ML20041C277

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Snupps Auxiliary Feedwater System Reliability Study Evaluation
ML20041C277
Person / Time
Site: Wolf Creek, Callaway, 05000000
Issue date: 01/31/1982
From: Roscoe B
SANDIA NATIONAL LABORATORIES
To:
Office of Nuclear Reactor Regulation
References
CON-FIN-A-1303 NUREG-CR-2458, SAND81-2596, NUDOCS 8203010028
Download: ML20041C277 (45)
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{{#Wiki_filter:_- NUREG/CR-2458 SAND 81-2596 SNUPPS Auxiliary Feedwater System Reliability Study Evaluation Prepared by B. J. Roscoe Sandia National Laboratories Prepared for l U.S. Nuclear Regulatory

Commission l

l r i l ( l l D D KOO 2 A PDR i

t n 4 NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. 1 Available from GPO Sales Program Division of Technical Inform 6 tion and Document Control U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Printed copy price: 13.00 and National Technical Information Service Springfield, Virginia 22161

a 'f NUREG/CR-2458 SAND 81-2596 1 i SNUPPS Auxiliary Feedwater System Reliability Study Evaluation o Manuscript Completed: August 1981 Date Published: January 1982 Prepared by B. J. Roscoe Sandia National Laboratories Albuquerque, NM 87185 Prepared for Division of Safety Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC FIN A1303 l

A Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room,1717 H Street., N.W. Washington, DC 20555 2. The NRC/GPO Sales Program. U.S. Nuclear Regulatory Commission, Washington, DC 20555 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda, NRC Office of Inspection and Enforce-ment bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG series are available fcr purchase from the NRC/GPO Sales Pro-gram: fortnal NRC staff and contractor reports, NRC-sponsored conference proceed:ngs, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commis-sion, forerunner agency to the Nuclear Regulatory Commission. Documents avaliable from public and special technical libraries include all open literature items, such as books, journal and periodical articles, transactions, and codes and standards. Federal Register notices, federal and state legislat:on, and congressional reports can usually be obtained from these libraries. Documents such as theses, dissertations, foreign reports and translations, t nd non-NRC conference pro-ceedings are available for purchase from the organization sponsoring the publication cited. Single copies of NRC draf t reports are available free upon written request to the Division of Technical Infor-mation and Document Control. U.S. Nuclear Regulatory Commission, Washington, DC 20555. 1

1 ABSTRAC7 4 The purpose'of this report is to present the results of the review of the Auxiliary Feedwater System Reliability Analysis

  1. or SNUPPS.

4 J D T 1 2 1 i 't I 4 7 i f I I I I i i 4 1 i I i I t f 5 4 111 4 h- [ , ~

~ ' ACKNOWLEDGEMENT The author appreciates the review and comments on the draft { provided by Jack W. Hickman of Sandia National Laboratories. l This report has extracted freely from the referenced documents. B V

i l l t CONTENTS i Page iii Abstract c Acknowledgement v Ex 6' List of Figures Summary and Conclusions 1 1. Introduction 2 1.1 Scope and Level of Effort 2 1.2 Specific Reviews 2 2. AFWS Configuration 3

2.1 System Description

5 2.2 AFWS Support 7 2.2.1 Power Sources 7 2.2.2 Alternate Water Sources 8 2.2.3 Steam Availability 8 2.2.4 Instrumentation and Controls 9 2.2.5 Initiation Signals for Automatic Operations 10 2.2.6 Testing 10 2.2.7 Technical Specifications 11 3. Discussion 33 T4 3.1 Mode of AFWS Initiation 13 3.2 System Control Following Initiation 13 3.3 Test and Maintenance Procedures Unavailability 13 3.4 Adequacy of Emercency Procedures 13 3.5 Adequacy of Power Sources and Separation of 14 Power Sources vii

CONTsNTS (Cont'd) Page 3.6 Availability of Alternate Water Sources -14 3.7 Potential Common Mode Failure 14 3.8 Application of Data Presented in NUREG-0611 14 3.9 Search for Single Failure Points 15 3.10 Human Factors / Errors 15 3.11 NUREG-0611 Recommendations Long and Short-Term 15 3.11.1 Short-Term Generic Recommendations 15 3.11.2 Additional Short-Term Recommendations 19 3.11.3 Long-Term Generic Recommendations 21 4. Major Contributors to Unreliability 23 5. Conclusions 24 6. Glossary of Terms 26 7. References 28 5 i i { viii

l i 5,IST OF FIGORES-Page 1. Auxiliary Feedwater System Simplified Flow 4 Diagram 2.- Comoarison of SNUPPS AFWS Reliability to 25 3-Other AFWS Designs in Plants Using the Westinghous NSSS l IX I 1 l 1 )

SUMMARY

AND CON'LUSIONS C The accident at Three Mile Island (TMI) resulted in many studies which outlined the events leading to the accident as well as those following. One of the important safety systems involved in the mitigation of such accidents was determined to be the Auxiliary Feedwater System (APWS). Each operating plant's Auxiliary Feedwater System was studied and analyzed. The results were reported in NUREG-0611.1 The licensee of each nonoperating plant was instructed 2 to perform a reliability analysis of his Auxiliary Feedwater System for three transient conditions involv-ing loss of main feedwater in a manner similar to the study made by NUREG-0611. The Union Electric Corporation (UEC), the licensee for two SNUPPS* units at Callaway, Missouri and the Kansas Gas and Electric Company (KG&E Co.) and Kansas City Power and Light (KCP&L), the licensees for the SNUPPS unit at Wolf Creek, Kansas, submitted a reliability report 3 to the U. S. Nuclear Regulatory Commission in June 1981. This report was reviewed by Sandia National Laboratories (SNL). The following conclusions resulted from the review: 1. UEC, KG&E Co. and KCP&L have satisfactorily complied with the requirement to make a relability study of their AFWS. 2. The AFWS of the SNUPPS has high reliability relative to the reliability of AFWSs of operating plants for Case I, Loss of Main Feedwater. Quantitatively, the unavailabil-ity of the system for this event is approximately 2 x 10-5 per demand. Qualitatively, the system is automat-ically initiated, highly redundant, has no observed sin-gle point vulnerabilities, and is tested periodically and following realignment to demonstrate availability of flow path to the steam generators. Failure on demand is dominated by triple failure events that involve pumps, test and maintenance, and the turbine driven pump (TDP) throttle valve or speed control valve. The unavailability the Case 2, Loss of Main Feedwater and Loss of Offsite Power, is 5.3 x 10-5 per demand, which places reliability in the high range. Failure on demand is dominated by triple failure events that involve the pumps, test and maintenance and the diesel generators. The unavailability for Case 3, Loss of Main Feedwater and Loss of all AC Power, is 1.4 x 10-2, which places the reliability in the medium range. The turbine-driven pump train has no identifiable ac power dependencies and is automatically actuated. Failure on demand is dominated by test and maintenance outage, failure of the TDP, and faults in the throttle valve.

