ML20214E655
| ML20214E655 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear, Crystal River, 05000000 |
| Issue date: | 04/30/1981 |
| From: | Enzinna R, Jeffery Lynch BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20214E454 | List: |
| References | |
| 32-1125434, 32-1125434-00, NUDOCS 8705220183 | |
| Download: ML20214E655 (133) | |
Text
%o 32-1125434 .00 Emergency Feedwater System
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[-p Upgrade Reliability Analysis .
13 i For The 1
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32-1125434-00 EMERGENCY FEEDWATER SYSTEM UPGRADE l RELIABILITY ANALYSIS FOR THE ARKANSAS NUCLEAR ONE NUCLEAR GENERATING STATION I UNIT NO. 1 (Contract 582-7179)
P. it Performance Engineering Babcock and Wilcox ,
Nuclear Power Generation Division P.O. Box 1260 Lynchburg, Virginia 24505 l
t Prepared by -
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Prepared by M d5 8 M acI w, :33 ; mcw. d e ILtMi
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32-1125434-00 CONTENTS Pace
1.0 INTRODUCTION
1 -1
2.0 DESCRIPTION
OF ANALYSIS 2-1 2.1 Fault Tree Analysis 2-1 2.2 Human Reliability Analysis 2-2 2.3 Failure Data 2-2 2.4 Mission Success Definition 2-2 -
2.5 Assumptions 2-4 3.0 RESULTS 3-1 3.1 Quantitative Results 3-1 3.2 Dominant Contributors to System Unavailability 3-1 4.0 RECOMMENDATIONS AND CONCLUSIONS 4-1 References Appendix A - Prevention of Main Feedwater Overfill A-1 Appendix B - Fault Trees 51
)
o Appendix C - Human Reliability Event Trees C-1 i
Appendix D ~ System Description Overview I D-1 SUPERSEDED 1
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32-1125434-00 LIST OF TABLES Pace 3-1 Unavailaoilities for ANO-1 EFWS 3-4 L
LIST OF FIGURES A-1 Failure to Terminate MFW Overfill A-6 A-2 Signals to Terminate MFW Overfill A-7 D-1 EFIC System Organization D-10 D-2 EFW Initation Logic D-11 j D-3 Typical (1x2)2 Logic Format D-12 D-4 SGA Vector Logic D-13
, D-5 OTSG Overfill Protection Logic D-14 D-6 EFW Control System D-15 0-7 AP&L EFWS P&ID D-16 I
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32-1125434-00
1.0 INTRODUCTION
This report presents a suntnary of the analysis methods and results of a reliability study of the proposed Arkansas Nuclear One (ANO-1) k Emergency Feedwater (EFW) _ and Emergency Feedwater Initiation and Control (EFIC) Systems. Tne objectives of this study were:
- 2. To identify dominant contributors to system unavailability.
- 3. To identify outlying system weaknesses, if any, and make appropriate reconsnendations for incorporation into the design process in order to ensure comparatively superior system reliability.
A brief description of the ANO-1 EFW and EFIC systems appears in Appendix D.
This report documents the analysis methods used including assumptions made-and analytical tools employed. The results are presented in both quanti-tative and qualitative terms. The numerical results include probability of failure. per demand of EFW initiate, EFW control, feed only good generator logic (FdGG), and EFW overfill protection. The qualitative ;
results include identification of major failure contributors, discussion of their significance with respect to system reliability, and conclusions f drawn from the results.
Appendix A contains a discuss'.on of prevention of main feedwater overfill.
I The EFIC system generates the signals used for terminating (by closing valves in the Main Feedwater System) Main Feedwater overfill should it !
occur. Based on the type of initiating event, probabilities of failing
( to terminate Main Feedwater overfill are presented.
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2.0 DESCRIPTION
OF ANALYSIS 32-1195434 99 Fault Tree analysis was the primary method used to evaluate the ANO-1 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 tnen used to quantify these cut sets. Human error probabilities used in the quantification were developed using the methodology described in Section 2.2.
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The results of this analysis include the dominant contributors to system failure. These contributors were reviewed to assess their effect on 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.
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.
The fault trees were constructed to a level of detail sufficient to identify '
i 4 all relevant connon hardware in the system. This level of analysis allowed identification of hardware which would, if failed, reduce designed redundancy
)
or cause the failure of other hardware. There was no attempt to account for connonalities imposed by external events such as fires, floods or earthquakes.
In addition to mechanical failure of hardware, failure causes due to human
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actions and test or maintenance activities were included. As indicated by the 2-1 < b.
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32-1125434-00 results, the human contribution to system failure probability frequently dominates the hardware contribution.
The computer code FTAD (2) was used for quantification of the fault trees, k.. ranking of basic event importance, and identification of major contributors to system failure.
2.2 Human Reliability Analysis The Human Reliability Analysis (HRA) was performed consistent with the methodology described in NUREG/CR-1278, 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; however, few data bases exist for human error rates. A technique developed at Sandia Laboratories (3) was used to quantify human error probabilities.
Probability tree diagrams for the human tasks of interest were constructed and are presented in Appendix C.
