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 C.cF E.
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 1.0 2.0 3.0 Object Conclusions Analysis Methodology PAGE 4.0 Bus Loss Summary Results and Chapter 15 Comparison 5
APPENDIX A APPENDIX B APPENDIX C Bus Tables Elimination Criteria Sample I,oad Table A-1 B-1 C-1 P~iuza Bus Tree ILLUSTRATIONS PAGE 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
S 4
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 t
previously reported limits of minisnua critical power r'atio (MCPR), peak 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.
7"0209
- 3. 0 ANALYSIS METHODOLOGY The division of responsibility in performing this analysis is as listed below:
TASKS ASSIGNED TO 4
DEFINE BUS STRUCTURE
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DEFINE CONTROL SYSTEMS
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IDENTIFY LOADS 6 EFFECTS DUE TO BUS TOSS
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DETERMINE CRITICAL IOADS 4
SUMMARIZE CRITICAL IOADS
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ANALYZE COMBINED EFFECTS COMPARE RESULT TO CHAPTER 15 4
ANALYZE EXCEPTIONS BRI BRI 8 GE BRI 6 GE BRI 6 GE BRI 8 GE GE GE
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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 Nuclear Boiler System Reactor Recirculation CRD Hydraulic Control System Feedwater Turbine Neutron Monitoring Process Radiation Monitor System Area Radiation Monitor System Reactor Water Cleanup Main Steam Condensate Turbine Control tube Oil P/0 TG, RFW Moisture Extraction Exhaust Steam Bleed (Extractiong Steam Heater Vents Feedwater Service Air Hydrogen Seal System Generator Cooling Air Removal Generator Hydrogen 8 C02 Purge Main Generator Excitation Off Gas Circulating Water Service Water RB Closed Cooling Water System TB Closed Cooling Water System Compressed Air Low'Conductivity Drains Primary Containment Instrumentation Heater Drains Miscellaneous Drains Sealing Steam 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-water capability).
The partial reduction of feedwater heating will gradually raise power to about 75$.'f the power exceeds feedwater capability, low level scram will occur and this event is bounded by the loss of all feedwater event already analysed in Chapter 815.
Due to the inaccessibility during power operation of the valve that.
isolates the air removal system the main condenser vacuum loss con-tinues until the main turbine trips and the reactor shuts down.
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If no operator action were
- taken, a
main turbine trip due to low condenser vacuum would occur more than a
hundred minutes after the power bus loss and at a reduced reactor power level.
This event is bounded by the turbine trip transient already analyzed in CESAR Chapter 15.
4.1.2 SM-2 {4.16KV)
The loss of this bus causes reactor feedwater pump 1B, condensate pump 1B, condensate booster pump 2B, and circulating water pump 1B to be inoperative.
There is also a partial loss of feedwater heating due to the extraction steam to the heaters partially bypassed to the main condenser, and a
possible slow loss of main condenser vacuum due to the loss of ejector/
condenser "B", leading to a
main turbine trip. If ejector/condenser "A" is in use at the time, or if it is manually started, the loss of ejector/condenser "B" would not result in a main turbine trip.
The worst case reduction in feedwater temperature has been calculated to be less than 334K.
Concurrent to reactor feedwater pump turbine lA trip and low reactor vessel water level, the reactor recirculation flow runback is initiated which is intended to reduce reactor power to about 68$ 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.-
APPENDIXA HANFORO CONTROL SYSTEM FAILUREANALYSIS PAGE Al DC 8US SYSTEM COh'IPONENT DESCRIPTION PRIMAAY EFFECT SECONDARY EFFECT COM8INED EFFECTS REACTOA F EEDWATER AEACTOR VESSEL HIGH WATER LEVEIBFPTAIP CHANNEL"C" CHANNEL"C"TRIPS ON LOSS OF POWER NONE, HIGHWATER LEVEL TRIP FUNCTION REOUIRES 2 OF 3 CHANNELSTO TRIP R FPT'S ANDMAINTUBSINE.
NONE;CHANNEL "C'NDICATES TRIPPED.
CONDENSEA COND.PCV.S VALVE COND PCV.S OPENS RFW FLOW REDUCED 6000 RPV WATER LEVEL3 SCRAM GPM. DECREASE IN CONDEN.
