . . . . . . . . . . . . . . A3 q A.3 1 DC Powe r S yste m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A5 l A.4 1 Component Cooling Water System. . . . . . . . . . . . . . . . . . . . . . . . . A6 ' f A.5 1 Reactor Protection System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A7 A.6 1 High Pressure Injection / Recirculation (HP!/HPR) . . . . . . . . . . . . . . . . A9 A.7 1 Primary Pressure Relief System . . . . . . . . . . . . . . . . . . . . . . . . . . A 12 l A.8 1 Auxiliary Feedwater System . . . . . . . . . . . . . . . . . . . . . . . . . . . . A 13 I A.9 1 Low Pressure injection / Recirculation (LPl/LPR) . . . . . . . . . . . . . . . . A 16 l A.101 Engineered Safety Features Actuation System (ESFAS) . . . . . . . . . . . . A 19 A.11 1 Refueling Water Storage Tank (RWST) . . . . . . . . . . . . . . . . . . . . . A.21 A.12 1 Condensate and Feedwater System. . . . . . . . .. . . . . . . . . . . . . . . . . A 22 ) A.131 Chemical and Volume Control System (CVCS) Emergency Boration . . . . A 23 APPENDIX B B.1 Plant Operations Inspection Guidance. . . . . . . . . . . . . . . . . . . . . . . B2 B.2 Surveillance and Calibration Inspection Guidance . . . . . . . . . . . . . . . B5" B.3 Maintenance Inspection Guidance . . . . . . . . . . . . . . . . . . . . . . . . . B 12 , APPENDIX C CONTAINMENT AND DRYWELL WALKDOWN . . . . . . . . C1 t 9 vi
. l
l'. INTRODUCTION , This inspection guide has been prepared to provide generic risk based inspection guidance for the Vogtle Electric Generating Plant based on revhw of several Probabilistic Risk Assess-ments (PRAs).W'7 8J 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 impection 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 LOCAs 2.1.1 Small or Intermediate LOCAs (Sequence 1) This accident sequence is initiated by a small ($ 5 2 in.) or intermediate (2 in < $ $ 6 in.) LOCA which does not depressurize the Reactor Coolant System (RCS) below the shutoff head of the low pressure Emergency Core Cooling System (ECCS). The Reactor Protection System (RPS) successfully scrams the reactor but the high pressure ECCS fails to provide adequate makeup, either in the injection or recirculation phases, resulting in core damage. Small LOCAs htye actually occurred in commercial nuclear power plants and consist primarily of stuck open PORVs and to a lesser degree reactor coolant pump (RCP) seal failures. Failures during the injection phase are dominated by valve failures in the common discharge or suction lines or in the mini flow lines. Failures during the high pressure recirculation modes of operation that can occur in the ECCS system, or in any of the support systems required for long term LOCA mitigations, are the dominant contributors to these sequences. The ECCS failures in the recirculation mode are dominated by operator failure to correctly realign the system from the injection mode or valve failures in the common discharge or suction lines or the mini flow line. Vogtle configuration takes suction from the low pressure recirculation pump discharge. The primary faults are suction (containment sump) valve and pump malfunctions. 2.1.2 Medium or Large LOCAs (Sequences 2,3) In accident sequences 2 and 3, an intermediate or large LOCA occurs which rapidly depressurizes the reactor coolant system, followed by successful scram of the reactor. Operation of the low pressure injection (LPI) mode of the ECCS is either successful but followed by failure of the recirculation system, or the LPI system itself initially fails, either of which leads to core damage. The initiating event is an intermediate or large prir.ary system pressure boundary failure from 6 to 29 in. In diameter. No actual failures of this r..agnitude have occurred. I
i Table 1 , Accident Sequences ! J Loss of Coolant Accident Sequences l l l 1. Small or medium LOCAs with failure of the emergency core cooling system high pressure l injection or recirculation. J
- 2. Medium or large LOCAs with failure of the emergeny core cooling system low pressure recirculation.
l 3. Medium or large LOCA with failure of the emergency core cooling system low pressure [ l injection. l l
- 4. LOCA outside containment.* f Transient Sequences
- 5. Loss of all CCW or ACCW initiator with a subsequent RCP seal LOCA. f I
l 6. Loss of 125V de bus initiator with failure of the .tuxiliary feedwater system (AFW). l l
- 7. Loss of offsite power (LOOP) initiator with failure of AFW and feed and bleed. )
- 8. Station blackout with loss of the AFW system.
- 9. Station blackout with a subsequent RCP seal LOCA. .
i
- 10. Loss of feedwater initiator (or a general transient with loss of feedwater) followed by loss ,
of AFW." i Anticipated Transient Without Scram (ATWS) Sequences i
- 11. Transient initiator with failure to automatically and manually scram with failure o.' timely ,
emergency boration. f
*Specified because of serious .offsite consequences.
"Specified based on a review of the studies that established precursors to potential severe core damage accidents (NUREO/CR 2497, 3591. 4674). j 2.1.2.1 Failures of the Low Pressure Recirculatico Mode of the ECCS (Sequence 2) ;
A major contributor to core dam' age for this u quence is the failure of the low pressure ! ECCS in the recirculation mode. This system failure is evenly divided between human errors and hardware failures. The dominant human error contributor is the failure to initiate recirculation I by manual realignment of the pump suction from the Refueling Water Storage Tank (RWST) to ! the containment sump.- l Since boron precipitation in the reactor vessel can be minimized or prevented and steam voids in the RPV head can be condensed by a backflush of cooling water through the core to : reduce boll off and resulting concentration of boric acid in the water remaining in the reactor {, t 2-
. t
vessel, a second operator error is the failure to manually switch the recirculation pump discharge from cold leg io hot les recirculation after about 14 hours following an accident. The common cause miscalibration cf the RWST level sensors is the only major failure not directly associated with the low pressure 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 falling to open, and operator failure to initiate recirculation cooling. Failures of recirculation system can also occur during the high pressuste operating mode as previously described in 2.1.1 above. 2.1.2.2 Failures in the Low Pressure Injection (LPI) Mode (Sequence 3) A major contributor to core damage for Sequence 3 is the failure to provide short term core injection due to accumulator or low pressure injection malfunctions. The accumulator failure is attributed to discharge line failures, primarily check valve failures to open or MOV plugging. The Low Pressure injection (LPI) system failure is dominated by pump failure to start or run, including common cause. Human error contributors are the failure to restore the system to operable status after testing and tiie failure to stop the pumps if the miniflow valve falls to open. The dominant system valve fallitre 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. 2.1.3 LOCAs Outside Containment (Sequence 4) The commonly designated V sequence, here called Sequence 4 or LOCA outside contain. ment,is initiated by a failure of any one of the pairs of series interface check valves that isolate the high pressure reactor coolant system (RCS) from the Low Pressure Injection (LPI) system. The resultant flow into the LPI system is assumed to rupture the low pressure piping or components outside the containment boundary. Although core inventory makeup by the high pressure injection and any available low pressure injection systems is initially available, the inability to switch to the recirculation mode eventually leads to core damage. The discharge of the LPI System consists of two low pressure injection lines with a normally open MOV. Downstream of this MOV (toward the RCS) the piping is rated for primary loop conditions. The discharge lines divide to connect to each RCS cold leg. Etch of these individual lines has one check valve 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 fail 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.
3
l
. j i
e 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 l valve occurs. l In addition, Vogtle has the accumulator discharge connected between the two interface f 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 LPI piping, thereby alerting the operator. l 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.' , t A potential recovery action has been included to accoum for operator action to isolate the l Interfacing LOCA by maeval closure of the LPI discharge MOV. The successful mitigation of ; this event is plant specific and is dependent on: !
- LPI pump separation 'o minimize the environmental impact of RCS blowdown on the second train. l
. The existence of two isolatable LPI discharge headers to enable the use of the other LPI f loop, or the ability to use another system for RCS makeup. !
- The capability of the LPI discharge MOV to isolaie the interfacing LOCA. The valve may not be designed to close against such a high differential pressure, thereby failing the i valve in the open position before depressurization can be effected.
2.2 Transient Sequences l l 2.2.1 Loss of All Component Cooling Water (Sequence $) l Sequence 5 represents a complete loss of the component cooling water (CCW) system that
- precipitates a reactor coolant pump (RCP) seal LOCA and also disables the high and low i pressure ECCS. The inability to provide high pressure makeup results in core damage. ;
One major contributor to the loss of CCW initiator is a pipe rupture that drains the system inventory before the bretk can be located and isolated. The second contributor is the common i cause failure of all operatlng CCW pumps, compounded by a failure of the standby pamp(s) to , start and run, i 2.2.2 Loss of DC Power (One 125V DC Bus) (Sequence 6) This sequence is initiated by a non ret.overable loss of a 125V DC bus. The DC power ! system provides control power to various systems. There have 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. The remainder of the
. precursors are due to equipment failures. A loss of one DC bus will typically disable the main
'M. Hibbs, et al., *NRC Studying implications of Unpublicized German Reactor incident," laside NRC, 10(23).1, December 3,1988. ,
4 6
- 9
- - --- --------- - - - -- _ _--- , , - - - - - , ,-r- - - - . . - , , , - ,,-----,,,,m,, -
feedwater system, a portion of the auxiliary feedwater system, and various DC dependent valves possibly including a PORV. This sequence postulates the failure of the remainder of the AFW system and the feed and bleed mode. Failure of secondary heat removal results in core inventory 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 the failure of two additional AFW trains. The major cause is system hardware failure including: 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. 2.2.3 Loss of Offsite Power / Station Blackout initiators Loss of offsite power (LOOP) accident sequences are characterized by loss of offsite pcwer followed by at least partial 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 Powei 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 I feedwater is unavailable due to the initiator. The Auxiliary Feedwater (AFW) system falls 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 LOOP initiator is one of the more common operating transients, comprising approxi-mately 21% of all precursors to potential core damage. Although some of these initiators are weather or 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 during or post maintenance and errors related to manual breaker operation. In addition, several initiators were caused by station transformer faults. The subsequent failure of one or more sources of emergency AC power, usually emergency diesel generators (EDGs) failing to start or run, is important because it disables a portion of the Auxiliary Feedwater (AFW) system. The major contributor to this sequence is the failure of the AFW system to provide sufficient flow to tht steam generators, partially caused by the failure of one or more (but not all) EDGs The remarider of the system falls due to a combination of unrelated faults, such as local failures (primarily valve related) of the AFW turbine steam admission line or the AFW pump discharge lines and local faults of the turbine driven (TD) pump. Another contribu. i tor is TD pump unavailability due to maintenance activities. The AFW system can also be subject to several common mode failures. One mode is undetected flow diversiont the seconde is sttam binding of the pumps due to main feedwater leakage through the AFW pump discharge check valves which flashes to steam in the AFW pump. 5 i
i i l ,
, )
The bleed and feed mode is the option of last resort. It is highly plant specific. In'some l plant PRAs, only one PORY is considered necessary for system success; however in other PRAs,
- 1 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 PORY block valve, given loss of the EDO. The block valve requires AC power ; ! to reopen. , 2.2.3.2 Station Blackout Sequences ! The dominant station blackout accident sequences begin with LOOP as described in 2.2.3.1 followed by failure of all onsite power sources. One short term station blackout has occurred, i 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, i
- l. 2.2.3.2.1 Station Blackout with Failure of Decay Heat Removal (Sequence 8)
The loss of all AC power results in an immediate failure of all decay heat removal systems except the turbine driven portion of the auxiliary feedwater system. The AFW system subse-quently falls 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 i unavailability of an EDO 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 1 failures are attributable to station battery depletion, that results in the loss of instrumentation and control pc,wer. Shon term failures are due to turbine driven pump, AFW discharge valve failures or the failure to manually open the pump discharge air operated valves. 2.2.3.2.2 St: tion Blackout with RCP Seal LOCA (Sequence 9) l In this sequence, the loss of all AC power disables all primary system injection, as well as , RCP seal coolir.g. A RCP seal LOCA occurs, accelerating the loss of the primary system, inventory and the onset of core damage. P 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 ger.erators (EDGs) or the unavailability of one EDO 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 inca 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 successfts restoration of the HP! component cooling. 2.2.4 Loss of Feedwater System or Transient Followed by Loss of Feedwater with Loss of Decay Heat Removal (Sequence 10) ' The loss of the condensate and feedwater system (or a transient followed by a loss of the i feedwater system) with the subsequent failure of the AFW system causes the primary system to overheat. The resulting system over pressurization causes PORY 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
. 6
, , , , 1,, --
c@sure failures, primarily due to positioner linkage p ablems, contributed approximately 15%. The remainder of the loss of feedwater precursors are fairly evenly divided among condensate pump trips, feedwater recirculation, control and bypass valve malfunctions, feedwater controller failures and miscellaneous contributors including multiple stuck open relief valves and main turbine trips v>hich induced feedwater 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 unavailabil-ity 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. 2.3 Anticipated Transient Without Scram (ATWS) Followed by Failure of Emergency Boration This sequence is initiated by a transient from high power followed by an RPS failure to automatically scram the reactor. The RPV has survived the initial pressure transient due to a negative moderator temperature coefficient. The attempts to manually scram are not successful P and emergency boration also fails. The initiator is a transient such as a MSIV closure, partial loss o' 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 failure 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 la dependent on system design. Some plants have an in line boric acid injection tank with redundant valving that is an integral part of the charging system /high pressure 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. Vogtle utilizes two boric acid pumps discharging through a common, normally closed high flow line to the charging pump suction header. Operator action is required to start a second pump (or switch a single operating pump to fast speed opertion); the normally closed MOV is opened based on the safety injection actuation signal. This configuration is more vulnerable to hardware failures related to the use of a single normally closed MOV and/or the system success criteria that requires both boric acid pumps to operate.
. 7
l
- 3. COMMON CAUSE FAILURES
- Certain common cause failures, either hardware or human related, appear as particularly i 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 fail to start or continue to run. ! c) Component Cooling Water or Auxiliary Component Cooling Water pumps fall to continue running. ! i d) Failure of high pressure injection discharge valves to open. c) Failure of low pressure injection pumps to start or continue running. !
- 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 e,ror. . b) Failure to switch from the Refueling Water S'.orage Tank (RWST) to the containment sump, i.e., failure to switch from the low p' essure injection phase to the low pressure . recirculation phase in response to a large or medium LOCA. l c) Failure to manually start locked out AFW pump, either turbine or motor driveh? ( 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 OUIDE 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 teneric insights. i In using the list, the inspector should select systems for inspection based on both knowl.
- edge 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. 8
, i l
Table 2 Systems Included in Generic RIOs for PWRs
- 1. Nuclear Service Cooling Water
- 2. Normal / Emergency AC Power
- 3. DC Power
- 4. Component Cooling Water (CCW)/ Auxiliary Component Cooling Water (ACCW)
- 5. Reactor Trip System
- 6. High Pressure injection /High Pressure Recirculation *
- 7. Primary Pressure Relief System (PPRS)
- 8. Auxiliary Feedwater (AFW)
- 9. Low Pressure Injection / Low Pressure Recirculation *
- 10. Engineering Safety Featuis Actuation System (ESFAS)
- 11. Refueling Water Storage Tank (RWST)
- 12. Condensate and Feedwater System
- 13. Emergency Boration/ Chemical and Volume Control System (CVCS)
- 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. Each of these is explained in detail below.
In uting these tables, however, it is essential to remember that other systems _and compo. nents we also important, if, through inattention, the failure probabilities of other systems were silowed to increase significantly, their contributloas 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 tables allow an inspector to concentrate on systems and components that are most significant to risk. In so doing, however, cogniaance of the status of systems performing other essential safety functions must be maintained.
- HPI includes the Safety injection System (SIS).
- Includes the Residual Heat Removal System (RHRS).
9
l ! 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 could be extracted from the system descriptions or other plant informa- i tion. (Note that this success criteria may be different from the more conservative success criteria j contained in the FSAR.) t 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 fall 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 fall 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. I 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 i 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, j 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: i O Nonnal and emergency operating procedures, check off lists, technical specifications, ; training, etc. S Periodic surveillance activities, procedures and training. a M Preventive or unscheduled maintenaace activities, technical specifications, procedures I and training. l T Periodic testing or in service inspectitn activities, procedures, and training. C Periodic calibration activities, proceduros, and training, j I ! Each failure mode is correlated to the appropriate accident sequence (s) described in Table ! ; and categorized as of high (H) medium (M) or low (L) importance, in nearly all' cases, the i importance categories are numerically based taking into account the event's contribution to the ! l eleven representative accident sequences. I Table AX 2 - Modified System Walkdown This table provides an abbreviated version of the licensee's system checklist, where ! i svallable, but includes only those items which are related to the dominant failure modes. It is ! generally.much less than the normal checklist. 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 10 i
- . , , . r-_., .- - - - ~ ~ * , --~ . , . - - -
. checklists, since they are based on certain versions of the licensee's system operating instrue.
tions. Valve numbers used are those identified in the licensee system checklists, or P&lD's. . Figure AX - Simpilfled 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 PalD'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 programs. Table 83 - 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 licensee's maintenance program. Important factors include the fre-quency and duration of maintenance as well as errors that degrade the component or render it inoperable when it is returned to service. APPENDIX C i Table C1 - Containment and Drywell Walkdown Table - Because they are normally inaccessible during operation, a separate walkdown checklist 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 these items when the opportunity arises.
- APPENDIX D (OPTIONAL)
System Dependency Matrix Whenever it is readily available, matrices showing the dependencies, and inter-dependencies, of front line FSF systems and support systems, and also of support systems to other support systems are provided to aid the inspector in determining what cther systems (or trains of systems) are affecte:1 when a particular system or train falls. This can be helpful in deciding the importance of syitems and in reviewing the adequacy of operator actions to restore the systems to service when they become inoperable (See Ref. 7.2.1).
. 11 a
s . .. l
- 7. REFERENCES i
1 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 Important Systems Components and Human Actions," BNL Technical Letter 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 1 cover letter to Dr. J.W. Chung, USNRC, dated November 7,1989. ;
- 3. R.E. Gregg and R.E. Wright, " Appendix Review for Dominant Generic Contributors," Idaho National Engineering Laboratory, Report No. BLB 3188, March 1988. j j 7.2 Other
References:
Plant Specific Risk Based Information ;
- 1. J.H. Taylor, et al. " Development and Use of Risk Based Inspection Guides," Brookhaven National Laboratory, NUREG/CR 5371. June 1989. ,
- 2. 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, i
- 3. A. Fresco, et al., " Indian Point Unit 3 Probabillstic Risk Assessment Based System Inspection Plans," Brookhaven National Laboratory, Technical Report A 3453 871, Rev. O, May 1987.
- 4. C.L. Atwood, et al., "PRA Applications Program for inspection at the Zion Nuclear Power i Station Draft Report," Idaho National Engineering Laboratory, Informal Report EGO EA
7304, June 1986. , i
- 5. M.F. Hinton and R.E. Wright, "PRA Applications Program for Inspection at Seabrook i Station Draft Report," Idaho National Engineering Laboratory, Informal Report EGG EA- j 7194, March 1986.
- 6. R.E. Gregg, et al., "PRA Applications Program for Inspection at the Surry Nuclear Power 1 Station Unit 1 Draft Report." Idaho National Engineering Laboratory, Informal Report i EGO REQ.7746, July 1987.
- 7. R.E. Gregg and R.E. Wright, "PRA Applications Program for Inspection at Millstone Unit 3 Draft Report," Idaho National Engineering Laboratory, Informal Report EGG SSRE 8016, l March 1988. l l
l 9 12 l l 1 1
- -- _ _ _I
ii is iii =niiis- .- i-4- 4 % e 4 APPENDIX A 4 e e
1 1
,' VOGTLE ELECTRIC GENERATING PLANT i
, RISK. BASED INSPECTION GUIDE -
Nuclear Service Cooling Water (NSCW) System Table A.1 1. Importance Basis and Failure Mode Identification 1 CONDITIONS THAT CAN LEAD TO FAILURE i Mission Success Criteria . S The NSCW system provides cooling water to various components in safety related systems including the Emergency Core Cooling System (ECCS), Containment Spray . System. Containment Heat Removal System and the Component Cooling Water System. e , Cooling water to various support systems is also provided by the NSCW These support ; systems include the auxiliary component cooling water system and heating, ventilation and - ' air conditioning systems for normal plant operation. The NSCW system consists of two independent trainst each train includes three pumps fo circulation of cooling water. The normal plant operating requirements are supplied by a single train. In addition. the plant emergency requirements for 30 days can be supplied by a single train. Within each train, total system capacity is provided by two of three available pumps. Thus, system success is accomp!!shed by operation of two out'of three pumps in one of the two available trains. The heat removed from the plant safety 'and support systems by the .NSCW is i transferred to the atmosphere through two cooling towers. Two cooling towers are , available, each of which has the capability to dissipate the plant heat loads generated during normal and emergency operation. Makeup water is available for the cooling tower , t basins from Well Water Storage Tank with river water as a backup. L Acc6 dent Importance 1mspect6on Dominent Failure Modes Sequence Catesory Activ6tles ,
- l. Failure of any of the following NSCW valves which iso- 1.5 H 5.M.T.C I
laios NSCW Gow to CCW heat exchansers CCW Heat Enchanser Train A Locked Open Valves .; Inlet: HV 11703 and HV 18704 ' Outlet: HV i1701 and HV 11702 I CCW Heat Enchanser Train B Locked Open Valves Inlet: HV 11616 and HV 11617 Ouuet: HV 11614 and HV 11615 ; 2 Pump train A or B out for maintenance 1.$ M M.T A 1. ' l 1 i I l
Accident Importante laspect> ' I Domiust Failure Medes Boquemee Category Activitles , ,
- 3. Pumps fall to start and run 1.5 M. S.M.T j Train A Pumps: P4-001 .
P4-003 l P4-005 , Train B Pumps: P4-002 N i P4-006 4, Pump discharge MOV or check vs!v: Ted to open or 1.5 M S.M.C , remain open ! Train A
- Pump P4 00l! Check Valve V4-025 MOV HW11600 (Normally Open) [
Pump P4-003: Check Valve V4 035 MOV HV.ll606 (Normally Open) l Pump P4 005: Check Valve V4-031' . MOV HWil605 (Normally Open) l Train B Pump P4 002: Check Valve V4 027 ; MOV HV.ll607 (Normally Opsn) Pump P4 004: Check Valve V4 037 1 MOV HV ll613 (Normally Open) r Pump P4-006: Check Valve V4-033 ' ' MOV HV.ll612 (Normally Open) l l s N t i
. A2 r
k
. VOGTLE ELECTRIC GENERATING PLANT RISK. BASED INSPECTION GUIDE .
