ML17341B400

Auxiliary Feedwater Sys Automatic Initiation & Flow Indication,Turkey Point Units 3 & 4, Technical Evaluation Rept

ML17341B400
Person / Time
Site:	Turkey Point
Issue date:	06/18/1982
From:	Kaucher J FRANKLIN INSTITUTE
To:	Kendall R NRC
Shared Package
ML17341B399	List: ML17341B400
References
CON-NRC-03-79-118, CON-NRC-3-79-118, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM TER-C5257-273-2, TER-C5257-273-284, NUDOCS 8209280530
Download: ML17341B400 (33)

Text

TECHNICAL EVALUATl0NREP0RT AUXILIARYFEED'ItII'ATER SYSTEM AUTOMATIC INITIATIONAND FLOW INDICATION (F-16, F-17)

FLORIDA POWER AND LIGHT CONPANY TURKEY POINT UNITS 3 AND fl NRC DOCKET NO.

50-250, 50-251 NRC TACNO.

42324, 42325 NRC CONTRACT NO. NRCC3-79-118 FRC PROJECT C5257 FRC ASSIGNMENT 9

FRC TASKS 273, 284 Prepared by Franklin Research Center 20th and Race Street Philadelphia, PA 19103 Author:

J.

E. Kaucher FRC Group Leader:

K. Fercner Prepared for Nuciear Regulatory Commission Washington, O.C. 20555 Lead NRC Engineer:

R Kenefa11 June 18, 1982 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..

8209280530 8209i5 PDR ADOCK 05000250 P

PDR Franklin Research Center A Division of The Franklin Institute The Benternin Franklin Parkway. Ptula. Pa. 19103 <2 i5) 4C8 I000

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TERM5257-273/284 CONTENTS Section Title Pacae IN TROD UCTZON

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1.1 Purpose of Review 1.2 Generic Issue

Background

l. 3 Plant-Specific

Background

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2 RVIEH CRITERIA

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3 TF HNZCAL EVALUATION...

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5 3.1 General Description of Auxiliary Feedwater System 3.2 Automatic Initiation.

3.2.1 Evaluation 3.2.2 Conclusion 5

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9 3.3 Flow Zndication-3.3.1 Evaluation 3.3.2 Conclusion

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10 3.4 Description of Steam Generator Level Indication 10 CONCLUSIONS 13 RE F~~~ NCES 14

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TERM5257-273/2S4 EOREHORD Tnis Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nucleaz Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions.

The technical evaluation was conducted in accordance with criteria established by the NRC.

Y". J.

E.

Kaucher contributed to the technical preparation of this report through a subcontract with HESTEC Services, Inc.

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TERM5257-273/284'

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INTRODUCTION 1.1 PURPosE oz REvIEw The purpose of this review is to provide a technical evaluation of the emergency feedwater system design to verify that both safety-grade automatic initia=ion cs.rcuitzy and flow indication are provided at Turkey Point Units 3

and 4.

Zn addition, the steam generator level indication available at these units is described to assist subsequent NRC staff review.

1 2 GENERIC ISSUE BACKGROUND A post-accident design review by the U.S. Nuclear Regulatory Commission (NRC) after the March 28, 1979 incident at, Three Mile Island (TMZ) Unit 2 has established that the auxilia y feedwater

{AFH) system should be treated as a

safety system in a pressurized water reac or (PWR) plant.

The designs of safety systems in a nuclear power plant are required to meet the general design criteria

{GDC) specified in Appendix A of 10C."-R50 [1].

The relevant design criteria foz the J2W system design are GDC 13, GDC 20, and GDC 34.

GDC 13 sets fozth the requirement for instrumentation to monitor variab'es and systems (over their anticipated zarfges of operation) that can affect reactor safety.

GDC 20 requires that a protection system be designed to initiate automatically in order to assure that acceptable fuel design limits a e not exceeded as a result of anticipated operational occur ences.

GDC 34 requires that the safety function of the designed

system, that is, the residual heat removal by the AFH system, can be accomplished even in the case of a single failure.

On September 13, 1979, the NRC issued a letter

[2] to each PWR licensee.

that defined a set of short-term requirements specified in NUREG-0578

[3] ~ It required that the A"-W system have automatic initiation and single failu e-proof design consistent with the requirements of GDC 20 and GDC 34.

Zn addition, AFW flow inaication in the control room must be provided to satisfy the zequire-ments se. forth in GDC 13.

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TERM5257-273/284 During the week of September 24, 1979, seminars were held in four regions of the country to discuss the short-term requirements.

On October 30, 1979, anothe letter was issued to each PHR licensee providing additional clarifica-tion of the NRC staff short-term requirements without altering their intent [4].