2.3 Failure Data [Fr.% 32 -ll1604/ - bo)
Point value estimates of the component failure data were obtained from references (4) through (15). The data obtained from the above references are indust / wide, i.e., 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 bt b d"* #
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ANO-1 experience. f DdA C'ILd d'h l
j 2.4 Mission Success Definition j In order to evaluate the impact of component failures on system reliability an explicit definition of mission success is required. Mission Success
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of EFW was divided into four parts; initiation, control, overfill protection, and F0GG. . . . .1m;p* l 2-2 -- i iQ f yJw<ch M b 10 l
32-1195434-00 For initiation and control, system unavailabilities were calculated for two cases; the first case represents strictly automatic initiation and control of the system, and the second case allows operators to intervene to correct system failures within 20 minutes. Or.ly automatic actuation of overfill protection and FOGG was evaluated.
- 1. Initiation Mission success 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 fail low or any one control valve fails higher than the rate limiter 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. FOGG 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 on? flow path to the good steam generator. Failure of FOGG in the cases where specifications call for feeding of both steam generators (i.e.,
spurious isolation of EFW) is included in the initiate evaluation.
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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.
2.5 Assumptions In evaluating the EFW reliability the following assumptions were made:
- 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 would open.
- 3. Degraded failures were not considered, that is, components were assumed to operate properly or were treate'd 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 reseat at a somewhat lower setpoint that may still be considered a success, and rarely do the code safety valves fail tc reseat entirely.
L 4. Mechanical failures of passive components (such as locked open manual valves) were disregarded due to their extremely low failure rates.
- 5. EFW overfill protection bistables fail untripped on loss of cabinet power supply.
- 6. All EFIC relays are solid state.
- 7. Sensor / transmitter failures will be repaired at next plant shutdown.
24 SUPERSEDED i
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32- 1 ' a5 4 r e - 00 !
- 8. Failure data used for the condensate storage tank are generic and do not account for the unusually active nature of the ANG-1 condensate storage tank.
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- 3.0 atsutTs M-1125434-00 3.1 Quantitative Results System unavailabilities were calculated for EFW initiation, control, overfill protection, and FOGG. 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 l operators to intervene to correct system failures within 20 minutes. Only automatic operation of overfill protection and FOGG was evaluated.
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 3.2.1 EFW Initiation (fully automatic initiation)
- 1. Pum; P7B unavailability is dominated by loss of offsite power and subsequent diesel generator failure.
- 2. Pump P7A unavailability is dominated by pump mechanical failure.
- 3. A major contributor to unavailability of either or both trains is inad-vertent failure to realign valve CV-2800 and/or CV-2802 after pump maintenance. Another major contributor to unavailability of either or both trains is failure to reclose pump 4 inch test lines after pump test.
- 4. If a LOOP occurs and both diesel generators fail to start, AC powered valves CV-X2 and CV-X3 will not open. Since CV-X1 and CV-X4 are DC
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powered and battery backed, they will open, providing flow paths from i
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32-1125434-00 the turbine driven pump to the steam generators. However, in order to provide motor pump protection, both recirculation valves CV-2815 and
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CV-2816 open automaticly. These valves are on the comon recirculation line for the EFW pumps. Even though the turbine driven EFW pump P7A i
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has open paths to both steam generators, its flow to the generators l Will be degraded by the open recirculation path. An analysis should be done to detemine if adequate flow will be available to the steam generators under this circumstance.
- 5. A dominant contributor to unavailability of both trains is flow I blockage of either check valve CS-98 or CS-99.
EFW Initiation (includes operator corrective action)
- 1. The major contributor to pump P78 unavailability is loss of offsite power and subsequent diesel generator failure. Secondary to this is pump mechanical failure.
- 2. Pump P7A unavailability is dominated by mechanical failure and being out of service for maintenance.
- 3. Contributor number (4) for the automatic initiation case is also a major contributor for this case.
- 4. Flow blockage of check valve CS-98 or CS-99 and subsequent failure l
'I of the operator to switch suction to the service water system is a
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dominant contributor to unavailability of both trains.
3.2.2 EFID Control 1 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
- 1. The most dominant contributor to FOGG failure is steam generator I pressure transmitter failure (either fails high or fails 1R
( 3-2 SUPE /.$
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- 2. A major contributor to FOGG failure is miscalibration of the P<600 psi bistables (a bistable set too low will fail to respond to a bad steam generator).
Another important contributor for the initial condition when both
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l steam generators are less than 600 psi and one if 150 psi greater than the other, is miscalibration of the 150 psi differential pressure bistables.
3.2.4 EFW Overfill Protection
- 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 I
the overfill condition. This condition does not apply to overfill protection circuits in cabinets C and D because these cabinets are not used ior EFW control..
- 2. In all four cabinets, level transmitters are the dominant contributors to failure of the overfill protection circuit trains.
- 3. Another major contributor to overfill protection circuit failure is bistable miscalibration.
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TABLE 3-1 )
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- UNAVAILABILITIES FOR AND-1 EFW '
1 L i Unavailability i g EFW Initiate includes failure to initiate EIV j g due to fluid system failure, spurious isolation by FOGG or overfill protection, and EFIC initiation failure.
