ML17277A616
| ML17277A616 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/30/1982 |
| From: | Eckert E, Scherer P, Wortham T GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML17277A615 | List: |
| References | |
| TAC-56674, NUDOCS 8306280456 | |
| Download: ML17277A616 (33) | |
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CONTROL SYSTEMS FAILURFS EVALUATION REPORT SEPTEMBER 1982 FOR WASHINGTON PUBLIC POWER SUPPLY SYSTEM WNP-2 NUCLEAR POWER STATION (HANFORD 2)
GENERAL ELECTRIC COMPANY, NUCLEAR ENERGY BUSINESS OPERATIONS
.San Jose, California 95125 Prepared By:
P. 'R. Scherer-Approved By:
T. R. Wortham, Manager, Technical Licensing Nuclear Control and Instrumentation Department E.
C.cF C. Eckert, Manager - Plant Transient Performance Engineering Nuclear Power Systems Engineering Department P. B. Kingston, Senior Licensing Engineer Safety and Licensing Operation 8306280456 83062i PDR ADOCK 05000897 E PDR 7-0209
CONTENTS PARAGRAPH PAGE 1.0 Obj ect 2.0 Conclusions 3.0 Analysis Methodology 4.0 Bus Loss Summary Results and Chapter 15 Comparison 5 APPENDIX A Bus Tables A-1 APPENDIX B Elimination Criteria B-1 APPENDIX C Sample I,oad Table C-1 ILLUSTRATIONS P~iuza PAGE Bus Tree 7-0209
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CONTROI SYSTEMS FAIIURES EVALUATION REPORT FOR WNP"2 NUCLEAR POWER STATION
- 1. 0 OBJECT.
This document constitutes:
4 An analysis in response to the HRC concern that the failures of power sources which provide power or electrical signals to multiple control systems could result in consequences outside the bounds of the . WNP>>2 Final Safety Analysis Report (FSAR)
Chapter 15 analyses and beyond the capability of operators or safety systems.
4 A positive demonstration that adequate review and analysis has been performed to ensure that despite such failures the WNP-2 FSAR Chapter 15 analyses are bounding," and no consequences beyond the =
capability of operators on safety systems would result.
A comprehensive approach was developed to analyze 'he control systems capable of affecting reactor water level, pressure or power in the WHP-2 plant.
This report with its attachments was prepared by the General Electric Company (GE) for the Washington Public Power Supply System (WPPSS).
Significant technical contribution was provided from Burns 8 Roe, Inc (BRI).
2.0 CONCLUSION
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This report, supplemented by the existing FSAR Chapter 15 analyses, docu-ments an evaluation of the WNP-2 Nuclear Power Station for system inter-action b ric 1'eans. The conclusion of this evaluate.on xs previously reported limits of minisnua critical power r'atio (MCPR), peak t
vessel and main steamline pressures, and peak fuel cladding temperature for the expected operational occurrence category of events would not be exceeded as a result, of common power source failures. Although new tran-sient category events have been postulated as a result of this study, the net effects have been positively determined to,be less severe than those of the original, conservative, Chapter 15 events. It should be noted that this study uses the event - consequence logic of the Chapter 15 analysis, but starts the logic chain from a specific source (e.g., a single bus failure) rather than a system condition :(e.g., feedwater runout). By approaching the study in this manner, a great deal of confidence can be placed in the study conclusions. The approach validated itself by uncovering previously unanalyzed interactions. The soundness of the total plant design is demonstrated by its being tolerant of these interactions.
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- 3. 0 ANALYSIS METHODOLOGY The division of responsibility in performing this analysis is as listed below:
TASKS ASSIGNED TO 4 DEFINE BUS STRUCTURE BRI
~ DEFINE CONTROL SYSTEMS BRI 8 GE
~ IDENTIFY LOADS 6 EFFECTS DUE TO BUS TOSS BRI 6 GE
~ DETERMINE CRITICAL IOADS BRI 6 GE .
4 SUMMARIZE CRITICAL IOADS
~ ANALYZE COMBINED EFFECTS BRI 8 GE COMPARE RESULT TO CHAPTER 15 GE 4 ANALYZE EXCEPTIONS GE
~ MODIFY/AUGMENT C1QZTER 15 IF NECESSARY
- 3. 1 DEFINE BUS STRUCTURE This step established the potential sources for system interaction by electrical means. Bus trees (see Figure 1) were constructed using one-line diagram information to show power distribution from the highest level not previously analyzed (the highest level previously analyzed is the loss of offsite power) down to the lowest level of plant distribution (Motor Control Center's, instrument. busses, etc.).
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- 3. 2 DEFINE CONTROL SYSTEMS This step established the scope of control systems to be analyzed. A complete list of WNP-2 plant systems and subsystems was compiled. This list was then reviewed to confine the analysis to only those systems with the potential to affect reactor pressure, water level, or power.
To ensure that all necessary systems were con'sidered, certain elimination criteria were established that documented the justification for not analysing that system further. If there was any uncertainty as to whether or not a system met the criteria, it was retained for further analysis.