  • SNUPPS - Standardized Nuclear Unit Power Plant System -_

1. Introduction The results of many studies pertaining to the TMI Nuclear Power Plant accident conclude that a proper functioning Auxiliary Feedwater System is of prime importance is the mitigation of such kinds of accidents. Therefore, a letter dated March 10, 19802 stating U. S. Nuclear Regulatory Commission (NRC) requirements o regarding the AFWS was sent to all operating license applicants with a Nuclear Steam Supply System (NSSS) designed by Westinghouse or Combustion Engineering. UEC, KG&E Co., and KCP&L, the applicants for operating licenses for SNUPPS plants which have Westinghouse-designed NSSS. provided a response in June 1981, in the form of a relia-bility analysis 3 which was prepared for them by Bechtel Power Corporation. The analysis was received by SNL on June 30, 1981. The analysis addresses the potential for failure of the AFWS to supply sufficient flow to three of four steam generators.5 The method utilizes a simplified fault tree approach. It takes into account component failure, outage due to test and maintenance, and human errors. The importance of certain failure modes are examined as part of the study. 1.1 Scope and Level of Effort This project undertakes a review of those portions of the reliability analysis which (1) satisfy requirement (b) of the letter 2 which states, " perform a reliability evaluaton similar in method to that described in Enclosure 1 (NUREG-0611) that was per-formed for operating plants and submit it for staff review," and (2) provide answers to the short-and long-term recommendations of NUREG-0Gli in response to requirement (c) in the letter. The 4 which was sub-review was conducted according to a Schedule 189 mitted by SNL to NRC. 1.2 Specific Review SNL reviewed the reliability analysis 3 submitted by UEC, KG&E Co., and KCP&L. Particular attention was directed toward determining that the analysis addressed in depth the reliability of the AFWS when subjected to three transient cases: (1) Loss of Main Feedwater, LMF, (2) Loss of Main Feedwater/ Loss of Offsite Power, LMF/LOSP, and (3) Loss of Main Feedwater/ Loss of all AC l Power, LMF/ LAC. Also, the methods used in NUREG-0611 were compared j to those used in the analysis. The specific findings are presented in Sections 3, 4, and 5. 2. AFWS System Configuration 5 consists'of two motor-driven pumps (MDP), one The AFWS otcam turbine-driven pump, and associated piping, valves, instru-msnts, and controls as shown on Figure 1. Each motor-driven auxiliary feedwater pump will supply 100 psrcent of the feedwater flow required for removal of decay heat from the reactor. The turbine-driven pump is sized to supply up to twice the capacity of a motor-driven pump.. Normal flow is from the condensate storage tank (CST) to the auxiliary feedwater pumps. Two redundant safety-related back-up cources of water from the essential service water system (BSWS) are provided for the pumps. In order to remove decay heat by the steam generators, auxil-iary feedwater must be supplied to.the steam generators (SG) in the event that the normal source of feedwater is lost. The minmum required flow rate is 470 gpm. Two auxiliary feedwater pumps are driven by ac-powered elec-tric motors supplied with power from independent Class 1E switch-gear busses. Each horizontal centrifugal pump takes suction from the condensate storage tank, or alternatively, from the ESWS. Pump design capacity includes continuoas minimum flow recircula-tion, which is controlled by restriction orfices. The turbine-driven pump provides system redundancy of auxiliary feedwater cupply and diversity of motive pumping power. The pump is a hori-zontal centrifugal unit. Pump design capacity includes continuous minimum flow recirculation. Power for all controls, valve opera-tors, and other support systems is independent of ac power sources. Steam supply piping to the turbine driver is taken from two of the four main steam lines between the containment penetrations and the main steam isolation valves. The piping from the ESWS to the suction of each of the auxili-ary feedwater pumps is equipped with a motor-operated butterfly valve, an isolation valve, and a nonreturn valve. Each line from the condensate storage tank is equipped with a motor-operated gate j vclve and a nonreturn valve. Each motor-driven pump discharges through a nonreturn valve and a locked-open isolation valve to focd two steam generators through individual sets of a locked-opsn isolation valve, a normally open motor-operated control valve, a check valve followed by a flow restriction orifice, and and a locked-open globe valve. 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2.1 System Description

The Auxiliary Feedwater System (AFWS) is a reliable source of water for the steam generators. The AFWS, in conjunction with safety valves in the main steam lines, is a safety-related system, the function of which is to remove thermal energy from the reactor coolant system by releasing secondary steam to the atmosphere. The AFWS also provides emergency water following a secondary side line rupture. Removal of heat in this manner prevents the reactor coolant pressure from increasing and causing release of reactor coolant through the pressurit.er relief and/or safety valves. The auxiliary feedwater system may also be used following a reactor shutdown in conjunction with the condenser dump-valves or atmospheric relief valves, to cool the reactor coolant system to 350*F and 400 psig, at which temperature the residual heat removal system is brought into operation. The system consists of two motor-driven pumps, one steam turbine-driven pump, and associate piping, valves, instruments, and controls. Each motor-driven auxiliary feedwater pump will supply 100 percent of the feedwater flow required for removal of decay heat from the reactor. The turbine-driven pump is sized to supply up to twice the capacity of a motor-driven pump. This capacity is suffi-cient to remove decay heat and to provide adequate feedwater for cooldown of the reactor coolant system at 50*F/hr within 1 hour of a reactor trip from full power. Normal flow is from the CST to the auxiliary feedwater pumps. Two redundant safety-related back-up sources of water from the ESWS are provided for the pumps. The condensate storage tank capacity allows the plant to remain at hot standby for 4 hours and then cool down the primary system at an average rate of 50*F per hour to a temperature of 350*F. Initially, sensible heat is removed from the reactor coolant system to reduce the temperature from a full power operation average temperature of 588'F to a nominal hot shutdown temperature of 500*F. Subse-quently, to bring the reactor down to 350*F at 50*F/hr, an initial makeup rate of 500 gpm is required. In order to remove decay heat by the steam generators, auxili-ary feedwater must be supplied to the steam generators in the event that the normal source of feedwater is lost. The minimum required flow rate is 470 gpm...._

