ML20050C810

Proposed Tech Spec Changes Re Emergency Relief from Drywell Inerting Limiting Condition for Operation,Time Extension for Rod Pattern Adjustments,Placement of Rod Block Clamp at 107% Rated Power & Extension for Operable LPCI Inverter

ML20050C810
Person / Time
Site:	Hatch
Issue date:	04/05/1982
From:	GEORGIA POWER CO.
To:
Shared Package
ML20050C807	List: ML20050C806 ML20050C810
References
TAC-47885, TAC-48295, TAC-48297, NUDOCS 8204090372
Download: ML20050C810 (14)

Text

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ATTACHENT 1 The proposed change to the Technical Specifications is confiImatory in nature; emergency relief was granted for Unit 1 on February 8, 1982. The attached pages document the basis for the requested, and approved, change.

Technical Specification 3.7. A.5.b (Unit 1) requires "within the 24-hour period subsequent to placing the reactor in the RUN mode following a shutdown, the ccntainment atmosphere oxygen concentration shall be reduced to less than 4% by volume and maintained in this condition." However, during the startup in progress on February 7, it became apparent that there was an insufficient amount of liquid nitrogen available to meet this 24-hour LCO, and that resupplies were not available prior to LCO expiration.

However, minimum levels (2000 gallons) of nitrogen were maintained to preserve operability of the Containment Atmosphere Dilution (CAD) system per Technical Specification 3.7. A.6.b. At that time a request was made to the NRC for an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, for a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> subsequent to entering the RUN mode, before oxygen concentration had to be reduced below 4% by volume.

The Plant Review Board and Safety Review Board have reviewed the proposed change and determined that it does not involve an unreviewed safety question. Because the 24-hour LCO is strictly an administrative control, as opposed to providing a safety function, margins of safety are not reduced and new accidents or malfunctions not previously analyzed are not created by extending the allowable time for inerting completion. The safety analysis report uses the CAD system as the basis for prevention of the combustion of gases in the drywell following a LOCA; and since the CAD system remained operable, the prooability or consequences of an accident or malfunction of equipment important to safety previously analyzed in the FSAR is not increased by this proposed change.

Your excellent cooperation with our staff in the expeditious resolution of the request was greatly appreciated. Oxygen concentration was brought below 4% in the additional time granted, averting an unnecessaIy shutdown cycle on the unit.

Included in this attachment are instructions for incorporation of the proposed changes and copies of affected pages.

8204090372 020405 PDR ADOCK 03000321 p PDR

ATTACENENT 1 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS l

The proposed change to t': Technical Specifications (Appendix A to Operating r

License DPR-57) would be incorporated as follows:

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1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREENTS 3.7.A.5 0xygen Concentration 4.7.A.5 0xygen Concentration

a. After completion of the shrtup The primary containment test program and demonst:ation oxygen concentration shall of plant electrical output, the be measured and recorded primary containment atmosphere daily in the main control shall be reduced to less than room.

4% oxygen with nitrogen gas dur-ing reactor power operation with reactor coolant pressure above 100 psig, except as stated in Specification 3.7.A.S.b.

b. Within the 24-hour

• period sub- *A 72-hour period is allowed sequent to placing the reactor for the startup in progress

! in the Run Mode following a shut- on February 7,1982.

, down, the containment atmosphere oxygen concentration shall be re-duced to less than 4% by volume and maintained in this condition. De-inerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.

6. Containment Atmosphere Dilution (CAD) 6. Containment Atmosphere Dilution (CAD)

a. Operability Requirements a. Functional Test After completion of the startup test The post-LOCA Containment program and demonstration of plant Atmosphre Dilution (CAD) electrical output and thereafter System shall be functionally whenever the reactor is in power tested once per operating operation, the post-LOCA containment cycle.

