ML20009E266

Mcguire Nuclear Station Unit 1 Auxiliary Feedwater System Reliability Study Evaluation

ML20009E266
Person / Time
Site:	McGuire
Issue date:	07/31/1981
From:	Bradley G SANDIA NATIONAL LABORATORIES
To:	Office of Nuclear Reactor Regulation
References
NUREG-CR-2096, SAND81-0676, SAND81-676, NUDOCS 8107270334
Download: ML20009E266 (59)

Text

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%AF l McGuire Nuclear Station Unit 1 Auxiliary Feedwater System Reliability Study Evaluation Prepared by G. H. Bradley, Jr.

Sandia National Laboratories U S Nuclear Regulatory Commission 4

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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 rach 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.

s Available from GPC Sales Program Division of Technical Information and Document Control U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Printed copy price: $3.75 and National Technical Information Service Springfield, Virginia 22161

NUREG/CR-2096 SAND 810S76 McGuire Nuclear Station Unit 1 Auxiliary Feedwater System L

Reliability Study Evaluation l

Manuscript Completed: Apr;l 19(:

Dato Published: July 1981 Prepared by G. H. Bradley, Jr.

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 A1121 l

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i ABSTRACT This report presents the results of the review of the Auxiliary Feedwater System Reliability Analysis for the McGuire Nuclear Station Unit 1.

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Table of Contents Page Summary and Conclusions 1

1. Introduction 2 1.1 Background 2 1.2 Review Activities 2 1.3 Content and Results of the Reliability Analysis 3 1.4 Scope and Level of SNL Effort 3

2. AFWS Configuration 4 2.1 General Description and Function of the AFWS 4 2.1.1 Emergency Operation 8 2.2 Component Design Classification 10 2.3 Power Sources 12 2.4 Instrumentation and Controls 2.4.1 Controls 14 2.4.2 Information Available to Operator 15 2.4.3 Initiation Signals for Automatic Operation 21 2.5 Testing 22 2.6 Technical Specifications 23

3. Discussion 25 3.1 Mode of AFWS Initiation 25 3.2 System Control Following Initiation 25 v

Table of Contents (Cont'd)

Page 3.3 Test and Maintenance Procedures and Unavailability 26 3.4 Adequacy of Emergency Procedures 26 3.5 Adequacy of Power Sources and Separation of Power Sources 27 3.6 Availability of Alternate Water Sources 27

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3.7 Pctential Common Mode Failure 28 3.8 Appilcation of Lata Presanted in NUREG-0611 28 3.9 Search for Single Failure Points 28 3.10 Human Factors / Errors 29 3.11 NUREG-0611 Recommendat.ans,Long- and Short-Term 29 3.11.1 Short-Term Generic Recommendations 29 3.11.2 Additional Short-Term Recommendations 34 3.11.3 Long-Term Generic Recommendations 38

4. Major Contributors to Unreliability 40

5. Conclusions 45

6. Glossary of Terms 47

7. References 49 vi

List of Figures Page

1. Simplified Flow Ltagram of Auxiliary Feedwater 5 System - William 8. McGuire Station Unit No. 1 -

Duke Power Company

2. Comparison of McGuire AFWS Reliability to other 43 AFWS Designs in Plants Using the Wes*.inghouse NSSS 4

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Summary and Conclusions The accident at Three Mile Island resulted in many studies which outlined

, the events leading to the accident as well as those following. One of

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the important safety systems involved in the mitigation of such accidents was determined to be the Auxiliary Feedwater System ( AFWS). Each operat-ing plant's AFWS was studied and analyzed. The results were reported in NUREG-0611 3 . Prior to obtaining an operating license, the licensae of each nonoperating plant is requiredl to perform a reliability analysis of his AFWS for three transient conditions involving loss of main feedwater in a manner similar to the study made by NUREG-0611. Duke Power Company, the licensee for William E. McGuire Nuclear Station Unit 1, submitted a reliability report 2 to the U.S. Nuclear Regulatory Commission (NRC) in July 1980. A supplement 7 to this report was submitted in February 1981.

These reports were reviewed by Sandia National Laboratories (SNL). The i

following conclusions resulted from the review:

1. Duke Power Company has satisf actorily complied with the requirement to make a reliability stiidy of their AFWS.

2. A comparison of the reported reliability of McGuire's AFUS to those of operating plants shows that McGuire's reliability is in the low to medium range of the AFWS reliability for operating plants.

If the flow requirements were reduced or if operator action could be taken to increase the quantity of auxiliary feedwater, McGuire's reliability would be in the medium to high range. Sandia is in agree-ment with these astessments. The reason for the low reliability is the fact that the AFWS at McGuire is mechanically throttled to provide protection from a break in the main feedwater line to or rupture of a steam generator.

1. Introductior.

1.1 Background

The results of many studies pertaining to the Three Mile Island Nuclear Power Plant accident conclude that a proper functioning Auxiliary Feedwater System (AFWS) is of prime importance in the mitigation of such accidents. Therefore, a letter dated March 10, 19801 stating NRC's requirements regarding the AFWS was sent to all operating license applicants with a Nuclear Steam Supply System (NSSS) designed by Westinghouse or Combustion Engineering.

The Duke Power Company (DP) Charlotte, North Carolina, the applicant for an operating license for the McGuire Nuclear Power Station Unit I which has a Westinghouse-designed NSSS, provided a response in the form of a reliability analysis 2

• which was prepared for them by Westinghouse Nuclear Energy Systems, Westinghouse Electric Corporation.

1.2 Review Activities This project undertakes a review of those portions of the reliability analysis which (1) satisf y requirement (b) of the letter which states, " perform a reliability evaluation similar, in method to that described in Enclosure 1 (NUREG-0611) that was performed for operating plants and submit it for staff review," and (2) provide answers to the short and long-term

recommendations of NUREG-0611 in response to requirement (c) in the letter. The review was conducted according to Schedule 1894 which was submitted by SNL to NRC.

1.3 Content and Results of the Reliability Analysis The reliability analysis 2 was submitted to NRC in July 1980 and was received by SNL on September 8, 1980. The analysis makes a detailed study of the failure of the AFWS to provide sufficient Auxiliary Feedwater (AFW) flow to three of the four steam generators and compares the results obtained with those obtained for the operating plants studied in NUREG-0611. The analysis places the McGuire Nuclear Station Unit I with those operating plants having medium to low AFWS reliability.

