ML19262C368
| ML19262C368 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 12/31/1979 |
| From: | Dorman R, Enzinna R, Weaver W BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19262C367 | List: |
| References | |
| NUDOCS 8002110387 | |
| Download: ML19262C368 (51) | |
Text
{{#Wiki_filter:, o EMERGENCY FEEDWATER SYSTEM RELI ABILITY ANALYSIS FOR THE THREE MILE ISLAND NUCLEAR GENERATING STATION UNIT NO. I BY W. W. Weaver R. W. Dorman R. S. Enzinna Revision 1 Decenter 1979 Babcock & Wilcox Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Va. 24505 1946 168 8002110 38 k
TABLE OF CONTENTS Section Page Executive Summary iii 1.0 Introduction 1 1.1 Background 1 1.2 Objectives 1
- 1. 3 Scope 1 1.4 Analysis Technique 2 1.5 Assumptions & Criteria 2 2.0 System Description 4 2.1 Overall Configuration 4 2.2 Supporting Systems 7 2.3 Power Sources 8 2.4 Instrumentation & Control 9 2.5 Operator Actions 10 2.6 Testing 10 2.7 Technical Specification Limitations 10 3.0 Reliability Evaluation 11 3.1 Fault Tree Technique 11 3.2 Comparative Reliability Results 11 3.3 Dominant Failure Contributors 12 Refe rences Appendix A A-1 Appendix B B-1 1946 169 i
LIST OF FIGURES
- 1. TMI-1 EPAS
- 2. River Water Supply to the EFWS - TMI-1
- 3. AC Power Distribution to EFWS Components - TMI-1
- 4. EFWS Initiation and Control Functional Logic Diagram (Simplified)-
IMI-1
- 5. Comparison of TMI-1 EPWS Reliability with NRC Result for Westinghouse Plants.
1946 170 ii
EXECUTIVE
SUMMARY
The NRC has requested all operating plants with Babcock & Wilcox (B&W) designed reacters to consider means for upgrading the reliability of their Emergency Feedwater Systems (EFWS). As a part of the response to this request, Metro-politan Edison / General Public Utilities and the other B&W Owners Group utilities have requested B&W to perforTn a simplified reliability analysis of existing emergency feedwater systems. This draft report presents the results of that reliability study for the EFWS for Three Mile Island. Unit 1 (TMI-1). The primary objective of this study was to evaluate EFWS reliability (defined as " point unavailability") using an approach which would produce results comparable to those obtained by NRC Staff analyses for Westinghouse and Combustion Engineering Plants. Another objective was to identify dominant failure contributors affecting system reliability. EFWS reliability was assessed for three cases: Loss of Main Feedwater (LMFW) with reactor trip, LMFW with Loss of Offsite Power (LMFW/ LOOP) and LMFW with Loss of all AC Power (LMFW/LOAC). System reliability was assessed by the construction and analysis of fault trees. The results of the study are on the following page. These results indicate that EFWS reliability for TMI-1, based on reliabilities obtained by the NRC for Westinghouse plants, is medium for all cases. Some of the dominant failure contributors which were identified in this study were,1) failures of EFWS initiation and control components, resulting in failure to obtain actuation for either EFWS train, 2) component unavailability resulting from preventive maintenance activities, 3) pump failures, and
- 4) human error associated with leaving manual valves closed 6fter testing or maintenance of the motor driven pumps.
A similar study is being performed for each Owners Group utility and additional plant specific draf t reports will be prepared. At the conclusion of the program, information contained in the plant specific reports will be collected and used to generate a generic reliability report comparing all B&W operating plants. 1946 171 iii
CASE 2: LOOP CASE 3: LOAC CASE 1: LMFW LOW MED HIGH LOW
- ME0* HIGH*
LOW ME0 HIGH b i TMl-1 15 @ @ C 30 Q g O ALL W PLANTS e O C OO O 8 MISSION SUCCESS WITHIN 5 MINUTES H RANGE OF W PLANTS MISSION SUCCESS WITHIN 15 MINUTES AS FOR CASES I & 2
$ $ MISSION SUCCESS WITHIN 30 MINUTES N COMPARISON OF TMI-1 EFWS RELIABILITY WITH NRC RESULTS FOR W PLANTS N
- 1. 0 Introduction
1.1 Background
This report presents the results of a reliability study for the Emergency Feedwater System (EFWS) for Three Mile Island Unit 1. The NRC is conduct-ing similar analyses for Westinghouse and Combustion Engineering plants. Preliminary results of the NRC study are available (Reference 1) and have been included in this report for comparison with the TMI-1 EFWS relia-bili ty. The approach employed in this study has been developed in close coordination with the NRC and is, therefore, expected to yield comparable resul ts . 1.2 Objectives The objectives of this study are: e To pc-form a simplified analysis to assess the relative reliability of the EFWS for TMI-1. It is intended that the results of this analysis be directly comparable to those obtained by the NRC for' Westinghouse ard Combustion Engineering Plants. This is assured by the use of the same evaluative technique, event scenarios, assumptions and reliability data used by the NRC. e To identify, through the development of reliability-based insight, dominant failure contributors to EFWS unreliability. 1.3 Scope The EFWS was analyzed as it will exist prior to startup, presently scheduled for early 1980. Three event scenarios were analyzed: o Case 1 - Loss of Main Feedwater with Reactor Trip (LMFM) e Case 2 - LMFW coincident with Loss of Offsite Power (LMFW/ LOOP) e Case 3 - LMFW coincident with Loss of all AC Power (LMFW/LOAC). These event scenarios were taken as given; that is, postulated causes for these scenarios and the associated probabilities of their occurrence were not considered. Additionally, external comon mode events (earthquakes, fires, etc.) and their effects were excluded from consideration. For each of the three cases, system reliability as a function of time was evaluated. 1.4 A_nalysis Technique The evaluation of reliability for the EPAS was based primarily on the construction and analysis of fault trees. This technique encourages the development of insights which permit identif' cation of the primary contributors to system unreliability. Application of this technique is described in detail in Section 3.1.