. Nuclear Service Cooling Water (NSCW) System TABLE A.12 MODIFIED SYSTEM WALKDOWN - Train A j Desired Actual Pow. Sup. Required Actual Description ID No. lAcation Position Position Breaker # Location Position Position 1 Pump 001 V4 A10 Pumphouse 1.uked - > Discharge By. Train A Open peas Valve NSCW Pump , Room Pump 003 V4 Al2 Pumphouse Locked - - - Discharge By. Train A Open ' pass Valve NSCW i Pump , Room Pump 005 V4.Al l Pumphouse Locked -- - Discharge By. Train A Open pass Valve NSCW Pump Room Cooling Tower HV.ll601 Pumphouse Locked - - - A Sprey Header Train A Open , Isolation Valve NSCW : Pump i Room Cooling Tower HV.ll602 Pumphouse Locked - - - A Spray Header Train A Open Isolation Valve NSCW Pump l Room ! Cooling Tower HV ll603 Pumphouse tecked - - - l- A Spray Hender Train A Open l laolatiou Valve NSCW l Pump Room Cooling Tower HV.ll604 Pumphouse tec,ked - - - A Spray Header Train A Open Isolation Valve NSCW ! Pump Room NSCW Transfer HV.1670 Pumphouse lacked - - - Pump 7 Dis. NSCW Open charge isolation XFER Valve Pump Room l A3 P n 4
- - ~ _ _ . -.
. . j VOGTLE ELECTRIC GENERATING PLANT ', '
RISK BASED INSPECTION GUIDE , Nuclear Service Cooling Water (NSCW) System I l TABLE A.12 MODIFIED SYSTEM WALKDOWN - Train A (Cont'd) Desired Actual Pow. Sep. Required Actual Description ID No. IAcation Poellion Position Breaker # Location Position Position l l Au Component HV.ll70s Aut. Bldg. lacked - - - I Cooling Water level 1 Throttled l Hut Enchanger Room 159 l i inlet laolation Valve i Aun. Component HV.ll707 Au. Bldg. Locked - - - Cooling Water level 1 - Throttled ; Hut Exchanger- Room 159 I I Oatlet isola. tion Velve , Aux, Component HV.ll?09 Am. Bldg. Locked - - - Cooling Water Level I Closed HX.I Bypass Room 159 Valve CCW HX.I In. HV.I1703 Aut. Bldg. Lochec - - - let Isolation Level 2 Open Valve Room 203 CCW HX.! In.- HV.ll704 Au. Bldg. Locke3 - - - let isolation Level 2 Open - Valve Room 203 _ , CCW HX1 HV.ll701 Au. Blds. Locked - - .-
, Outlet laolation Level 2 Opm Valve Room 203 HV.ll?02 Au Bids. ; Locked
, CCW MX1 - - - Outlet teolation Level 2 Open Valve Room 203 , Cooling Tower HV.ll637 Aux Bids. '.ock;*d - - - Basin Supply Level B Upon laolation Valve Room 202 Cooling Tower HV.ll6sg Aus. Bids. Locked - - - Basin Return level B Open , laolation Valve Room 202 NSCW Train B V4497 Aua. Bids. Open - - - l 10 Train A In- Room 322 ter Tie teolation r Valve i NSCW Pump P4 001 NSCW - 1 AA02-04 Control Bldg. Racked in l 001 Pumphouse level A Room i RA4s i 1 1 A-4 l
)
. VOGTLE ELECTRIC GENERATING PLANT !
RISK BASED INSPECTION GUIDE- l Nuclear Service Cooling Water (NSCW) System 4 TABLE A.12 MODIFIED SYSTEM WALKDOWN - Train A (Cont'd) J Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaket # Location Position Position , N5CW Pump P4-003 NSCW - I AA02 08 Control Bldg. Racked in 003 Pumphouse bvel A Room RA48 NSCW Spare P4 005 NSCW - l AA0212 Control Bldg. Racked in Pump Pumphouse hvel A Room , ILA48 Pump P4St HV.ll600 NSCW Open lABB.22 Aux. Bldg. Closed ) Outlet MOV Pumphouse hvel 1 Room
!!8 l Pump P4 005 HV.ll605 NSCW Open IABB.35 Aut. Bldg. Closed Outlet MOV Pumphouse Level 1 Rocra 118 Pump P4 003 HV ll606 NSCW Open IABB.36 Aus. Bldg. Closed Outlet MOV Pumphouse uvel 1 Room 118 i
f J f ( l 1 ! 1 i l i A5 1
VOGTLE ELECTRIC GENERATING PLANT
- RISK BASED INSPECTION GUIDE ,
Nuclear Service Cooling Water (NSCW) System TABLE A.12 MODIFIED SYSTEM WALKDOWN - Train B Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position dreaker # Location Position Position Pump 002 V4 A16 Pumphouse Locked - - - Dischany By. Train B Open pass Valve NSCW Pump Room Pump 004 V4-tA Pumphouse Locked - - - Discharge By. Train B Open pass Valve NSCW Pump Room Pump 006 V4-A17 Pumphouse Locked - - - Discharge By- Train B Open pass Valve NSCW Pump Room Cooling Tower HV-Il608 Train B Locked - - - B Spray Header Pumphouse Open isolation Valve Floor Level Cooling Tower HV Il609 Train B Locked - - - B Spray Header Pumphouse Open Isolation Valve Floor Level Cooling Tower HV Il610 Train B Locked - - - B Spray Header Pumphouse Open Isolation Valve Floor Level Cooling Tower HV Il611 Train B Locked - - - B Spray Header Pumphouse Open Isolation Valve Floor Level NSCW Transfer HV 1671 Pumphouse Locked - - - Pump 7 Dis- NSCW Open char 8e Isolation XFER Valve Pump Room Aux. Component HV-Il622 Aux. Bldg. Locked - - - Cooling Water Level 1 Throttled Heat Exchanger i Inlet isolation Valve A-6
.- m <
1 2 VOGTLE ELECTRIC. GENERATING PLANT : iRISK BASED INSPECTION GUIDE Nuclear Service Cooling. Water (NSCW) System E . . '. , I TABLE A.I 2 - MODIFIED SYSTEM WALKDOWN - Train B (Cont'd) Desired Actul Pow. Sup. Required Actual ! Description ID No. Location' Position Position Breaker # Location - Position Position .- Au. Comp w nt HV.ll621 Au. Bldg. Locked r Cooling Weier Level i Throttled Hest Exchanger .i I Outlet Isola. tion Valve Aut Component HV.ll61 Au. Bldg. Locked. - -- - Cooling Water Level 1 Close HX.1 Bypass - 1 Valve c CCW HX 1 In. HV.ll617 Au. Bids. Locked - - - j i let isolation Level 2 Open Valve CCW HX.I In. HV.ll616 Au. Bids. Locked - - - let isolation Level 2 Open Valve i CCW HX.1- HV.ll615 Au. Bldgi Locked - - - Outlet isolation Level 2 Open , Valve , a' CCW HX.I HV.ll614 Aux. Bldg. Locked , Outlet Isolation Level 2 Open i Valve Cooling Tower - HV.ll689 Au. Bldg. Locked - - - Basin Supply Level B - Open Isolation . Valve Room B.it 'l Cooling Tower HV.ll690 Au. Bldg. Locked: - - - Basin Return Level B Open - Isolation Valve Room B.11 NSCW Pump P4-001 Pumphouse - l AA02 04 Control Bids. Racked in 001 Level A Room RA48 NSCW Pump P4-003 Pumphouse - I AA02 08 Control: Bldg. 003 Level A Room Racked in. RA48 NSCW Spare P4-005 Pumphouse - I AA0212 Control Bldg. Pump Level A Room Racked in RA48 l ~ . \ l A7 I 1 l l ! 1_--_--._______ _ __ _ _ _____
an' s . 7-s- 8 LmJ ..
=J :q; = "
_c.o. . :r
.r _
~
w . w .c 4 -
.s.
c : - -
- ... . . , , ~. .
. 5F M.',p. j + 49 >,+m. ,. . v*-
m
. . . .. . p-e.,s.e .
= ,n:7::M: ,Mn. .- s( .= ... . =>+ " pper -
.. J
, . i, y, ?e:w.
<q
. w v. , .n. 3;s:,
- 2;
.. i Y.r 3 m mr 3-- ;&r a - g ;q .u v 3,.
. n &.
. w ~ ~c ta.O.eL ; e. .,
~-
t.
) ~.
., r. F . .e nwn.q
- w
.) i:
p _.r,e .
- i. ? .
1 - .T _ p1 -
'""", **"w dL : 31 3+t
^
.41 <,m
' * ,. . c. rm .,
E ***E**'I'*
- M*
I ; g ,~. o_
~ i ' i. , , _
@ f .J J h, e'k.A * -i..'l y,\" (% .~.$ i a,'; %.V - " E
~,,,-s
' : - l "l=*
.& .; ;,, z a.
a s, ,,. a. .
.. i; ya;ee scm ,
ec.W.i M et?.4 W, <; P O 4,- ====== = l-l= = e-L, wq_
.g~ ~ '~
l . j w-wg o ,. ,ya.
- _ ;a. - o . .
- a. v m" .
E .- a v.1 . I 1 ; .m '
'l ~,* mn)-
1/r :- ? Q(q. zw-;,p g.t . 45 ' g
.: ..I :- GrQ;? . g .. ..
MKW)y;' J .* 1
-.$={ '. "' ):>b,,,
y;n gf. .
- c gn -
a
;-, -..- , .. Le , ,-
. . .w g.t.. u ,s .
_ - . " .p "*" " If,c w:.y - - -
~~
r, p s i i..e-. i . V-
,..sb . .%
n ..< . . - :..r.m,. . ,: .
.h _ . , <
~ . -
.w , m
- _ ;
,,"" : ' 5 0+ ,
- -1 ..-
, Y %,[; jy
.;Q, rY #* W. 4,
] . . v' w' * . '
aussesso ****"r n!; ' f, O i.. '#s
, u.
n n ..
.u.
.m_2. ,. m
.. o,.y n. .
w- -
. - ~ ' -- -'
p Wkp #_
~
- h. . - .h 9 .
*h'-Q iQ.y #
*E ca.
l , ;P :, - l MMLWQM 9 c ; *' i %1*K. ir-t m . Wl M- * ~ M;. > - 2- x::M.rnua =% : .v> r . I.? ":7 - l - I ,.
. / , t .L ~ ,
. . 7: u- <Wom43 >
%.s , s 1- _Q.-
1 j .~w, x.- y A + u. ne4.s "u ,~,.'- . ,; ;- .;m, -~a .e.
,A. ,;n--e s.
e./ 4 i w ~, us: 1 .s.
?q ,., .; 'Sp . 'sfy .+.M s 1c; lc t -- g.c r"X_ t - .,. g ,
. . g Figure A.1-1; Vogtle Nuclear Service Cooling Water System (Train A) Showing Ceumnent Iecations
~
(Reference SAIC 89/1540 DRAFT. Figure 3.9-2)
. e '. "
4 e
.,_......__,...____________,____._,;....___,,,,,__,__y
.a...
-,,# . . , , 7..n...;.. n..c.._,g., . _ _ . . , , . , , . ,
.. .sp,.+ .....;..;..,.s.., .r,,, _
___.a __
e < ==3 o k===l t=J
~: .. ..
-] - _
p- I
*- , , . , ==j i --.i
,y-;
., _ -i- - -
/, - uw--. . . -; ,
~ s
--TC
- t. ,,,=
s ; e
.{ _
- .s.e ;. - ;
'E i
t
,I -
- ; 2 -
- ' M- ' - r-t:.i n, 7 S
Q .. - g
@T 1:
~
t
' - a 1= _
jy_ _" t f .. 4 -
*- l_ a __ g .-
q ,nc C- 'g 9s e a w g{ - . i c g :,.e-s
;J -- _ - - t - ; ' 1, 7 _. , s ._; -
_va , .3- p, -
; f M.,
e ' I .. t=3 t= D._ s - ,,
;.;j,; , nr . ' 1-..,* . - q ..
-vt - R 7
v n ,,. ; : cv g, , ,. ,
& 3r. g ,. 2.g :. . q . ., ..._ , . .. im .m ,a 3;;;;) ~ q A.
mm . W;M
.m
~~
, f. e Gi' , :a
,. M 7-b ;,
~ .
'~
- n. , / ?
-.e. -. .
. . - ,.~ _=._
s s . _, ._.._
- ~-i - e-
, -.. -a . ,- .
~ ~A W ?
.g g a g g T gal.....lllll
~~ -
.. >w -- l l l l l J.. L . -
_ 9 - - w a- _ _._ l, l q m
.p n .- .-
= ,
.. . . , ~
~ ' '
; g n a9 ;-
y
--r.,
- ;+ s}
; - %., - ;s
~
5
, F"f,,, ~::::-
]i e .
- 1. l
-s
- y. e .- ar :9? - < e .c. .
t g.
- .7,,- :;,, .; y..
e;x.
. ./.. . e-n; > y; . l :q ' lnc
.7 9-l ., . ;. -.
,_ .4 .
g f., ., s i . 4 Figure A.1-2 Vogtle Nuclear Service Cooling Water System (Train B) Showing Component I;xations. j (Reference SAIC 89/1540 DRAI T, Figure 3.9 3) - i e _ _ _ _ _ _ _ _ . _ _ _ _ . . . _ _ _ _ _ _ _ . _ _ _ . --_____--___-_-______--_--._______--_________________-_____.-_______._-__-r_ _ _ _ _ _ . _ - _ . _ _ _ _ _ _ _ , _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ - . . _ _ _ _ . . - _ _ _ _ . _ - . _ _ _ _ - _ _ _ _ . - _ - - - . _ . . = . . . _-
TABLE A.12 (Cont'd) ', . ! REFERENCE DOCUMENTS _ a TITLE I.D. NO. REV DATE'
- 1. Vogtle Units I and 2 Final Safety i Analysis Report Section 9.2' i
i -2. Vogtle Electric Generating Plant Procedure 11150 1 3' 11/19/98
- Nuclear Service Cooling Water System Alignment"
{Vogtle Electric Generating Plant Procedure 13150 1 3 2/8/89
" Nuclear Service Cooling Water System" l
l
.I i
l
\
l a 4 A-10 ' I 1 l
.f a . ' - 1 J
.' ' VOGTLE ELECTRIC GENERATING PLANT- j RISK. BASED INSPECTION GUIDE e l J
i l Normal / Emergency AC Power System j Table A.2-1.- Importance Basis and Failure Mode Identification ; l CONDITIONS THAT CAN LEAD TO FAILURE l Mission Success Criteria The Normal / Emergency AC Power System consists of eight 4160 VAC buses, several- . 480 VAC buses,120 VAC vitalcinstrumentation' buses,125 VDC, buses,' two _ diesel generators,;and-their associated motor control centers, breakers, transformers, chargers, . , inverters, and batteries. The 4160 VAC buses are subdivided into two Engineered Safety Feature (ESF) buses . and six non ESF buses. The ESF buses supply power to the plant ESF loadsLand-are reduced to 480 and 120 volts for' distribution to smaller loads and the insuumentation and. control power systems. Examples of the electrical loads supplied by the ESF buses include:- - l 1. Nuclear Service Cooling Water Pumps
- 2. ' Component ' Cooling . Water . Pumps ;
1
- 3. Safety Injection Pumps
- 4. Residual Heat Removal Pump-
- 5. Containment Spray Pump
- 6. Centrifugal Charging Pump -
1
- 7. Motor Driven ' Auxiliary Feedvater Pump q
- 8. Battery Chargers
- 9. Various Room Coolers
- 10. Nuclear Service Cooling Tower Fan In safety related systems with redundant trains, power is supplied to each train from different 4160 ESF buses.
Power for on-site distribution is taken from the two auxiliary transformers (VAT) or' the two reserve auxiliary transformers (RAT). The 4160 VAC ESF buses- are fed from the i l RATS and the emergency diesel generators. In addition, each 4160 VAC ESF bus has a . crosstic to the RAT of the opposite train.'Following a loss of voltage on the 4160 VAC buses, each bus will she'd its loads and the crosstic breakers will trip.-An'undervoltage ' condition on the 4160 VAC ESF buses will automatically start the diesel generators and close the output breakert A-l 1 li
,o C__________ -
- i.
- L s
.. '{
j I Accidest Importance Inspectha ' Activities l' Dominaat Failure Modes Sequesee- Category ,.
- 1. . Failure of EDO: to start & run 7,8,9 H O.S.M.C Diesel generators I A or IB ,. ]
,2. EDOs unavailable due to test or maintenance 7,8,9 H M 7hs unavailabilltv of one EDO requires additional testing j of th'. rasisimng diesel with I hour and every additional 8 - '
hours, l 3. Failure to restore AC power after stesion blackout with 8,9 . H O. L concurrent RCP. seal LOCA
- 4. Failure of automatic bus transfer to backup source of AC 10 H M.S.C.
- ' power resulting in loss of vital AC bus
, , e-l ' Six 120 VAC vital instrument panels are normally supplied power hem inverters with backup power supplied by.a transformer,7he normal supply and backup breakers are . nechanically interlocked such that only one breaker can be q I closed.
-1 Normal Supply Alternate Supply Panel ~ Breaker Breaker j lAYlA lAYlA-02 1 AYl A 01 IAY2A IBYlB l AY2A 02 IBYlB-02 I AY2A 01 -
IBYlB ] IBY2B IBY28-02 IBY2B 01 l ICYlA ICYl A 02' ICYl A 01 l IDYlB IDYlB.02 IDYlB 01 5 - Improper EDO post maintenance valve or breaker lineup. .7,8,9 M O.M - Circuit. Breaker IAA0219 Train A (Switchgear I AA02) and Breaker IBA0319 Train B (Switchgear IBA03) must-be open during normal operation ' .. Circuit Breaker IAA0205 Train A (Switchgear IAA02) and Breaker IBA0301 Train B (Switchgear IBA03) must be racked in during normal operation
~ Manual operated valves HV Il705 and HV ll706 (dol A) .
and HV.11619 and HV 11620 (DOlB) must be locked
'open for normal operation. (Diesel generator cooling water -
valves.)
- 6. Cooling water valves for EDO closed or blocked 7,8,9 = L- S,M,C Im:ked open manual valves DOIA: HV 11705 (Supply)
HV.ll706 (Return) DOlB: HV.ll619 (Supply) e
' HV.11620 (Return)
- 7. Failure of EDO output breaker to close 7,8,9 L- S,M,C -
DOIA: Breaker l AA0219 (Normally Open) DOlB: - Breaker IBA0319 (Normally Open) s
- 8. Failure to transfer to reserve source of AC power or failure 7,8,9 L S.M of EDO start signal l
, .j i- -
A-12 L
,s .,n ,,.
4.
,' Accident lenportance laspection Dooninant Failure Medes ~. Sequence - Category Activities ,
9, Fallwe of Inveriet 6,7,10- L, M.S.C -l Circuit Breaker - i laverter (Racked le) I 1ADill l AYl A 01 1 ADill! IAY2A-02 1B0112 IBYlB 02 IBDill2 IBY28 02 1CDil3 ICYlA42 IDDil4 IDYlB-02 I i I l 1 1-I
-1 i
t d A-13
VOGTLE ELECTRIC GENERATING PLANT - RISK. BASED INSPECTION GUIDE , , !. Noimal/ Emergency'AC Power System. i TABLE A.2 2 MODIFIED SYSTEM WALKDOWN Desired ' Actual Pow. Sup. Required Actual , Description ID No. - Location Position Posidon Breaker # Location Position Position DO 1 NSCW HV.ll705 DO 1 A - tecked. - - - Cooling Supply. Room Open isolation Valve DO 1.NSCW HV.11706 DO l A' Locked, - - - Cooling Retum Room Open isolation Valve DO 2 NSCW HV.ll619 DO IB Locked - - - Cooling Supply Room Open Isolation Valve DO 2 NSCW HV.ll620 DO lB Locked. - .- - Cooling Return Room Open isolation Valve Supply Circuit Breakers to 120 V AC Vital in. strument Panels lAYlA - - - - l AYl A 02 Control Build. Closedl ing Level B-LAY 2A - - - - I AY2A 02 Aux Building ClosedI - - Level i IBYlB - - - - IBYlB 02 Coritrol Bulld.- Closed I ing Level B IBY2B - - - - IBY28-02 Aux. Building Closed-I Level i ICYlB - ,
- - ICYlB-02 Control Build. Closed l
ing Level B IDYlB - -
- - IDYlB-02 Control Build. Closed l
ing Level B Alternate Supply Breakers to 120 V AC Vital in-strument Panels lAYlA - - - - l AYl A 01 Control Build. Open2 . Ing Level B 1AY2A 2 1 AY2A 01 Aux. Building Open Level i I Assumes panel is energized 2 Assumes primary breaker is closed.
. A-14 i
I 1 J
0 . 4
. ,' VOGTLE ELECTRIC GENERATING PLANT ,
, RISK BASED INSPECTION GUIDE: ..
Normal / Emergency . AC Power System .
,f TABLE A.2 2' MODIFIED SYSTEM WALKDOWN-(Cont'd)
'i Desired Actual Pow. Sup. . .
Required Actual Location - Position Position
'Descriptlon i ID No.' Location Position Position Breaker #
]
IBYlB = .
- - .- IBYlB 01 l Control Buildi Open 2 .
Ing Level B ] ' 2 Aux. Building IBY2B ' - - - - IBY2B OlL Open r Level I ICYlB - - - - ICYl A 01 Control Build. ing Level B Open2 ] ; IDYlB - - - - IDYlB 01 ' Control Build. Open 2
] '
, ins Level B . "e Normal Power - - - - I AA02 05 Control Bulld. Racked in to 4160 V ing Level A ' Room A48 Switchgear Normal Power - .
- - IBA03 01 Control Bulld. Racked in to 4160 V- ing Level A Switchgear Room A50 i Emergency - - .- - IAA0219 Control Bulld. Open
[ Power to 4160 ing Leiel A V Switchgear Room A48 ' I Emergency -
- - - IBA0319 Contrcl Build- Open _
i Power to 4160 ing Level A l' V Switchgear Room A50 l l
. a t
L 2 Assumes primary breaker is closed l P . A-15 m ( ,. . .g , l i
" , - - - + - , . , _ _ ,
1 L l c- -
=a ..
= =.g= . == === ~g *=
gj:ceee: s u., wee x w
= w-ce. ..
a, . : _: I I) lI (I m,.hgM .
> . .s ..- i _
.s a
- ( . - , ,
m - ;;ae. .
,,. e + .e e.