1.4 Scope and Level of SNL Effort SNL reviewed the reliability analysis 2 submitted by DP.

Particular attention was directed toward determining that the analysis addressed in depth the reliability of the AFWS when subjected to three transient cases (1) LMFW, Loss of Main Feedwater, (2) LMFW/LOSP, Loss of Main Feedwater/ Loss of Offsite Power and (3) LMFW/ LAC, Loss of Main Feedwater/ Loss of all ac power. Also the methods used in the analysis were compared to those used in NUREG-0611. The specific findings are presented below in Sections 3, 4 and 5.

Comments and questions were recorded during the review and submitted to NRC on'the 3rd of October 1980. These questions a

were forwarded to DP by NRC. DP and its contractor Westinghouse Electric Corporation met with representatives from NRC and SNL on the 27th and 28th of October at the McGuire Nuclear Station Unit 1.

At this meeting a review of the McGuire AFWS and the AFWS reliability analysis was given by Westinghouse and a tour of the AFWS was conduct-ed by DP. During the tour, observations were made to facilitate the discussion period which followed. In the discussion period each of 4

the 33 questions asked was answered and discussed in detail. The 1

j official responses 5 were forwarded to NRC on the 10th of November.

No exact verification of the results could be mad,e since the basic events and the failure allocations associated with each event were not published in the caport2,

2. AFWS Configuration 2.1 Ceneral Description and Function of the AFWS Figure 1 is a simplified flow diagram of the AFWS for McGuire Unit.

No. 1. The AFWS for McGuire Unit No. 2 is of identical design.

Each unit's AFWS consists of two motor-driven pumps (450 gpm @ 3200 ft. head) and one turbine-driven pump (900 gpm @ 3200 ft. head).

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supplies water to steam generators C and D. The turbine-driven AFW pump supplies water to all steam generators. The criterion for successful performance of the AFWS is the delivery of a minimum of 450 gpm to three of the four steam generators.

Water to the suction of each unit's AFWS is supplied from several non-safety grade water sources and one safety grade water source.

The water is supplied f rom these sources on a priority based on water quality as follows:

Source Safety Grade Maximum Capacity

1. Upper Surge Tanks No 85,000 gallons

2. Auxiliary Feedwater Nc 42,500 gallons Condensate Storage Tank

3. Condenser llot Well No 170,000 gallons

4. Nucicar Service Water Yes Nuclear Service Water Pond (1.8 x 10 8 gal)

An additional 30,000 gal. (maximum) is available from the condensate storage tank when the condensate storage tank pumps are available to fill the upper surge tanks. A source of non-steam generator quality water is also availabic through the steam generator quality water lines f rom the Standby Shutdown Facilities (SSF). Layout of piping for this source of water is constructed to run through designated

" vital" areas of the plant with tight security controlled access to provide a source of water that is potentially safe against acts of

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7-sabotage by outside persons not employed at the nuclear site. Only the redundant isolation valves and the associated piping cre

- safety grade components.

The non-safety grade water sources are headered into a common line in the Service Building.- This single line is then routed to the AFW pumps located in the Auxiliary Building. The safety-grade Nuclear Service Water System (NSWS) is connected to the AFWS of 4

each unit such that redundant nuclear service water channels A and l 4

B are aligned to the turbine-driven pump of each unit; channel A '

is also aligned to motor-driven AFW pump A while channel B is aligned to motor-driven AFW pump B of each unit. Safety class isolation valves are provided in the AFW pump suction lines to isolate the non-safety grade sources when supply from the NSWS is i

1 required.

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}. The discharge from each pump flows through an air-operated control valve and a motor-operated (remote controlled)-isolation valve in individual feedlines to each steam generator. The discharge from each AFW pump also has a loop for full flow pump testing that is I

also used as a minimum flow loop for protecting the pump during low flow operation. A locked closed interconnection for long-term use only in the case of a LOCA is provided between the AFW motor-driven pump discharge lines. This permits flow from either pump i to be fed to all four steam generators. The ficw from the AFh3 y,- , ---y. -- . . , . --e- , -n- .-.n--. - - , , , - , .w - g ---- ,, ,--

8-enters the steam generators through individual nozzles on each generator.

2.1.1 Emergency Operation Start-up of the AFW pumps is automatic. As an accident initiated cooldown of the reactor progresses, the AFWS is controlled manually from the Control Room or locally at the pumps if the Control Room is not available. At start-up, motor-driven AFW pumps supply 170 gpm feedwater to each steam generator and the turbine-driven AFW pump supplies an additional 180 gpm per steam generator. As less water is needed to maintain the water level in the steam generators, AFWS flow is diminished by adjusting the motor-driven AFW pump discharge control valves and/or adjusting the turbine speed or pump discharge control valves on the turbine-driven pump. As the accident progresses the turbine-driven AFW pump is removed from service by the operator to minimize condensate loss to the atmosphere. The AFWS is allowed to function during an accident shutdown until the reactor coolant temperature is reduced to 3500F ami a pressure of 425 psig. At this point, the Residual Heat Removal System (RHRS) is placed into operation and the AFWS taken out of service.

Water supply for the AFWS during emergency plant operation is normally from the auxiliary feedwater condensate storage tank.

! Under highest flow condition (two unit blackout), each unit has a r

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l-10 minute minimum supply of condensate quality water. Before depletion of this source, the operator is expected to take ranual control of the AFWS and trip the turbine-driven pump. If this pump is needed, the operator can open a bypass valve (ICA152) to provide a parallel flow path from the upper surge tank and additional feedwater to prevent automatic switchover to the NSWS source. The operator also has the option of defeating the auto switchover after taking manual control of the system.