- 1. 5 Assumations and Criteria Assumptions and criteria were made in consultation with the NRC staff and were selected to assure that the reliability evaluation results will be comparable to those obtained by the NRC for the Westinghouse and Combustion Engineering analyses.
- 1) Criterion for Mission Success - In order to evaluate the overall reliability contribution of system components, it is necessary to establish whether or not failure of those components will prevent successful accomplishment of the EFWS mission. Thus, it is necessary to explicitly define the criterion for mission success. The criterion adopted for this study was the attainment of flow from the turbine-driven pump or from both motor-driven pumps to at least one steam generator.
System reliability was calculated at times of 5,15, and 30 minutes to allow for a range of operator action. These times were specifi-cally chosen because NRC-supplied operator reliability data for these times was available; however, these times are reasonable and con-sistent with LMFW mitigation for B&W plants. In their study, the NRC staff has used steam generator dryout time as a criterion for success-ful EFWS initiation, and the 5-minute case represents a comparable result for B&W plants since emergency feedwater delivery within 5 minutes will prevent steam generator dryout. However, steam generator dryout itself does not imply serious consequences; a more appropriate criteria is the maintenance of adequate core cooling. Recent ECCS analyses (Reference 2) have shown that adequate core conling can be maintained for times in excess of 20 minutes without EFWS operation, providing that at least one High Pressure Injection Pump is operated. 1946 174
- 2) Power Availability - Tt a following assumptions were made regarding power availability:
LMFW - All AC and DC power was assumed available with a probability of 1.0. LMFW/ LOOP - OC power was assumed available with a probability of 1.0. One diesel generator was assumed unavailable with a prob-abili ty of 10-2 The other diesel generator was available with a probabili ty of 1.0. LMFW/LOAC - DC and battery backed AC were assumed available with a probabili ty of 1.0.
- 3) NRC-Supplied Data - NRC supplied unreliability data for hardware, operator actions and preventive maintenance were assumed valid and directly applicable. These data are listed in Appendix B.
- 4) Coupled Manual Actions - Manual initiation of valves with identical fcnction was considered coupled. Such valves were assumed to be both opened manually or both not opened. The case in which.one valve was opened and the other valve was left closed was not considered.
- 5) Degraded Failure - Degraded failures were not considered; that is, components were assumed to operate properly or were treated as failed.
- 6) Initiation and Control Reliability - Although initiation and control circuits for the two emergency feedwater trains are largely separate, both within and outside the Integrated Control System (ICS), common dependencies require the assumption of a single source for initiation and control. A failure probability of 7 x 10-3 was assigned for this sou rce.
- 7) Pump Suction Strainers - An NRC-supplied unreliability of 1 x 10-4 was assigned to the strainers in the EFW pump suction lines.
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2.0 System Description
2.1 Overall Configuration A diagram of the TMI-1 EFWS is presented in Figure 1. The system consists of two feed trains supplied by one turbine-driven pump and/or two motor-driven pumps with a common suction source. This system can feed emergency feedwater to either or both steam generators under automatic initiation and control . 2.1.1 Suction The primary water source for the EFWS consists of two interconnected condensate storage tanks. Each tank has a capacity of 250,000 gallons; each tank is required by Technical Specifications to contain a minimum of 150,000 gallons for EFWS use. A common suction header for all three EFWS pumps is supplied with water via two 10-inch lines to the two condensate storage tanks. These lines contain normally-open AC motor-operated valves (COV10A and B), and check valves (C0V16A and B). Another water source is the 165,000 gallon condensor hotwell. Water can be acquired from this source by manually opening either air-operated valve COV8 or AC motor operated valve COV12 and breaking condensor vacuum. A backup source of river water is available via the Reactor Building Emergency Cooling Pumps. This source is described in Section 2.2. The common suction header contains normally-open AC motor-operated sectionalizing valves (EFV1A and B). The suction connection for each pump contains a flow tub ' use in controlling recirculation flow, locked open valves, and stra:ners. 2.1.2 Pumps and Discharge Cmsstie Emergency feedwater is supplied to a common discharge crosstie by three pumps: a turbine-driven pump (EFP1) and two motor-driven pumps (EFP2A and B). The turbine-driven pump is rated at 920 gallons per minute at 1020 psig with the recirculation control valve closed. Each .notor-driven pump is rated at 460 gallons per minute at 1020 psig, with the recirculation control valves closed. 1946 176 A recirculation path is provided for each pump. This path consists of a lh-inch line containing a check valve, an air-operated control valve, a flow orifice and a normally-open valve. All three recirculation lines are connected to a comon recirculation line. Recirculation for each pump is controlled by a flow tube in the pump suction line which provides a signal to the air-operated recirculation control valve. As pump suction flow decreases below a certain point, the control valves open to maintain a minimum recirculation. The control valves fail open on loss of control air. However, the utility has calculated that, with these valves fully open, pump EFP1 can continue to provide a flow of 385 GPM to each steam generator (total flow of 770 GPfi) and either pump EFP2A or EFP2B, can provide a 400 GPM discharge (total flow of 800 GPM when both motor driven pumps are operatii g). Each motor-driven pump discharges t a common discharge crosstie via check valves and nonnally-open valves. The turbine-driven pump discharges directly to the crosstie between normally-open AC motor-operated section-alizing valves (EVF2A and B). The crosstie permits any of the three pumps to feed either or both of the steam generators. 2.1.3 Flow Control Valves The flow of emergency feedwater to each steam generator is controlled by air-operated modulating flow control valves (EFV30A (train A) and EFV30B (train B)). Positioning of these valves is controlled by electric to pneumatic converters that receive control signals from the Integrated Control System (ICS). The valve positions are controlled to maintain desired steam generator water levels; ccntrol for these valves is described further in Section 2.4.2. The valves are also interlocked with pressure switches so that emergency feedwater (and main feedwater) is cut off to a given steam generator if a low pressure (< 600 psig) is detected within that generator. 