- Provisions are in'corporated in the AFWs design to allow for periodic operation to demonst' rate performance and structural and leaktight integrity. Leak detection is provided by visual examina-tion and in the floor drain system. Two auxiliary feedwater pumps are driven by ac-powered electric motors supplied with power from independent Class lE switcIplear busses. Each horizontal centrifugal pump takes suction fros' the c condensate storage tank, or alternatively, from the ESWS. i$mp design capacity includes continuous minimum flow recirculatia, n which is controlled by restriction orifices. A tutbine-driven pump provides system redundancy of auxiliary feedwater supply and diversity of motive pumping power. The pump is a horizontal centrifugal unit. Pump bearings are cooled by the pumped fluid. Pump design capacity includes continuous minimum l flow recirculation. Power for all controls, valve operators, and other support systems is independent of ac power sources. Steam supply piping to the turbine driver is taken from two of the four main steam lines between the containment penetrations and the main steam isolation valves. Each of the steam supply lines to the turbine is equipped with a locked-open gate valve, normally closed air-operated globe valve with air-operated globe bypass to keep the line warm and two nonreturn valves. Air-operated globe valves are equipped with de-powered solenoid valves. These steam supply lines join to form a header which leads to the turbine through a normally closed, de motor-operated mechanical trip and throttle valve. All piping in the AFWS is seamless carbon steel. Welded joints are used throughout the system, except for flanged connections at l the pumps. The AFWS is not required during normal power generation. The pumps are placed in the automatic mode, lined up with the condensate storage tank, and are available if needed. In addition to remote manual-actuation capabilites, the AFWS is aligned to be placed into service automatically in the event of an i emergency. The common water supply header from the condensate storage tank contains a locked-open, twelve-inch, butterfly isolation valve. Correct valve position is verified by periodic surveillance. In case of failure of the water supply from the condensate storage tank, the normally closed, motor operated butterfly valves from l the ESWS are automatically opened on low suction header pressure. Valve opening time and pump start time are coordinated to assure l adequate suction pressure with either onsite or offsite power available. l If a motor-driven pump supplying two of the three intact steam generators fails to function, the turbine-driven pump will automat-ically start when a low-low level is reached in two of the four l t

l steam generators. During all of the above emergency conditions, the normally open control valves are remote manually operated. During all of the above emergency conditions, the normally open motor-driven pump control val /es are automatically operated to limit runout flow under all secondary side pressure conditions. This is required to prevent pump suction cavitation at high flow rates. The turbine-driven pump design includes a lower net positive suction head (NPSH) requirement. Therefore, the turbine-driven pump control valves are remote manually operated. Low pump discharge pressure alarms will alert the operator to a secondary side break. The operator will then determine which loop is broken by observing high auxiliary feedwater flow, using control room flow indication, and close the appropriate discharge control valve. This can be accomplished within 10 minutes after pump start. The AFWS is located in the auxiliary building. This building is designed to withstand the effects of earthquakes, tornadoes, hurri-canes, floods, external missiles, and other appropriate natural phenomena. In addition, the AFWS is designed to remain functional after a safe shutdown earthquake (SSE). Complete redudancy is provided and no single failure will compromise the system's safety functions. All vital power can be supplied from either onsite or offsite power systems. 2.2 AFWS System Support 2.2.1 Power Sources Each SNUPPS unit is provided with a Class lE ac and de power system. The Class lE ac system distributes power at 4.16 kV, 480 V, and 120 V to all safety-related loads. Each Class lE 4.16 kV load group (two in each unit) is supplied by a separate preferred power supply feeder and one diesel generator (standby) supply feeder Each 4.16 kV bus supplies motor loads and 4.0 kV/480 V load center transformers with their associated 480 V buses. The Class lE ac system is divided into two redundant load groups per unit (load groups 1 and 2). For each unit, each ac load group consists of a 4.16 kV bus, two 480 V load centers, 480 V motor control center, and lower voltage ac supplies. No provisions exist for automati-cally connecting one Class lE load group to another redundant Class lE load group or automatically transferring load between load groups. The Class lE switchgear, load centers, and motor control centers ( for the redundant load groups are located in separate rooms of the i control building to ensure physical separation. l l ~

The de control supplies for switchgear breaker operation are separate and independent so that Class lE de load group 1 supplies Class lE load group 1 switchgear. The battery chargers for de load j group 1 are fed from the same load group switchgear. Class lE dc load group 2 supplies Class lE ac load group 2 switchgear. The standby power supply for each safety-related load group q consists of one diesel generator complete with its accessories and i fuel storage and transfer systems. One diesel generator is connected exclusively to a single 4.16 kV safety feature bus of a load group. The diesel generators are housed in separate rooms of a seismic Category I structure which ensures physical separation for fire and missile protection. Power and control cables for the diesel gen-erators associated switchgear are routed to maintain physical l separation. Four independent Class lE 120 V ac vital instrument power sup-plies are provided to supply the four channels of protection systems and reactor control systems. Each vital instrument ac power supply l consists of one inverter, one distribution bus, and one manual transfer switch. Each inverter is supplied by a Class lE battery system. The Class lE de system provides de electric power to the Class lE de loads and for control and switching of the Class lE systems. Physical separation, electrical isolation, and redundancy are pro-vided to comply with the requirements of IEEE-308. Subsystems 1 and 4 provide control power for ac Load Groups 1 and 2, respectively. Each Class lE de power subsytem consists of one 125 V battery, one battery charger, one inverter, and distribution switchboards. The Class lE batteries, chargers, and de switch gear of each separate group are located in separate rooms of the seismic Cate-gory I control building. Chargers and de switchgear are in sepa-i I rate rooms from the batteries. 2.2.2. Alternate Water Sources Normally, water to the AFWS pumps is supplied from the CST. However, two redundant safety-related backup sources of water from the ESWS are provided. Should the CST water supply to the pump suction be disrupted, the system will automatically switch from the CST to the ESWS on low pump suction pressure. 2.2.3 Steam Availability i Steam supply piping to the turbine driver is taken from two of l the four main steam lines between the containment penetrations and the main steam isolation valves. The steam lines contain provisions to prevent the accumulation of condensate. The turbine driver can operate with steam inlet pressures ranging from 100 to 1,250 !

