ML20090F628

Section 17.3 Re Responses to Resolution Issues Related to Reliability Assurance Program

ML20090F628
Person / Time
Site:	05000605
Issue date:	03/11/1992
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20090F622	List: ML20090F617 ML20090F628
References
23A6100AQ, 23A6100AQ-RB, NUDOCS 9203110138
Download: ML20090F628 (21)

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ment is required to included in an 0-RAP are: determine suggested O-RAP activides that will assure acceptable performance during plant life. Such activiues

1. RethbilityPerformance Monitorine: Measurement may consist of periodic surveillance inspections or tests, of the performance of equipment to determine that monitoringofSSCperformance and/orperiodicprevendve it is accomplishing its goals and/or that it will maintenance (Ref.1). An example of a decision tree that continue to operate with low probability of failure.

would be applicable to these activitics is shown in Figure 17.3.7. Asindicated,someSSCsmv requireacombination

2. Relbbility Methodoloev: Methods by which the of acuvities to assure that their performance is consistent plant owner / operator can compare plant data to the with that assumed in the PRA.

SSC data in the PRA. Amendment [U-3 ABWR =w Standard Plant uvn 9 1 Miem Pnoritiratiom identificauon, foi cach of De plant owner / operator's O-RAP will address the the nsk significant SSCs, of the imponance of interfaces with construcuan, startup testmg, operanons, thalitem as a contnbutor to its sy stem unavailabihty maintenance, engineering, safety, beensing, quaht) and awignment of prienties to problems that are anurance and procurement of replacement equipment. detected with su,h equipment. 17,3,11 D R AP Implementat!on 4. RootCause Analuis: Deu tminauon,for problems that occur regarding reliability of risk significant Anexampleofimplementationof the D RAPisgiven SSCs, of the root causes, those causes w hich,af ter for the standby hquid control sy stem (S LCS ). The purpose correction, will not recur to again degrade the of the SLCS is to inject neutron absorbing poison into the reliabihty of equipment. reactor, upon demand, providing a backup reactor shutdow n capability independent of the contn,i rods. De system is

5. "orrective 9.ction Determination: Idenuficanon capable of operating oser a wid? range of reactor psessure of corrective actions needed to restore equipment condiuons. The SLCS may or may not be identified by the to it; required functional capabihty and rehability.

final PR A as a significant contnbutor to CDF or to offsite based on tne results of problem identification and risk. For the purpose of this e xample it is assumed that the root cause analys sf SLCS is idendfied as a sigmficant contnbutor to CDF or to offsite nsk.

6. Corrective Action Imnlementaurg Carrnng out idenufied corrective acuan on risk significant 17,3,11,1 SLCS description equipment to restore equipment to its intended function in such a way that plant safety is not During normat operation the SLCS is on standby.only compromised dunng work, to funcuon in event the operators are unable to control reacdvity with the normal controt rods. De SLCS con >ists 7 Cggective Action Venficauon: Post-corrective of a boron solution storage tank, two posidve displacement actton tasks to be followed after maintenance on pumps, two motor operated injection valves (provided in risk sigmficant equipment to assure that such parallel for redundancy), and associated piping and valves equipment will perform its safety funcuens, used to transf er borated water f rom the storage tank to the reactor pressure sessel(RPV).

& Plant Acme: Some of the n. k significant s equipment is expected to undergo age related he borated solution is discharged through the 'B' deg ad: Lion that will require equipreent high pressure core Gooder (HPCF) subsystem sparger, A replacement or refurbishment. sci emauc diagram of the SLCS, showing major system components,is presented in Figure 17.3.8. Some locked

9. Feedback to Desienm The plant ow ner/ operator open maintenance valves and some check valves are not will penodically compare performance of nsk-shown Key equipment performance requirements are:

sigmficant equipment to that specified in the PRA and D-RAP,as mentionedinitem 1,abose,and at

a. Pump Dow 50 epm per pump its discretion, may feedback SSC performance

b. Maximum reactor pressure 125'O psig data to plant or equipment designers in those cases (for injecdon)

~ that consistently show performance below that t Pumpble volume in 6100 U.S. gal specified. storage tank (mmimum)

10. Procrammgic Interfaces: Reliability assurance Design provisions to permit system testing include a interfaces related to the work of the severa!