1946 177 2.1.4 Steam Supply for the EFWS Turbine Steam for the EFWS turbine is obtained from a tie between the two main steam lines from each steam generator. The steam supply line from each steam generator contains a normally-open AC motor-operated valve (MSV2A or 2B) and a check valve (MSV9A or 9B). Also connected to the steam lines are the ICS-controlled atmospheric dump valves (MS4A & B) and turbine-bypass valves. The steam supply lines are connected to the turbine inlet line via a network of steam admission valves. As shown in Figure 1, steam can be admitted to the turbine by opening either of the air-operated valves MSV13A or MSV138 or by opening either of the DC motor-operated valves MSV10A or MSV10B. The air-operated valves are in 2-inch lines which are designed to admit a sufficient quantity of high pressure steam to permit a smooth startup of the turbine and avoid turbine overspeed. The air-operated valves are opened automatically by the EFWS initiation logic discu sed in Section 2.4.1. The air-operated valves are interlocked to prevent their opening if steam pressure in the corresponding steam generator is below 100 psig. Valve MSV13A opens preferentially; if Steam Generator A pressure is above 100 psig, MSV13A will open on initiation and MSV13B will remain closed. MSV138 opens only if there is low pressure in Steam Generator A and the pressure in Steam Generator B is above 100 psig. Low pressure in both steam generators will cause both valves to remain closed. As decay heat is removed arJ steam generator pressures decrease, it may become desirable to obtain a less restrictive steam admission path to the tu rbi n e. This can be accomplished by manually opening either of the two DC motor-operated valves MSV10A or MSV10B which are in 6-inch lines. Auxiliary steam is normally unavailable and was not considered for the purpose of this study. After the steam admission valves, steam for the turbine passes through a pressure control valve, MSV6, that controls steam pressure to 200 psig, to the turbine governor valve. Two overpressure relief valves (MSV22A and B) are providec .with setpoints of 495 and 505 psig, respectively. Turbine exhaust is veni.ed directly to the atmosphere.
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2.1.5 Valve Operation and Indication Table I contains information concerning the operation, position indication, motive power source and control for active valves in the EFWS and support-ing systems. All motc.-operated valves are AC powered with the exception of MSV10A and MSV10B which are DC powered. All motor-operated valves are position indicated and controllable from the control room. The power for position indication and control for each valve is derived from the power source for the valve operator. Air-operated valves position indicated in the control room are shown in Table I. Certain air operated valves are on a backup air supply which will assure valve operability under degraded AC power conditions during which normal control air is lost. (Refer to Section 2.2) 2.2 Support Systems and Backup Water Source The EFWS tuttine, pump motors, and pumps are self-contained entities wi ;hout dependence on secondary support systems. However,' the pump mr' ors cannot run indefinitely without forced ventilation. For this s'. dy, this dependency was ignored because the pumps can operate for in excess of 2 hours without ventilation and the pump room fans are powered from diesel generator-backed source. A support system affecting EFWS reliability is the air supply for certain EFWS air-operated valves. The TMI-1 air supply system consists of two 60 hp compressors, I A-P1A and I A-P1B, and two back-up 5 hp compressors, IA-P2A and I A-P2B. Power for compressors I A-P1A and I A-P2A is derived from diesel generator backed 4160 VAC bus "10". The power for compressors I A-P1B and I A-P2B is derived from diesel generator backed 4160 VAC bus "1E". In an emergency, air can also be supplied from the two station service air compressors provided off-site power is available. Unde r nonnal conditions instrument air is supplied to the emergency feedwater and turbine plant components by the two main instrument air compressors IA-P1A and IA-P18. During loss of off-site power conditions, the main instrument air compressors will continue to be supplied power from the diesel generators provided an ESAS actuation has not occurred. Under loss of off-site power with concurrent ESAS actuation, these instrument air 1946 179 compressors are automatically shed from the diesel generator. When the normal air supply is not available, the key emergency feedwater and turbine plant equipment automatically receives air from the back-up 5 hp air compressor and their 80 gal. air reservoirs. These back-up compressors remain loaded on the diesel generator supplied bus even during loss of off-site power and coincident ESAS actuation conditions. Compressor IA-P2A feeds a comon supply line which provides air to the flow control valves (EF-V30A and B), the atmospheric dump valves (MS-V4A and 4B), the EFW turbine pressure control valve (MS-V6), and the turbine driven pump recirculation control valve (EF-V8B). The other back-up air compressor supplies air to the turbine bypass valves. A backup supply of river water is available from the Reactor Building Emergency Cooling pumps. This water supply enters the EFWP common suction header between sectionalizing valves EFVIA and EFV1B. The backup water supply diagram is shown in Figure 2. A large number of manual actions are required to access this backup water supply. Motor-operated valves EFV4 and EFV5 are normally kep t locked closed and the motor control center breakers for these valves are locked open; these locks must be removed and the breakers closed. Than an RB emergency cooling pump must be started to satisfy interlocks .: thin EFV4 and EFV5. Finally, these valves must be opened along with pump discharge valves (RRV1A and B). 2.3 Power Sources A simplified diagram showing the AC power t'stribution to some EFWS compo-nents is provided in Figure 3. As shown, AC power for EFWS components necessary to achieve emergency feedwater flow is derived from diesel generator-backed 4160 VAC busses 1D" and "1E". Normally (Case 1) power for these busses is obtained from the switchyard. However, in the event of LMFW/ LOOP (Case 2), the diesel generators are automatically started and critical EFWS components will remain operable: ( All AC MOV's required for EFWS initiation are normally open and do not require a change of position. However, operation of some AC MOV's is required to obtain suction from the condensor hotwell; this operation would include opening of the condensor vacuum breaker valve, and opening of C0V12 if COV8 is inoperable). 