SER VACUUM 0
INVERTER REACTOR FEEDWATEA AELAYTT.X-IARFPT TRIP INTER.
LOCK TO REACTOR RECIRC SYSTEM BELAYTT.X.1ADEENERGI2ES PAOVIDINGRFPT TRIP SIGNAL TO SYSTEM NONE ~ 80TH RFPT TRIP AND APV LOWWATER LEVEL SIGNALS AEOUIRED FOR AECIBC BUNSACK ALTER.AC STEAM LEAKDETECTION TEMPEAATUAESWITCHES E3I.N604A8. NSISA8 REACTOR FEEOWATER RFW FCV 2A DISASLE HIGHTELIPERATUAE INPUT TO hlSIV TRIP RFW.FCV.2A FULLOPEN NONE RPV WATER LEVEL3 SCRAM
APPENOIX A HANFORD CONTROL SYSTEM FAILUREANALYSlS PAGE A2 DC 8US SYSTEM COMPONENT DESCRIPTION PRIMARY EFFECT SECONDARY EFFECT COMBINED EFFECTS O
REACTOR FEEDWATER REACTOR VESSEL HIGHWATER LEVEL, RFP TRIP CHANNEL"8" CHANNEL "8"TRIPS ON LOSS OF POWER.
NONE, HIGHWATER LEVEL TRIP FUNCTION REQUIRES 2 DF 3 CHANNELSTO TRIP RFPT'S ANDMAINTURBINE NONE, CHANNEL"8" INDI~
CATES TRIPPED Pt INVERTER ALTER. AC REACTOR FEEDWATER RFW-FCV.28 CIRCUIT RFW LOCKUP CIRCUIT RELAYTT X.IB, RFPT TRIP INTERLOCKTO REACTOR BECIRC. SYSTEM STEAM LEAKDETECTION TEMPERATURE SWITCHES E31.NB ISCD, N604CD FCV.28 OPENS STOP RFP'S IF CONTROL SIGNAL LOST RELAYTT-X.18 DEENEBGIZES PROVIDING RFPT TRIP SIGNAL TO REACTOR RECIRC. SYSTEM DISASLE HIGH TEMPERATURE INPUT TO MSIVTRIP
'SCRAM AT RP V LOW LEVEL3 SCRAM'T~LOW LEVEL 3 IF RFP CONTROL.
'SIGNALLOST f NONE.BOTH RFPT TRIP 5 RP LOWWATER LEVELSIGNALS REQUIRED FOR RECIRC. BU 8ACK NONE REACTOR SCRAM AT BPV LOW LEVEL3
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APPENOIX A HANFORO CONTROL SYSTEM FAILUREANALYSIS PAGE A3 DC BUS SYSTEM COMPONENT DESCRIPTION PRIMARY EFFECT SECONDARY EFFECT COMBINED EFFECTS AEACTOA FEEDWATEA AFP TURBltlE"A"SOLENOID TRIP CIRCUIT AFP TURBINE"A"TRIP RPV WATER LEVELLOWER RPV VIATERLEVELLOWER AND REACTOA AECIAC AUN-AND REACTOA RECIRC AUN-BACKTO 68% POWEA BACKTO 68% POWER CD cP 0
AIAAEMOVAL AEACTOAFEEDWATEA AR V-IVALVE AR V 2 VALVE RFP TURBINE"8"SOLEtIOID TRIP CIRCUIT LOSE CONDENSER VACUUM SLOWLY RFP TURBINE"8"TAIP MAINTURBINETRIP
&100 MINUTES MAINTURBINETRIP ON LOWCONDENSER VACUUM
~ 100 MINUTES RPV WATER LEVELLOWER RPV WATER LEVELLOWER AND REACTOR RECIRC AUN.
AND AEACTOR RECIRC RUN.