; e. e ,
- a u n- ~:3 x
p,; -: y.h em. . s ;, -
,,. < . .' I ,g y o o
< 'cc p;e o- o o_ 1 - o 3wa*=o(;
- . ew 2, ~ f . .e.>
; r, > ;, -
o :
- f. [. [ t
+
- e. . ; ..
~ ' y . g
+s *
- - c=. - -
w. Figure A.2-1-- Vogtle Station Electric Power Showing Component locations l (Reference SAIC 89/1540 DRAFT Figure 3.6-2); e
-p
_~ ., -. _ . , , . w_ _, , __ __ _ _ _ , _
-<--.m.~. n.- +
e
j - i
.,j
. I 8 ,
'l 4.
****~' E ~
.s n.
esset p -
..m
-vammaman
~
g.] L i..'& **"* I.,' '~ +.., . ,-. ;m\ f.1 k M.EW8,88 , eo ~. ** . t - ~--- . .
,,, e,,"- ,.,
c., 6
, ,, % .51 -j . ,- +-
-(i,.3, -
, ,:1 ' .,c.
g . . . , 9 -
? $@O M + * ~ .1I .
- j. ,,_ . , Y' ;; sp r m.,. .. . . , ..
- y, y, . G. .a s2-
, gP95 eamag .
s < * . : - '
- t. . ,.c_. ~
1
-m , . . >:
.lF,,, - .
447), ;I I .,. . - , l : ll7- u w
- 3. r .ll s'y.+ - 1 il -
1 4
..et' ,
_ _.. . _ , _ ~ , _ .- ,
,.-.e _... .a
, . e = = .1 -
u
, n -
+ -g; 2.~ 4
.m o .
-; o?.
J'o' _
. 2 - o 3 t_; ,
c... ,
- i. a eeB.3WI.OGAA USD. .sg.9es.0EE. b4 4EB.9muSDGam enSEB agg.9mpeDa05mm ess.e
%, e. co- ; --
.c=-
s 8"" $
- i. N
- A~ ,r +
4 s w o 2
-n-, .
t o.
- o_ n.. o ._
i , +m 1 , 1
* ' ' ' . ; - - i p ; 3 e=smoes l l soo -~ '
4 l- ,,r _ , _( , o >.
. m w . o.
- o 6m _m y ,
.ce .a -
x
-+. . 3.c 1-
~ ,. 3. z .r.
. . . y== .. -
- u. - . . . c. , . m. , a. .. ,
5~. x ~ n_
. n- 7 , _ ,
, .~_
4 s 1, , l . +-
. ' p1.} s. r 1 q.
7,-. ,a s <jn , - i r .,
+1 i'< . -,: -- . - ~
ip; i
,i i ., ii ip
- - -, , ':,, - : r, r * ,: . .J' ,
.. g' .
.-+ .n. % ,, i j + =. - ce .cc .cc
) * . . , . emar sees eagg em.f 1
_ -x. . ;,- ) e = ==> e = ~==
, - e=~
*, . .o , .:
. . . ,. ; . ~. . ..
l
..c.
- e.
- J
.- , ... i i m esm. ases.n ev ueessswints casis ammaam mes=aanno 1
d t 1 , Figure A.2-2 Vogtle 14160 and 480 VAC Electric Power Systems (Train A) Showing Component locations 4 (Reference SAIC 89/1540 DRAFT, Figure 3.6-4) l l .I f-i
n;.
. < - sarte -
- . * , .+-. .. ~j; - ;ief, ". !;L'.;; 3,
;;;,;w - g . . sesst
,4 =mesonmme y .
w i .. s -", ,' _ + sende
.,. sa 4 ,
,7., ,+4 ';n' #;a S.S.Insom
~ geoges e, 2
' a
+ .
T
,- ,- a Wa , -7ercsesausu 1 i ,
_; .u 'f ' ,",;' ,
' ', - s ':
f
-T S f; -
&,..; c3 .g w-. c.a a go es m d ' , tr u 4' t
~
G ;" ...e j.. pi _ w :-/.,t ql*
-y[
[f-'g . 7 ! *
- , L /' q
/Y
_1 p- .. . , + *
.1 g y ,
f% - I, eenmoes . aee
... .easese _m eem , - **=s** m ,
Mw_uA gaggg , .- M m Wees ,
N'* M arw' M em .
g , ::x a y ,# : e y; - >,
"b.1 emmensen - anvase ns. :; " - armes eg
.. , - . . . ; t
., % w9A 2-. .g.. j *k - - - '> ~
- h, l
+ ,
. .s
^
1 I -
' h (l .
1 - i Ii ~/.s.: > ll 1
.a.
asse sse.cneast asse s msgoman eeses assesanoman gesy .- je smene seas ese - as as meseg - assegpendoe,rs og esseas,esemn
%. 4, W ,
g *
++~- - . , +- , ;'$ .:,
({ , ._
;1 B' i I, .;1 I*, -I l;- 1l -
r - .!,,]qI l/j . .. ,
.f -
+
4 h
- _z.
=
y r,; c;i 4 #.
$ 4
,% e \ ;. p fp 3 ' '. h , ','T I e :1I - t f '-
e #,. --
.s, jYr, ; J- 1 #
s c - QQ:.. , + . .;.'.sa,.. 3, h W" .* ~ \ > '
',4 g.> u.1.r- . b - [.L 1u. - s am ans,-cc .-
- y ,,.em
.m n m..e La c y ,-
, -:, y e- ,, 2
.m
~
- 7) ; "1-a l- {
s ' ? -- - a , e o~ * .ii
. o - <, 4 r ;
!. . . ...; . ,. .c..
I 7 ,1 -% y?,-:.-e s ; 2 ' . , + . .-? ,- { u am
.{.. : ? 1g- [; ,F ' (< , g.m. Ejd.;!; ; ? '.' ~-'i., gh. assy < s a - t g . - .
-a .
. .eeu, x; ans,v escC ess,,w ,,ests .. ur eggms.s .
4 q .tA y , j ., ~ _ f; ., ...s, ,
;;?'~ azy g~! ~ :j.&f} ;5g-R.h:5 .f: ' *
.a c . e
' Figure A.2-3 Vogtle 14160 and 480 VAC Electric Power Systems '(Train B) Showing Component locations (Reference SAIC 89/1540 DRAFT, Figure 3.6-6)
. - . S .
e
' - - - ' ' - - " " ' ' ~^'
~1 TABLE A.2 2l(Cont'd):.
REFERENCE DOCUMENTS TITLE I.D. NO. 'REV DATE , { ; l- - 1. Vogtle Units 1- and 2 Final Safety Section 8 - Analysis Report l , j
- 2. Vogtle Electric Ger.erating Plant Procedure = 13427-1 9 11/15/88 "4160 V AC 1E Electrical Dietribution .
System" l
- 3. Vogtle Electric Generating Plant Procedure 11427-1= 4 2/2/89
. "4160 V AC- IE Electrical Distribution: _ System Alignment"
- 4. Vogtle Electric Generating Plant Procedure - 11431-1 '2 9/16/87 "120-V AC IE Vital' Instrument Distribution
= System Alignment for Startup and Normal Operation"
, 5. Vogtle Electric Generating Plant Procedure - 13431-1 3' 12/20/87 "120 V- AC IE Vital Instrument Distribution ' System"
- 6. Vogtle Electric Generating Plant Procedure 11150 3 11/19/87
" Nuclear Service Cooling Water
) l System Alignment"
]
L o 1 4i l l 4 1 1 e
- A-19 l l
4h , ,. i
+ VOGTLE ELECTRIC GENERATING PLANT: '
RISK BASED INSPECTION GUIDE ~
- DC Power System .
Table A.3-1. Importance Basis and Failure Mode Identification i ( CONDITIONS THAT CAN LEAD TO-FAILURE
, Mission Success Criteria.
Thell25 V DC Power System provides a continuous source of power for control, instrumentation; and operation'of DC motors. There are four safety. features 125 V DC
. systems (designated as A; B, C, and D). Each system includes's battery, switchgear, two ;
redundant battery chargers, inverter (system A and B have two inverters),- and distribution .. _ panels. Systems A and C form the Train A safety features DC system, while systems B and -
- D form Train B. Trains A and B are each. supplied power by two Class 1E-480 volt AC- 3 motor control centers through battery chargers to the 125 volt DC switchgear. Backup power to the DC switchgear is also available from the station batteries. DC power is fed from the switchgear to the DC distribution panels, the DC. motor control center and the inverters (for supplying power to the 120 volt AC vital instrumentation panels).
Accident Importance laspection - Dominant Fallare Medes Sequence Category . Activities
- 1. loss of 125 V DC bus 6 H - O.S.M.T.C --.- ,
Normally Closed Circuit Breakers '
)
System A:
. l AD106 (Battery charger I ADICA to SWG l ADI) .
LAD 107 (Battery charger lADICB to SWG IADI) 1 IAD101 (Battery to SWG ~1ADI) ~ ' System B: . , , IBD106 (Battery charger IBDICA to SWG IBDI) IBD107 (Battery charger IBDICB to SWG 18D1)
. 188101 (Battery to SWG 18D1)
System C:
. lCD106 (Battery charger ICDICA to SWG ICDl) i ICD 107 (Battery charger ICDICB to SWG ICD 1)
ICD 101 (Battery to SWG ICD 1) System D:
, 1D0106 ,(Battery charger IDDICA to SWG IDDI)
IDDIM - (Battery charger IDDICB to SWG IDDI) IDD101 (Battery to SWG IDD1)
>1 e
A-20 q u ,.
~ Accidest Importance laspection ;
- Desninant Failure Modes Sequence Category Activities .
l 2. Failure of on line charger and failure of spare to energize 6 ~M' S.M,T,C on demand i ! ~ Sysum A: Batwry chargers IADICA and IADICB l [ System B: Batory chargers IBDICA and IBDICB ' l' Sysum C: Batwry chargers ICDICA and ICDICB
' 3.. Operational sest or rnainwnance error resulting in 6. L O.M -
a) doenergizing or cascading of DC power supplies b) failure to properly restore bensties or charger aner ; maintenance
- 4 Pallure of Batteries 6 L M,S T Beneries are susceptible to failure during long term opers. (
tion: for example, following station blackout. System A:-- I ADIB System B: IBDIB a- System C: ICDIB Sysum D: IDDIB 5, Loss of bettery room. ventilation 6 L , M.S.T . This failure mode is more significant during events requir-Ing long term use of baneries for emergency power. v I I- , J l d g l . - s. 5 . A-21 4 .
. , - r.., , . . - . . ,~, - - n -
^ '
VOGTLE ELECTRIC GENERATING PLANT . RISK BASED INSPECTION GUIDE '- DC Power System - i
,3 TABLE A.3 2 MODIFIED SYSTEM WALKDOWN =1 l
. . Desired Actual '. Pow, Sup, Required Actual :
Description. ID No. location Position Position Breaker e location Position ~ Position i Supply Power to 125 Y DC Switchgear . (1AD1) i From:
- 1) Battery- - - -. -- 1 ADl Ol' Control Build- Racked in 1 Ing Level B . ')
Room B52~
- 2) Battery
- - -- - I ADl' 06 Control Build- Racked In Charger ing Level B (IADICA) Room B52 ,
- 3) Battery Charger I ADl.07 Control Build- Racked In j ins Level B i (IADICA) Room B52 Supply Power 1 to 125 V DC Switchgear .
(IBDI) , From:
~
{
- 1) Battery - - - -
1BD101 Control Build- Racked In (IBDIB)' ing Level B 1 Room B47
- 2) Battery - - - - IBDI 06 Control Build . Racked in !
Charger ing Level B. (IBDICB) Room B47
- 3) Battery - - - ..--- IBDI 07 Control BuilS } Racked in- .l Charger ing Level B '
Room B47 -: Supply Power to 125 ' V DC 1 Switchgear (ICD 1) From:
- 1) Battery - - - - ICDl-01 Control Build- Racked in (ICDIB) ins Level B Room BS$
!, 2) Battery - - -- --- ICDl.06. Control Build- Racked in Charger ing Level B (ICDICA) Room B55 i e A-22
+
e VOGTLE ELECTRIC, GENERATING- PLANT i- , RISK BASED INSPECTION GUIDE , l; DC Power System \: TABLE A.3 2 MODIFIED SYSTEM .WALKDOWN (Cont'd) 1 Desired - Actual Pow. Sup. Required Actual , Description ID No. . Location Position Position Breaker # . Location Position ' Position
]
- 3) Banery - - - - ICDI.07 Control Build. Racked in '
Charger - ins Level B (ICDICB) Room BS5 Supply Power to 125 V.DC - ,
]
Switchgear l (IDDI) 'l 1 From:
- 1) Battery - - - - IDDI.01 Control Bulld- Racked in ,I l
ing Level B ' Room B48 , 1
- 2) Banery - - - - 1DD106 Control Build. Racked in .
Charger ing Level B , Room B48 j
- 3) Battery - - - - IDDl 07 Control Build . Racked In J Charger ~ ing Level B Room B48 I
\
l' l , l l 4 i emes. 'eme Am23 k-h
~
N..
.m . .. .
- r. -
t 4-
**" m g ,g . g I 5** I ,
I
+
OIE M N
- j
, b .,':
ONEM
- WP es sests ggg g . 4M p ggM.g.
Se
.'* } - SP St em 2
....4.
. ~<.v.. .- ur "
..n.,.* -
j 2 4 -, -= , ., l l...'l ., l * ' ,i# s ll' y l[ 8 ,j l lg gg y e P* %
- , j-
- ,, j , ., - 4
- a 7os .
~
_-:..~ .-oe
. :c .; -
.enm.
. . .e.n.ma.r- -
c mc. .. .e.m., M.c.. is.oc. .i ,.m . :a c. = = =.ca. o.c. - eis r-c , . c..
. amac=> 1, .. : -
l ..k i i 4
.wik n /A n u ~ ,
.Tp.
pT N M ~7 'nf n ,
'-l n) N; . M9:
- 5 %% ~ - T .
.J- ' . ; y C, v.c ,e > -
,f1 I- >c : ~4 1M/ I Ii]R ;L . _ .r-II i I^ 4
,jt iI' 1
F$ h . j -
~ e.- N.
a c. au ; ' oo. . a:. 2. .
' ', s,,,-
r es.cm,. ' se as enens ..., .,(
%y + a,
;;w -
,, c. . :: .m n n n . , .- v. -
- wJuctA o,3
~ i.: ,
N ;
~
e- - h m%,in f.y v: WE.n; .x , p:it.w*; = . a
+-
a-9t;. . g '.
.n .
~
- n'
^
gJ c;w u% + n c;.P n ,,7.n:. . m.. ;. . r
..n.++ ,.<
w ,y Q ~I n
- x. . ) , m,~<.~, .4. '
s rvb- e .- J
~ [' ,r
%g ,
p- 4 . " -@%*'" .D o
,k." ', -*
[.'kQ 3"", - a<' .,Q.
^ .P,y ' , jj k. b4' . d. i * .
'>ll
. . ~y7
, .f ' I *: ..;7.. _d i.l . g[ .,
c [{ '.hI,", $ i* ,e *F f
= .[t - ; ,e,'...
, , (g'[ ., - ,
,7 7.^"* ..m,".%-
- .x 5
, ' "l
. .Re . a,a v.: u ;;: n , s ;
sts tg g . :e g L
.ec RRogs a , y. e;
+ n c'd*
- m~ - . .c.
."'+. , :
w .v. - . <., s. , .. v m.4.:: .p . ,-.x.n :: - z, e.o.,m, ;.> y g c-,, ? ;
- x. . ; . ,. - -~ i
* ""E'
- x. ,,['h.
g M I. .,_((.. . . , . 2 w-- -8N W . ** ~ .=ms .L* -. = r .
.m,h Nn
-Y W .S.EWS, e
== =
.-,.v= m p/m+.e :m. . .
w ,~ ;- - y , . e . e, - 44 - 4 . nyit.m. . py'; . M.,ia G ., p ... , n . yw --%.a. ,; p -,. 4.h.mm&. W,N. ,- .W., ; y:a(G. 1 r w ;- 9 ,-~y l l r -- ,
> .m 1 \ -
n d .e, -- +: .-a ,u
;; e u u . ,
w.w~ cu s , . a .. A = = = ~
% ;:-f .=====.-
y ac (.w;;.. -
=====,c= :
.+
n**.at .avsee
- ,; n y};; .a. . - % s, _
-* ,7 c ..
- u. ....- a
.. s move teseanwnoranverse.ri.uscasts sameas-__ _
Figure A11 Vogtle 1125 VDC and~120 VAC Electric Power Systems (Train A) Showing Component locations (Reference SAIC 89/1540 DRAFT, Figure 3.6-8) . -
. e e-4 a
x._ _ . - -- .- n. , - , - 'a s .. ..y-- .,-..~......u .m ....,w-. ,v. m - m,y p w ____w...y_ :_, .
.,m - ~, , - . . -
l --
- i
******* I '** I I = _I l = g g . g ,
- a c u: c c ac ma
-l__ m..ac -cc .ma l :. yl ..ac an.z . l l l l .. l o;- o
_ o - o; -
. o. 4, 4
+ , ,
^ .
- ~
..t. ._. ,
- " g ,
, _ 1; ,
I
;-f ., .,
, es.m..er . c =ze E -
m o.cma.c.ae
. ===.ar . ,.e. nmo.v s
' t e.s.me.r ; .
- c. .
- c. . 3,
; se ..- c. . . .. .e t +
lisiI h]i. . h3,5 _ li:E .? ' fh]i, h] .. ._ 1 .
- . s,
~; , m .
,[];~ ._ i -
i:.. .-, :. . .s c, or y .: ~ , . p oc 3 - n. y;:.? f. , j o:
- n. . o . , , .
a. x: ,
'l l *
'iI l?
l gIB45c.sustcastaan etDe -
.g gP es appig l- i
*l .3m act SP SS sawet.a.scs 3ee na esoe g'- [ ~..a 7
-4 ~5 Y: ,.
- n . .
h>
~ ~
'^h.
e
D . . .D ; _
e }'~ ,
'N '
N , 'il a i l , ,,) ,I. . ll . [ lI a" I n" [' "._:, i y c 7 z u , ,
- + y.,
4
- _w
.. . , s.
- e -
" t r. H - - -
{ l Ji t- ' ,-l (JF% ,;L s. ? , f.]/-) p .; -
- Ii g,; -
' 2 ,
' , , , i1 ',
I i }< !. w. ,
. :n y , 5; ' y- ; -
,g. l WCE tED.If
*, I. 'T 7,
*.- l .stE m,c sel l [g 1T_S..A y
^
r
,. e j (T [ ~ Q$ 1 , '"
. 'o. = . = =w.
;. 4.m=,=c=o.a ci. .- .
J. --c-
. ia s: . .
-ni-- e .s.m.. ..,..
t
,. ._g_ .;
< = , s. ; ~ ' ' , -,
, 4
< t n-
- s,II- -
wo 3 i s' ..
. - + .1,
.u
.- l , 11 -
r,
< ,. + s, ,
i
.c . . n . c si m ..
c . t *:- -- l
, = .
, - _l-
.. ,_ l .. ,;
. I. <=- l. a.
i, i
- m. 6
.s
.t c .
4 f $ Figure A3-2 Vogtle'l 125 VDC and 120 VAC Electric Power Systems (Train B) Showing Cei.4,Gecat locations ~ j' (Reference SAIC 89/1540 DRAFT, Figure 3.6-10) I a
-~...%.v.%._ m ._ _ . ~ . , . , _ _ . _ _my . _ . .
. TABLE A.3 2 (Cont'd) '.
- REFERENCE DOCUMENTS TITLE 1.D.- NO. REV DATE 1..Vogtle Electric Generating Plant Procedure 11405.I' 3 11/4/88 "125 V DC IE Electrical Distribution System Alignment"
- 2. Vogtle Electric Generating' Plant Procedure 13405 1 5 5/11/88
- "125 V DC IE Electrical Distribution System"
- 3. Vogtle Units -1 and 2 Final Safety : Section 8.3 Analysis Report 4
A-26 h.
c 3 . VOGTLE ELECTRIC GENERATING PLANT ' o RISK BASED INSPECTION GUIDE' Component Cooling Water System (CCWS) Table A.41. Importance Basis and Failure Mode Identification CONDITIONS THAT CAN LEAD TO FAILURE
+
Mission-Success Criteria The component cooling water system is a non-radioactive, demineralized water, closed . loop cooling water system 1which removes heat from various plant components and-transfers it to the Nuclear Se vice Cooling Water System. The system consists of two trains, designated as Train A and Train B. Each train utilizes three pumps and one heat ' exchanger; each of the three pumps is rated at 50% of the required system capacity. Thus, mission success is provided by the operation of two pumps in one of the trains . The reactor components cooled by the CCWS include the Spend Fuel Pool (SFP) Heat-Exchanger, the Residual Heat Removal (RHR) Heat Exchanger, and the RHR Pump Seal Coolers There are automatic isolation valves in the CCWS flow path; thus, flow is continuously maintained through the SFP and RHR systems. However, heat removal by the CCWS is required only when SFP or RHR are in operation. After removing the heat loads, the CCW flow passes through a heat exchanger that is cooled by the NSCWS and the coolant returns to the suction side of the CCW pumps. Each train of the CCWS has a surge tank that provides for expansion and contraction of the. coolant due to temperature changes.--
. . Accident importance laspectico Dominant Failure Modes . Sequence' Category : Activities
- 1. Pumps fall to start & run 1,5 H- S.M T.C -
Train A Pumps: P4-001 P4-003 P4 005 (Train A Standby) Train B Pumps: P4-002 P4404 P4 006 (Train B Standby)
- 2. t.ocal fault of heat exchanger valves that isolate or se. 1 M S,M T.C verely reduce CCW flow or NSCW coolant flow A-27 m -
+
[ .
,i l Acendest .