1946 180 In the event of LMFW/LOAC (Case 3), EFW may still be initiated because startup and operation of the EFWS turbine pump is not AC dependent. Critical air-operated valves will open on eventual loss of normal and backup air, but these valve openings do not prevent EFW initiation. (Flow control valves EFV30A and EFV30B, steae u...:ssion valves MSV13A and MSV13B, tuttine pressure control valve MSV6, and recirculation control valve EFV8 will fail open.) 2.4 Instrumentation and Control 2.4.1 Initiation Logic A functional logic diagram for EFWS initiation is shown in Figure 4(A). The diagram is simplified and does not show all redundancies actually in the hardware. As shown, the motor-driven pumps will be started upon loss of all four reactor coolant pumps or low differential pressum across both main feedwater pumps, providing that voltage is available on the respective 4160 VAC bus for each pump and following any delay caused by an engineered safeguards actuation. The pump motor start signals for pumps A and B are also sent to the EFWS turbine steam admission valves (MSV13A and B). The steam admission valves may be opened by the pump motor start signals or by turbine trip signals from both Main Feedwater Pump turbines. As explained in Section 2.1.4, valves MSV13A and MSV13B are interlocked with steam supply line pressure switches so that MSV13A opens preferentially. The logic shown in Figure 4( A) is supplied with battery-backed power. 2.4.2 EFWS Flow Control The flow of emergency feedwater to the steam generators is controlled by air-operated control valves EFV30A and EFV308. These valves modulate the flow to obtain desired steam generator levels under direction of ICS-controlled electric / pneumatic converters. The battery-backed ICS logic for valve control is shown in Figure 4(B). If all four RC pumps are tripped, the valves will open and control to the setpoint for RC pump trip. If at least one RC pump is operating, but both main feedwater pumps have tripped, the valves will open and control to a lower setpoint. If at least one RC pump and one main feedwater pump is operating, both valves are directed to remain closed. Manual control of valve position is also available. JOff jg1 _9
As shown in Figure 4(B), operation of the flow control valves is interlocked with steam line rupture detection logic. If a steam generator pressure of less than 600 psig is detected by this logic, the respective flow control valve will close to shut off emergency feedwater flow to that generator. 2.4.3 Ins trumen ta tion EFWS instrumentation in the control room, in addition to valve positions previously described, includes: e condensate storage tank level for both tanks. e EFW pump discharge pressure for all three pumps. e Emergency feedwater flow to each steam generator. e Steam generator pressure and level for each steam generator. 2.5 Operator Actions Assuming no component failures have occurred and the system is correctly configured, no operator actions are required to obtain emergency feed-water flow in Cases 1, 2 or 3. Several operator actions may be required to maintain flow after some interval following system initiation. These actions include opening of valves MSV10A and MSV10B to assure a continuing adequate steam supply
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to the EFWS turbine and opening of COV8, COV12 and the condensor vacuum breaker valve if required to obtain EFWS suction from the condensor hotwell. The manual actions described in Section 2.2 are required to obtain suction from the backup river water source. 2.6 Tes ting The three EFWS pumps are tested on a monthly basis. The test demon-strates that each pump can be started and will operate on recirculation for one hour. This test also demonstrates operability of the valves associated with the turbine steam supply. Flow control valves EFV30A and EFV308, steam supply valves MSV10A, f1SV10B, MSV13A and MSV13B, and turbine pmssure control valve MSV6 are cycle tested on a quarterly basis. 2.7 Technical Specification Limitations Technical specifications require that all three EFWS pumps be operable before the reactor coolant can be heated above 2500F. However, one pump is permitted to be out of service for maintenance for a period of up to 48 hour,. Af ter this time period has expired, the plant must be placed in cold shutdown within 12 hours. Technical Specifications also require that a minimum of 150,000 gallons be maintained in each of the two condensate storage tanks. 3.0 Reliability Evaluation 3.1 Fault Tree Technique The TMI-1 EFWS reliability was evaluated by constructing and analyzing a fault tree. The f -lt tree developed during this study is contained in Appendix A. The top level event in this tree is failure to achieve mission success; from this point, the tree branches downward to a level of detail corresponding to NRC-supplied data. This level is generally indicated by basic event circles. As indicated on page A-1, system failure can result from preventive maintenance related failures or component failures. Component failures consist of failure of the turbine-driven pump train coupled with a failure of one of the motor-driven pump trains. The combinations of failures which can defeat the cump trains are addressed in the following pages of the appendix. The techniques used in fault-tree construction and the symbols shown in Appendix A are consistent with those used in WASH-1400 (Reference 3). Following completion of the tree, hand calculations were performed to obtain system unavailability for 5,15 and 30 minutes for each of the three event scenario cases. 3.2 Comparative Reliability Results The results of the analysis are presented in Figure 3. Indicated in this figure are the system reliability results for each of the three cases and for each time 5,15 and 30 minutes. The basic format for this figure, including the characterizat.on of Low, Medium, and High reliability, was adopted from infonnation presented by the NRC in Reference 1. Because the NRC-supplied input data were often unverified estimates of component and human reliability, absolute values of calculated system reliability must be de-emphasized; results have significance only when used on a relative basis for purposes of comparison. Accordingly, the intent of Figure 5 is to show the relative reliability standing of the TMI-1 EFW3 for each of the three cases and also to compare these results to the NRC results for Westinghouse plants. The Westinghouse results and numerical values permitting construction of Figure 5 were all obtained from Reference 1. It should be noted that there is a scale change for the Case 3 results; reliability results for Case 3 cannot be cross-compared with Cases 1 and 2. As shown in Figure 5, relative to Westinghouse, TMI-1 has low to medium reliability for Cases 1, ? and 3. The underlying causes for these reliability results are described below. Some general observations may be made regarding the results in Figure 5. As the time for operator action increases from 5 to 30 minutes, the probability of mission success improves slightly. Itst of the improve-ment occurs 'sie 4en 5 and 15 minutes, reflecting a difference in the NRC-supplied operator reliability data for these times. On the other hand, there was little difference in the operator reliability data between 15 and 30 minutes and this is reflected in the system unavailability results. The small difference in the results for Cases 1 and 2 is primarily caused by the relatively improbable loss of one diesel generator. The Case 3 results mainly reflect the loss of both motor-driven pumps. 3.3 Dominant Failure Contributors 3.3.1 Case 1 - LMFW Dominant failure contributors for Case 1 include:
- 1. Failure to obtain emergency feedwater to either steam generator caused by the failure of actuation circuit components common to both feedwater trains (e.g., failures within the ICS which would prevent both flow control valves from opening).
