Text

Amends 149 & 86 to Licenses DPR-57 & NPF-5,respectively, Revising Tech Spec Reporting Requirements for 10CFR50.72 & 50.73 & Primary Coolant Iodine Spiking
	ML20237C775
Person / Time
Site:	Hatch
Issue date:	12/01/1987
From:	Jabbour K; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20237C778	List: ML20237C775; ML20237C779;
References
	GL-83-43, GLF-5-A-86, TAC-55713, TAC-55714, NUDOCS 8712220165
	Download: ML20237C775 (69)
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UNITED STATES NUCLEAR REGULATORY COMMISSION.

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E W ASHINGTON, D. C. 20555

%J GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE

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Amendment No. 149 License No. DPR-57 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant.

Unit 1 (the facility) Facility Operating License No. DPR-57 filed by Georgia Power Company, acting for itself Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated February 13, 1987, complies with the standards and req)uirements of the Atomic Energy Act of1954, as a lations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8712220165 871201 PDR ADOCK 05000321 l

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. 2. -Accordingly, the license is amended by changes to the Technical Specif1-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.149, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Kahtan N. Jabbour, Acting Director Project Directorate II-3 Division of Reactor Projects-1/II

Attachment:

Changes to the Technical Specifications Date of Issuance:

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. ATTACHMENT TO LICENSE AMENDMENT NO.149 FACILITY OPERATING LICENSE NO. DPR-57 1

DOCKET NO. 50-321 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment nunber and contain vertical lines indicating the areas of change.

Remove Insert Page Page vi vi 1.0-10 1.0-10 3.6-4 3.6-4 3.6-5 3.6-5 3.6-9 3.6-9 3.7-6 3.7-6

-3.7-6a 3.7.6a 3.11-2a 3.11-2a 3.11-5 3.11-5 6-7 6-7 6-10 6-10 6-12 6-12 6-13 6-13 1

6-14 6-14 6-15

'6-15 i

6-15b 6-15b 6-15d 6-15d q

6-16 6-16 6-17 6-17 6-18 6-18 Replace the following pages of the Appendix 8 Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

4-1 4-1 4-2 4-2 5-8 5-8 i

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e Section Section Pace LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.14 RADIOACTIVE EFFLUENT 4.14 RADIOACTIVE EFFLUENT INSTRUMENTATION INSTRUMENTATION 3.15 RADIOACTIVE EFFLUENT 4.15 RADI0 ACTIVE EFFLUENT CONCENTRATION AND DOSE CONCENTRATION AND DOSE 3.16' ENVIRONMENTAL 4.16 ENVIRONMENTAL MONITORING PROGRAM MONITORING PROGRAM 5.0 MAJOR DESIGN FEATURES 5.0-1 5.0-1 A.

Site 8.

Reactor Core 5.0-1 C.

Reactor Vessel

5. 0-1 0.

Containment

5. 0-1 i

E.

Fuel Storage 5.0-1 F.

Seismic Design 5.0-2 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Responsibility 6-1 6.2 Organization 6-1 6.3 Unit Staff Qualifications 6-6 6.4 Training 6-6 6.5 Review and Audit 6-6 6.6 Reportable Event Action.

6-13 l 6.7 Safety Limit Violation 6-13 6.8 Procedures 6-14 6.9 Reporting Requirements 6-15 6.10 Record Retention 6-23 6.11 Radiation Protection Program 6-25 l

6.12 High Radiation Area 6-25 i

6.13 Integrity of Systems outside Containment

6-26 6.14 Iodine Monitoring 6-27 I

6.15 Post Accident Sampling and Analysis 6-27 l

t 6.16 Offsite Dose Calculation Manual 6-27 l

l HATCH - UNIT 1 vi Amendment No. 149

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1.0 DEFINITIONS (Continued) i EEE. MILK ANIMAL i

A cow or goat that is producing milk for human consumption.

i FFF.

DOSE EOUIVALENT IODINE The DOSE EQl'IVALENT I-131 shall be that concentration of I-131 (microcurie / gram), which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in table III of TID-14844 or those in NRC Regulatory Guide 1.109 Revision 1, October 1977.

GGG. ACTION ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

HHH.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

III. STAGGERED TEST BASIS a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.

b.

The testing of one system, subsystem, train or other designated components at the beginning of each subinterval.

JJJ. REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

l HATCH - UNIT 1 1.0-10

' Amendment No. 149

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LIMITING CONDITIONS FOR OPERATION SUPv{lttANCE REQUIREMENT 5 3.6.F.

Reactor Coolant themistry 4.6.F.

Reactor Coolant themistrv 1.

Radioactivity 1.

Radioactivity Whenever the reactor is critical the limits on activity concen-During equilibrium power s.

trations in the reactor coolant operation an isotopic shall not exceed the equilibrium analysis, including value of 0.2 wci/gm of dose quantitative measurements equivalent

I-131.

for at least 1-131. 1-132, 1-133, and 1-135 shall be performed monthly on a coolant liquid sample, b.

During equilibrium power operation an isotopic analysis, including quantitative measurements for at least Xe-133 and Xe-135 shall be performed monthly on a steam jet air ejector of f-gas sample, Additional coolant samples c.

shall be taken whenever the reactor coolant dose equiva-lent I-131 concentration exceeds 0.2 9 C1/gm and any if activity concentration >0.2 i

w C1/gm dose equivalent 1-131 but of the following conditions are met:

during one continuous time interval,t 4.0 v C1/gm for more

1) During startup or >4.0 Ci/ge, be in at least NOT SHUTOOWN with the main steam line2) Following a power change isolation valves closed within 12 exceeding 15% of rated hours.

thermal power in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months averaged for(net change

That I-131 concentration which alone would 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ),

produce the same thyroid dose as the

3) The off-gas level quantity and iodine mixture actually the SJAE increase,s by at

present, more than 10,000 wCt/sec in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at release rate 1 75,000 pC1/see, or

4) The off-gas level at the SJAE,, increases by more than 155 in I hour at release rate > 75,000 wCi/sec.

5) Whenever the reactor coolant dose equivalent I-13) concentration ex-ceeds 4.0 wC1/ge.

HATCH,- UNIT 1 3.6-4 Amendm nt No. 149

LIM 111NG CONDifl0NS FOR OPERATION SURV[lLLANCE REOUIRCMENTS The first additional coolant sample shall be taken between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following the change in therwel power or off-gas level. Additional coolant liquid samples shall be taken at 4-hour intervals for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or until a stable iodine concentra-tion below the limiting value of 4.0 pCi/gm is established. An isotopic analysis shall be performed for each sample, and quantitative measurements made to determine the dose equivalent I-131 concen-tration. If the total iodine activity of the sample is below 0.2 pC1/ge, an isotopic analysis to determine equivalent I-131 is not required.

All data obtained from normal and any additional samples shall be included in the annual report.

1 HATCH - UNIT 1 3.6-5 kmendment No. 149

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4 llMITING CONDITIONS FOR OPERATION SURV[]lLANCE REQUIREMENTS 3.6.H.1.

Relief / Safety Valves 4.6.H.1.

Relief / Safety Valves a.

When one or more relief / safety a.

End of Operatina tvele valve (s) is known to be f ailed *** an l orderly shutdown shall be initiated Approximately one-half of all and the reactor depressurized to relief / safety valves shall be less than 113 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , benchchecked or replaced with Prior to reactor startup from a a benchchecked valve each re-cold condition all relief / safety fueling outage. All 11 valves valves shall be operable.**

will have been checked or re-placed upon the completion of every second operating cycle.

b.

With one or more relief / safety b.

Each Operatina Cvele valve (s) stuck open, place the reactor mode switch in the shutdown position.

Once during each operating cycle, at a reactor pressure

> 100 psig each relief valve shall be manually opened until thermocouple downstream of the valve indicate steam is flow-ing from the valve.

j c.

With one or more safety / relief valve c.

Intearity of Relief Valve tailpipe pressure switches of a Bellows

saf ety/ relief valve declared inoperable and the associated The integrity of the relief valve safety / relief valve (s) otherwise bellows shall be continuously indicated to be open, place the monitored and the pressure j

reactor mode switch in the Shut-switch calibrated once per I

down position.

operating cycle and the accu-mulators and air piping shall be inspected for leakage once per operating cycle, d.

With or.e safety / relief valve tailpipe d.

Relief Valve Maintenance pressut; switch of a safety / relief valve declared inoperable and the asso-At least one relief valve shall ciated safety / relief valve (s) otherwise be disassembled and inspected indicated to be closed, plant operation each operating cycle, may continue. Remove the function of

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that pressure switch from the low low e.