BACKTO 68% POWER BACKTO 68% POWER On O
MAINTURBINE CONTROL VOLTAGE AEGULATORCONTROL CIRCUIT MAINTURBINETRIP ON LARGE LOADCHANGE NO LOADFOLLOWING
APPENDIX A HANFORD CONTROL SYSTEM FAILUREANALYSIS PAGE A4 AC BUS SYSTEM
'I COMPONENT DESCRIPTION PRIMARY EFFECT SECONDARY EFFECT COMBINED EFFECTS REACTOR RECIRC RECIRC PUMP C0018 PUMP C0018 TRIP TO LOW SPEED REACTOR LOW POWER I.EVEL LOW POWER LEVEL REACTOR RECIRC RECIRC PUMP COOIA PUMP COOIATRIP TO LOW SPEED REACTOR LOW POWER LEVEL LOW POWER LEVEL O
O OFFGAS OFFGAS F05 IAVALVE F05'IB VALVE REGEN. BLOWERS LOSE CONDENSER VACUUM
- SLOWLY, LOSE CONDENSER VACUUM SLOWLY MAINTURBINETRIP
%100 MINUTES I MAINTURBINETRIP m 100 MINUTES MAINTURBINETRIP ON LOW CONDENSER VACUUM W 100 MINUTES MALNTURBINE TR(P ON LOW CONDENSER VACUUM
~ 100 MINUTES
APPENDIXA
)IANFORD CONTROL SYSTEM FAILUREANALYSIS PAGE AC BUS SYSTEM COMPONENT DESCRIPTION PRIMARY EFFECT SECONDARY EFFECT COMBINED EFFECTS REACTOA FEEDWATER CIRC WATER REACTOR AECIRC REACTOR FEEDWATER COND. PUMP IA COND. BOOSTER PUMP ZA CIRC. WATER PUMP IA LFMG SET SOOIA RFP TNG IA)TURNING GEAR)
BFP.MOP IA)MAINOILPUMP)
PUMPS INOPERATIVE PUMP INOPERATIVE LFMG S001A INOPERATIVE HDTTNG-RFP IATAIP STOP MOP ~ BFP IATRIP STOP AOP-AFP IATAIP DEGAEASE IN F LOWTO SUCTION OF RFP'S DECREASE IN COOLING FLOWTO MAINCONDENSEA NONE ATFULLPOWER SEE SECTION 4.0 RPV WATER LEVELI.OWER RPV WATER LEVELLOWER AND REACTOR RECIRC. AUN AND AEACTOA RECIRC. AUN.
BACKTO 88% POWER BACKTO 889f POWEA REACTOA FEEDWATER 991X.A ITURBINESPEED SW)
~ IATRIP RPV WATER LEVELLOWER BPV WATER LEVEL LOWER AND AEACTOR RECIRC. AUN.
AND REACTOR AECIRC. RUN.
BACKTO 881I POWER BACKTO BBKPOWER AIR RELIOVAL BLEED STEAM AR-SPV-2A AR SPV 28 AR.SPV. I I AB SPV.1.2 BS V.39A VALVE BS V.36A VALVE BS-V4)4 VALVE LOW CONDENSEA VACUUMDUE TO AIR EJECTOA "A"LOSS LOWCONDENSER VACUUMDUE TO AIR REMOVAL SYSTEM ISOLATED BS V-39A OPEN BS V.3SA OPEN BS V 84 OPEN MAINTUABINETRIP IN'03 IMINUTES MAINTURBINETRIP IN 103 MINUTES
,NOI)EI DECREASE FEEDWATER TEMPERATURE DECREASECONDENSER VACUUM MAINTURBINETRIP IN103 MINUTES O
HEATER DRAINS HEATER VENTS SEALING STEAM OFFGAS COMPRESSED AIR BS.V.4A VALVE BS.V 6A VALVE BS.V4IA VALVE HVV.20A VALVE SSV 12A VALVE OG.V 129A VALVE INSTR. AIACOMPAESSOR 'IA COMP. 18 ON BUS SM 3, COMP.
