ML19350E510
| ML19350E510 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/19/1981 |
| From: | Enzinna R, Jeffery Lynch BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19350E505 | List: |
| References | |
| TASK-2.E.1.2, TASK-TM NUDOCS 8106230206 | |
| Download: ML19350E510 (103) | |
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I EMERGENCY FEEDWATER SYSTEM UPGRADE l- RELIABILITY ANALYSIS FOR THE CRYSTAL RIVER NUCLEAR GENERATING STATION UNIT No. 3 (Contract 582-7179) i i f 4 Plant Performance Engineering Babcock and Wilcox J-Nuclear Power Generation Division P.O. Box 1260 Lynchburg, Virginia 24505 Prepared By /dmst6 [ '/Mc - Prepared -
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-L CONTENTS ;j.
4 Page
1.0 INTRODUCTION
1
2.0 DESCRIPTION
OF ANALYSIS 2-1 i 2.1 Fault Tree Analysic 2-1 2.2' Human Reliability Analysis 2-2
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2.3 Failure Data 2-2 2-2 2.4 Mission Success Definition 2.5 Assumptions 2-4 3.0 RESULTS 3-1 3.l' Quantitative Results 3-1 3.2 Dominant contributors to System Unavailability 3-1 i 4.0 RECOMMENDATIONS AND CONCLUSIONS 4-1 References Appendix A - System Description Overview A-1 Appendix B - Fault Trees . B-1 Appendix C - Human Reliability Event Trees . C-1 ie I: ! i. 1 I
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l LIST OF TABLES: PAGE 3- 1. Unavailabilities for FPC Crystal River-3 EFWS J-3 LIST OF FIGURES A-1 EFIC System Organization A-9 A-2 EFW Initiation Lc,gic A-10 A-3 Typical (1x2) 2 Logic Format A-ll A-4 SGA Vector Logic A-12 A-5 OTSG Overfill Protection Logic A-13 A-6 EFW Control System A-14 A-7 Crystal River-3 EFWS P&ID A-15
( i; l.0L INTRODUCTION This report presents a ' summary of lthe analysis methods and results of a reliability study of the proposed Crystal River-3 Emergency Feedwater (EFW) and' Emergency Feedwater Initiation' and Control (EFIC) Systems. The 'bjectives of _-thi:. st#j were:
- 1. To perform a detailed analysis to assess the reliability of the
' proposed EFW and EFIC systems.
- 2. To identify dominant contributors to system unavailability.
3.- To identify outlying system weaknesses, if any, and make appropriate recommendations for incorporation into the design process in order to ensure comparatively superior system reliability. A brief description of the Crystal River EFW and EFIC systems appears in Appendix A. This report documents the analysis methods used including assumptions made and analytical tools employed. The results are presented in both quantitative and qualitative terms. The numerical results include probability of failure per demande of EFW initiate, EFW control, feed only good generator logic (F0GG), and EFW overfill protection. The qualitative results include identification of major failure contributors', discussion of their significance with respect to system reliability, and c.onclusions drawn from the results. 1-1 I
( a. 4 . J,, , 2.0 . DESCRIPTION OF' ANALYSIS Fault Tree analysis was the primary method used to evaluate the Crystal ; River-3 EFW and EFIC systems. Once the fault trees were constructed important = minimal cuts sets were found using the Fault Tree Analysis Program, FTAP (2). Failure data obtained from references (4) through (15) were then used to quantify these cut sets. Human error j probabilities used in the quantification were developed using the
! ~ methodology described in Section 2.2.
The results of this analysis include the dominant mtributors to system fail .re. These contributors were reviewed to assess their effect on i successful system operation as defined in Section 2.4. Successful system operation, fault tree analysis, human reliability analysis, and failure data are discussed in the following sections.
, s i 2.1 Fault Tree Analysis Fault Tree Analysis consistent with the methodology described in the Fault Tree Handbook, NUREG 0492 (1) was used to evaluate the reliability of the EFWS. The complete fault trees for the EFWS are included in Appendix B.
1 h The fault trees were constructed to a level of detail sufficient to identify all relevant common hardware in the system. This level of f analysis allowed identification of haroware which would, if failed, reduce designed redundancy or cause the failure of other hardware. There was no ettempt to account for commonalities imposed by external events { such as fires, floods, or earthquakes. In addition to mechancial failure of hardware, failure causes dse to human actions and test or maintenance l activities were included. As indicated by the results, the human
, contribution to system failure probability frequently dominants the a i hardware contribution.
.c The computer code FTAP (?) was used for quantification of the fault trees, ranking of basic event importance. and identification of major contributors to system failure. 2-1
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i* . 2.2 HumanReliabilityAnaks3 j The Human Reliability Analysis (HRA) was performed consistent with the methodology described in NUREG/CR-1273, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications (3). The basic human error rates are found in Chapter 20 of Reference 3. Numerous data bases for component (hardware) failure rates are available; p 4 however, few data bases exist for human error rauas. A technique developed at Sandia Laboratories (3) was used to quantify human error probabilities. Probability tree diagrams for the human task of interest were constructed and are presented in Appendix C. 2.3 Failure Data Point value estimates of the component failure data were obtained from references (4) through (15). The data obtained from the above references are industry wide, ie. generic. These data are not directly applicable for the plant specific probabilistic analysis and were, therefore, updated, or made plant specific. by using the component failure data obtained from Crystal River-3 experience.
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2.4 Mission Success Definition l In order to evaluate the impact of component failures on system reliability an explicit definition of mission success is required. Mission Success of EFW was divided into four parts; initiation, control, overfill protection, and F0GG.
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' For initiation and control, system unavailabilities were calculated for two cases; the first case represents strictly automatic initiation and g
control of the systems, and the second case allows operators to intervene to correct system failures within 20 minutes. Only automatic actuation j of overfill protection and F0GG was evaluated. l l l 2-2 I L
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- 1. Initiation!
Mission succcess is defined as attainment- of adequate flow from at' least one pump to at least one steam generator. Failure to initiate includes: fluid system failure, EFIC initiation failure, spurious-isolation of EFW by FOGG,- and _ spurious isolation of EFW by overfill protection.
- 2. Control Given successful initiation,. control system failure is defined as all four control valves failing. low or. any one control valve failing higher than the rate limiter of the parallel valve to the same steam generator can.. compensate for. For. the purpose of this analysis, control valve fails high is defined as a valve failure that eventually would lead to an overfill or overcooling condition if unchecked. Control failure does not include credit for the mitigating effect of the overfill protection circuit.
- 3. ,F0GG Successful operation of F0GG is defined as isolation of both EFW flow paths to the bad steam generator and allowance of flow through at least one' flow path to the good steam gener ator. Failure of F0GG in the cases where specifications . call for feeding of both steam generators (i.e., spurious isolation of EFW) is included in the initiate evaluation.
- 4. EFW Overfill Protection Successful operation is defined as isolation of a flowpath that has failed to meet the mission success criterion for control (in the high -
state) given above. i 2-3 l t i
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-f 2.5 ,Ayumptions.