. Importance I-;:h .-
I Deniaant Fallere Medes- Sequence Category - Activities Manually operated, locked open valves Train A' Heat Exchanger
> Inlet: HV 11807-Outlet:. HV 11806 ,
Train B Heat P.schanger inlet: HV 11815 Outlet: HV 11814 -
- 3. Pump discharge and auction valves fall to open or remain ' 5 L S.M.T.C.
oPen. 3 Trala A:- j ' Pump P4 001 -. 1
- Section: HV.ll800 Manual, locked open Discharge: HV.!1803 Manual, locked open -'
030 Check valve
]
Pump P4-003 ,
- Section: HV.ll801 Manual, locked open + l Discharge: -HV ll804 Manual. locked open ;
032- Check valve 'l Pump P4-005 . Suction: HV.ll802 Manual, locked open Discharge: HV ll805 Manual, locked open 034 Check valve Train B: ! Pump P4 002 i Suction: HV Il808 Manual, locked open Discharge: HV ll811 ' Manual, locked open 055 Check valve Pump P4-004 l Suction: HV.11809 Manual locked open i Discharge: HV ll812 - Manual, locked open .t 057 Check valve ! Pump P4006 Suction: HV Il810 Manual, locked open Discharge: HV Il813 Manual, locked open 059 . Check valve
- 4. Pumps out for maintenase 5- L M,T Train A Pumps:
P4 001 P4 003 P4 005 (Train A Standby) Train B Pumps: P4 002 - P4 004 . P4-006 (Train B Standby) A-28 I
l i
.VOGTLE ELECTRIC GENERATING PLANT
/ '
-. ~ RISK. BASED INSPECTION GUIDE ~ -
Component Cooling Water System 1 J' TABLE A.4 2.. MODIFIED SYSTEM WALKDOWN l Desired Actual. Pow. Sup. Required Actual Description . ID No- Location Position Position Breaker # < Location Position Position ; Racked in
~
, Train A CCW - P4 001 Aux. . Bldg. - I AA02 03 Control Bids.
- Punips - Level A Level A Room q
' A76 3 I P4 003 Aux. Bldg. - -- I AA02 07 ' Control Bldg. Racked in Level A - IAvel A Room m, A76-l* I- P4 005 Aux. Bldg. - I AA02.ll Control Bldg. Racked In Level A Level' A Room A76 ; Traln B CCW P4 002 Aux. Bldg. .- IBA03 08 Contml Bldg.- Racked in Pumps Level A Level B Room , 1 i B61 P4-004 Aux. Bids. -- IBA0312 Control Bids. Racked in Level A LAvel B Room L. B61 l-P4-006 Aux. Bids. .- IBA0316 Contml Bids, Racked in , Level A Level B Room B61 .I l Inlet Isolation LIV ll807 Aux. Bldg. Locked. -- - - Valve NSCWS Level A Open l Heat Exchanger Train A Outlet Isolation HV Il806 Aux. Bldg. Locked - - - lj Valve NSCWS Level 2 Open Heat Exchanger l Train A : Inlet Isolation HV Il815 Aux. Bldg. Locked - - - Valve NSCWS Level A Open H:st Exchanger Train B Outlet Isolation - HV.ll814 Aux. Bids. Locked. - - -- 1 Valve NSCWS Level 2 Open 1 l'._ Heat Exchanger .l Train B l Inlet isolation HV ll816 Aux. Bldg. Locked - - - Valve- -Spent . Level A Open Fuel Pool Heat Exchanger- .i Train A " l i l A-29 J 1 l l i
, e w - e ,-6 e w~ , - -- s e - - - - -- ---m* - - - - - - -
.. <.. 1
[ h.
- VOGTLE ELECTRIC' GENERATING PLANT!
. RISK. BASED -INSPECTION GUIDE ' '
Component Cooling. Water System j l TABLE A.4 2' MODIFIED SYSTEM WALKDOWN.(Cont'd) Desired Aceual . Pow. Sup. .. Required Actual Description ' ID No. Location Position Position ' Breaker # ' location' Position Position Outlet isolation HV l1817 Aux. Bldg. Locked- .
- .1 - -
Valve-4 pent ~ level A Open Puel Pool Heat Exchanger 4 Train A - Inlet isolation - HV.11920 Aux. Bids. lacked - - -- - Valve--Spent Level A . Open ' Puel Pool Heat i Exchanger Train B '- Outlet Isolation - HV.ll821 Aux.' Bldg. Locked; a - - - Valve--Spent level A ' Open ' Puel Pool Heat . Exchanger l Train B Inlet Isolation. HV.ll818 Aux. Bldg. Locked- - :- - Valve--RHR Level B Open - l Heat Exchanger Train A Aux. Bldg. -
~
Outlet Isolation . HV.11819 Locked- ~ - - - Valve-RHR Level C Open Heat Exchanger Train A Inlet Isolation HV.ll822 Aux. Bldg. Locked - - - l Valve-RHR level B Open Heat Exchanger Train B ' Outlet isolation HV.I1823 Aux. Bldg. Locked . - - -
-[
Valve--RHR Level C Open Heat Exchanger Train B Inlet isolation - V4-086 Aux. Bldg. Locked - - '- [ H Valve--RHR 14 vel D Open '! Pump Seal ' Cooler Train A Outlet Isolation V4-088 - Aux. Bldg. Locked - - - c Valve--RHR Level D Open Pump Seal , Cooler Train A i t A-30 t i 4
.. e 1
J
.' ' VOGTLE ELECTRIC GENERATING PLANT-
- RISK. BASED-INSPECTION GUIDE -
i . Component Cooling Water System TABLE A.4 2 MODIFIED' SYSTEM WALKDOWN (Cont'd) ,
# ^ I Desired Actual Pow. Sup. Required - Actual .
Description - ID No. Location Position Position . Breaker # Location Position Position inlet !aolation V4108 Au. Bldg. Locked - - - 1 Valve -RHR; L4 vel D Open - Pump Seal -i g , Cooler Train ~B Outlet. lsolation V4110 Am. Bldg. Locked. - - - Valve- RHR Level D Opn Pump Seal . Cooler Train B CCWL Pump HV Il800 Au. Bids. Locked - - - ) P4-001 Inlet level A Open , Isoladon Valve HV.Il803 Au, Bids. CCW Pump Locked - - - l P4-001 Outlet - Level A Open ' i ( laolation Valve CCW Pump HV.ll801 Am. Bldg. Locked - P4 003 Inlet level . A - Open lactation Yalve l CCW Pn.:::p HV.11804 Au. Bids. Locked - - -- P4 003 Outlet Level A Open Isolation Valve , CCW Pump - HV ll802 Au. Bids. Locked - - - P4-005 Inlet- Level A Open isolation Valve l CCW Pump HV 11805 Au. . Bldg. Locked - -- - P4005 Outlet IJvel A Open l 1 solation Valve HV ll808 A ax. Bldg.' Locked CCW Pump - .- - P4 002 Inlet level A Open l l Isolation Valve i l CCW Pump HV.ll8tl Au. Bids. Locked - ' - - l P4 002 Outlet Level A Open - laolation Valve l CCW Pump HV Il809 Au. Bldg. Locked - - - P4 004 Inlet Level A Open isolation Valve CCW Pump HV Il812 Au. Bldg. Locked - - - l P4-004 Outlet level A Open isolation Valve , l
.l l
A-31 i , 4 I , 1
\
9
- , - ~ <m,-, , e,~ - '
. .i e
- VOGTLE ELECTRIC GENERATING PLANT ',- '
-RISK BASED INSPECTION GUIDE:- - -
Component Cooling Water System l; TABLE A.4 2 MODIFIED SYSTEM ,WALKDOWN (Cont'd) i Desired Acts ' Pow. Sup. Required Actual i Description . ID No. Location - Position Positiou Breaker # . Location Position Position
. CCW Punp HV.ll810 Au. Bldg. Locked .
P4-006 Inlet Level A Open
~
Isolation Valve l' CCW Pump HV 11813 Am. Bldg. Locked- .-- . i P4 006 Outlet Level . A - Open Isolation Valve f l~ h L , i I 4 s r 0 4 . r t s
,1 A-32 a .
~.
~: ..
me. g B *** I E *** I commewcomme.
*****wt== :j. .a. ,, ,g .* ,,
C ,'l- {: , .f'
,. ,. i; T ,. > .m*; , _ ' ~ , .
. n ..w 11;
.,y
,e L- 7.
e
--c-*
.. - m.e=.m*us.
w .
,- g. , y ~, -
Id.-.
- s. N 1-
,, 7,.. .u..,, - , ,
'gg
?..
.to es. _ , :,
i ? -ae . .e e. a
,t
~
A, - w
. a. ,,,,,a. .
).
E
-- m.Ew.
4
- .. r.e.
~
I.- lE hi. ..: _( i
- m > .. - -.. -
.o m
y
~
s.
.. m "
5
. 9538 .s e
. #- .- - :. L. <
; 4*M,.e
.. . P. Lear -
= _7 -
8'
%, 6. .%s * .: .M gff--
-g A ; r.
- , l M 2
u'7 -
;w
.,y _.z.
,1 esa,. usa,,,. l
- .a s.es t..t 7 f r .- e n, q
,_...w,- . :g-< ,
4
~
p.. O ~* O e ;, f.? _ :.i ,-.9oL o.s J+ ,
~1
'm.- i ,K m.e, :...,,m-
%: ,-7mm;N - w > ,. 14
,7 s,t ,.
5
,g g; 3 a ,t *4
. ~ )=~ e
- r ~e '
- ,u m e#
,p. ; . .4 -
-s 7 w
- n. +
~ #~ ,. . yc..-y y . . ;
- c. = .
_- _ 4 w a
,i m., . y, e ;.e os.r;: 4.c.o . , m. - '.,
c . 4
.y.
- , ,m. -4... >
- m. e. . . . .
p,
... 2 ,. .
v.. ..' I " < * ' B . , . ,,, L _ ge E [7 s: alQ *p n 1 - n. nv., ? ~ a.. 1
..:t.
l -
+, 4
,. ... 4 ... .. - . .r
. . .. c.- -
-e - - - .n '-.
~ .,
m b *** "-~ ; , i i .
..n ., - .A .~
b -
~,. m,o ^
.? .h :
..~r ,.., -
._,,. ..r._;
5
- r. ., .u..
's ee . +
. ,.\,
'~ + . 1_
s w' ' ~
. . M* M '.f. = %,
m 'C.'. 7 Stween
.pg,a. x,.
~ .. , ,. ..-
. u.t1 +~
..; r..( ..
, r *r. _ _-
- m. ; x,: s-
- c. v.; w , a _
~e ~
c .-.
- w. -
m , .- f, .. n ;: , _ .- r . . Fijlure A.4-1 Vogtle 1 Component Cooling Water System Showing Component Locations
- Reference SAIC 89/1540 DRAFT. Figure 3.8-2) 0 F ..
, w-y ,, n- --_w, mw ' ^ ' ' - -~ * " " ' " " * ' '" ' '
TABLE A.4 2 (Cont'd)
- REFERENCE DOCUMENTS TITLE I.D. NO. REV DATE
- 1. Vogtle Electric Generating Station Procedure 11715 1 3 2/9Ai9
" Component Cooling Water System -
Alignment"
- 2. Vogtle Electric Generating Station Procedure 13715 1 3 5/16/88
" Component Cooling Water System"
- 3. Vogtle Units I and 2 Final Safety Section 9.2.2 Analysis Report 44, W
-d i
e i f 5 e A-34
~
-q _
-1 e , . s. '
l
' .' VOGTLE ELECTRIC GENERATING PLANT
. RISK BASED INSPECTION GUIDE -
l ! I i Reactor Trip System i 1 I Table A.5-1. Irnportance Basis and Failure Mode Identification ) ~ 1 i' CONDITIONS THAT CAN LEAD TO FAILURE I l Mission Success Criteria j 4 The Reactor Trip System automatically keeps the reactor operating within a safe H region by shutting down the reactor whenever 'he limits of the region are exceeded (or i j reached). When.-ver a direct process or calculated variable reaches a setpoint the reactor i j will be shutdown in order to protect against either gross damage to fuel cladding or loss of I t system integrity which could lead to release of radioactive fission products into the Containment. l l The following' systems typically make up the Reactor Protection System. ) . a. Process Instrumentation and Control System
- b. Nuclear Instrumentation System I
- c. Solid State Logic' Protection System
- d. Reactor Trip Switchgear -
1 c. Manual Actuation Circuit The reactor trip system contains sensors, which when connected with analog circuitry l consisting of two to four redundant channels, monitor various plant parameters. The Reactor Trip System also contains digital circui.try, 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. The following parameters initiate reactor trip sipals. *
- 1. High neutron flux: Power range
- 2. High nection flux: Intennediate range ;
- 3. _High neutron flux: Source range
- 4. Power range high positive neut7n flux rate ,
i
- 5. Power range high negative neutron flux rate ;
- 6. Overtemperature Tn , .
- 7. Overpewer T,y
- 8. Pressurizer low pressure i
- 9. Pressurizer high pressure
- 10. Pressurher high water level
- 11. Low retetor coolant flow
- 12. Reactor coolant pump bus undervoltage i L
! -i A 35 4
?
-, 4 - . n s ,,v -. < , -
' . j 1 13. Reactor coolant pump bus underfrequency , ,
- 14. Low feedwater flow
- 15. Low low steam generator water level
- 16. Turbine generator :
- n. Low auto stop oil pressure
- b. Turbine stop valve close 3
- 17. Safety injection signal 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,LB.YB 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. Duririg plant power operation, a- , DC undervoltge 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 j 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. Aeeldest importance laspection Dominant Failure Medes Sequence Category Activities i ) 1. Instrument failure due to calibrationhnalntenance error, or 11 H- S.M.C 4 random failure which inhibits laitiation of reactor trip + signal -
- 2. Reactor trip breaker or trip bypass breaker fails to open 11 M S.M C ,
l.osic Train A Trip Breakers l 52/RTA I 52/BYB > tagic Train B Trip Breakers , 52/RTB - 52/BYA
- 3. Operator failure to manually scram reactor following 11 L. O f A'!WS i
l i I t t l
. I l
o , A 36 j i j , ,
. , , . _ . _ , , , . , . . - .. . - - - - . , . ~ - . . . , - ,
. . . .-_ .__. . _ _ - _ . . _ - -- -. . - - . - ~ . .._ - _ _
I
,' VOGTLE GENERATING STATION
, RISK. BASED INSPECTION GUIDE - ;
Reactor Trip System l I TABLE A.S.2 MODIFIED SYSTEM WALKDOWN i The Reactor Trip System is a normally energized system and operability is assured by ;
- ~
extensive sutveillrnce testing. Observations during this testing will provide the inspector j i with direct Iriput 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: COMPI NENT REQUIRED STATUS' ACTUAL STATUS
\
- 1. Reactor Trip Breakers RTA Closed l RTB Closed
- 2. Reactor Trip Bypass Breakers BYA Open BYB Open
- 3. Solid-State Protection System Power Supply Breaker Closed .
Breaker INY4N 07 (Panel INY4N) r
- 4. Main Board Demultiplexer Power Supply Besaker Closed Breaker INY3N.04 (Panel INY3N)
- 5. Annunciator Panel - Reactor Trip System No windows illumitiated
- 6. R111 System Status Monitorins Panel No bypass lishts illuminated 1
l < 1 I l A 37 , _______________.________L _ . . _ . . . _ , . .. .
TABLE A.5 2 (Cont'd) '. ' REFERENCE DOCUMENTS TITLE 1.D. NO. 'REV DATE
- 1. Vogtle Electric Generating Plant Procedure 13503-1 2 8/26/88
" Reactor Control Solid. State Protection-System"
- 2. Vogtle Units 1 and 2 Final Safety Section 7.2 Analysis Report h
M A 38 1
VOGTLE ELECTRIC GENERATING PLANT I
, RISK. BASED INSPECTION GUIDE Emergency Core Cooling System /High Pressure Injection l 4
Table A.6-1. Importance Basis and Failure Mode Identification l i l CONDITIONS THAT CAN LEAD TO FAILURE I Mission Success Criteria ;
- The Emergency Core Cooling System (ECCS) includes high and intennediate head ,
- injection pumps to deliver borated water from the RWST to the reactor coolant system i
! (RCS) following small and intermediate loss of coolant accidents where the RCS does not ;
- depressurize sufficiently to allow injection via the residual heat removal pumps. ;
i y The high head portion of the system is provided by two centrifugal charging pumps (CCPs) that can deliver borated water to the RCS at pressures up to approximately 2510 l j pais. The maximum flow rate from each pump is 550 gpm. The CCPs are started
- automatically bre,ed on a safety injection (SI) signal, and are automatically aligned to take suction from the RWST during the injection phase. For long term operation in the i recirculation phase, suction is provided by the residual heat removal pump discharge flow.
l Mission success for the operation of the CCPs is provided by the operation of one charging l pump. . Two safety injection (SI) pumps are provided for the intermediate head requirements-of the ECCS. Each SI pump can deliver a maximum flow of 650 gpm and has a shutoff I head of approximately 1500 psia. These pumps deliver water from the RWST during the injection phase and from the containment sump (via the RHR pumps) during the recircula-tion phase. The switchover to the recirculation phase is performed manually based on low water level in the RWST. Mission succes. for de safety injection pumps is provided by the operation of one pump during the h.section and recirculation' phases. l Accident imp 6ctance lanpection 1 Dominant Failure Modes ty = Catesory Activit6es
]
- 1. Failure to switch from RWST to the containment sump via 1- H 0 the residual heat system. This switchover is manually initiated bened on low RWST water level.
- 2. Pallure of discherse valves to open, includins common i H S.M,T,C -
cause failures (includes check valves) Motor Operated Valves: . St Hot Leg injection: 8802 A and B 1 SI Cold tag injection: 8835 CCP Injection: 84s3 A and B: s43s I A-39
-*,- - w <w .W-e ,
Aec6 dest leportance leapeettie , Deeninast Fallere Medes Sequence Category Activities Check Valves:
$1 Pump Cold Leg Loop 1: ' 143 81 Pump Cold Leg Loop 2: 144 81 Pwnp Cold Les Loop 3: 145 St Pump Cold Lag Loop 4: 146 31 Pump Hot Lag Loop It 120
$1 Pump Hot lag Loop 2: 123 81 Pump Hot Leg Loop 3: 122
' St Pump Hot Lag Loop 4: 121 St Pump Discharge: 098 (Pump A) 51 Pump Dlacharge: 099 (Pump B)
CCP trgjection: 013
, 3. Failure of pump suction valves to open, including common 1 M S,M,T C asuse fallwe (includes check valves)
Motor operated valves:
$1 Pump Suction from RWST: 8806 8923 A and D' CCP Suction from RWST: LV.Il2 D and E 8807 A and B Check Valves:
St Pwnp Suction: 090
- 4. Pallure of pump mlniflow valve to open falls operating 1 M S.M T.C Pump Motor operated valves:
St Pun Miniflow: 8813,8814 and 8920 CCP Miniflow: 8110 ) 8111 A and B CCP Miniflow (alternate): 8508 A and B 1
$. Electrical fallwes (power cable / breaker) disable $1 pump 1 M SM room ventiladon
- 6. Fallwe of Nuclear Service Cooling Water System valve to 1 M S M,T open or remain open disabling pump cooling ,
Manual valves, locked open
$1 Pump Cooling Inlet isolation Valve: V4100 Oudet laoladon Valve: V4101 Inlet teolation Valve: V4131 Oudet teoladon Valve: V4132
- 7. Local fault of pumps / pumps fail to start or run 1 M 5,M,T.C Safety injection Pumps 15003 P6 004 Centrifugal Charging Pumps P6 001 p6 002 e
A-40 i
s . Accident importance Inspection , Dominnet Failure Medes Sequence Category Activities
- 8. Failure of valve in the common portion of the suction line i M S.M.T.C from the RWST.
Locked open manual valve 207.
- 9. Baron irdection tank (BIT) section or discharge valves fall 1.11 L 3.M.T.C ]
BIT Suction: Normally Open MOVs: 8803A 8803B - Brr Discharge: Normally Closed MOVs: 8801A , i 88015 l 10 Local pump failures ! L S.M
- failure of control cable to MCC 1
- failure of pump breaker to close St Pump Motor Breakers P6 003 1 AA0216 ;
15004 IBA0317 l 1 CCP Motor Breakers , ir P6 001 l AA0213 P6.(C2 IBA0313 l
- 11. Pump in maintenance i L M Two CCP or $1 pumps should not be simtitaneously in maintenance i 12. Operator failure to realign pumps on low low RWST level 1 L O )
l Centrifugal charging pumps and Si are manually aligned in - series with the RHR pumps during recirculation, This alignment is a manual action. ( l' l l I l l - L . A-41
. , , . ,,-e - + - - . _ _ . . _ . , -- y ,
VOGTLE ELECTRIC-GENERATING PLANT - RISK. BASED INSPECTION GUIDE . Emergency Core Cooling System /High Pressure Injection System TABLE A 6 2 MODIFIED SYSTEM WALKDOWN Desired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Breaker e Location Position Position RWST Outlet V4207 Vicinity of Locked - - - - Manual leolation RWST= Open Vaive St Pwnp A V4135 Au. Bldg. Locked - -- -- - Manual Section Room BIS Open teolation Valve $1 Pump A V6100 Au. Bldg. Locked - - - - Manual Dis- Room Bl$ - Open charge Isolation Valve $1 Pump B V4136 Aux. Bids. Locked - - - - Manual Suction Room B19 Open Isolation Valve
$1 Pump B V6101 An. Bldg. Locked - - - -
Manual Dio. Room B19 Open charge Isolation Valve Hot 143 injec. HV 8802A Au. Bldg. Clo*ed lABD17 An. Bldg. Closed tion Header Iso. Level A Level C lation Valve: Loops I and 4 Hot Leg HV 88028 Fuel Bldg. Closed IBBD 17 Aux. Bldg. Closed Indoction Level A Level B Hender Isolation Valve: 1. mops 2 and 3 SIS Cold Lag HV 8835 Au. Bldg. Open lABB.21 Aux. Bldg. Closed Header Isolation Level A Level i Valve i St Pump Inlet HV 8923A Au. Bldg. Open I ABD.18 Aux. Bldg. Closed Isolation Valve: L4 vel B Level C . Train A St Pump Inlet HV 8923B Au. Bldg. Open IBBD 18 Au. Bldg. Closed ' Isolation Valve: Level B Level B Train B I Charging Pump HV.843g Aux. Bldg. Open IBBD-37. Au. Bldg. Closed Discharge isola. Level C Level B tica Valve v A42 . 1 1 I
n
..----.l L
e , VOGTLE ELECTRIC GENERATING PLANT l , RISK BASED INSPECTION GUIDE 1 Emergency Core Cooling System /High Pressure Injection System i h i ' TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd) l I" Desired Actual Pow. Sup. Required Actual Description ID No. lacation Poeltion Position Breaker # 14 cation Position Nition i c Chargmg Pump NY.848$8 Au. Bldg. Open 188D.40 Au. Bldg. Closed . 8 Discharge Level C Level 3 [ luolation Valve l Charging Pump HV 8485A Am. Bldg. Open I ABD-40 Au. Bids. Closed a A Discharge Level C Level C ' j isolation Valve
- , .1 Pump Suc. HV.8806 Au. Bldg. Open IBBD12 Am. Bldg. Closed i an isolation Level B L4 vel B j Valve
{ RWST to LV.Oll2D : Am. Bldg. Closed IA9D-08 Au. Bldg. Closed Charging Pump level C Level C laolation Valve RWST to LV.0ll2E Am. Bldg. Closed IBB D 08 Am. Bldg. Closed I l Charging Pump L4 vel C Level C 1 l Isolation Valve
- 31 Pump suc. HV 8807A An. Bldg. Closed IABD.21 An. Bldg. Closed tion to Charging L4 vel B L4 vel C Pumps isolation l
Valve
- l. l
[ 31 Pump Suc. HV.8807B An. Bldg. Closed IBBD 21 Au. Bldg. Closed l' , Lion to Charging Level B level C i Pumps isolation l Valve l' St Pump B HV 8920 Au. Bldg. Open IABD-12 Am. Bldg. Closed Mininow Isola. Level B Level C l tion Valve St Mininow HV 8813 Am. Bldg. Open IBBD.33 Au. Bldg. Closed
- leolation Valve Level B Level C
- i. ~
l- St Miniflow HV 8814 Au. Bldg. Open IABD.21 Au. Bids. Closed L p 1 solation Valve level B 14 vel C Chargmg Pump HV.8110 Au. Bldg. Open IABDIl Am. Bids. Closed Mininow Isola. L4 vel C L4 vel C 1
- tion Valve J
; Charging Pump HV 8111A An. Bldg. Open 188D.11 Am. Bldg. Closed
[1 l Miniflow isola. Level C LAvel C tion Valve . l Charging Pump HV.811IB Am. Bldg. Open 1BBB.1i Am. Bldg. Closed 1 Miniflow isole. L4 vel C level I ! tion Valve . l 4 * !' A-43 l 1 5
- w: . _
%O O 4
VOGTLE ELECTRIC GENERATING PLANT ', ' RISK BASED INSPECTION GUIDE . Emergency (. >re Cooling System /High Pressure Injection System TABLE A.6 2 MODIFIED SYSTEM WALKDOWN (Cont'd). Desired Actual Pow. Sup. Required Actual Description ID No. L4 cation Position Position Breaker # l.acation Position Position Alt. Charging HV 8508A - l ABB 16 Am. Bldg. Closed Pump Miniflow Level i Valve Alt. Charging . HV.tS08B - 1898 16 Am. Bids. Closed Pump Mininow Level i Valve Bl? Discharge H%8801A Am. Bldg. Closed IAB D 36; An. Bldg. Closed leolation Valve 14 vel B 1.evel C BIT Discharge HV 8401B Au, Bldg. Closed 188D 36 Au. Bldg. Closed leolation Valw Level B Level C Brr inlet Isola. H%8803A Am. Bids. Open - - - tien Valve 14 vel B BIT inlet laola. H W88033 Am. Bldg. Open - - - tion Valve Level B i SI Pump P6 003 Au. Bldg. - - l AA0216 Connel Bldg. Racked in Level B 14 vel A
$1 Purnp P6 004 Au. Bids. - - IBA0317 Contml Bldg. Racked in level B level A -
i A-44
e ,e ? i t e i ! I i
~
.lli".l!i e
.. l! . .
ili :lli , m:j
- i - m:31 g jiw - Q , c:. 4 ,' J , ._
di .,; .