test tank and associated piping and valves. De tank can organizations and personnel groups working on be supplied w ith de mineralized water w hich can be pumped nsk significant SSCs n a closed loop through either pump or injected into the reactor. O Amenament 17 bi 1 ABWR mm Standard Plant , RlW H Q N-The SLCS uses a dissolved solution of sodium 17.3.11.2 SLCS Operalton i pentaturate as the neutron. absorbmg poison. This wlution ) is held m a heated storage tant to maintain the soluuon The SLCS is nitiated by one of three meane abos e its saturation temperature. The SLCS soluuan tank, (a) manually minated from the main control room, a test water tank, the two pouuve displacement pumps. (b) automatically minated if conditions of high reactor and awociated sahing are hicated in the secondary pres sure and pow er lescl nol below dic ATWS permisuve contunment on the floor elevauon below the operaunS po,er lesel cust for 3 minutes, or (c) automatically Ikur. This is a Seismic Category I structure, and the SLCS iniuated if conditions of RPV water level below the level equipment is protected from phenomena such as 2 setpoint andpowerlevelnot below the ATWS permissive earthquakes tornados, hurncanes and floods as well as power level exist fo 3 rn.nutes. The SLCS provides f rom intemal postulated avc ide nt phenomena. In this area, borated wata to the reactor, ore to introduce negause the S LCS is not subject to conditions such as missiles pipe reacunty eflects durmg the required condiuons. whip, and dischargmg Duids. The purnps are capabic ol ptrd uc ma disc harge pressure TomecJts neJauve reacuyh obje : Live,it isnecessary to inject the soluuon mto the reactor w hen the reactor is at for the SLCS to iniect a quanuty of t iron w hich produces high pressure condiuons correspondmg to the system """"'rnum concentrat on of 850 ppm of natural boron in i relief valve actuation. Signah indicatmg storuge tank the reactor core at 68 F. To allow for potennal leakage and hquid level, tank outlet valve postuon, pump discharge imperlect mmng m the reactor sy stem, an addiuonal 259c pressure and injection vahe position are available in the (220 ppm) margm is added to the above requirement. The control room required concentrauon is achieved accounung for diluuon in the RPV with normal water level and including the The pumps, heater, valves at d controls are powered volume in the residual heat removal shutdown coohng from the standby power supply or normal offsite power. P' ping. This quanuty of boron solution is the amouru The pumps and valves are powered and controlled from which is above the pump suction shutoff level m the separate buses and circuits so that single active failures storage tank thus allowmg for the portion of the tank will not prevent system operauon. The power suppli:d to volume which cannot be injected one motor operatea injecuon valve, storage tank discharge sahe, and injecuon pump is from Division I,480 VAC. The power supply to the other motor operated injection 17.3,11,3 Major Differences Frorn valve, storage tank outlet valve, and injection pump is ()perating BWRs from Disision 11,480 VAC The power supply to the tank heaters and heater controls ts connectable to a standby The SLCS design is very similar to that of operaund pow er source. The standby pow er source is Class 1 E from BWRs. Automatie actuation of the ABWR SLCS is an on site source and is mdependent of the atf-site power. similar to that incorporated m some operaung BWRs. Because of the larger ABWR RPV volume, the pumping All components of the system which are required for capacity has been increased from 43 to 50 gpm per pump. injection of the neutron absorber into the reactor are injection of SLCS soluuan through the HPCF sparger has classified Seismic Category 1. All major mechanical been shown by boron mixing tests to give better mixing components are designed to meet ASME Code than the operating plant injecuon through a standpipe. reuuirements as shown below, Ceamenent ASME Design conditions Injection valves of operating planis are leak proof explo: ave Code Class htauts Iemsututt vahes to keep boron out of the reactor durmg SLCS tesung. In the ABWR the injection valves are motor Storage Tank Sutie Head 150*F operated and a sucuon pipe fill system keeps the lines Pump M60 psig 150*F filled with distilled water at slightly higher pressure than injecnon Vah es 1 1560 psig 150 *F that of the boron storage tank to preclude entry of boron into the reactor, The motor operated injection va.ves Pipmg Inhottd of proude the following advantages over explosise sahes; Inguon Valves 1250 psig 575*F Amendment gy 4 ABWR unm Standard Plant mn O a Radiation exposure to personnel is potennally to start upon demand could result from electrical or reduced during testing and maintenance because mechanical problems at the pumps or the tr con trol citcuits. less work will be required at the sah es. Two AC electncal system failures that contribute to

b. - Post-mjecdon coatamment isolation capability SLCS sptem lailure are i&nufied in Table 17.3.2. No is enhanced because the motor operated vahes further details of electncal system failures or maintenance can be closed following toron injecdon.

