ML20034C520

Proposed Tech Spec 3/4.7.9,enhancing RHR Sys Reliability by Removing Automatic Isolation Requirement for RHR Suction Valves

ML20034C520
Person / Time
Site:	North Anna
Issue date:	04/27/1990
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20034C518	List: ML20034C515 ML20034C520
References
NUDOCS 9005040065
Download: ML20034C520 (25)

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ATTACHMENT ONE m

l PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS 3/4.7.9 RESIDUAL HEAT REMOVAL SYSTEM l

t NORTH ANNA POWER STATION UNITS 1 AND 2 1

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VIRGINIA ELECTRIC AND POWER COMPANY r

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i PLANT SYSTEMS t

3/4.7.9 RESIDUAL HEAT REMOVAL SYSTEM (RHR) l OPERATING l

' LIMITING CONDITIONS FOR OPERATION 3.7.9.1 Two RHR subsystems shall be OPERABLE.

l APPLICARlLITY-MO D E S 1, 2 and 3.

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ACTION:

With one RHR subsystem inoperable, restore the inoperable subsystem to OPER ABLE status within 7 days or be in HOT SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

t SURVEILLANCE REQUIREMENTS l

4.7.9.1 Each RHR subsystem shall be demonstrated OPERABLE by:

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a. Verifying Isolation of the RHR system prior to the Reactor Coolant System pressure exceeding 500 psig, by closing, de energizing both remote operated RHR suction isolaten valves and locking the associated breakers.

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b. At least once per 16 motC.3, during shutdown, i

1. Cycling each, remote or automatically operated valve in the subsystem flowpath through one complete cycle of full travel.-

2. Verifying that each RHR pump is OPERABLE per Specification 4.0.5.

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NORTH ANNA UNIT 1 3/4726 4

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PLANT SYSTEMS RESIDUAL HEAT REMOVAL SYSTEM. (RHR) l

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' LIMITING CONDITIONS FOR OPERATION I

3.7.9.2 As a minimum, ork RHR subsystem shall be OPERARLE.

l APPLICABILITY: MODES 4 and 5.

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ACTION:

With no RHR subsystems OPERABLE,immediately restore at least one l

RHR subsystem to OPER ABLE status or maintain the Reactor Coolant System T vg ess than 350'F by use of attemate heat removal methods.

l The provisions of Specification 3.0.3,3.0.4 and 4.0.4 are not applicable.

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SURVEILLANCE REQUIREMENTS l

l 4.7.9.2 The required RHR subsystem shall be demonstrated OPERABLE by:

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a. Verifying isolation of the RHR system prior to the Reactor Coolant System pressure exceeding 500 psig, by closing, de energizing both remote operated RHR suction isolation valves and locking the associated breakers.

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b. At least once per 31 days:

1. Cycling each testable, remote or automatically operated valve in the subsystem flowpath through at least one complete cycle, and

2. Verifying the correct position of each manual valve in the subsystem flowpath, not locked, sealed or otherwise secured in position, and i

3. Verifying the correct position of each remote or automatically operated i

valve in the subsystem flowpath,

c. At least once per 18 mor.ths:

1. Cycling each, remote or automatically operated valve in the subsystem flowpath through one complete cycle of full travel.

2. Verifying that the RHR pump,in the subsystem flowpath,is OPERABLE per Specification 4.0.5.

NORTH ANNA UNIT 1 3/4727 y

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s PLANT SYSTEMS 3/4.7.9 RERIDtJAL HEAT REMOVAL SYSTEM. (RHR) l OPER&IlNS-l LIMITING CONDITIONS FOR OPERATION 3.7.9.1 Two RHR subsystems shall be OPERABLE l

APPLICABILITY: M O D E S 1, 2 and 3, ACTlON:

With one RHR subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS

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4.7.9.1 Each RHR subsystem shall be demonstrated OPERABLE by:

a. Verifying isolation of the RHR system prior to the Reactor Coolant System pressure exceeding 500 psig, by closing, de energizing both remote operated RHR suction isolation valves and lockin0 the associated breakers.

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b. At least once per 18 months, during shutdown,

1. Cyclin 0 each, remote or automatically operated valve in the subsystem

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flowpath through one complete cycle of full travel.

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2. Verifying that each RHR pump is OPERABLE per Specification 4.0.5.

NORTH ANNA UNIT 2 3/4723

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PLANT SYSTEMS 1

REalDlJAL.HEAI SEMOVAL SYSTEM - (RHR) l SHUTDQ3Mw l

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i CMITING CONDITIONS FOR OPERATION 3.7.9.2 As a minimum, one RHR subsystem shall be OPERABLE.