ICON BUSSM2 BS V-4ACLOSES BS V 6ACLOSE5 BS.V BACLOSES HV4/.29A OPENS SSV '12AOPENS OG-V. 129A OPENS COMPRESSOR IA INOPERATIVE DECREASE IN FEEDWATER TEMPERATURE AND CONDENSER VACUUM NONE NONE NONE (
NONE ~ COMPAESSORS 18 &,
NONE IC MEET REQUIREMENTS
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APPENDIX A IIANFORD CONTROL SYSTEM FAILUREANALYSIS PAGE AC BUS SYSTF M COMPONENT DESCRIPTION PRIMARY EFFECT SECONDARY EFFECT COMBINED EFFECTS REACTOR RECIRC SUBLOOP HYD.PWR UNIT D003A ISEE MCQD Al IF OTHER LOOP NOT AVAILABLE FCV 60A LOCKS UP NO LOAD FOLLOWING NO LOADFOLLOWING O
6 REACTOR FEEDWATER RFPT GOV.IA
'EEDWATER SYSTEM CONTROL CIRCUITRY REACTOR VESSEL HIGHWATER LEVEL, RFP TRIP CHANNEL "A" RFW PUMP IAINOPERATIVE F EEOWATER PUMPS AT LAST SPEED CHANNEL"A TRIPSONLOSSOF POWER RPV WATER LEVELLOWER AND REACTOR RECIRC. RUN-BACKTO 88% POWER NO LOADFOLLOWINGWITH REDUCED POWER NONE, HIGH LEVELTRIP FUNCTION REQUIRES 2 OF 3 CHANNELSTO TRIP RFPT'S ANDMAINTURBINE RPV WATER LEVELLOWER AND REACTOR RECIRC. RUN-BACKTO 88% POWER RPV HIGHWATER LEVELTRIP CHANNEL"A"INDICATES TRIPPED REACTOR RECIRC LOOP A 5 8 FLOW CONTROLLERS FLUXCONTROLLER FLUX ESTIMATOR LOCKUP OF FC VALVES60A 5 608 HIFTCONTROL TO MANUAL ODE NO LOAD FOLLOWING MANUALLOAD FOLLOWING
APPENDIX A IIANFORO CONTROL SYSTEM FAILUREANALYS)S PAGE Al AC BUS SYSTEM COMPONENT DESCRIPTION PRIMARY F.FFECT SECONDARY EFFECT COMBINED EFFECTS C4 CO REACTOR FEEDWATER CIRC. WATER COhlPBESSED AIR REACTOR FEEDWATEB REACTOR FEEOWATER MAINSTEAM BLEEDSTEAM HEATER VENTS AIR BEhlOVAL BLEED STEAM hlAINSTEAIh HEATER DRAINS COND. PUMP 18 COND. BOOSTER PUMP 28 CIRC. WATER PUMP 18 INSTR. AIR COMPRESSOR IC RFP-TNG.IB ITURNINGGEAR)
RFP.MOP. IB {MAINOILPUMP)
(TURBIkE SPEED SW) 99 TIX.I8 MS-V.1428 VALVE BS.V448 VALVE BS.V 458 VALVE BS.V 1168 VALVE HV.V.298 VALVE AR.SPV 2C,D BS.V.308 VALVE BS.V.68 VALVE MS.V.1338 VALVE MS V-1378 VALVE BS V 48, 58,68 VALVE PUMPS INOPERATIVE COMPRESSOR IC INOPERATIVE ICOMP. 1A ON BUS SM.I, COMP.
18 ON BUS SM.3)
HOT TNG RFP'IB TRIP STOP MOP ~ RFP 18 TRIP,.
STOP AOP - RFP 18 TRIP RFP 18 TRiP MS V.1428 OPENS BS.V 448 OPENS BS V 458 OPENS BS.V-1168 OPENS HV.V.298 OPENS LOSECONDENSER VACUUM SLOWLY BSV 398 OPENS BS V 69 OPENS MS V.1338 OPENS MS V 1378 OPENS BS V.48, 68, 68 CLOSES DECREASE IN FLOWTO SUCT)ON OF RFP'S DECREASE INCONDENSER COOLING FLOW NONE, COMPRESSOBS 'IAIh 18 MEET REQUIREMENTS BPV WATER LEVELLOWER AND REACTOR BECIBC. RUN.
BACKTO 68% POWER RPV WATER LEVELLOWER AND REACTOR RECIRC. RUN.
BACKTO 66ICPOWER NONE t
I NONE NONE MAINTURBINETRIP IN I83 MINUTES NONE DECREASE IN FEEDWATER TEMPERATURE ANDCON.
DENSER VACUUM SEE SECTION 4.0 BPV WATER LEVELLOWEA AND REACTOR RECIRC. RUN BACKTO 68% POWER RPV WATER LEVELLOWER AND REACTOR RECIBC. RUN.