are included here. That leaves the five components noted Explosise valves cannot be rec!osed to provide abov e for specialattention w ith regard to reducing the nsk contamment isolation, of ;ystem failure. 17.3.11.4 SLCS Fault Tree

a. Redesign De top level fault tree for the SLCS is shown m The design evaluadon of Figure 17.3.2 is used by the Figure 17.3.9, with the top gate defined as failure to designer. The design assessment shows that the component deliver 50 gpm of torated water from the storage tank to failure rates are the same as those used in the PR A, so there the RPV. Details providmg input to most of the events in

s no need to recalculate the PRA. Also, no one SSC has Figu.e 17,19 are contained in the several addiuonal a m ajor im pact on SLCS system unavailability, so redesign branches to the fault tree.

er teselection of components is not required and the seven components are identified for consi:leration by the it is assumed that the SLCS has been idenufied by the 0 RAP. PR A as a systern making significant contribution to CDF. A listing of the SLCS components or events by Fussell. Redesign considerations, if they had been required, Vesely importance was made, and those S SCs with greate3t w ould im e included trying to idenufy more reliable relief importance are given in Table 17.3.1. No SSCs appear to valves and pumps and suedon lines less likely to rlug. The be nsk significant becasse of aging or common cause latter might be achieved by using larger diameter pipes or consideradons. De seven most signifkant components muldple sucuen lines. Pump and valve reliability might are listed in Table 17.3.2, so these SSCs should be be enhanced by specinc design changes or by selecuon of considered as nsk-significant candidates for 0 RAP a different compenent. Any such redesigri would has e to actis iues. be e v aluated by balancing the increase in reliabihty agamst the added comphcation to plant equipment and layout. 17.3.11.5 System Design Response

b. Failure Mode Identification The seven SLCS nsk-significant components idenuried in Table 17.3.2 as having high importance in the SLCS If redesign is not necessary,or after redesign has been fault tree are now considered for redesign or for 0-RAP completed, the appropriate 0-RAP activides would be activities, as noted above. The Dow chart of Figure 17.3.2 idendfled for the three SLCS component types identified guides the designer, by the fault tree and discussed above. This begins with determining the likely failure modes that will lead to loss Two of the events in Table 17.3.2 result from now of of function, following the steps in Figure 17.3.3. The SLCS Ruid being diverted through relief valves back to components of SLCS have adequate failure history to pump suedon rather than into the RPV. Since gate and idendfy cridcal failure modes,so Assessment Paths A and check vahe failures (which could result in relief valve C (Figures 17.3.4 and 17.3.6, respectively) would be operation) are accounted for by ecparate events, the relief followed to define the failure modes for consideration.