Those systems that met the criteria for elimination were removed from the complete system list to produce the final list of control systems for analysis. This final list, reviewed by GE and BRI, is as follows:
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3.2 DEFINE CONTROL SYSTEMS (Continued)
SYSTEMS SYSTEMS Feedwater Control System Hydrogen Seal System Nuclear Boiler System Generator Cooling Reactor Recirculation Air Removal CRD Hydraulic Control System Generator Hydrogen 8 Feedwater Turbine C02 Purge Neutron Monitoring Main Generator Excitation Process Radiation Monitor System Off Gas Area Radiation Monitor System Circulating Water Reactor Water Cleanup Service Water Main Steam RB Closed Cooling Water System Condensate TB Closed Cooling Water System Turbine Control Compressed Air tube Oil P/0 TG, RFW Low'Conductivity Drains Moisture Extraction Primary Containment Exhaust Steam Instrumentation Bleed (Extractiong Steam Heater Drains .
Heater Vents Miscellaneous Drains Feedwater Sealing Steam Service Air Plant Service Water 3.3 IDENTIFY LOADS This step provided the data base necessary to determine which electrical loads were to be analyzed. A set of load tables comprised of all elec-trical loads of the control systems in Paragraph 3.2 ,was assembled by GE and BRI, each'roviding information on the loads within their respective scope of supply.
Each load was listed with its power bus source, its unique Master Parts List system number, circuit description, and failure mode on power loss with primary and secondary effects. A sample of a load table is included in Appendix C.
3.4 DETERMINE CRITICAL LOADS This step constituted the first analytical step in sorting out the loads with the potential for initiating events affecting reactor pressure, water level and power. The elimination criteria established earlier for the system list was refined in Appendix B for use in the component review for determining which individual loads were worthy of further consideration or could be deleted from the analysis. If there was any uncertainty as to whether or not a load met the elimination criteria it was retained 'for further analysis. The code associated with an elimination criterion was assigned to each eliminated load in the load tables discussed in the previous step.
3.5 SUMMARIZE CRITICAL LOADS The non-critical loads were deleted from the load tables, and the remaining loads are grouped together by their common power busses. These tables are shown in Appendix A.
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- 3. 6 ANALYZE COMBINED EFFECTS This step provided the basis for determining the worst, case combinations of load and system failures that are credible events considering the interconnection by power distribution. Using the combined effects at the lowest level bus as a starting point,, the next higher bus was postulated to fail and the total effects at that level analyzed. This process was continued up to the highest bus level. The combined effects at the lowest bus level are included in the Appendix A tables. Horst case effects at the higher levels are summarized in Section 4.. The combined effects at intermediate bus levels less severe than their associated higher bus combined effects were analyzed but not included in Section 4. Combined effects at intermediate bus levels which were more severe than their associated higher bus combined failures were analyzed and included in section 4.
3.7 COMPARE RESULTS TO CHAPTER 15 This step evaluated the consequences of all potential system interaction events initiated Py electrical means. A review of the information in the Appendix A tables was conducted "in 'the course of developing the bus summaries of Section 4. At each bus level of the combined effects analysis, the review evaluated the effects as being bounded by a specific Chapter 15 transient analysis or not. Section 4 includes these evaluations considering the worst case effects.
3.8 ANALYZE EXCEPTIONS The purpose of this step was to determine if a failure scenario not directly covered by a Chapter 15 transient analysis would be bounded by one with more severe effects. The cases of this type are included in the Section 4 descriptions of worst case failures.
3.9 MODIFY/AUGMENT CHAPTER 15 IP NECESSARY This step was not necessary in the WNP-2 analysis.
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4.0 BUS LOSS
SUMMARY
RESULTS AND CHAPTER 15 COMPARISON 4.1 AC BUS 4.1.1 SM" 1 (4.16KV)
The loss of this bus causes reactor feedwater pump lA, condensate pump 1A, condensate booster pump 2A, and circulating water pump 1A to be inoperative, and the air removal system to be isolated.
There is a partial loss in feedwater heating since the extraction steam motor-operated valves to the feedwater heaters fail open (as" is) on loss of motive power, but steam is partially bypassed to the condenser via the dump valves which also fail open.
There is also
. a slow loss in the main condenser vacuum due to the isolation'f the air removal system and loss of one circulating water pump, Concurrent wit+the 'feedwater pump,,lA trip and a reactor vessel low water level, "the reactor recircul'ation flow runback reduces reactor power to about 68$ NBR to stay within the remaining feedwater capacity and avoid scram.
If no operator action were taken, a main turbine trip due to low condenser vacuum would occur more than ninety minutes after the power bus loss and at a reduced reactor power level. 4 This event is bounded by the turbine trip already analyzed in FSAR Chapter 15 and bounded by the loss of lower bus, PP-lB-A.
4.1.1.1 PP"1B-A (120V)
The loss'of this bus causes the reactor feedwater pump turbine 1A to be inoperative. There is also a partial loss of feedwater heating, and a slow loss of main condenser vacuum due to the isolation of the air removal system, leading to a main turbine trip.
The worst case reduction in feedwater temperature has been determined to be considerably less than 83 P.
Concurrent with the feedwater pump trip a'nd a low reactor vessel water level, the reactor recirculation flow runback occurs which is intended to reduce reactor power to about 68$ NBR. (Within the remaining feed-gradually raise power to about 75$ .'f water capability). The partial reduction of feedwater heating will the power exceeds feedwater capability, low level scram will occur and this event is bounded by the loss of all feedwater event already analysed in Chapter 815.