,) ,g .
i [,
- m
- lim:l 'i i
L,4 :+4 ;,4 ' a -
+4 i i i l kh,I
.~ <
r: - 1 l 4 1 ., k I . fhiI ff,I \
. e.
I ! ! g l i I fI fI ! f ,I ' I. ,I I !
- i. iI I
-lil t,ir li'I g, i li j i 1:['. p- 1 I' q-.i e
,. li[' g rji :
,g :
4
- i 41 '
c d! : c +cdl r e g
. l ejm
.- - 4 , . ,
.a n _,
]< . .
, o
'. ;;w s .,l h ""
. ,: . }, , .e .,
h s h.Q,A , .j', c<> , . e> .. m, . ,+ , , 4 , , . :,.....p.,. .gf. i ' . v .. . . q . r,m. , r u wyn- y'<,; , l , .-
- (' 1
<f. ]7( , .h
. .hs ' \ ' ' ,$
?' P #*Ii='
I *
~'
+' "2
. Y ( {j,3 ,#[
.. .y; ;j gg d; /*" [ ' j-Y
; {j,g ' a () '
': . : 3 ,_
^
a 3 1j i
- y. . !1 y:. g .
n.
,. ,c 4 -
! 1
>p -
F '4' ; l,i - li 43 p. 4:3.'f (( ii J. g .g~ : i 'p e- - 2 cu, T m ,lm,,31 i '. I I I _f, . N. N . I-g5 a., ' ' fi,I x . tXI ! 1 " i
,g, -
... . e _. ; cq
,i n,. m:1.
i -
= ,i .I .
~ !
t
, ' .. ?[..,
l . Q h A-45 F
.. , , , , , . t o- y --~ w--. . . ~ - - _ _ . , , ., , , - . , . , . - . _ < . -,--./ , - :. , .-
l l i TABLE A.6 2 (Cont'd) . f REFERENCE DOCUMENTF i \
' TITLE I 1.D. NO. REV DATE i I
13105 1 9 11/30/88 :
- 1. Vogtle Electric Generating Plant Procedure
" Safety injection System"
- 2. Vogtle Electric Generating Plant Procedure 11105 1 8 !!/23/88 l
" Safety Injection System Alignment" j
- 3. Vogtle Units 1 and 2 Final Safety Section 7.3.2 i t
Analysis Report i I f r P j i i t
-[
i f ('
.i
~
f Y s i l f A-46 j l-i
?
VOGTLE ELECTRIC GENERATING PLANT RISK. BASED INSPECTION GUIDE . Primary Pressure Relief System Table A.7-1. Importance Basis and Failure Mode Identification CONDITIONS THAT CAN LEAD TO FAILURE Mission Success Criteria The Primary Pressure Relief System (PPRS) provides protection from overpressurization of the prirnary system to ensure that primary integrity is maintained. he 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 SRVs have combined relief capacity of 1.26 x 10' lbm/hr and are set to open at 2485 psig. The PORVs provide RCS pressure relief at a set point of 2335 psig and have a combined capacity of 420,000 lbm/hr.'In addition to automatic opening on high pressure, the PORVs can be manually opened by. the reactor operator. The PORVs discharge to the pressurizer relief tank. Each PORY is provided with a motor operated block valve. The block valves are normally open unless a PORV is leaking. The PPRS i*,
dependent on the AC power buses for motive and control power to the PORY block valvt s and vital AC power for control power to the PORVs. . t ne success criteria for the PPRS vary depending on the application. The success criterion for the PPRS following a transient event demanding PORY opening is that the PORVs sw:cessfully reclose. The success criterion for the PPRS following a transient and failure of t'ie AFWS is that both PORVs successfully open on demand. Other success criteria invcive the actuation of both PORVs and up to three SRVs depending on the transient anil the extent of other system failures. Accideot Importaxe leapecties Dealeaet Failure Medes Sequence- Category Activities
- 1. PORV fails to open for bleed & feed mode 6.7 H S.M.T.C PV 455A and PV 456A .
- 2. Failure of PORV/SRV to ressat causing small 1OCA I H M Safety Relief Valves:
PSV 8018A PSV 80188 PSV 8018C Power Operated Relief Valves: PV 455A PV 456A A-47 i l
. -t 4 . .. $
t ei. t Accident c importance lespect6ise i Dominant Fellert Modes Segeence Categwy Activitnes
- i t
- 3. PORY bloct valve closed 7 M O.M !
G- MV $000A and $0008 ! 5
- 4. Operster er w in blood a food activides causes lack of - 6 M O l
RCS cooling ,
t f 6 h 1 I 4 t., . t E F l r s f r r i i 0 1 L 1 ! i u \ 1
- 1 2 \ 0 A-48 ;
1
.I<
, 45 . . - .- - _ . , - . . . .
, -. VOGTLE ELECTRIC GENERATING PLANT RISK BASED INSPECTION GUIDE Primary Pressure Relief System TABLE A.7 2 MODIFIED SYSTEM WALKDOWN Dsaired Actual Pow. Sup. Required Actual Description ID No. Location Position Position Besaker # Location Position Position Presentiser PV 4$$A Containment comed lABE Control Bldg. Closed PORVs Level B PV 456A Containment Closed ISBE Control Sidg. Closed level B -
PORY Block HV 8000A Containment Open lABE Control Bldg. Closed Valves lavel B HV 80008 Cortainment Open IBBE Control Bldg. Closed. level A Safety Relief PSV 8010A Containrrant Closed - - - - Valves PSV 80108 Containment Closed - - -- - PSV 8010C Containment Closed - - - - 4 A49 4
l l 1 l o s., --
~~
s P ~
- ~ ~ -
2'g _ . , . , = _ S._ < - n= ; _ .
. hi
~
4. il P ' -,9,_.9.-::::::, '
.,_ 64. .-
^
- A_ -
v f% 6
. I -
g ,m-
+ ,
. 4 . ;.[ / : - .x.
y ,e emee _-,- -
?
- -~s 2 _ ..
- y. ..- . .- -, r. , ?. ~
. : ;; ;;: g .1, x, 1,
... ., . ; . n ,n ga. ,- _ n".,-
l
'* ?
, , - . , t .
e i
.J am, re au ,e 1 -d---- - * .y y'
' . =M .'
.. A
=:n, '
> a. - .
t s_
?--
- ., . - sS _c= x .
c , l-_c % -- Me
, - u -
* ~
- ." c. .
o -
' !:U
.; (. +>
,xn ;5g_Jz?
. + -
,y *
; ; r-
. d;: fz
,.y -
, g; n p 4= pg, : ..
^
~ ..
.. ...a - - _-
. 1 .
. ~, :. , . , .
, . ,, . 4.. - .
p
._w..... ..
~ , .. ,
-;,x^- .q._--y
- _ 4. -
.. .-.;. . -4 3o y
t
; =.g --.--
- u. 4
'r i
E **- E Figure A.7-1 Vogtie 1 Reactor Coolant System Showing Component locations (Reference SAIC 89/1540 DRAFT, Figure 3.1-3) 6 e l
$r n .% ,-
.nww . .r.-- . ,-.: - .* w , .. , _ _ _ _ . _ . _________________m._. _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ , _ _ . _ , , _ . _ . _ _ _ . - . ___m._ _ , _ _ _ _ _ _ . , _ _ , _ , , _ _ . _ _ _ _ _ . . ____._m_____ _ _ _ _ _ _ _ . _ . _ _ , .. _ . _
, TABLE A.7 2 (Cc t'd)
^
REFERENCE DOCUMENTS TITLE I.D. NO. REV DATE.
- 1. Vogtle Electric Generating Plant Procedure 11001 1 8 11/21/88
" Reactor Coolant System Alignment" ,
- 2. Yogtle Units 1 and 2 Final Safety Section 5.4 Analysis Report 0
9 9 1 42WS A 51 4
VOGTLE ELECTRIC GENERATING PLANT . RISK. BASED INSPECTION GUIDE , Auxiliary Feedwater System Table A.81. Importance Basis and Failure Mode Identification CONDITIONS THAT CAN LEAD TO FAILURE' 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 provides the safety function of early core heat removal following a transient or small LOCA. The AFW system is a three train system-Trains A, B, and C-including two motor d.-iven pumps (MDPs)-Trains A and B-and one turbine driven pump (TDP)-Train C. Each MDP discharges to two of the SGs and is sized to remove 100% of the plant decay heat. The TDP is approximately twice the capacity of the MD pumps and discharges to all of the SGs. Each pump takes suction through a common header from either of two Condensate Storage Tanks (CSTs).' Each of the two CSTs has a capacity of approximately 480,000 gallons. The Technical Specifications require one CST to be operational with at least 340,000 gallons of water. Each flow path from an AFW pump discharge to an SG has two check valves, a normally open motor operated valve, and locked open manual valves. The two MDPs start automatically on receipt of an AFW actuation signal. This signal is generated in response to any of the following conditions: low-low SG water level, safety injection signal or trip of both main feedwater pumps. A station blackout signal also starts the MDPs and causes the TDP throttle / trip valve to open automatically starting the TDP. The AFW system depends on AC power for motive power to MDP motors and the motor-operated valves, and Class IE DC power for control power to MDPs and TD pump. As noted above, either of the MDPs has the espability of supplying a minimum of 100 ' percent of the feedwater requirements for safe cooldown of the RCS In addition, the TDP has the capability to deliver 200 percent of the feedwater flow required for decay removal. Thus, mission success for this three train configuration can be accomplished assuming a postulated failure in the discharge piping of one train, concurrent with a single active component failure in another train. A-52 -
4 i j
)
i: . Accident importnace leapection Dessleast Failure Modes
- ,. Segeance Category Activities t
- 1. Pallure to manually start locked out standby pump 7,10,6 H O :
Procedure 136101. Auxiliary Feedwater System, describes i the required alignment for vanous modes of operation i ! 2. Lacal fault of valve in turbine. driven pump discharge to 6,3,7 H S.M,T > j seeam generator I i TD Pump Discharge Valves-015 Locked open, manual valve 014 Check valve i To Steam Generator i
; 019 Locked open, manual valve 020 Check valve HV $122 Motor operated, normally open
- To Siesm Generator 2
- l '022 Locked open, manual valve
! 023 Check valve HV $125 Motor operated, normally open l To Steam Generator 3
- . 025 Locked open, manual valve
, 026 Check valve HV $127 Motor operated, normally open
; To Steam Generator 4 l 016 Locked open, manual valve
, 017 Check valve j HV 5120 Motor operated, norma!!y open [ 3. Pallure to manually start pump given auto start failure 10.6,7 H' O Manual startup of MDPs is described in Procedure No. ' 136101 (under normal operating conditions) l l- 4. Turbine driven pump fails to start or run, or out of service 10,8,7,6 H S.M,T,C . l due to maintenance or testing l Turbine driven pump: P4-001 (Train C) ! 5. . Motor driven pump falls to start or run, or out of service 6,10,7 H S.M T.C due to maintenance or testing P4 002 (Train B) P4 003 (Trals A)
- 6. Local fault of valve in motor driven pump discharge to 6,7 H S.M.T steam generator Train A: MD Pump Discharge Valves
- Check valves: 001,043 and 046 Lacked open, manual valve: 035,042 and 045 Motor operated valves, normally open: HV 5137 and HV 5139 Train 8: MD Pump Discharge Valves Check valve: 002,037 and 040 Locked open, manual valve: 060,036 and 039 "
Motor operated valves, normally open: HV $132 and HV 5134 v A 53
, , . ,, _ , - . . _.,....,i , , , , - . - . - . .-.y... ., r,. -. - , . . - . . ..
i Accident importance leapeetion l Denisset Failure Medes Sequence. Category Activities . - 4 I ,
- 7. Steam supply valves fait closed (or other valve faults in 10,7,6 H S.M,T,C -l steam admissior, line) for turbine driven pump q Motor operated valves: - 3009, 3019 and $106 ' ]
Locked open manual valves: 005 and 007 l Check valves: 004,006 and 008 ; Trip and trottle valve: 5129 l Speed governing valve: 5133 le 8. 14 cal fault of suction valve from the condenarse aiorage 10,7 M O.S.M tank (CST) - Train A: CST 001: Locked open manual valves: 092 and 095 Check valve: 033 CST 002: 1.acked open manual valve: - 099 - l Motor operated valve: 5119 i 8 Check valve: 052 . l Train B: CST 001: tecked open manual valves: 091 and 094 j Check valve: 058 i CST 002: 1.acked open manual' valve: 098 ! l Motor. operated valve: 5118 Check valve: 061 q i Train C: ) CST 001: Locked open manual valves: 090 and 093 ; Check valve: 013 CST 002: = 1.acked open manual valve: 097 ,; Motor operated valve: 5113 l Check valve: 051 - ,
- 9. ' 'V now control valve in maintenance falls delivery 7,6.10 M M I (D or MD pumps :
L Motor Operated Valves: Train A: 5137 and 5139 . Train B 5132 and 5134 ! Train C: 5120, 5127, 5125 and 5127 l i
- 10. Undetected now diversion 7 M O Motor. operated valves $154 and $155 regulate the i mininow lines of Tiain A and B and require calibration of i the now element. >
l-11, Undetected FW leakage back through pump discharge 7 'M , O ! valves causes steam binding [ Check valves 001,002 and 014 prevent backnow to the i pump discharge piping. j
- 12. Lacal fault of motor driven pump power breaker 10,7 M S.M i Circuit Breaker Switchgear [
IBA03 21 IBA03 (Pump 2) I 1 AA02 7 1AA02 (Pump 3) ; i l I v i i A-54 [ . t l a-
.t:
,' Accident Importanee lampacties Deelmani Pellere Medes Segmenee Category Activities -
- 13. Failwe to increase flow to steam generator given plugging 6 L 0 of valve in other swam generator
- 14. Lacal fault of APW actuation signal logic falls to actuate 10 L 8,C MD pump and/or TD pump steam valves Actuation signals for motor <triven APW pump:
- 1) Law. low level in any steam generator
- 2) Safety impction signal
- 3) Trip of both main feed pumps
- 4) Complete loss of electric power Actuation signals for turbine driven AFW pump:
- 1) Low. low level la two stern genere ors
- 2) Complete loss of electric power
- 15. Falls to restore 1D pump discharge valve after test 6 L 0 Locked open manual valve: Ol$
e e A 55 l
t . . i VOGTLE ELECTRIC GENERATING PLANT ' l RISK. BASED INSPECTION GUIDE , ) Auxiliary Feedwater System f
)
TABLE A.S.2 MODIFIED SYSTEM WALKDOWN i Desired Actual Pow. Sup. Required Actual ) Description ID No. Location Posidon Poettion Breaker e Location Position Position i
]
TD Pump i HV 5090 CST.1 Locked - - - Suction froen Valve Room Open
)
1 CST 1 laolation i Valve
]
MD Pump 2 HY.5091 CST.I Locked - - - j Section from Valve Room Open ; CST 1 Isolation ' i 2 Valve l MD Pump 3 HV.5092 ' CST 1 Locked - - - ' Suction from Valve Room Open CST 1 Isolation Valve TD Pump i HV.5097 CST 2 Locked - - - Section from Valve Room Open CST 21 solation i Valve MD Pump 2 HV 5098 CST.2 Locked - - - . Suction from Valve Room Open : CST 2 leolation . , Valve MD Pump 3 HV.5099 CST.2 tecked - - - l Suction from Valve Room Open < a CST 2 leoladon Valve l MD Pump 3 HV.5095 APW Pump Lacked - - - l Section from Room Train Open CST.I Isolation A Valve MD Pumps V4 054 AFW Pump Lacked - - - Train A Recirc. '
. Room Train Open isolation Valve A .
m CST 2 MD Pumps V4 035 APW Pump Locked - - - Train A Dis- Room Train Open charge Isolation A Valve 1' v L A 56 9 g
. ._.. . . . _ _ _ _ . . _ . _ . . . . ____ , _ . . _. ~ _ _ _ . . ___.
l, 1
, VOGTLE ELECTRIC GENERATING PLANT l
, RISK. BASED INSPECTION GUIDE l i
Auxiliary Feedwater System ; TABLE A.S.2 MODIFIED SYSTEM WALKDOWN (Cont'd) t l-Desired Actual Pow. Sep. Required Actual , Description ID No. Location Position Position Breaker # 14 cation Position Position , MD Pumps V4 0$6 APW Pump 1.acksi - - - Cross Connect Room Train Closed i Train A to B t Train B leola. L
; tion Valve !
i MD Pump 2 HV.5094 AFW Pump Locked - - - Suction from Room Train Open ; ! CST 1 Isolation B Valve i MD Pump Train V40$3 AFW Pump tecked - - - B Rectre. Isola. Roose Train Open
- l tion Valve to B CST.2 MD Pump Train V4 060 AFW Pump Locked - - - i B Discharge Room Train Open !
k isolation Valve B i Cross Connect V40$$ AFW Pump Locked - - - , Train A to Room Train Closed , Train B B
- TD Pump V4 028 AFW Pump Locked - - -
- Recirc. Isolation Room Train Open Valve to CST.I C i TD Pump Dis. V4-Ol$ AFW Pump Locked - - -
, charge isolation Room Train Open i l Valve C Main Steam TD V- '33 AFW Pump - Closed - - - Pump Aut. IC Turbine j Steam Stop Drive Check Isalation Valve TD Pump Aus. VA264 APW Pump Clesed - - - I Steam Supply IC Turbine Isolation Valve Drive > Main Steam In- V4-013 APW Pump Open - - - let isolation IC Turbine i Valve Drive Main Steam V4 083 AFW Pump Open -- --- - Outlet isolation IC Turbine j Valve Drive
. I A 57 i
4
, . , . ,. . . _ . - ,,,...-~_...4
I
'o o
, VOGTLE ELECTRIC GENERATING PLANT ' '.
- RISK. BASED INSPECTION GUIDE- -
i Auxiliary Feedwater System l TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd) , 1 Desimd Actual Pow. Sup. Required Actual l
. Description ID 140. Location Position Position Breaker e Location Position Position Main Steam to V4-014 APW Pump Open - -
J TD Pump IC Turbine . Drive i Main Steam V400$ Aua. Bldg. Locked - - -
$01 to TD Leveli Open .
Pump South Maln j Steam Valve Room MD Pump Train V4-045 A u. Bids. Locked - c- - 5 A to SG 1 Level A Open South Main l Feedwater Valve Room SO1&4 MD Pumpe V4 046 A u. Bids. Locked - - - Train A Stop Level A Open . Check Valve In- South Main - jaction to SG 1 Feedwater Valve Room SO1&4 MD Pumps V4 042 Au. Bldg.' Locked - - - Trala A Injec- Level A Open tion to SG-4 South Main isolation Valve Feedwater Valve Room 301&4 MD Pumpe V4-043 A u. Bldg. Locked - - - i Train A !$. Level A Open - tion to 50-4 South Main Stop Check Foodwater Valve Valve Room SOI&4 r TD Pump Injec. V4-019 A u. Bldg. Locked '- - - tion to 301 Level A Open leolation Valve South Main Feedwater Valve Room l SOI&4 A 58 l .
- 1. .