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APPLICABILITY: MODES 4 and 5.

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i ACTION:

With no RHR subsystems OPERABLE, immediately restore at least one RHR subsystem to OPERABLE status or maintain the Reactor Coolant System T vg ess than 350'F by use of allemate heat removal methods, l

i The provisions of Specification 3.0.3,3.0.4 and 4.0.4 are not applicable, l

SURVElLLANCE REQUIREMENTS l

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4,7.9.2 The required RHR subsystem shall be demonstrated OPERABLE by:

Verifying isolation of the RHR system prior to the Reactor Coolant System a.

pressure exceeding 500 psig, by closing, de energizing both remote operated RHR suction isolation valves and looking the associated breakers,

b. At least once per 31 tisys:

1 Cycling each testable, remote or automatically operated valve in the subsystem flowpath through at least one cor@lete cycle, and l

l 2, Verifying the correct position of each manual valve in the subsystem flowpath, not locked, sealed or otherwise secured in position, and I

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3. Verifying the correct position of each remote or automatically operated valve in the subsystem flowpath, t

c. At least once per 18 months:

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Cycling each, remote or automatically operated valve in the subsystem flowpath through one complete cycle of full travel.

2. Verifying that the RHR pump,in the subsystem flowpath,is OPERABLE per Specification 4.0.5.

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NORTH ANNA UNIT 2 3/4724 l

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i ATTACHMENT TWO s

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CHANGES TO PROPOSED MERITS TECHNICAL SPECIFICATIONS f

3.3.9 RESIDUAL HEAT REMOVAL SYSTEM i

NORTH ANNA POWER STATION UNITS 1 AND 2 f

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i VIRGINIA ELECTRIC AND POWER COMPANY t

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e' RCS PIV Leakage 3.3.16

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1 3.3 REACTOR COOLANT SYSTEM (RCS) 3.3.16 El.pretsure isolation valve Leakaoe

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LCO 3.3.16 The leakage from each RCS Pressure Isolation Valve (PlV) shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a RCS pressure 1 2215 and s 2255'psig.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS l

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more RCS P!Vs A.1 Restore RCS PIV leakage

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4 outside leakage limit, to within limit.

Completion Time is on a 4

per valve basis 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> DB A.2 Isolate the high 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> pressure portion of the affected system from the low pressure portion by use of at least two-closed manual or-deactivated automatic valves.

B.

Required Action not 8.1 Be in MODE 3, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> met within required i

Completion Time.

mfd B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> L

i North Anna Units 1 & 2 3.3 38 Amendment Nos xxx & yyy I

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RCS PlV Leakage l

0 3.3.16 SURVE!LLANCE REQUIREMENTS SVRVE!LLANCE FREQUENCY SR 3.3 16.1

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SR 3.0.4 is not applicable for entry into MODE 3 and 4.

Verify leakage of each RCS P!V within 18 months limits.

A!iD 2.

In accordance with tection XI of the ASME Code, tett pressure between Prior to entry 150 psig and 2235 psig are allowed, into MODE 2, whenever the unit has been in MODE 5 for

> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been performed in the previous 9 months AliD t

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve s

actuation due i

to automatic or manual action, or flow through the valve g

i Per SR 3.0.5

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(continued)

North Anna Units 1 & 2 3.3 39 LAmendment Nos. xxx & yyy 1

o RCS P!V Leakage o

3.3.16

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SURVE!LLANCE REQUIREMENTS (continued)

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SURVEILLANCE

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3 FREQUENCY SR 3.3.16.2 Verify Residual Hett Removal (RHR) System 18 months./

l ppen interlock prevents the inlet valves l

from being c ened with a simulated or actual RCS pressure signal 1 418 psig.

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.e W ify RHR System au re interlock M nth;s

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CROSS REFERENCES TITLE i'

NUMBER-RCS Loops Mode 4 i

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RCS Operational Leakage l.13.6 e

Accumulators U33.15 ECCS Trains Tavg > 350'F a3.4.1

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3.4.2 Containment Integrity i

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ATTACHMENT THREE i

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DISCUSSION AND SAFETY EVALUATION i

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FOR PROPOSED CHANGr.3 TO 4

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TECHNICAL SPECIFICATIONS i

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3/7.7.9 RESIDUAL-HEAT REMOVAL SYSTEM j

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NORTH ANNA POWER'BTATION UNITS 1 AND 2 l