Fully automatic initiation 3.9 x 10~4 Includes operator corrective action within 5.4 x 10-5 4
20 minutes J
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i EFIC Control
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) Fully automatic control 8.0 x 10 t
Includes operator corrective action within 1.2 x 10 ~0 20 minutes FOGC
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A > 600 psi B < 600 psi (feed only A) 3.8x10,f A < 600 psi B > 600 psi (feed only B) 3.8 x 10,3 A < 600 psi B < 600 psi A 150 psi > B (feed 1.5 x 10
[f A < 600 psi only A) ,3
'I B < 600 psi B 150 psi > A (feed 1.5 x 10 only B)
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EFW Overfill Protection 6.2 x 10"#
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i 32-1105434-00 4.0 RECOMMENDATIONS AND CONCLUSIONS Section 3.2.1 identifies the pump recirculation configtration as a dominant contributor to unavailability. In the postulated scenario, failure to j open of the motor pump train flow control valves (due to loss of AC power) results in automatic opening of recirculation valves CV-2815 and CV-2816 which consequently diverts flow from the turbine driven pump. We recommend that an analysis be performed to determine if adequate flow will be available to the steam generators under this circumstance. If analysis indicates that adequate flow will not be available from the turbine train to the steam generators, then we recommend that an alternative to the proposed recirculation configuration be implemented. It is important to note that it. the postulated scenario (loss of offsite and onsite AC power) the motor driven pump is inoperative and recirculation for the protection of the motor driven pump is unnecessary. ,
Due to the active nature of the ANO-1 condensate storage tank and the identification of check valves CS-98 and CS-99 as dominant contributors to unavailability, we recomend further analysis of the ANO-1 EFW primary water source in order to ensure existence of an adequate water supply for the EFWS, Fro:n this analysis we can conclude that except for the two concerns described above the proposed Ap&L ANO-1 EFWS design has no outlying weaknesses. The proposed upgrade analyzed 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 4-'
SUPERSEDED s
I 32-1125434-00 f by the overfill protection and FOGG features. Fluid system changes that represent significant reliability improvements include: pump ,
discharge piping reconfiguration with parallel flow control valves; turbine steam supply changes such as parallel steam admission valves, addition of pressure regulating valves, and check valves to separate
-steam generator A and B steam supplies; elimination of AC power dependencies in the turbine driven train; and additional condensate storage tank level indicators.
The reliability study revealed that a major failure c6ntributor for EFW control, F0GG, and overfill protection was miscalibration of setpoints.
The importance of human errors on system failure should not be underestimated.
Even though a system is as well designed as the EFIC system and adequate redundancy is built-in, human error represents a potential comon mode
- failure. Since maintenance personnel or other technicians . interface with tne entire system there is a potential for design redundancy to be reduced through human error.
Human errors contribute to the EFW and EFIC system unavailability in at least three ways. First, miscalibration of setpoints in FOGG, overfill protection, or control represents a dominant failure contributor in each of
? these functions. Seconc, a major contributor to EFK pump unavailability is a
leaving the pump valved out after preventive maintenance. Third, inadvertently leaving recirculation valves open after pump test is a major contributor to pump unavailability.
Therefore, it is recomended that administrative procedures for the maintenance, test, and calibration of the EFW and EFIC systems be carefully i prepared. These procedures should confom to accepted human factors I
engineering pnactices in order to minimize the e ror probpb.11.14y;..apsod#tedm
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with their use. Additionally, administrative controls should be applied !
l to ensure that these well-written procedures are effectively used by plant personnel during EFIC and EFW maintenance. -
1 l
A major contributor to failure of FOGG and EFW overfill protection is
)
level and pressure transmitter failure. It is expected that this would be the dominating hardware contributor 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 I which in turn causes it to dominate other hardware contributors. Although the transmitters are the major hardware failure contributor for FOGG and EFW overfill protection they are not limiting factors to EFW unavailability.
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 transmiters may be the initiating cause of the control failure resulting in the overfill condition, and consequently reduced 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.
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32-1125434-00 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 EFN overfill protection device, the orobability 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 protet!. ion probability in Table 3-1 already accounts for the comonality discusse'd 1 .
the preceding paragraph). The result is that for the automatic case the probability of AFW overfill is 5.0x10-6 per EFW demand, and including operator corrective action the probability of EFW overfill is negligibly small.
In conclusion, further analysis is recomended for two concerns. These concerns involve the proposed common pump recirculation line with automatic opening of valves CV-2815 and CV-2816 upon loss of AC power, and adequacy ,
of the EFWS water supply. Once these two concerns have been satisfactorily ,
addressed this analysis will show that the proposed ANO-1 EFW and EFIC systems are reliably designed. '
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32-1125434-00 REFERENCES i
- 1. N. H. Roberts, and D. F. Haasi, Fault Tree Handbook, NUREG-0492, November 1978.
- 2. FTAP 2, Comeuter-Aided Fault Tree Analysis, Babcock and Wilcox
- Document Number NPGD-TM-536, February 6, 1980.
I 3. A. D. Swain, H. E. Guttmann, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, NL7.EG/CR-1278, October 1960.
- 4. W. H. Sullivan and John P. Poloski, Data Summaries of Licensee Event Reports of Pumos 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 Sumaries of Licensee Event Reports of Diesel Generators at U.S. Comercial Nuclear Power Plants, NUREG/CR-1362 EGG-EA-5092, March 1980
- 7. Reactor Safety Study: An Assessment of Accident Risks in U.S. Comercial Fuelear Power Plants, Appendix III, WASH-1400, NL9tEG-75/014, October 1975.
- 9. IEEE Guide to the Collection and Presentation of Electrical Electronic and Sensing Cc ponent Reliabilitv Data for Nuclear-Power Generr. tion Stations, IEEE Std. 500-1977.
- 10. J. W. Pegram, Come11ation of Failure Rates for Use in Ouantitative g Reliability Analvses Compiled froc Publiclv Available Sources, g Babcock and Wilcox Company, Nuclear Power Generation Division, Lynchburg, Virginia, Septe=ber 1975.
y 11. Nuclear Plant Reliability Data System, 1979 Annual Reports of Cumulative
. System and Component Reliability, NUREG/CR-1635, Septe=ber 1980.