I ' VOGTLE ELECTRIC GENERATING PLANT 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 Posttion Breaker # Location Position Position TD Pump Injec- V4 020 Auz. Bldg. Lacked -- - - tion to 301 14 vela Open Stop Check South Main Valve Feedwater Valve Room 501&4 TD Pump Irtjec- V4-016 Aua. Blog. Locked - - - tion to $0 4 IAvelA Open 1 solation Valve South Main Feedwater Valve Room 501&4 TD Pump Injec. V4-017 Aua. Bids. Locked - - - tion to $0 4 L4 vel A Open Stop Check South Main Valve Feedwater Valve Room . 501&4 APW Stop V4in3 Ana. Bids. Locked - - - Check Valve to level A Open 50 1 South Mein Foodwater Valve Room sol &4
$01 Inlet Test V4081 Auz. Bids. Locked .-- - -
Line Isolation Level A Closed Valve South Main Feedwater , Valve Room I SO1&4 APW Stop v4 ll6 Ana. Bldg. Locked - - - Check Valve to Level A Open . SG4 South Main ' Feedwater Valve Room SOI&4 30-4 Inlet Test V4-084 Ana. Bids. Locked - - - Line Isolation Level A Closed Vdn Sa& Mdn Feedwater Valve Room 501&4 A 59 i
l VOGTLE ELECTRIC GENERATING PLANT '.
- RISK. BASED INSPECTION GUIDE ,
Auxiliary Feedwater System TABLE A.S.2 MODIFIED SYSTEM WALKDOWN (Cont'd) Desired Actual Pow. Sup. Required Actual Description ID No. Lasetion Poaltion Position Breaker 8 1.ocation Position Position 60 2 Main V4 007 Control, Lacked - - - siens Supply to Bids. Level Open m Pump 1 North Main Steam Valve Room MD Pump Train V4 036 Connel Locked - - - B to 50 2 loo- Bldg. Level Open lation Valve A Nonh Main Peedwater Valve Room 502&3 MD Pump Train V4-037 'Conool Lacked - - - B for 50 2 Vids. Level Open Stop Check A Nonh Valve Main Peedwater Valve Room 802&3 MD Pump Train '9 Control Locked - - - B to 50 3 Iso- Bids. IAvel Open ! lation Valve A North Main Peedwater Valve Room l SO2&3 1 MD Pump Train V4-040 Control - Lacked - - - i B to 50 3 Stop Bldg. L4 vel Open 1 Check Valve A North Main . Peedwater i Valve Room SO 2 & 3 . m Pump Injec. V4 022 Control Lacked - - - tion to $0 2 Bldg. Level Open teolation Valve A North Main Peedwater j Valve Room i s 302&3
~
A 60
VOGTLE ELECTRIC GENERATING PLANT ; RISK. BASED INSPECTION GUIDE Auxiliary Feedwater System l l - TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd) j .. Desired Actual Pow. Sup. Required Actual , i, Description ID No. Location Position Position Breaker # Locatiot Position Position , TD Puenp Irtjoc. V4023 Control Locked - - - l tion to 80 2 Bids. Level Open Step Check A North Valve Main Peedwater i Valve Room ' SO2&3 P TD Pump lajec. V4-023 control Locked - - - tion to 50 3 Bids. Level Open 1 Isolation Valve A North Main Foodwater , Valve Room ! SO243 TD Pump injec- V4-026 Control Locked - - - tion to 50 3 Bids. Level Open L Stop Check A North ', i Valve Men Feedwater i Valve Room 502&3 SG-2 Inlet Test V4-082 Control Locked -- - .- Line Isolation Bldg. Level Closed i Valve A North i Main Foodwater i Valve Room 502&3 . ! 6
$0 3 Inlet Test V4 083 Control Locked - - -
l Line isolation Bids. Level Closed Valve A North Main Feedwater , Valve Room 302&3 e Aux. Poodwater P4-003 Aus. -- - I AA0217 Control Bldg. Racked in i Pump 3 Feedwater 14 vel A 1 Pump House Train A AFW Pump Room o 4- ! A 1
.w ,
=,e a, r, ,...,,e-, + , , . . . , - . .. , . ~ ,
m, , - . , ~ . ,
m . 0,+ ' ;,. e a. VOGTLE. ELECTRIC GENERATING PLANT- - i RISK BASED INSPECTION GUIDE , .
. Auxiliary Feedwater System -
- TABLE A 8 2 , MODIFIED SYSTEM WALKDOWN (Cont'd)-
%y Desired Actual Pow.' Sup. .
- Required Actual -
Description - ID No. Location' Position Position Breaker # - . Location Position Position
- y
- *' Am. Peedwater - P4-002 : Au .- - IBA03 21 - Control Bldg. Racked _in Pump 2 Peedwater .- Level A Pump House g Train B. . . g i AFW Pump Room o' , i APW Pump 3 - Discharge lsola. HV 5139 - A u. Ridg. 14 vel a 1 Open - l ABB 17-- Aux. Bldg.' Level 1 Closed . 5 L tion Valve South Steam E Valve Room I APW Pump 3 HV 5137 A u. Bldg., Open ' 1ABB 37 Aux. Bldg. Closed' Discharge Isola. Level'1 Level l-l - T tion Valve South Steam Valve Room [ I APW Pump 2 - Discharge Isola-HV 5132 Contml Bldg. North Open 1B88 38 Au. Bldg.' 14 vel 1 - Closed [ tion Valvc - Steam Valve Room AFW Pump 2 HV Sl34 Control - Open IBBB 39 Aux. Bldg.' Closed,
~
*- Discharge Isola .. Bids. North level 1 g tion Valve Steam Valve
. Room 1 APW Pump 3 FV 5155 . Au. Open l ABF 20 - Diesel Gener. Closed Minimum Flow Feedwater stor Bldg.
_ Valve Pump Hovse - > Tr & A L AFW Pump , " Room E AFW Pump 2 - FV 5154 Aux,- Open - IBBF-06 Diesel Gener. Closed [ Minimum Flow Feedwater stor Bldg. g Valve : Pump House p,_ Train B . . , R APW Pump n Room . E' i E CST 2 to AFW. }{V5118 Au. ' Closed IBBF 12 Diesel Gener. Closed [ Pump 002 Isola. Feedwater stor Bldg. k tion Valve Pump House I Train B AFW Pump Room W IE I. A-62 E = = . E
.- u, y
.I VOGTLE ELECTRIC GENERATING PLANT y
. RISK. BASED INSPECTION GUIDE, '
j Auxiliary Feedwater System 7 TABLE A.8 2 MODIFIED EYSTEMiWALKDOWN (Cont'd) Desired Actual ' . Pow. Sup.; ' Required Actual Description ID No. Location _ Position Position c Breaker # Location Position Position CST 2 to APW- HV.5119 ' Au. > Closed 1 ABF 12' Diese! Gener. Closed Pump 3 Isola. Feedwater stor Bldg. tion Valve ' Pump House . Train AL l
~
APW Pump Room' CST 2 to AFW HV.5113 - Au. Closed , [CDIM 06 Control Bldg. Closed Pump 1 isola- Feedwater u vel B q tion Valve Pump House 1 TD AFW Pump Room Steam Supply to HV 5106 Am. Closed ICDil105 Control Bldg. Closed TD Pump Peedwater Level B Pump House ; TD AFW 4 Pump Room Steam Supply HV.3009 A u. Bldg. Open Control Bldg. Closed IBD1M-01' ; (SG.1) to TD South Steam uvel B i Pump Valve Room i ficam Supply HV 3019 - Control Open I ADIM 01-- Control Bids. Closed - - (SG 2) to TD Bldg. Level Level B Pump i North l Steam Valve Room ! TD Pump to HV 5122 Au. Bldg. Open ICDIM-02 Control Bldg. Closed , Steam Generator South Steam 14 vel B ' ,
! Valve Room 1D Pump to HV 512S Connel
- Open ICDIM Control Bldg. Closed Steam Generator Bldg. North Level B 2 Steam Valve ,
Room ; TD Pump to- HV 5127 Connel Open ICDIM 04 Control Bids. Closed Steam Generator Bids. North 14 vel B . 3 Steam Valve i Room TD Pump to HV 5120 A u. Bids. Open ICDIM-01 Control Bldg. Closed Steam Generator South Steam Level B 4 Valve Room
- i A-63
o y 7, , ,1 p + i
' 4 4 3;3
- -c3--
}
[, .s a- VOGTLE ELECTRIC GENERATING PLANT' - RISK BASED' INSPECTION GUIDE , Auxiliary Feedwater System-TABLE A.8 2 MODIFIED SYSTEM WALKDOWN (Cont'd)
.p Desired Actual Pow Sup. Required - Actual Description ID No. Location Posidon Position Breaker # Location Positien ', Position
. TD Purny Trip 5129 Auz. Open' ICDIM . _
Connel Bids. Closed' and Throttle = Peedwater 1Avel B A . Valve Pump House TD APW Pump Room V
\'
1 A-64 I e d k. .
, s. . -
-if $
g
?
.t I
e , a i 1
.e lll lll .
lgl
^
(-;.Il j *- T ro t - e ' e l oHl . % roe- o:: %roKlioni %-- Ki 01:1 %-- i . I 44 .' M .3 :
-p 3 M 'h .i, ,
L j
.o-< i .3 m . ' '. . . 3 4
, .w ,S . y.
s 3
.. q
- -il,' ' $ If l .' J. Y, f-( J, h, . , . 5 "", b. U.' 'i
/. _;
' -- n
. . . . . .g: 7
-4. . ...: .,y.r -
4 - i
. =:n C .
.__gl-
.c-s }i)d .L '.cf, ,+
l ,
. -a,- W .;J .
)c,(
' 1, c; -
) :- )
e i ;. 9 e.p. 's :g[
- 3 -
l [7':. . ej;;%p r i ;;g %w i. a. e i Ow
~
. . - .., .w , .
I t' I , M. I i' I- I . O ': g l ,1 l. ,1 '- "] ,I g y} .' ? lt{. .'- It{ ltl , g
. s ....
,s l >. .
e .. . P. Q 1,- Lt I J 'I - I f (tSI [ j. I I i,
-f 'I ti,I '""
l } d l(4 .g .d !.hl hf- - 0! lO' lI t o! lLjdjj,[__ _t; g
'! I .$ 3 I = .? ! I-
! I$ I I
m f Q 3 I ,
'I i
il it .i
'], ,::jtsf?f! S s kI ,
75Ic ii .,s : t .
-{
i
'l .,g ' T f jl1 'PQ
-a - - .
'g
'"I
' Jf, wg .
I l .
,f"I T[ g. E'.c
<o.
, m _; . l 'M _g { ,e { l. ,,
#..-t . "' -
r
- a: '
I '.t i I I
~
i I
, j. j , %9 :
IL n.', "2.ianlL
- I t I M - 7 l
t
. : -s i
- 3 mis -
u a - i_ l -
'i m
, o, i o l- " g 33 CD C3
~
a_ - I $ 8,[ 1"
-
- s 4 l
,, ... n. L
. . a ' ,,
i ,_; g' l . j y I g_i; 1- o m. - i it i 31 - ' a . -- l i ,s l
- A-65 i
i.
+
- i, TABLE A.8 2 (Cont'd)
REFERENCE DOCUMENTS TITLE - I.D. NO. - REV DATE 1.1Vogtle Electric Generating Plant Procedure 11610-l* 5 2/20/89, "Auxiliar, Feedwater System Alignment" >
"2. Vogtle Electric Generating Plant Procedure 13610-li 8 5/30/85'
" Auxiliary Feedwater System"
' 3.' Vogtle Units 1 and 2 Final Safety -- Section 7.3 Analysis Report
- 4. Fresco, A., et al.~,," Review of the Vogtle Units
- NUREG/CR-4228 5/85 1 and 2 Auxiliary Feedwater System BNL--
Reliability Analysis" A-66 9 iF
Il e 4 J. + VOGTLE ELECTRIC' GENERATING PLANT H+ <
' RISK BASED INSPECTION GUIDE -
1 Emergency-Core Cooling System / Low Pressure Injection (LPI) l Table A.9-1. :Importance Basis and Failure Mode Identification-i CONDITIONS THAT CAN LEAD _TO FAILURE ': 1 i Mission Success Criteria ~. q
-1
- The LPI system injects borated water from the Refueling Water Storage Tank (RWST) '
into the RCS to provide l core cooling water during the injection phase of a large break - , LOCA. The residual heat removal (RHR) pumps are utilized to provide the injection flow. 1 In addition, four Accumulators are available to flood the core with borated. water-
~
immediately following a large break LOCA. They are designed to minimize core damagei l' until the safety injection pumps can provide adequate water for core cooling. Each tank i,s 3 pressurized with nitrogen at 650'psig and contains a nominal water volume of 950 gl lwith ' a minimum boron' concentration of 2000 ppm.. . i The accumulators are self contained, self-actuating, and passive in nature. Each tank
~
is connected to the RCS at one of the reactor inlets (cold legs). A check valve, held closed by RCS pressure, provides isolation during normal operation. The tanks can be isolated by -
- motor-operated valves during plant shutdown and depressurization. The accumulators are not dependent on any support systems. Three of the four tanks provide sufficient water to - i
~
' cover the core following a Design Basis' Accident (DBA), assuming the contents of one of -
the four tanks spilled throbgh the break. The LPI system can be aligned to take suction from the containment sump an'd-- ! maintain a borated water cover over the reactor core for extended periods of time in'the recirculation phase. System alignment for recirculation is accomplished through both manual and automatic actions. The LPI system consists of two residual heat removal pumps taking suction from the ; RWST discharge header. Upon receipt of a SIAS, the two pumps will start automatically.
. When RCS pressure drops below 200 psig, the LPI will begin to deliver flow to the cold legs. Mission success is provided by operation of one of the two'RHR pumps.. j A-67
Importance laspech9e ' Accident Dominant Failure Modes Sequence Category Activities ,
- 1. Accumulator failure including check valve failure or plug- 11 S.M.T ging of MOVs Check Valves:
Loop 1: 079
- 1. mop 2: 080 Loop 3: 081 loop 4: 082 Normally Open MOVs:
loop 1: HV 8808A Loop 2: HV 8808B Loop 3: HV 8808C Loop 4: HV 8808D
- 2. Operator failure to isolate interfacing LOCA 4 H O RHR Train A: HV 8701 A and B RHR Train B: HV 8702 A and B These MOVs are controlled from the main control room and are interlocked such that they cannot be opened if RCS pressure exceeds 425 psig and automatically close before RCS pressure increases to 750 psig.
- 3. Operator failure to sucesssfully switch from LPI to the 2 H O recirculation mode including valve alignment errors Based on low low RWST level, MOVs 8811 A and B open automatically, operator action is then required to:
Cose MOVs 8812 A and B Cose MOVs 8813, 8814 and 8920 Cose MOVs (8508 A and B) and (8509 A and B) Cose MOVs 8716 A and B Open MOVs 8807 A and B Open MOVs 8804 A and B Cose MOVs LV ll2 D and E Cose MOV 8806
- 4. LPI pump (s) fail to start or fail to run including common 1,3 H S.M T.C cause failure Train A: P6-001 Train B: P6-002
- 5. Failure of recirculation auction (containmei.: sump) 1,2 H S,M,T,C valve (s) to open MOVs 8811 A and B open automatically based on low-low water level in the RWST
- 6. Failure of LPI suction valve from RWST to close 1,2 M S M,T,C MOVs: 8812 A and B Operator action is required to close these valves.
- 7. Failure to realign system after testing 3 M O
- 8. Cold leg isolation valve falls to close 2 M S.M,T,C MOV: 8809 A and B Operator action is required to close these valves.
h A-68
t t n-
., C Accident importance laspection
_ Dosissat Failure Modes Sequence Category Activities , _ 9. , Pump disch rge crossover valve fails to close 2 .M - S.M.T.C . MOVs: 8716 A and B
. Operator action is required to close these valves as part of the procedure for switchover to the recirculation mode.
- 10. Failure to switch from cold leg to hot les recircula' tion 2 M S.M T- !
laitiation of hot les recirculation is an operator action-required at approximately 16'h after the accident. The i following valve actions are required: Close MOVs 8809 A and B Open MOVs 8716 A and B - Open MOV 8840 Close MOV 8821A -
- Open MOV 8802A f j Close MOV 8821B ;
Close MOV 8835 j
-.Open MOV 8802B {
11 LPI pump retuns line (miniflow) valve fails to open (or 1,3 M- S.M,T.C 1 mmain open) and operator falls to stop pumps, o: < q MOVs: 0610 and 0611 These valves open when flow is less than 500 gpm and . 1 close when flow is greater than 1000 gym. I
- 12. Containment sump plugs 1 1, S,M !
- 13. l.P hot leg recirculation discharge valve fai's tr, open 2 L ShlT 4 Train A: MOV 8802A -
Train B: MOV 8802B-
- 14. Heat exchanger cooling water valves closed or plugged 2 L S.M.T.C !
Manual, locked open valves: . l Train A: HV.ll818 and HV.ll819 Train B: HV 11822 and HV Il823 15, Irdection isolation valves fall to remain open, rupture, or 1,2,3 L S.M,T,C fall to remain closed Cold leg isolation valves: 8809 A and B Hot leg isolation valves: 8802 A and B
- 16. Recirculation suction valves rupture / fall to remain closed 2 L S M,T Valves that must close during the recirculation mode of l operation include-l MOVs: 8812 A and B (RWST to RHR pump suction) l 17, injection check valves rupture, fall to open, or fail to 4 L 'M T mmain open Cold les: Loop 1: 147 Hot leg: Loop 1: 128 Loop 2: 148 loop 2: 123~
Loop 3: 149 Loop 3: 122 Loop 4: 150 1Aop 4: 129
-A-69 l
J s-
x . .. Accident importance laspec' tion _ .
~'
Demissat Failure Modes : : Sequence Category Activities , . e
'18. Pumps ravallable due to malnenance ..
1,2,3 - L, M ! (Only r.,as RHR should be out for maintenance at any time) . .l
'fral'n'A: P6 001 . , . il Train B:~ P6 002 19 Lifting of system relief valve below set point . 3 L S.T Relief valves: PSV 8708 A and B (Relief capacity of 20 gym at 600 psig) l 1
I
'i A-70
W. l')
.f+ ~
in
^
.' . VOGTLE ELECTRIC GENERATING = PLANT RISK. BASED INSPECTION GUIDE y, Emergency Core Cooling System / Low Pressure Injection TABLE A.9 2 MODIFIED SYSTEM WALKDOWN
,' Desired Actual- Pow. Sup. Required Actual Description ID No. 14 cation Position Position Breaker # Location Position Position RHR Heat Ex. HV 0606 Aux. Bldg. Open - -. - -
changer Outlet lavel C Isolation - R C90 Valve--Train A RHR Heat Ex. V6-019 Aux. Bldg! Locked - - - changer inlet level- C Open i isolation ~ R C90 i Valve--Train A /
' ij RHR Heat Ex. HV 0607 Aux. Bldg. Open - - --
changer Outlet , lavel B_ . Isolation R C91- i Valve--Train B RHR Heat Ex. V6-020 Aux. Bldg. . Locked - - - changer W.et Level B- Open Iso': < ,n R C91 Vr < -Train B Rl; Test , V6-029 Aux Bldg. Locked - - - Recire. Isolation Level D Closed Valve R D48 RHR Test V6 226 Aux, didg. Locked - - - Rectre. Isolation L4 vel D Closed 1 Valve R N8
> RHR Pump Mo- P6-001 - Operable I AA02 09 Control Bldg. Racked in tor-Train A Level'A. i R A48 RHR Pump Mo- P6 002 - Operable IBA0310 Control Bldg.- Racked in tor-Train B Level A:
R.A50 RHR Pump In- HV 8812A- - Operable IABD-09 Aux. Bldg., Closed let Isolation Level C Valve-Train A
- RHR Cold Leg ' HV 8716A . Operable I ABD Aux. Bldg. Closed Isolation Level C Valve--Train A Containment HV 8811A - Operable I ABD 29 Aux Bldg. Closed Sump Isolation
- Level C Valve-Train A O
A-71
.VOGTLE ELECTRIC GENERATING PLANT--- ',
RISK. BASED INSPECTION GUIDE: , Emergency Core Cooling System / Low- Pressure Irijection TABLE A.9 2- MODIFIED SYSTEM WALKDOWN (Cont'd) Desired - Actual Pow. Sup. Required Actual Description-ID No. - Location Position Position Breaker #- Locauon Posi6 ion Position 4 RHR Pump. PV 0610 Operable I ABD 38 - Aux. Bldg. Closed Mininow Level C' Valve--Train A . RHR Pump la. HV 88128 - Operable IBBD-09 Aux. Bldg. Closed > let Isolation Level Bi 7 Valve-Trrin B 3 RHR Cold bg ' HV 8716B - Operable , IBBD20 Aux. = Bldg. Closed Isolation Level B Valve-Train' B - Containment ilV 8811B - Operable IBBD 29 Aux. Bldg. Cosed-Sump Isolation - Level B Valve -Train 11 RHR Pump -- FV 0611 - - Operable 1BBD38 Aux.L Bldg. ' Closed : Mininow ~ !avel B 1 Valve-Train B RHR laolation , HV 8804A - Operable IABB-05 Aux. Bldg. Cosed-Valve --Heat Level 1-Exchanger Train A to Charging , _ . Pump RHR isolation HV 8809A Operable IABB.14 Aux.~ Bldg. Closed Valve--Train A Level-1 to Cold Leg RHR ! solation HV 8804B Operable IBBB.05 ~ Aux. Bldg. Cosed Valve-Tmin B Level I Heat Exchanger , to inject Pump ' RHR laolation - HV.8809B Operable IBBB.14_ Aux. Bldg.- Cosed Valve- Train B- 14 vel I to Cold Leg RHR Isolation HV 8840 Operable 1888 24 - Aux. Bldg. Closed Valve-Hot Leg Level 1 Injection Crosa-over Hot lag Isola. HV 8702A Containment Closed IBBE Control Bldg. Cosed tion Level A Valv 'm 4 A-72
ik5 l VOGTLE ELECTRIC AENERATING PLANT RISK BASED INSPECTION GUIDE Emergency Core Cooling Sy.nem/ Low Pre sure Injection TABLE A.9 2 ' MODIFIID SYSTEM WALKDOWN (Cont'd) l Desired - Actual Pow. Sup. Required ~ Actual Description ID No. Location Position Position Breaker # Location Position Po:ltion i i HV-8702B Containment Closed IDDI Control Bldg. Closed-L4 vel B Bat. tery Room Hot Les isola. HV 8701 A Containment Closed IABB Control Bldg. Closed tion - Level B Valves.-Loop 1 HV 8701B Containment Closed ICD 1 Control Bldg. Closed Level B Bat- j tery Room j Accumulator HV 8808A Containment Open '
! solation --
Valve--Laop 1 i Accumulator- HV 8808B Containment Open isolation Valve--Lacp 2 Accumulator HV 8808C '- Containment Open isolation Valve--Loop 3
~
Accumulator
~
HV 8808D Containment Open Isolation , Valve--Loop 4 I 4
. 1
. A-73 1
1
n -
...; ~-
~ .,
g sa 3 **** amamm ee s F ensee
='**
- n -
' u__
h : -:h,:
~,,-
. ;3 . r ,_ .c.c .-
H. - _ a= I ,- _. e- 'e " A a areasname , ' , . J , g '
, W.*,,,
7, . x' 2
.,L M-[M s.Sf a p* p 1; o
.w , a
[g
+ '
g - ' .,, . *g A. , e .,
;. , . se se '.