psia. Exhaust steam f, rom t,he turbine driver is vented to the ctmosphere above the auxiliary buildin'g roof. 2.2.4 Instrumentation and Controls The AFWS instrumentation is designed to facilitate automatic operation and remote control of the system and to provide contin-3 uous indication of system parameters. Redundant condensate storage tank level indication and alarms are provided in the control room. The backup indication and alarm use auxiliary feedwater pump suction pressure by converting it to available tank level. Both alarms provide at least 20 minutes for operator action (e.g., refill the tank) assuming the largest capac-ity auxiliary feedwater pump is operating. Pressure transmitters are provided in the discharge and suction lines of the auxiliary feedwater pumps. Auxiliary feedwater flow to each steam generator is indicated by flow indicators provided in the control room. If the condensate supply from the storage tank fails, the resulting reduction of pressure at the pump suction is indicated in the control room. Flow transmitters and control valves with remote control etations are provided on the auxiliary feedwater lines to each eteam generator to indicate and allow control of flow at the auxili-l ary shutdown panel and in the control room. Flow controllers for j the motor-driven pump control valves position the valves to limit l the flow to a preset value through the full range of downstream l operating pressures. l l l In addition to control room instrumentation showing the water I level and pressure in individual steam generators, the operator is i provided with the following information relating to the auxiliary feedwater system: Control Room Indication / Control Control Room Locali Alarm Condensate storage tank suction MOV valve position X X ESWS suction MOV valve position X X i Condensate storage tank level X X X Condensate storage tank suction header pressure X Low pump suction pressure X X X Low pump discharge pressure X X X Pump flow control valve operation X X Pump flow control valve position X X Auxiliary feedwater flow X X tLocal Means Auxiliary Shutdown Panel - _ _ _ - _ _ _ _ _ _

2.2.5 Initiation Signals.for Automatic Operation The AFWS is designed for automatic actuation in the event of an emergency. Any one of the following conditions will automatically start both motor-driven pumps: A. Two out of four low-low level signals in any one steam generator. d B. Trip of both main feedwater pumps. C. Safeguards sequence signal (initiated by safety injection signal. D. Class lE bus loss of voltage sequence signal (i.e., loss of offsite power). The turbine-driven pump is automatically actuated on either of the following signal s : A. Two out of four low-low level signals in any two steam generators. B. Under voltage conditions on any two out of four reactor coolant pump feeder potential transformer cubicles. Additionally, the AFWS is capable of remote-manual actuation. In case of failure of the CST water supply, the normally closed, motor-operated butterfly valves from the ESWS are automatically opened on low suction header pressure. 2.2.6 Testing The AFWS is testable through the full operational sequence that brings the system into operation for reactor shutdown and for design basis accidents (DBA), including operation of applicable portions of the protection system and the transfer between n.ormal and standby power sources. The safety-related components, i.e., pumps, valves, piping, and turbine, are designed and located to permit preservice and inser-vice inspection. Testing and maintenance activities which remove components and/or systems from service can be significant contributors to the overall AFWS unavailability. The most common forms of valve main-tenance performed during power operation are packing adjustments and repairs to the motor-operated valve (MOV) and air-operated valve (AOV) control circuits and operators. Nearly all these maintenance activities are performed with the valve in the failsafe position during the maintenance interval. Therefore, maintenance of MOVs and AOVs was not considered to be a contributor to valve unavailability. Check valves and manual valves are expected to require very little maintenance. The low test and maintenance impact on this part of the AFWS was the basis for not including a human error contributor to unavailability for the manual valves in the individual SG flowpaths (i.e., fore and aft of the MOVs and AOVs). Although test and maintenance contributions were not treated for the valves associated with the branch flowpaths to a specific Steam generator, the unavailability due to testing and maintenance of the pump subsystems was treated. Additionally, it was assumed that coincident test and/or main-tenance of components of more than one AFWS pump and its associated flow paths, while the reactor is at full power, is in violation of ] the plant technical specifications; therefore, minimum cut sets containing such coincident basic events are treated as not being credible and are discarded in the quantitative evaluation. 2.2.7 Technical Specifications A review of the Standard Technical Specifications indicates that for power, start-up, or hot standby plant status, the limiting condition of the AFWS for plant operation include: At least three independent auxiliary feedwater pumps and associated flow paths shall be operable with: a. Two motor-driven AFWS pumps, each capable of being powered from separate emergency busses. I b. One turbine-driven AFW pump capable of being powered from an operable steam supply system. Required action is: a. With one auxiliary feedwater pump inoperable, restore at least three AFW pumps (two capable of being powered from separate emergency busses and one capable of being powered by an operable steam supply system) to an OPER-ABLE status within 72 hours or be in at least HOT STANDBY condition within the next 6 hours'and in a HOT SHUTDOWN within the following 6 hours. b. With two auxiliary feedwater pumps inoperable, be in at leact HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. c. With three auxiliary feedwater pumps inoperable,imme-diately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. __-__ -__

To satisfy specified surveillance requirements, each auxiliary feedwater pump shall be demonstrated OPERABLE: 1. At least once per 31 days by: a. Verifying that each motor-driven pump develops a dis-charge pressure of greater than or equal to (later)* psig at a: flow of greater than or equal to (later) gpm. c b. Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to (later) psig at a flow of greater than or equal to (later) gpm when the secondary steam supply pressure is greater than (later) psig. c. Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. l d. Verifying that each automatic valve in the flow path is in the full open position whenever the auxiliary feed-water system is placed in automatic control or when above 10 percent of RATED THERMAL POWER. 2. At least once per 18 months during shutdown by: a. Verifying that each motor-driven pump starts automatically upon receipt of each of the following test signals: 1. Loss of both main feedwater pumps. 2. Safety injection signal. 3. Two out of four low-low level signals in any one steam generator. 4. Class lE bus loss of voltage sequence signal. b. Verifying that the steam turbine-driven pump starts automatically upon receipt of each of the following i test signals: I 1. Undervoltage conditions on any two out of four reactor coolant pump feeder potential transformer cubicles. 2. Two out of four low-low level signals in any two steam generators. c. Verifying that the valve in the suction line of each auxiliary feedwater pump from the ESWS automatically _ actuates to its full open position on a low suction pressure test signal.