Atmosphere Dilution (CAD) System must be operable and capable of supplying nitrogen to the primary containment for dilution if required

. by post-LOCA conditions. If this specification cannot be met, the system must be restored to an operable condition within seven days or the reactor must be taken out of power operation.

b. Seven-Day rutrogen Supply Seven-Day Nitrogen Supplv i

After completion of the startup test The level in the li v id

, program and demonstration of plant nitrogen storage tanks electric output and thereafter shall be recorded twice

, whenever the reactor is in power operation, the CAD System shall con-tain a minimum of 2000 callons of liquid nitrogen. If this specification cannot be met, the minimum volume will be restored within seven days or the reactor must be taken out of power operation.

3.7-9

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Whenever the reactor is in power Instrumentation Survellance is operation, there shall be at least listed in Table 4.2-11.

one CAD System H2 and 02 analyzer serving the primary containment.

If ore H2 and 02 analyzer is in-operable, the reactor may remain in operation for a period not to exceed seven days.

d. Post-LOCA Repressurization Limit The maximum post-LOCA primary con-tainment repressurization limit allowable using the CAD System shall be 30 psig. Venting via the SGTS to the main stack must be initiated at 30 psig follouing the initial post-LOCA pressure peak.

7. Drywell-Suppression Chamber, 7. Drywell-Suppression Chamber Differential Pressure Differential Pressure Differential pressure between the The pressure differential drywell and suppression chamber shall between the drywell and be maintained equal to or greater suppression chamber shall than 1.5 psid except as specified in be recorded once each shift.

(1) and (2) below: If this specifica-tion cannot be met, and the olfferential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.

1) This differential pressure shall be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> * *A 72-hour period is allowed after having placed the Mode for the start-up in progress Switch ir, the RUN mode. The dif- on February 7,1982.

ferential pressure may be removed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to achiev-ing a shutdown.

2) This differential pressure may be decreased to less than 1.5 psid for a maximum of four hours during required operability testing of the HPCI system pump, the RCIC system pump, and the dIywell-pressure suppression chamber vacuum breakers.

3.7-10

ATTACFf4ENT 2 The proposed change would extend the amount of time allotted for corrective actions to restore the APRM flow referenced simulated thermal power scram and control rod block trip setpoints to within allowable values. Present requirements call for a reduction in thermal power if the necessary pattern adjustments are not completed in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Making pattern adjustments and determining the affects of these adjustments are iterative processes which are difficult to complete within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The proposed change would extend this time limit to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The addition of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> will permit better determination of the base distribution and the thermal limits. All requirements to adjust APRM trip levels will remain the same.

The Hatch Unit 1 Technical Specifications place no time limit on this corrective action.

The Plant Review Board and Safety Review Board have reviewed the proposed change and determined that it does not involve an unreviewed safety question. The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety are not increased above those analyzed in the FSAR due to this change. The possibility of an accident or malfunction of a different type than analyzed in the FSAR does not result from this change because no new modes of operation are introduced.

Included in this attachment are instructions for incorporation of the proposed change and copies of affected pages.

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i ATTACt+ENT 2 i

NRC DOCKET 50-366 i l

! OPERATING LICENSE.tFF-5 EDWIN I. HATCH NUCLEAR PLANT UNIT 2 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed change to Technical Specifications (Appendix A to Operating License NPF-5) would be incorporated as follows:

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POWER D15TRIBUT10'i LIMITS

}/4.2.2 APRM SETPOINTS LIMITING CONDIT10'1 FOR OPERAT10fl 3.2.2 The APRM flow referenced simulated thermal power scram trip set-point (S) and cor. trol rod block trip setpoint (SRB) shall be established

• according to the following relationships:

5 i (0.66W + 51".)

Spg 1 (0.66W + 42L) where: 5 and S pr a re in percent of RATED THERMAL POWER, and W = Loop' recirculation flow in percent of rated flow.

ADPLICABILITY: C.0l;D1T10N 1, when THERMAL POWER 1 25% of RATED THERMAL P0HE r. .