Lin evaluating the EFWS reliability the following assumptions were madei
- 1. One code safety valve stuck -open was conservatively considered a failure of that steam generator to supply adequate steam to drive the
' turbine: driven pump.
- 2. For a loss of 'Offsite power (LOOP) transient or a non-LOOP transient it was assumed that all code safety valves on both steam generators vould open.
- 3. Degraded failures were not considered, that is, components were i
assumed to operate properly or-were treated as failed. For example, the failure rate for code safety valve fails to reseat does not distinguish between degraded failures and catastrophic failures. This is a conservatism since experience has shown that in most
.( instances where code safety valves do not reseat at the specified setpoint, they do rescat at a somewhat lower setpoint that may still
) be considered a success, and rarely do the code safety valves fail to reseat entirely.
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- 4. Mechanical failures to passive components (such as locked open manual i valves) were disregarded due to their extremely low failure rates.
- 5. EFW overfill protection bistables fail untripped on loss of cabinet
- power supply.
l 6. .All EFIC relays are solid state. [ 7. Steam generator level sensor / transmitter failures will be repaired as soon as the failure is discovered; other sensor / transmitter failures will be repaired at next plant shutdown. 2-4
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- 3.1 Qudntitative Results ISystem unavailabilities were calculated for EFW initiation, control, overfill ' protection, and F0GG. For initiation and ' control, system unavailabilities were calculated for two cases; the first case represents strictly automatic initiation and control; and the second case allows ..
operators to -intervene to correct system failures within 20 minutes. Only automatic . operation of ' overfill protection and F0GG was evaluated.
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y Point estimates of system unavailabilities are presented in Table 3-1 for each of the mission success conditions described in~Section 2.4. A
^ discussion of'the dominant failure contributors for each mission success condition follows.
3.2 Dominant Contributors to System Unavailability
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t 3.2.1 EFW-Initiation (fully automatic initiation) 1 Pump EFP-1 unavailability is dominated by loss of offsite power and subsequent diesel generator failure. I
- 2. Pump EFP-2 unavailability is dominated by mechanical failure of steam admission valve ASV-5 or pump mechanical failure.
- 3. A major contributor to unavailability of either or both trains is inadvertent failure to realign pumps after pump maintenance.
t f ' EFW Initiation (includes operator corrective action) l i 1. The major contributor to pump EFP-1 unavailability is loss of offsite power and subsequent diesel generator failure or pump mechanical failure. l
- 2. Pump EFP-2 unavailability is dominated by being out of service for maintenance or pump mechanical failure.
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- f C 3.2.2 EFIC Control r
For both the automatic and operator corrective action. cases the dominant contributor to-failure is miscalibration of steam generator level setpoints which results in'a control valve failing high.
-3.2.3 FOGG i 1. The most dominant contributor to F0GG f ailure is steam generator pressure transmitter failure (either fails high or fails low).
- 2. A major contributor to F0GG failure is miscalibration of the P< 600
.) _ psi .bistables ( a bistable set too . low will fail to respond to a bad steam generator).
- 3. Another important contributor for the initial condition when both steam generators are less than 600 psi and one is 150 psi greater
[ than the other, is miscalibration of the 150 psi differential pressure bistables. 3.2.4 EFW Overfill Protection i
- 1. Steam generator level transmitters used for overfill protection in
., cabinets A and B are the same as used for control and there is a possibility that these transmitters may be the initiating cause of the overfill condition. This condition does not apply to overfill protection circuits in cabinets C and D because these cabinets are not used for AFW control.
- 2. Another major contributor to overfill protection circuit failure is
- - bistable miscalibration.
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l TABLE 3-1 UNAVAILABILITIES FOR FPC CRYSTAL RIVER-3 EFWS Unavailability
. EFW Initiate includes failure to initiate EFW due to fluid system failure, spurious isolation
= by F0GG or overfill-protection, and EFIC
' initiation failure.
. Fully automatic initiation 9.9 x 10-5
. Includes operator corrective action within 2.9 x 10-5 20 minutes EFIC Control:
l Fully automatic control 8.0 x 10-3 Includes operator corrective action within 1.2 x 10-6
'20 minutes
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FOGG A > 600 psi B < 600 psi (feed only A) 3.8 A < 600 psi B > 600 psi (Feed only B) 3.8 xx10-10-l A < 600 psi B<600 psia 150 psi >B(feed 1.5 x 10-3 only A) A < 600 psi B < 600 psi B 150 psi > A (feed 1.5 x 10-3
' only B)
I EFW Overfill Protection 1.1 x 10-4 1
- ' 3-3 l
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, 4.0 RECOMMENDATIONS AND CONCLUSIONS From this analysis we can conclude that the proposed FPC Crystal
+ River-3 EFWS design has no outlying weaknesses. The proposed upgrade
' represents a considerable improvement over the pre-upgrade reliability reported in the B&W report " Auxiliary Feedwater Systems Reliability Analysis", BAW-1584 (17). This improvement is primarily
'due to fluid system reconfigurations and the addition of the EFIC system.
The proposed safety grade EFIC system provides an improved approach to initiation and control with additional reliability provided by the T overfill protection and F0GG feature.. Fluid system changes that represent significant reliability improvements include: self cooling of the turbine driven pump, automatic loading of pump EFP-1 onto the i diesel generator upon loss of offsite power, flow control valve recon-figuration, and installation of a condensate storage tank standpipe. Mechanical failure of steam admission valve ASV-5 was identified as a i dominant contributor to unavailability of pump EFP-2. The estimated t failure probability of ASV-5 is, however, slightly less than that of the pump itself but on the same order of magnitude. Therefore, we conclude that ASV-5 mechanical failure is not an outlying contributor to unavailability. { In our judgement, the reliability of the proposed design is adequate;
. however, it was estimated that addition of a valve in parallel with steam admission valve ASV-5 would provide approximately a 28% decrease in calculated unavailability.
The reliability study revealed that a major failure contributor for EFW control F0GG, and overfill protection was miscalibration of setpoints.