valve failures of concern can be considered to be valve body failures or inadvertent opening of the relief valves. For the S LCS relief valves past experience with similar Plug;ing of the suction lines from the storage tank could valves show s that the major failure modes are fluid leakage result from some contamination of the tank Huid or from the valve body and a spunous opening as result of collection of foreign matter in the tank. The pump failures failure of the spnng, the spring fastener, the valve stem or O Amendme,* g34 ABWR umow Standard Plant ma ( ) v the disk. Past pump failures fallinto tw o general categones, maintenance activities and frequencies are show n inTable electrical problems resulting in failure to start on demand 17.3.3 for each identified failure mode. De D RAP will and mechanical problems that cause a running pump to include documentation of the basis for each suggested stop or fail to provde rated flow. The pluggmg of Guid 0 RAP activity. lines genertdly results from presence of sediment or precipitation of compounds from saturated fluid. 17.3,12. Glossary of Terms Following the flow chan of Figure ' ' 3 A the designer MlFS Anucipated Transient Without S ram. would determine more details about each failure mode, includmg pieceparts most likely to fail and the frequency CDF The core damage frequency as calculated of each failure mode category or piecepart failure. This by the PRA. w culd result in a list of the dominant failure rnodes to bc considered for the O-R AP. AS ME Section Xi require ment D RAP Design Reliability Assurance Program for mservice inspection and other mandated inspecuans performed by the plant designer to assure and test would be identified, as indicated in Figure 17.3_6. that the plant is designed so that it can be i operated and maintained in such a way Examples of the types of failure modes that could that the reliability assumptions of the impact reliability of these idenu fted components are shown PRA apply throughout plant life. in Table 17.3.3. De table is not a complete lisung of important failure modes, but is intended to indicate the ~ fussel!- A measure of the component contribution types of failures that would be considered. Vescly to system unavailability. Numerically, Imporance the percentage contnbution of component c. Identification of Maintenance to system unavailability. (O Requirements () GE NE GE Nuclear Energy, ABWR plant For each identified failure mode the appropriate designer. mamtenance tasks will be idenufied to assure that the failure mode will be (a) avoided,(b) rendered insignificant, Ca ncr/ The uulity or other organization that owns or (c) kept to an acceptably low probability. The type of Operator and operates the ABWR following maintenance and the maintenance frequencies are both construction, important aspects of assunng that the equipment failure rate will be consistent with that assumed for the PR A. As 0-RAP Operauonal Reliabilh Assurance indicated in Figure 17.3.7, the designer would consider Program performed by the plant owner / pe riodic testin g, performance testing or periodic preventi ve operator to assure that the plant is operated maintenance as possible O RAP activities to keep failure and maintained safely and in such a way rates acceptable, thatthe reliability assumptionsof the PR A apply throughout plant life. For the SLCS relief valves, which normally have no cycles during operation. A visual inspection for leakage Piecepart Aportionof a(rtsk significant) component and periodic inspections of internals are judged to be whose failure would cause the failure of appropriate. De pumps can be functionally tested the component as a whole. The precise periodically for ability to start and run and vibration can be definition of a "piecepart" will vary measured during functional tests to detect potential between component types, dependmg mechanical problems. Detailed disassemb;y,inspecuon upon their complexity, and refurbishment would be done less frequently. To prevent line plugging the storage tank can be sampled for PRA Probabilistic risk assessment priorned sedimen t and/or liquid saturation, with appropriate c lean mg to identify and quanufy the nsk associated or temperature increase as necessary. Examples of with the ABWR. OV Ameniiment g7 3a ABWR maman Standard Plant nu O Rist. Those SSCs u hich are identified as Sigmficant contribudng significantly to the sy stem unavailabihty. SSCs Structures, systems and components identified as being important to the plant operation and safety. 17.3.12 Reference (1) E. V. Lofgren, et. al.,"A Process for Risk-Focused Maintenance", S AIC, NUREG/CR 5695, March 1991. l l i l O l l I 1 l 9 Amendment g7 pg ABWR memw Standard Plant nov a O Table 17.3.1. SLCS Components with Largest Contribution to System Unavailability COMPONENT FUSSELL-VESELY IMPORTANCE OVF00lHW Flow Divened Through Relief Valve F003A 0.50 OVF002HW Flow Diverted Through Relief Valve F0038 0.50 OFLOOOHW Plugged Suction Lines From Tank 0.24 OPM00lHW SLCS Pump A (C001 A) Fails to Operate 0.05 OPM002HW SLCS Pump B (C0018) Fails to Operate 0.05 ECA003H AC Power Cable 03 Failure 0.05 ECA013H AC Power Cable 13 Failure 0.05 O Table 17.3.2. Risk-Significant SSCs for SLCS Relief Valves F003A and F003B Suction Lines from Tank Pumps C001 A and C001B AC Power Cable 03 AC Power Cable 13 O Amendment g7 3_n ABWR -m St.alld..ard Plant. mn TA B LE 17.3.3. EXAMPLES OF SLCS FAILURE MODES & 0-RAP ACTIVITES COMIONENT FAILURE RECOMMENDED MAINTENANCE B ASIS

• MODE /CAUSE MAINTENANCE INTERVAL Rehef Valve Body leakage Visual inyecuon N months Expenence Spurious opening, inspect closure for 10 years Low failure rate; spnng failure breaks; measure ASME Code 151 spnng constant; replace spnng.

Spurious opening. Visual inspection of 10 years Low failure rate; spnng tastener failure spring fastener; ASME Code 151. replxe if necessary. Spurious openmg. Visual and penetrant toyears infrequent use, low failure of salve stem inspecuon of stem, failure rate, ASME or disk ultrasonic inspecuon Code 151. of stem; replace if netentry. Pump Fails to start, Functional test of 6 mor'hs Expenence with electncal problems pump with suction other electncal from test tank, no pumps. flow from storage rank Fails to run, Measure pump 6 months infrequent use, httle mechanical problems vibrauon dunng wear. pump operauon in functional test. Disassemble / inspect 5 years Infrequent use. tow pump for corrosion, failure rate MME wcar. Refurbish as CodeISI. necessary. 4.r= Sucuan Lines Lines plugged by Sample storage tank 6 months Clean system, little sediment water for sediment; chance of sediment. clean tank as necessary. Lines plugged by Sample storage tank I month Saturated solution precipitated boron for degree of most Itkely source of compounds saturatin of boron line plugging. compound < Increase tank temperature as necessary. All SLCS components have been used in operating BWRs, so there is much experience to guide owners / operators m care of the equipment. Amendment 173-10 w-- ABWR