Post-TNI analyses of primary system response to feedwater transients and reliability of installed AFW systems also established that, in the long term, the AFW system should be upgraded in accordance with safety-grade require-ments.

These long-term requirements were clarified in the letter of September 5,

1980

[5).

This letter incorporated in one document, NUREG-0737 [6), all TMZ-related items approved by the commission for implementation at that time.

Section ZZ ~ E.l.2 of NUREG-0737 clarifies the requirements for the AFd system automatic initiation and flow indication.

l. 3 PLANT-SPECIFIC BACKGROUND Zn Reference 2, the NRC informed the Licensee, Florida Power and Light Company (FPL), that it would have to meet the requirements of NUREG-0578.

Reference 4 clarified and reiterated this requirement.

On November 21, 1979 [7),

FPL replied to the two NRC letters on the subject of short-term requirements.

Comments in FPL's letter relative to the AFR system centered on interim control-grade automatic initiation and flow indica ion systems.

On January 14, 1980

[8) r FPL provided detailed information on the AM design, citing specific items of Sections 2.1.7.a and 2.1.7.b of NUR="G-0578.

On February 3, 1981

[9)> FPL sent a letter to the NRC describing proposed A=-w system changes in detail.

On April 13, 1981 [10], FPL submitted a revision to the Turkey Point Technical Specifications to the NRC Director, Division of Licensing.

By letter cated July 23, 1981 ['1), FPL submitted additional information conce ning the Turkey Poin Units 3 and 4

AFW system.

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TER"C5257-273/284 2

REVE R CRITERIA To improve the reliability of the AM system, the NRC required licensees to upgrade the system, where necessary, to ensure timely automatic initiation when requirec.

The system upgrade vas to proceed in two phases.

In the short

tern, as a m'nimum, control-grade signals and circuits vere to be used to automatically initiate the AFH system.

This control~rade system was to meet the folloving requirements of NUREG-0578, Section 2.1.7.a

[3]:

"1.

The design shall provide for the automatic initiation of the auxiliary feedwater system.

2.

The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedvater system function.

3.

Testability of the initiating signals and circuits shall be a feature of the design.

4.

The initiating signals and circuits shall be powered from the emergency buses.

5.

Manual capability to'initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a

single failure in the manual circuits vill not result in the loss of sys" em function.

6.

The ac motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.

7.

The automatic initiating signals and circuits shal'e designed so Nzt their failure will not result in the loss of manual capability to initiate the AFH system from the control room."

In the long term, these signals and circuits were.to be upgraded in accor-dance with safety-grade requirements.

Specifically, in addition to the above rec"irements, the automatic initiation signals and circuits must have indepen-dent channels, use environmentally qualified components, have system bypassed/

inoperable status

features, and conform to control svstem inte action criteria, as stipulated in IEEE S d 279-1971

[12}.

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TERM5257-273/284 The capability to ascertain the AFW system performance from the control room must also be provided.

In the short term, steam generator level indica-tion and flow measurement were to be used to assist the operator in maintaining the required steam genezator level during AM system operation.

This syst: em vas to meet the following requirements from NUREG-0578, Section 2.1.7.b,[3]:

"1.

Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the contzol room.

2.

The auxiliary feedwa ter flow instrument channels shall be powered from the emezgency buses consistent vith satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9 f133.

The NRC staff has determined that, in the long term, the overall flowrate indication system for Westinghouse plants must include either one AW flowrate indicator with one vide-range steam generator level indicator for each steam generator, or two flovrate indicators.

The flovrate indication system must be environmentally qualified, powered fzom a highly reliable, battery-backed non-Class 1E power source, periodically testable, part of the plant's quality assurance

program, and capable of display on demand.

The operator relies on steam generator level instrumen ation and AFW flow indication to monitor AiW system perfozmance.

The requiremen s for this steam generator level instrumentation are specified in Regulatory Guide 1.97, I

Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess P'ant and Environs Conditions During and Folloving an Accident" [14).

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TERM5257-273/284 3

TECHNICAL EVALUATION 3 ~ 1 G NERAL DFSCRIPTION OF THE AUXILIARYFEEDWATER SYSTEM The auxiliary feedwater (AFW) system at Turkey Point Units 3 and 4

supplies water to the secondary side of the steam generator for z'eactor decay heat removal when normal feedwater sources are unavailable due to loss of offsite power or other malfunct:ions.

The system consists of three steam turbine-driven pumps (600 gpm at 2775 feet of water) capable of supplying feedwater to any or all of t)le six steam generators in. the two units.