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.o The importance of human errors on system failure should not be under-estimated. Even though a system is as well designed as the EFIC system
' and adequate redundancy is built-in, human error represents a potential i'
common mode fa;sure. Since maintenance personnel or other technicians interface with the entire system, there is a potential for design redundancy to be reduced through human error. Human errors contribute to the EFW and EFIC system un:vailability in at least two ways. First, miscalibration of setpoints in FOGG, overfill
. protection, or control represents a dominant failure contributor in each of these functions. Second, a major contributor to EFW pump unavailability is leaving the pump valved out after preventive maintenance.
t Therefore, it is recommended that administrative proceduras for the mainten3nce and calibration of the EFW and EFIC systems be carefully prepared. These procedures should conform to accepted human factors engineering practices in order to minimize the error probability associated with their use. Additionally, administrative controls should be applied to ensure that these well-written procedures are effectively used by plant personnel during EFIC and EFW maintenance. A major contributor to failure of F0GG is pressure transmitter failure. It is expected that this would be the dominating hardware contributor l li because, due to location, the failed transmitters are not repaired until the next plant shutdown. The large mean time to repair contributes to the
; unavailability of this component which in turn causes it to dominate l
other hardware contributors. Although the transmitters are the major l hardware failure contributor for F0GG they are not limiting factors l to EFW unavailability. 4-2
Some-steam generator level transmitters used for control are also used for overfill protection in Cabinets A and B. There is a small probability that these transmitters may be the initiating cause of the control failure resulting in the overfill condition, and consequently reduceo redundancy of the EFW overfill protection circuits. However, this concern is not serious because of inherent compensating features. For one, analysis of the control system indicates that control failure is dominated by bistable
'miscalibration and not transmitter failure. Also, transmitter failure is most likely to occur sometime before the EFW demand; in this case the operator has the opportunity to put the control valve in the safe mode before a demand occurs by switching the valve control to closed bias. Third, Cabinets C and D overfill protection operates on the isolation valves and is independent of the control function in Cabinets A and B.
Overall, the probability of an EFW overfill is small. Since the overfill is caused by a control valve failing high and is mitigated by the EFW overfill protectt . * . ice, the probability of an overfill can be obtained by multiplying together the values in Table 3-1 for EFIC control failure and EFW overfill protection failure (the overfill protection probability in Table 3-1 already accounts for the comonality discussed in the preceding paragraph). The result is that for the automatic case i the probability of EFW overfill is 8.8 x 10-7 per EFW demand, and including operator corrective action the probability of EFW overfill is negligibly small. ,t
; In conclusion, this analysis shows that the proposed FPC Crystal River-3 EFW and EFIC systems are reliably designed.
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!l-REFERENCES l 1. N. H. Roberts, and D. F. Haasl,. Fault Tree Handbook, NUREG-0492, 1 --November 1978.
- 2. FTAP 2, Computer-Aided Fault Tree Analysis,' Babcock and Wilcox
. Document Number NPGD-TM-536, February 6,1980.
.3. A.~ D. Swain, H. E, Cuttmann, Handbook of Human Reliability Analvsis with Empbssis on Nuclear Power Plant Applications, NUREG/CR-1278,
'0;ist:r 1980.
j; '4. k. H. Sullivan and John P. Poloski, Data Summaries of Licensee Event Reports of Pumps at U.S. Commercial Nuclear Power Plants,
~
NUREG/CR-1205, EGG-EA-5044, January 1980.
- 5. Warren H. Hubble, and Charles Miller, Data Summaries of Licensee Event Reports of Valves at U.S. Commercial Nuclear Power Plants, NUREG/CR-1363, EGG-EA-5125. Volumes 1 through 3, June 1980. -
- 6. J. P. Poloski, and W. H. Sullivan, Data Summaries of Licensee Event Reports of Diesel Generators at U.S. Commercial Nuclear Power Plants, NUREG/CR-1362, EGG-EA-5092, March 1980 7.- Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendix III, WASH-1400, NUREG-75/014, October 1975.
- 8. uSurvey of Feed Pump Outages, EPRI FP-754, Project 641, Final Report April 1978.
- 9. IEEE Cuide to the Collection and Presentation of Electrical Electronic and Sensing Component Reliability Data for 9uclear-Power Generation Stations, IEEE Std. 500-1977.
- 10. J. W. Ptgram, Compilation of Failure Rates for Use in Quantitative Reliability Analyses Compiled from Publicly Available Sources, Babcock and Wilcox Company, Nuclear Power Generation Division, Lynchburg, Virginia, September 1975.
- 11. Nuclear Plant Keliability Data System, 1979 Annual Reports of Cumulative System and Component Reliability, NUREG/CR-1635, September 1980.
I 12. Military Standardization Handbook, Reliability Prediction of Electronic Equipment, MIL-HDBK-217C, Departner. of Defense, April 1979.
- 13. Report on Reliability Survey of Industrial Plants, Part 1: Reliability of Electrical Equipment, IEEE Committee Report, IEEE Transactions on
- Industry Applications P. 213, Vol. lA-10, No. 2, March / April 1974.
- 14. H. Hnatyshyn, Reliability Program Report for Engineered Safety Features Actuation System II, WNP Units 1 and 4, Automation Industries, Inc.,
Vitro Laboratories Division, Silver Spring, Maryland, May 1979.
- 15. Reactor Protection System, Topical Report BAW-10085P, Rev. 6, Babcock and
[ Wilcox, Nuclear Power Generation Division, April-1979. 4 7 i
t.
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' REFERENCES-(continued).
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.. 16. ' System Description Emergency Feedwater System for FPC Crystal River-3; 1121209-00, Babcock and Wilcox, Nuclear Power Generation Division,
, November 1980. ,
- 17. _W. W. Weaver, R. W. norman, R.'S. Enzinna, Auxiliary Feedwater Systems-
' Reliability Analyses for Plants with Babcock and Wilcox Reactors, b BAW-1584, Babcock and Wilcox, December, 1979. .t. .f t' .k '
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7 APPENDIX A SYSTDi DESCRIFIION OVERVIEW
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-4 A.0 SYSTEM DESCRIPTION OVERVIEW The-following system description and accompanying diagrams represent a condensed version of the more complete description found in the " System Description of the Emergency Feedwater System for FPCo Crystal: River-3" (16).
A.1 Summary Description The emergency feedwater (EFW) system consists of two interconnected trains, capable of supplying EFW to either or both steam generators (SGs) from either water source under automatic or manual initiation and control. A piping and' instrumentation' diagram is included as Figure A-7 of this report. The EFW pumps take suction from either the condensate storage tank or from the ccm;ensor hotwell and discharge to the SGs. In the flow path between th_ FW pumps and the SGs there are isolation valves, check valves, control valves, flow instrumentation, and pressure instrumentation to control the flow of EFW to the SGs. The fluid system design is described in Section A.2. The instrumentation system design is described in Section A.4. A.2 Fluid System Design A.2.1 Suction The primary water source for the EFW trains is the Seismic Category I condensate storage tank. Water is supolied from this tank through a common line with a locked open manual valve to separate lines containing normally open motor operated valves. Although there are other connections to the condensate storage tank they draw through an internal' stand-pipe which asssures that a minimum of 150,000 gallons is held in reserve exclusively for the EFW system. This reserve is further verified by redundant, safety grade level indication in the control room. In addition, safety grade low level alarms are provided to alert the operator to perform a suction transfer.
&-D
The r.ain 'condensor hotwell is the primary alternative. suction source available for- the EFW system. Separate lines with normally closed
.DC; powered valves draw suction through a common line with a locked open manual valve ar.d a check. valve. The DC-powered valves are interlocked such that they.can be opened only if at least one of two DC powered vacuum. breaker valves is open.