PROPOSED TECHNICAL SPECIFICATION ; CHANGE RESIDUAL HEAT REMOVAL SYSTEM i

he PROPOSED CHANGES n

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1. CHANGING THE AUTOMATIC ISOLAYlON OF THE RESIDUAL HEATy REMOVAL zSYSTEM FROM' THE REACTOR COOLANQ SYSTEM TO' BE (

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- ANf ADMINISTRATIVELY CONTROLLED OPEHATION." ; Y

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l-T.S. ?/43S.1' RESIDUAL HEAT REMOVAE - OPERATING 4 i

}i ss-Specification 4,7.9.1.a will be modified to' require isolationfof the RHR system from -

y the Reactor Coolant System when the RCS pressure is above 500 psig_by closing 1

and deenergizin.g the RHR suction isolatio6' valves ratheMhan requiring automatic zl L

isolation of the RHR suction isolation valves when the RCS pressure ls above 660 psig..

3 DISCUSSION The Residual Heat Removal (RHR) System is usol a provide core cooling wher$

the Reactor CoolantSystem (RCS) is operated below approximately' 350 F and 450.psig. The system, which is located entirely within the' Reactor Containment build!r.g, consists of two pumps, two heat exchangers, and appropriate piping and M<

is not part of the Safety injection System and does not perform any Emergency t

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Core Cooling System functions. The RHR System piping inside the SHR isolation valves is designed for 600 psig at 400*F. RHR is connected to the "A".RCS hot leg

- by one pipe supplying suction to, both RHR subsystems. The suctio'n line has two

-1 moto'r operated valves (MOVs), MOV 700 and MOV-7 RHR System from the RCS,

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t The RHR suction line val es. MOV 700 and MOV 701, operate almilarly. These I 3

valves have no automatic opan functions and they may be remotely _ opened using the control room pushbutton only !f the'RCS pressure is not greater than the open 1

permissive interlock setpoint,418 psig. The suction valves may be reinotely dosed by using the control room pushbuttons or automatically closed whenever the RCS pressure is above the autoclose setpoint, 582 psig.

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A, PROPOSED RHR Ts NAPS uNirs 1 ANo 2 1

Two relief valves, RH RV 721 A and RH RV 721B, provide overpressure protection for the RHR system piping. One valve is located on the suction piping to each RHR pump with' lift setpoints of 467 psig.. Although each relief' valve is capable of relieving the tiow from three charging pumps, Interlocks permit only two pumps to operate simyltaneously and Technical Specification 3.5.3 and station procedures -

only allow orie charging pump to be opsrable when the RCS cold leg' temperature j

is less than or equal to 324 F on Unit 1 and 340 F on Unit 2. Therefore, the relief i

valves are fully capable of' relieving the maximum mass input to the Reactor Coolant System (RCS).

j in addition, RHR overpressure protection may also be provided by the Low l

Temperature Overpressurization Protection' system (LTOP) from the Pressurizer l.

Power Operated Relief Valves (PORVs), RC PCV-455C and RC PCV 456 when the f's RHR system is in service.

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BASIS FOR TECHNICAL SPECIFICATION CHANGE l

A disadvantage of the autoclosure feature is the possibility of an inadvertent valve closure during RHR operation resulting in the loss of decay heat removal capability.

The Westinghouse Owners Group has evaluated the removal of the RHR suction j.

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valve autoclosure interlock. The results of this study (WCAP-1_1736 A) shows an increase in RHR availability and overall improvement in plant safety if the autoclosure interlock is removed. North Anna Unit 1 was used as one of the four lead plants for this project.

L' The WCAP reviews many references including a 1985 internal NRC memorandum f

vrhich lists issues to be considered when a utility is' requesting removal of the autoclosure interlock. These issues are:

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1. The means available to minimize Event V concerns.

Response

Means are available to minimize Event V concerns. Event V is defined as an r

intersystem LOCA which bypasses containment. With the entire North Anna RHR System located inside of containment, an Event V accident can not take place with the RHR System. Removing the RHR suction valve autoclosure interlock will not change r3 probability of an Event V for North Anna.

Intersystem LOCAs between RHR and RCS are characterized as small break LOCAs inside of containment for North Anna. The RHR System provides three i

potential RCS LOCA paths, one through the suction line and one through each of the two discharge lines. The frequency of an intersystem LOCA associated with the RHR suction MOVs was quantified by Westinghouse for the current Page 2 of 15

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sN 90183 -

PROPOSED RHR Ts CHANGE l ~

NAPS uNrrs 1 ANo 2 North Anna design as 9.28 E 7. events per year. This intersystem LOCA probability does not significantly change after the autoclosure interlock has been removed.

t 2, The alarms available to alert the operator of an improperly positioned valve.