If the AFW condensate storage tank is not available, feedwater is next supplied from the Upper Surge Tank (UST). Makeup is required to the UST if it is used as the AFWS water source through an entire cooldown operation. Normal make-up to the UST is from the condensate storage tank of the Condensate System or f rom the Demineralized Water S ys t em. A maximum of 30,000 gallons can be supplied to the UST if two condensate storage tank pumps are available. Each pump supplies flow at 300 gpm. If two demineralized water pumps are available, a maximum of 127,500 gallons from the filtered water tanks at 475 gpm flow from each pump can be supplied to the UST. Operator action is required to align the UST to these sources. In the event the AFW Condensate Storage Tank and the UST are both unavailable, supply to the suction of the AFW pumps is next supplied from the condenser hotwell. The total gallons per minute flow from the hotwell is j limited based on condenser vacuum and water level. Operator action 1 I

is required to limit total flow from the AFW pumps.

Should all the previously mentioned non-safety grade feedwater sources be unavailable to the suction of the AFW pumps, the assured source of wrier is then supplied from the Nuclear Service Water System (NSWS). The NSWS water source is aligned automati-cally when the suction pressure of the AFW pumps drops below 2 psig for 3 seconds. The NSWS pumps are required to supply cooling water to the motor-driven AFW pumps. The suction lines of the AFW pumps are piped into the NSWS pump discharge piping downstream of all NSWS cooling control valves; thus, the NSWS pumps do not have to operate to supply water to the AFW turbine-driven pump.

For non-seismic plant conditions and in the event water is not avail-able f rom the NSWS or from the Condensate System as explained above, feedwater may be supplied to the steam generators via the Standby Shutdowr. Facilities (SSF). The isolation valves that isolate the supply of water f rom the SSF open automatically when the suction pressure of the AFW turbine-driven pump drops below a preset value to align the AFW pumps to this source of feedwater.

2.2 Component Design Classification The AFWS of each unit including its primary water supply f rom the NSWS are engineered safeguards systems. The major components of s

these systems are designed according to seismic and other requiremente as given in the following table:

ASME-B/PV Seismic S ys tem / Component Code Section O BJ., DBE

1. AFWS - Turbine- III - Class 3 Yes Yes Driven Pumen

2. AFWS - Motor- III - Class 3 Yes Yes Driven Pumps

3. AFWS - Valves III - Class 2 Yes Yes III - Class 3 Yes Yes

4. NSWS - Pumps III - Class 3 Yes Yes

5. NSWS - Strainers VIII Yes Yes

6. NSWS - Valves III - Class 2 Yes Yes III - Class 3 Yes Yes The components listed above are also designed for tornado, wind and i 1

missile protection. Piping for the safety-related portions of AFWS and NSWS is designed accordingly. The motors of the motor-driven pumps of the AFWS and NSWS for each unit are designated Electrical-Safety - Class 2E. This same classification is given to the motora of valve motor operators of these systems. Electrical equipment of 2E classification requires seismic qualification to a safe shutdown carthquake criterion and are so des *gned.

The NSWS is designed to provide cooling water for various Auxiliary Building and Reactor Building heat exchangers during all phases of station operation. Each unit has two redundant " safety-related"

i headers serving two trains of equipment necessary for a safe plant shutdown and a "non-essential" header nerving equipment not required for a safe shutdown. Water is normally supplied to the system from a lake (Lake Norman). Should a seismic event cause a loss of the lake, a Standby Nuclear Service Water Pond (SNSWP) that is designed to meet seismic loads provides a source of water to the NSWS. As an Engineering Safeguards System the SNSWP is automatically va.1ved to provide feed to the channels of the NSWS of both units following a safety injection signal from either unit.

2.3 Power Sources The turbine-driven AFW pump of each unit is supplied with steam from redundant feedlines. One feedline is supplied steam from the unit's steam generator "B" outlet header upstream of its main steam isolation valve (MSIV) and the other from the unit's steam generator "C" outlet header upstream from its MSIV. This assures steam to the turbine-driven AFW pump even with these two MSIV's closed.

Each unit of the station is equipped wi+h an Essential Auxiliary Power System (EAPS) ths.t includes onsite 4160 V, 600 V, 120 V ac and 125 V de power. This system supplies power necessary for a safe shutdown of the reactor, containment isolation, containment spray and cooling, auxiliary feedwater flow, and emergency core cooling following an accident. It consists of redundant switchgear, load l

t centers, motor control centers, panelboards, battery chargers, i

batteries, inverters, diesel-engine ac generators (two per unit),

protective relays, control devices, and interconnecting cable supplying two redundant load groups of each unit.

The 120 V ac and the 125 V de Vital Instrumentation and Control Power Systens of the EAPS supply continuous power for control and -;

instrumentation in the Reactor Protection and Control System. The EAPS of each unit is designed to meet the criteria set forth in the NRC Ceneral Design Criteria (CDC 17, CDC 18), IEEE 279-1971, IEEE 308-1971 and Regulatory Guides 1.6, 1.9, and 1.32.

The motor-driven pumps of a unit's AFWS receive power from their EAPS via two identical but separate 4160 V emergency buses. In the event of a loss of offsite power, the pumpi receive power via the emergency buses f rom two diesel ac generatots (4160 V) designat-ed "A" and "B". Diesel generator "A" provides power to the emergency bur that feeds the unit's AFW motor-driven pump designated "A" and diesel generator "B" provides power to the bus feeding AFW pump "B".

Redundant motor-operated valves and other electrical equipment designated "A" and "B" receive power in a similar manner.

2.4 Instrumentation and Controls The controls and instrumentation of the AFW9 are designed to meet NRC-imposed safety class separation requirements.

2.4.1 Controls The control of auxiliary feedwater flow and steam generator water level is accomplished f rom the main control room by use of air-operated valves that automatically maintain a correct ,

position, (set manually in the control room or at a local control panel) for the required auxiliary feedwater flow.

Since the instrumentation used in the automatic control of valve position is not qualified for all accident events, safety grade solenoids are provided to assure that the AFW control valves are in the " fail-safe open" position following the automatic start ,

of a corresponding upstream AFW pump. The pump minimum-flow valves are likewise provided with safety grade solenoids to isolate pump minimum flow on the same automatic start signal.

Af ter any automatic start, the operator can reposition the solenoid valves from the control room and use the non-safety control if oper-abic. If repositioning of the solenoid valves causes a flow upset, as indicated by flow indication on enca st oun generator, the solenoid i valves must be placed back in the fail-safe position, and the contral valves manually throttled locally at the valve using handwheels provided.