A.2.2 Pumps and Discharge Cross-Connect EFW Train B pump (FWP-2) is a full capacity turbine driven p: imp, EFW Train A pump (EFP-1) is a full capacity motor-driven pump. The Train A and B pumps discharge through check valves and motor operated stop-check valves into cross-connected discharge lines. The separate cross-connects contain normally open motor operated valves. These cross-connects permit either pump to feed either or both steam generators. A.2.3 Emergency Feedwater Flow Control Valves The flow of EFW to each steam generator is controlled by normally closed pneumatically operated control valves in parallel paths. These control valves s e ' designed to fail open on loss of air. Initiation and control instrumentation for these valves is described in Section A.4 of this report. A.2.4 Steam Generator EFW Isolation Valves Each steam generator can be isolataed from EFW flow by normcily-open motor-operated valves. These valves are located in the parallel lines upstream of the EFW control valves. Initiation and control instrumentation.for these* valves is described in Section A.4 of this report. A-2
A.2.5 Recirculation siid Test' Lines Recirculation lines are connected to the discharge piping of the EFW
.p;mps.- Recirculation'for pump protection is accomplished with normally open flow paths to the condensate storage tank consisting of small lines with check' valves.and locked-open manual ~ valves.
EFW pumps can be operability test'ed using the normal- recirculation flow paths to confirm the pump and pump drive capability to operate and produce the required discharge pressure. No change to the normal EFW system valvt lineup is required to perform this testing. A.2.6 Steam Supply for the EFWS Turbine Steam supply for the EFW turbine pump is obtained from both steam generators through lines containing check valves and normally-open-DC
.notor operated stop-check valves. The check valve and motor operated valve provide redundant isolation capability to preclude blowing down the good steam generator in the event of steam line or feed line reak.
Downstream of these valves the lines join to_ form a common supply to the pump turbine. Upstream of the turbine is a normally closed DC motor operated valve. A description of the controls for this valve is contained in Section A.4 A.3 Supporting Systems The EFW turbine driven pump and turbine are self-contained entities without dependencies on secondary support systems. The bearings on the turbine and pump are lubricated by slinging oil from reservoirs near the bearings. Lube oil cooling is accomplished by heat transfer to the pumped fluid. The EFW motor driven pump and pump motor bearings are lubricated by slinging oil from reservoirs near the bearings. Lube oil cooling is provided by the nuclear service closed cycle cooling system. Two of the five cLoling water pump: receive diesel-backed power, i A-3
+
g A.3.1 Power The two EFW trains are powered from diverse power sources. EFW pump EFP-1 is turbine driven and EFW pump EFP-1 is AC power motor driven with back-up power from the diesel generator. Valves required to operate the
'EFW system are also on AC power with back-up pover from the diesel generator.
To ensure EFW flow in the event of loss of all AC power, the turbine driven pump train derives its power from the steam generators for the pump and from a battery-backed DC buss for its steam supply valves. A.4 Inst _rumentation Description The emergency feedwater initiation and control system (EFIC) is an instrumentation system. designed to provide the following:
- 1. Initiation of emergency feedwater
- 2. Control of EFW at appropriate setpoints
- 3. Level rate cohtrol when required to minimize overcooling
- 4. Isolation of the main steam and main feedwater lines of a depressurized steam generator
- 5. The selection of the appropriate steam generator (s) under conditions of steam line break or main feedwater or emergency feedwater line br.eak downstream of the last check valve
- 6. Termination of main feedwater to a steam generator on approach to overfill conditions
- 7. Termination of EFW to a steam generator on approach to overfill conditions, and
- 8. Control of atmospheric dump valve tu a predetermined setpoint.
The EFIC-See Figure A-1-consists of four channels (A,B,C,D). Each of the L four channels are provided with input, initiate, and sector logics. Channels A and B also contain trip logics and control logics. Each channel moniturs inputs by means of the input logic, ascertains whether action should be initiated by means of the initiate logic and determines wnich SGs should be fed by means of the vector logic. A-4
' Channels A and B monitor initiate signals;from each of the four initiate
. logics-by means of the trip logics to transmit trip signals when required. Channels A and B also exercise control of emergency feedwater flow to the SG by means of control logics to maintain SG 1evel at
. prescribed values once EFW has been initiated. In addition, Channels A and B also monitor SG A and B overfill signals originating in the Channel A,'B, C and D initiation 10gics. By means of trip logics, Channels A and B terminate main feedwater to a steam generator that is approaching overfill.
A.4.1 Input Logic The input logic is located in each of the channels.. The i.1put logic:
- 1. Recevies the input signals
- 2. Provides input buffering as required-
- 3. Compares analog signals to appropriate setpoints to develop digital signals based on analog values
- 4. Provides for the injection of test stiFJII
- 5. Provides buffered Class lE signals and isolated non-lE signals A.4.2 Initiate Logic The initiate logic, depicted in Figure A-2 is located in each channel.
The initiate logic derives its inputs from the input logic and provides signals which result in the issuance of trip signals via the trip logics in Channels A and B. The initiate logic issues a call for EFW trip (to the trip logic) when:
- 1. All for RC pumps are tripped
- 2. Both main feedwater pumps are tripped
- 3. The level of either steam generator is low
- 4. Either steam generator pressure is low
- 5. Either of two ar,ticipatory trips (trips not yet assigned) are present A-5
.c Other functions of the initiate logic are:
~
- 1. Issue'a call for SG A main feedwater and' main steam line isolation-when SG A pressure is low 2.- . Issue'a call for SG B main feedwater and main steam line isolation when SG B pressure is low
- 3. . Signal' approach to SG A overfill when SG A level exceeds a high level setpoint
- 4. Signal approach to SG B overfill when SG B 1evel exceeds a high level setpoint
~5. Provide for manually initiated individual shutdown bypassing of RC pumps, main feedwater pumps, and SG pressure initiation of EFW as a function of permissive conditions. The bypass (es) are automatically removed when the permissive condition terminates 6., Provide for maintenance bypassing of an EFIC initiate logic A.4.3 Trip Logic The trip logic of the EFIC employs a 1-out-of-2 taken twice format. A typical example of this logic is depicted in Figure A-3. The trip logic is provided with five 1-out-of-2 taken twice trip networks.
These networks monitor the appropriate outputs of the initiate logics in each of the channels and output signals for tripping:
- 1. Emergency Feedwater
- 2. SG A main steam line isolation
- 3. SG B main steam line isolation
- 4. SG A main feedwater isolation
- 5. SG B main feedwater isolation For each trip function, the trip logic is provided with two manual trip switches. This affords the operator with a means of manually tripping a selected function by depressing both switches. The use of two trip
' switches allows for testing the trip switches and also reduce the possibility of accidential manual initiation.
A-6
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. Trip signals are transmitted out of the EFIC by-activating a' relay thereby gating-power onto trip busses. In this manner, the Er'IC provides power to_ energize the control relays whose contacts form the AND gates-in the controllers.