Response

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Although in the WCAP, the. Westinghouse Owners' Group recommends installing an additional annunciation to specifically, alert the control room operators to an improperly positioned RHR valve, no such new alarms are needed at North Anna. Existing control room annunciators at North Anna will.

adequately inform the control room operator of an overpressurization event-when RHR is in service.

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- Flow through the RHR relief valves, the letdown header relief valve or the Pressurizer PORVs will cause the Pressurizer Relief Tank annunciators to.

alarm.

' The RHR discharge piping high pressure annunciator will alarm The LTOP annunciators will inform the Control. Room Operator of.

overpressurization events, i

Leaving both suedon valves open will cause the PRT annunciators to alarm when the RHR pressure opens the relief valves at 467 psig which is sufficiently below the RHR design pressure of 600 psig. The RHR discharge piping tigh l

pressure annunciator will provide additional warning if pressure ~ reaches 560 l

psig.

if one suction MOV is left open and the.other closed,-the RCS could be fully i

pressurized without causing an annunciator to alarm. Then a single failure of the closed valve would cause an intersystem LOCA within containment. The l

unit would have to be depressurized until the intact MOV could be closed.

Westinghouse is recommending alarms and status lights be. installed at power stations considering autoclosure interlock removal. These alarms are intended e

i to reduce the probability of leaving a single suction valve-open.

Two i

annunciators, one for each MOV, would decrease the probability of leaving one valve open while the other is closed to approximately the probability of both MOVs falling.

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PROPOSED RHR Ts CHANGE NAPS UNITS 1 AND 2 In addition to existing annunciators and existing procedures, the proposed I

Technical Specification change will protect from RHR overpressurization. When RHR is remo'ved from service, the operator is required to close the suction valve, de-energize the motor by opening'the power supply breaker, and lock the breaker in the open position. Independent operator verification is used for these actions.

At North Anna, because of the requirements cited above,' the current probability l

of leaving one valve open and the other falling is likewise reduced to the.

I probability of both valves falling.. Additionally, design differences between North Anna and power stations which have RHR outside of containment or which have the Safety injection' System as part of the RHR System result'in-different consequences. For the latter utilities these' consequences necessitate.

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the installation of annunclators which warn of leaving one valve open. Because of existing operating procedures, existing control room alarms and a different RCS, RHR, and Si configuration, no new annunciators or status lights are needed at North Anna.

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3. Adequacy of the RHR suction relief valve capacity.

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Response

Each RHR suction relief valve has the car,acity to relieve the flow from at least one charging pump. Because electrical irderlocks permit only two pumps to be operable at a time, the RHR relief valves are of sufficient capacity to relieve the flow from the available charging pumps. The capacity.of the RHR relief valves is sufficient for any pressurization event.-

4. Means other than the ACI to ensure both MOVs are closed (e.g., single switch actuating both valves).

Response

The means to be used to ensure the MOVs are closed is t'hrough_ independent -

operator: verification.

When RHR is-~ removed from service the existing procedures require closing the suction-valve, de-energizing the' motor by opening the power supply breaker, and locking the breaker in the open:

position. Independent verification will be used to' ensure that these actions are properly performed.

Page 4 of 15 i

i SN 90-183 PROPOSED RHR TS CHANGE '

I l-NAPS UNirs 1 AND 2 1

~ 5. Assurance that the function of the open permissive circuitry is not affected by the proposed change,

Response

The open permissive interlock circuitry will be unaffected and the current open i

permissive setpoint will remain 418 psig. This interlock permissive prohibits the remote opening of the RHR suction valves unless the RCS pressure is less than 418 psig. Testing of the open permissive interlock will be performed after:

i completion of the. wiring changes necessary to remove the autoclosure interlock, and at regular testing intervals thereafter..

6. Assurance that MOV position indication will remain available in the control room.

Response

The current control room position indication for the RHR suction valve MOVs will not be changed. The current control room indication provides valve' position lights for full open, full closed and both. lights illuminated for intermediate position. These Indicators are de energized when the valve is de-energized.

_3 Assurcnce that the' suction and discharge MOVs are closed when the valves are l

de-energized will be through the use of independent verification of the correct-

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performance of the applicable operating procedures.