Due to the inaccessibility during power operation of the valve that.
isolates the air removal system the main condenser vacuum loss con-tinues until the main turbine trips and the reactor shuts down.
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If no operator action were taken, a main turbine trip due to low condenser vacuum would occur more than a hundred minutes after the power bus loss and at a reduced reactor power level.
This event is bounded by the turbine trip transient already analyzed in CESAR Chapter 15.
4.1.2 SM-2 {4.16KV)
The loss of this bus causes reactor feedwater pump 1B, condensate pump 1B, condensate booster pump 2B, and circulating water pump 1B to be inoperative.
There is also a partial loss of feedwater heating due to the extraction steam to the heaters partially bypassed to the main condenser, and a possible slow loss of main condenser vacuum due to the loss of ejector/
condenser "B", leading to a main turbine trip. If ejector/condenser "A" is in use at the time, or if it is manually started, the loss of ejector/condenser "B" would not result in a main turbine trip.
The worst case reduction in feedwater temperature has been calculated to be less than 334K.
Concurrent to reactor feedwater pump turbine lA trip and low reactor vessel water level, the reactor recirculation flow runback is initiated which is intended to reduce reactor power to about 68$ NBR. The effect of the slightly colder feedwater is to raise this final power slightly but probably .still within the feedwater capability.
In the unlikely event of a main turbine trip (more than a hundred minutes after the power bus loss, and at reduced reactor power), the effects of this event are similar to those of the loss of SM-1.
4.1.2.1 PP-2P-A The loss of this bus causes a trip of feedwater pump 1B, a partial loss of feedwater heating, a possible slow loss of main con-denser vacuum, due to the loss of ejector/condenser "B". A potential delayed (~100 min.) main .turbine trip due to low condenser vacuum.
The worst case reduction in feedwater temperature has been calculated to be considerably less than 8347. Concurrent with the feedwater pump trip and a low reactor vessel water level, the reactor recirculation flow run back will reduce reactor power to 68~ NBR (which is within the remaining feedwater capability).
The partial reduction of feedwater heating will gradually. raise power to about 75$ . If the power exceeds feedwater capability. Low level scram will occur and this event is bounded by the loss of all feedwater event already analysed in Chapter 15.
Since the reactor is operating at reduced power, should a main turbine trip occur, the event is bounded by the turbine trip transient already analyzed in CESAR Chapter 15.
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The loss of this bus causes condensate pump 1C, condensate booster pump 2C, and circulating water pump 1C to be inoperative. There is also a partial loss of feedwater heating and a potential delayed (>90 min.)
main turbine trip due to low condenser vacuum.
The worst case reduction in feedwater heating is less severe than that due to the loss of power at a lower bus because the motor-operated extraction steam valves fail-open on loss of motive power; this re-duction in temperature has been calculated to be no more than 23~F.
The condensate pump 1C trip causes the remaining condensate booster pumps 2A and 2B to trip on low suction head, and causes both feedwater pumps to trip.
The reactor scrams on low water level. The scram occurs rapidly and precedes the main turbine trip and the effects of the loss of feedwater heating on the reactor.
This event is bounded by the loss of feedwater transient already analyzed in FSAR Chapter'5 and by the loss of the lower bus, PP-3A-A.
4.1.3.1 PP-3A-A The loss of this bus causes a partial loss of fehdwater heating, a loss of the main turbine oil temperature control valve, and a potential main turbine trip due to vibration.
The worst case reduction in feedwater temperature has been calculated to be no more than 51 F. This reduction in feedwater heating will increase reactor power by less than eleven percent nuclear boiler rated (NBR) power.
The worst case scenario is the unlikely event of a loss of feedwater heating and a delayed main turbine trip.
A computer analysis was performed to determine the reactor parameters as a consequence of a main turbine trip at approximately 111 percent of initial steady state power (turbine trip at ~ll6 NBR power). The results yielded a delta critical power ratio (ECPR) of less than 0.15, and a maximum vessel pressure of 1177 psia which are less severe than the consequences of the loss of feedwater heater, manual flow control, and,.the feedwater controller failure-max. demand at high power tran-sients analyzed in FSAR Chapter 15. This event is then, although previously not analyzed for the WNP-2 plant, still bounded by existing analyses.
4.2 DC BUS 4.2.1 Sl"7 The loss of this bus causes a trip of both reactor feedwater pump turbines and a potential delayed trip of the main turbine.
Following the trip of both feedwater turbines, the reactor. vessel water level lowers and the reactor scrams on the low water level.
The scram occurs rapidly and precedes the would-be main turbine trip.
This event is bounded by the loss of feedwater transient already analyzed in FSAR Chapter 15.-
PAGE Al APPENDIX A HANFORO CONTROL SYSTEM FAILURE ANALYSIS COh'IPONENT PRIMAAY SECONDARY COM8INED DC 8US SYSTEM DESCRIPTION EFFECT EFFECT EFFECTS REACTOA F EEDWATER AEACTOR VESSEL HIGH WATER CHANNEL "C" TRIPS ON LOSS OF NONE, HIGH WATER LEVEL NONE;CHANNEL LEVEIBFP TAIP CHANNEL "C" POWER TRIP FUNCTION REOUIRES TRIPPED.