A.4.4 Vector Logic - The. vector logic-Figure A appears in each of.the EFIC channels Figure A-1. The vector logic monitors * ,
- 1. SG pressure signals
- 2. SG (A and B) overfili signals (see Figure A-5 for overfill protection logic)
- 3. EFW trip signals _(vector enable) originating in Channel A and B trip logics.
The vector logic develops signals for open/close control of steam generator A and B emergency feedwater valves.
- - When enabled and with no overfill signals present, the valve open/close commands are determined by the relative values of steam generator pressures as follows
l-l SG A Valve SG B Valve Pressure Status Comand Comand SG A & B > Setpoint Open Open SG A > Setpoint & SG B < Setpoint Open Close l SG A < Setpoint & 53 8 > Setpoint Close Open SG A < Setpoint & SG B < Setpoint and SG A & B within 150 Open Open SG A 150 psi > SG B Open Close SG B 150 psi > SG A Close Open l A-7
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; A;4.5 Control Logic I
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The control logic is depicted in Figure'A-6. For each SG (A'and B) there are.two-controls which are selectable by trdnsfers within the EFIC j system.' The two; foot level setpoint control is automatically selected when an EFW trip occurs with one or more re' actor coolant pumps operating.1
~ A level rate' control with a twenty-foot setpoint is selected when an EFW trip ocurs with no reactor coolant pumps operating.
e 1 i-4 k A-8 t-I
'Mf- h t- tpt . *- -w - - , - -.-s--+ em / e- & *--
FIGURE A-1 EFIC System Organization CHANNEL A y INITIATE EFW COINCI DENCE
= ISOLATE SGA MAIN STEAM LOGIC = ISOLATE SGB MAIN STEAM
' = ISOLATE SGA MFW RC PUMP TRIP & . = lcSLATE SGB MFW INPUT
. MFU PUMP TRIP = SG A CONTROL e LOGIC CONTROL
- B N OL S G. PRESSURE .
NEUTRON FLUX : VECTOR PEN 1 SGA EFW VALVES i CWSEj LOGIC h OPEN SGB EFW VALVES y CLOSE CHANNEL B INITI ATE EFW COINCIDENCE
& ISOLATE SGA MAIN STEAM LOGIC & ISOLATE SG'B MAIN STEAM RC PUMP TRIP - > ISOLATE SGA MFW
. INPUT - -
- ISOLATE SG B HFW MFW PUMP TRi? _
S.G. LEVEL
- LOGIC CONTROL 5 SGA CONTROL S.G. PRESSURE LOGIC & SGB CONTROL MEUTRON FLUX VECTOR OPB l* SGA EFW VALVES CLOSE LOGIC
& OPEN 1
& CLOSEJ SGB EFW VALVES CHANNEL C
, RC PUMP TRIP VECTOR = OPEN MFW PUMP TRIP ,& LOGIC & CLOSE
, INPUT S.G. LEVEL LOGIC - OPD e S.G. PRESSURE O ~
SGB EFW VALVES MEUTROM FLUX ,
& CLOSE CHANNEL D
{ RC PUMP TRIP - OPEN ' MFW PUMP TRIP SGA EFW VALVES
- VECTOR S.G. LEVEL ;" INPUT LOGIC ,, S.G. PRESSURE *7 LOGIC = OPEN NEUTRON FLUX GB E W VAMES CLOSE A-9
FIGURE A-2 EFW Initiation Logic NC PUMP Al TRIP ? RC PUMP t.2 TRIP ? RC PUMP B1 TRIP ? , '! RC PUMP B2 TRIP ?
# < % 55 F.P. ?
^
T SHUTDOWN BYPASS M oC O RESET ~ SHUTOOWN BYPASS i r i NFW PUMP A TRIP ? MFW PUMP B TRIP ? , J e < 2 55 F.P. 7 SHUTDOWN BYPASS g RESET _ SHUT 00VN BYPASS ge
.L m INITIATE EFW v
SG A LEVEL < % 1' ? SG B LEVEL < 0 l'? SHUTDOWN BYPASS r h
- OC C-SHUTDOWN BYPASS RESET
[C 9
,,,a
, SG A & SG B PRESS < % 750 7 l l SG A PRESS < 0 600 ? m SG B PRESS < 800 7 v A-10
- . . , .. . - .- . .- . . - i; R
] FIGURE A-3 Typical l1x2}x 2 Logic Format 1 IWPUT iND I l 1 CONTROLLER M i COINCIDENCE
- SENSOR INITI ATE LOGICS l logic l CONTROLLER I TYP. OF CONTROLLER I g l i i l I i l l l l
FTEST l , gg l l 9 Cj J, l I c? Ias j I I l5 F TEST 1 1 -FTRIP l l y C'i l I J B ! l l l p+s)(c.o) l l l l - TRIP THI S l l l I DW CE l? FT eST 3i I O, _T l C! l J C l - n TEST CONFIRM i
~
l TO EFIC ! l4 l FTEST l _9 ^ RESET
.C&D l j
O'i J o l _ l l l 1 l l l l -l TRIP l 1 I i l A-ll
FIGURE A-4 . . SGA Vector Logic SG A OVERFILL PER THIS CHANNEL ? SG A PRESS < 600? .C 1 SG B PRESS < 600? , , g ,J N , m 4 OPEN s
- V SGA EFW
> VALVES
, 4 CLOSE SG A PRESS 150 PSI > SG B PRESE?
b d ' ' SG B PRESS 150 PSI > SG A PRESS L_,, I, i i EFW TRIP ISSUED BY CHANNEL A? EFW TRIP ISSUED BY CHANNEL B? A-12
FlGURE A-5 OTSG Overfill Protection Logic . SG A LEVEL > % 31' ? SGA APPROACH SG A LEVEL > 28' ? TO OVERFILL TO EFW VECTOR LOGIC _ J ISOLATE MFW TO SGB SG B LEVEL > ~ 31'? SGB APPROACH SG B LEVEL > ~ 28'? TO OVERFILL TO EFW VECTOR LOGIC v I A-13 I
FIGURE A-6 EFW Control System T NEDN - SG A LEVEL 1 : A - - _ T,2 K+) ' 1 ~ CONTROL VALVE J L JL J L
~;:: 2 '
SP SG A LEVEL 2 : K4 HAND
;; = A J L I f
~ 12" RATE LIMITED # LOW T , 20' SELECTOR 4 ~SP BIAS FOLLOWER l J L .
~ 4"/ MIN ON INCREASING LEVEL
~ 200"/ MIN ON DECREASING LEVEL ::: 30' SP A-14
,r_
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. . . i . . .
P00R ORGINAL
~'
f . . < l
$ i
( k n i APPENDIX B FAULT TREES I. f a ( i
9 9 EW INITIATION FAULT TREE I 1 l i l l B-1
--w@4- 6pe=ge c.e >9.aw e - m= g + p e W._sM m Sam m e as *4 --e e er hh > }:
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=
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6 e 4 S tan esm W w W *= ta e m
.a em 2 E -
mumme u eum e* ts= W8 ** b4 a e S I en g , e i p
.- - . .... . .. . . - _ . - - - . ~ . . . . - . .