Doerability of Tailoioe set logic circuitry until the next COLD Pressure Switches SHUTDOWN. Upon COLD SHUTDOWN, restore the pressure switch (es) to OPERABLE The tailpipe pressure switch status before STARTUP.

of each relief / safety valve shall be demonstrated operable e.

With both safety / relief valve tailpipe by performance of a:

pressure switches of a safety / relief valve declared inoperable and the asso-1.

Functional Test:

ciated safety / relief valve (s) otherwise indicated to be closed, restore at least a.

At least once per 31 i

one inoperable switch to OPERABLE status days, except that all l

within 14 days or be in at least HOT portions of instrumen-SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months tation inside the prl-and in COLD SHUTDOWN within the mary containment may be following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

excluded from the functional test, and "Does not apply to two-stage Target Rock SRVs

- The Reflief/ Safety valves are not required to be operable for performance of inservice hydrostatic or pressure testing with reactor pressure greater than 113 psig and all control rods inserted. Overpressure protection will be provided as required by ASME Code.

- - The failure or malfunction of any safety / relief valve shall be reported by telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office or his designee no later than the first working day following the event; and a written follow-up report within 30 days. The written follow-up report should be completed in accordance with 10 CFR 50.73 or other applicable requirements.

HATCH - UNIT 1 3.6-9 Amendment No. 149

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0UIREMENTS 4.' 7. A. 2. e. Tvo c B Te s t - Leak Tests of Pen 3-trations with Senis and Bellows Tcontinued)LTables 3.7-? and 3.7-3)

(1) Primary containment components which seal or penetrate the pressure containing boundary of the containment shall be tested at a pressure not less than P These components shall$e. tested at each major refueling shutdown or at intervals not to exceed two years.

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(2) (a) The personnel air lock shall be tested at in-I tervals not to exceed j

six months at P by pres-5 a

surizing the compartment between the.two air lock doors.

During intervals of door use when containment integrity is required, the door seals shall be tested at 10 psig after each opening.

(b) Personnel air lock leakage shall not exceed 0.05 La-f.

Type C Tests-Lot;. Leak Tests of Containment Isolation Valves-(Tables 3.7-1 and 3.7-4)

Type C tests shall be performed under the program established in Appendix J of 10 CFR Part 50.

Containment isolation valves (except for main steam line iso-lation valves) shall be test.ed at a pressure not less than P

a Type C tests shall be performed at each major refueling shutdown or at intervals not to exceed two years.*

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All Type B and Type C Leakage Tests (i.e., Local Leak Rate Tests) that fail (i.e., test leakage is such that an LER would be required) during an outage shall be reported according to 10 CFR 50.73 by one, 30-day written report that is due within 30 days of the first leakage test failure in the outage.

All other leakage test failures discovered during the outage will be reported in a revision to the original report due within 30 days af ter the end of the outage.

HATCH - UNIT 1 3.7-6 Amendment No. 149

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS g.

Acceptance Criteria for Tvoe B i

and TvDe C Tests The combined leakage rate of components subject to Type B and C tests shall be determined under the program established in Appendix J of 10 CFR Part 53 and shall not exceed 0.6 L

a h.

Main Steam Line Isolation yelves The main steam line isola-tion valves shall be tested

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at a pressure of 28 psig for leakage at least once per operating cycle.

If a total leak rate of 11.5 scf per hour for any one main steam line isolation valve is exceeded, repairs and retest shall be performed to correct this condition.

All Type B and Type C Leakage Tests (i.e.

Local Leak Rate Tests) that fail (i.e., test leakage is such that an LER would be required) during an outage shall be reported according to 10 CFR 50.73 by one, 30-day written report that is due within 30 days of the first leakage test failure in the outage.

All other leakage test failures discovered during the outage will be reported in a revision to the original report due within 30 days af ter the end of the outage.

HATCH - UNIT 1 3.7-6a Amendment No. 149

'I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.C.

Minimum Critical Power Ratio (MCPR)

For single-loop operation, the MCPR limit is increased by 0.01 over the comparable two-loop value.

If at any time during operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the steady state MCPR is not returned to within g

the prescribed limits within two (2)

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hours, then reduce reactor power to less than 25% of rated-thermal power within the

' four(4)' hours.

If the Limiting Condition for Operation 1

is restored prior to. expiration of the specified time interval, then t

further progression to less than 25% of rated thermal power is not required.

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. HATCH'- UNIT 1 3.11-2a Amendment No. 149

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BA$hT, QB_WlilN6 CON 0lil0NS FOR OPERATION ANC SURVE!LLANCE RE0VIREM[NTS

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3.11.C.

MinQmCriticalPowerRatio(MCPR) (Continued) 3, 3

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At core thermal power levels less than or equal to 255, the reactor will be operating 'at minimum recirculation pump speed and the moderator void content will be very ses11'. For all designated control EW patterns which may be en-

. ployed at this point, operating plant experience and thermal hydraulic analy-sis indicated that the rwsuit'ing,MCPR value is in ?stess of requirements by e cmit.erable margin. With this low void conteit; any inadvertent core flow j

increase would only v. lect operation in a more con 3rrvative mode relative to

.e NCIR- -Ouring initial-start-up testing of the plant, a MCPR evaldation will be unde at the 255 thermal power level with minimum recirculation pump speed.

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The NCPR margin vill thus be demonstrated such that future NCPR evaluations below this power level will be shown to be unnecessary. The daily require-V cent for calculating MCPR above 255 ated thermal power is sufficient since m

power distribution shif ts are very sloA%en thert,_have not been significant power or control rod c.hanges. The re(strement for calculating NCPR when a limiting control rod ~>ettern is, approached ensures that MCPt will be known following a change 13 lmer or power shape (regardless of angnitJde) that

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HATCH - UNIT 1 2.11-5

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Amendment Nc 149 1

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ADMINISTRATIVE CONTROLS l

MEETING FREQUENCY 6.5.1.4 The PRB shall meet at least once per calendar month arid as convened by the PRB Chairman or his designated alternate.

QUORUM 6.5.1.S The minimum quorum of the PRB necessary for the performance of the PRB responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and three voting members including alternates.

RESP 0NSIBILITIES 6.5.1.6 The Plant Review Board shall be responsible for:

a.

Review of all procedures required by Specification 6.8 and changes thereto, except those for the Radiological Environmental Monitoring Program, any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b.

Review of all proposed tests and experiments that affect nuclear safety.

c.

Review of all proposed changes to Appendix "A" Technical Specifications.

d.

Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.

e.

Investigation of all reportable violations of the Technical Specifications including the preparation and forwarding of reports,

covering evaluation and recommendations to prevent recurrence to the Vice President-Plant Hatch, the Senior Vice President-Nuclear Operations, and to the Safety Review Board (SRB).

f.

Review of all REPORTABLE EVENTS.

g.

Review of unit operations to detect potential nuclear safety

hazards, h.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Menger or the SRB.

HATCH - UNIT 1 6-7 Amendment No.

149

ADMINISTRATIVE CONTROLS OUORUM 6.5.2.6.

The minimum quorum of the SRB necessary for the performance of the SRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least a majority.

of the members. No more than a minority of the quorum shall have line

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responsibility for operation of the unit.

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REVIEW 6.5.2.7.'

The SRB shall be responsible for the review of:

a.

The safety evaluations for (1) changes to procedures, equipment or systems and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions-did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59,10 CFR.

c.

Proposed tests or experiments which involve an unreviewed safety q'estion as defined in Section 50.59, 10 CFR.

u d.

Proposed changes to Technical Specifications or this Operating License.

e.

Violations of codes, regulations, orders Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f.

Significant operating abnormalities or deviations f rom normal and expected performance of unit equipment that af fect nuclear safety.

g.

All REPORTABLE EVENTS.

l h.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could af fect nuclear safety.

i. Reports and meetings minutes of the Plant Review Board.

l HATCH - UNIT 1 6-10 lbaendment No. 149

ADMINISTRATIVE CONTROLS k.

The Radiological Environmental Monitoring Program and the results thereof annually.

1.

The Offsite Dose Calculation Manual. Process Control Program, and implementing procedures at least once per 24 months.

l AUTHORITY 6.5.2.9.

The SRB shall report to and advise the Senior Vice President -

Nuclear Operations on those areas of responsibility specified in Sections 6.5.2.7. and 6.5.2.8.,

RECORDS

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6.5.2.10.

Records of SRB activities shall be prepared, approved and 7

distributed as indicated below:

a.

Minutes of each SRB meeting shall be prepared, approved and forwarded to the Senior Vice President-Nuclear Operations within 14 days following each meeting.