_ ; t '. ~. -. ,9- z _ t e, f- ~
=
,, u, .., . .
s J -, *
- _L .=
. 4 ;,
~,4g., e . e 6 e y; ;
~f
*' ,-- ..;y,-g*
- y i
.g 7 p 4 ; e :, --- ,.4.,,*
c.
, t ,
-8 .._. . p s-.-
o g '*,?
, - .l._, ****
^y' - '- e.eme ~t}
'""*g *T, g ane g _, . , . ,
^
f.,=,,^ Q^ em,o
= = -
,== - - U em.es - -
; . ese . , . . , , . . .
..g l emO.
, ~
...r -
'w{ l. g __ > ?,k._
.e+
, .v. ...,,,,,, . ~ . . +y .,,
.1 n 2
,_ r+.,#. .s . t.-c 4
;, ., p
- en.e
+ "I' h
! tfl'l ^^, ', .$:L H ; %y. -.', _A s' ~ ,,,,t -
,' e.moe ;-
" 3 g f . ****
3; g f ' , 3 $,.2,-j.u
',,wJ{ :
2 sPr
.-4 f i' ,N ..}}., ~ -y)(see 4 t h-' i ,...,_r!4 t.- ' esesses *% ,r y _.g"'y
; p>>a
.-.g i ..
e .- ,*- f,a . -.. > , , , s -
** me. ;&
-q'- . xq.,,y 3
; i
,e_,.3,y.* I .7. g
,< ' ~+ , , t, _
**** r L, r,
.s _ A. . +o .c .,, . ,
.x. +.
r* g s
, + g-af' S -
;;g- .._ agggg g gg , '
ec ::- gwmg .,c.; - . w - . . . 4 s.- -
~7a :
5
' _ g. w.. ~.gi ,' y, Q. ., i c :::
4 r , ,- ,, m 5 ' .w .-19. J:-& _
- ,, 3. . . m. . . . . ,
s , r%.y^..
. . ,--. , ,h, ... A
.v 3 . ,,.,.M-
. .-t . O
- a. . n. . "
' . ;, ; rf<
)A y !iyi
,,~.... q.
+- *
.i - ". vp" , ; 4 f .d j .& *
%. w' t ,.
a .n pwc:n a: . , n . <rs ~ -- - ~. j[ w} '. f Q
),
; +s, yo - .Q Q e, en e p jy j0, 4-fgp
.- , .c
[,, p . , . c
-= w i *. . . . *
' ' h6 7".eg! f "~fD , @-'. 5').je e _ %_ *=_;. --
g,-
. .. x e i =r. .: y 4, g,
. -.,., n(
, y - .r . . - ,,
'g_,so s.
_ w -. - 4n4- .
'- U8 '
/
e64 D ~
~_
M. t
- e. . -
~& g ? % 3.
,, q
- r..e <-
wm . a .-
, o ,
i 3 i
-. pm1 __
- h. _ O ..._
~ Figure A.9-1 Vogtle 1 Residual Heat Removal System Showing Component locations (Reference SAIC 89/1540 DRAFT, Figure 3.3-4)
O e e .r
- - .. 2 --
->a.-'.sa.-- ..r-s-,.-, i i , , . . ,
' . _ _ . _ _ ' _ ___ - -.u- -
* ,' TABLE A.9 2 (Cont'd) -
REFERENCE DOCUMENTS. TITLE I.D. NO. ~REV DATE
- 1. Vogtle Electric Generating Plant Procedure 11011 6 10/7/88
" Residual Heat Removal System Alignment"
- 2. Vogtle~ Electric Generating Plant Procedure. 13011 1 11 1/21/89
" Residual Heat Removal System"
- 3. Vogtle Units 1 and 2 Final Safety Section 7.3.2 Analysis' Report, <
~
- 4. Vogtle Electric Generating Plant Procedure 11001-1 8 11/21/88 ..
" Reactor Coolant System Alignment"
-I I
l 1 4 e O A-75
y ' e
- A u
4 . VOGTLE ELECTRIC GENERATINGiPLANT .- i t RISK. BASED INSPECTION GUIDE- _, .
, Engineered Safety: Features Actuation System (ESFAS) g Table A.10-1. Importance Basis and. Failure Mode Identification i CONDITIONS THAT CAN LEAD TO-FAI, LURE [
1 Mission. Success Criteria , l The Engineered Safety Features Actuation System (ESFAS).~1s designed to' sense - . i selected plant parameters and determine whether the prescribed safety. limits are being-exceeded. If safety _ limits- are exceeded,. logic' combinations are formed and actuation signals are sent to the ESF components that are required to' respond to the particular plant . [ condition. The ESPAS consistst of two _ distinct portions of circuitry: ~ Analog circuitrycthat: 4 !! provides three to four redundant channels that ' monitor various plant parameters, and digital
~
circuitry that provides two redundant logic trains that receive input from the- analog ~ channels and provide logic to activate the requuired ESF systems. Each digital train is capable of actuating all ESF equipment. The ESFAS depends on the electric power system to provide 120 V AC for instrumentation and 125 V DC for instrumentation and logic circuits. - L The specific functions that- rely on the ESFAS ter initiation include: 9 . " 4 reactor protection system. -
- 1. Reactor trip, if not previousiy nievided .by .he.
L .2. Proper sequencing of ESF power ic:c . !:.tleding:
- a. Cold leg injection isolation valves. ,
- b. Cnarging pumps, safety injection pumps, residual heat removal pumps and ,
associated valving. . 3
- c. Motor driven auxiliary feedwater pumps.
-l 3. Phase A containment isolation. '
- 4. Steam line isolation.
5.- Main feedwater'line isolation.
- 6. Start emergency diesels to assure backup power supply.-
,' 7. Containment spray actuation.
- 8. . Nuclear Service Cooling Water and Component Cooling Water pump start.
- 9. Actuation of containment air cooling units.
- 10. Isolation of the control room normal HVAC and. actuation of the ESF HVAC for .
L the control room, control building, auxiliary building and the auxiliary feedwater pump house. A-76 4.m , ll 1 . ltn. ,, - - -
. . - . . . n . . . .. = .- --- . _ . . _ _ .
' ~ 1-
, o .11. Containment purge isolation. '
~ 12. Actuation of the reactor c:vity post accident purge . units and the ESF-chilled ,
water pumps and chillers. Accident importance laspection
- Dominant Failure Modes Sequence Category ' Activities -.,
i
~ 1.' Failure of automatic initiation logic (most criti al for aus. 6.10 tilary feedwater (AFW) initiation) through fol'. wing sce-natio.-
a) lastrurN't failure through calibration or maintenance M. O.S.M.T.C .
, : ete
- Actuation signals for the motor driver AFW pumps:
- 1. Low-low level in any steam gener stor (2/4 coincl. I
.dence)
- 2. Safety injection signal
- 3. Trip of both main feed pumps
- 4. Complete loss of electric power 'j Actuation signals for turbine driven AFW pumps:
- 1. law low level in two steam generators
- 2. Complete loss of electric power b) Logic nisys fall to close M T.M
-l c) Failum of 120 V vital AC (see Table A.21) L M.S.C
.i i
l 1
'f A-77 c
4 VOGTLE ELECTRIC GENERATING PLANT - ' RISK. BASED INSPECTION GUIDE ' , Engineered' Safety Features . Actuation System (ESFAS)- b> TABLE A.10 2- MODIFIED SYSTEM WALKDOWN The ESFAS is a normally energized system that is subject to extensive surveillance !
+ testing to assure operability. Observation of these surveillance tests and review of the test -
results will provide the inspector'with irnportant information regarding the operability of the system. System walkdown during power operation will only ascertain the' proper iii 1 alignment of certain circuits;-The following items are important during these alignment .
. checks:
- 1. Ensuring that channels are not bypssed or Inadvertently left in test;- and I 2, Determining that root valves associated with sensors and instrumentation related to -
the ESFAS are open, o
- li' \
li
~
h: ' t . i L;' ; \ E
l l' i
~
A-78 _l I t x: y ,.
i. s 1
~
i' TABLE - A.10 2 (Cont'd) . 3 REFERENCE DOCUMENTS ; i TITLE I.D. NO. - REV DATE~ i 1
- 1. Vogtle Units 1 and 2 Final Safety -
Section 7.3 j Analysis Report .. ;
.j l
i h l
- i l
m e 0 A-79
.- d. -
VOGTLE ELECTR$C GENERATING PLANT. ',
- LRISK BASED INSPECTION GUIDE .
Refueling Water Storge Tank (RWST) Table - A.11;1. . Importance Basis and Failure Mode Identification CONDITIONS THAT CAN-LEAD TO FAILURE -
- l" Mission Success Criteria ~ ,
The Refueling Water Storage Tank (RWST) is a passive component that provides the -l source of water during' the operation of four safety significant systemsi safety injectien --
-centrifugal- charging, contairunent; spray, and low pressure injection. It. is also critica. - during the switchover from the injection phase to high or low pressure recirculation from the containment sump upon receipt of an RWST low low level-signal. 'j l
The RWST contains 345,000 gal, of water that is available for the injection mode of operation. This provides a water supply sufficient for at least 19 minutes of operation of all - , pumps operating at maximum flow rates. An additional. water volume is provided below J the minimum submergence level for vortex prevention.- , During the injection phase,Lif primary system. pressure remains above the safety injection pump shutoff head, the pumps discharge to the RWST through the minimum flow recirculation lines .until' thel RCS pressure is sufficiently reduced to allow inflow. Accident importance laspection Dominant Failure Modes ' Sequence Category Activities Common cause miscalibration of RWST level season that 1.2 M O.S.C d fall realignment of high and low pressure ECCS i LT 0990 LT 0992 :i LT 0991 LT 0993
- 2. Failure to rralign system after refueling outage. (Refueling I L 0,M procedurer should be reviewed)
' ?
A-80
'l l
1
, VOGTLE ELECTRIC GENERATING PLANT ..
RISK BASED INSPECTION. GUIDE Refueling Water Storage Tank i i TABLE A.112 MODIFIED SYSTEM.WALKDOWN Desired Actual _ Pow. Sup, Required Actual l Description ID No. Location Position Position - Breaker #. . Location Position Position ; I L4 vel Transmit, X.152 RWST ' Open NA' ter Roof Valves X.153 RWST Open NA i X 154 RWST Open NA: ;, RWST NA !
^
X 155 Open t 4
- 1
-1
~.
+
I
. A-81
. s= ,
'Y!.> *
;" ' TABLE A.112 (Cont'd)
- REFERENCE DOCUMENTS .
TITLE I.D. NO. REV DATE
- 1. Vogtle Units 'I and 2 Final Safety Section 6.3 Analysis Report
- 2. Vogtle Electric Generating Plant Procedure - 13105-1 8 11/23/88
" Safety Injection System Alignment" ;
4 s l i i l O A-82 l 1 e c
. .e >-
-VOGTLE ELECTRIC GENERATING PLANT- .
RISK BASED INSPECTION GUIDE Condensate and Feedwater System Table A.12-1. Importance Basis and Failure Mode Identification _ , i CONDITIONS THAT CAN-LEAD TO FAILURE i 1 1 - Mission Success Criteria The Condensate and Feedwater system provides feedwater flow to the steam gener-ators during normal operation and following a transient. The system consists of two turbine - ; driven main feedwater pumps, three motor driven condensate pumps, the condensate - l booster pumps, and_ the hotwell. Feedwater flow for heatups, cooldowns, and hot standby. operation is provided by the motor driven auxiliary feedwater pumps. The inventory of the
' hotwell (with the CST as a backup supply) is assumed sufficient for all mission times of -i interest.
During normal operation, between 20% and 100% power level, operation .of the feedwater system is automatic. Water level in the feedwater heaters, the heater drain tank, and the condenser hotwell_is automatically controlled by a modulating flow controlLvalve. i The feedwater flow control. system is a three element system ' utilizing feed flow, steam flow and steam generator level. The main feedwater pumps 'are tripped based on a high high level in one steam generator or actuation of the safety injection _ system. The condensate and feedwater system is dependent on non-class lE DC power and instrument ~ i air. The success criterion is restoration of flow from one or.more main feedwater pumps to
- one or more steam generators.
' Accident - Importance - Inspection Dominant Failure Modes Sequence Category ' Activities
- 1. tass of Power Conversion System is a. important t:6nsient to -H event when coupled with loss of AFW. Failure modes for the PCS are:
a) FW line break with failure of operator to isolate O.T b) Failure of main FW or condensate pumps to continue S.M.T sunning Main Feedwater Pumps' Condensate Pumps - P4 004 P4 001 P4 005 - P4-002 i P4 003 c) Failure of main FW and condensate pumps to start and O run following loss of DC bus (see Table A.3-1) A-83
4
'l
,, VOGTLE-ELECTRIC GENERATING PLANT ',- -
-[
RISK BASED INSPECTION GUIDE , Condensate and Feedwater System. s TABLE A.12 2 MODIFIED SYSTEM WALKDOWN + Desired . Actual Pow. Sup.- , Required Actual i Description : ID No.. ' Location Position Pos6 tion Breaker # Location Position Position ' Condensate P4 001 ' Puenps__ ,; P4-002 pg3 ,
,[
1 Main Feedwater P4-004 I Pumps , e P4 005 ' i a
~ .
4 i e l 1 1 A 84 l
- i- +e --c , '* .-_-_____ _______. _____ _ ___.-
e .
' TABLE A.12 2 (Cont'd)
REFERENCE DOCUMENTS TITLE I.D. NO. REV DATE
- 1. Vogtle Units 1 and 2- Final Safety Section 10.4 -
i Analysis Report i
- 2. Vogtle Electric Generating Plant Procedure 11615-1 7- 9/21/88
" Condensate and Feedwater Systems- i
' Alignment"
- 3. Vogtle Electric Generating Plant Procedure 13615 1 12 '12/2/88
" Condensate and Feedwater Systems" a
l eut e t A-85
- y. . .
o VOGTLE ELECTRIC GENERATING PLANT ',. - RISK BASED-INSPECTION. GUIDE- ,. . Chem, 1 cal and Volume Control' System (CVCS) Emergency Boration Tablei A.I3-1. Importance Bash and Failure ' Mode Identification i 1 CONDITIONS THAT CAN LEAD TO FAILURE.
. t h
Miasion Success Criteria:
'Ihe Chemical and. Volume Control System (CVCS) provid_es several major functions
- during startup, normal operation, emeigency operation, and shutdown of the reactor. The
'RCS boron concentration'is normally controlled by the makeup portion' of the' CVCS However,lthere are occasions when it is necessary tolborate at a rate that exceeds the normal; maximum capability of the makeup system.fint these; situations, the CVCS is manually initiated to rapidly inject concentrated boric l acid into the ,RCS.
a
<One boric acid storage tank and two boric. acid pumps are provided to supply boric acid to the RCS cold legs during the emergency injection phase of the CVCS operation. A L batching. tank is provided for preparing boric acid for makeup to the storage tanks; The 1 boric acid pumps are started manually. For emergency boration,-a boric acid direct feed '
MOV .is' provided that comes off the common boriciacid pump discharge and' supplies , concentrated boric acid directly to the charging pump suction header. This valve may be : f opened by either a SIAS or a handswitch on the control panel. 1 It is assumed that l' out of 2 boric acid pumps are requireil to provide concentrated - boric acid ~ to 1.out of 2 charging pumps for; injection into the RCS.
' Accident? Importance laspection ' , Domlaast Failure Modes Sequence . Category Activities i
- 1. Failure to initiate and perform emergency boration. This is . tl . H O-a manual operator . action. -
- 2. Single valve failure to open, preventing boric acid now 11 M S.M.T
, MOV: HV 8104
- 3. Failure of boric acid pumps to provide sumcient now 11 M= 3.M.T
' Pumps: P6 006 P6007 4, Charging pumps unavailable due to maintenance or failure !I M- S.M.T l to run t
6 A-86
'.' i, }
4 l__-- --- -_-_l-----_-------------L---------------~--- ' ~ - - ~ ~ ' ~ ' ** ~~~ ~ ~~~'*'~~ * * ' " *
-j s> ;
- c. Li VOGTLE' .'LECTRIC' GENERATING PLANT P RISK BASED INSPECTION GUIDE .
Chemical and Volume Control /Immediate Boration System TABLE A.13 2 MODIFIED SYSTEM WALKDOWN Desired Actual . Pow Sup, . Required Actual Description ' ID No. Looetion - Position Position Breaker # Location Nition Position Boric Acid HW8104 - Closed 1 ABB 24 -- Au. Bldg . Closed i Flow Isolation -* . Level 1 -! Valve _j Boric Acid P6 006 Pumps; -t P6 007 v Centrifugal P6 002 ' 1 AA0213 Control Bldg. Racked in l Charging Pumps Level A l P6 003: 1BA0313 Control Bldg. Racked in -l 1Avel A
?
a
~ .
t
- ,h
.1 e i A-87
.'n
~
~
..~
-l
~- 3 :
1"- =::=. g. .,::::. .. g.-
.>
- d. : '
1
.. ., a. .
-p ~ '
m$ I u. *,G .
-.9 9,.
esse s. _9. esas e.e . 9_ j a h}g 2 ,
~M
" t _, '., , A
.-+4.-54 d.*. - -
_ . 9 _ _ # Eo W.,, ,
, , -m_:: . ,,,
_ w u. n . e t . -. . .s. . . . u=- .&.
, ~1 , > z . +w. .*.-
~~ .- . y. :;
- 9 .
. > ;a n .,
,~ .; i ..-.
,'x*
. 4-
.41 , , s. c. w 9., 7.~u 1. r. ,.
1 ,a . , , ..
~. , . , i .
-$e.-.g'? -
w
- a. _. g _. , ,
2 3 ; 9
' .k,-
a-M _.--
- , ;.,. m,
.a*q
. ,,m-i
, >,s
,.k. wg
> , ~1 .7a,;. m <
l fA6,. - . E 3' UN, #%N. -N N+I- y* a OM.M.g~. ., .S m
.[ -
M-- :::: i $E m.- r p%a),Ml'*$m.ND L
. . . g 2% : _
,- i
+
c .,
- m a - g3m . + c.-gs;n q ,.c ; 4..u; m qU,%: a *s m ... mm,
.f
- u. , s ., - ~,q' . e. ;:
~= %a)y m, a. _ ~*.- -... .- % {_ u1,. : _;c.; 7;pjt "
o. w
* , ; 1 t.
%, ...w... > . .,
m . . . -
%p.-p: m:_ i:,; g;:. 2*M, Q MgMi'%y,
~
. - cq .:; e.o,,- .a m, n _
~v ^
x'i :
.* y' 3%
.Tl 9 ;,. . gim M N <
w -
- p.
oo
. M y.---. -
n 1.
+ - e . mts e.ym
% s; . W
# y%
D W.;? -
- ,. .re :.
; 2e gx u
* ;.2
.,.:s :ee: b-..
- 7 w
u.n
- n. p.....
~~..
g,~. oo v, w. . w e . . .
.n:
n
..- . a . ""* p 4-h.
cg' 5 -+ 3 N 5 4. r -. f .J @.4 4/f*Ey.7 - ' , $ ,g,g,, g.g g , . aa. *==*-
; '.' ,3 7 p ?_ . ,2 ,s.-=
O 'F ,
..m-- a s n- +-
.c.- 1/;g:n' C y;p
- g, m _g_4
., qf u W,.u, T" n;M,. , ~
- -M - -
'l.._[e=f m.;- h;.-
q
-::r, ;
;m awv "
~ ~ G. p...
a- e r,.i.
.y v _....v gr w! p ::=. :: 3.;r
&. .m ,~
- i-
% m- Q,,,,,,, ' ~-, ~m~ m - uz :y (d ' ;~:#
- .w- g:pa.
y 7 n i m p. Q' g ,
' 7c
,- -~,.. -
g,s
~
g ;g ;.v:,, r. a- rgvm
,,, v.,,.
3- <!w
. m.. ,
,n., -. , b. +
e ",' n r- - - , ,.,..t
'"J Q Q, ..') afg j tQ,y*: G,y,g.s- , ik. .ys W,, m'.
,' -,';n # Q.',y-}:i.'k
,. y, , n~. #.
. cy.
.. .- .. . . u~ m;.
~.
,s.
.+,m,me~,.
g,. .
. +..
~ _ .
- u. -
. 0:
u,,
#, - ; o
. ~
w ~m. ,
, g p n u. . n
.- .m ,.,,,.o.
2 zw <9 -- . p r' m ;-
**.
- 7V -
% 4. . .
i.'1;-
%a3.. . _@.. .
/ D s9 - -
,:#~ .,( ~9 ,%lg. "~, ,
u z. ._ . .
~ .-
- a. .s: n y g,.; .. . ~. : .c %
m r- - s, N, s . - . .
., y.i.
. p ,3 4 y ^2
.~ q 7*. ,._
,$ 1 ., t.
~ M. - .
& gp 4, .,QU, ( ; & J., m. ~&s
%. , _ 3 ., .~e -
e
- ; r :--
n..~ uT'%
,. n :, u %' ,:M. *, , ' .