A DISCHARGE PATH AN9 A1/B2
' FAILS H9 rm
[ A0V CHECK VALV FWV-Yl/FWV-X1 ['EFV 33/EFV-14 EFV-16/EFV-15 NECHANICALLY l FWV-43/FWV 44 . ROW FLOW FAILS FAILS TO BLOCKAGE BLOCKAGE CLOSED OPEN FWVYlZTO FWV43ZFB i EFV33ZFC EFV16ZFB - FWVXIZTO FWV44ZFB EFV14ZFC EFVl5ZFB i 9
- -, . ._,._.,i
- I ,
.t. i I
- i 10 OlSCHARGE II PATH A2/B1 I
FAILS N10 Nil rm 1 [EFV 32/EFV 11 CHECK FV-18/EFV-17 i FLOW I llECHAN CAL FAILURE FW-161/FW-162 pyy FLOW BLOMAGE j y 44 FAILS ( BLOCKAGE FAILS TO OPEN CLOSED \ -. FWV43ZFB EFVIBZFB g .FWV161TO EFV32ZFC FWV44ZFB EFV17ZFB FWV162TO EFV11ZFC 9 i
... ,_..... ,~c., . _ _ _ , , , , . _ _ . .
1 TURBINE PUNP EFP 2 FAILS N12 rm I I TURBINE PUNP TURBINE STEAN PUNP EFP-2 TRAIN 8
- . EFP-2 VALVES Fall SUCTION IN P.N FAILURE l URE
- ACTUAT ON RCulT)
A A n A N13 CHANNEL I B l AFW FAILURES ACTUATION FAILS B-5 AFWTRPOP l i l i
- . . . - , . . . . - , . , - , , . , - . . . + - - . . -
... .. W , is . L .,
~
..h.;._......~. .
e O 6 a.J b d@ E O O h
- w >= N N
E E E m e D N 2l) lN E e k b W =.J b h w w e w 8 % had 9' e. J w e Q U N w e
*g - ==
E 4 3 w EA ,,, 4
= m J w 4 O $#3 U O te.
== w >= >= * %
2 E E N aC > d 3 N
- N l' u ,M = e + *- 8 U M b M m.
ha
>Z had 53 t#9 tad e D t#> M W 4" 4's. ,eg N E **= W
% R W .D a. _Q.JQ J~ED b had EU >
Q D e-- g ha. M 4/3 W
*W
==
3 tad b N E
- D W IlD w A .mmme. 4 LS E w ==J bad e#9 med a - >= =
=53 ha.
O D r3 E
>== 2 %
agn
. Eh
> tad h N
he. LS hf had W 4 N w U Q RK l 4!"% I had had au > "
*= Z > ta. Q h g
4 Q l > = MM e had W W sumussa *== 4.8 G --
> M 4#9 M EE-tfl
==
Z O N
>= e L
gw g4 M had had M meer - had D g
>= tad M >
M'" W . RE- Eg y E=- 4 ha.d A m.-- e- " M g 4 E B-6 - M
' ~ '
4 NO STEAN ~ FRON CONNON STEAN PATH N18 I ! I i STEAN ADNISSION OPERATOR VALVE ASV-5 LOCAL FAILS TO OPEN CORRECTIVE ACTION FAILS NI9 - w ASY5ZZOP p DC POWER ASV.5 TO VALVE FAILS f NECHANICALLY t t FAILS TO OPEN ASV5ZZTO
o . i
?
w I-t- 1 A2I NO STEAN FRON SGA/SGB [I f i N20 a, N21 P 1 eN N j ANY OF OV HECKVALVE'g
- 4 CODE SAFETY , STOP CHECK ,
NSV-186/ 1 VALVES ' NSV-55/MSV-56 NSV-187 { l l
\
Fall TO RESEAT j SPURIOUS CLOSURE FLOW BLOCKAGE ~
\ /
i tsVSG$RS NSV55ZSC MSVl86FB CSVSCBRS MSV56ZSC MSVl87FB
- l j
i
b 4 O " J - 5 k ( EDTOR PUNP EIP-1 FAILS [/]N22 i T I I L POWER NOV PUNP EFP-I PUNP FAILURE STOP CHECK EFP-1 ( 1 EFV-7
'"' ' CHANICAL TO EFP-1 i g FAILURE TO a CL 5 EFV7ZZSC EFPlZZFR CHECK NO PUNP PUNP TRAIN A ,
Y VALVE EFV-6 EFP-1
' EFP-1 SUCTION FLOW , NECHANICAL ACTUATION FAILURE FAILURE BLOCKAGE TO START EFP!ZZFS EFV6ZZFB l b
r.. . - --' - . --. n. .. . . ; 2 - . . . 1 6 1 e POWER 2 FAILURE TO I EFP 1 N23 i s ' DIESEL
/ GENERATOR ,
LOOP DG3A FAILS TO START OR
, IN P.N.
LOOPZZZZ B-10 DG3AZZFS i 1
...i
\
c'" _n.
- . .,......e..., , . \>
I i e
/\ NO PUNP
- /N24\ EFP-1
[ ACTUATION (PER CHANNEL A) N24 ., g - l CHANNEL A PERATOR FAILURES Apg ACTUATION FAILS B-11 AFWTRPOP e
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N NS . asetausp% ~ M6 , N ' 6 N, N. .
'. T l TRIP CABINET
.A I FAILURES 3',
) N39 i
e m t' I 1 120V.AC BOTH CABINET j ' RELAYS Fall BUSS FAILS [ TO CLOSE : C; N40 . r % F DC POWER CABINET A INVERTER FAILS, FAILS . i BREAMER FAILS , OPEN 129V AC 120V AC ABIAET A TCAMBRF0 TCAINZAM
~
. / 1,
! ' #/ I'
ff171-~~~~,~ ,T;&: , ?L's.; fua &: , a ,:in :- --+.u s : , '.i . . ,, .'...~:-
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.- = e
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. ? ' ~ .^ ! !!; itij 'j. ' ~
v T l E l N A E I F S A B O
- I A S L C Y C A
H L O T E T
- O R B
5 4 N T E N S I E B R R
-- A B U l E C L W I O P A P I F R C
- T D 6
4 C B N R A M C L S TET L V E S T A Z s R 0 N N I A I E A 2 I A V F 1 B I l V F B
-- N A C 0 I C T 2 S 1 S U
B - N
. T R E C E E P A 0 r N I
KA O V F R . B E S O B A R L M C B I A 1 B
- C B F T
~
- ' . i
~~ -
- . .~ : - .-
f P 4
.?:
4g BOTH CABINET l, RELAYS FAIL TO CLOSE N47 I N48 ' I l i , I -l e , RELAY A RELAY B FAILS TO CLOSE FAILS TO CLOSE s a f 4 N49 N51 4" M50 N52 e% r% i oaREARYASH REBRYASH BAARYAFD BABRYAF0 RYARYATC RYBRYATC cy REARYBSH REBRYBSH e qy BAARYBF0 BABRYBF0 RYARYSTC RYBRYPTC i e a
. n....-----.~.- -. . ~ . . . . -
l l 4 i, 1 t r i N 4 N5 DC POWER FAILS , i' H53 N54 4 i DIESEL 4 i BATTERY GENERATOR 3A/3B OG3A/DG3B i
, LOOP I
$. FA!LS Faits l
i BA3AZZAN LOOPZZZZ DG3AZZFS BA3BZZAN LOOPZZZZ B-19 DG3BZZFS . i j 1 i { j k i.