- 12. Military Standardization Handbook, Reliability Prediction of Electronic Equipment, MIL-HDBK-217C, Department of Defense, April 1979.
l
- 13. Report on Reliability Survey of Industrial Plants, Part I: Reliability of Electrical Equipment, IEEE Comittee Report, IEEE Transactions on Industry Applications P. 213, Vol. IA-10, No. 2, March / April 1974.
- 14. M. Hnatyshyn, Reliability Program Report for Engineered Safety Features Actuation System II. WNP Units 1 and 4, Automation Industries, Inc.,
Vitro Laboratories l>ivision, Silver Spring, Maryland, May 1979.
- 15. Reactor Protection System, Topical Report BAW-10085P, Rev. 6. Babcock and Wilcox, Nuclear Power Generation Division, April 1979. . ,,
( I 32-1125434-00 l 16. System Description Emergency Feedwater System for AP&L ANO-1 I 15-1119467-01, Babcock and Wilcox, Nuclear Power Generation Division February 1981.
- 17. W. W. Weaver, R. W. Dorman, R. S. Enzinna, Aux 111arv Feedvater Systems Reliability Analyses for Plants with Babcock and Wilcox P.eactors BAW-1584, Babcock and Wilcox, December 1979.
(( . 3-Abmed 32-ll1604l? I
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, R.Ju e n t I.s f evps& fu b fuen c.e #10: , * *J.W.ftysm , C u m a:Id*%
- f f* U"' (' W * [- ? O cus+Atu t e\ A;l.'+ y *F' A n<lne. s (c ~ oiles (<a n Pu b l.d.,
h<. law .Snvezs , Laheug. s,l( W,'Isoy S.[1] Earles, D. R.
" Reliability Application and Analysis Guide" The Mar. tin Company, MI-60-54 (Rev.1); July 1961.
/D.[2] VDEW Problem and Damage Statistics VDEW, Frankfurt, Ger=any, 1966 - 1970. -
82-1125434-00
/0- [3] Balfanz, H. P. -
"FalJure Ra te Compilation" IRS-W-8, December 1973.
Institute for Reactor Safety, Germany.-
/$. [4] Sexton, 1. B.
" Failure Rate Estimates of some Nuclear and Conventional Power Plant Cenponents" Atomics International AL-Memo-15 .(Revised) 1968.
/0-[5] Green, A. E. and A. J. Bourne
" Safety Assessment with Ref erence to Protective Systems for Nuclear Reactors" UKAEA, AHSB(S) R117, 1966.
/O[6] Best1, W. , Gieseler, H. M. and G. Teniglia
" Reliability Analysis in Reactor Design and Safety Assessment"
-0 ECD, Water-cooled Reactor ,, Safety, Fanj, May 1970.
6 -
/0.[7] ucleonics Week, Vol.14, No. 8, p. 6.
/0.[8] " Failure Data Handbook for Nuclear Power Facilities".
Liquid Metal Engineering Center DIEC Memo-69-7, Vol. 1, August 1970.
/0. [9] Orvay. H. J. and R. C. Erd:.an
" Reactor Siting and Design from a Risk Viewpoint" Nuclear Engineering and Design. Amsterdam i Vol. 13, No. 2 August 1970. .
/0.[10) f .kler,11. C. , B. J. Carrick et.al.
" Reliability Analysis of !!uclear Power Plant Protective Systems" Holmes and Narver, Inc. , Los Angeles HN-190, May 1967.
/i
/0. [11) Zylstra, E. H. van
Recemended Failure Rates for' Relf ability Analysic" l1 May 1960. '
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-f- VALVE p DD p=SiNNAL
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LOGIC i 1 s U A 1r B C to D tit k CO A e- o L" TRIP LOGIC 9.! I E D F D g g g 'g 5 A. - C,_,A C
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32-1125434-00 APPENDIX B r= nzzs I E i l i I ( L i l j i p, p r; p a pw., p4 7w. q. p? 2 Mud $~;~ E il' b J1 1 J 15.7'.. lad 35
l l 32-1125434-00 l COMPONENT IDENTIFICATION CODE i l 1 COMPONENT CODE Buffler amp or battery BA (batteries have numbers, e.g. , BAD 07) Breaker BR I Cabinet main circuit breaker CB Charger CH Controller CN , Pressure sensor, delta P DP FOGG module FC FOGG module FG Inverter IN Level sensor LS Level transmitter LT Valve or human operator OP Proportioner PR Resistor RE
' Rate follower RF Relay RY-Setpoint signal SP Sensor power supply transformer SPT Subtractor SU Trip cabinet TC Timer circuit TI Valve stem VS i
SUPERSEDED 3Y l
i 32-11'25434-00 COMPONENT FAILURE NDDES l IDENTIFICATION, CODE l FAILURE ) ODE CODE All Modes AM Catastrophic Failure CF Erroneously Trips ET Flow Blockage FB Fails Closed FC Fails to Energize FE Fails to Reinitialize FI Fails Low FL Fails Open FO Tails to Run FR Fails to Start FS Fails to Switch FW Indicates Trip IT Left Closed LC Lef t Open LO Mechanical Failure MF Miscalibrated High MH Miscalibrated Low HL No Signal NS In Preventive Mai'tenance n PM Fails to Reseat RS Spuriously Closes SC Shorts SH Spurious Signal SS Fails to Close TC Fails to Detect Trip TD Fails to Open TO In Test TS Unable to Trip UT fI!O h c.