"C'NDICATES 2 OF 3 CHANNELS TO TRIP R FPT'S AND MAINTUB SINE.
CONDENSEA COND.PCV.S VALVE COND PCV.S OPENS RFW FLOW REDUCED 6000 RPV WATER LEVEL 3 SCRAM GPM. DECREASE IN CONDEN.
SER VACUUM REACTOR FEEDWATEA AELAYTT.X-IARFPT TRIP INTER. BELAYTT.X.1A DEENERGI2ES NONE ~ 80TH RFPT TRIP AND 0 LOCK TO REACTOR RECIRC SYSTEM PAOVIDING RFPT TRIP SIGNAL APV LOW WATER LEVEL INVERTER TO SYSTEM SIGNALS AEOUIRED FOR AECIBC BUNSACK STEAM LEAK DETECTION TEMPEAATUAE SWITCHES DISASLE HIGH TELIPERATUAE NONE ALTER. AC E3I.N604A8. NSISA8 INPUT TO hlSIV TRIP REACTOR FEEOWATER RFW FCV 2A RFW.FCV.2A FULL OPEN RPV WATER LEVEL 3 SCRAM
PAGE A2 APPENOIX A HANFORD CONTROL SYSTEM FAILURE ANALYSlS COMPONENT PRIMARY SECONDARY COMBINED DC 8US SYSTEM DESCRIPTION EFFECT EFFECT EFFECTS REACTOR FEEDWATER REACTOR VESSEL HIGH WATER CHANNEL "8" TRIPS ON LOSS OF NONE, HIGH WATER LEVEL NONE, CHANNEL "8" INDI~
LEVEL, RFP TRIP CHANNEL "8" POWER. TRIP FUNCTION REQUIRES CATES TRIPPED O 2 DF 3 CHANNELS TO TRIP RFPT'S AND MAINTURBINE REACTOR FEEDWATER RFW-FCV.28 CIRCUIT FCV.28 OPENS 'SCRAM AT RP V LOW LEVEL3 REACTOR SCRAM AT BPV LOW LEVEL 3 RFW LOCKUP CIRCUIT STOP RFP'S IF CONTROL SIGNAL SCRAM'T ~LOW LEVEL LOST 3 IF RFP CONTROL.
'SIGNAL LOST f RELAYTT X.IB, RFPT TRIP RELAY TT-X.18 DEENEBGIZES NONE. BOTH RFPT TRIP 5 RP Pt INTERLOCK TO REACTOR PROVIDING RFPT TRIP SIGNAL LOW WATER LEVEL SIGNALS INVERTER BECIRC. SYSTEM TO REACTOR RECIRC. SYSTEM REQUIRED FOR RECIRC. BU 8ACK STEAM LEAK DETECTION TEMPERATURE SWITCHES DISASLE HIGH TEMPERATURE NONE E31.NB ISCD, N604CD INPUT TO MSIV TRIP ALTER. AC
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PAGE A3 APPENOIX A HANFORO CONTROL SYSTEM FAILURE ANALYSIS COMPONENT PRIMARY SECONDARY COMBINED DC BUS SYSTEM DESCRIPTION EFFECT EFFECT EFFECTS AEACTOA FEEDWATEA AFP TURBltlE "A"SOLENOID AFP TURBINE "A"TRIP RPV WATER LEVEL LOWER RPV VIATER LEVEL LOWER TRIP CIRCUIT AND REACTOA AECIAC AUN- AND REACTOA RECIRC AUN-BACK TO 68% POWEA BACK TO 68% POWER AIA AEMOVAL AR V-I VALVE LOSE CONDENSER VACUUM MAINTURBINE TRIP MAINTURBINE TRIP ON AR V 2 VALVE SLOWLY &100 MINUTES LOW CONDENSER VACUUM CD
~ 100 MINUTES cP 0 AEACTOA FEEDWATEA RFP TURBINE "8" SOLEtIOID RFP TURBINE "8" TAIP RPV WATER LEVEL LOWER RPV WATER LEVEL LOWER TRIP CIRCUIT AND REACTOR RECIRC AUN. AND AEACTOR RECIRC RUN.