All manual valves in the main flow paths of the AFWS are mechanically

" lock-opened" or " lock-closed" in their normal system operation posi-tion. The motor-operated valves in the flow path are designed to fail in their *as is" position.

I 2.4.2 Information Available to Operator The important information available to the operator at the control room and locally near the component's location includes for each AFW pump the suction pressure, suction flow and diocharge pressure; l

i and for each steam generator the flow, water level and pressure.

l Additional information is given in the following instrumentation i

table:

l l NSSS Physical Location l Component Control System Control Room Local l 1. Motor (AFW Pump) - start /stop AFWS X X pump

2. Turbine (AFW Pump) - start /stop Main Steam X X pump

3. Turbine (AFW Pump) - raise / Main Steam X X j lower spe.ed I

i 4. AFW pump auto defeat - on/off AFWS X X

5. UST supply motor-operated AFWS X X isolction valve (ICA4) -

open/close

6. Condenser hotwell supply AFWS X X motor-operated isolation valve (ICA2) - open/close AFWS X X

7. AFW condensate storage AFWS X X tank supply motor-operated isolation valve (ICA5) -

open/close

NSSS Physical Location Component Control Sys tem Control Room Local

8. Nuclear ses Ice water supply AFWS X X motor-operated isolation valves (ICA85A, ICA86A, ICA18B, ICAll68, ICA15A and ICAll7B) -

open/close

9. AFW pump suction motor- AFWS X X operated isolation valves (ICA7A, ICA9B, and ICA11B) -

open/close

10. Feedwater air-operated AFWS X X flow control valves (ICA40, ICA44, ICA56, ICA60, ICA36, ICA48, ICA52, and ICA64) -

flow position ,

11. Feedwater motor-operated AFWS X X isolation valves (ICA42B, ICA46B, ICA58A, ICA62A, ICA3BB, ICA50B, ICAS4A, and ICA66A) - open/close

12. NSWS to AFW pump suction AFWS X auto switch-over switch-defeat.

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NSSS Physical Location Component Control S ystem Control Room Local

, 13. Alternate SSF supply AFWS

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motor-operated isoletion valves (ICA161C and ICA162C)

Process / Component NSSS Physical Location Status Indicator System Control Room Local

1. AFW pumps motor - on/off AI US X X

2. AFW pump turbine - on/off AFWS X X

3. AFW pump turbine - speed Main Steam X X

4. AFW pump turbine - steam Main Steam X X pressure

5. AFW punp. auto start defeat - AFWS X on/off

6. Control room overidden AFWS X by local control

7. Upper surge tank - Condensate X water 1cvel

8. Steam condensate storage Condensate X tank - water level

9. Condenser hotwell - Main Condenser X water icvel

10. AFW condensate storage AFWS X tank - water level

11. NSWS pond - water level NSWS X

12. UST supply motor-operated AFWS X X isolation valve (ICA4) -

open/close

13. Condenser hotwell supply AFWS X X motor-operated isolatior.

valve (ICA1) - open/close

Process / Component NSSS Physical Location Status Indicator S ys t e,m, Control Re 3, Level

14. AFW condensate storage AFWS X X tank supply motor-operated isolation valve (ICA6) -

open/close

15. NSWS supply motor-operated AFWS X X isolation valves (ICA85A, ICA15A, ICA116B and ICA117B) - open/close l 16. AFW pump suction motor- AFWS X X operated isolation valves (ICA7A, ICA98, and ICA11B) -

open/close

17. AFW air-operated min! mum AFWS X t

flow valves (ICA20, ICA27, and ICA32) - open/close

18. AFW air-operated feedwater AFWS X X flow control valves (ICA40, ICA44, ICA56, ICA60, ICA36, ICA48, ICA52, ICA64) open/

I close and 0-100 percent open

19. AFW feedwater flow motor- AFWS X X operated isolatfon valves (ICA42B, ICA46B, ICA58A, ICA62A, ICA38B, ICA50B, ICAS4A, and ICA66A) - open/close

20. Main feedwater pressure Main Feedwater X

Process / Component NSSS Physical Location Alarms S ystem Control Room

1. AFW turbine stop valve - Main Steam X tiosed

2. AFW turbi.ne spee.' setting - Main Steam X Ices than 3600 rpm

3. UST - low water level condensate X

4. ULT supply motor-operated AFWS X isolated valve (ICA4) -

closed

5. AFW pump suction motor- AFWS X operated isolation valve (ICA7A, ICA9B, ICA11A) - A closed

6. Individual AFW pump - AFWS X low suction pressure

7. Individual AFW pump AFWS X low suction flow

8. AFW pump air-operated AFWS X minimum flow valves (ICA20, ICA27, and ICA32) - open

9. Feedwater flow air-operated AFWS X control valves (ICA40, ICA44, ICA56, and ICA60) -

open or closed

Process / Component NSSS Physical Location A1.cma S ys tem Control Room

10. Feedwater flow air-operated AFWS X control valves (ICA36, ICA48, ICA52, and ICA64)-above or below setpoint

11. Feedwater flow motor operated AFWS X isolation valves (ICA42B, ICA46B, ICA58A, ICA62A, ICA38B, ICA50B, ICA54A, and ICA66A) - closed 0,4.3 Initiation Signals for Automatic Operation f

The AFW motor-driven pumps start automatically on the following signals:

1. Two out of four low-low water level signals in any steam generator, l

2. Loss of all main feedwater pue;m. l I

3. Initiation of a safety injection "S" signal, 5.

4. Loss of offsite power (station blackout).

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An auto-start-defeat twitch is provided for items 1 and 2 above for uso during periods not requiring an automatic motor-driven AFW pump start. During periods which require automatic start, the defeat feature is automatically removed. The AFW turbine-

driven pump starts automatically upon the generation of two out of four low-low water level signals in any two steam generators or upon loss of offsite power.

The piston-operated isolation valves (ISA48 and ISA49) that control steam to the turbine-driven AFW pump are held closed with air through redundant normally-energized electrical solenoid control valves connected in a series configuration. De-energizing one of the redundant control solenoids will ven air from the " fail-open" valve operator allowing the isolation valve to open and admit steam to the turbine of the AFW turbine-driven pump.