All three pumps are interconnected on the discharge sade to two common discharge

lines, one line to each unit.

These common discharge lines each branch into three supply lines for the three steam generators in each unit.

The AFW lines to each steam generator contain two normally closed, dc, air-operated flow control valves in parallel.

3.2 AUTOMA IC INITIATION 3.2.1 Evaluation Auxiliary feedwater flow to the steam generators is a tomatically initiated when preset levels of any of the following parameters are exceeded:

Turbine-Driven Pumps 1.

Safety injection (0 of 3) 2.

Low steam generator level in any one steam generator (2 of 3) 3.

Loss of voltage on both 4160 V buses 4.

Loss of both main feedwater pumps.

All initiating signals and circuits are supplied from redundant, Class 1E r vital power supplies, as is the contro 1 power for al1 AFW valves.

In addition, all ac-operated valves are automatically loaded onto the diesel generators.

Tne normal valve configuration for the AW system is all AM pump suction valves open, discharge flow control valves closed>

and the steam admission

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TZRM5257-273/284 valves to the turbine-driven pumps closed.

The steam admission valves to two of the three AFN pumps are being modified so that they are dc~perated;

thus, two of the Abort pumps will start independently of ac power availability.

However, all three APR pumps are turbine-driven,. and the APR system, therefore, does not meet the pump power supply diversity requirement.

The AW pumps discharge control valves are dc solenoid/airwperated valves.

The air supply for all valves is backed by a seismically qualified nitrogen supply that automatically initiates on loss of normal air supply.

The operation of any one AFH pump will provide the necessary capacity for removing decay heat to prevent overpressurization of the reactor coolant system and to maintain steam generator levels.

All three APW pumps start upon automatic system actuation, and automatic isolation of a leaking steam generator is a design feature of the system and is provided by the main steam isolation system.

The primary source of water for the AFH system is the 250,000-gal, Seismic Category 1, condensate storage tanks (CST) of both units.

Sufficient water inventory (185,000 gal) is maintained in the tanks to bring the plant to hot standby, hold there for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, and subsequently cool down to the resiaual heat removal system entry temperature of 350'F.

Indication of CST level is proviaed in the main control room, and annunciation and alarm of CST low water level is providea.

The backup water supply for the AZW system uses water from the plant water treatment system to resupply the CST; this method could not be used if the CST were not available.

The Licensee further stated that a non-safety-grade 500,000-gallon deaerated water storage tank is being constructed and will be available to supply the CSTs.

A review of initiation logic and wiring diagrams revealed no credible single malfunction that would prevent protective action at the system level,.

when ecuired.

Xn.addition, the Licensee has stated that the design of the A=-8 svs em initiation logic meets I="~K Std 279-1971 in that no single component failure will prevent the automatic start signal from being initiated, and the ini ia"ing signals and circuits are po~ered from safety-grade power suppl'es.

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TERM5257-273/28 4 Manual operation of the AFH system is provided in the control room and at the local station.

Each control circuit is independent so that a single failure in one train will not affect the redundant train.

ln addition, the automatic initiating circuits are designed to be e3.ectrically independent from the control zoom manual start circuit so that the failure of the automatic initiating signals does not affect the control room manual capability of AFH pumps.

None of the protection signa3.s for the automatic initiation of AFrl are used as control signals; consequently, there is no control and protection system inte"action.

Seismic requirements for the emergency feedwatez system were not considered in the single failuze analysis because the NR will address this issue separately.

A determination of whether components aze qualified for accident and post-accident environments was not conducted.

The environmental qualification of safety-related

systems, including AZW system circuits and components, is being determined separately by the NRC and is not within the scope of ~Ais review.

Review of the initiation circuit diagrams revealed no credible single malfunction that would prevent proper system action when required.

The electrical isolation and physical separation of elements of the proposed auxiliary feedwater actuation system design comply with the require-ments of NUREG-0578

[3] and EEEE Std 279-1971

[12).

Concerning

bypasses, the Licensee has stated the following:

Channel Bypasses o

Trio Cnannel B

ass - This bypass is provided for periodic testing of the system and to remove a channel from service due to a component failure.

This bypass is manua3ly initiated and manually removed.

Only one channel can be bypassed at a time, and the coincidence logic is 2 of 2 while in test.

Ooezatina B passes o

The Licensee has stated that the system contains no operating bypasses.

The design of the AFAR/ control valves is such that the initiation signal operates a so3.enoid valve in series with the control air signal to each

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TERM5257-273/284 control valve.

There are no overzides in the contzol ci cuit for the solenoid valve; however, the air signal to the control valve can be controlled automatically or manually by the operator in the control room via hand indica ing controllers mounted on the main control consoles 3 and 4.

