ML20098H491

Proposed Tech Specs Reflecting Addition of Accumulator Leakage Rate Limits to Automatic Depressurization Sys Operability Criteria

ML20098H491
Person / Time
Site:	Hatch
Issue date:	09/05/1984
From:	GEORGIA POWER CO.
To:
Shared Package
ML20098H490	List: ML20098H489 ML20098H491
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.28, TASK-TM TAC-56047, TAC-56048, NUDOCS 8409130004
Download: ML20098H491 (6)

Text

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

. 3.5.F Autmatic Depressurization Systen 4.5.F Autunatic Depressurization System (ADS)

(ADS)

1. ' Normal Syste Availability 1.

Normal Operational Tests

%e seven valves of the Autmatic a.

A simulated autmatic actuation Depressurization System shall be test shall be performed on the operable:

ADS prior to startup after each refueling outage. Surveillance a.

Prior to reactor startup frm a of all relief valves is covered cold stiutdown, or' in Specification 4.6.H.

b.

When there is irradiated fuel in b.

A leak rate test of each ADS the reactor vessel and the valve acx:tnulator, check valve, reactor is.above 113 psig except and actuator assembly shall be as stated in Specification performed during each refueling 3.5.F.2.

outage at a pressure of 90+18 psig. %e leakage rate shall be verified to be6 4.5 S G H.

2.

Operation with Inoperable 2.

Surveillance with Inoperable cmponents cmponents If one of the seven ADS valves is When it is detemined that one of known to be incapable of autmatic the seven ADS valves is incapable of operation, the reactor may reain in autmatic operation, the HPCI systs operation for a period not to exceed and the actuation logic of the other seven (7) days, provided the HPCI ADS valves shall be d monstrated to

-syste is operable.

(Note that the be operable innediately and daily pressure relief function of these thereafter until all seven Ar3 valves is assured by Specification valves are capable of autmatic 3.6.H; Specification 3.5.F only operation.

applies to the ADS function).

3.

Shutdown Requirenents If Specification 3.5.F.1 or 3.5.F.2 cannot be met, an orderly shutdown will be initiated and the reactor pressure shall be reduced to 113 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9409130004 840905 3.5-9 i

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' 3.5. F.1 Normal System Availability (continued)

. Specification 3.6 -states the requirenents for the. pressure relief

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function of the valves.

It is possible for any neber of the valves c

assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures yet be fully capable of performing their pressure relief function.

Because the automatic depressurization systen does not provide makeup to the reactor primary vessel, no credit is taken for the steen cooling of-the core caused by the system actuation to provide further conservatism to the Core Standby Cooling Systens.

l Se' ADS valve accoulators are sized such that, following ' loss of the

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pnematic supply, at least two valve actuations will be possible with the drywell at -70% of its design pressure.

%is drywell pressure F

results from the largest break which could lead to the need for rapid

depressurization through the ADS valves.

ne allowable accmulator

. leakage criterion ensures the above capability for 30 minutes following loss of the pneumatic supply.

2. Operation with Inoperable Caponents h

With one ADS valve known to be incapable of autmatic operation. six valves renain operable to perform-their ADS function.

However, since the ECCS Ioss;of Coolant Accident analysis for anall line breaks assmed that all seven AD6 valves were operable, reactor operation with one ADS

. valve inoperable is only allowed to continue for seven -(7) days provided that the HPCI system is denonstrated to be operable and that the iactuation logic for the (renaining) six ADS valves is demonstrated to be

. operable.

3.;Minime Core and Containnent cooling Systens Availability

%e purpose of this Specification is to assure that adequate core cooling sluipnent is available at all times. ' If, for 'exanple, one core spray loop were out of service and the diesel which powered the opposite core spray were out of service, only 2 RHR peps would be available.

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. Specification 3.9 must also be consulted to determine other requirenants for the diesel generators.

In addition, refer to definition 1.0.00 for i

Cunulative Downtime requirements.

his specification establishes conditions for the performance of major

' maintenance, such as draining of the suppression pool. % e availability of the shutdown cooling subsysten of the RHR systen and the RHR service "wata_r system ensure adequate supplies of reactor cooling and energency

.me'wp water when the reactor is in the Cold Shutdown condition.

In aJdition this specification provides that, should major maintenance be rerformed, no work will be performed which could lead to draining the

-water fran the reactor vessel.

3.5-18

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ENCLOSURE 2 NRC DOCKET 50-366 OPERATING LICENSE NPF EDWIN I. HA701 NUCLEAR PIANT UNIT 2 imuutST TO AMEND 7ECENICAL SPECIFICATIONS The proposed change to the Unit. 2 Technical Specifications (Appendix A - to

' Operating License NPF-5) would be incorporated as follows:

Renove Page Insert Page 3/4 5-3 3/4 5-3 B 3/4 5-2 B 3/4 5-2

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BGBGDICY 00RE COOLING SYS11!MS

• 3/4.5.2 AIFKMATIC DEPRESSURIZATION SYSTEM LIMITING C31DITION FOR OPERMPION

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3.5.2 %e Autmatic Depressurization System (ADS) shall be OPERABLE with at least seven OPERABIE ADS valves.

APPLICABILITY:. CONDITIONS 1,

2 and 3 with reactor vessel stem dme pressure >150 psig.

ACTION:

a.

With 'one of the above required ADS valves inoperable, POWER OPERATION may continue provided the HPCI, GS arxl LPCI systens are OPERABIE; restore. the inoperable ADS valve to OPERABIE status within 14 days or be in at least HOT SHlTID0hN within the next 12

. hours =and reduce reactor vessel steam dme pressure to.4150 psig

'within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i b.