Whenever any AFW pump starts automatically, the Steam Generator Blowdown System (SGBS) is automatically isolated by closing piston operated isolation valves. The SCBS does not isolate automatically on an operator induced AFU pump start so that operation of the SGBS  !

is maintained during all normal modes of plant operation.

2.5 Testing The AFW pumps are periodically tested to meet inservice surveillance requirements. A full flow test loop to the UST is provided at the discharge of each AFW pump. Adequate instrumentation is provided i to verif y pump performance. The motor-driven AFW pumps may be used during plant startup in their normal alignment to the steam generators. Pump performance and automatic feedwater flow control can be verified during this mode of operation. The turbine-driven AFW pump performance and its discharge control valve travel stop settings can also be verified during this mode of operation.

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( 2.6 Technical Specifications A review of the Technical Speci ication indicates that for power, etart-up, or hot standby plant status the limiting conditions of the AFWS for plant operation include:

1. At least three independent auxiliary feedwater ;_ttps and associated flow paths shall be operable with:

a. Two motor-driven AFW pumps each capable of being l

powered from separate emergency buses and,

b. One turbine-driven AFW pump capable of being powered f rom an operable steam supply system.

2. With one auxiliary feedwater pump inoperable, restore at

least three AFW pumps (two capable of being powered from i

separere emergency buses and one capable of being posered by an operable steam supply eystem) to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot standby condition within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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The Technical Specification requires all valves of the AFWS to be given inservice tests and inspections in accordance with the ASME Boller and Pressure Vessel Code (Section XI ar.d applicable Addenda) for Safety Class 1, 2 and 3 components. Additional surveillance requiremene.s include:

t

1. At least once per 31 days

a. Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1210 psig at a flow of greater than or aqual to 450 gpm.

b. Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1210 psig at a flow of greater than o< equal to 900 gpm when the secondary steam sup,1y pressure is greater than 900 psig.

c. Verifying that each non-automatic valve in the flow path that is not locked, seal.ed, or otherwise secured in position is in its correct position.

d. Verifying that each automatic valve in the flow path is in the fully open position whenever the auxiliary feedwater sysh m is placed in automatic control or when above 10 percent of RATED TilERMAL POWER.

2. At least once per 18 months during shutdown

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, (3) Steam Cencrator Water Level -- Low-Low from one steam generator.

1 I

b.' Verifying that the steam-turbine-driven pump starts automa-I tically upon receipt of each of the following test signals:

(1) Loss of offsite power, (2) Steam Generator Water Level -- Low-Low from two steam l

. generators.

c. Verifying that the valve in the suction line of each auxiliary feedwater pump _from the Nuclear Service Water

! System automatically actuates to its full open position within less than or equal to 10 seconds on a low suction pressure test signal.

l

3. Discussion 3.1 Mode of AFWS Initiation The AFWS is initiated automatically. The MDP's will start on (1) two out of four low-low water level signals in any steam generator, (2) loss of all main feedwater pumps, (3) initiation of a safety injection signal, and (4) loss of offsite power.

The TDP starts on the generatiori of two cut of four low-low water level signals in any two steam generators or upon loss of offsite pover. In the event of low suction pressure the NSWS is automati '

cally aligned to the pumps and suction pressure restored.

3.2- System control 7311owing Initiation Af ter initiation, ; proper flow is established by adjusting the MDP discharge control valves and/or adjusting the TDP sneed ci discharge

L control valves. When the reactor coolant condition is reduced to 3500F and 425 psig the 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 '.nspections in accordance with the ASHE Boiler and Prcssure Vessel Code (Section XI and applicable Addenda) for Safety class 1, 2, and 3 components. Also every 31 days there are (1) pump discharge pressure and flow tests (2) non-automatic valve position verification test and (3) automatic valve position verificotion when the AFWS sytem is in automatic control. The pumps and system are available on demand during all tests. During shutdown the automatic starting i

of each pump and the functioning of the automatic valves from closed to full open in 10 seconds in the action line of each AFW pump trom the NSW3 are checked: there are no coincident tests or maintenance of components within the AFWS. There was no evidence that the actual Test and Maintenance Procedures were reviewed in detail tn assure that the abovi guidelines had been observed.

3.4 Adequacy of Emergency Procedures The Emergency Procedures were not reviewed or included in the analysis by DP. Emergency operation was discussed and it was assumed that the emergency procedures had been written to implement the emergency operations. However, answers (6) 1

by McGuire to the recommendations of NUREG-0611 imply that there may be no emergency procedures for the AFWS.

3.5 Adequacy of Power Sources and Separation of Power Sources The EAPS is designed to meet the criteria set forth in the NRC General Desigri Criteria (CDC17, GDC18), IEEE 279-1971, IEEE 308-1971 and Regulatory guides 1.6, 1.9, and 1.32. The motor-driven pumps, associated motor-operated valves and other electrical equinment receive power from two identical but separate 4160 V emergency buses. 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. ine TDP 1

is not dependent upon ac power. Redundant power sources enhance l system reliability as does the separation of these power sources which eliminates many common cause failure events.

3.6 Availability of Alternate Water Sources DP has many alternate water sources. For water of steam generator quality the main source (non-safety grade) is the upper surge tank.

Backup to this source is the AFW condensate Jtorage tank, the upper surge tank and the condenser hotwell. The primary alternate water source is the NSWS which is safety grade but not of steam generator quality. This snurce is automatically available in the event the NPSil falls below 2 psig for 3 seconds. Valve operation time is 10 seconds or less. A fina? alternate s spply is the SSF. Every

attempt is made to operate with steam generator quality feedwater.

Automatic switch over to the NSWS is fast enough to prevent pump failure because of no water supply 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. DP reports that no common cause failures were discovered through the analysis that would result in both the TDP and the two MDP's not meeting the AFW flow requirements.

3.8 Application of Data Presented in NUREG-0611 i

The report (2) did not contain a table which included the Fault tree events. The fault ttee was checked and all applicable components as shown in Figure 1 were properly included. Although the report states that the data in NUREG-0611 were used end the tables from NUREG-0611 reproduced as Appendix B, there was no way from the report review to verify this. At the meeting at McGuire, evidence was made available which showed that the analysis was made in great detail and that NUkdG-0611 data were used.

e.