I b.

Reports of reviews encompassed by Section 6.5.2.7. above, shall be prepared, approved and forwarded to the Senior Vice President-Nuclear Operations within 14 days following completion of.the review.

c.

Audit reports encompassed by Section 6.5.2.8. above, shall be forwarded to the Senior Executive Vice President, the Senior Vice President-Nuclear Operations and to the management positions responsible for the areas audited within 30 days af ter completion of the audit.

6.6.

REPORTABLE EVENT ACTION l

6.6.1.

The following actions shall be taken for REPORTABLE EVENTS:

l a.

The Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the PRB and the results of this review shall be submitted to the SRB, the Vice President-Plant Hatch, and the Senior Vice President-Nuclear Operations.

6.7.

SAFETY LIMIT VIOLATION 6.7.1.

The following actions shall be taken in the event a Safety Limit is violated:

a.

The unit shall be placed in at least HOT SHUTDOWN within two hours.

b.

The Safety Limit violation shall be reported to the Commission as soon as practical and in all cases within one hour of occurrence.

The Vice President-Plant Hatch, the Senior Vice President-Nuclear Operations and the SRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

HATCH - UNIT 1 6-12 kmendment No. 149

ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued) c.

A Licensee Event Report shall be prepared pursuant to 10 CFR 50.73.

I d.

The Licensee Event Report shall be submitted to the Commission in accordance with 10 CFR 50.73, and to the PRB, the SRB, the Vice President-Plant Hatch, and the Senior Vice President-Nuclear Operations within 30 days of the violatio i.

6.8.

PROCEDURES 6.8.1.

Written procedures shall be established, implemented and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

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b.

Refueling operations.

f c.

Surveillance and test activities of safety related equipment.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation, g.

PROCESS CONTROL PROGRAM implementation, h.

OFFSITE DOSE CALCULATION MANUAL implementation.

6.8.2 Each procedure of 6.8.1. and other procedures which the Plant Manager or Plant Support Manager has determined to af fect nuclear safety, and changes thereto, shall be reviewed by the PRB and approved by the appropriate member of plant management, designated by the Plant Manager or Plant Support Manager prior to implementation. The Plant Manager or Plant Support Manager will approve administrative procedures, security plan implementing procedures, and changet thereto. The Manager-Plant Training and Onsite Emergency Preparedness shall approve the emergency plan implementing

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procedures and changes thereto. All other procedures of this specification and changes thereto will be approved by the department head designated by the Plant Manager or Plant Support Manager. The procedures of this specification shall be reviewed periodically as set forth in administrative procedures.

6.8.3.

Temporary changes to procedures of 6.8.1. above may be made provided:

a.

The intent of the original procedure is not altered.

HATCH - UNIT 1 6-13 Amendment No. 149 l

ADMINISTRATIVE ( CONTROLS 6.9.

REPORTING RE0VIREMENTS ROUTINE REPORTS l

6.9.1.

In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Of fice of Inspection and Enforcement unless othervise noted.

STARTUP REPORT 6.9.1.1.

A sumery report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license. (2) amendment to the license involving a planned increase in power level.

-(3) installation of fuel that has a dif ferent design or has been manu-E factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant.

6.9.1.2.

The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other comitments shall be in-cluded in this report.

6.9.1.3.

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of comercial operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORT 1d 6.9.1.4.

Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be subeltted prior to March 1 of the year following initial criticality.

i F A single submittal may be made for a multiple unit station. The submittal should combine those sections that are cosmon to all units at the station.

HATCH - UNIT 1 6-14 Amendment No. 149 l

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ADMINISTRATIVE CONTROL $

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ANNUAL REPORTS (Continued) 6.9.1.5.

Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other ptrsonnel, including contractors, receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions,a e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and re f ueli..g. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external l

sources shall be assigned to specific major work functions.

b.

Documentation of all challenges to safety / relief valves.

c.

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.6.F.1.

The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis af ter the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

d.

Any other unit unique reports required on an annual basis.

1 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT (a) 6.g.1.6 Routine radiological environmental surveillance reports covering the radiological environmental surveillance activities related to the plant during the previous calendar year shall be submitted prior to May 1 of each year. A single report may fulfill this requirement for both unitst 6.g.1.7 The Annual Radiological Environmental Surveillance Report shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the reporting period, including (as appropriate) a comparison with the preoperational studies, operational controls, previous environmental surveillance reports, and an assessment of any observed impacts of the plant operation on the environment. The reports shall also include the a.

A single submittal may be made for a multiple-unit station. The submittal should combine those sections coninon to all units at the station.

8this tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

HATCH - UNIT 1 6-15 Amendment No. 149

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4 ADMINISTRATIVE CONTROLS f

e.

Type of container, e.g., LSA, type A, type 8, large quantity.

f.

Solidification agent, e.g., cement.

The Radioactive Effluent Release Report shall include (on a quarterly basis) unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents that were in excess of 1 Ci, excluding dissolved and entrained gases and tritium for liquid ef fluents, or those in excess of 150 Ci of noble gases or 0.02 Ci of radiciodines for gaseous releases.

The Radioactive Effluent Release Report shall include any changes to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MANUAL made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10.

Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Of fice of Management and Program Analysis, U. S. Nuclear Regulatory Commission, Washington,.

D. C. 20555, with a copy to the Regional Office of Inspection and Enforcement no later than the 15th of each month following the calendar month covered by the report.

l HATCH - UNIT 1 6-15d Amendment No. 149

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(This page is intentionally left blank.)

l i

HATCH - UNIT 1 6-16 Amendment No. 149 W

A---

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(This page is intentionally left blank.)

HATCH - UNIT 1 6-17 Amendment No. 149 d

___.________.__a

ADMINISTRATIVE CONTROLS i

SPECIAL REPORTS 6.9.2.

Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period i

specified and for each activity shown in Table 6.9.2-1.

Special reports for fire protection equipment operating and surveillance requirements shall be submitted, as required, by the Fire Hazards Analysis and its Appendix 8 requirements.

6.10.

RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

~

6.10.1.

The following records shall be retained for at least five years:

a.

Records and logs of unit operation covering time interval at each power level.

I b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety, c.

ALL REPORTABLE EVENTS submitted to the Commission.

l d.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e.

Records of changes made to the procedures required by Specification 6.8.1.

f.

Records of radioactive shipments, g.

Records of sealed source and fission detector leak tests and results.

h.

Records of annual physical inventory of all sealed source material of record.

6.10.2.

The following records shell be retained for the duration of the unit Operating License:

a.

Records and drawing changes reflecting unit design modifi-l cations made to systems and equipment described in the Final Safety Analysis Report, b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

l I

HATCH - UNIT 1 6-18 Amendment No. 149

?

I m____-

- _ _.. - - ~

-. fi -

4.0 Special Surveillance and Study Activities 4.1 Erosion Control Inspection I

4.2 Unusual or Important Events Requirements Requirements l

The licensee shall be alert to the occurrence of unusual or I

important events. Unusual or important events are those that cause potentially significant environmental' impact, or could be of public interest concerning environmental impact from plant operation. The following are examples:

unusual or important bird impaction events on cooling tower structures or

, meteorological towers, onsite plant or animal disease outbreaks, unusual mortality of any species protected by the Endangered Species Act of 1973, fish kills near the HNP site, and significant violations of relevant permits and certifications.

Action Should an unusual or important event occur, the if censee shall make a report to the NRC as required by 10 CFR 50.72 or 10 CFR -50.73.

Bases Prompt reporting to the NRC of unusual or important events, as described, is necessary for responsible and orderly regulation of the nation's system of nuclear power reactors. The information thus provided may be useful or necessary to others concerned with the same environmental resources. Prompt knowledge and action may serve to alleviate the magnitude of environmental impact or to place it into a perspective broader than that available to the licensee. The NRC also has an obligation to be responsible to inquiries from the public and.the news media concerning potentially significant environmental j

events at nuclear power stations.

4.3 Exceedina 1.inits of Other Relevant Permits Requirements The licensee shall notify the NRC of occurrences exceeding the limits specified in relevant permits and certificates issued by other Federal, State, and local agencies that are reportable to the agency that issued the permit. This requirement shall apply only to topics of NEPA concern within the NRC area of responsibility as identified in the Environmental Technical Specifications (ETS).

l l

^**"d*"'**

HATCH - UNIT 1 4-1

This requirement shall' commence with the date of issuance of the operating license for Unit 2 and continue until approval for modification or termination is obtained from the NRC in accordance with section 5.6.3.