BACK TO 68% POWER BACK TO 68% POWER O
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MAINTURBINE CONTROL VOLTAGE AEGULATOR CONTROL MAINTURBINE TRIP ON NO LOAD FOLLOWING CIRCUIT LARGE LOAD CHANGE
PAGE A4 APPENDIX A HANFORD CONTROL SYSTEM FAILURE ANALYSIS
'I COMPONENT PRIMARY SECONDARY COMBINED AC BUS SYSTEM DESCRIPTION EFFECT EFFECT EFFECTS REACTOR RECIRC RECIRC PUMP C0018 PUMP C0018 TRIP TO LOW REACTOR LOW POWER LOW POWER LEVEL SPEED I.EVEL REACTOR RECIRC RECIRC PUMP COOIA PUMP COOIA TRIP TO LOW REACTOR LOW POWER LOW POWER LEVEL SPEED LEVEL O
O OFFGAS F05 IA VALVE LOSE CONDENSER VACUUM MAINTURBINE TRIP MAINTURBINE TRIP ON F05'IB VALVE SLOWLY, %100 MINUTES I LOW CONDENSER VACUUM W 100 MINUTES OFFGAS REGEN. BLOWERS LOSE CONDENSER VACUUM MAINTURBINE TRIP MALNTURBINE TR(P ON SLOWLY m 100 MINUTES LOW CONDENSER VACUUM
~ 100 MINUTES
PAGE APPENDIX A
)IANFORD CONTROL SYSTEM FAILURE ANALYSIS COMPONENT PRIMARY SECONDARY COMBINED AC BUS SYSTEM DESCRIPTION EFFECT EFFECT EFFECTS REACTOA FEEDWATER COND. PUMP IA PUMPS INOPERATIVE D EGA EASE IN F LOW TO SEE SECTION 4.0 COND. BOOSTER PUMP ZA SUCTION OF RFP'S CIRC WATER CIRC. WATER PUMP IA PUMP INOPERATIVE DECREASE IN COOLING FLOW TO MAINCONDENSEA REACTOR AECIRC LFMG SET SOOIA LFMG S001A INOPERATIVE NONE AT FULL POWER REACTOR FEEDWATER RFP TNG IA )TURNING GEAR) HDTTNG-RFP IATAIP RPV WATER LEVEL I.OWER RPV WATER LEVEL LOWER .
BFP.MOP IA )MAINOIL PUMP) STOP MOP ~ BFP IATRIP AND REACTOR RECIRC. AUN AND AEACTOA RECIRC. AUN.
RFP AOP-IA )AUX OIL PUMP) STOP AOP- AFP IATAIP BACK TO 88% POWER BACK TO 889f POWEA REACTOA FEEDWATER 991X.A ITURBINE SPEED SW) RFP ~ IA TRIP RPV WATER LEVEL LOWER BPV WATER LEVEL LOWER AND AEACTOR RECIRC. AUN. AND REACTOR AECIRC. RUN.
BACK TO 881I POWER BACK TO BBK POWER AIR RELIOVAL AR-SPV-2A LOW CONDENSEA VACUUM DUE MAINTUABINETRIP DECREASE FEEDWATER TO AIR EJECTOA "A" LOSS IN'03 AR SPV 28 IMINUTES TEMPERATURE DECREASECONDENSER AR.SPV. I I LOW CONDENSER VACUUMDUE MAINTURBINE TRIP IN 103 VACUUM AB SPV.1.2 TO AIR REMOVAL SYSTEM MINUTES MAINTURBINE TRIP IN103 ISOLATED MINUTES BLEED STEAM BS V.39A VALVE BS V-39A OPEN ,NO I)EI BS V.36A VALVE BS V.3SA OPEN BS-V4)4 VALVE BS V 84 OPEN HEATER DRAINS BS.V.4A VALVE BS V-4ACLOSES DECREASE IN FEEDWATER BS.V 6A VALVE BS V 6ACLOSE5 TEMPERATURE AND BS.V4IA VALVE BS.V BACLOSES CONDENSER VACUUM HEATER VENTS HV V.20A VALVE HV4/.29A OPENS NONE SEALING STEAM SSV 12A VALVE SSV '12AOPENS NONE OFFGAS OG.V 129A VALVE OG-V. 129A OPENS NONE (
O COMPRESSED AIR INSTR. AIA COMPAESSOR 'IA IA COMPRESSOR NONE ~ COMPAESSORS 18 &, NONE COMP. 18 ON BUS SM 3, COMP. INOPERATIVE IC MEET REQUIREMENTS ICON BUSSM2
'i PAGE APPENDIX A IIANFORD CONTROL SYSTEM FAILURE ANALYSIS COMPONENT PRIMARY SECONDARY COMBINED AC BUS SYSTF M DESCRIPTION EFFECT EFFECT EFFECTS REACTOR RECIRC SUBLOOP HYD. PWR UNIT IF OTHER LOOP NOT AVAILABLE NO LOAD FOLLOWING NO LOAD FOLLOWING D003A ISEE MCQD Al FCV 60A LOCKS UP REACTOR FEEDWATER RFPT GOV.IA RFW PUMP IA INOPERATIVE RPV WATER LEVEL LOWER RPV WATER LEVEL LOWER AND REACTOR RECIRC. RUN- AND REACTOR RECIRC. RUN-O BACK TO 88% POWER BACK TO 88% POWER 6
'EEDWATER SYSTEM CONTROL F EEOWATER PUMPS AT LAST NO LOAD FOLLOWING WITH RPV HIGH WATER LEVELTRIP CIRCUITRY SPEED REDUCED POWER REACTOR VESSEL HIGH WATER CHANNEL "A TRIPSONLOSSOF NONE, HIGH LEVELTRIP CHANNEL "A"INDICATES LEVEL, RFP TRIP CHANNEL "A" POWER FUNCTION REQUIRES 2 OF 3 TRIPPED CHANNELS TO TRIP RFPT'S AND MAINTURBINE REACTOR RECIRC LOOP A 5 8 FLOW CONTROLLERS LOCKUP OF FC VALVES 60A 5 NO LOAD FOLLOWING 608 FLUX CONTROLLER HIFT CONTROL TO MANUAL MANUALLOAD FOLLOWING FLUX ESTIMATOR ODE
PAGE Al APPENDIX A IIANFORO CONTROL SYSTEM FAILURE ANALYS)S COMPONENT PRIMARY SECONDARY COMBINED AC BUS SYSTEM DESCRIPTION F.FFECT EFFECT EFFECTS REACTOR FEEDWATER COND. PUMP 18 PUMPS INOPERATIVE DECREASE IN FLOW TO SEE SECTION 4.0 SUCT)ON OF RFP'S COND. BOOSTER PUMP 28 CIRC. WATER CIRC. WATER PUMP 18 DECREASE INCONDENSER COOLING FLOW COhlPBESSED AIR INSTR. AIR COMPRESSOR IC COMPRESSOR IC INOPERATIVE NONE, COMPRESSOBS 'IA Ih BPV WATER LEVEL LOWEA ICOMP. 1A ON BUS SM.I, COMP. 18 MEET REQUIREMENTS AND REACTOR RECIRC. RUN 18 ON BUS SM.3) BACK TO 68% POWER REACTOR FEEDWATEB RFP-TNG.IB ITURNING GEAR) HOT TNG RFP'IB TRIP BPV WATER LEVEL LOWER RFP.MOP. IB {MAINOIL PUMP) STOP MOP ~ RFP 18 TRIP,. AND REACTOR BECIBC. RUN.