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= . -
32-1125434-00 HUMAN RELIABILITY ANALYSIS Human reliability event trees for the operator actions of interest. are presented in Appendix C. The basic human error probabilities (BHEPs) , associated with each branch of the trees may be found in. Chapter 20 of Reference (3) in Tables 20-1 through 20-24.
- The. failure paths are designated with capital letters.and the success paths with lower case letters. The failure probability for the tree is found by suming the paths that end with failures (capital letters).
For.a discussion of the effects of stress on the BHEPs see Chapter 17 of i Reference (3). The doubling rule is also discussed in Chapter 17 of i Reference (3). Once quantified the values for operator action are inserted into the fault tree (see Appendix B) and are treated as basic events in the fault tree quantification process. 4 e t 1 t l hO C2E(D$ {_. l 1
~
l . 1 AT% MOP = /k to' /M 32-1125434 90 l 1 Operator fails to manually initiate Auxiliary Feedvater I a
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i AFLTRPOP (cont'd) 32-1125434-00 A = operator fails to recognize the need for AW, low MW pump bead alarm l A = operator reads message incorrectly B = operator fails to recognize the need for A W , lov Steam Generator alarm B = operator reads message incorrectly C = operator fails to initiate task of manually actuating A W once the need is recognized l - l I C-2 SUPERSEDED h7
l l 32-1125434- 00 l l CV2800RE = /.h/o'#/4.,, ,g l Operator f ails to open CV2800 upon EW demand. Valve had been left closed af ter motor pump preventive maintenance. Q o*?
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D' . A = operator fails to notice low pump t.P alarm A operator incorrectly reads low AP alarr message B = operater selects wrong switch to trip off pu=p C = auxiliary operator fails to initiate task of opening valve once given order by control room operator. D - auxiliary operator manipulates incorrect valve. 1
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Eh: ' C-3 lO O
32- 11254 3 4- Ou CV2802RE : A/or#/p Q ) I Operator fails to open CV2802 upon EFW demand. Valve had been left closed after turbine pump preventive maintenance. o c o*' A+ a A E E <? 1 a 1, , . o ' I # b 0*,oS g4 r h c- o.o* b r 8 I d ,o.8' 9 4
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I i ! A 1 A = operator fails to notice low pump LP alarm I l AI = operator incorrectly reads low LP alam message B = operator selects wrong switch to trip off pump C = auxiliary operator fails to initiate task of opening valve once given order by control room operator, j D = auxiliary operator manipulates incorrect valve E = auxiliary operator fails to properly reset turbine trip SUPERSEDED C-4 lof
i CV2800LC/CV2802LC = 3.3vo-3/w 32-117.5434- 0 i Valve ETV-3/EFV-4 inadvertently lef t closed af ter preventive maintenance on pump. I \ # C* 6 o-0 o9 ,o b oN
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CV2800LC/CV2802LC (cont'd) 32-1125434-00 B = failure of checker (2nd operator or maintainer) to discover error of omission C = f ailure to use check-off provision of procedure properly [ D = operator connaits error of emission using check-off procedure properly D = operator comunits error of omission using check-off procedure improperly a = valves sticks during restoration E = operator fails to restore a sticking valve completely I I d E l
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P7ARECOP/P7BRECOP 5 98 /v#/d%g 32- 112 54 3 4- d0 An EFW pump is in test when an EFW demand occurs. 9 o"
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. ,f P7ARECOP/P7BRECOP (cont'd) 32-1125434-00 A = operator fails to respond to low AP alarm B = operator incorrectly reads low AP message . C = operator fails to respond to low SG level alam D = operator incorrectly reads low SG level message E = auxiliary operator fails to initiate valve restoration task once given the order by the control room operator G = auxiliary operator fails to select proper valve to close I
I J 1 I SUPalSEDED C~8
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P7ARECLO/P7BRECLO = 7.3 vo'#/0er,,a / d Four inch recirculation line left open after test c
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32-1125434-00 l P7ARECCA, P78RECCA = .7.luo #/q Operator fails to realign recirculation valves, which had been inadvertently left mispositioned, upon AFW demand. I 3 5' a A
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A = operator fails to realyze that recirculation valve is closed, fails to respond to low EFW discharge pressure. 7 B = operator in auxiliary building fails to initiate the task of closing Y recirculation valve upon having received the instruction to do so. C = operator in auxiliary building fails to close proper valve.