- m o
~
D Figure A.131 Vogtle 1 Charging System'(CVCS) Showing Component locations l (Reference SAIC 89/1540 DRAFT, Figure 3.4-2): - e 9
..,.,..,.y , . - . . . , , , , , . , , , , . . . - . . , ., . .,,.,.._,_,_..,..,,g,,. %,_._,.
y ... -:....# W
- - - +. . _ _ _ _ _ _ _ _ _ _ _ _ _ _
n!. . - i )
- t - .,
1 I APPENDIX B
'l i
l TABLES OF q L
- 1) PLANT OPERATIONS INSPECTION GUIDANCE; j
. 2) SURVEILLANCE- AND. CALIBRATION INSPECTION GUIDANCE 1
- 3) MAINTENANCE INSPECTION GUIDANCE ,
1 l l 1
=l 4
i I t I I 5 I l I e I n
>w+- .- --+ -+,, --
_ _ -,- - , , w ,e.* - ,, ,,. , -w,* , y
j i 4 l VOGTLE ELECTRIC GENERATING PLANT , RISK BASED INSPECTION GUIDE . ! Table B.1 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. . l System Failure Discussion Nuclear Service Cooling Pump train A or B out for main. Table A.1 1, item 2 Water tenance , Table A.21, item 1 ' Normal & Emergency Failure of Emergency Diesel Gen. AC Power crators (EDGs) to start or run ( [DG 1A or IB) , Failure to restore AC power after Table A.21, Item 3 i station blackout w/ concurrent RCP seal LOCA i Improper EDO post maintenance Table A.21, Jtem 5 ! valve or breaker lineup Valves: Locked Open i DGIA: HV Il705, HV 11706 DGlB: HV Il619. HV-Il620 - Breakers: > 1 AA0219, IBA0319 (Open) 1 AA0205, IBA0301 (Racked In) . DC Power Loss of 125V DC bus Table. A.31, Item 1 , Operational test or maintenance er. Table A.31, item 3 i ror resulting in a) de energizing or . cascading of DC power supplies l b) failure to properly restore bat. ' teries or charger after maintenance l l Reactor Trip System Instrument failure due to Table A.51, Item 1 } calibration / maintenance error Operator failure to manually scram. Table A.51 Item 3 l reactor following ATWS . l ECCS/High Pressure Failure to switch from RWST to Table A.61, item 1 ! Injection the containment sump via the Low Pressure Recirculation System - Operators failure to stop pumps on Table A.61, item 13 RWST low low level alarm t Y B1 I
i l
,' Table B.1 Plant Operations inspection Guidance (Cont'd)
System Failure _ Discussion Primary Pressure Relief PORY block valve closed Table A.71, item 3 System HV-8000A and HV-t'J008 l Operator :rror in breed and feed Table A.71, item 4 l activities causes lacit of RCS cool. ! ing Auxiliary Feedwater Failure to manual'y start locked Table A 81, Item 1 ; I out standby pump ! i MD pumps P4 002 and P4 003 l 4 , Failure to manually start pump Table A.81, Item 3 ! ! given auto start failure , Failuto to restore TD pump from Table A 81, item 4 ' testing. ; TD pump P4 001 (Train C) Local fault of valve 1.1 motor. Table A.81, item 6 driven pump discharge to steam generator Motor operated valves, locked open manual valves and check valves Local fault of suction valve from Table A.81, item 8 the condensate storage tank (CST) Motor-operated valves, locked ' ~ open monual valves and check valves.- Unde?ceted flow diversion Table A.81, item 10 l- Undetected FW-leakage back Table A.81. Item 11 : i through pump discharge valves ! causes steam binding Failure to increase flow to steam Table A.81, Jtem 13 generator given plugging of valve in other steam generator Failure to restore TD pump dis. Table A.81, item 15 i charge valve after test
- l Locked open, manual valve: 015 7
i ECCS/ Low Pressure Operator failure to isolate interfac. Table A.91, Jtem 2 Injection ing LOCA Motor operated valves: HV-8701 A and B HV 8702 A and B i B2
o , y Table B.1 Plant Operations Inspection Guidance (Cont'd) . .. System Failure Discussion , Operator failu's to successfully Table A.91. Item 3 switch from '.PI to recirculation ! including valve allgmnent errors ] Failure to realign system after Table A.91, item 8 I j testing
- Operator failure to stop pumps if Table A.91, item 11 pump return lin
- (n:!niflow) valve i falls to open or remain open Motor operated. valves: FCWO610 i and FCWO611 j Engineered Safety Fea. Failure of automatic initiation Table A.101, item i j ture. Actuation System logic by instrument failure through 3 calibration or maintenance error i Refueling Water Stor. Common cause miscalibration of Table A.ll 1, Item'1 l age Tank (RWST) RWST level sensors failing auto i or manual realignment of high and low pressure ECCS j LT-0990, LT-0991 LT-0992 and J LT 0993 Failure to realign system after re. Table A.ll 1, Item 2 ;
fueling outage . ; Condensate and Loss of feedwater (and AFWS) by 7:bie A.12-1. Item 1 Feedwater a) FW line break with operator (a) & (c) ! failure to isolate break ] c) Failure of MFW and condensate pumps to start or run following loss of DC bus ( ! Emergency Boration Operator failure to initiate and' Table A.131, item 1 . perform emergency boration
+
l l 1 1 j 1 B3 , i
j
- j i
VOGTLE ELECTRIC GENERATING PLANT ] RISK BASED INSPECTION GUIDE Table B.2 Surveillance and Calibration Inspection Guidance ; l The listed components are the risk significant components for which surveillance and/or ! calibratio.1 should minimite failure.
- System Failure Discussion ]
. Nucleet Service Cooling Failure of valves which isolate NS Table A.1 1, Item' I j Water flow to CCW heat exchangers i Train A: Locked operi va.!v:: ;
i HV li703, HV Il704, HV-!!701, j HV-11702 Train B: Lock'ed open. valves j HV 11616. HV-617. HV Il614, i l HV 11615 i l Pumps fall to start or run Table A.1 1. Item 3 ! 4 Train A: ( P4-001, P4 003, P4-005 Train B:
- P4-002. P4 004 P4-006 I Pump discharge MOV check valve Table A.1-1, item 4 -
or header isolation valve fails to i i open or remain open Normal and Emergency Emergency diesel generators Table. A.2 1, item 1 - AC Power (EDGs) fall to start or run DGIA and DGlB , Failure of automatic bus transfer Table A.21, item 4 to backup source of AC power re-sulting in loss of a vital AC bus ; Cooling water valves for EDO Table A.21, Item 6 closed or blocked DGIA: HV 11705, HV 11706 l DGlB: HV 11619 HV-Il620 l Failure of EDO output breakers to Table A.21, item 7 i close j Breakers IAA0219 and IBA0319 , Failure to transfer to reserve Table A.21, Item 8 source of AC power or failure of EDO start signal , Failure of inverter Table A.21, item 9 1 l 4 l 1 B-4
+
o . j Table B.2 Surveillance and Calibration Inspection Guidance (Cont'd) ' . System Failure - Discussion . l DC Power Loss of 125V DC bus Table A.31. Item 1 Failure of ofline chstger and fail- Table A.31. Item 2 ure of spare to energize on demand - l Failure of batteries Table A.31. Item 4 ! I Loss of battery room ventilation Table A.31. Item 5
. y Component Cooling Pumps fail to start or run Table A.41, item 1 l' Water System P4 001, P4 003, P4 005 Local fault of heat exchanger Table A.41, item 2 valves which isolate or' severely i reduce CCW flow or SW coolant i flow , l Pump discharge or suction valves Table A 41, Item 3 !
fall to open or remain open j
' Reactor Trip Instrument failure due to Table A.51, Item 1 'I
, 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 Train A: 52/RTA 52/BYB Train B: 52/RTB 52/BYA ECCS/High Pressure Failure of HPI discharge valves to Table A.61, Jtem 2 Injection open including common cause fail-ute (includes check valves) , Failure of recirculation suction . Table A 6-1, Item 3 valves to open including common cause failure (includes check valves) Failure of pump mintflow valve to Table A.61 Item 4 , open fails operating pump ! Electrical failures (power cable / Table A.61, Item 5 breaker) disable SI pump room 1 ventilation j Failure of NSCW system valve to Table A.61, Item 6 I open or remain open disables SI l pump room cooling ) Iocal fault of pumps / pumps fall to - Table A.61, Item 7 start or run a St Pumps: P6 003, P6-004 ) CCP: P6 001, P6 002 B-$ 1
. . _ _ _ _ _ _ . . - - - _ -~ _ ._
)
Table B.2 Surveillance and Calibration Inspection Guidance (Cont'd) . System Failure Discussion Failure of valve to open in the Table A.61, Item 8
- common portion of the HPI suc-tion line from the RWST ;
- Locked open manual valve 207 , l Boron injection tank valves or Table A.61, Item 9 l boration suction valves to charging ,
4 pump suction fail to open l' t BIT Suction Valves: l HV-8803A and HV-8803B BIT Discharge Valves: HV 8801A and HV 8801B , l Local pump failures: Table A.61, item 10
- failure of control cable to MCC 3
- failure of pump breaker to close Pump Breaker SI .
I P6 003 1 AA0216 e , P6 004 IBA0317 i CCP P6 001 1 AA0213 P6-002 1B A0213 . 4 Primary Pressure Relief PORY falls to open when required Table A.71, item 1 ! System for feed and bleed mode PV-455A and PV-456A - 4 Auxiliary Feedwater Local fault of valve in turbine. Table A.81, Item 2
- driven pump discharge to steam generator Turbine driven pump falls to start Table A.81,-Item 4 ,
or run l' ' P4 001 f Motor driven pump falls to start Table A.81, item 5 or run . P4 002 and P4-003 Local fault of valve in motor- Table A.81, Item 6 driven pump discharge to steam generator Steam supply valve or throttle / trip Table A.81 Item 7 valve falls to open (or other valve faults in steam admission line) for turbine-driven pump -l Local fault of suction valve from Table A.81, Item 8 I the CST B-6 l
.-% y - -4 - , _d ,,,_,_c- , -- - .- - v- .-. - - - -
i Table B.2 Surveillance and Calibration Inspection Guidance (Cont'd) , l System Failure Discussion - l Local fault of motor driven pump Table A.81, item 12 ! power breaker Local fault of AFW actuation sig- Table A 81, Item 14 i nal logic fails to actuate MD ,
^ '
pump and/or TD pump steam valves ! ECCS/ Low Pressure In- Accumulator fr.ilure, including Table A.91, item 1 jection (LPI) check. valve failure or plugging of MOVs , LPI pump (s) fall to start or run Table A.91, item 3 including common cause failure LPI pumps fall. to start or fall to run Table A.91, item 4 including common cause failures , P6 001 and P6 002 Failure of recirculation suction Table A.91, Item 5 :
~
(containment sump) valves to open i MOVs: 8811A and 8811B ; Failure of LPI suction valve from Table A.91. Item 6 RWST to close l MOVs: 881'!A and 8812B Cold leg isolation valve fails to close Table A.91, item 8 MOVs: 8809A and 8809B Pump discharge crossover valve ' Table A.91. Item 9 - [ fails to close MOVs: 8716A and 8716B Failure to switch from cold leg to . Table A.91,' Item 10 hot les recirculation - LPI pump return line (miniflow) Table A.91, Item i1 ' l valve falls to open or remain open, including common cause I l- Containment' sump plugs Table A.91 Item 12 LP hot leg recirculation dischars: Table A.9-1, Item 13 , valve falls to open MOVs: 8802A and 8802B Heat exchanger cooling water Table A.91, Item 14 l valves fall to open s Train A: HV l!818 and HV ll819 I Train B: HV Il822 and HV-Il823 4 . t B-7 h
.~~,w-w- ..w,, ~ , - , . . . . , _ , .
a 1 . Table B.2 Surveillance and Calibration Inspection Guidance (Cont'd) System Failure Discussion , injection isolation valves fall to Table A.91, item 15 remain open Cold Leg: 8809 A and B Hot Leg: 8802 A and B
^
Recirculation auction valves - Table A.91, item 16 rupture / fall to remain closed - MOVs: 8812A and 8812B Injection check valves rupture, fall Table A.91, item 17 to open, or fall to remain open Lifting of system relief valve be. Table A.91. Item 19 - low setpoint Engineered Safety Fea- Failure of automatic initiation Table A.101, Item I tures Actuation System logic by: (ESFAS) a) instrument failure through call- l bration or maintenance error i b) logic relays failing to close c) failure of 120V vital AC Refueling Water Stor- Common cause miscalibration of Table A.ll 1, Item 1 age Tank (RWST) RWST level sensors which falls auto or manual realignment of , high and low pressure ECCS ! LT 0990 LT 0991, LT 0992 and LT-0993 Condensate and Loss of feedwater (& AFWS) by Table A.121 Item i I Feedwater System failure of MFW or condensate , pumps to continue running ; 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 HV 8104 Failure of boric acid pumps to Table A.131 Item 3 provide sufficient flow ; Pumps: P6 006 and P6 007 '
. Charging pumps unavailable due Table A.131,' Item 4 to maintenance or failure to run q i
i 9 B-8
,. (
s , , i i VOGTLE ELECTRIC GENERATING PLANT *
. i RISK BASED INSPECTION GUIDE ., .
Table B.3 Maintenance Inspection Guidance f
'lhe components listed here are significant to risk because of unavailability for mainte- l nance. The dominant contributors are usually frequency and duration of maintenance, with !
some contribution due to improperly performed maintenance. [ f, System Failure Discussion { Nuclear Service Cooling Failure of valves which isolate Table A.1 1, Item 1 . Water SW flow to CCW heat exchangers I Train A t Inlet: HV-11703 and HV-Il704 t Outlet: HV 11701 and HV-11702 i Train B ! Inlet: HV ll616 and HV Il617 i Outlet: HV Il614 and HV ll615 i Pump train out for maintenance Table A.1 1, item 2 f Pumps fall to start or run Table A.l.1, item 3 l l Pump discharge MOV,, check valve Table A.1-1, Item 4 or header isolation valve falls to open or remain open [ Normal and Emergency Emergency diesel generators Table A.21, item I h AC Power (EDGs) fail to start or run _, ! DGIA and DG1B l EDOs unavailable due to mainte- Table A.21, Item 2 f nance ' Failure of automatic bus transfer Table A.21, Item 4 to backup source of AC power re- ! sulting in loss of a vital AC bus j 1mproper EDG post maintenance Table A.21, Item 5 , valve or breaker lineup i l Cooling water valves for EDG fail Table A.21, item 6 to o p n - ' DOIA: HW11705 and HV Il706 ; l DGlB: HW11619 and HW11620 ( Failure of EDO output breakers to Table A.21, item 7 [ close DGIA: lAA0219 DGlB: IBA0319 i t B9 ! t 9
,, , w Table B.3 Maintenance Inspection Guidance (Cont'd) l 1
- System Failure Discussion Failure to transfer to reserve Table A.21. Item 8 ' !
i source cf AC power or failure of ) l EDO start signal
- Failure of inverter or MG set Table A.21, Item 9 DC Power Loss of 125V DC bus , ,
Table A.31 Item 1 I Table A.31, Item 2
~
! Failure of on line charger and fall. ure of spare to energize on de-j mud
- . System A: lADICA and 1ADICB l
- System B
- IBDICA and IBDICB i
! System C: ICDICA and ICDICB 1 , Operational test or maintenance er. Table A.31, Item 3 I i ror resulting in I a) de energizing or cascading of DC power supplies i b) failure to properly restore bat- l 4 teries or charger after mainte-
- nance l Failure of batteries Table A.31, item 4
, Batteries: l ADlB, IBDIB, ICDlB , j and IDDIB , ( Loss of battery room ventilation Table A.31. Item 5 ; l Component Cooling Pumps fail to start or run Table A.4-1, item 1 I Water System i Local fault of heat exchanger Table A.4-1, Item 2 - valves which isolate or severely reduce CCW flow or SW coolant , flow Train A: HV Il806 and HV Il807 Train B: HV Il814 and HV 11815 - Pump discharge or suction valves Table A.41 Item 3 fall to open or remain open
- Pumps out for maintenance Table A 4-1, item 4 Reactor Trip System Instrument failure due to Table A.5-1 Item l' L
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 ! l breaker fails to open B 10 t
- - - , --- -=,---.-w-wr ,*.vw-. ,--ee,, .w - - , - - - - , y:-- -
s * *
+
Table B.3 Maintenance Inspection Guidance (Cont'd) ', . I System Failure ' Discussion - i ECCS/High Pressure j Failure of HPI discharge valves to Table A.61 Item 2 l Injection open including common cause fall. ; [ ure (includes check valves)' l Failure of pump suction valves to Table A.61. Item 3 ! open including common cause fail-l , ute (includes check valves) i Failure of pump return line ~ Table A 61. Item 4 (miniflow) valve to open fails op. - erating pump Electrical failures (power cable / Table A 61, Item 5 breaker) disable HPR pump room ventilation 1' Failure of nuclear service cooling Table A.61. Item 6 , water system valve to open or re. main open disabling pump cooling Local fault of pumps / pumps fall to Table A.6-1 Item 7 t- start or run Failure of valve to open in the . ' Table A 6 2, Item 8 common portion of the HPI suc. tion line from the RWST. Locked open, manual valve: 207 Boron injection tank suction or Table A.61. Item'9 ' discharge valves fail to open ' Local pump failures: Table A.61. Item 10
- failure of control cable to MCC
- failure of pump breaker' to close ;
Pump in maintenance Table A.6-1 Item 11 Primary Pressure Relief System PORY falls to open when required Table A.7-1. Item 1 for feed and bleed mode j PV-455A and PV-456A Failure of PORV/SRV to rescat Table A.7-1, Item 2 causing small LOCA " PORV block valve closed Table A.7-1 Item 3 ' ' HV 8000A and HV-8003B Auxiliary Feedwater Local fault of valve in pump dis- ' Table A.81, Item 2 charge to steam generator + Turbine-driven pump fails to stan Table A.8-1 Item 4 or run, or out of service due to ' maintenance or test P4-001 N B-11
.l
{ - i 1
.E' Table B.3 Maintenance Inspection Guidance (Cont'd)
\ l System Failure Discussion f Motor-driven pump fails to start Table A.8 l, Item 5 , or run, or out of service due to maintenance or test P4-002 and P4 003 , i Local fault of valve in motor- Table A.81, Item 6 l i driven pump discharge to steam t generator . Steam supply valve or throttle / trip Table A.81, Item 7 valve falls to open (or other valve i faults in steam admission line) for i turbine driven pump , Local fault of suction valve from Table A.81, Item 8 , the CST - , t AFW flow control valve in mainte- Table A 81, Item 9 ' i nance falls delivery from TD pumps Local fault of motor driven pump Table A.81 Item 12 ' power breaker a ECCS/ Low Pressure In- Accumulator failure including check Table A.91 Item 1
- jection valve failure or plugging of MOVs LPI pump (s) fall to start or run Table A.9-1, Item 4 including common cause failure P6-001 and P6-002
" i Failure of recirculation suction (con- Table A.9-1. Item 5 tainment sump) valve (s) to open HV-8811 A and liv-8811B j
- Failur. . LPI suction valve from ' Table A.9-1 Item 6 I RWSt to close MOVs: 8812A and 8812B l Cold leg isolation valve fails to close Table A.91, Item 8 MOVs: 8809A and 8809B
. Pump discharge crossover valve Table A.9-1, Item 9 '
fails to close ' ., MOVs: 8716A and 8716B ' Failure to switch from cold leg to Table A.91, Item 10 - hot leg recirculation I LPI pump return line (miniflow) Table A.9-1, Item 11 valve fails to open or remain open MOVs: 0610 and 0611 , 4 Containment sump plugs Table A.91. Item 12 B-12 6 ___..1._z_________________._____-_ _ ,.- _ , _ _ _ _ _ . , _ _ _ , . , ,,, ., ,. . , --. , --.. . r,
Table B.3 Maintenance Inspection Guidance (Cont'd) , System Failure Discussion ' l LP hot leg recirculation discharg: Table A.91 Item 13 , - valve fails to open ; t MOVs: 8802A and 8802B Heat exchanger cooling water Table A.91. Item 14 valves fall to open i Train A: HV-11818 and HV-Il819 Train B: HV 11822 and HV-11823 I Injection isolation valves fall to Table A.91, item 15 remain open MOVs:8809A and 8809B 7 8802A and 8802B i
- Recirculation suction valves Table A.91, Item 16 rupture / fail to remain closed '.,
t 4 MOVs: 8812A and 8812B Injection check valves rupture / fall Table A.91. Item 17 to opers, or fall to remain open Pumps unavailable due to mainte. Table A.91 Item 18 l nallCC P6 001 and P6 002 Premature lifting of system relief Table A.91 Item 19 valve PSV 8708A and PSV-8708B Engineered Safety Fea- Failure of automatic initiation Table A.10-1, Item l_ ! tures Actuation System logic by: a) instrument failure through call-bration or maintenance error b) logic relays failing to close c) failure of 120V vital AC i Condensate Feedwater Loss of PCS (& AFWS) by Table A.121, Item 1(b) System b) failure of MFW or condensate pumps to continue running Emergency Boration Single valve failure to open pre. Table A.131, Item 2-(CVCS) venting boric acid flow due to s power or control, circuit fault - HV 8104 Failurn of boric acid purnps to Table A.131, item 3 provide sufficient flow P6-006 and P6 007 : Charging pumps unavailable due Table A.131, Item 4 to maintenance or failure to run B-13
-------.---_______-________.u- , ,,- ..--,..__-....,,,._.% , - , . . - , w .. _,, .,c,- .m -.<2 w ..#-.. ..., t
.- 4. -o APPENDIX C CONTAINMENT WALKDO%W e e 4 4
't e
e
*- o 4
VOGTLE ELECTRIC GENERATING PLANT
,* RISK BASED INSPECTION GUIDE I Table C.1 Containment Walkdown !
Discussion Since the containment is generally inaccessible during normal plant operation, those j components listed in the preceding tables which are located within the containment are ; listed below' l Desired Actual .l Description ID No. Location Position Poaltion ; i l l RCS RHR ! solation Val > HV.8701 A Containment Closed Loop ! HV 8701B i 1 l RCS.RHR ! solation Valve HV.8702A 'I Containme.it Closed Loop 4 HV.8702B Pressurizer PORV HV.455A Containment Closed Pressurizer PORV HV.456A Containment Closed Pressurizer Safety Relief Valve PSV.8010A Containment Osgged Pressurizer Safety Relief Valve PSV.8010B Containment Gagged Pressurizer Safety Relief Valve Not PSV.8010C Containment g, , l ? Pressurizer PORV Block Valve HV.8000A _ Containment Open (PCV.455A) !- Pressurize PORV Block Valve HV.8000B Containment Open C-1 I
. - - . _ - - - , . , , ,- - - - , , n - .. - , , , , , , - , , , , .i,. . . . < .
Table C.1 Ccntainment Walkdown (Cont'd) Desired Actual . Description ID No Location Poaltion Position Accumulator Isolation Valve Loop 1 HV 8808A Containment Open Accumulator Isolation Valve HV 8808B Containment Open Loop 2 - Accumulator Isolation Valve HV.8808C Containment Open Loop 3 Accumulator Isolation Valve HV 8808D Containment Open Loop 4 k
. t i
C2 1 i
*}}