7. Assessment of the proposed changes effect on RHR system reliability, as well as the effect on Low Temperature Overpressurization (LTOP) concerns -

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Response

Past experience shows that unnecessary actuation of the autoclosure Interlock has caused a significant number of losses for decay heat.remo' val capability L

when overpressurization transients were not in progress. Inadvertent isolation of the-RHR System also increases the probability for overpressurization 3

because the letdown system, which provides RCS pressure. control, is connected to the RHR System.

Deletion of the autoclosure interlock-will j

therefore, increase the reliability of the RHR.

The performance of the North Anna LTOP will be enhanced _with the deletion of o

1 the RHR autoclosure interlock. The RHR suction line relief valves will provide relief capability in addition to the existing Pressurizer PORVs. This additional y

relief capacity could be defeated by to the autoclosure interlock which would isolate the RHR relief valves from the RCS.

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,y sN 90183 PROPOSED RHR Ts CHANGE

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NAPS uNifs 1 AND 2 ACCIDENT SEQUENCE EVALUATION All RHR overprdssurization accidents may occur in one of four possible situations.

The four situations are: 1. placing RHR in service, 2. RHR in service, 3. removing RHR from service, and 4. RHR not in. service. The key points of each accident category are discussed as follows:

1. While the RHR System is being placed in service, overpressurization of the RHR l

System will be avoided by the open interlock of the suction MOVs. Each MOV has independent prevent 'open interlock circuitry, including separate RCS 3

pressure transmitters.- If overpressurization were to occur, then the,RHR relief valves would provide adequate capacity and redundancy to prevent RHR pipe-

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damage. The operator would be informed of the pressurization event by the 1

Control Room annunciators and' indicators. The. RHR MOV may then, be -

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reclosed or RCS pressure decreased to terminate the transient.

2. When the RHR is in service, RHR and RCS overpressurization is prevented by the low temperature overpressurization protection (LTOP)' Technical-l Specifications. These requirements allow' only one Charging Pump to be-operable and limit the temperature difference between the Steam Generator primary and secondary liquid temperatures. If an overpressurization event does-occur, the Control Room Operator will be informed by the PRT annunciators, LTOP related annunciators and the RHR. discharge piping annunciator.

Mitigating the effects of the overpressurization will be'the RHR relief valves and-the Pressurizer PORVs. _lf the RCS temperature is above the LTOP system applicable RCS temperature, the RHR relief. valves have adequate capacity and redundancy to handle the pressure transient until operators correct the cause of the event.

3. When the RHR System is being removed from service, overpressurization of the RHR System will be prevented by the use of. operating procedures. These q

procedures will use independent operator verification to ensure the RHR -

suction MOVs are fully closed, the power supply breaker to the motor operators d

opened, and the breakers locked in the open position. Leaving both MOV suction valves open will cause the RHR relief valves to lift preventing further

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RCS pressurization. The operators will be able to identify that the suction valves are open by-the PRT annunciators, PP.T indicators and RHR discharge -

piping annunciator (if pressure is high enough). The operators will then be able l

to properly remove RHR from service or depressurize the RCS until the relief i

valves reseat.

Page 6 of 15 l

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PROPOsEo RHR Ts NAPS UNITS 1 ANo2 I

4. When the RHR System is not in service, overpressurization of the RHR System is prevented by the suction line isolation valves open permissive interlock and the discharge line isolation valve and check valve on each discharge line. With i

the RCS at a significantly higher pressure than the RHR System; an I

overpressurization event due to an intersystem LOCA between these systems is considered to be the worst type of RHR overpressure accident.. As previously discussed in the other three acciderit situations, leaving both suction line MOVs open is prevented by current procedural and proposed Technical Specification -

controls. Opening an RHR suction isolation valve is functionally prohibited unless the RCS pressure is below 418 psig by the open permissive circuitry.-

l The following are the possible accident sequences with the RCS at high pressures:

Both suction MOV discs rupturing.

One suction MOV left open and the other MOV disc rupturing.

l One discharge MOV left open and the check. valve in the same line rupturing.

One discharge line check valve leaking and the MOV in the same line rupturing or leaking.

t l-The resulting intersystem LOCA and rapid depressurization of the RCS will be a

identified and responded to according to guidance in the applicable Emergencyf Operating Procedures.

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Because of the currently existing overpressure p_rotection features, it has been concluded that the automatic closure of the RHR suction isolation valves can be replaced by a Technical Specification requirement to manually isolate _the-RHR from the RCS whenever the RCS pressure is above 500 psig, without adding new annunciators or status lights. This change will not significantly increase the probability of an RHR overpressure event and will increase'the reliability and availability of the RHR system during cold shutdown conditions.