  • (later) signifies that the numerical values have not yet been established..

3. Discussion 3.1 Mode of AFWS Initiation The AFWS is initiated automatically. The MDPs will start an (1) two out of four low-low water level signals in any steam gener-ator, (2) loss of both main feedwater pumps, (3) initiation of a cafety injection signal, or (4) loss of offsite power. The TDP starts on the generation of two out of four low-low water level cignals in any two of four steam generators or upon loss of offsite power. The automatic starting of any AFWS pump initiates automatic isolation of CST from all its other users. In the event of low CST level, make up water is automatically supplied. 3.2 System Control Following Initiation After initiation proper flow is established by adjusting the MDP discharge control valves and/or adjusting the TDP speed or discharge control valves. When the reactor coolant average temper-ature is reduced to 350*F the residual heat removal system (RHRS) is placed into service and the AFWS taken out of service. 3.3 Test and Maintenance Procedures'and Unavailability The technical specifications require that all valves be given in service tests and inspections in accordance with the ASME Boiler and Pressure Vessel Code (Section XI and applicable Addenda) for Safety Class 1, 2, and 3 components. Als'o every 31 days there are (1) pump discharge pressure and flow tests (2) non-automatic valve position verification test and (3) automatic valve position verifi-fication when the AFWS system is in automatic control. The pumps and system are available on demand during all tests. During shut-down the automatic starting of each pump and the functioning of the automa' tic valves from closed to full open in the suction line of each AFW pump from the ESWS are checked. There are no coincident tests or maintenance of components wit'..in the AFWS. The above

procedures are standard for Westinghouse designed units.

These -procedures were used as input to the reliability analysis. The procedures have been written for SNUPPS and are currently in the process of being approved by the applicants. 3.4. Adequacy of Emergency Procedures The Emergency Procedures were not reviewed or included in the analysis by Bechtel. Emerhency operation was discussed as part of the SNL review and it was assumed that the emergency procedures would be written to implement the emergency operations. The SNUPPS license and safety office stated that it was committed to write procedures to cover operator verification of pump suction alignment to the service water.

3.5 Adequacy of Power Sources and Separation of Power Sources The motor-driven pumps, associated motor-operated valves and other electrical equipment receive power from two identical but separate 4160 V emergency busses. One bus, "A", supplies one pump and "B" the other. In the event of loss of offsite power the two diesel generators each supply one bus in a like manner. The TDP is supplied with steam from two steam generators. The TDP is not c dependent upon offsite or diesel-generated ac power. Redundant power sources enhance system reliability as does the separation of these power sources which eliminates many common cause failure events. 3.6 Availability of Alternate Water Sources For water of steam generator quality the preferred source is the Nuclear Safety Class 3 Condensate Storage Tank. The primary alternate water source is the Essential Service Water System which is safety grade but not of steam generator quality. This source is available by way of automatic controlled valves. Switchover to the ESWS is fast enough to prevent pump failure because of no water sup-ply at the pump intake. 3.7 Potential Common Mode Failure A common mode, or more generally common cause, failure is a group of component failures, with or without the same failure mode, that are the direct result of the same event, cause or condition and that leads directly to a specific system failure. The evaluation of common cause failures was approached by Bechtel in two phases. In the first phase those areas which historically have been associated with common cause failures, e.g., diesel generators and common location of components belonging to different trains, were examined. In addition to examining piping and instrumentation drawings and other drawings, extensive examination was made of the SNUPPS model (a completely detailed one-sixteenth scale model). It was concluded that human error was the only significant potential common cause and that other common causes, such as grit, temperature, manufacture, vibration, etc., would be insignificant in comparison. In the second phase, an attempt was made to quantify the effect and importance of human error induced common cause failures. 3.8 Application of Data Presented in NUREG-0611 l 3 l The report contains Appendix B which includes the failure data or basic event probabilities. The fault tree was checked and all applicable components as shown in Figure 1 were properly included. Although Appendix B includes the data in NUREG-0611 in the table of basic events, there is also data included from NUREG/ CR-1362 on diesel generators, NUREG/CR-1363 on control circuits,.

and NUREG/CR-1464 on loss of offsite power. These data were used to quantify the fault tree. 3.9 Search for Single Failure Points l There were no single failure points (SFP) associated with Case 1, LMF, or Case 2, LMF/LOSP. For Case 3, LMF/ LAC, there were many SFPs since Case 3 describes a single channel system. Any SFP has a major effect on the reliability of a redundant system and if i any are found, they should be thoroughly reviewed. 3.10 Human Factors / Errors Human Factors / Errors were considered by Bechtel where appropri-ate in the fault tree. Test and maintenance outages were found to be important contributors to system unavailability, but in combina-tion with pump and pump-driven failures. 3.11 NUREG-0611 Recommendations, Long-and Short-Term 3.11.1 Short-Term Generic Recommendations I. Technical Specification Time Limit on AFWS Train Outage Recommendation GS-1 The licensee should propose modifications to the Technical Specifications to limit the time that one AFWS pump and its associated flow train and essential instrumentation can be inoperable. The outage time limit and subsequent action time should be as required in current Standard Technical Specifications; i.e., 72 i hours and 12 hours, respectively.

Response

The limiting conditions for operation related to the auxilary feedwater system will be addressed in the proposed Technical Specifications for the SNUPPS plants. The proposed Technical Specifications will be submitted l approximately one year before the scheduled fuel load for the first SNUPPS unit and will be based on NUREG-0452, Rev. 3, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors." 1._

l i t II. Technical Specification Administrative Controls on l Manual Valves-Lock and Verify Position Recommendation GS-2 l The licensee should lock open single valves or multiple valves in series in the AFWS pump suction piping and lock open other single valves or multiple valves in series that could interrrupt all AFWS flow. Monthly inspections should be performed to verify that these valves are locked and in the open position. These inspections should be proposed for incorpora-tion into the surveillance requirements of the plant Technical Specifications. See Recommendation GL-2 for the longer term resolution of this concern.