This design allows for considerable operational flexibility, but in effect allows the operator to override an actuation signal by taking manual control of the flow control valves and thus does not meet the requirements of ZEEE Std 279-1971.

The salient points are that where operating requirements necessitate automatic or manual bypass of a protective function, the design should be such that the bypass will be removed automatically whenever permissive conditions are not met; continuous indication of the bypass condition in the control room is required;.and a means for administratively contzolling the bypass should be provided.

The A=-8 pump discharge lines and turbine-driven A% pump steam supply lines for each unit combine into single lines through which all water and

steam, respectively, from either unit must flow.

A pipe break in either of these single flow paths would cause loss of the capability to provide AFW flow to all the steam generators of one unit.

The Licensee has agreed to develop operating procedures to provide direction to the operators regarding isolation of AM system steam supply lines oz feedwater line piping breaks.

Steam and feedwater piping modifications are also being developed to ensure redundancy in the common AFH dischazge header and the common steam supply header to the.

I AFR pump turbines.

The Turkey Point Technical Specifications reauire that each AFR pump be tested once each month.

AFn flow is initiated by manually opening valves (from the control room) to admit steam to the ARf pump turbine and therefore establish Av'H flow to the steam gene ators.

Channel unctional tests are required at least once every 62 days, and initiating signals and circuits are tested during the integrated safeguards test performed during each refueling outage.

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TERM5257-273/284 3.2.2 Conclusion It is concludec that the initiation signals, logic, and associated circuitry of the Turkey Point Units 3 and 4 AFR system comply with the long-term safety-grade requirements of Section 2.1.7.a of NUREG-0578

[3] and the subsequent clarification issued by the NRC with the following exceptions:

o Annunciation of channel

bypasses, in the control room, is not provided.

o The manual bypass capability for bontrolling the AFN flow control valves should be designed in accordance with IEEE Std 279-1971>

to provide automatic removal of the bypass when permissive conditions are not met, continuous indiction in the control room of the bypass condition>

and a means for administratively controlling the bypass switch.

3

~ 3 FLOP INDICATION, 3.3.1

=-valuation Each of he AW pump headers to each steam generator 's ecuipped with a flow transmitter with output indicated in the control room and locally at the AFH control va've location.

In addition, wide-range, non-safety-grade steam generator level indication is provided.

Both flow and level are cont'nuously displayed in the control room.

The AFN flow indication system is powered from the vital bus sytem, which is a Class 1E power source.

The AFN flow signal is also used as an input to the MR flow control system.

The Licensee has stated that the AW flow indication system is part of the plant quality assurance program.

AFN flow indication system channel checks are performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> P

ana channel functional tests are performed monthly.

Channel calibration is performed each refueling outage.

The environmental qualification of flow measurement and indication equipment is being reviewed separately by the NRC and is outside the scope of this review.

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a TERM5257-273/284 3.3. 2 Conclusion Zt is concluded that the sensors, transmitters, indicators, and recorders of the Turkey Points Units 3 and 4

AFW flow measurement system comply with the requirements of Section 2.1.7.b o

NUREG-0578 and the subsequent clarification issued by the NRC.

3.4 DESCRZPTZON OF ST~~

GENERATOR LEVEL ZNDZCATZON Steam generator level instrumentation at Turkey Point Units 3 and 4 serves several purposes in addition-to control room panel indication.

There are th ee safety-related measuretaent channels and two non-safety-related channels for each of the three steam generators in each nuclear unit.

One non-safety-related channel in each steam generator employs a wide-range sensor for indication and recording only (one 3-pen recorder in the control room).

Safety-related channels employ narrow-range sensors that provide signals for the following:

1.

reactor trip, turbine trip, feedwater pump t.ip, and automatic initiation of AFR system based on low-low levels 2.

turbine trip and feedwater pump trip based upon high-high levels 3.

control of main feedwater flow control valves through an isolation device.

The remaining non-safety-related channel in each steam generator is available as an alte nate means for control of the main feedwater flow control valves.

All safety-related channels are powered from emergency buses.

A'll are independent and separated to the extent that cables are run in separate raceways.

Non-safety-related channels are powered from normal 120-Vac non-class 1E buses.

Low and high level alarms are provided on the main annunciator panels for each steam generator.

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T RM5257-273/284 Safety-related channels are checked every 31 days as part of engineered safety features actuation system surveillance.

Calibration is performed during scheduled refueling outages (12-to 3.8-month intervals).

Separate control room panel indicators are provided for each safety-related channel of measurement (nine for each nuclear unit).

A selector switch permits the operator to record any one of the channels for each steam generator..