BACKTO 68% POWER DECREASE IN FEEDWATER FLOW DECREASE IN FEEDWATER TEMPERATURE DECREASEINCONDENSER VACUUM MAINTURBINETRIP INI83i MINUTES
0 AI
APPENOIX A HANFORD CONTROL SYSTEM FAILUREANALYSIS PAGE AB AC BUS COMPONENT DESCRIPTION PRIMARY EFFECT SECONDARY EFFECT COMBINED EFFECTS OFF GAS SEALING STEAM OG V.1288 VALVE SS V.12$
SS V-18 OG V 1298 OPENS SS V.128 OPENS SS.V-IB OPENS
%c6 6
MAINSTEAM BLEED STEAM SEALING STEAM BLEEDSTEAM MS.V.1258 BS V.60 VALVE BS V 31 VALVE BS V.528 VALVE SS.V 30 VAI.VE BS.V-18 VALVE BSV BD BS.V BE BS-V-IC BS V.10D 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 MS V-1258 OPEN BS.V.SB OPENS BS-V.31 OPENS BS V 628 OPENS SS.V40OPENS BS.VALVESFAILCLOSED OROPEH IER VALVE TYPE
>NONE
'ECREASE IN FEEDWATER TEMPERATUREAND CONDENSER VACUUM DECR ASE IN FEEDWATER TEMPERATURE. DECREASE IN CONDENSER VACUUM "
APPENDIX A HANFOAD CONTROL SYSTEM FAILUREANALYSIS PAGE AS AC BUS SYSTEM COMPONENT DESCRIPTION PRIMARY EFFECT SECONDAfIY EFFECT COMBINED EFFECTS REACTOR FEEDWATEA COND. PUhIP IC PUMPS INOPERATIVE DECREASE IN FLOWTO SUCTION OF RFP'S SEE SECTION 4.0 CP Ch CIRC. WATEA REACTOA RECIAC REACTOR AECIAC COND. BOOSTER PUMP 2C CIRC, PUMP IC LFMG SET S0018 SUBLOOP HYD PWA UNITD0030 LFMG SET S0018 INOPERATIVE IF OTHER LOOP NOT AVAILABLE FCV408 LOCKS UP DECREASE IN CONDENSER COOLING F LOW NONE ATFULLPOWER NO LOAD FOLLOWING NO LOADFOLLOWING REACTOA REClflC SUBLOOP HYD PWA UNITD0038 IF OTHER LOOP NOT AVAILABLE FCV 608 LOCKS UP NO LOADFOLLOWING NO LOADFOLLOWING REACTOR FEEDWATEA COMPRESSED AIR AEACTOA FEEDWATER TURBINE GOVERNOR IRFPTNOV.IB)
INSTR. AIR COMPRESSOR IB ICOMP. IABUS ON SM.I, COMP.
IC BUS ON SM.2I REACTOR FEEDWATER TURBINE "8"TRIP COMPRESSOR 18 INOPERATIVE APV WATER LEVELLOWER AND REACTOR BEGIRD'UN BACKTO 66% POWER NONE, COMPAESSORS IA5 IC MEET REOUIAEMENTS RPV WATEA LEVELLOWER AND REACTOA RECIRC. RUN.
BACKTO 68II POWE A NONE
APPENDIX A HANFORI3 CONTROL SYSTEM FAILUREANALYSIS PAGE A10 AC BUS SYSTEM BLEED STEALI TURBINE SERVICE WATE COMPONENT DESCRIPTION BS V BA,4IB BS.V-10A BS V-108 BS-V.13A BS.DV.6A BS-DV4A BS.DV 2Al BS-DV 2A2 BS DV 3A1 BS-DNA BS.V.27 BS VMS BS V 628 BS.V 62A PLUS OTHER BS VALVES i MAINTURBINEOILTEMPERATURE CONTROL VALVE PRIMARY EFFECT BS V.BAJACLOSED IF FLOW LOST BS.V-IOACLOSED IF FLOW LOST BS V.108 CLOSED BS V.13A CLOSED BS.DV 5A OPEN BS-DV BAOPEN BS.DV.2A1 OPEN BS-DV 2A2 OPEN BS-0 VIIOPEN BS-DV4AOPEN BS.V.27 OPEN BS.VRB OPEN BS.V 628 OPEN BS.V 62A OPEN TSW.TCV4 CLOSED SECONDARY EFFECT DECREASE FEEDWATER TEMPERATURE AND ICONDENSER VACUUM.