Response

This item is not applicable to SNUPPS because the design does not include single valves or multiple valves in series that could interrupt auxiliary feedwater pump suction or all auxiliary feedwater flow. III. AFWS Flow Throttling-Water Hammer Recommendation GS-3 The licensee should reexamine the practice of throttling AFWS flow to avoid water hammer. The licensee should verify that the AFWS will supply on demand sufficient initial flow to the neces-sary steam generators to ascure adequate decay heat removal following loss of main feedwater flow and a reactor trip from 100% power. In cases where this reevaluation results in an increase in initial AFWS flow, the licensee should provide sufficient infor-mation to demonstrate that the required initial AFWS flow will not result in plant damage due to water hammer.

Response

Throttling auxiliary feedwater flow to avoid water hammer will not be utilized. The system design pre-cludes the occurrence of water hammer in the steam i generator inlet as described in Section 10.4.7.2.1 of I the SNUPPS FSAR. -

~ l IV. Emergency Procedures for initiating Backup Water ~ [ Supplies Recommendation GS-4 Emergency procedures for transferring to alternate sources of AFWS should be available to the plant l operators. These procedures should include criteria to inform the operators when, and in what order, the transfer to alternate water sources should take place.

Response

4 The SNUPPS design includes an automatic transfer to the alternate sources of supply. Procedures will pro-vide guidance to the operator concerning alternate water sources. The normal supply from the condensate storage tank (CST) is through a locked-open, butterfly valve. Peri-odic surveillance will verify valve position. Opening of valves from the backup ESWS and starting of auxil-iary foedwater pumps are timed such that an AFWS start with no suction from the CST is not a mode for common failure of all auxiliary feedwater pumps. V. Emergency Procedures for Initiating AFWS Flow Following a Complete Loss of Alternating Current Power Recommendation GS-5 The as-built plant should be capable of providing the required AFWS flow for at least two hours from one AFWS pump train, independent of any ac power source.

Response

The turbine-driven pump in the SNUPPS design is capable of being automatically initiated and operated independent of any alternating current power source for at least two hours. Essential controls, valves opera-tors, other supporting systems, and turbine lube oil cooling for the turbine-driven pump are all indepen-dent of alternating current power. VI. AFWS Flow Path Verification Recommendation GS-6 l The licensee should confirm flow path availability I of an AFWS system flow train that has been out of ser-l vice to perform periodic testing or maintenance as follows: l ~ (1) Procedures should be implemented to require an operator to determine that the 'FWS valves are properly aligned and a..econd oper-ator to independently verify that the valves are properly aligned. (2) The licensee should propose Technical Specifi-cations to assure that, prior to plant startup following an extended cold shutdown, a flow test would be performed to verify the normal flow path from the primary AFWS system water source to the steam generators. The flow test should be conducted with AFWS valvos in their normal alignment.

Response

(1) Valve lineups and independent second operator verification of valve lineups will be required on the auxiliary feedwater system after main-tenance. Verification of operability will be included as part of functional testing on return from extended cold shutdown. (2) SNUPPS has the latest version of the Standard Technical Specifications which provide ade-quate assurance of the operability of the auxiliary feedwater system. VII. Non-Safety Grade, Non-Redundant AFW System Automatic Initiation Signals Recommendation GS-7 The licensee should verify that the automatic start AFWS signals and associated circuitry are safety-grade.

Response

The SNUPPS AFWS is designed so that automatic initiation signals and circuits are redundant and meet safety-grade requirements. VIII. Automatic Initiation of AFWS Recommendation GS-8 The licensee should install a system to automatically initiate AFWS flow. l l

Response

See response to Reconmendation GS-7. l l l I 3.11.2 Additional Short-Term Recommendati'ons I. Primary AFWS Water Source Low Level Alarm Recommendation The licensee should provdic redimdant levet indica-tion and low level alarms in the contrcl room for the AFWS primary water supply, to allow the operator to anticipate the need to make up water or transfer to an alternative water supply and prevent a low pump suc-tion pressure condition from occurring. The low level alarm setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity AFWS pump is operating.

Response

The existing SNUPPS design provides the following redundant control room indication for condensate storage tank level.* a) LI-4A shown on Figure 9.2-12. b) P1-24A, Pl-25A of P1-26A-Class lE auxiliary feedwater pump suction pressure indication shown on Figure 10.4-9. Direct correlation between pump suction pressure and tank level is achieved by simple conversion. Exclusion of dynamic piping losses from the conversion results in a conservative determination of tank level. Redundant control room tank level alarms are as follows: a) LAHL-7 shown on Figure 9.2-12. b) LAL-24-Class lE auxiliary feedwater pump low suction pressure alarm shown on Figure 10.4-9. Setpoints for both alarms will allow at least 20 minutes for operator action assuming that the largest capacity auxiliary feedwater pump is operating. II. AFWS Pump Endurance Test I Recommendation The licensee should perform a 48 hour endurance j test on all AFWS pumps, if such a test or continu-ous period of operation has not been accomplished tx> date. Following the 48 hour pump run, the punp should

  • See SNUPPS FSAR, Revision 3, dated 4/81..-

be shut down and ' cooled down'and ' hen restarted and t run for one hour. Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing / bearing oil l temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed envi-ronmental qualification limits for safety-related equipment in the room. l l

Response

l l SNUPPS will perform a 48-hour, in situ endurance l test on all auxiliary feedwater pumps as part of the pre-operational test program. l III. Indication of AFWS Flow to the Steam Generators Recommendation l The licensee should implement the following requirements as specified by Item 2.1.7.b on page A-32 of NUREG-0578: (1) Safety-grade indication of AFWS flow to each steam generator should be provided in the control room. (2) The AFWS flow instrument channels should be powered from the emergency busses consistent with satisfying the emergency power diversity requirements for the AFWS set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9. l

Response

The SNUPPS auxiliary feedwater design provides i safety-grade (Class lE) indication in the control room l of auxiliary feedwater flow to each steam generator. l The design utilizes four independent Class lE power j supplies. The safety-grade steam generator level l indication provides a backup method for determining the auxiliary feedwater flow to each steam generator. IV. AFWS Availability During Periodic Surveillance Testing Recommendation I Licensees with plants which require local manual realignment of valves to conduct periodic tests on one AFWS train and which have only one remaining AFWS train available for operation should propose Technical Specifications to provide that a dedicated individual who is in communication with the control room be sta-tioned at the manual valves. Upon instruction from the - - - --

I ~ controi room, this op'erator would realign the valves in the AF'W system from the test mode to its operational alignment.