L CONCLUSIONS The conclusion of this safety evaluation is that the overall safety of North Anna Power Station Unit 1 and Unit 2 will be improved. The Technical Specification surveillance requirement for the RHR autoclosure interlock should be' modified as L

i stated. The open interlock associated with RHR suction valves will be retained.

Additional station modifications, such as adding annunciators or status lights warning of open valves, are not necessary. An increase in RHR availability will result from removing the automatic closure feature of the suction MOVs.

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sN 90183 :

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PROPOSED RHR Ts CHANGE NAPS UNITS 1 AND 2 BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION 4

The proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92 because operation of North Anna Units 1 and 2 in 3

accordance with this change would not:

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1. result in a significant increase in the probability or consequence of an accident j

previously evaluated. Replacing the Residual Heat Removal System suction-valve automatic closure interlock requirement with a requirement for manual

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isolation will-not res' ult in._ a significant change in the probability = or consequences of an intersystem LOCA :This change will however decrease the -

probability for a loss of decay heat removal accident. The interlock has initiated past events where the suction valves have inadvertently closed. This accident sequence could lead-to core damage if not corrected in a timely manner.

Removing the autoclosure Interlock significantly decreases, but-does not eliminate, the possibility for this accident sequence.

The two RHR discharge lines do not have autoclosure interlock protection. The probability of leaving an RHR suction valve open and rupture of the remaining suction valve is approximately the same as leaving an RHR discharge valve open and rupture of the single discharge check valve. An intersystem LOCA between the high pressure RCS and the low pressure RHR System can occur through the suction line or either of the two RHR discharge lines. Removing the suction valve autoclosure interlock will not.have a significant-effect on the intersystem LOCA probability associated with the suction line or discharge

lines, The consequences of all UFSAR accidents remains the same because the RHR i

System is located entirely within containment and there are no changes to the radiological barriers within containment.

With or without the autoclosure interlock, the worst case accident associated with a rupture of the RHR isolation valves is an intersystern LOCA. The consequences of this LOCA are similar to the consequences of LOCAs already-analyzed in the UFSAR.

The consequences for a loss of decay heat removal event remain the same after the.

autoclosure interlock is removed.

2. create the possibility of a new or different kind of accident from any accident previously identified. Administrative control utilizing independent verification to

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close and de-energize the RHR suction valves is not significantly_different from t

using an autoclosure interlock to ensure both suction valves are closed when y

the RCS is pressurized above the RHR design pressure. Because there are no.

significant changes to the RHR System, the possibility for a different type of L

accident or malfunction are not created. The possible accident sequences have been previously evaluated.

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PROPOSED RHR Ts CHANGE j

NAPS UNITS 1 ANo 2

3. result in a significant reduction in a margin of safety. This change does not alter L

the conditions or assumptions of the accident analysis or the basis of the current

- Technical Specification.

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l RHR suqtion valve autoclosure interlock is not specifically addressed in.any I

Technical Specification basis. Removing the RHR suction valve autoclosure interlock will increase the availability of the RHR System to remove decay heat during shutdown conditions.

When the RHR System Is in service, the RCS low temperature overpressurization protection (LTOP) system will provide an adequate margin of -

safety to protect the RCS and the RHR Systems. In addition to the LTOP.

system, the two RHR relief valves are each sized to provide protection against -

RHR overpressurization.

l Therefore, pursuant to 10 CFR 50.92, based on the~above considerations, it has been determined that this change does not involve a significant hazards consideration.

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.C PROPOSED RHR Ts CHANGE NAPS UNITS 1 ANo 2

2. MODIFICATION OF: RHR PUMP SURVEILLANCE REQUIREMENTS l

T.S. 3/4.7.9.1 RESIDUAL HEAT REMOVAL - OPERATING -

T S. 3/4.7.9.2 RESIDUAL HEAT REMOVAL - SHUTDOWN u

The RHR pump testing change replaces the requirement to demonstrate that'each pump can develop greater than or equal to 123 psl: differential pressure with a i

requirement that the pumps be tested in accordance with Specification 4.0.5 which refers to Section XI of the 1980 ASME Boller and Pressure Vessel Code.

DISCUSSION The current test requires the pump discharge valve to be throttled, with the system discharge motor operated valves open, to obtain at least 123 psi differential pressure which in effect places the pump in a mini flow recirculation condition.