~ ;.
- f. .s.- . , . . . . . , , . . _ , .
l I
- I l
l' E W CONTROL FAULT TREE I i l I 'I i' 'I
x . ,. ;a ~ ;u .,;,. . .; a. .,n- _ . . a ,a. a.. .~ , . - I-y . 8 1 L C0tfR0t . , FAILURES i Cl 1 i 1 l 1 OUT OF 4 gUi 0F 2 4 OUT Of 4 h VALVE 1 Fall CAslNET POWER gag'Vt3 Fall ul&M $ PL IL L0s (2CV's'(Fall MIGM) [ C2 [D C3 () C4 I I PERATDA ERATOR fills TO FAIL 1 TO 13CLATE OPEN 110 LATE OPE PATM PAINS v&LVE FAIL $ CAslNET P08E8 OPFAILPi
- F LFH CPEN IUPPLV FAMURE CS CE ,
r% r% i i i i 1 OUT OF 4 2 OUT OF 4 CAllNET A CA8 thel B CONTROL VALVE CONTROL VALVE 1 Posti FAILURE - POWER FAILURE FAILutE! Fall MIGH IFWV.yjAND (F9V.1$1 AND Ffy Il FAIL Ftv.162 NISM) Fall MIGM) -i 00 0 01 0 .i FBV 41 FAILS LOW F8V Isl FAILS 78V 182 FAIL $ FtV Il%L$ 'l B-21 ss a Cn. A to, ss Cn a t0e ss A Cn : LO' ssA Cn A 4 0 1 P00RBRMA l I
p 4 V, i
'k.
1 DUT OF 4 CONTROL VALVES F'All HIGH g i 1/4 ' C201 I i FWV-YI FWV-161 FWV-162 FWV-X1 FAILS HIGH FAILS HIGH FAILS HIGH FAILS HIGH l SG B CH A SG B CH B - SG A CH B SG A CH A 03 10 B-22 10 0 i j r l' 1 i'
<- = . .
l
.- i
- l L -!
w D
)'
>= O i J >= d .J W
='=
h I 1 w WM M J l N
===
,J E ** H 4 4 >
g b >= d 4
~
W D
>= O
- J >* ed a W ~
W
" " " " * > W G4 l
. m W M .d
-ggy
==8 E =.
4 4 > E b >
# med i
E w D e- O a N em= .md b O .g .m. e== N w S E4 CD U > M -J == w 3 == > a 4 4 >= O E u me
- 1. >= >"
M wW DD dOED ga wO p
- s= Z >= ==4 g 6 C.S ** N a a - (.D
>CZWZ ll'l3 E M-p.= . E -J g g .J. 4 == E .-d b
W I N uO4E4 w >= m ' >" D O N
*=
J e-= A e w - m M l .- > W to =J J J E 4 6" E 6
>=
m l t
\
I i & E w 67) E O 6 mid
' gllll3 W E4 --
>= W i
B-23 I = ** = = q
. _, _ g W
A 4 O w D
- >" aus O E
e i I f I l' f l
; 2 , .;*....-, .. ,,../ - ._,. ... , , _ , . _
t'{ :, . *'
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.I ;1 I
101 102 CABINETA/BPOWER FAILURE
- )
\ ,,
I __ , _._ , A,, P_ .A _ AC ,A,L, { TRANSFORMER I FAILS OPEN f ,, TCAMBRF0 TCASPTAM B-24 TCBINZAM TCOMBAFO 18
,i TC85PTAM 3 t
. -1 p .
I L 6 i I
, _ .,_ A. . ~u e. .;
[ . . . ., e e F8V Yl/F9V 161/ F9 162/F9V Il FAIL 5 HIGN I i Cl , ClO Cll i Cl2 i j VALVE E/P L CONVERTER STICR1 HIGH ' . AFTER VALVE HA$ L8 E FA b 'k I V ALREADY HIGH OPENEO . V412EIFL v'41ZV3FH
'U$
Vl61V5FH vl62V5FH VrilV3FH t c -l l i FAlts CONTROL FAILS CONTROL FAILS CONTROL HIGH. AT HIGH. ALL RC HIGH, AT LEAST LEAST ONE RC PueP3 TRIPPED ONE RC 1 Puer CN (LOOP > PusP DN ~ 4 8 C22 CIS I Cig C23 i C16 C20 C24 l l l
\.
( i E R E ONE RC PUNPS ONE RC L PuuP DN TRIPPED pump (TUDF) (LOOP) RUNNING ($HOULO (SHOULO (tgree CONTROL
" l 20 LOOP) .I TO 2 FT) .
py) LFWNLOOP ERR 0NEOUS LOOPZZZZ LFtNLOOP CONTROL TO CONTROL TO B 0P CONTROL TO LFONLOOP 20 FT FAILS 0 Z 2 Fi FAILS p ,'y 20 FT. LFsNLDOP HlGH LOOPZZZZ HIGH B-25 i . P00R BRfilNAL 1
r- . ~ . . . . . , . .. ___-.- ~ . .. . . . .
~
I t 1 I14 15 ' 16 I ERR 0NEOUS CONTROL TO 20 FT (FWV-V1/ FWV-161/FWV-162/
, FWV-XI)
\ .