3.9 Search for Single Fai? Points There were no single failure pointo (SFP) associated with case 1, LMF, or Case 2, LMF/LOSP. For Case 3, LMF/ LAC, there were many I

\

sn

R I ^

i l .29 -

i SFPs since Case 3 describes a single-channel system. Any SFP has a major effect on the reliability of a redundant system and if _ any are found, they. should be thoroughly reviewed.

3.10 Human Factor / Errors Human Factors / Errors were considered by DP where appropriate in the fault tree. Automation is a major factor in decreasing the

~

effect on reliability of these types of event.

I i

3.11 NUREG-0611 Recommendations, Long- and-Short-Term 3.11.1 Short-Term Generic Recommendations I. Technical Specification Time Limit on AFW System Train Outage Recommeadation CS-1 i

The licensee should propose modifications to the Technical Speci-fications to limit the time that one AFW system pump and its associated flow train and essential instrumentation can be in-operable. The outage time limit and subsequent action time should be as required in current Standard Technical Specifica-tions; i.e.,'72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, respectively.

Response

McGuire Nuclear Station has Standard Technical Specifications and, as such, already has these requirements included in the Technical Specifications.

1 II. Technical Specification Administrative Controls on Manual Valves - Lock and Verif y Position.

Recommendation GS-2 The licensee should lock open single valves or multiple valves in series in the AFW system pump suction piping and lock open other single valves or multiple valves in series that could ititerrupt all AFW flow. Monthly, inspections should be performed to verify that thesa valves are locked and in the open position. These inspections should be proposed for incorporation into the surveillance requirements of the plant Technical Specifica-tions. See Recommendation GL-2 for the longer-term resolution of this concern.

4

Response

All manual valves in the auxiliary feedwater flowpath are checked monthly to verify that they are lvcked open. This requirement is included in the McGuire Technical Specifications.

III. AFW System Flow Throttling - Water Hammer Recommendation GS-3 The licensee sbauld reexamine the practice of throttling AFW system flow to avoid water hammer.

I

4 l

? ,

I l The licensee should verif y that the AFW system will supply on demand sufficient initial flow to the necessary steam generators to assure adequate decay heat removal following loss of main l

, feedwater flow and a reactor trip from.100% power. In cases f

I where this reevaluation results in an increase in initial AFW system flow, the licensee should provide sufficient information i

to demonstrate that the required initial AFW system flow will

' 'e not result in plant damage due to water hammer.

Response

Auxiliary feedwater flow is not throttled initially to prevent water hammer. The required flow rate is available within 60 seconds following the initiating event.

IV. Emergency Procedures for Initiating Backup Water Supplies Recommendation GS-4 Emergency procedures for transferring to alternate sources of AFW supply should be available to the plant 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

Transfer of the auxiliary feedwater supply from the normal to the safety gra'e assured supply occurs automatically when suction pressure drops below an acceptable limit. The instru-

mentation and controls utilized in the switchover logic are safety grade.

V. Emergency Procedures for Initiating AFW Flow Following a Complete Loss of Alternating Current Power k

Recommendation CS-5 The as-built plant should be capable of providing the required AFW flow for at least two hours from one AFW pump train, independent of any ac power source.

Response

The auxiliary feedwater system at McGuire is capable of automatic initiation and of providing the required flow for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> independent of any ac power source. This is accomplish-ed by means of the turbine-driven auxiliary feedwater pump and de motor-operated / solenoid valves at appropriate locations in the system.

V1. AFWS Flow Path Verification Recommendation CS-6 The licensee should confirm flow path availability of an AFW system flow train that has been out of service to perform periodic testing or maintenance as follows:

(1) Procedures should be implemented to require an operator to determine that tie AFW system valves are properly I

t l aligned and a second operator te independently verify that the valves are properly a*.tgned.

(2) The licensee should propose Technical Specifications to assure that, prior to plant startup following an extended ,

i" cold shutdown, a flow test would be performed to verify the normal flow path from the primary AFW system water l

source to the steam generators.' The flow test should be l

l conducted with AFW system valves in their normal alignment. l l f Response 3 l

I (1) Procedures will be developed to provide for double verifi- i cation of the auxiliary feedwater system alignment following maintenance activities. For normal periodic testing of the -

l

>?

system, no realignment of manual valves is required so no verification of system status is necessary.

(2) McGuire Nuclear Station has the latest version of the Standard Technical Specifications which provide adequate {

I p

assurance of the operability of the auxiliary feedwater [

L s ys tem. I f

VII Non-Safety Grade, Non-Redundant AFW System Automatic Initiation f Signals Recommendation GS-7 l

The licensee should verify that the automatic start AFW system i L

signals and associated circuitry are safety grade.  ;

Response

e The McGuire auxiliary feedwater system employs safety grade automatic igitiation signals and circuits. Automatic initiation of the system is discussed in the Final Safety Analysis Report, Section 7.4.1.1.

VIII Automatic Initiation of AFWS Recommendation GS-8 The licensee should install a system to automatically initiate AFW system flow.

Response

See response to Recommendation GS-7.

3.11.2 Additional Short-Term Recommendations I Primary AFW Water Source Low Level Alarm Recommendation The licensee should provide redundant level indication and low level alarms in the control room for the AFW system primary water supply, to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

The low level alarm setpoint should allow at least 20 minutes i

L for operacor action, aesuming that the largest capacity AFW pump is operating.

Response

As noted in the response to Recommendation GS-4, the McGuire design utilizes an automatic transfer of the auxiliary feed-water supply to the assured supply, the nuclear service water system. In addition to this, single channel, non-safety-grade level indication and low level alarms are provided in the control room for each of the normal auxiliary feedwater sources (upper surge tank, condenser hotwell and filtered water storage tank).

11 AFW Pump Endurance Test Recommendation The licensee should perform a 72-hour endurance test on all AFW I

system pumps, if such a test or continuous period of operation has not been accomplished to date. Following the 72-hour pump run, the pumps should be shut down and cooled down and then l restarted and run for one hour. Test acceptance criteria should I

include demonstrating that the pumps remain within design limits with respect to bearing / bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equipment in the room.

r I

i

Response

It is our understanding that the Staff has modified this recommendation to perform a 48-hour endurance test on all auxiliary feeda. iter pumps in lieu of the 72-hour test. The motor-driven auxiliary feedwater pumps were run several days during the hot functional test period. The exact time period and system configuration will be documented. The turbine-driven auxiliary feedwater pump has not been run for an extended period. A 48-hour test of this pump will be performed after the heatup following initial fuel loading.

l III Indication of AFW Flow te the Steam Generators Recommendation 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 AF;.' flow to each steam generator should be provided in the control room.

l (2) The AFW flow instrument channels should be powered from the emergency buses consistent with satisfying the emergency power diversity requireuents for the AFW system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Stardard Review Plaa, Section 10.4.9.