MAINTURBINETRIP ON HIGH OILTEMPERATURE COMBINED EFFECTS MAINTURBINETRIP ON HIGHOILTEMPERATURE PECREASE IN FEEDWATER:
TEIIPEIIATIIIIE OECREASEIN CONOENSER)
ACUUM O
O IO OFFGAS REACTOR RECIRC.
OG-V.'I26A OG.V.I49A SUBLOOP HYD PWR UNITA D003A DECREASE IN CONDENSER VACUUM DECREASE INCONDENSER VACUUM IF OTHER SUBLOOP NOT AVAIL..NOLOAD FOLLOWING ABLE, FCV 60A LOCKS UP DECREASE IN CONDENSER VACUUM NO LOADFOLLOWING
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(k.lto VV)
TR-5(V)
SR-1-'ll 51 -Il (45OV)
MC 1C 5H.1t 4.1664) a a
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HC t5 PP 15-A MC 1F
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HC B
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FROH HC 1F P
IF A,
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HC SC MC.SB HC.85 MC BBA IhC.88m MC BC PP BA.C zPI'BA C A CIC.bC 8 Olh-b2 CIC-BR fO pp.le PP-6R RPS (8)
PP bAS PP. BA 8 z=
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r PP BCA.A BA'CYQRV 51.2 81-2 CI.Z C. SI. ZD BATTtRV BI E Cl-4 PP-SA MC-Bhk
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'5.AA 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 5V5'T5HB FAIl.uRBS ANAI (SIS BILS 'Tlh68 INHP-Z FIGURE I SggOF4
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(6.'I Wv)
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FAILlIRQ5 ANAL'(515 SII5 TR85 WHP 2 FIGURE ) SH 4 OF 4
APPENDIX B ELIMINATIONCRITERIA 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.
POMER BUS (AE OES IGNATION SYSTEH INCL. HPLI SYSTEH CONtROL AND INSTR UHENTAT ION LOADS OH BUS GENERAL EFFECTS LIHITATIONS ON THE SYS. '5 CAPABILITY TO PERFORH ITS PRINCIPAL FUtlCTION OUE TO LOSS OF BUS J ~ =
r! S
~
EFFECT EFFECTS OH SYSTEH OH OTHER SUBFUNCTIONS SYSTEHS EFFECTS (MITHIH C
10 Hlti OH 0
REACTOR MATER D
LEVEL, PRESSURE E
OR CPR SPECIFIC EFFECTS OF BUS LOSS TO SYSTEH EPL (a
r INPUTS REQUIRED FROH DTHER SYSTBt5 AND EFFECT ON LOSS OF INPUT
'F Power Supply 399 in 80-Gll (PP-BA-A)
(CKT-33)
RFM (Reactor Feedwater 822)
Loop 04-A51 RFM-PT"18 P 1-18 SRU-2 A0102 Hone Indicator on BO-A Hone 5 process computer indicate min.
pressure gp D
Hone RFM-P-18 discharge
~
'ressuro indication lost in control room.
Loop 04-A53 RFM-FT-28 SQRT-28 SRU-8 FIC-28 E/P-28 RFM-FCV"28 will go full open, by passing feedwater to wain condenser feedpumps will increase to 115K flow to try to main-tain feedwater. flow to vessel Low flow indi-cation and signal to RFM-FIC-28 on 80-A causing sin flow air to RFM-FCU-28; (fully opening valve)
H620-504.2 Reactor Vessel A
level will drop slowly, opera-tor aust take action to lower reactor power o prevent scram t Level 3 RPV level'ill drop High condenser level annunciated in contr'ol room (Cond-LS-ZN).
Loop 04-T52 RFM-OPT"38 OPI-38 SRU-56 Hone Indicator on BD-T indicates ain.
diff. pressure Hone D
None hP across HP heater 8
indication lost.
~ 4tlO~ I01i01<<<<IOOit<<1<<0 APPEHDIZ C
~ 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 h]r/C06294-44"
<<"Code Classification for Effects on Reactor Parameters:
"A" - Immediate (cl minute) and Direct "8" - Immediate but Indirect "C" - Effect is Delayed
<<0" - No Fffect an Reactor Parameters
(<10 minutes) "
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