- 2. Preventive maintenance outages affecting one pump coupled with component failures affecting either or both of the other pumps.
1946 184
- 3. Pump failures may involve control circuit failures, mechanical failures or plugging of the suction strainers. In addition, failure of the turbine-driven pump may be caused by steam supply inadequacies such as stuck-open safety relief valves, or failure of MSV13A to open. Steam supply valve interlocks could prevent MSV13B from opening should there be a mechanical failure of MSV13A, necessitating manual actuation of steam supply valves MSV10A and MSV108.
- 4. Another important cont ibutor to system unavailability involves human error associated with inadvertently leaving manual valves EFV10A and EFV108 closed after testing or maintenance of the motor driven pumps.
3.3.2 Case 2 - LMFW/LOAC Failure contributors for Case 2 are the same as for Case 1. The loss of offsite power will cause some motor-operated suction valves to become inoperable from the control room. However, these valves are normally kept in the correct alignment, and loss of valve operability has little effect on overall system availability. Loss of one diesel generator reduces system availability because of the loss of the associated motor driven pump; however, dominant failure contributors for the remaining portion of the system are unaffected. 3.3.3 Case 3 - LMFW/LOAC ihe dominant failure contributors for Case 3 are the same as those of Case 1 which pertain to the turbine-driven pump. An additional potential contributor is failure of the pressure control valve, MSV6, to the fully open position upon loss of both normal and backup air. This may cause a turbine overspeed trip or a degraded steam supply if the inlet relief valves MSV22B and MSV22A are continuously open. A major finding of this study is the lack of AC dependencies which would prohibit EFWS mission success on loss of all AC power. 1946 185 TABLE I EfWS AND BACKUP WATER SUPPLY VALVES Position on Posi tion on Pos. Ind. Locotion Normal Loss of Loss of and For Manual Valve No. Posi tion Control Signal Motive Power Location Motive Power Source Opera tion EF-VlA Operi as-is as-is Yes/C.R. lA-ES 480V Control Center (5) EF-VlB Open as-is as-is Yes/C.R. 18-ES 480V Control Center (5) E F-V2A Open as-is as-is Yes/C.R. lA-ES 480V Control Center (5) EF-V2B Open as-is as-is Yes/C.R. 1B-ES 480V Control Center (5) EF-V4 Closed as-is as-is Yes/C.R. IC-ES 480V Control Center (5) EF-V5 Clos 9d as-is as-is Yes/C.R. 1C-ES 480V Control Center (5) EF-V8A Opentl) Open Open No Instrument Air /125/250 VDC Dist Panel IC E F-V8B Open Open Open No Instrument Air /125/250 VDC Dist Panel IC EF-V8C Open(l) Open Open No Instrument Air /125/250 VDC Dist EF-V30A Closed (2) Half Open Full Open Hand /Autd6)PanelIC Instru n{3; Air /A Inverter thru (5) Station EF-V30B Closed (2) Half Open Full Open Hand / Auto (6) ICS/NNI Instrum9 Air /A Inverter thru (5) Sta tion ICS/NNIL C0-V10A Open as-is as-is Yes/C.R. lA-480V Turb Plant Control Center (5) (Non-ES) CO-V10B Open as-is as-is Yes/C.R. 1A-480V Radwaste Control Center (5) (Non-ES) C0-V14A Open as-is as-is Yes/C.R. lA-ES 480V Control Center (5) C0-V14B Open as-is as-is Yes/C.R. 1B-ES 480V Control Center (5) C0-V111A Open as-is as-is Yes/C.R. 10-Turbine Plt Control Center (5) (Non-ES) C0-V111B Open as-is as-is Yes/C.R. 1D-Turbine P1 t Control Center (5) (Non-ES) MS-VIA/B/C/D Open as-is as-is Yes/C R. IC-ES Valve 480V Control Center (5) MS-V2A Open as-is as-is Yes/C.R. 1C-ES Valve 480V Control Center (5) MS-V2B - Open as-is as-is Yes/C.R. 1C-ES Valve 480V Control Center (5) MS-V8A e Open as-is as-is Yes/C.R. IC-ES Valve 480V Control Center (5) MS-V8B A Open as-is as-is Yes/C.R. 1C-ES Valve 480V Control Center (5) MS-V10A Os Closed as-is as-is Yes/C.R. 125/250 VDC Dist Panel IC (5) MS-V10B