Res ponse This recommendation has been implemented as indicated by our response to item 11.1.1.2 from NUREG-0694 contained in Duke Power Company " Response.to TML Concerns" submitted initially on May 23, 1980.

IV AFWS Availability During Periodic Surveillance Testing -

Recommendation Licensees with plants which require locol manual realignment of valves to conduct periodic tests on one AFW system train and which have only one remaining AFW train available for operation should propose Technical Specifications to provide that a dedicated individual who is in communication with the control room be stationed at the manual valves. Upon instruction from the control room, this operator would re-align the valves in the AFW syst:m f rom the test mode to its operational alignment.

Response

The auxilf'.ry feedwater system design is such that no manually operated valves need to be repositioned during periodic testing of the system. Those valves which must be repositioned can be operated from the control room. In the event the system is automatically actuated, these valves will be actuated to their f

" safety" position.

I

. 11.3 Long-Term Generic Recommendations 1 Automatic initiation of AFWS Recommendation GL-1 For plants with a manual starting AFW system, the licensee should install a system to automatically ir.itiate the AFW systeu flow. This system and associated automatic initiation signals should be designed and installed to meet safety grade requirements. Manual AFW aystem start and control capability should be retained with man.ual start serving as backup to automatic AFW system initiation.

4

Response

See response to Recommendation GS-7.

II Single Valves in the AFWS Flow Path ,

Recommendation GL-2 Licensees with plant designs in which all (primary and alternate) water supplies to the AFW systems pass through valves in a single flow path should install redundant parallel flow paths (piping and valves).

Response

The McGuire auxiliary feedwater system design has redundant flow paths via redundant pumps, valves and piping.

1

l III Elimination of AFWS Dependency on Alternating Current Power Following a Complete Loss of Alternating Current Power I

L Recommendation GL-3 I

At least one AFW system pump and its associated flow path and essential instrumentation should automatically initiate AFW i

system 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

i l

l See response to Recommendation GS-5.

IV Prevention of Multiple Pump Damage Due to Loss of Suction Resulting f rom Natural Phenomena Recommendation GL-4 Licensees having plants with unprotected normal AFW system water supplies should evaluate the design of their AFW systems to determine if automatic protection of the pumps is necessary folloving 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 problem and taking action should be considered in determining whether operator action can be relied on to prevent pump damage.

. Consideration should be given to providing pump protection by means such as automatic switchover of the pump suctions to the alternate safety-grade source of water, automatic pump trips on

low suction pressure, or upgrading the normal source of water

/

to meet se.smic Category 1 and tornado protection requirements.

Response

Auxiliary feedwater system pumps are protected by automatic switchover to the safety grade source of water following any loss of normal source resulting from natural phenomena or other causes.

V Non-Saf ety Grade, Non-Redundant AFWS Automatic Initiation Signals Recommendation G1-5 The licensee should upgrade the AFW system automatic initiation signals and circuits to meat safety grade requirements.

Response

t See response to Recommendation GS-7.

4. Major Contributors to Unreliabilit y Duke Power lists the following major contributors to unreliability for each case.

Case No. 1 - LMFW The dominant (controlling) contributors to system unreliability were found to be the loss of the motor- and turbine-driven pump

)

systems as caused by such f ailure modes as the pumps f ailing to start and run due to a pump component failure; the motor-driven pumps fail to start or run as caused by an open pump circuit breaker or a fault in the electrical control circuit used for automatic closing of a pump circuit breaker; and the turbine-driven pump fails

-to start and run due to faults attributed to the turbine control stop valves and the turbine speed control valve.

Other dominant contributors to AFWS system unreliability were found to be unscheduled maintenance of pumps and the testing of valves in the feedlines to each steam generator f rom the motor-driven pumps and the turbine-driven pump.

The redem'%:cy employed in the design of the McGuire AFWS was found to be of the type whereby no obvious single faults (active components, manual valves or human errors) were identified that dominate the unreliability of the AFWS for a loss of main feedwater transient.

Case No. 2 - LMFW/ LOOP Because the sources of ac power are redundant the dominant failure modes for Case 2 are the same as Case 1. The reduction in AFWS reliability for this transient is caused by the loss of redun-dancy in ac power sources that results from a loss of offsite power.

t i

j Case No. 3 - LMFW/LOAC In this transient, loss of both offsite and onsite ac power is postula-ted to occur with the coincident loss of main feedwater flow, so that j A

the available operatting pump subsystems of the AFWS are teduced to only the steam turbine-driven pump train. Thus, any single failures in this pump train alone would be sufficient to fail the AFWS for this transient.

The dominant contributors to system unreliability for this case were found to include: (1) the turbine-driven pump is offline for unscheduled maintenance; (2) the pump train is down due to testing of the motor-operated valve (ICA7A) in the pump's suction line; (3) the manual valve (ICA19) in the pump suction line fails closed due to hardware failure or human error causing a loss of NPSil at the pump's suction; (4) various turbine-driven pump faults (i.e., turbine / pump hardware component failure, the turbine control stop valve fails closed or the speed control valve fails closed) causes loss of dischatge flow from the pump; and (5) the manual valve (ICA21) or check valve (ICA22) in the pump's discharge line fails closed blocking flow from the turbine-driven pump.

SNL agrees with the above listing. No quantification of results was made by DP in their report nor were resulta quantified in NUREG-0611.

The quantitative estimates obtained f rom notes provided by DP are for case 1 5.1x10-4, for Case 2 1.2x10-3, and for Case 3 1.2x10-2, These values are plotted in Figure 2 along with the operating plant ratings which were derived from NUREG-0611. The McGuire AFWS han medium reliability for Case No. 1, LMFW; low reliability (high end) for Case No. 2, LMFW/LOCP; and medium reliability (high end) for

_ i: 3_

l .