I C T I C.*.

t it h 5 or 5-. e n:eecine the allov atle valur .1r.111 ate corrective action ri t ni n 15 riihu t e 3 anc continue corrective action so that 5 and S with1r the reoutrec lir.its* within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce THERMAL POWtn EEto" l less than 25' of F.ATED THERFAL l'0WER within the r. ext 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURvf 1LLAt:CE FEC*I EMENT5 4.2.2 The CMFiF2 shall be deternir.ed and the #PP.r flow referenced simulated ther El povier scram ar.d control rod block trip setpoints or LPRM readincs ad,usted, as required:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Whenever THERM L POWER has been increased by at least 15! of RATED THERP.AL POWER and steady state operating conditions have been established, and

c. Initially and at least once per 12 hcurs when the reactor is operatine with a CMFLPD 3 FRTP.

7'ith t CORE MAA!M.M FRACTION OF LIM; TING POWER del;ilTY (CMFLPD) greater than the fracticn of RATED THERMAL POVER (FRTP), ,

up to 951 of R;!ED THERP%L POWER, rather than adjusting the APRM setpoints, the APRM gain may be ad,)usted such that APRM readings are creater than or coual to 1001 times CHFLPD, provided that the adjusted APRM reading does not exceed 100', of RATED THERMAL POUER and the require-gein adjustment increment does not exceed 101 of RATED THERMAL POWER.

HATCH - Uti1T 2 3/4 2-5

ATTACfNENT 3 The proposed change would place a rod - block clamp at 107% of rated power. This would prevent operation above 107% power when the recirculation flow rate is above its rated value. Below 100% rated core flow the present

- RBM flow referenced rod block already prevents an unsatisfactorily high power level. Core analysis for Unit 1 has determined tnat a rod block at j this power level is necessary to protect the fuel integrity in an above rated core flow condition. This analysis, which is based on a Rod

' Withdrawal Error transient is contained in Unit 1 - Reload 4 Supplemental

! Licensing Submittal (Y1003J01A13) dated July 1980 and the Unit 1 - Cycle 5

Safety Review of Conditions Above Rated Flow (NEDO-24308-1, rev. 1) dated July 1981.

The Plant Review Board and Safety . Review Board have reviewed the proposed change and have determined that it does not involve an unreviewed safety question. The probability and consequences of an accident or

malfunction involving equipment important to safety are not increased above those analyzed in the FSAR, because this change would actually create a more conservative restriction on outward control rod motion. However, since normal operations do not occur above 100% rated power, this rod block would not affect normal control rod manipulations. The possibility of an accident or malfunction different from those analyzed in the FSAR does not result t

from this change since this rodblock would create no new failure modes as it would not affect art / of the present reactor flux limitations. Furthermore, the Technical Specifications margin of safety is not reduced since this change would represent an increase in the thermal limit margins for operating conditions above rated flow, and would not affect any of the present rod block or reactor trip setpoints.

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ATTACHENT 3 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed changes to Technical Specifications ( Appendix A to Operating License DPR-57) would be incorporated as follows:

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Table 3.2-7 (Continued)

Required Operable Ref. Trip Channels '

No. condition Per Trip

(a) Instrument Nomenclature System (b) ' Trip Setting Remarks 3 APRM Downscale 2(e) 3/125 of full scale Not required while performing low

, power physics test at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.

12% Flux 2(e) 12/125 of full scale This function is bypassed when the Mode Switch is placed in the RUN position.

High Flux 2(e) 0.66W + 42% W is , the loop recirculation flow i

." rate in percent of ' rated. Trip 4

';* level'. setting is, in percent -of

;;; rated power. Not/ required ,while'

) performing low power ph/sics fests' at atmospheric pressure during . ~ or after refueling.at power [ levels not to exceed 5 MWt. '

s l A RBM Inoperative 1(e)(f) Not,applicatile , Inoperative trio produc'e@ by switch i not in operate, circuit ' boards not in circuit, fails.. to null, less than required number 'of LPRM inputs 1 ,

.for rod selected. >

s 3 4-i :Downscale i 1(e)(f) 3/125 of full scale  ?

'(g)(h) ,  ;

High Flux 1(c)(f) , 0.66d 3.741% W is the loop recirculation flow

- .Not to exceed 107% l rate in pe: cent of ~ rated. Trip'

,~i "4 level- setting is .in percent of s

, e rated thermal power. ,

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ATTACHENT 4 This proposal' would change the action for Unit 2 Technical Specifications 3.8.2.1.e to allow one inverter to be inoperable for a period 1

not to exceed 7 consecutive days providing the other inverter is operable.