Action The licensee shall make a report to the NRC in the event of a REPORTABLE EVENT exceeding a limit specified in a relevant permit or certificate issued by another Federal, State, or local agency. The report shall be submitted within the time limit specified by the reporting requirement of the corresponding certification or permit issued pursuant to Section 401 or 402 of. PL 92-500. The report will consist of a copy of the report

_ made to the Georgia Department of Natural Resources, Environmental Protection Division.

Bases The NRC is required under NEPA to maint'ain an awareness of environmental impacts causally related to the construction and operation of facilities licensed under its authority.

Further, some of the ETS requirements are couched in terms of compliance with relevant permits, e.g., the NPDES permit, issued by other licensing authorities. The reports of exceeding limits of relevant permits also alert the NRC staff to environmental problems that may require mitigative action.

i l

l HATCH - UNIT 1 4-2 Amendment No. 149

5.6.2 Nonroutine Reports (Deleted.

Refer to 10 CFR 50.72 and 10 CFR 50.73 for reporting requirements.)

i r

f i

5.6.3 Changes in Environmental Technical Specifications and Permits 5.6.3.1 Changes in Environmental Technical Specifications Requests for changes in ETS shall be submitted to the NRC for review and authorization in accordance with 10 CFR 50.90. The request shall include an evaluation of the environmental impact j

of the proposed change and a supporting justification.

l Implementation of such requested changes in ETS shall not commence prior to incorporation by the NRC of the new l

specifications in the license.

HATCH - UNIT 1 5-8 Amendment No. 149 i

8[

fo UNITED STATES

+

g NUCLEAR REGULATORY COMMISSION n

5

-l WASHINGTON, D. C. 20555 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 86 License No. NPF-5 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated February 13, 1987, complies with the standards and req)uirements of the Atomic Energy Act of 1954, as amended (the Act, and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the l

public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

1 l 2.

Accordingly, the license is amended by changes to the Technical.Specifi-cations as indicated in the attachment to this license amendment, and l

paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby anended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

86, are hereby incorporated in the

. license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

\\

15 Kahtan N. Jabbour, Acting Director Project Of rectorate II-3 Division of Reactor Projects-I/II

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Attachment:

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Changes to the Technical Specifications pW##

Date of Issuance:

December 1, 1987 g

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ATTACHMENT TO LICENSE AMENDMENT N0. 86 FACILITY OPERATING LICENSL N0. NPF-5 DOCKET NO. 50-366 l

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Corresponding overlaaf pages are provided to maintain document completeness.

Remove Insert Page Page iia iia XV XV XVI XVI 1-9 1-9 1-10 1-10 3/4 3-47 3/4 3-47 3/4 3-48 3/4 3-48 3/4 4-3 3/4 4-3 3/4 4-4 3/4 4-4 3/4 4-7 3/4 4-7 3/4 4-8 3/4 4-8 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-11 3/4 4-11 3/4 5-1 3/4 5-1 3/4 5-2 3/4 5-2 3/4 5-4 3/4 5-4 3/4 5-7 3/4 5-7 3/4 5-8 3/4 5-8 I

3/4 6-5 3/4 6-5 3/4 6-6 3/4 6-6 3/4 6-7 3/4 6-7 3/4 6-8 3/4 6-8 3/4 8-7 3/4 8-7 3/4 8-8 3/4 8-8 6-6 6-6 6-9 6-9 6-11 6-11 6-12 6-12 6-13 6-13 6-14 6-14 6-14a 6-14a 6-14b 6-14b 6-14d 6-14d 6-15 6-15 6-16 6-16 6-17 6-17 Replace the following pages of the Appendix B Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Corresponding overleaf pages are provided to maintain document completeness.

4-1 4-1 4-2 4-2 5-8 5-8

1 INDEX' 1

l DEFINITIONS SECTION' f

1.0 DEFINITIONS (Continued)

PAGE REPORTABLE EVENT 1-9 TABLE 1.1, SURVEILLANCE FREQUENCY NOTATION 3-10 TABLE 1.2, OPERATIONAL CONDITIONS 1-11 h

HATCH-UNIT 2 iia Amendment No. 86

_____________-__A

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IuDEx ADMINISTRATIVE CONTROLS

,n SECTION PACE 6.1 RESPONSIBILITY

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6.4 TRAINING i

6-5 6.5 REVIEW AND AUDIT 6.5.1

i PLANT REVIEW BOARD (PP3)

'1 Function 6-5 1 s, Composition 6-5

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Alternates 6-5 j{

Meeting Frequency i1l:

6-6 Quorum l

6-6 1,

Responsibilities 6-6 j

Authority 1

6-7 g

Records a

s 6.5.2 6-7 SAFETY REVIEW BOARD (SRB)

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6-8 4

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Meeting Frequency

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6-8 i

Quorum 6-9 Review 6-9 NATCH-UNIT 2 XV Amendment No. 48 0D9PcCt' ten f ? tt r <cA B -2,w:

~~-

_ _ _ _ _ _ - - - - ~ ~

e INDEX ADMINISTRATIVE CONTROLS l

SECTION PAGE SAFETY REVIEV BOARD (Continued)

Audits 6-10 Authority 6-11 Records 6-11 6.6 REPORTABLE EVENT ACTION 6-11 l

~

f 6.7 SAFETY LIMIT VIOLATION 6-11 6.8 PROCEDURES 6-12 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6-13 l

STARTUP REPORT 6-13 o

ANNUAL REPORTS 6-13 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT 6-14 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-14a MONTHLY OPERATING REPORT 6-14d SPECIAL REPORTS 6-17 6.10 RECORD RETENTION 6-17 6.11 RADIATION PROTECTION PROGRAM 6-18 6.12 HIGH RADIATION AREA 6-18 HATCH-UNIT 2 XVI Amendment No. 86 J

_.d...

_._.m

~~

1.0 DEFINITIONS (Continued)

VENTING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, j

concentration, or other operating condition in such a manner that replacement air or gas is not provided or required during VENTING. The term " vent" used in system names does not imply a VENTING process.

MILK ANIMAL

~

A cow or goat that is producing milk for human consumption.

' REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

1 l

l l

)

l HATCH - UNIT 2 1-9 Amendment No. 86 u_._____.____

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TABLE 1.1 k

~

SURVEILLANCE FREQUENCY NOTATIONS I

Netation Definition Precuency i

S once per shift Cnce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 1

D Daily once per 24 hcurs 2

W Weekly once per 7 days M

Monthly once per 31 days

~

Q Quarterly once per c2 days

,7 SA Semi-annually once per 184 days

+

j R

REFUELING Once per 18 months 5/U STARTUP Prior to each reacter startup P

Prior Completed prier to each release i

NA Not applicable Not applicable 1 (-

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BATCH UNIT 2 1-10,-

Amendment No. 48

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INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION 1

l LIMITING CONDITION FOR OPERATION 3.3.6.2 The seismic monitoring instrumentation shown in Table 3.3.6.2-1 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

l a.

With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit l

a Special Report to the Commission within the next 10 days outlining the cause of the ma'Ifunction and the plans for restoring the instrument (s) to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS e

4.3.6.2.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the fre-quencies shown in Table 4.3.6.2-1.

4.3.6.2.2 Each of the above required seismic monitoring instruments actuated during a seismic event shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 30 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.

l A Special Report shall be prepared and submitted to the Com-mission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety.

HATCH - UNIT 2 3/4 3-47 Amendment No. 86

~

TABLE 3.3.6.2-1

's s k

SEISMIC MONITORING INSTRUMENTATION NINIMUM MEASUREMENT INSTRUMENTS RANGE OPERABLE INSTRUMENTS AND SENSOR LOCATIONS Triaxial Time-History Accelerographs(a) (c) 1.

Diesel Generator Building El 130'0" 0-0.5g 1

a.

(2L51-N021 )

b.

Reactor Building 87' Level on Drywell Pedestal (2L51-N020) 0-0. 5g.

1 Drywell - Feedwater Inlet to RPV 0-0.59 1

l l

c.

(2L51-N004)

+.

Switchyard (c) (1L51-N005) 0-0.5g 1

d.

2.

Triaxial Peak Recording Acceleromegys Diesel Generator Base Support 0-1.0g 1

a.

(lL51-N007)

Intake Structure (c)(IL51-N006) 0-1.09 1

b.

Control Buying Main Control 0-1.09 1

c.

Room Floor (1L51-N008)

'N Control ~ Building Floor El 112,(c)'

0-1. 0g -

1 d.

s,, j (2L51-N028)

Reactor Bldg Refueling Floor 0-1.0g 1

e.