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Closed as-is as-is Yes/C.R. 125/150 VDC Dist Panel ID (5) MS-V13A Closed N/A Open Yes/C.R. Instrument Air /125/250 VDC ES Panel (5) O lE os
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Position on Position on Pos. Ind. Location No rnal Loss of Loss of and For Manual Valve No. Posi tion Control Signal Motive Power Location Motive Power Source Operation MS-V13B Closed N/A Open Yes/C.R. Instrument Air /125/250 VDC ES Panel (5) IF MS-V3A-F Closed llalf Open Closed Yes/C.R. Ins tru n (5) ICS/NNI 3{ Air /A Inverter thru 1 MS-V4A/B Closed Half Open Closed Yes/C.R. Instrument Air /A Inverter thru (5) ICS/NNI(3) MS-V6 Initially Open Yes/C.R. Ins tru Open ICS/NNI np 3 Air /A Inverter thru C0-V6 Closed N/A Closed Yesg(4)i) TI Instrument Air /125 VDC Relay C0-V7 Closed N/A Open Yes R Instrument Air /125 VDC Cab CO-V8 Closed N/A Open Yes(4) Instrument Air /125 VDC XCC CO-V12 Closed as-is as-is Yes/C.R. lA ES Valve MCC (5) C0-V13 Open as-is as-is Yes/C.R. 1C Turbine Plant MCC (5) (1) Position during normal operations. Valve closes upon sensing pump flow of 100 or 200 gpm. (2) Position during nonnal operations. (3) A Inverter fed from 1A ES Motor Control Center. Alternate power from A battery thru inverter lA. (4) Indication on solenoid demand not valve position. (5) Equipped wi th handwheel. (6) Not true position indication, demand signal only. 4 N
REFERENCES
- 1. " Auxiliary Feedwater Reliability Study," an NRC staff presentation to the ACRS at the ACRS Meeting of July 26,1979,1717 "H" Street, Room 1046, Washington, D.C.
- 2. " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," May 7,1979.
- 3. WASH-1400 (NUREG-75/014), " Reactor Safety Study (Appendix II)", USNRC, October 1975.
t I 1946 188
Fig. 1 TMI-1 EFRS TtF8IPE Af t05 A 7505 e # OUIP Od'E W 5 u Cs ggy 4g uSV AA
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TO EFWP SUCTION I I EFV-5) L.C. EFV-4) L.C. RRV-8A [RRV-8B HS OR ES ACT - [ [ - HS OR ES ACT RRV-1A RRV-18 4" 4" 12" LARGE MESH I2" L.C. u STRAINE O DISCHARGE DISCHARGE TO RIVER . . TO RIVER RRP-1A R8 RRP-18 EMERGENCY COOLING PUMPS
+ n 4 --_
RIVER WATER FIG. 2. RIVER WATER SUPPLY TO EFWS - TMI-l 1946 190
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APPENDIX A INSUFFICIENT FLCT FRCN EMERGENCY FEE 0fATER SYSTER O PREVENTIVE WAINTENANCE RELATED FAILURES A A EITHER WOTOR- TURBINE-DRIVEN PutP ORIVEN PUNP TRAIN FAILS
- TRAIN FAILS TUR81NE. SUCil0N OlSCHARGE FAILURE TO FAILURE OF ORivEN PUNP TURBINE-TURBINE.
FAILURE CRIVEN PUMP ORIVEN T9AIN SUCTION ;UCTION 40 TOR -0 RIVEN MOTOR-0 RIVEN DISCHAllGE FAILURE TO FAILURE TO PUNP A PUMP 8 FAILURE OF 40 TOR -0Ri v EN y0TCR 0 RIVEN FAILURE FAILURE TRilh5 PuuP 4 pggp g A-1 1946 194
l E l B G A R F A N A gi I H S A C E R S V T I L D A V i' l E l A G A R F N I A C E R S V T A H S l O A L V A l E l G A B R F A T N H S AI R C E S V I D A L V A N EF E GOV n I S A A ERN HRDI CU - A SLRR l I OT OAT FO M l E l G A A R F A A N H S I C E A S V R I L T D A V D E B S E 2 O l V L A i F C S E S S O E L R V I C L A A F V A D E 2 E l V S i F O A S E L S C O E R V S C L L 1 A AI T V F _ b& L
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A l DISCHARGE FAILURE OF TURBINE-DRIVEN TRAIN O t D (h
-- -~
CROSSTly TRAIN 8 TRAIN A yg{y{ { yg DISCHARGE DISCHARGE VALVE EFV23 FAILS CLOSED VALVES Fall VALVES Fall FAILS CLOSED A A A A h
3 4 e TRAIN A 8 OlSCHARGE VALVES Fall AIR-CPERATED VALVE EFV30A,, VALVE EFV308 FAILS EFV12A/ EFV128 TO OPEN e fD f) INTERLOCK MECHANICAL FAILURE dB CIRCUlf 6B ACTUATE VLV dB dB A-4 1946 197
5 6 CROSSTIE VALVE EFV2A EFV2B FAILS CLOSED [h ECHANICA FAILURE L I INGERTENTL'j % VALVE FAILS 14CTUATE VALVE I TO OPEN A 3 MANUAL { l 8 7 T T I I FAILURE l circuli l l OPERATOR I A-5
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A TURB INE - DRIVEN PUMP FAILURE