NRC Bulletin 88-04,-" Potential Safety Related Pump Loss", identified. concerns with minimum flow designs of safety related pumps and requested licensees to

. c investigate these concerns and correct them where applicable.- Virginia Electric and Power Company responded to these concerns in a letter to the NRC on August 8,1988 (Serial No. 88 275B) and subsequently determined that. full flow testing in accordance with Specification 4.0.5, which refers to Section XI of the 1980_ ASME l

Boiler and Pressure Vessel code, to be ' prudent and to provide the most reliable data on pump performance while reducing pump operation under-mini flow conditions.

The minimum differential pressure requirement of Technical Specification 3.7.9.1 was based on the assumption that the RHR pumps receive'an Engineered Safety Feature (ESF) actuation and are required for proper Safety injection _ (SI) flow. The -

RHR system is not required.for emergency core cooling under any accident conditions and is not capable.of providing Si flow to the core. 'Therefore, the minimum differential pressure, that is required in the STS'for SI pumps, is not i

applicable for North Anna's RHR pumps.

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. l NAPS UNITS 1 ANo 2 BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION j

The proposed change in the RHR pump surveillance testing criteria does not i

involve a significant hazards consideration as defined in 10 CFR 50.92 because operation of North Anna Units 1 and 2 in accordance with this change would not:

1. result in a significant increase in the probability or consequence of an accident i

previously evaluated. Testing of the RHR pumps in accordance with ASME XI requirements instead of the 123 psi differential pressure test requirement will not increase the consequences of any previously evaluated accidents. North Anna's RHR system is not part of the ECCS. The RHR pumps are not required l

to flow to the core during a Safety injection.

2. create the possibility of a new or different kind of accident from any accident previously identified.' Because there are no signifl cant changes to the RHR System, the possibility for a different type of accident or malfunction are not' created. The possible accident sequences have been previousl' evaluated.-

y

3. result in a significant reduction in a margin of safety. This change does not after the conditions or assumptions of the accident analysis or the basis of the current Technical-Specification. Since the RHR system does not_ perform any ECCS l

functions, testing of the RHR pumps in accordance'with: ASME Section XI instead of the current 123 psi d.ifferential pressure test requirement does not.

reduce any margin of safety and does enhance overall system reliability by eliminating pump wear from operation in a minimum. recirculation flow condition.

i Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has i

been determined that this change does not involve a significant hazards l

consideration.

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PROPOSED RHR Ts CHANGE l

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3. ADMINISTRATIVE CHANGES T.S. 3/4.7.9.1 RESIDUAL HEAT REMOVAL - OPERATING The Surveillance Requirements 4.7.9.1 is being reformatted for consistency _and clarity. Since the surveillances are. required at 18 month intervals, this will be stated in the common header rather than restating it in each Individual surveillance requirement. This is consistent with other. Technical Specificatioris.

T.S. 3/4.7.9.2 RESIDUAL HEAT REMOVAL - SHUTDOWN Currently, Specification 4.7.9.2 cites Specification 4.7.9.1l for the month surveillance requirements. This change will reiterate these requirements in their.

_j entirety rather than requiring referring to a separate specification.

This will j

eliminate confusion and clarify the requirements, i

T.S. 3/4.7.9.1 RESIDUAL HEAT REMOVAL - OPERATING T.S. 3/4.7.9.2 RESIDUAL HEAT REMOVAL - SHUTDOWN

'l Throughout specifications 3/4.7.9.1 and 3/4.7.9.2, the term " residual heat removal" ls being replaced by "RHR". "RHR" is currently defined in the title of 3.7.9.1 and is.

.l currently used in the title of 3.7.9.2.

This change: brings uniformity by ' defining "RHR"in the title of each specification and thereafter using "RHR" throughout the j

two specifications.

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.i DISCUSSION

-l These Changes are administrative'in nature only and involve'no~ changes to the plan; equipment or operating practices.

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- NAPS UNITS 1 ANo 2 PROPOSED RHR Ts BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION We have reviewed. the proposeo administrative changes to Technical Specificatiorts 3/4.7.9.1 and 3/4.7.9.2 and have found that these changes will not involve a significant hazards considerations because the changes conform to 48 FR 148764 (14870)" Examples of Amandments That Are Considered Not Likely To Involve Significant Hazards Considerations...", example j']"A purely administrative:

change to technical sp9cifications: for example a change to achieve consistency.

throughout the technical specifications, correction of an error, 'or a change in nomenclature." The proposed changes will not:

1.

result in a significant increase in the probability or consequences of an-accident previously evaluated, because, the changes are only. of an administrative nature intended to clarify the 'affected technical specification and provide for greater consistency. These Changes-involve no changes to the plant equipment or operating practices and procedures.