(2L51-N029) l f.

Reactor Pedestal Inside Biological Shield (2L51-NO35) 0-2.0g 1

Feedwater Inlet Reactor Piping 4) 0-2.0g 1

g.

to RPV (2L51-NO3 Triaxial Seismic Switches (b) 3.

Reactor Building 87' Level on 0.025-0.25g i

a.

Drywell Pedestal (2L51-N022)

Reactor Building 185l Level Out-b.

side Biological Shield (2L51-N024) 0.025-0.25g Triaxial Response Spectrum Recorder (a)2-26 Hz 1

4.

Hatch - Unit 1 Coginment Foundation El 87' (1L51-N105) 0-0.5g l

a.

With main control room indication and annunciation.

a DWith main control room annunciation.

'~

Shared with Hatch - Unit 1.

l Amendment No. 11 3/4 3-48 HATCH - UNIT 2 7_._.____..___...____.

SEP 111979 A-----

REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP-LIMITING CONDITION FOR OPERATION 3.4.1.3 An idle recirculation loop shall not be started unless the temperature differential between the reactor coolant within the der:e and the bottom head drain is 1 145'F, and a.

The temperature differential between the reactor coolant l

within the idle loop to be started up and the coolant in the reactor pressure vessel is 1 50*F when both loops hava been idle, or b.

The temperature d5 #erential between the reactor coolant within the idle and operating recirculation loops is < 50*F when only one loop has been idle, and the operating liiop flow rate is 1 507, of rated loop flow.

7 APPLICABILITY: CONDITIONS 1, 2, 3 and 4.

ACTION:

With temperature differences and[er flow rate exceeding the above limits,-

suspend startup of any idle recirculation loop.

SURVEILLANCE REQUIREMENTS 8

4. 4.1. 3 The temperature differential and flow rata shall be determined to be within the limit within 30 minutes prior to startup of an idle recirculation loop.

HATCH - UNIT 2 3/4 4-3 I ^)

1 4

q

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 The safety valve function of the following reactor coolant system safety / relief valves shall be OPERABLE with the mechanical lift settings within 11% of the indicated pressures *.

4 Safety-relief valves 9 1090 psig.

4 Safety-relief valves 9 1100 psig**.

3 Safety-relief valves 9 1110 psig**.

APPLICABILITY: CONDITIONS 1, 2 and 3.

ACTION:

a.

For low-low set valves, take the attion required by Specification 3.4.2.2.

For ADS valves, take the action required by Specification 3.5.2.

b.

With one or more safety / relief valves stuck open, place the reactor mode switch in the Shutdown position.

c.

With one or more S/RV tailpipe pressure switches of an S/RV declared inoperable and the associated S/RV(s) otherwise indicated to be open, place the reactor mode switch in the shutdown position, d.

With one S/RV tailpipe pressure switch of an S/DV declared inoperable and the associated S/RV(s) otherwise indicated to be closed, plant operation may continue. Remove the function of that pressure switch f rom the low low set logic circuitry until the next COLD SHUTDOWN.

Upon COLD SHUTDOWN, restore the pressure switch (s) to OPERABLE status before STARTUP.

e.

With both S/RV tailpipe pressure switches of an S/RV declared ir.op-erable and the associated S/RV(s) otherwise indicated to be closed, restore at least one inoperable switch to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

f.

The failure or malfunction of any safety / relief valve shall be reported by telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designee no later than the first working day following the event; and a written followup report within 30 days.

The written followup report should be completed in accordance with 10 CFR 50.73 or other applicable requirements.

SURVEILLANCE RE0VIREMENTS 4.4.2.1 The tail-pipe pressure switches of each safety / relief valve shall be demonstrated OPERABLE by performance of:

\\

a.

CHANNEL FUNCTIONAL TEST:

1.

At least once per 31 days, except that all portions of the channel inside the primary containment may be excluded from the CHANNEL FUNCTIONAL TEST, and 2.

At each scheduled outage of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during which entry is made into the primary containment, if not performed within the previous 31 days.

b.

CHANNEL CALIBRATION and verifying the setpoint to be 85 psig, with an allowable tolerance of +15 psig and -5 psig, at least once per 18 months.

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperature and pressure.

- Up to two inoperable valves may be replaced with spare OPERABLE valves with lower setpoints of 1090 and 1100 psig, respectively, until the next refueling outage.

3 HATCH - UNIT 2 3/4 4-4 Amendment No. 86

i REACTOR COOLANT SYSTEM-3/4.4.4 CHEMISTRY

^

LIMITING CONDITION FOR OPERATION

~

.]

s 3.4.4 The chemistry of the reactor coolant system shall be maintained sp within-the limits specified in Table 3.4.4-1.

1H APPLICABILITY: At all times.

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3 ACTION:

a.

In CONDITION 1, 2 and 3:

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With the conductivity or chloride concentration o ceeding the 1.

limit specified in Table 3.4.4-1,. but less than 10 pmho/cadat 25'C and~1ess than 0.5 ppm, respectively, operation may continue for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and this need not be reported to the Commission, provided that operation under these conditions l

shall not exceed 336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months ' per year. If operation under these conditions exceeds 336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months per year, in lieu of any other l

report required by 10 CFR 50.73, prepare and submit a Special Report to the Consnission pursuant to Specification 6.9.2, within

.y 30 days, outlining the course of the limit violation and the plans for restoring the conductivity or chloride concentration to within the limit. The provisions of Specification 3.0.4 are not applicable.

2.

With the conductivity or chloride concentration exceeding the limit specifie4 in Table 3.4.4-1 for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during one confinuous time interval or with the conduct;ivity exceeding 10 umho/cm at 25'C or chloride exceeding O'.S ppm, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

At all other times:

1.

With the conductivity of the reactor coolant in excess of the limit specified in Table 3.4.4-1, restore the conductivity to within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the chloride limit of Table 3.4.4-1 exceeded for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , perform an engineering evaluation to determine-the effects of the out-of-limit condition on the structural integrity of the reactor coolant system. Determine,that the structural integrity of the reactor coolant system remains acceptable for continued operation prior to proceeding to CONDITION 3.

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HATCH - UNIT 2 3/4 4-7 Amendment No. 86

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4 4.4 Therecctof). coolant shalVbe determined to be within the specified t

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chlorides at least once psr 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and

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Continuously recording t' e conductivity of the reactor coolant, h

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Analyzing a sample of the reactor toolant for conductivity at least onth per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the continuous recording con-a y

ductivity monitohis inoperable.

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I RE ACTOR C00L AW7 SYSTEM 3/4.4.5 SPECIFIC ACTiv!TY LIN1 TING CONDITION FOR OPERATION 3.4.5 The specific activity of the reactor coolant shall be limited to:

a.

5 0.2 pCi/ gram DOSE EQUIVALENT !-131, and b.

$ 100/E wCi/ gram.

APPLICABILITY: CONDITIONS 1, 2, 3 and 4 ACTION:

a.

In CONDITIONS 1, 2 and 3, with the specific activity of the reactor coolant; E

l 1.

> 0.2 uti/ gram DOSE EQUIVALENT I-131 but 5 4.0 wC1/ gram for l

more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or > 4.0 wC1/ gram, be in at f east HOT SHUTDmei with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

> 100/l pC1/ gram, be in at least NOT SHUTDOWN with the main l

steamline isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD

$HUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In CONDITION 1, 2, 3 or 4, 1.

With the specific activity of the primary coolant > 0.2 wC1/ gram DOSE EQUIVALENT I-131 or > 100/E wC1/ gram, perform the sampling and analysis requirements of Ites 4b of Table 4.4.5-1 at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months until the specific activity of tM primary coolant is restored to within its limits.

MATCH - UNIT 2 3/4 4-10 Amendment No. 86 l

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RE AC10R COOL ANT SYSTEM j

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L!uiTING CONDITION FOR OPERATION (Continued)

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ACTION: (Continued) 2.

With:

a) THERMAL POWER changed by more than 15% of RATED THERMAL POWER in one hour, or b) The of f-gas level, at the SJAE, increased by more than 10,000 v Ci/sec. in one hour at release rates less than 15,000 v C1/sec, or c) The off-gas level, at the SJAE, increased by more than 155 in one hour at release rates greater than 75,000 y C1/sec.,

perform the sampling and analysis requirement of Item 4C of Table 4.4.5-1.

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$URVEILLMCE RE0UIREMENTS 4.4.5 The specific activity of the reactor coolant sh411 be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.

NATCH - UNIT 2 3/4 4-11 Amendment No. 86 1

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM

)

LIMITING CONDITION FOR OPERATION 3.5.1 The High Pressure Coolant Injection (HPCI) system shall be 1

OPERABLE with:

a.