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- Ih4= Q Sk'SUCTOP r Vr f to # y Operator fails to properly switch suction from Condensate Storage Tank to alternate An' Suction 9
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,0-I A = fails to notice low CST meter teading B = ignores AP pump suction alarm C = fails to read message indicating no pump AP D = fails to carry out plant procedure to switch suction when CST level is low I E = does not switch suction valves in correct order, fails to follow a logical procedure G = selects wrong HOV switches s <, esr y n e m t*~ #
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32-1125434-b0 d CV2803CA/CV2806CA/CV3850CA/CV3851CA/CV2803LA/CV2806LA - 4" 'O' Operator local corrective action fails at valve l gi o,e 9
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32-1125434-00 CV2803CA/CV2806CA/CV3850CA/CV3851CA/CV2803LA/CV2806LA (cont'd) A = operator fails to respond to low AP alarm B = operator incorrectly reads low AP message C = operator fails to respond to low SG 1evel alam D = operator incorrectly reads low SG 1evel message E = auxiliary operator fails to initiate valve restoration task once given the order by the control room operator I G = auxiliary operator fails to select proper valve to close I I l 1 I . SUPERSEDED C-13
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l CW1TPOPrlo80*Y M 32-1125434-00 I l Operator fails to manually open steam admission valve ! 9 9 i 9 9 g O' 4 o0 a 0* g, I i ,
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9 I cn n , ccont d> 32-1125434-00 . A = operator fails to recognize the need to open valve, no flow indication from pump - A l= reads low flow alarm message incorrectly B = oparator fails to recognize the need to open valve, low pump AP alarm Bl = reads low pump AP alarm massage incorrectly I C = auxiliary operator fails to initiate task of opening valve once given the instruction by the control room operator D = auxiliary operator manipulates wrong valve I - B 7
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N 3 l . C-15 SUPERSEDED I (1 -
0?TAILPT/0PTAILPH = /.fseo * *[7.fe 4' *fu
,J 5434-00 Operator f ails to isolate open path, one control /two control valves, have failed high E
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OPFAILPT/OPFAll?H (cont'd) 32-1125434-00 A = operator fails to notice that a control valve (s) has f ailed high, from its position indication B = operator fails to recognize the need for AFW control valve closure, fails to respond to high SG 1evel alarm
~
C = operator attempts to close isolation valve to isolate path but manipulates wrong POV switch D = operator fails to verify corrective action by checking Steam Generator level I l t I I . l C-17 SUPERSEDED
/N
32-1125434-00 O?FAILSM = 0 /t/0' l Operator fails to put control string in safe mode (select close bias) upon failure of any element in the instrumentation string
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o .0 g D* i l t A = operator fails to respond to annunciator, indication of level transmitter failure (SG' level alarm) B = operator reads message incorrectly C = f ailure to check other indications to insure t ht initial condition is actually an instrumentation failure and not an incorrect Steam Generator I level D = operator fails to manipulate proper switch on control cabinet (select close bias switch) to put control valve in safe mode l l , ,h [$J l e A .ChQ)CE~&jg';g l' C-18 !
//s i
32-1125434-U0 OPESE1.28 = /> / o - "/M Operator erroneously selects 28 ft. level control, he believes a LOCA l exists. l a' e 9 o# I i r A* I p.% 5 B ,0 *' b l l i A = operator misreads containment radiation monitor, believes high containment radiation exists. B = fails to follow plant policy to check for other LOCA indications before switching to 28 ft. level control I I I I C-19 SUPERSEDED l l i w Ilb
i 32-1125434-00 l Miscalibration (bistables, setpoints, etc.): # 84<0"# M e d [A miscoAb!"f4*
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,' c.L ygi) c A = failure of maintenance t'echnician to initiate calibration task B = failure of technician to properly adjust calibration device, involves reading a digital display C = f ails to correctly adjust bistable, setpoint, etc.
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32-1125434-00 MAIK FEEDWATER OVERFILL = M e /v,,/M y "Arwet c * [ Y '"
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,a d D, Pg 3 l "A" = f ails to detect deviant display, one of one, moderately high stress due to Rx trip "B" = manipulates wrong switch to trip pump off "C" = f ails to respond to alarm, moderate stress "D" = manipulates wrong switch to trip pump, high stress due to short time I
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l l I l I 32-1125434-00 l l l i i 1 3 l I l APPENDIX D l SYSTEM DESCRIPTION OVERVIEW
]
l i I I 1 ! I 1 1 I i D-1 W -~DED Il4
32-1125434-00 0.0 DESIGN DESCRIPTION The following system description and accompanying diagrams represent a condensed version of the more complete description found in the
" Emergency Feedwater System Description for Arkansas power and Light" (16).