RFP AOP 18 IAUXOIL PUMP) STOP AOP - RFP 18 TRIP BACKTO 68% POWER REACTOR FEEOWATER (TURB IkE SPEED SW) RFP 18 TRiP RPV WATER LEVEL LOWER RPV WATER LEVEL LOWER 99 TIX.I8 AND REACTOR RECIRC. RUN. AND REACTOR RECIBC. RUN.
BACKTO 66ICPOWER BACK TO 68% POWER MAINSTEAM MS-V.1428 VALVE MS V.1428 OPENS NONE DECREASE IN FEEDWATER t FLOW C4 I
BLEEDSTEAM BS.V448 VALVE BS.V 448 OPENS NONE DECREASE IN FEEDWATER CO BS.V 458 VALVE BS V 458 OPENS TEMPERATURE BS.V 1168 VALVE BS.V-1168 OPENS DECREASEINCONDENSER VACUUM HEATER VENTS HV.V.298 VALVE HV.V.298 OPENS NONE MAINTURBINE TRIP INI83i MINUTES AIR BEhlOVAL AR.SPV 2C,D LOSECONDENSER VACUUM MAINTURBINE TRIP IN I83 SLOWLY MINUTES BLEED STEAM BS.V.308 VALVE BSV 398 OPENS NONE BS.V.68 VALVE BS V 69 OPENS hlAIN STEAIh MS.V.1338 VALVE MS V.1338 OPENS MS V-1378 VALVE MS V 1378 OPENS HEATER DRAINS BS V 48, 58,68 VALVE BS V.48, 68, 68 CLOSES DECREASE IN FEEDWATER TEMPERATURE AND CON.
DENSER VACUUM
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PAGE AB APPENOIX A HANFORD CONTROL SYSTEM FAILURE ANALYSIS COMPONENT PRIMARY SECONDARY COMBINED AC BUS DESCRIPTION EFFECT EFFECT EFFECTS OFF GAS OG V.1288 VALVE OG V 1298 OPENS SEALING STEAM SS V.12$ SS V.128 OPENS SS V-18 SS.V-IB OPENS MAINSTEAM MS.V.1258 MS V-1258 OPEN >NONE DECR ASE IN FEEDWATER TEMPERATURE. DECREASE BLEED STEAM BS V.60 VALVE BS.V.SB OPENS IN CONDENSER VACUUM "
BS V 31 VALVE BS-V.31 OPENS BS V.528 VALVE BS V 628 OPENS SEALING STEAM SS.V 30 VAI.VE SS.V40OPENS BLEEDSTEAM BS.V-18 VALVE BS.VALVES FAILCLOSED 'ECREASE
%c BSV BD OROPEH IER VALVE TYPE IN FEEDWATER 6 BS.V BE TEMPERATUREAND BS-V-IC CONDENSER VACUUM BS V.10D 6 BS.V 128 BS V.2C2 BS V4IF BSV 10E BS.V 2CI BS.V 2C2 BS.V.3A2 BSV3CI BS.V.10C BS.V-3C2 BS-VRC BS-V 68 BS.V.281 BS-V.282 BS.V-38 'I BS V 382 BSV48 BS V.I IF BS V 12C BS VRC PLUS OTMER BS VALVES
PAGE AS APPENDIX A HANFOAD CONTROL SYSTEM FAILURE ANALYSIS COMPONENT PRIMARY SECONDAfIY COMBINED SYSTEM DESCRIPTION EFFECT EFFECT EFFECTS AC BUS COND. PUhIP IC PUMPS INOPERATIVE DECREASE IN FLOW TO SEE SECTION 4.0 REACTOR FEEDWATEA SUCTION OF RFP'S COND. BOOSTER PUMP 2C CIRC. WATEA CIRC, PUMP IC DECREASE IN CONDENSER COOLING F LOW REACTOA RECIAC LFMG SET S0018 LFMG SET S0018 INOPERATIVE NONE AT FULL POWER REACTOR AECIAC SUBLOOP HYD PWA UNIT D0030 IF OTHER LOOP NOT AVAILABLE NO LOAD FOLLOWING NO LOAD FOLLOWING CP FCV408 LOCKS UP Ch REACTOA REClflC SUBLOOP HYD PWA UNIT D0038 IF OTHER LOOP NOT AVAILABLE NO LOAD FOLLOWING NO LOAD FOLLOWING FCV 608 LOCKS UP REACTOR FEEDWATEA AEACTOA FEEDWATER TURBINE REACTOR FEEDWATER TURBINE APV WATER LEVEL LOWER RPV WATEA LEVEL LOWER GOVERNOR IRFPTNOV.IB) "8" TRIP AND REACTOR BEGIRD'UN AND REACTOA RECIRC. RUN.