One OPERABLE high pressure coolant injection pump, and b.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor pressure vessel.

~

APPLICABILITY: CONDITIONS 1*, 2* and 3* with reactor vessel steam dome pressure > 150 psig.

ACTION:

a.

With the HPCI system inoperable, POWER OPERATION may continue and the provisions of 3.0.4 do not apply *, provided the RCIC system, ADS, CSS and LPCI system are OPERABLE; restore the inoperable HPCI system to OPERABLE status within 14 days or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to 5150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

With the surveillance requirements of Specification 4.5.1 not performed at the required frequencies due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applic-able provided the appropriate surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the tests.

SURVEILLANCE REQUIREMENTS 4.5.1 The HPCI shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water, and

See Special Test Exception 3.10.5 HATCH - UNIT 2 3/4 5-1 Amendment No. 86

EMERGENCY CORE COOLING SYSTEMS _

T}

t SURVEILLANCE REQUIREMENTS (Continued)

Verifying)that each valve (manual, power o'perated orin th 2.

automatic or othemise secured in position, is in its correct position.

b.

At least unce per 92 days, by verifying that the system develops a flow of at least 4250 gpm for a system head cor-responding to a reactor pressure of > 1000 psig when steam is being supplied to the turbine at < 1600 psig.

~

c.

At least once per 18 months by:

+'

1.

Perfonning a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.

2.

Verifying that the system develops a flow of at least y,b 4250 gpm for a system head corresponding to a reactor

(

1 pressure of > 165 psig when steam is being supplied to N/

the turbine at 165 + 15 psig.

Verifying that the suction for the HPCI system is auto-3.

matica11y transferred from the condensate storage tank to the suppression chamber on a condensate storage tank low water level signal and on a suppression chamber high water, level signal.

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f HATCH - UNIT 2

.3/,4 5-2

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EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 LOW PRESSURE CORE COOLING SYSTEMS CORE SPRAY SYSTEM i

LIMITING CONDITION FOR OPERATION 3.5.3.1 Two independent Core Spray System (CSS) subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE CSS pump, and b.

An OPERABLE flow path capable of taking suction from at least one of the following OPERABLE sources and transferring the water through the spray sparger to the reactor vessel; 1.

In CONDITION 1, 2 or 3, from the suppression pool.

2.

In CONDITION 4 or 5*;

a) From the suppression pool, or b) When the suppression pool is being drained, from the condensate storage tank containing at least 150,000 gallons of water.

APPLICABILITY: CONDITIONS 1, 2, 3, 4, and 5*.

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-a.

In CONDITION 1, 2 or 3; 1.

With one CSS subsystem inoperable, POWER OPERATION may

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continue provided both LPCI subsystems are OPERABLE; restore the inoperable CSS subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With both CSS subsystems inoperable, be in at least HOT i

SHUTDWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The core spray systes and the suppression chambtr are not required to be OPERABLE provided that the reactor vessel head is removed and the cavity is flooded, the spent fuel pool gates are removed, and the water level is maintained within the limits of Specification 3.9.9 and 3.9.10.

HATCH - UNIT 2 3/4 5-4 Amendment No. 86 9

EMERGENCY CORE COOLING SYSTEMS LOW PRESSURE COOLANT INJECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.5.3.2' Two. independent Low Pressure Coolant Injection (LPCI) sub-systems of the residual heat removal system (RHR) shall be OPERABLE with each subsystem comprised of:

l a.

Two OPERABLE RHR pumps, b.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor pressure vessel.

APPLICABILITY: CONDITIONS 1, 2, 3, 4* and 5*, **.

ACTION:

a.

In CONDITION 1, 2 or 3; 1.

With one LPCI subsystem or one LPCI pump inoperable, POWER OPERATION may continue provided both CSS subsystems are OPERABLE; restore the inoperable LPCI subsystem or pump to OPERABLE status within 7 days or be in at least HOT SHUTDJWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT -

DOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With both LPCI subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and either be in COLD SHUTDOWN or maintain reactor coolant temperature s 400*F by use of alternate heat removal methods within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With the LPCI system cross-tie valve open or power not removed from the valve operator, be in at least HOT SHUTDOWN with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In CONDITION 4* or 5*, ** with one or more LPCI subsystems inoperable, take the ACTION required by Specification 3.5.3.1.

The provisions of Specification 3.0.3 are not a>plicable.

Not applicable when two CSS subsystems are OPERABLE per Specification 3.5.3.1.

- Not applicable when the CSS is not required to be OPERABLE per Specification 3.5.3.1.

HATCH - UNIT 2 3/4 5-7 Amendment No. 86

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l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.2 Each LPCI subsystem shall be demonstrated OPERABLE:

a.

At least once-per 31 days by:

i 1.

Verifying that the system piping from the pump dis-charge valve to the system isolation valve is filled with water, 2.

Verifying that each valve (manual, power operated or o.o automatic).'in the flow path th;t is not locked, sealed, or otherwise secured in position, is in its correct position, and 3.

Verifying that the subsystem cross-tie valve is closed with power removed from the valve operator, b.

At least once~ per 92 days by verifying each pair of LPCI pumps discharging to a common header can be started'from the control room and develops a total flow of at least 17,000 m

gpm against a system head corresponding to a reactor vessel "j.-

pressure of _> 20 psig.

~

c.

At least once per 18 months by performing a system func-tional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.

s

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HATCH - UNIT 2 3/4 5-8

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CONTAINMENT SYSTEMS SURVEILLANCE RE0UIREMENTS (Continued) b.

If any periodic Type A test f ails to meet either.75 La, or

.75 Lt. the test schedule for subsequent Type A tests shall be reviewed and approved by the Connission.

If two consecutive Type A tests fail to meet either.75 L or.75 L. a Type A l

t a

test shall be performed at least every 18 months until two l

consecutive Type A tests meet either.75 La or.75 L. at I

t which time the above test schedule may be resumed.

l l

c.

The accuracy of each Type A test shall be verified by a supplemental test which:

1.

Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A

{

test data is within 0.25 L or 0.25 L '

a t

2.

Has a duration sufficient to establish accurately the change in leakage rate between the type A test and the supplemental test, and 3.

Requires the quantity of gas injected into the contain-ment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P, 57.5 psig, or P 28.8 a

t psig.

d.

Type B and C tests

shall be conducted at Pa, 57.5 psig, at l

intervals no greater than 24 months except for tests involving:

1.

Air locks, which shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3, and 2.

Main steam line isolation valves, which shall be leak tested at least once per 18 months.

e.

All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be per-formed to select a balanced integrated leakage measurement system.

f.

The provisions of Specification 4.0.2 are not applicabita.

All Type B and Type C Leakage Tests (i.e., Local Leak Rate Tests) that fail (i.e., test leakage is such that an LER would be required) during an outage shall be reported according to 10 CFR 50.73 by one 30-day written report that is due within 30 days of the first leakage test failure in the outage All other leakage test failures discovered during the outage will be reported in a revision to the original report due within 30 days af ter the end of the outage.

HATCH - UNIT 2 3/4 6-5 Amendment No. 86

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CONTAINMENT SYSTEMS

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1 J) j PRIMARY CONTAINMENT AIR LOCK

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LIMITING CONDITION FOR OPERATION 3.6.1.3 The primary containment air lock shall be OPERABLE with:

)

a.

Both doors closed except when the air lock is being used for nomal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.

An overall air lock leakage rate of s 0.05 L, at P,, 57.5 psig.

APPLICABILITY: CONDITIONS 1, 2* and 3.

ACTION:

o a.

With one primary containment air lock door inoperable, maintain at least the OPERABLE air lock door closed; restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed; operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 30 days. The provisions of Specification 3.0.4 are not applicable.

b.-

With the primary containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6 c.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS

~

4.6.1.3 The primary containment' air lock shall be demonstrated OPERABLE:

a.

- After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying no detectable seal leakage when the gap between the door seals is pressurized to L 10 psig fo~r at least 15 minutes, b.

At least once per 6 months by conducting an overall air lock leakage rate test at P overall air lock leaka$e, 57.5 psig, and by verifying that the rate is within its liir.it.

c.

.At least once per 6 months by verifying that only one door in the air lock can be opened at a time.

See Special Test Exception 3.10.1.-

o* Exemption to Appendix J of 10 CFR 50.