D.1 Summary Description The emergency feedwater (EFW) system consists of two interconnected trains, capable of supplying EFW to either or both steam generators (SG) from { either water source under automatic or manual initiation and control. piping and instrumentation diagram is included as Figure D-7 of this A report. The system pumps (EFW pumps) take suction from either the condensate storage tank or from the service water syste- and discharge to the SGs. In the flow path between the EFW 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 D.2. The instrumentation system design is described in Section D.4. D.2 Fluid System Desion D.2.1 Suction The primary water source for both EFWS trains is the condensate storage 9 tank, T-41. This tank is required by technical specifications to contain N a reserve of 107,000 gallons for EFWS use. Water is supplied from this 7 tank to a comon suction header via a single line containing a locked , open valve and check valves. An alternate EFWS suction is available from the nuclear service system, loops one and two. Suction may be manually transferred from the conden-sate storage tank to the nuclear service water system by means of motor-operated valve pairs. A comon control switch for each pair causes the valves to assume opposite positions; that is, if one valve is open the other valve is closed and vice versa. Operators are alerted to perform this suction transfer by a low suction pressure alann on the comon suction header. . j SUPERSEDED D-2 {1 0
32-11254g;.99 D.2.2 Pumos and Discharoe Cross-Tie EFWS train A uses a turbine-driven pump P7A, and the train B pump P78, is motor-driven. Either pump has a capacity sufficient to supply adequate water to either or both SGs. These flows include a nomal recirculation flow and under low system flow conditions recirculation L flow paths open to allow additional recirculation flow. The pumps discharge through check valves into cross-connected discharge lines. The cross-connection line contains two nomally-closed motor-3 g operated valves. This cross-tie pemits either pump to feed either or both steam generators. D.2.3 Emeroency Feedwater Flow Control Valves The flow of EFW to each SG is controlled by redundant nomally-open modulating solenoid motor operated control valves in parallel paths. These control valves are designed to fail "open". Initiation and control instrumentation for these valves is described in Section D.4 of this report. D.2.4 Steam Generator Isolation Valves Each SG can be isolated from EFW flow by nomally-open motor-operated ; valves. These valves are located in the parrilel lines downstream of the EFW control valves. Initiation and control instrumentation of these valves is described in Section.D.4 of this report. l f D.2.5 Recirculation and Test Lines Recirculation and test lines are connected to the discharge piping of I J both pumps. Nomal recirculation for pump protection is accomplished i D with nomally-open flow paths consisting of small lines, restricting orifices, and needle valves as shown in Figure D-7. Additional recircu- ' lation flow is provided for pump protection when the flow paths to the steam generators are closed. This additional recirculation flow path is provided by interlocking motor-operated valves CV-2815 and CV-2816 flow elements in the parallel lines to the steam generators. If low flow is sensed by the flow elements upstream of valves CV-X1 and CV-X2, valve CV-2816 is comanded to open. If low flow is sensed by the flow elements upstream of valves CV-X3 and CV-X4, valve CV-2815 is comanded to open. c4 j D-3 l2 ( l
32-1125434-00 Simultaneous opening of valves CV-2815 and CV-2816 will allow increased recirculation flow. In addition to the nomal recirculation path, a recirculation test line and its interconnected valves (FWil A, FWilB, FW12A and FW128) are used to perfom full flow tests of the EFWS pumps by diverting flow to the condenser hotwell or the condensate storage tank. I D. 2. '6 Steam Supply for the EFWS Turbine Steam supply for EFW Pump p7A turbine is obtained from both steam generators via locked-open valves. Downstream of these valves the pipes g join to fom a conrnon supply to the pump turbine. A check valve is
~
installed in each line to preclude blowing down the good steam generator ! in the event of a steam line or feed line break. Upstream of the turbine are redundant DC motor-operated nomally-closed valves. These valves are opened automatically on EFWS initiation. They may also be manually opened. A description of the controls for these valves is contained in Section D.4. Steam from these valves passes through a redundant pressure r' educing station and on to the turbine governor valve. A turbine overspeed trip valve is also provided to trip the turbine if turbine speed becomes excessive. Turbine trip is alamed in the control room; the trip valve ; must ce reset locally. ' f Two overpressure relief valves are connected to the steam supply line upstream of the turbine governor. These valves will protect the piping I and turbine downstream of the pressure reducing valves in the event of a ;
. pressure reducing valve failure.
m 1 Turbine exhaust is vented to the atmosphere. D.3 Supporting Systems The EFW pumps, pump motor and turbine are self-contained entities without dependencies on secondary support systems. The bearings on the turbine and both pumps are lubricated by slinging oil from reservoirs near the ? bearings._ Coolingisaccomplishedbyheat$bsgtQes 0 f 2b '
D.3.1 Power 32'-1125434-00 , The two EFW trains are powered from diverse power sources. The motor I driven pump (P78) train is powered by AC power. AC power for all components needed to obtain emergency feedwater flow is derived from diesel generator-backed 4160 VAC busses. To ensure EFW flow in the event of a loss of all AC power, the turbine driven pump (P7A) train derives its power from the SGs for the pump and from a battery-backed DC buss for its valves. 1 0.4 Instrumentation Description The emergency feedwater initiation and control system (EFIC) is an instrumen-tation system designed to provide the following:
- 1. Initiation of emergency feedwater
- 2. Control of EFW at appropriate setpoints
- 3. Level rate control when required to minimize overcooling
- 4. Signals for temination of main feedwater to a steam generator on approach to an overfill condition
- 5. The selection of the appropriate steam generator (s) under conditions of steamline break or main feedwater or emergency feedwater line break downstream of the last check valve
- 6. Temination of EFW to a steam generator on approach to overfill conditiqns
- 7. Control of atmospheric dump valve to predetemined setpoint, and g 8. Signals for isolation of the main steam and main feedwater lines of 6 a depressurized steam generator, f The EFIC - see Figure D consists of four channels (A.B.C. & D).
Each of the four channels are provided with input, initiate, and vector logics. Channels A and B also contain trip logics and control logics. Each channel monitors inputs by means of the input logic, ascertains whether action should be initiated by means of the initiate logic and detemines which SGs should be fed by means of the vector logic. 1 , SUPERSEDED I?h
i j 32-1105434-00 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 i SG by means of control logics to maintain SG level 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 Channels A, B, C and D initiate logics. By means of trip logics, Channels A and.B terminate ; main feedwater to a steam generator that is approaching overfill. l l D.4.1 Input Logic The input logics are located in each of the channels. The input logic:
- 1. receives analog input signals l-
- 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 stimuli
- 5. provides signals to the remaining channel logic.