Table 1 lists the safety-related channels for all three steam generators of each nuclear unit; Table 2 lists non-safety-related narrow-range instrumen-tation for the three steam generators of each nuclear unit; and Table 3 lists non-safety-related vide-range instrumentation for the three steam generators of each nuclear unit.

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TERM5257-273/284 Table 1

Safet -Related Level S'team Generator Instrum en ts Tan No.

Channel Transmitter Range (inches of

~ater column) 1 1

1 2

2 2

3 3

3 LT-474 LT-475 LT-476 LT-484 LT-48 5 LT-486 LT-494 LT-495 LT-496 III,IIIIIIII ZZ ZZ ZZ

30.13-138. 22 Table 2

Non-Safety-Related Level Level Instruments Narrow Rance Steam Generator No.

Tao No.

Sa fety Channel Transmitter Range (inches of water column) 1 2

3 LT-478 LT-488 LT-498 NSR e

0-143 N

Table 3

Non-Safet -Related Level Level Instruments Wide Rance Steam Generator No.

Tao No.

Transmitter Range (inches of water column 1

2 3

LT-477 LT-487 LT-4 97 0-513 N

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CONCLUSIONS It is concluded that the initiation signals, logic, and associated circuitry of the Turkey Point Units 3 and 4 auxiliazy feedwater system comply with the long;term,. safety-grade requirements of Section 2.1.7.a of NUREG-0578 j3) and the subsequent clarification issued 'by the NRC with the following exceptions:

o Annunciation of channel

bypasses, in the control room, is not,provided.

o The manual bypass, capabil'ity for controlling the AEW flow control valves should be designed in accordance with IEEE Std 279-1971, to provide automatic removal of the bypass. when permissive conditions-are not met, continuous indication, in 'the con zol room of the bypass condition, and a means for administratively contzoll'ing the bypass switch.

It is concluded that the sensors, transmitters, indicators, and recorders of the Turkey Point. Units 3 and 4 A~A flow measurement systera comply with the requirements of Section 2.1.7.b of NU1KG-0578 and the subsequent clazification issued by the NRC.

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REPERENCES Code of Federal Regulations, Title 10, Office of the Federal

Register, National Archives and Records Service, General Services Administ:ration, Revised January 1,

1980 2 ~

NRC generic letter to all PWR licensees.

Subject:

Short-term Requirements Resulting from Three Mile Island Accident NRC, September 13, 1979 3 ~

NUREG-057 8 TMI-2 Lessons Learned Task Porce Status Report, and Short-term Recommendations NRC, July 1979 4 ~

NRC generic letter to all PWR licensees.

Subject:

Clarification of Lessons Learned Shor -term Requirements NRC, October 30, 1979 5.

NRC generic letter to all PViR licensees.

Subject:

Shoat-term Recuirements Resulting from Three Mile Island Accident NRC, September 5, 1980 6 ~

NUREG-0737 Clarification of TMX Action Plan Requirements

NRC, November 1980 7 ~

R. Z. Uhrig (FPL)

Letter to D.

G. Eisenhut (Division of Operating Reactors

[DOR]

NRC)

Subject:

Short-te m Recuirements-FPL Responses November 21, 1979 8.

R. E. Uhr'ig (FPL)

Letter to A. Schwencer (DOR, NRC)

Subject:

Reply to NRC Recommendations 2.1.7.a and 2.1.7.b of NUBEG-0578 January 14, 1980 9 ~

R.

E. Uhrig (FPL)

Letter to D.

G. Eisenhut (DOR, NRC)

Subject:

Detailed Description of FPL Replies to NUR"=G-0578 February 3, 1981 10'.

E. Uhrig (FPL)

Letter to D. G. Eisenhut (Director, Office of Nuclear Reac or Reculation, Division of Licensing)

Subject:

Proposed Changes to the Technical Specification for Turkey Point Units 3 and 4

April 13, 1981 Franklin Research Center A S~SIOI: d'sic Ft~.a Inaa1ute i~i,

TERM5257-273/284 ll.

R. E. Uhrig (PPL)

Letter to S.

A. Varga (DOR, NRC),

Subject:

Responses, to Questions Raised in Reference 14 July 23, 1981 12 IEEE Std 279-1971

'Criteria for Protection Systems for Nuclear Power Generating Stations" Institute of, El'ectrical and Electronics. Engineers, Inc.,

New York, NY 13 -

NUREG-75'/087 Standard Review Plan, Section 10.4.9f Rev.

1 NRC 14.

Regulatory Guide 1.97 (Task RS 917-4)

Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an

Accident, Rev.

2

NRC, December 1980

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