C25 C26 f . C27 g3 C2B RCP RELAYS Fall , j RCP RELAY (INDICATE ALL RC - FAILS, SWITCHE PUNPS TRIPPED) - SPURIOUSLY TO , i INDICATE ALL R UNPS TRIPPE C29 I TIRY41SS C30 I TIRY61SS C31 TIRY62SS C32 TIRYX1SS I i 1 RELAY l REl.AY 1 i RELAY l I RCP141l T RCP241lT RCP2411T RCP41tT RCP161lT RCP2611T RCP361lT ,h RCP1621T RCP611T RCP2621i B-26 RCP3621T RCP621T ! RCPlXIIT RCP2X11T RCP3X11T RCPX11T ( l e
.._.s..-..,.~. .<
[. .- .,.;- f .2 . I CONTROL TO 20 FT g C33 l l' C34 i C35 C36 20 FT 20 FT L8 20 FI 20 FT Rt31$10R j,UFI, $UBTRACTCR A+ f (([Ci( Fill $ I Log FAlts LOW (AILS.LOS $1K.10
$HORT (OPEN)
SIE
$U2041FL PR20 AIFL LS2041N$
R(20411H BA2041FL
$U2081FL PR2061FL LS2061N$
Rf20813N BA2081FL SU2082FL PR2062FL L$2062N3 Rf2062$H IA2062FL 320xlFL PR2011FL L32011NS RE20113H BA20llFL l _ CONTROLS 10 28 FT 20 FT $tTPOINT RH[N CONTROL 10 FAIL $ VALVE 20 FT OteANDED NIGH i 42 0 C40 g% { T ITPOINT PERAfD 28 Fi LEVEL SfiPolNT SET 100 ERRCNE00$tt $tiP0lNT Buff 0N RELAT FAIL $, MIGH EITREBE $1GNAL j $ ELECTS 28FT INelCAft$ 28 FT ElICAllIRAIl0- FAILS LEVEL CONTROL BUTTON MAS OfEN l gigg DEPRi$$E0 WHEN IT ACTUALLY MAS NOT pgs[L2g OPlu2041 $P204t FH yE$fL28 OPlu2061 $P2061FH OPEu2062 $P2062FM % f OPl!!L28 OPlu20XI $P2011FH cPtstL23 RY2841As RT286Lau Rv2s62A' B-27 RY28tlA5 t [ i
\
i l
T I -- .. ..... . i 121 22 l 23 24 i CONTROL TO 2 FT FAILS HIGH (FvV.yl/FIV 1El/ ' FWV 162/FIV-X1) C45 C46 C47 C48 gs p 2 FT SETPOINT IF F FAILS VALVE HIGH BUFFER ; RESISTOR K+ l SUBTRACTOR I. PR0241FL~ SUO24tFL / C49 BA0241FL RE0241SH
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' 4, APPENDIX C HUMAN RELIABILITY EVENT TREES i
s i I o r t i
=
i f f . s
y AFVTRPOP Operator fails to manually initiate Dnergency Feedwater a k. (
'l A
c C b B d 9 C b B c 1 B 1 b i C c i i C-1
r , _.~., . .. ..__._..__.__c.. . - . . . r AIVTRPOP_(cont'd) A = operator fails to recognize the need for AIV, low MEV pump head alarm Al= operator reads message incorrectly B = operator fails to recognize the need for AIV, low Steam Generator alara B = operator reads message incorrectly C = operator fails to initiate taak of manually actuating AFW once the l need is recognized i t f ( 1 l t I 1 C-2
.~. -.- . . ...:.~-~-...~ .-.. .
.g-
-r
( ASV5IZ6? Operator fails to manually open steam admission valve A i k: a 1 ( l C b c I b B d D B c C 1 b e d D c C
,. d D
q I i I ! I' !l I l
.t
-k C-3
, . ...m.. . . _ . _ _ . . . _ . . . . . _ , .
t 4
\
ASV5ZZ9P (cont'd) A = operator f ails to recognize the e. ed to open valve, no flow indication from pump Al= reads low flow alarm message incorrectly B = operator fails to recognize the need to oper valve, low pump AP alarm 31= reads 109 pump AP alarm message incorrectly C = auxiliary operator f ails to initiate taak of opening valve once given the instruction by the control room operator D = auxiliary operator manipulates wrong valve 1 L l I* t i f_ i C-4
. _ , _ . . ._a.__...._. . . . . ?
SWSUCTOP Operator fails to properly switch suction from Condensate Storage Tank to alternate AFW Suction A 0 D e E b g is e l
/s E
h Ie fs t i i A = fails to notice low CST meter reading B = ignores AP pump suction alarm C = fails to read message indicating no pmp AP D = fails to carry out plant procedure to switch suction when CST level is low E = does not switch suction valves in correct order, fails to fcilov a logical procedure G = selects wrong MDV switches l l I l C-5
1 EFv3ZZRE/EFV4ZZRE [ Operator fails to open valve EFV-3/EFV-4, which had been left closed after p op preventive maintenance. A i 1 A 1 a B b C C d A = operator ignores alarm, low pump AP alarm A = operator fails to read alarm message correctly B = operator selects wrong switch to trip off pump C = operator in Auxiliary Building fails to initiate task of opening valve upon receiving instruction to do so D = operator in Auxiliary Building fails to locate proper valve I i a C-6
r= . . ,...:...... ._....u.-..-..... .. . EFV3ZZLC/EFV4ZZLC
.f Valve EFV-3/EFV-4 inadvertently left closed after preventive maintenance on pop.
C i C d d D 6 h b B e E 6 A B
\
\
e E 6 A l t e E
-l-t
~.
1 1 C-7
- f. . . . . . . . _ , . . . _ _ _ _ . . . . . . . . ..
, l l
1 l EFV3ZZLC/EFV4ZZLC (cont'd) B = failure of checker (2nd operator ur maintainer) to discover error of omission C = failure to use check-off provision of procedure properly D = operater commits error of emission using check-off procedure p'roperly D = operator commits error of caission using check-off procedure improperly A = valves sticks during restoration E = operator fails to restore a sticking valve completely 1 i i C-8
.l OPFAILPT/0FFAILPH Operator fails to isolate open path, one control /tvo control valves, have failed high I4 8
C b 4 D C d D c C
\
i C-9
. . . . . . . . .. . .-.... . +. . .
=
lt OPFAILPT/0PFAILPH (cont'd) A = operator fails to notice that a control vahe(c) has failed high, from its position indication B = operator fails to recognize the need for AW control valve closure, fails to respond to high SC level alara - C = operator attempta to close isolation valve to isolate path but manipulates wrong POV switch D = operator fails to verify corrective action by checking Steam Generator level e-C-10
r , . . .. . . . _ ~ . .. , i f l 6 OPFAILSM Operator fails to put control string in safe mode (select close bias) upon failure of any element in the instrumentation string l a A b B (- C I
/
l D A = operator fails to respond to annunciator, indication of level transmitter failure (SG 1evel alara) B = operator reads message incorrectly l C = failure to check other indications to insure taht initial condition is ! actually an instrumentation failure and not an incorrect Steam Generator level
- r. D = operator fails to manipulate proper switch on control cabinet (select close bias switch) to put control valve in safe mode l'
I C-11
....-...a _a.._.......
r _. . . . - _ P: j
\
l t: OPESI2,28 Operator erroneously selects ',8 f t level control, he believes a thCA maists. a A B b l t A = operator misreads containment radiation monitor, believes high containment radiation exists. B = fails to follow plant policy to check for other 1.0CA indications before switching to 28 ft. level control I I , i l C-12
T -' . _,m......,.. _ ~ . . . . i l i c P Miscalibration (histables, se2 points, etc.) I A N B b r i l C C A = failure of maintenance technician to initiate calibration taak B = failure of technician to properly adjust calibration device, involvec reading a digital display C = fails to correctly adjust bistable, setpoint, etc. 1 4 v P v l l I I C-13 _ . _ -}}