D.4.2 Initiate Logic The initiate logic, depicted in Figure D-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 four RC pumps are tripped <
l y 2. both main feedwater pumps are tripped-4 U
- 3. the level of either steam generator is low
- 4. either steam generator pressure is low
- 5. Flux to MFW flow ratio trip is present Other functions of the initiate logic are:
. 1. issue a call for SG A nmin feedwater and main steamline isolation when SG A pressure is low
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32-1125434 I 2. issue a call for SG B main feedwater and main steamline 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 level 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.pemissive conditions. Thebypass(es)are automatically removed when the pemissive condition teminates
- 6. provide for maintenance bypassing of an EFIC initiate logic D.4.3 Trip Looic The trip logic is illustrated in Figure D-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 I
- 2. SG A main steamline isolation
- 3. SG B main steamline isolation
- 4. SG A main feedwater isolation
- 5. SG B main feedwater isolation It should be noted, for the later discussion of the vector logic, that f the trip logic outputs a signal when 1-out-of-2 taken twice trip of EFW occurs. Also, note the presence of the vector enable switch.
i For each trip function, the trip legic 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 reduces the possibility of accidental manual initiation, i l SUPERSEDED i o_7 12 5
32- 11254 3 4- do Trip signals are transmitted out of the EFIC by activating a relay thereby gating power onto trip busses. In this manner, the EFIC provides power to energize the control relays whose contacts fonn the AND gates in the controllers. D.4.4 Vector Logic The vector logic - Figure D appears in each of the EFIC channels - I Figure D-1. The vector logic monitors:
- 1. SG pressure signals
- 2. SG (A and B) overfill signals
- 3. EFW trip signals (vector enable) originating in Channel A and B trip logics The vector logic developes signals for open/close control of steam generator A and B emergency feedwater valves.
The vector logic outputs are in a neutral state until enabled by trip signals (vector enable) froni the channel A or B trip logics. Once enabled, the vector logic will issue close comands to the valves associated with any SG for which any overfill signal exists. 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
'6 SG A Valve 'SG B Valve pressure Status Comand Consnand 2 {
( SG A & B > Setpoint Open Open l l SG A > Setpoint & SG B < Setpoint Open Close 2 SG A < Setpoint & SG B > 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 f 'SG A 150 psi < SG B O Clos SU! ERSEDnED D-8 12.6
i 1
$ D.4.5 Control Locic 32-11S5434-U0 The control logic is depicted in Figure D-6. For each SG (A and B)-
there are two controls which are selected by transfers within the control circuit. The three foot level setpoint control is automatically 4 selected when an EFW trip occurs with one or more reactor coolant pumps operating. A level rate control with a twenty foot setpoint is selected when an EFW trip occurs with no reactor coolant pumps l operating. I I i l t ,7 . 7 p , r...e,, ( ..1;7
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FsGuaE 0-1 32- 11-0 54 3 4- d0 EFIC System Organization i CHANNEL A y INITIATE EFW ColNCI DENCE ' ISOLATE SGA MAIN STEAN LOGIC = ISCLATE SG8 NAIN STEAN ISOLATE SGA NFW RC PUMP TRIP r 10 SLATE SG8 NFW INPUT NFW PUMP TRIP = SG A CONTROL OGIC CONTROL { S.G. LEVEL S. G. PRESSU RE e LOGIC O SG 8 CONTROL-NEUTRON FLUX : YECTOR SGA EFW VALVES CLDSE logic 50PENl>SG8EFWVALVES
- CLOSE CHANNEL B INITIATE EFW-COINCIDENCE
& ISOLATE SGA MAIN STEAM LOGIC z ISOLATE SG'8 MAIN STEAN RC PUMP TRIP -
O ISOLATE SGA NFW j NFW PUMP TRIP
- INPUT C ISOLATE SG B NFW l . LOGlC S.G. LEVEL CONTROL 5 SGA CONTROL S.G. PRESSURE LOGIC & SGB CONTROL 4 VECTOR m
> SGA EFW VALVES LOGIC OPS
-CLOSE)"SGBEFWV
- CLOSE, CHANNEL C RC PUMP TRIP VECTOR = OPEN
- MFW PUNP TRIP ,7 IN PUT LOGIC SGA EFW VALVES
& CLOSEJ S.G. LEVEL ; LOGI C
= OPD S. G. PRESSURE -
SGB EFW VALVES' f NEUTRON FLUX O CLOSEJ CHANNEL D RC PUNP TRIP
=
MFW PUMP TRIP OPDI\SGAEFWVALVES S.G. LEVEL ;= lNPUT VECTOR LOGlC
- CLOSEJ
~
! S.G. PRESSURE *= LOGIC r OPEN MEUTRON FLUX
= CLOSE
? SUPER)SEDED i 6
.. _. . _ - _ _ _ - . - - --/LR . -
"""E -2 32-1125434-d0 ,
l EFW Initiation ~ Logic , RC PUNP At TRIP ? RC PUNP A2 TRIP ? RC PUMP B1 TRIP 7 RC PUMP B2 TRIP 7 I 9 < = 5$ F.P. ? { SHUTDOWN BYPASS RESET g ; g
~
SHUTDOWN BYPkSS I r : MFW PUMP A TRIP ? " I MFW PUMP B TRIP 7 m e < 0 5r, F.P. 7 s
; SHUTOOWN BYPASS RESET 3C C U' SHUTDOWN BYPASS OC
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V SE A LEVEL < l' ? SS B LEVEL < l'? i s SHUTDOWN BYPASS y L OC T h
=
SHUTOOWN BYPASS RESET gC 0 A 9
- a SG A & SG B PRESS < % 750 7 SG A PRESS < t 600 ?
i-x r SG B PRESS < 0 800 7 v '; SUPERSEDED
/29 -.l
T I W 32-1125434-00
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