2.

create the possibility of a new or different kind of accident, because, the changes are only of an administrative nature' intended to clarify the'affected ~

technical specification and provide for greater consistency. These Changes involve no changes to the plant equipment or operating ; practices-and L

procedures.

3.

result in a significant reduction in the margins of safety because, the changes are only of an administrative nature intended to clarify the affected technical specification and provide for greater. consistency.7These Changes involve no changes to the plant equipment or operating practices and procedures.

Therefore, pursuant to 10 CFR' 50.92, based on the above consideration, it has been determined that op' eration of North Anna' Power Station Units 1 and 2, in accordance with these changes, will not involve a significant hazards consideration.

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O PROPOSED RHRTs CHANGE E

NAPS UNITS 1 AND 2 j

References 1.

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,

" Safety, Evaluation Report, North Anna-Power Station Units 1 and 2, Docket Numbers 5Q 338 and 50-339." NUREG 0053,' Original report issued June 1976.-

Supplement Number 7 issued -August 1977 section 5.2,8 Reactor: Coolant

' Pressure Boundary, Overpressurization Protection. Supplement 11 issued August 1980.

e

-2.

Virginia Electric and Power Company, North Anna Power Station, "L!censee Event.

Reports" (LERs). Unit 1 Docket Number 50 338: 87-007-00i 83 003/03L-0.'

Unit 2, Docket Number 50-339: 83 036/03L-0, 83 023/03L-0, 80-001/03L-0.

l 3.

U.S. Nuclear Regulatory Commission, Office = of: Nuclear. Reactor-Regulationi.

1

" Standard Technical _ Specifications for1 Westinghouse Pressurized ' Water Reactors", NUREG 0452, revision 4 issued Fall 1983.

4.

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, 4

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear.

Power Plants", NUREG-0800, revised June 1987.

4 1

5.

United States, Government. " Code. of Federal Regulations", Title -10, Energy.

i revised January 1,1988.

l-6.

U.S. Nuclear Regulatory Commission, Office of ~ Standards Development, Regulatory Guide 1.139, " Guidance For Residual Heat Removal", May 1978.

7.

Sandia National Laboratories for NRC's Office of Nuclear Regulatory.Research, l

" Analysis of Core Damage Frequency: From Internal Events: Surry, Unit ]",

NUREG/CR 4550,. revision 1, Volume 3, Part 1. Draft For Comment. Printed-(

September 1988.

8.

Pacific Gas and Electric Company, Diablo Canyon Units 1 and 2, Docket numbers 50-275 and 50 323, " Removal of RHR System Autoclosure Interlock Function", PG and E Letter No. DCL,-87187, August 4,1987.

9.

Westinghouse, " Residual Heat Removal System Autoclosure Interlock Removal i

Report For Diablo Canyon Nuclear Power Plant", WCAP-11117, Revision 2.0, July 1987.

.10. Westinghouse, " Residual Heat Removal System Autoclosure Interlock Removal I

Report For Westinghouse Owners Group", WCAP-11736-A, Revision 0.0, October -

1989.

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-ld,t PROPOSED RHRTS CHANGE NAPS UNITS 1 ANo2 11.-U.S. Nuclear ' Regulatory Commission, Office for Analysis and Evaluation of Operational Data, " Case Study Report:-- Decay Heat Removal Problems-at U.S. Pressurized Water Reactors", December 1985, AEOD/C503.

12. U.S.Nucleat Regulatory Commission, Office of Nuclear Reactor Regulation, " Loss of Residual Heat Removal (RHR) While The Reactor Coolant System (RCS) is 1

' Partially Filled (Generic Letter 87-12)", July 9,1987.

j

11. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, " Loss H

of Decay Heat Removal (Goneric Letter No.' 88-17) 10 CFR 50.54(F)", October 17, i

1988.

14. Institute of Nuclear Power Operations,- Significant Operating Experience-Report.

85-4, " Loss or Degradation of Residual Heat Removal Capability In PWRs", August 28,1985.

15. Institute of Nuclear Power Operations, Significant Operating-Experience Report 88 3, " Losses of Residual Heat Removal With Reduced Reactor Vessel Water Level", October 19,1988, i

16. U.S. Nuclear Regulatory - Commission, "An Assessment. of Accident Risk in-U.S. Commercial Nuclear Power Plants", WASH 1400, NUREG-75/014, October 1975. See Appendix 1, " Accident Definitions and Use of Event Trees", Section 4.1.6.

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