BACK TO 66% POWER BACK TO 68II POWE A COMPRESSED AIR INSTR. AIR COMPRESSOR IB COMPRESSOR 18 INOPERATIVE NONE, COMPAESSORS IA 5 NONE I COMP. IA BUS ON SM.I, COMP. IC MEET REOUIAEMENTS IC BUS ON SM.2I
PAGE A10 APPENDIX A HANFORI3 CONTROL SYSTEM FAILURE ANALYSIS COMPONENT PRIMARY SECONDARY COMBINED DESCRIPTION EFFECT EFFECT EFFECTS AC BUS SYSTEM BLEED STEALI BS V BA,4IB BS V.BAJACLOSED IF FLOW LOST DECREASE FEEDWATER BS.V-10A BS.V-IOA CLOSED IF FLOW LOST TEMPERATURE AND BS V-108 BS V.108 CLOSED ICONDENSER VACUUM . MAINTURBINE TRIP ON BS-V.13A BS V.13A CLOSED HIGH OIL TEMPERATURE BS.DV.6A BS.DV 5A OPEN PECREASE IN FEEDWATER:
BS-DV4A BS-DV BA OPEN TEIIPEIIATIIIIE BS.DV 2Al BS.DV.2A1 OPEN OECREASEIN CONOENSER)
BS-DV 2A2 BS-DV 2A2 OPEN BS DV 3A1 BS-0 VII OPEN ACUUM BS-DNA BS-DV4A OPEN BS.V.27 BS.V.27 OPEN BS VMS BS.VRB OPEN BS V 628 BS.V 628 OPEN BS.V 62A BS.V 62A OPEN PLUS OTHER BS VALVES i TURBINE SERVICE WATE MAINTURBINE OIL TEMPERATURE TSW.TCV4 CLOSED MAINTURBINE TRIP ON CONTROL VALVE HIGH OIL TEMPERATURE OFFGAS OG-V.'I 26A DECREASE IN CONDENSER DECREASE INCONDENSER DECREASE IN CONDENSER OG.V.I49A VACUUM VACUUM VACUUM O REACTOR RECIRC. SUBLOOP HYD PWR UNIT A D003A IF OTHER SUBLOOP NOT AVAIL..NOLOAD FOLLOWING NO LOAD FOLLOWING ABLE, FCV 60A LOCKS UP O
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O 5H-I (k.l to VV) 5H.1t 4.1664) a a TR-1-1S fR I 5M 15 5I. 15 (06OV) 5I.-II lkeav)
SR-1-'ll TR- l5- IZ 51 -Il (45OV) MC-1A HC t5 MC 1F MC IA HC-IC MC 1%
TR 1Z 15 SM IZ PP-1A- B PP 15-A 'fO II4-1 MC 1C PP-1A.S Pp-1K B P IF A, MC. IS HC ID n
HC B HC IR FAoH FROH HC 1F PP 1F-B PP.IS A
.HC 15 h MC"IC 5 pp 15 MC R8 u5-PP FROH 5Z.I HC 1C A PP-1R I)5 PP-A BATTSR'/
bl I 51- t b A'f T 6 IIV Sl-Ct PP 1CA.A Q. ~ 8 E<1.5 fO 52-1 II4 5 DP.51 IA pp-7A. C 'TO II4 I PP-1A BA'TTSRY-BZ.I = HC-52 IA - I-IQ pp-1A.C A FROM RP5 II CTZ.F001 p 1A.A MC.52 IS DP 51-ID BATTYR'f HC- wA-A Bo IA HC ID AI4)40)4CIA'TIOt4 CO BREAAGR TRIP P P-1A.A.A LS PP 50" A PP 7A 5 BATTSR'f Bu5 ove RI.OAD TRIP SO.IB Qw COt4TROI 594tTBM5 PP-Ta-f FAII.uR65 At)At.9515 Bu5 TRGQ 9 su5 uuOsR vOiTASe W t4P-2 FI6UREl SQLOF4
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IR'8'81 VIL- 85.82 51.81( 4 eOV) HC. SA HC BF HC SA HC SC MC.SB TR 82-85 Olh- b2 ~PP.SA-/L, P.'lC B ~PP.'55.5 HC.85 MC BC MC BBA PP BA.C RPS (8) FROM Cl2- POOI PP-BF A.A T P ~aC.A z . CIC-BR PP bAS T. .