(

j-HATCH - Unit 2 3/4 6-6

2N E_ ~

CONTAINMENT SYSTEMS

,,. )'

iSIV LEAXAGE CONTROL SYSTEM s>

s

_IMITING CONDITION FOR OPERATION 3.6.1.4 Two MSIV Leakage Control System (LCS) subsystems shall be DPERABLE.

APPLICABILITY: CONDITIONS 1, 2 and 3.

i ACTION:

a.

With one MSIV leakage control system subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The pro-e visions of Specification 3.0.4 are not applicable.

h.

With both MSIV leakage control system subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT-DOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS

{

)

4.6.1.4 Each MSIV Leakage Control System subsystem shall be demonstrated OPERABLE:

a.

At least once per 31 days by starting the blower from the control room and operating the blower for at least 15 minutes.

b.

Each COLD SHUTDOWN, if not perfomed within the previous 92 days, by cycling each bleeder valve through at least one complete cycle of full travel.

c.

At least once per 18 months, by performance of a functional test which includes simulated actuation of the subsystem i

throughout its operating sequence and verifying that each auto-matic valve actuates to its correct position and the blower starts and developes:

1.

For inboard MSIVs - 100 scfm at a vacuum of 60" H20, and 2.

For outboard MSIVs - 240 tefm at a vacuum of 50" H 0.

2 a.

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HATCH - UNIT 2 3/4 6-7 S.

.s.

CONTAINMENT SYSTEMS pRIMARYCONTAfNMENTSTRUCTURALINTEGRITY LIMITING CONDITION FOR OPERATION b

3.6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Speci-fication 4.6.1.5.

APPLICABILITY: CONDITIONS 1, 2, and 3.

^

ACTION:

I With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212*F.

SURVEILLANCE REQUIREMENTS 4.6.1.5 The structural integrity of the primary containment shall be determined during the shutdown f.or each Type A containment leakage rate test by a visual inspection of the accessible interior and exterior surfaces of the containment and verifying no apparent changes in appear-ance of the surfaces or other abnormal degradation.

Amendment N. 86 HATCH - UNIT 2 3/4 6-8 9..

-_O

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) j p

4.8.1.1.4 Reports - All diesel generator failures, valid or non-valid, shall be reported'to the Commission pursuant to 10 CFR 50.73 or Specification

.6.9.2, as applicable. If the number of failures in the last 100 valid tests, on a per nuclear unit basis, is 2.7, the report shall be supplemented to include.the additional information recommended in Regulatory Position C.3.b of Regulatory Guide, 1.108, Revision 1, August 1977.

p i

1 Amendment No. 86 HATCH - UNIT 2 3/4 8-7 l

)

D

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TABLE 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE Number of Failures In Last 100 Valid Tests

Test Frequency

<1 At least once per 31 days 2

At least once per 14 days 3

At least once per 7 days

>4 At least once per 3 days

. m,

Criteria for determining number of fa'ilures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.100, Revision 1 August 1977, where the last 100 tests are determined on a per nuclear unit basis.

_ps HATCH - UNIT 2 3/48-8

~

ADMTNTSTRAT1VE CONTROLS i

MEETING FREQUENCY 6.5.1.4 by the PRB Chairman or his designated alternate.The PRB r

QUORUM 6.5.1.5 PRB responsibility and authority provisions of these Te and three voting members including alternates. Specificati RESP 0NSIBILITIES 6.5.1.6 The Plant Review Board shall be responsible for:

Review of all procedures required by Specification 6.8 and a.

changes thereto Monitoring Progr,am, any other proposed procedures thereto as determined by the Plant Manager to affect nuclear saf e y.

b.

Review of all proposed tests and experiments that affect nuclear safety.

Specifications. Review of all proposed changes to Appendix "A" Techni c.

d.

equipment that affect nuclear safety. Review of all pro Investigation of all reportable violations of the Techn'ical e.

~

covering evaluation and recommendations to the Vice President-Plant Hatch, the Senior Vice President-Nucl Operations, and to the Safety Review Board (SRB).

ear f.

Review of all REPORTAELE EVElrIS.

l Review of unit operations to detect potential nuclear safety g.

hazards.

h.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Manager or the SRB HATCH - UNIT 2 6-6 Amendment No. 86

_ _ _ _ _ - - - - - _ - - - - - - - ~ ~ ~ ~ ~

_ _ _ _ _ - - - - - - - - ~ ~ ~ ~

ADMINISTRATIVE CONTROLS 1

I OUORUM 6.5.2.6 The minimum quorum of the SRB necessary for the performance of the SRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least a majority of the members.

No more than a minority of the quorum shall have line responsibility for operation of the unit.

REVIEW 6.5.2.7 The SRB shall be responsible for the review of:

a.

The safety evaluations for (1) changes to procedures, equipment or systems and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verity that such actions did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

c.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d.

Proposed. changes to Technical Specifications or this Operating License.

e.

Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f.

Significant operating abnormalities or deviations f rom normal and expected performance of unit equipment that af fect nuclear

safety, g.

All REPORTABLE EVENTS.

h.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.

i. Reports and meetings minutes of the Plant Review Board.

t l

HATCH - UNIT 2 6-9 Junendment No. 86

-ADMINISTRATIVE CONTROLS k.

The Radiological Environmental Monitoring Program and the results thereof annually.

1.

The Offsite Dose Calculation Manual. Process Control Program, and implementing procedures at least once per 24 months.

AUTHORITY

]

I 6.5.2.9 The SRB shall report to and advise the Senior Vice President -

Nuclear Operations on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of SRB activities shall be prepared, approved and distributed as indicated below:

a.

Minutes of each SRB meeting shall be prepared, approved and forwarded to the Senior Vice President-Nuclear Operations I

within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be i

prepared, approved and forwarded to the Senior Vice President-Nuclear Operations within 14 days following completion of the review.

c.

Audit reports encompassed by Section 6.5.2.8 above, shall be l

forwarded to the Senior Executive Vice President, the Senior i

Vice President-Nuclear Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE EVENT ACTION l

6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

l a.

The Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the PRB, and the results

~

of this review shall be submitted to the SRB, the Vice President-Plant Hatch, and the Senior Vice President-Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION l

6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The unit shall be placed in at least HOT SHUTOOWN within two hours, b.

The Safety Limit violation shall be reported to the Commission as soon as practical and in all cases within one hour of occurrence.

The Vice President-Plant Hatch, the Senior Vice President-Nuclear Operations and the SRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l HATCH - UNIT 2 6-11 Amendment No. 86 L

ADMINISTRATIVE CONTROLS l

SAFETY LIMIT VIOLATION (Continued) c.

A Licensee. Event Repcrt shall be prepared pursuant to 10 CFR 50.73.

l d.

TheLicenseeEventReportshallbesubmittedtotheCommissionin accordance with 10 CFR 50.73, and to the PRB, the SRB, the Vice President-Plant Hatch, and the Senior Vice President-Nuclear Operations within 30 days of the violation.

6.8 PROCEDURES 6 '. 8.1 Written procedures shall be established, implemented and maintained i

covering the activities referenced below:

a.

The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33. Revision 2, February 1978.

b.

Refueling operations.

f c.

Surveillance and test activities of safety related equipment.

d.

Security Plan implementation, e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

g.

PROCESS CONTROL PROGRAM implementation.

h.

OFFSITE DOSE CALCULATION MANUAL implementation, 6.8.2 Each procedure of 6.8.1 and other procedures which the Plant Manager or Plant Support Manager has determined to af fect nuclear safety, and changes thereto, shall be reviewed by the PRB and approved by the appropriate member of plant management, designated by the Plant Manager or Plant Support Manager, prior to implementation. The Plant Manager or Plant Support Manager will approve administrative procedures, security plan implementing procedures, and changes thereto. The Manager-Plant Training and Onsite Emergency Preparedness shall approve the emergency plan implementing procedures and changes thereto. All other procedures of this specification and changes thereto will be approved by the department head designated by the Plant Manager or Plant Support Manager. The procedures of this specification shall be reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a.

The intent of the original procedure is not altered.

HATCH - UNIT 2 6-12 Jbnendment No. 86 w-___-_____---__________-_-.

ADMINISTRATIVE CONTROLS 6.9 rep 0RTING REQUIREMENTS ROUTINE rep 0RTS l

6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

START-UP rep 0RT 6.9.1.1 A sunmary report of plant start-up and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a plarned increase in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant.

6.9.1.2 The start-up report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be in-cluded in this report.

6.9.1.3 Start-up reports shall be submitted within (1) 90 days following completion of the start up test program, or (2) 90 days following resumption or commencement of commercial power operation, or (3) 12 months following initial criticality, whichever is earliest.