Response

This recommendation is not applicable to the SNUPPS design. 3.11.3 Long-Term Generic Recommendations I. Automatic Initiation of AFWS Recommendation GL-1 For plants with a manual starting AFWS, the licensee should install a system to automatically ini-tiate the AFWS flow. This system and associated automatic initiation signals should be designed and installed to meet safety-grade requirements. Manual AFWS start and. control capability should be retained with manual start serving as backup to automatic AFWS l initiation.

Response

See response to Recommendation GS-7. II. Single Valves in the AFWS Flow Path Licensees with plant designs in which all (primary and alternate) water supplies to the AFWS pass through i valves in a single flow path should install redundant i parallel flow paths (piping and valves). l

Response

The alternate water supply (essential service water) connects to the auxiliary feedwater pump suction piping downstream of the single, normally locked-open valve in a single flow path from the. primary water source (condensate storage tank). Valves from the alternate supply automatically open on low pump suction pressure. l Refer to the response to GS-2 and GS-4. I III. Elimination of AFWS Dependency on Alternating Current Power Following a Complete Loss of Alternating Current Power I Recommendation GL-3 At least one AFWS pump and its associated flow path and essential. instrumentation should automatically !

~ initiate AFWS flow *and be capable of being operated independently of any ac power source for at least two hours. Conversion of de power to ac power is acceptable.

Response

See Response to recommendation GS-5. IV. Prevention of Multiple Pump Damage Due to Loss of Suction Resulting from Natural Phenomena Recommendation GL-4 Licensees having plants with unprotected normal AFWS water supplies should evaluate the design of their AFWS determine if automatic protection of the pumps is necessary following a seismic event or a tornado. The time available before pump damage, the alarms and indications available to the control room operator, and the time necessary for assessing the probblem and taking action should be considered in determining whether operator action can be relied on to prevent pump damace. Consideration should be given to providing pump protection by means such as automatic switchcVer of the pump suctions to the alternative safety-grade source of water, automatic pump trips on low suction pressure, or upgrading the normal source of water to meet seismic i Category 1 and tornado protection requirements.

Response

As discussed in the response to GS-4 and GL-2 above, the SNUPPS design includes automatic transfer to the alternate water source. The alternate source (essential service water) is protected from tornados and is seismic Category I. V. Non-Safety Grade, Non-Redundant AFWS Automatic Initiation Signals Recommendation Gl-5 The licensee should upgrade the AFWS automatic initi-ation signals and circuits to meet safety-grade requirements.

Response

See response to Recommendation GS-7.

4. Major Contributors to Unreliability Bechtel lists the following major contributors to unreliability for each case. 1. Loss of Main Feedwater with Offsite Power Available - No single failures that would result in insufficient auxil-iary feedwater flow were identified. The dominant failure modes are composed of triple failure events. These failures involve the motor-driven pumps A and B (MDP A and ADP B), the turbine-driven pump TDP), and test and maintenance acts. The three most important triple failure events are: 1) MDP A And MDP B And TDP Do Not Start Due to Driver or Pump Faults, 2) MDP A Unavailable Due to Test or Maintenance And MDP B And TDP Do Not Start Due to Driver or Pump Faults, and 3) MDP A And MDP B Do Not Start Due to Driver or Pump vaults And Throttle Valve or Speed Governor Valve Faults. The unavailability of the AFWS for Case 1 is 2.0 x 10-5 per demand, which places this system in the high reliability group relative to operating PWRs. 2. Loss of Main Feedwater and Loss of Offsite AC The dominant failure modes are composed of triple failure events in this case and they include diesel generators A and D. The three most important triple failure events are:

1) MDP A And TDP Do Not Start Due to Driver or Pump Faults And Diesel Generator B Faults; 2) MDP A Unavailable Due to Test or Maintenance And TDP Does Not Start due to Driver or Pump Faults And Diesel Generator B Faults; and 3) MDP A Does Not Start Due to Driver or Pump Faults And TDP Unavailable Due to Test or Maintenance And Diesel Generator B Faults.

The unavailability of the AFWS for Case 2 is 5.3 x 10-5 per demand, which places this system in the high reliability group relative to operating PWRs. 3. Loss of Main Feedwater and Loss of All AC - If all ac power is lost, there is only the turbine-driven pump available. In this case, the dominant failures are single failure events. The three most important events are: 1) TDP Does Not Start Due to Driver or Pump Faults, 2) TDP Unavailable Due to Test or Maintenance, and 3) TDP Fails Due to Throttle Valve Faults. The unavailability is 1.4 x 10-2 per demand, which places this system in the medium reliabiity group relative to operating PWRs. SNL agrees with the above findings. For Case 1 the unavail-ability of AFWS is 2.0 x 10-5 per demand while for Cases 2 and 3 the unavailability is 5.3 x 10-5 per demand and 1.4 x 10-2 per demand, respectively. These valves are plotted in Figure 2 along with the operating plant ratings which were derived from NUREG-0611. l The SNUPPS AFWS has high reliabil.ity for Case NQ. 1, LMFW; high reliability for Case No. 2, LMFW/ LOOP; a,nd me,dium reliability for Case No. 3, LMFW/LOAC. Sandia agrees with these ratings. 5. Conclusions It is concluded on the basis of this review that the applicant has completed requirement (b) of the March 10, 1980 letter. The AFWS of the SNUPPS has high reliability relative to the reliability of AFWSs of operating plants for Case 1 Quantitatively, the unavailability of the system is approximately 2 x 10-5 per i demand. Qualitatively, the system is automatically initiated, i highly redundant, has no observed single point vulnerabilities, and is tested periodically and following realignment to demonstrate availability of flow path to the steam generators. Failure on i demand is dominated by triple failure events that involve pumps, test and maintenance,-and the TDP throttle valve or speed control valve. The unavailability for Case 2 is 5.3 x 10-5 per demand, l which places reliability in the high' range. Failure on demand is dominated by triple failure event.s that involve the pumps, test and maintenance and'the diesel generators. The unavailability for Case 3Lis 1.4 x 10-2, which places the reliability in the medium range. The TDP train has no identifiable ac power dependencies and is~ automatically actuated. Failure on demand is dominated by test and maintenance outage, failure of the TDP, and faults in the throttle valve. a l l l t l..