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STE l MECHANICAL i CONTROL l SUPPORTIVE DRIV l ' SY m MS TURBINE TRIP VALVE STOPS PUMP F3 1946 200 hs)ITWBIE kuh RESET A-7
R 00 O 1 ET S VA ', SRLR OI AE A V P L N E
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GPiti44 f aits L 9ChL gasual [ Opining I. fails Pest e a0 fos f ailget OMeaf0E f 4'L1 CDetagL PWII ccNIIR 'IM g;ggy,y f ailvat CI RCUlf 3Ptesfee
,gqg faits Faill A-9 1946 202
- sammmmmmmmmmmmmme
N E I N D I L N E L C A E R E 3P U Y nD O S L LE 'V S PA K SG E PVC ET R U O P S !P L. ES V L H MER L I G A EE A A I E KT V F RO G H T N OT N SA I T I A S,N R L , E EI F PA 0 OF N 0 I O ll T A II U U A A T y - C A N D B E E 3 P V T l C C L Q A V OEVD A R S O LR E E E T R USS P E LPO O E S T I E L V NAE L R L I F K A A I L ,GNS AI V F A I L C NI O E A L PF C R OSD T EA S
- I RAILE A ET L O
P C O L A E C R I U N A IL H A C F E N V L R0 V G OTEN T I I A SA N RL UE EI T P PACO F A N D A E E 3 P N T l G C O A V l A R S 0 lS E N 1 G A L P i UI O E s U T A V l A CF L A RI A Al A V F y K C t OE yE L R S RU$O E L pL T I g NA g I F g L LGS .- r A A NL CUI ONNAl
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A SUCTION FAILURE TO TURB I NE - DRIVEN PUMP O A A INSU FF ICI ENT INSUFFICIENT FLOW FROM FLOW FROM VALVE EFV18 VAiVE EFVIA MOTOR 0PERATED INSUFFICIENT MOTOR-OPERATED INSUFFICIENT VALVE EFV18 FLOW FROM VALVE EFVIA FLOW FROM FAILS CLOSED C A0 C A AND FAILS CLOSED 3. INSUFFICIENT FLOW FROM EMERGENCY RIVER WATER SOURCE MERGENC I EFV3 l l RIVER 44TER l 205 A-12
*l) lu!UF F ICl!Ilf FLOW FROS CST A 9 A#9
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FAILURE IN LINE FROM HOTIELL TO CST A MOTCR-OPERATED MOTCR OPERATE 0 VALVE COVl4A VALVE COV12 FAILS CLOSE0 FA LS TO n n f% f% MECHANICAL MECHANICAL T A FAILURE FAILURE CTUATEVLV} PENIN
, LOC l MANilAL LEFT OPENING %
CLOSED FAILS R VALVE FAILS 7 ACTUATE VLy TO OPEN l l l OPERATOR l FAILURE o l 4%circuli j b hC MANUAL { T TO CIRCULI FAILURE } l l l 0PERATOR I
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F AILURE I4 Ll4E FR05 N0ffELL TO C$f 8 Q v0f 0R -0Pt R ATE 0 gg7;g gpggAIES A lt-OPER Af t0 v4 Lyt Ccyl As TALVI CVOS VALVE Carl 3 I AILS CLOSED FAILg gtgggg F AILS 70 Q Q Q s(CM A 4 t C AL gnaggggt EECMatlCA, gpg gg FAILURE F AILURE FAILURE falls CLOSE0 GPERATOR WAtti ,gg,g FAILS TO N ihADWUrDCLY P hG CLOSED fI'f. ggg OPERAf0R W ALV( FAILS .P(Rif 04 VALVE FAILS FAILE TO FAILS TO ACTU Af( WLy to OPE 4 CTUATE vt 70 CPE4 CPEhlt OPEhlt O i LOCAL LOCAL gamLAL gaggat QPlu t kG C,AIL, pet t hG FA,<> 1946 208 P08f 8 C0hfROL sof0R PCTE R CONTROL uGTCR I CIRCUlf F A IL URE Ef", OPERAf0R FAllyRg OPERATOR falls F AIL 1 Fall 1 A-15
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Q b CEO adC E LM A-16 1946 209
A MOTOR-DRIVEN PUMPS IN PREVENTIVE MAINTENANCE F7 TURBINE-DRIVEN PUMP y '" TRAIN FAILS C .
.1 MOTOR-MOTOR D 'VEN DRIVEN P ?? A PUMP B IN P.M. IN P.M.
4 N
f /
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N E P R R P N V M I U M P N U O I T E PVE Ml C
- UiN E P N ;;
A N E N I NVE BE ET RVRN UI PI TR A DNM I R P S OM L T U I O P A A M F N R E N E V I H I A T R R I D T E PE N
- AO NW
A . DISCHARGE TRAIN A IN PREVENTIVE MAINTENANCE TURBINE- [\ VALVE DRIVEN PUMP TRAIN FAILS EFV30A P.M. Y G DISCHARGE SUCTl0N TURBINE. FAILURE OF FAILURE TO DRIVEN PUMP TURBINE. TURBINE DRIVEN PUMP DRIVEN TRAIN FAILURE
~
CROSSTIE TRAIN 8 A VALVE EFV28 OfSCHARGE FAILS CLOSED VALVES Fall N A 4 6 N
A . DISCHARGE TRAIN B IN PREVENTIVE MAINTENANCE O [\ VALVE TURBINE-DRIVEN PUMP EFV308 TRAIN FAILS P. .
?
E TURBINE- SUCTION DISCHARGE FAILURE TO FAILURE Of DRIVEN PUMP TURBINE-TURBINE-FAILURE DRIVEN PUMP DRIVEN TRAIN CROSSilE TRAIN A ~ VALVE EFV2A DISCHARGE < FAILS CLOSED VALVES Fall h ~
~ e
e A C0hCE 454 f t 1TCHASE fehl 414 PREVE4flVE sal 4fE44hCE
;$f 4 valvt C3tlCA GA
.gvl4A 14 E I TME R WO TOR .
