ML20212Q833

Proposed Tech Specs Re RWCU Leakage Detection Sys Setpoints for Area temp-high Instruments

ML20212Q833
Person / Time
Site:	Hatch
Issue date:	08/25/1986
From:	GEORGIA POWER CO.
To:
Shared Package
ML20212Q805	List: ML20212Q803 ML20212Q833
References
1152T, TAC-62757, TAC-62758, NUDOCS 8609080064
Download: ML20212Q833 (7)

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Georgia Power A ENCLOSURE 1 NRC DOCKETS 50-321 AND 50-366 OPERATING LICENSES DPR-57 AND NPF-5 EDWIN I. HATCH NUCLEAR PLANT-UNITS 1 AND 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS CHANGES T0 RWCU LEAXAGE DETECTION SYSTEM BASIS FOR CHANGE REQUEST-Currently the Plant Hatch Units 1 and 2 Technical Specifications specify an allowable value of <. 1240F for the Reactor Water Cleanup (RWCU) aubient temperature-high allowable value. This value can be found in Table 3.2-1 of the Unit 1 Technical Specifications and in Table 3.3.2-2 of the Unit 2 Technical Specifications. This value was derived, for both units, from an analytical limit of 1300F.

The current plant setpoint of 1200F (developed from the above allowable value) has repeatedly resulted in undesirable isolations of the RWCU system due to high ambient temperatures in the RWCU room. In order to minimize this adverse impact on plant operation, GPC has performed an analysis to justify a higher analytical limit (and resultant allowable value) for this isolation function. This proposed analytical limit of 1570F has been analyzed and found to be acceptable from the standpoints of isolation capability, prevention of inventory loss, and equipment qualification.

The design basis for RWCU isolation is to provide protection against incipient pipe rupture (leak before break). General Electric has provided documentation which shows that the plant is designed to withstand a break of the largest pipe in the RWCU system, resulting in an inventory loss of 6624 gallons prior to system isolation. This inventory loss value provides a bound for selecting time intervals, and thus temperature setpoints', at which RWCU will isolate for smaller pipe breaks. The temperature setpoints are based on prevention of excessive inventory loss in the event of a pipe break resulting in 25 gpm or greater leakage into the RWCU room. For postulated pipe breaks resulting in leakage cf less than 25 gpm, an alarm function, set at a lower temperature than the isolation function, is provided in the control room.

For the RWCU system isolation function, the inventory loss associated with the proposed higher temperature limit will be greater due to the delayed isolation. However, the potential additional inventory loss associated with the new limit, for a 25 gpm break, has been analyzed and found to be well within the 6624 gallon limit. Based' strictly on this design criteria, an analytical limit of up to 1800F would be acceptable. However, environmental qualification requirements constrain the proposed analytical limit to 1570F, El-1 1152t 08/25/86 8609090064 860825 PDR ADOCK 05000321 P PDR ,

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l Georgia Power d ENCLOSURE 1 (Continued) i NRC DOCKETS 50-321 AND 50-366 OPERATING LICENSES DPR-57 AND NPF-5 EDWIN I. HATCH NUCLEAR PLANT-UNITS 1 AND 2 RECUEST TO REVISE TECHNICAL SPECIFICATIONS CFANGES TO RWCU LEAKAGE DETECTION SYSTEM BASIS FOR CHANGE REQUEST l

i as explained below. Therefore, this new limit of 1570F reduces the l conservatisms associated with the value currently in the Technical Specifications while maintaining the original design basis.

The only equipment in the RWCU room which performs a safety function, and thus

! requires environmental qualification, is the instrumentation which performs i the RWCU isolation function. In the event of a postulated break, delayed l isolation at the higher temperature limit creates a harsher environment in the l RWCU room prior to and following isolation valve closure. The. proposed analytical limit of 1570F was analyzed and found to assure that l

environmental qualification requirements for this equipment were not exceeded. Environmental qualification is the limiting factor in determination of the revised limit.

Using the analytical limit of 1570F, GPC has applied Regulatory Guide 1.105 criteria and has obtained an allowable value of 1500F for the Units 1 and 2 Technical Specifications.

El-2 1152t 08/25/86 7C3775

I GeorgiaPower d l

1 ENCLOSURE 2 l NRC 00CKETS 50-321 AND 50-366 OPERATING LICENSES DPR-57 AND NPF-5 EDWIN I. HATCH NUCLEAR PLANT-UNITS 1 AND 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS CHANGES TO RWCU LEAKAGE DLItCTION SYSTEM 10 CFR 50.92 EVALUATION Pursuant to 10 CFR 50.92, Georgia Power Company has evaluated the aitached proposed amendments for Plant Hatch Units 1 and 2 and has determined that their adoption would not involve a significant hazard. The basis for this determination is as follows.

Proposed Change:

Provide for adjustments to allowable values for the Reactor Water Cleanup (RWCU) leakage detection system.

Basis:

Georgia Power Company has reviewad the proposed change and has determined that it does not involve a significant hazards consideration for the following reasons:

The proposed change does not significantly increase the probability or consequences of an accident previously evaluated because the change has been determined to be conservative with regard to the plant design basis. Evaluations have been performed to demonstrate acceptability of the new limit in regard to the design criteria of isolation on a 25 gpm pipe break, to the potentici additional inventory losses due to raising the limit, and to the environmental qualification requirements for the equipment in the area.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because this

' change does maintains the function of the RWCU isolation system and does not result in new modes of plant operation.