If the Start-up Report does not cover all three events (i.e., initial criticality, completion of start-up test program, and resumption or commencement of commercial operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL PEPORTSS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

A single subsittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

Amendment No. 86 HATCH - UNIT 2 6-13

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) 6.9.1.5 Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other personnel, including contractors, receiving exposures greater than 100 mrem /yr and their associated man rem exposure i

according to work and job functions,8 e.g., reactor operations and surveillance inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received f rom external sources shall be assigned to specific r..ajor work functions.

b.

Documentation of all challenges to safety / relief valves.

c. 'The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5.

The following information shall be included:

(1)-Reactor power history starting.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine was reduced to less than the limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in

'which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

~

d.

Any other unit unique reports required on an annual basis.

l ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT <a>

6.9.1.6 Routine radiological environmental surv'eillance reports covering the radiological environmental surveillance activities related to the plant during the previous calendar year shall be submitted prior to May*1 of each year. A single report may fulfill this requirement for both units.

6.9.1.7 The Annual Radiological Environmental Surveillance Report shall include summaries, interpretations, and statistical evaluation of the a.

A single submittal may be made for a multiple-unit station. The submittal should combine those sections common to all units at the station.

8This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

HATCH - UNIT 2 6-14 Amendment No. 86

l l

t ADMINISTRATIVE CONTROLS results of the radiological environmental surveillance activities for the i

reporting period, including (as appropriate) a comparison with the

)

preoperational studies, operational controls, previous environmental I

surveillance reports, and an assessment of any observed impacts of the I

plant operation on the environment.

The reports shall also include the j

results of the land use surveys required by Specification 3.16.2 of Unit 1 i

Technical Specifications and the results of licensee participation in the 1

interlaboratory comparison program required by Specification 3.16.3 of Unit 1 Tc:hnical Specifications.

I The Annual Radiological Environmental Surveillance Report shall include summarized and tabulated results in the format of table 6.9.1.7-1 of all radiological environmental samples taken during the report period, with the exception of naturally occurring radionuclides which need not be reported.

In the event that some results are not available for inclusion with the report, the report shall be submitted, noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as practicable in a supplementary report.

The reports shall also include the following:

a.

Summary description of the radiological environmental monitoring program.

j b.

Map of all sampling locations as keyed to a table indicating distances and directions from main stack.

c.

Results of the licensee participation in the Interlaboratory Comparison Program.

SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORTeae

~

6.9.1.8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

Any changes to the 00CM shall be submitted with the next semi-annual report in which the change (s) was made effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall Be submitted with the monthly operating report for the period in which the evaluation was reviewed and accepted by the Plant Review Board.

a.

A single submittal may be made for a multiple-unit station.

The submittal should combine those sections that are common to all units at the station; however, the submittal shall specify the releases of radioactive material froe each unit.

HATCH - UNIT 2 6-144 Amendment No. 86 l

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ADMINISTRATIVE CONTROLS e.

Type of container, e.g., LSA, type A, type B, large quantity f.

Solidification agent, e.g., cement.

The Radioactive Effluent Release Report shall include (on a quarterly basis) unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents that were in excess of 1 C1, excluding dissolved and entrained gases and tritium for liquid effluents, or those in excess of 150 Ci of noble gases or 0.02 C1 of radiotodines for gaseous releases.

^

The Radioactive Effluent Release Report shall include any changes to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MANUAL made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management.

and Program Analysis, U. S. Nuclear Regulatory Commir.sion, Washington, D. C. 20555, with a copy to the Regional Office of Inspection and Enforcement no later than the 15th of each month following the calendar -

month covered by the report.

i 1

HATCH - UNIT 2 6-14d Amendment No. 86 _

i

II

>1 i

l i

(

is page is intentionally lef t blank.)

FRTCH - UNIT 2 6-15 Amendment No. 86 c.

j I

i l

f

/

(This page is intentionally left blank.)

k-i e

HATCH - UNIT 2 6-16 Amendment No. 86

s' ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office 1l

'of Inspection and Enforcement Regional Office within the time period specified for each report. Special reports for fire protection equipment operating and surveillance requirements shall be submitted, as required, by the Fire Hazards Analysis and its Appendix B requirements.

3 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for

, 73 at least the minimum period indicated.

7 6.10.1 The following records shall be retained for at least five years:

a.

Records and logs of unit operation covering time interval at each power level, b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c.

ALL REPORTABLE EVENTS submitted to the Commission.

l d.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e.

Records of changes made to the procedures required by Specification 6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

h.

Records of annual physical inventory of all sealed source

.s material of record.

i 6.10.2 The following records shall be retained for the duration of the unit Operating License:

a.

Records and drawing changes reflecting unit design modifi-cations made to systems and equipment described in the Final Safety Analysis Report.

b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

h nde nt N. %

HATCH - UNIT 2 6-17 i

a-_--___-___

.l \\ -

t i

4.0 Special Surveillance and Study Activities 4.1 Erosion Control Inspection 4.2 Unusual or Imoortant Events Requirements Requirements The licensee shall be alert to the occurrence of' unusual or important events. Unusual or important events are those that cause potentially significant environmental impact or could be of public interest concerning environmental impact from plent operation. The following are examples: unusual or important bird impaction events on cooling tower structures or meteorological towers, onsite plant or animal disease outbreaks, unusua? mortality of any species protected by the Endangered Species Act of 1973, fish kills near the HNp site, and significant violations of relevant permits and certifications.

Actions Should an unusual or important event occur, the licensee shall make a prompt report to the NRC as required by 10 CFR 50.72 or 10 CFR 50.73.

Bases Prompt reporting to the NRC of unusual or important events, as described, is necessary for responsible and orderly regulation of the nation's' system of nuclear power reactors. The information thus provided may be useful or necessary to others concerned with the same environmental resources. Prompt knowledge and action may serve to alleviate the magnitude of environmental impact or to place it into a perspective broader than that available to the licensee.

The NRC also has an obligation to be responsive to inquiries from the public and the news media concerning potentially significant environmental events at nuclear power stations.

4.3 Exceedino Limits of Other Relevant permits

-Requirements The licensee sail.aotify the NRC of occurrences exceeding,he limits specified in rebvant permits and certificates issued by other Federal, Stite, ano local agencies that are reportable to the agency that trued the permit. This requirement shall apply only to topics of NbM concern within the NRC area of responsibility as ident1Ned it, the f. environmental Technical Specifications (ETS).

^"*d"**"'

HATCH - UNIT 2 4-1

4 j.

a f

c, i

o

.f Y

s This requirement shall commente with the date of issuance of the

., /

c; orating licensw* for Unit < and continue until approval for

,'s modification or termination is obtained,frde r,he NRC in 1A a'cprdan@?with section 5.6.3.

't

/

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')

Action n

J t

1 he j h easse shall make a report to the NRC in the event of a i

'!QOP.TAEi.E EVENT exceeding a limit specified in a relevant l

permit or certificate issued by another Federal, State, or local i

agency. The report shali be submitted within the time limit specified by the reporting requirements of the corresponding certification or permit issued pursuant to Section 401 or 402

- to the Ge ngia'Cr[e report will consist of a copy of the report made of PL 92-500. Th artment.of Natural Resources, Environmental Protection Division.

~

Bases The NRC is required under NEPA to maintain an awareness of environmental impacts causally related to the construction and operation of facilities licensed under its authority.

Further, some of the ETS requirements are couched in terms of compliance with relevant permits, e.g.

the NPDES permit, issued by other licensing authorities. The reports of exceeding limits of relevant permits also alert the NRC staff to environmental problems that may require mitigative action.

l I

Amenament n. 86 HATCH - UNIT 2 4-2

r l

l f-l l

5.6.2 Nonroutine Reports-Deleted. Refer to 10 CFR 50.72 and 10 CFR 50.73 for reporting requirements.

l h~

l 5.6.3 Changes in Environmental Technical Specifications and Permits

~

5.6.3.1 Changes in Environmental-Technical Specifications Requests for changes in ETS shall be submitted to the NRC for review and authorization in accordance with 10 CFR 50.90.

The request shall include an evaluation of the environmental impact of the proposed change and a supporting justification.

Implementation of such requested changes in ETS shall not comeAnta prior to incorporation by the NRC of the new

" specifications in the license.

HATCH-UNIT 2 5-8 Amendment No. 86

_ - _ -

v t e Letter Sequence Other
TAC:55713, (Open) TAC:55714, (Open)
Initiation Request Acceptance... Administration Questions Answers Results Approval Other: ML20212H439, ML20237C775
MONTHYEAR1987-12-01 [Table View]

ML20237C775

Text

Attachment:

Attachment:

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