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. Standard Plant REV. H r 1 '%Y GE NUCLEAR ENERGY VICE PRESIDENT AND GENERAL M ANAGER NUCLEAR SERVICES AND NUCLEAR PROJECTS DEPARTMEST OPERATIONS NUCLEAR QUALITY ASSURANCE OTHER DESIGN i ADVANCED SECTIONS PRODUCT REACTOR QUALITY PROGRAMS ASSURANCE PROCUREMENT ENGINEERING SERVICES f% SYSTEM Q/A coshy DESIGN DTIS SYSTEMS RELIABILITY PROCUREME'T INTEGRATION AND AND ANALYSIS PERFORMANCE SERVICES ENGINEERING ABWR PROJECT MANAGER D-RAP ~ ^ REL ILITY ENGINEERING SERVICES GENE EXTERNAL SUPPORT ORGANIZATIONS Figure 17.3.1. Typical GE-NE Organization for an ABWR Project r% } %,/ Amendment }7 31l ABWR

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_an ard Plant St d ^ uvn O RISK SIGNIFICANT SSCs IDENTIFED BY PRA SYSTEM 1 FAULT TREE COMPONENT p RECALCULATION REDE51GN Y RELIABILITY ASSESSMENT YES ARE PRA RESULTS YES IN DESIGN PHASE: y SIGNIFICANTLY CHANGED y ARE FAILUPE RATES > BY HIGHER FAILURE THOSE IN PRA? RATE? NO NO 1 Y DOES SSC FAILURE HAVE YES IS COMPONENT YES A LARGE IMPACT ON y iN A I LE. y A A AND OST SYSTEM UNAVAILABILITY? EFFECTIVE? NO NO 1 Y SSCs FOR O-RAP Figure 17.3.2. Design Evaluations for SSCs O Amendment 373_ g; e ABWR

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Standard Plant uvn n. m,.' RISK-SIGNIFICANT SSCs FOR O-RAP Y ASSESSMENT PATH A YES DOES FAILURE HISTORY i IDENTIFY CRITICAL FAILURE MODES AND PIECEFARTS? NO y ^ ^" ASSESSMENT PATH B IDENTIFY EXISTING [ IDENTIFY CRITICAL FAILURE MAINTENANCE-RELATED i t/ MODES AND PIECEPARTS ACTIVITIES AND USING ANALYTICAL METHODS REQUIREMENTS Y Y V DEFINE DOMINANT FAILURE MODES TO IDENTIFY MAINTENANCE REQUIREMENTS DEFEND AGAINST l l Figure 17.3.3. Process for Determining Dominant Failure Modes of Risk-Significant SSCs r% ,] Amendment t73 I3 9 ABWR

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INFORMATION NEEDED ASSESSMENT PATH A DATA ASSESSMENT TO INPUT FROM ACCEPTED ESTABUSH FAILURE HISTORY INDUSTRY DATA BASES CONSULTATION WITH l KNOWLEDGEABLE ENGINEERING, OPERATIONS 1y AND MAINTENANCE PERSONNEL DETERMINE THE ANALYSIS ROCT CAUSE ANALYSIS BOUNDARY (INDIVIDUAL DESIGN REVIEWS COMPONENT, COMPONENT TYPE + SYSTEM WALKDOWNS IN SIMILAR APPLICATIONS. ETC ) = Y FROM FAILURE HISTORY, CONSTRUCT LIST OF FAILURE MODES /CAUSES AT PIECEPART LEVEL g Y IF APPROPRIATE, DEVELOP FAILURE MODE CATEGORIES AND ASSIGN EACH PIECEPART F AILURE TO A CATEGORY Y OBTAIN OCCURRENCE FREQUENCY OF EACH CATEGORY (OR PIECEPART FAILURE) Y DEFINE THE DOMINANT FAILURE MODE LIST FROM DATA CONSIDERATIONS Figure 17.3.4. Use of Failure History to Define Failure Modes O Amendment g 7,y g 4 AIRVR 23 = w Standard Plant s kFv n ,m ( ) %J INFORMATION NEEDED ASSESSMENT PATH 8 QUALITATIVE AN^L I AL ENGINEERING DIAGRAMS OF ENT CRITICAL COMPONENT UNDER ASSESSMENT V PERFORM A FAULT TREE OR FMEA ANALYSIS ON COMPONENTS TO PIECEPART LEVEL Y IDENTIFY: SINGLE PIECEPART FAILURES THAT Fall THE COMPONENT'S FUNCTION (AND THAT ARE LIKELY TO OCCUR), O t LATENT PIECEPART FAILURES NOT C DETECTED THROUGH ORDINARY DEMAND

TESTING, PIECEPART FAILURES THAT HAVE COMMON CAUSE POTENTIAL, INCLUDING BY AGING OR

WEAR, PIECEPART FAILURES THAT COULD CASCADE TO MORE SERIOUS FAILURES.