With two or more of the above required ADS valves inoperable, be in at least HOT SHLTIDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor stem dme pressure to 150 psig within the next-24 hours.

c..

With the Surveillance Requirenent of Specification 4.5.2.b not performed at the required ' interval due to low reactor stean pressure, the provisions of' Specification 4.0.4 are not applicable

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provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor stean pressure is adeguate to perform the tests.

SURVEILIANCE REQUIRDtENTS 4.5.2. %e ADS shall be denonstrated OPERABIE at least once per 18 months by:

~ a.

Performing a - systen functional test which includes simulated automatic actuation of the systen throughout its energency operating seguence, but excluding actual valve actuation.

b.

Manually opening each ADS valve when the reactor stean dme pressure is 2 100 psig and observing that either; 1.

'%e control valve or bypass valve position responds accordingly,. or 2.

%ere is a corresponding change in the measured stean flow.

c.

Performing a leak rate test of each ADS valve accunulator, check valve, and actuator assenbly at a pressure of 90+18 psig.

%e leakage rate shall be verified to be$4.5 SCni.

F HA101 - UNIT 2 3/4 5-3

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EMERGENCY CORE COOLING SYS' IBIS

=

BASES AU100LTIC DEPRESSURIZATION SYS'ITM (Continued)

>ADSLautmatically controls seven selected safety-relief valves although the* hasards analysis only takes credit for six valves.

It is therefore appropriate to permit one valve to be out-of-service for 14 days without materially; reducing systen reliability.

%e ADS -valve. accmulators are sized such that, following loss of the pnematic supply, at least two valve actuations will be possible with the drywell at 70% 'of _its design. pressure.

%is drywell pressure results fra the largest break which could lead to the need for rapid depressurization through the ADS valves. %e anowable acctmulator leakage criterion ensures the above capability for 30 minutes fonowing loss of the pnematic supply.

l-3/4.5.3 LOW PRESSURE CDRE COOLING SYS'IEMS

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3/4.5.3.1 CORE SPRAY SYS'IEM

% e~ core spray system (CSS) is provided to assure that the core adeguately, cooled fonowing a loss-of-coolant accident.

'Iwo subsys provide adeguate core cooling capacity for all break sizes -fra 0.2 ftpens up to and ' including the double-ended reactor recirculation line break, and for enaner breaks.fonowing depressurization by the ADS.

%e CSS -specifications are applicable during an OPERATIONAL CDOITIONS because the CSS is a primary source of.energency core co7 ling after the

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reactor vessel is depressurized and to provide a source for flooding of the core in. case of accidental draining.

When in- 00teITION 1, 2 or 3 with one CSS subsysten inoperable, the OPERABILITY of the redundant full capacity (ES subsysten and the full capacity low pressure coolant injection mode of the RHR systen provides assurance of adequate core cooling and justifies the spacified 7 day out-of-service period.

Whm in COteITION 4 or 5 with neither CSS subsysten OPERABIE, prohibition of all operations which have a potential for draining the reactor. vessel minimizes the probability of energency core cooling being required. Se required OPERABILITY of both LPCI subsystens or, in CONDITION 5 only, requiring the reactor vessel to be flooded with the fuel pool gates removed, provides assurance of adequate core flooding and the restrictions on operations are not applicable.

%e surveinance requirenants provide adequate assurance that the CSS will be OPERABLE when required. Although all active caponents are testable and full: flow can be denonstrated by recirculation during reactor operation, a coplete functional test requires reactor shutdown.

%e pmp discharge piping is maintained full to prevent water hamner danage to piping and to start cooling at the earliest moment.

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DiCIOSURE 3 lac DOCIG55 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 HNIN I. HATCH NUCLEAR PIJNP UNITS 1, 2 REQUEST TO AMEIO TECHNICAL SPECIFICATIONS Pursuant 1to 10 CFR 50.92, Georgia Power Company has evaluated the attached

. proposed amendments and has determined that their adoption would not involve

' a significant hazard.; %e basis for this determination is as follows:

a.

PROPOSED CHANGE

Add to the ADS surveillance requirenents (Unit 1 Tednical Specification 4.5.F.1 and. Unit 2 Technical Specification 4.5.2) a leak rate. test of each ADS accumulator system which is to be performed at least once per

- operating cycle. %e' test is to be performed at a pressure of 90 + 18 psig and the. acceptance criterion for leakage is to be 4.5 SCEE or less.

BASIS'

%is change constitutes an additional restriction not presently included

'in the Technical Specifications.

%is change does not affect the probability or consequences of an accident or malfunction analyzed in.

the FSAR. % is change does not create the possibility of an. accident or malfunction.of. a different type than any analyzed in the FSAR.

The margin of safety as defined in the basis for any Technical Specification ils not affected by this change.

%e effect of this change is therefore

-within the acceptance criteria and the change is consistent with Itan (ii) of " Examples of Junendnents that are Considered Not Likely to

' Involve Significant Hazards Considerations" listed on page 14,870 of the

. April 6,1983 issue of the Federal Register.

b.

PROPOSED CHANE:

Change bases to reflect the above change.

-BASIS his is a purely administrative change to the Technical Specifications.

i This change does not ' affect the probability or consequences 'of an 4

accident or-malfunct. ton analyzed in the FSAR.

%e margin of safety as defined ~ in. the basis for any Tednical Specification is not affected.

The effect of this change is therefore within the acceptance criteria and the change - is consistent with Item (i) of the " Examples of Junendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6,1983, issue of the Federal Register.

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