P'5CC IhC.88m PI'BA C A CIC.bC 8 PP-6R fO pp.le z= PP. BA PP- SA 8
HC 'SD BATTtRV 2 FP-BA. 2 E Cl-4 BI MC-Bhk CO ZA BATTBRV BO- CA PP SP.A BATTBRV BA'CYQRV bO 'ZB HC 5P.A 51.2 81-2 ~PP.BA.A A CO 25 r PP BCA.A CI.Z DP-SO-B P '5 .AA FROM PP BA-F C. SI. ZD I4C ZC - HC ZC PP SP A b OP I ZD PP BAZCA IN 2 FROM 51-2 PP.SI.2A pP- BA-A DP 51.25. PP.BA g COHTROI 5V 5'T5HB FAIl.uRBS ANAI (SIS BILS 'Tlh68 INHP- Z FIGURE I SggOF4
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FAILlIRQ5 ANAL'(515 SII5 TR85 WHP 2 FIGURE ) SH 4 OF 4
APPENDIX B ELIMINATION CRITERIA Elimination Criterion Basis Components whose failure effects, are clearly bounded by a dominant failure effect on the same bus can be eliminated by inspection. An example would be the loss of several trips such as feedwater turbine ovezspeed trip on the same bus as the solenoid that, controls all remote trips. The solenoid loss is clearly the dominant effect. Also in the case of identical components, only one of the components on that bus need be listed.
N2 Instrumentation with no direct or indirect controlling function or passive input (such as a permissive) into congfol logic. Instrumentation and other dedicated inputs to the process computer," as well as the computer itself, may be excluded. Operator actions as a result of indications are not considered control functions for the control systems failure analysis.
N3 Control systems and controlled components (pumps, valves) which have no direct or indirect interaction with reactor operation/parameters. Examples are communications, most unit heaters and controls, lighting controls, ventilation control systems for exterior buildings, machine shop equipment, refueling or maintenance equipment controls,-
etc.
Control systems and controlled components (pumps, valves) that do interact or interface with reactor operating systems but which cannot affect the reactor parameters (water level, pressure or reactivity) either directly or indirectly. Examples are : some offgas components, area radiation monitors. Valves that fail as is and in a normal full open or close position are also in this category.
N5 Systems which are not used during normal power operation.
For example, eliminate start-up, shutdown or refueling systems not used during normal operation.
N6 Some lube oil pumps are powered from AC busses but have a back-up pump powered from a DC source. Since a single electrical failure cannot disable the lube oil function these components can be eliminated from the analysis.
Requires further analysis.
" In some cases more than one of these criteria may apply.
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EPL GENERAL EFFECTS SPECIFIC EFFECTS OF BUS LOSS TO SYSTEH (a LIHITATIONS ON THE EFFECTS (MITHIH r
SYS. '5 CAPABILITY C 10 Hlti OH INPUTS REQUIRED POMER SYSTEH CONtROL AND TO PERFORH ITS EFFECT EFFECTS 0 REACTOR MATER FROH DTHER SYSTBt5 BUS (AE SYSTEH INSTR UHENTAT ION PRINCIPAL FUtlCTION OH SYSTEH OH OTHER D LEVEL, PRESSURE AND EFFECT ON LOSS OES IGNATION INCL. HPLI LOADS OH BUS OUE TO LOSS OF BUS SUBFUNCTIONS SYSTEHS E OR CPR OF INPUT
'F Power Supply RFM Loop 04-A51 Hone Indicator on BO-A Hone gp D Hone RFM-P-18 discharge ~
399 in 80-Gll (Reactor RFM-PT" 18 5 process computer indication 'ressuro (PP-BA-A) Feedwater P 1-18 indicate min. lost in control room. .
(CKT-33) 822) SRU-2 pressure A0102 Loop 04-A53 RFM-FCV"28 will go Low flow indi- Reactor Vessel A RPV level'ill High condenser level RFM-FT-28 full open, by passing cation and signal level will drop drop annunciated in contr'ol SQRT-28 feedwater to wain to RFM-FIC-28 on slowly, opera- room (Cond-LS-ZN).
SRU-8 condenser feedpumps 80-A causing tor aust take F IC-28 will increase to 115K sin flow air action to lower E/P-28 flow to try to main- to RFM-FCU-28; reactor power tain feedwater. flow (fully opening o prevent scram to vessel valve) t Level 3 H620-504.2 Loop 04-T52 Hone Indicator on BD-T Hone D None hP across HP heater 8 RFM-OPT"38 indicates ain. indication lost.
OPI-38 diff. pressure SRU-56
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~ 0<<1<<<<00 F 0 ' ~ <<1 '01<<
REFEREttCES:
FSAR 10.4, Table 10.4-2 H634 04-ASl/2 H504, Rev. 36 , H634 04-A53/3 H620/504-2, Rev. 1 H634 04-T52/0
<<"Code Classification for Effects on Reactor Parameters:
"A" - Immediate (cl minute) and Direct "8" - Immediate but Indirect h]r/C06294-44" "C" - Effect is Delayed
<<0" - No Fffect an Reactor Parameters (<10 minutes) "
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