TRANSIENT EVENT 3 LMFW LMF W/ LOOP LMFW/ LOST DF ALL AC' PLANTS LOW MED HIGH LOW MED HIGH LOW MED HIGH WE STINGHOUSE l

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/ medium range for Case 1, the low range for Cane 2 and the medium range for Case 3. Sandia is in agreement with these assessments and with the additional assessments made with the alternate assump- , tions. The long-term modification suggested by Duke Power in reference 7 to simplif y operation of the system, which in turn should increase system reliability, should be followed by NRC. f f l l l i i l L s

6. Glossary of Terms ac alternating current i

AFW Auxiliary Feedwater AFWS Auxiliary Feedwater System ASME American Society of Mechanical Engineers B/PV Boiler and Pressure Vessel CST Condensate Storage Tank DBE Design Basis Earchquake de direct current DP Duke Power Company EAPS Essential Auxiliary Power System FSAR Final Safety Analysis Report gpm gallons per minute IEEE Institute of Electrical and Electronic Engineers LAC Loss of all ud power LMFW Loss of Main Feedwater LOCA Loss of Coolant Accident LOOP Loss of Offsite Power MDP Motor Driven Pump MSIV Main Steam Isolation Valve NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System o Glossary of Terms (Cont'd) 4 NSWS Nuclear Service Water System OBE Operational Basis Earthquake . psig pounds per square inch gage RilR3 Residual lleat Removal System SFP Single Failure Point SGBS Steam Generator Blowdown System SNL Sandia National Laboratories SNSWP Standby Nuclear Service Water Pond SSF Standby Shutdown Facilities TDP Turbine Driven Pump THI Three Mile Island UST Upper Surge Tank V Volt 1

7. References

1. Letter to all Pending Operating License Applicants of Nuclear Steam Supply Systems Designed by Westinghouse and Cambustion Engineering from D. F. Ross. Jr., Acting Director Division of Project Management Office of Nuclear Reactor Regulation, Subject, Actions Kequired from Operating License Applicants of Nuclear Supply Systems Designed by Westinghouse and Combustion Engineering Resulting from the NRC Lulletins and Orders Task Force Review Regarding the Three Mile Island Unit 2 Accident, dated March 10, 1980.

l 2. WCAP-9751 Reliability Analysis of the Auxiliary Feedwater System l for the McGuire Nuclear Station Unit #1 dated July 1980 by W. E. Shopsky Westinghouse Electric Corporation.

3. NUREG-0611 " Generic Evaluation of Fee dwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants"sdated January 1980.*

4. Schedule 189 No A1121-0 Title, " Review of Auxiliary Feedwater System Reliability Eval 2ation Studies for Diablo Canyon 1, McGuire 1, Summer 1, San Onofre 2, and Palo Verde" dated August 6, 1980.

5. Memorandum William O. Parker Jr., Vice President Steam Production Duke Power Company to Harold R. Denton, Director NRR, USNRC

, Re: McGuire Nuclear Station Docket No. 50-369 dated November 10, l 1980. l

6. Memorandum William O. Parker Jr., Vice President Steam Production Duke Power Company to Harold R. Denton, Director NRR, USNRC Re: McGuire Nuclear Station Docket No.s 50-369, 50-370, dated September 18, 1980.

7. Memorandum William O. Parker Jr., Vice President Steam Production Duke Tower Company to Harold R. Denton, Director NRR, USNRC Re: McGuire Nuclear Station Dockets Nos. 50-369, 50-370, dated February 4, 1981.

• Availabic for purchase from the NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and/or the National Technical Information Service, Springfield, VA 22161.

I Distribution: U.S. Nuclear Regulatory Commission (130 copies for AN) Distribution Contractor (CDSI) 7300 Pearl Street Bethesda, MD -20014 Armand Lakner U.S. Nuclear Regulatory Commission - Washington, DC 20555 1222' C. H.-Bradley 3141 L. J. Erickson (5) 3151 W. L. Carner (3) (for DOE / TIC) 3154-3 C.11. Dalin (25) (for NRC distribution to NTIS) I 4400 A. W. Snyder 4414 J. W. Hickman 8214 M. A. Pound \ t~;C Poru 335 1. f EPORT NUMBER (Assspied by DOC / ,,,,, U.S. CUCLEMI KEGULATORY COMMIS$PON _ g BIBLIOGRAPHIC DATA SHEET SAND 81-0676'

4. T'TLE MD SUBT6TLE (Add Vosume Na, sf garer,eart 2. (teave blaint McGuire Nuclear Station Unit 1 Auxiliary Feedwater System Reliability Study Evaluation 3. RECIPIENT'S ACCESSION NO.

7. AUTHOHIS) 5. D ATE REPORT COMPLETED i G.H. Bradley, Jr. l^"

I "Apri o" T."l 1981 9, PE RFORMING ORGANIZATION NAME AND MAILING ADDRESS (/nclude lip Code) DATE REPORT ISSUED MONTH Sandia Natianal Laboratories . July l YEAR 1981 Albuquerque, NM 87185 s it , ,, u.n.,

8. (Leave Nankl

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (inctuor I,0 Codel 10 PROJECT / TASK / WORK UNIT NO.

Division of Safety Technology Office of Nuclear Reactor Regulation 11. CONTR ACT NO. U.S. Nuclear Regulatory Commission Washington, DC <0555 ' FIN A1121 13 "YPE OF RE PORT PE RIOD COVE RE D (/nclusive defes) 15 SUPPLEMENTARY NOTES 14 (Leeve o/atal 16 AUSTR ACT 000 wruds or sessf This report presents the results of the review of the Auxiliary Feedwater System Re, liability Analysis for the McGuire Nuclear Station Unit 1. 17 KE Y WORDS AND DOCUMENT ANALYSIS ' 7a DESCRIPTORS 17b IDENTIFIF RSJOPE N ENDED TERMS AV AIL ABILtTY ST ATEMENT 19 SE CURITY CLASS (Th,s recorrt 21 NO OF PAGES Unclassified Unlimited 20 SE Cuai: < Ct AS,S tra,i p, i 22 price Unclassified S NRC F ORY 33% t7 774 .