CRivEm PuuP TUR9 l hE TRA14 Fell 3 DRIVE 4 PLMP TR414 fills 30fCs entvin sof 0# 04tVE4 Ftsp 4 015CMa#GE ' PetP g TURll4E - FaithAE CF SUC71C4 OtscwaaGE F & l L URE E070R ORivE4 gqivgg FalLURE TO FalLURE FAIWRE OF gggggg PygP FAILURE f uall4E - f uP 81 %E - gorygg egg, cRivre TRang
$UCT10m 54CTIC4 FalLU8E to Fattuat in votes.CRivt4 u0f04 C#lvt4 PLup 9 PtuP 4 l
l lhSVFFICIENT INSUFFICIENT FLCf FRCN ELOI IIOI t utLF F ICLE %T WALVE EFVit FLCS F ACt r ..!
~-
14 tit EFVla 1450FFICIE 41 FLCf FRCE C3f I A40
*0fsFL t 40TCA O'ER4 FED IkSUFFICIENT VALVE EFVII FLCs FaC5 FAILS CLOSEO E,NERGENCF RIVER 4fER SOURCE A-21 g4g 21k y
24 CowCin54 f t STGAAGE TAht I 14 PttetMilit ga l tfit & 4CE A C11 I univt CCfl08 OR C0fi48 in P e Tullint - Elf MER EQICA-aniytm PtsP 04lvlie PuuP f t414 F AIL 1 79614 falls Ol!CNARGE N0fDs.On vig 50 TOR DalvtM OISCHipGE fuRIltt - SUCTICM IdllUNI UI IA U #I II Ptup a Pts # 9 ORIVtm PURP Tut tist . 708 tint - BOTOR Clittu *e . vi g e._ge Onivia Taalt FalLURE F AILbpt F AILutt tanius SUCIl0'l $0Cfl04 F AILURE TO FaltuRE f0 u0f04 DRivtM uafga Otivtle INIUFF ICit%I Peup 3 14IUFF IClt%I Pter a 'LO' 0' C FLCS FRCE tuttGtwCT DivtR VALVE (FVl4 gays en, get
~ ~ ~
i41LF F ICi tti FLC8 FRCE vatyt (Frig 143UF F I Ci tNI U FLC1 FACE Cst A AND 40fftLL
#0T OR -OPla titD INSUF F ICitNT VAlyt EFyl4 FLCe FR0s EBERGENCY Riitt Fall 1 CLD510 etTFs $0tett A A A-22 1946
AFN ;; DIX B_ lH:C-SUPPL IED DT TA !GE" "0? PU"'>nHS OF ""r4CTillG A C0::PAU.T!E _ 5S:2:ii OI- *. I ST : ' AP.IS KS I S :t ! TIL.IP RTE *:TI,'L : :L SILITIES Point Vai va Es tir.e. .e
, of Probcuili ty o -'
Failure on E:rund __
- i. Comperst (uardware) Failure Data
- a. ValvrA:
lianucl Vrives (Plugged) s1 x.10 -4 Check Valves N1 x 10 f' l:otor Operated Valves
. Ibchanical Cor..penents s1 x 10,f3 Pluoging Contribution s1 x 10 '
Control Circuit (Local to Valve) w/ Quarterly Tests s6 x 10-3 w/f'onthly Tes ts s2 x 10-3
- b. Pug i: (1 Pump) ltechanical Cocpanents s1 x 10-3 Control Circui t
. w/ Quarterly Tests N7 x 10 7 w/iiontnly Tests s4 x 10-3
- c. Actuation Locic N7 x 10-3
' Error factors of 3-10 (up and down) about such values are not unexpected for basic data uncertainties. 1946 216 B-1
3 II. Human Acts & Errors - Failure Octa:
+ Estimated Human Error / Failure Prcbabilities +
+ Madi fying Factors & Situations --
With Local Walk-With Valve Position Around & Double Indication in Control Room Check Procedures w/o Either Point Est on Poi nt Est on Point Est vr. Value E rror Value E rro r Value E rror Es timate Factor Es tima te Factor Es tica te Factor A) Acts & Errors of a Pre- - Accident Nature
- 1. Valves mispositioned during test / maintenance.
a) Specific single 1 1 1 1 10 10-2 x1x 10 m valve wrongly selected 2U.x 10-2 xy 20 73.x 10-2 x7 N out of a population of valves during conduct af a test or maintenance act ("X" no. of valves in population at choice). b) Inadvertently leaves 45 x 10-4 20 s5 x 10-3 10 s10-2 10 correct valve in urong posi tion.
- 2. More than one valve is s1 x 10-4 20 s1 x 10 -3 10 %3 x 10-3 10 affected (coupled errors).
~
cts rs) ' 1
Appendix B II . Hu.r.a n Ac t s & E rro r: - Failure Da ta (Con t'd) .
+ Estimated liuman Error / Failure Probabilities -+
Estimated Failure Prob. for Primary Tir.c Actuation Opera tor to Actuate fleeded AP.lE Co . nonen ts B) Acts f. Errars of a Post-Accide:.t lla ture
- 1. Manual actuation of s5 min, s5x10j AfUS f rom Control N15 nin. s1 x 10 3 Rocm. Considering s30 min. s5 x 10-
"nen-dedicated" operator to actuate AGl5 and possible backup actuation of AFWS.
Ill . Iaintenanc;? Ou tare Contribution Maintenance outage for pumps and E"0V5: 3* 0.22 (dhours/raintenance act) g Maintecance 120 1946 218 B-3 i k}}