Y DEflNE THE DOMINANT FAILURE MODE LIST FROM ANALYSIS CONSIDERATIONS Figure 17.3.5. Analytical Assessment to Define Failure Modes Ih w Amendment 17 3-15 ABWR ummAo Standard Plant REv i4 O INFORM ATION NEEDI D ASSESSMENT PAT}{ C IDESTIFY PLANNED

• ASN'E SECTION XI REQUIRDt1NTS

^ A. AND VENDrJR RECOMMENDATIONS a E 1 TS . EQ RI-QUIREMENTS TIiCINCAL SPECIDCA110N FOR TESTING A CAUBRA110N Y ODIER REGULATORY-MANDATED LIST ALL MAlhTENANCE REQUIRB1ENTS REQUIREMENTS AND RECOMMENDATIONS FROM ALL SOURCES Y PARTITION LIST INTO THOSE MAISTENANCE REQUIREMENTS AND RECOMMENDATIONS ACTUALLY PLANNED AND THOSE THAT ARE NOT V MAINTENANCE MAINTENANCE RECO A UT ACTUALLY PLANNED V Y RECORD RATIONALE RECORD RATIONALE FOR PERFORMING THE FOR NOT PERFORMING MAlhTENANCE THE MAINTENANCE Y IDENTIFY FAILURE IDENTITY FAILURE MODES AFFECTED AND MODES NOT PROTECTED FREQUENCY OF B Y MAINTENANCE MAINTENANCE (IF ANY) Y DEFINE THE DOMINANT FAILURE MODES Figure 17.3.6, Inclusion of Maintenance Requirements in the Definition of Faihre Modes O Amendment 173-16 4 ABWR uma&w Standard Plant nov ri %. Y DOMINANT Fall URE MODES OF RISK-SIGNIFICANT 55CS v YES ) DOES SSC R QL RE PERIODIC SPECIFY REQUIRED TESTS i 1 NO 1 Y YES DOES SSC REQUIRE SPECIFY PERFORMANCE TESTING? PERFORMANCE MONITORING NO A %_) y YES DOES SSC REQUIRE PERIODIC PREVENTIVE MAINTENANCE? SPECIFY PERIODIC PM NO 1 V DOCUMENT, FOR OWNER / OPERATOR. MAINTENANCE ACTIVITIES AND B ASES, PLUS UNCERTAINTIES FORTilE RISK-SIGNIFICANT SSCS. Figure 17.3.7. Identification of Risk-Significant SSC O-RAP Activities f~h Amendment ABWR uunuw Standard Plant ur n O \\ LTT COs'tANtaT h35 ~~ 2 g "Pa'F l M" ~ ';,' W a " 2 1 - s - -GHG)" ),h 3 jstEATra r 1 -.. ~. HPCF 'B' I i ..M.. p

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-i i I [ con g t._____________i l'- - trsT t A3x Figure 17.3.8. Standby Liquid Control System (Standby Mode) O Amendment 3 7,3.g g ABWR nonow Standani Plant ava o N,] SLC FAILS TO DELIVER $0 GPM BORATED WATER G1 F AILURE OF BOTM SLC BOTH PUMP BOTH PUMP E ORATED PUMPS TO SUCTON DISCHARGE WATER NOT DELIVER VALVES FAIL VALVES Fall AVAILABLE AT DORATED CLOSED CLOSED PUMP WATER TO SUCTION OtSCH G2 G7 G10 GIS FAILURE TO INITIATE /D \\ s) s_ MANUAL F AILURE TO AUTOMATIC F AILURE INITIATE TO WiTIATE G 25 OPERATOR SIGNAL TO FAILS TO INITIATE INITIATE FAILURE O A SLCOO1HE G23 1.00 E-o1 Figure 17.3.9. Standby Liquid Control System Top Level Fault Tree / \\ i j '% / Amcmiment 17 3-19