The proposed changes do not involve a significant reduction in a margin of safety because the design basis of the plant is being maintained.

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I GeorgiaPower A ENCLOSURE 2 (Continued)'

NRC DOCKETS 50-321 AND 50-366 OPERATING LICENSES DPR-57 AND NPF-5 EDVIN I. HATCH NUCLEAR PLANT-UNITS 1 AND 2 REC'UEST TO REVISE TECHNICAL SPECIFICATIONS U TNGES TO RWCU LEAKAGE DEIECTION SYSTEM 10 CFR 50.92 EVALU'ATION This change _ is consistent with item (vi) of the " Examples of Amendments That l are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the Federal Register, April 6,1983. Example (vi) of actions involving no significant hazards consideration is a change which may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria. This change raises the current Plant Hatch Units 1 and 2 Technical Specifications value from 4- 1240F to 1500F.

' This change is being proposed to reduce the potential of spurious isolations of the RWCU rooms. This change is within the original design basis of the plant which specifies that isolation of these rooms should occur in the event of a 25 gpm pipe break. The increased time to isolation has been analyzed by General Electric and the potential additional inventory losses were found to

- be well within acceptable limits. An evaluation lias also determined that there is no impact on the environmental qualification results for the associated equipment in those areas.

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Georgia Power d ENCLOSURE 3 NRC DOCKETS 50-321 AND 50-366 OPERATING LICENSES DPR-57 AND NPF-5 EDWIN I. HATCH NUCLEAR PLANT-UNITS 1 AND 2 ,

REQUEST TO REVISE TECHNICAL SPECIFICATIONS lTAETS TO RWCU LEAKAGE DETECTION SYSTEM PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS The proposed changes to the Units 1 and 2 Technical Specifications

( Appendices A to Operating Licenses DPR-57 and NPF-5) would be incorporated as follows:

1 Unit Remove Page Insert Page 1 3.2-3 3.2-3

. 2 3/4 3-17 3/4 3-17 a

1152t 08/25/86 700775

Table 3.2-1 (Cont.)

z:

3= Requi red i$ Ref. Trip Dpe rab le Action to be taken If oc No. Instrument Condition Channels Trip Setting number or channels is *

, (a) Nomenclature per Trip not met for both trip Remarks (4)

Systeo (b) systems (c)

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,, 4 Main Steam Line High 2 53 times normal Initiate an orderly losd initiates Group 1 Radiation full power back- reduction and close MSIVs isolation.

g round ao within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. -

5 Main Steam Line Low 2 2825 psig Initiate an orderly load Initiates Group 1 Pressure reduction and close Isointion. Only .

MSIVs within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, required in RUN mode, therefore activated when Mode Switch is in RUN position.

6 Main Steam Line High 2 5138% rated flow initiate an orderly load initiates Group 1 Flow (5115 psid) reduction and close MSIVs isolation, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

7 Main Steam Line High 2 5194*F Initiate an orderly load Initiates Group 1 Tunnel Temperature reduction and close MSIVs isolation, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

I# 8 Reactor Water High 1 20-80 gym Isolate reactor water Final trip setting DO Cleanup System ., cleanup system. will be determined Dif ferential Flow during startup test

[a p rog ram.

9 Reactor Water High 2 5150 F isolate reactor water Cleanup Area cleanup system.

Tempe ra tur e 10 Reactor Water High 2 567'F isolate reactor water Cleanup Area c!eanup system.

Ventilation Differential Tempe ra ture 11 Condenser Vacuum Low 2 27" Hg. vacuum initiate an orderly load- Initiate Group 1 reduction and close MSIVs isolation within 8 hrs.

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3 IABLE 3.3.2-2 (C ntinued) I ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 3=

-4 o All OWABLE Z *

' TRIP SETPOINT VALUE __

TlitP FUNCTION c -

3. REACTOR WATER CLEANUP SYSTEM ISOLATION

-A 5 79 gpe $ 79 gpm g

a. A Flow - High Area Temperaturo-High s 150"F s 150 F b.

Area Ventilation A Temperature - High 5 67*F ' 5 67'F c.

NA NA

d. St.CS Initiation 2 -55 inches * ' 2 -55 inches *

e. Reactor Vessel Water Level-Low Low (Level 2)

4. HICH PRESSURE COOLANT INJECTION SYSTEM ISOLATION HPCI Steam Line Flow-High i 307% of rated Flow 1 307% OF rated Flow

a. 2 100 psig 2 100 psig

b. HpCI Steam Supply Pressure - Low

c. HFCI Turbine Exhauss Diaphragm 5 20 psig i 20 psig -

Pressure-High

d. HPCl Pipe Penetration Room 5 169'F Temperature - High 5 169'T

e. Suppression Pool Area Ambient 5 169'F $ 169*F M T empe ra tu re-H i gh < 42.5'F A f. Suppression Pool Area AT - High < 42. 5'.F 9 Suppression Pool Area Temperature NA w T imer Relays ' NA e

h. Emergency Area Cooler Temperature'- 5 169'F $ 169'F.

Q High 5 1.85 psig 5 1.85 psig

i. Drywell Pressure - High NA Logic Power Bus Monitors NA J.

'See Bases Figuro B 3/4 3-1. ,

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