Currently , action step a. calls for an inoperable inverter to be I returned't'o an operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 ' hours and in COLD SHUT 00WN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.- The inverters provide the normal puer for motor control centers

. (MCC) 2R24-S018 A&B which supply power to 1.PCl injection valves 2E11-F015A &

8, respectively. These injection valves r..ust be operable in order for their I -

recpective~ RHR loops to be considered operable. Since the limiting condition for operation _(LCO) for the R$ loops on an indivio_31 basis is 7 consecutive days, the LCO -for the LPCI inverters should not be more restrictive than the system whose operability they affect. Furthermore, the existing LCO precludes normal maintena. ce for the inverters. For these reasons this action step should be changed as proposed.

. Thb imple$entation of this change does not constitute an unreviewed safety question (10 CFR 50.59) as explained below.

_ .The probability of occurrence and the consequences of an accident or

• "_ malfunction of equi & ment important to safety are not increased above those analyzed in FSAR due to this change because the inverters are only used for l

Rm system operation during emergencies, and the R$ system was analyzed in

! the FSAR.

The possibility of an accident or malfunction of a different type than that analyzed in the FSAR does not result from this change because a failure of a RHa system was analyzed and the failure of an inverter could result in an identical condition.

,- The margin of safety as defined in the Technical Specification is not reduced due to this change because the applicable safety margin for the

~1nverters should be the same as that for the RHR system, since the only safety function of the inverters is to supply RHR emergency power.

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ATTACHMENT 4 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN I. HATCH ffJCLEAR PLANT UNIT 2 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed change to Technical Specifications (Appendix A to Operating License NPF-5) would be incorporated as follows:

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'ELECTRXCAL F0 DER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING C0tOITION FOR OPERATION 3.8.2.1 The following A.C. distribution system buses, inverters and motor-generator (MG) sets shall be OPERABLE with breakers open between redundant buses:

a. 4160 volt Essential Buses 2E, 2F, and 2G,

b. 600 volt Essential Buses 2C and 20,

c. 120/208 volt Essential Cabirets 2A and 28,

d. 120/208 volt Instrument Buses 7A and 28, and

e. A.C. Inverters 2R44-S002 and 2R44-5003.

APPLICABILITY: CONDITIONS 1, 2 and 3 ACTION:

a. With one of the inverters in 3.8.2.1.e inoperable, restore the inverter to an OPERABLE status within a period not to exceed seven (7) consecutive days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With one of the above required A.C. distribution system buses; inoperable, restore the inoperable bus to OPERABLE status within 88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. With two or more of the above required A.C. distribution system buses l or inverters inoperable, restore at least all except one of the inoperable buses and inverters to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.1 The aDove required A.C. distribution system buses and inverters shall be detennined OPERABLE:

a. At least once per 7 days by verifying correct breaker alignment and indicated power availability, and

b. At least once per 31 days be determining that the 250 volt DC/600 volt AC inverters 2R44-5002 and 2R44-S003 are OPERABLE by verifying inverter output voltage of 600 volts 1 5 while supplying their respective buses.

HATCH - UNIT 2 3/4 8-10 Amendment No. 23

h ATTACHMENT 5 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, rPF-5 EDelIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS

)

Pursuant to 10 CFR 170.22(c), Georgia Power Company has evaluated the i attached proposed amendment to Operating Licenses OPR-57 and NPF-5 and hp detennined that:

a) The proposed amendment does not require the evaluation of a new Safety Analysis Report or rewrite of the facility license; b) The proposed amendment does not contain several complex issues, does not involve ACRS review, and does not require an environmental impact statement; c) The proposed amendment does involve more than one safety issue, namely ; 1) relief from drywell inerting LCO for Unit 1, 2) extension of time for corrective actions for making rod pattern adjustments for Unit 2, 3) addition of RBM rod block clamp at 1077.

of rated power, and 4) extension of the LCO for an inoperable LPCI inverter for Unit 2.

d) The proposed amendment is therefore a single Class XV emendment.