TR ANSIE NT EVE NT5 LMFW L M F W/ LOOP LMFW/ LOSS OF ALL AC* l f MED f HIGH LOW MED HIGH LOW MED HIGH PLAN TS LOW W E STING HOUSE H AOD AM NECK S tb 4> H SA~ O~OF R E h PR AIRBE ISL AND f 4> 'O IHD d> ALEM I ZION 4 3 4> Y ANKE E ROWE S TROJAN (> S (> INDI AN POIN T tl 9 F B (b K E W ANE E I G (> H B. ROBINSON BE AVE R VALLEY 41 e d i G g () GINNA PT.BEACM S S 4> 9 COOK TUR KE Y POINT S 9 4p FARLEY G S t> SURRY S G NORTH ANN A G G G S S S SNUPPS h-ORDE R OF M AGNITUDE IN UN AVAILABILITY REPRE SENTED. ' NOTE. THE SCALE FOR THIS EVE NT IS NOT THE SAME AS THAT FOR THE L M F W A N D L M F WIL OOP. Figure 2. Comparison of SNUPPS AFWS Reliability to other AFWS Designs in Plants Using the Westinghouse NSSS. 6. Glossary of Terms AC Alternating Current ac alternating current AFWS Auxiliary Feedwater System l ASME American Society of Mechanical Engineers CST Condensats Storage Tank DBA Design Basis Accident i DC Direct Current dc direct current ESWS Essential Service Water System FSAR Final Safety Analysis Report ft feet i spm gallons per minute hr hour IEEE Institute of Electrical and Electronic Engineers in inch KCP&L Kansas City Power and Light KG&E Co. Kansas Gas and Electric Company LAC Loss of all AC power i LMF Loss of Main Feedwater LOSP Loss of Offsite Power l MDP Motor Driven Pump NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System psia pounds per square inch absolute psig pounds per square inch gage.

l I l P&ID Piping and Instrumentation Drawing SFP Single Failure Point SG Steam Generator SNL Sandia National Labaratories SNUPPS Standardized Nuclear Unit Power Plant System i SSE Safe Shutdown Earthquake SSF Standby Shutdown Facilities T Temperature TDP Turbine Driven Pump Tavg Average Temperature TMI Three Mile Island UEC Union Electric Corporation V Volt t

  • F Degrees Fahrenheit i

l I

) \\ I 7. References 1. NUREG-0611 " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants" dated January 1980. 2. Letter to all Pending Operating License Applicants of Nuclear Steam Supply Sy. stems Designed by Westinghouse and Combustion Engineering from D. F.

Ross, Jr.,

Acting Director Division of Project Management Office of Nuclear Reactor Regulation, Subject, Actions Required from Operating License Applicants of Nuclear Supply Systems Designed by Westinghouse and Combustion Engi-neering Resulting from the NRC Bulletins and Orders Task Force Review Regarding the Three Mile Island Unit 2 Accident, dated March 10, 1980. 3. Reliability Analysis of the SNUPPS Auxiliary Feedwater System, submitted to NRC by N. A. Patrick, SNUPPS, on June 8, 1981. 4. Schedule 189 No. 1303-1 Title, " Review of Auxiliary Feed-water System Reliabilty Evaluation Studies for Comanche Peak 1 and 2, Waterford 3, Watts Bar 1 and 2, and Midland 1 and 2," dated May 7, 1981. 5. " Auxiliary Feedwater System," part 10.4.9 of the SNUPPS FSAR, Revision 3, April, 1981. l __ =. I Distribution: US NRC Distribution Contractor (CD6I) 7300 Pearl Street i Bethesda, Maryland 20014 130 copies for AN 25 copies for NTIS Author selected distribution - 1 copy ( List available from author. ) 4400 A. W. Snyder 4412 J. W. Hickman (5) 4412 B. J. Roscoe (2) 8214 M. A. Pound 3141 L. J. Erickson (5) 3151 W. L. Garner (3) For DOE / TIC ( Unlimited Release) i i l l l I

U.S. NUCLEAR REGULATO.RY COMMISSIW ' UEY/CRN5Y# y BIBLIOGRAPHIC DATA' SHEET SAND 81-2596 l 4 TITLE AND SU8 TITLE (Add Volume No.arfsperminant

2. (Leave b/wkl

( SNUPPS Auxiliary Feedwater System Reliability Study Evaluation

3. RECIPIENT'S ACCESSION NO.
7. AUTHOR (S)

$. DATE REPORT COMPLETED l B.J. Roscoe l YEAR XO~Tsugust 1981 l

9. *ERFORMING ORGANIZATION NAME AND MAILING ADDRESS (/ne/uar 2,p Codel DATE REPORT ISSUED MONTH l YEAR Sandia National Laboratories January 1982 c

Albuquerque, NM 87185 s.(t,,v,e,,nki i l

8. (Leave blankl
12. SPONSORING ORGANIZATION NAME AND MAILING ADORESS //nclude Zea Codel PRMGMSWRK M W Division of Safety Technology Office of Nuclear Reactor Regulation 11 CONTRACT NO.

U.S. Nuclear Regulatory Commission FIN A1303 Washington, DC 20555 13 TYPE OF REPORT PE RIOD COV E RE D (/nclus.ve dams /

15. SUPPLEMENTARY NOTES 14 (Leave o/mkl 16, ABSTR ACT 000 words or less)

The purpose of this report is to present the results of the review of the Auxiliary Feedwater System Reliability Analysis for SNUPPS. i 17 KE Y WORDS AND DOCUME NT AN ALYSIS 17a DESCRIPTORS l l 17b IDENTIF4E RS.'OPEN ENDE D TERMS

18. AV AILADILITY STATEMENT 19 SECURITY C(ASS (TM,s reporr/

21 NO OF PAGES Unclassified Unlimited za SEYcin LASS (ra.s pa,> 22 PRICE assing S NRC FORY 335 47 77)

UNITE 3 STATES I l PdUCLE A51 RESULATORY COMMISSION g WASHINGTON. D. C. 20555 y POSTAGE AND FEES PAID m OFFICI AL SUSINESS U.S. NUCLga n ngGubA TOR Y g cOuussalON N PE N ALTY FOR PRIVATE USE. $300 NA C1 CD Ch 2 Co -Q Ch> CX >o n m m O G> Hm T Ch Cn HmE m m 1 m\\ _H M C O m<> r-CD H_ O 2 c>2 I i C: xl l ~< a W l N-}}

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