Text

Amend 87 to License NPF-29,incorporating Programmatic Controls Into Administrative Controls Section of TS in Accordance W/Requirements of 10CFR20.106,40CFR190, 10CFR50.36(a) & App I to 10CFR50
	ML20086U043
Person / Time
Site:	Grand Gulf ; (NPF-29-A-087, NPF-29-A-87)
Issue date:	12/18/1991
From:	Larkins J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20086U048	List: ML20086U043; ML20086U051;
References
	NUDOCS 9201070178
	Download: ML20086U043 (48)
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UNIT ED sT ATE s

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' 'g NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D

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e ENTERGY OPEEATIONS, INC.

SYSTEM ENERGY RESOURCES, INC.

SOUTH MISSIS $1PPI ELECTRIC POWER ASSOCIATION HISSISSIPPI POWER AND LIGHT COMPANY DOCKET N0. 50-416 GRAND GULF NUCLEAR STATION UNIT 1 AMENDMENT,TO FACIL!TY OPERATING _ LICENSE Amendnent No. 87 License No. NPF-29 1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee)datedAugust 13, 1991, complies with the standards and requirements nf the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth ir 10 CFR Chapter I; B.

The facility will operate in conformity with the epplication, the provisions of the Act, and the rules and regulatieris of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be (.onducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in complience with the Comission's regulations;,

D.

The issuance of this amendment will not be inittical to the common defense and security or to the health and safet,y of the public; and E.

The issuance of this amendment is ii. -cordance with 10 CFR Part 51 of the Comission's regulations and ali oplicable requirements have been sitisfied.

1 9201070170 911210 ADOCK050g16 DR

2 2.

Accordingly, the 1Icense is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendmcntt end paragraph 2.C.(2) of facility Operating License No. NPF-29 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 87. are hereby incorporated into this license.

Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment it effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION u

John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects III, IV, and V Office of Nuclear Reactor Regulation r

Attachment:

Changes to the Technical Specifications Date of Issuance: December 18. 1991

.~

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. NPF-29 fgCKETNO.50416 i

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

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE PAGES INSERT PAGES 1

1 11 11 Y

Y xi xi xiii xiii xvi xvi avit xvii xX Xx 1-3 1-3 1-5 1-5 1-6 1-6 1-8 1-8 3/4 3-91 thru 3/4 3-96 3/4 3-91 3/4 3 97 thru 3/4 3-105 3/4 3-92 3/4 3 106 3/4 3-93 3/4 3-107 3/4 3-94 3/4 3-108 3/4 3-95 3/4 3-109 3/4 3 96 3/4 3-110 3/4 3-97 3/4 3-111 3/4 11-1 thru 3/4 11-7 3/4 11-1 3/4 11-8 thru 3/4 11-16 3/4 11-2 3/4 11-17 3/4 11-3 3/4 11-18 thru 3/4 11-19 3/4 12-1 thru 3/4 12-12

" 3/4 3-6 B 3/4 3-6 b 3/4 11-1 B 3/4 11-1 B 3/4 11-2 thru B 3/4 11-5 B 3/4 11-2 B 3/4 11-6 B 3/4 12-1 thru B 3/4 12-2 6-15 6 15

6. d.r 6-ibb 6-17 6-17 6-18 6-18 i

6-19 6-19 6-20 6-20 6-22 6-22 6-23 6-26

u un.>

INDEX

. DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTI0N...............................................................

1-1 1.2 AVERAGE PLANAR EXP05URE..............................................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...........................

1-1 1.4 CHANNEL CALIBRATION..................................................

1-1 1.5 CHANNEL CHECK........................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST..............................................

1-1

1. 7 CORE ALTERATION......................................................

1-2 1.8 CRITI'.AL POWER RATI0.................................................

1-2 1.9 DOSE EQUIVALENT I-131................................................

1-2 1.10 DRYWELL INTEGRITY....................................................

1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY......................................

1-3 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME...................

1-3 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.

1-3 1.14 FRACTION OF LIMITING POWER DENSITY.....................

1-3 1.15 FRACTION OF RATED THERMAL P0WER......................................

1-3 1.16 FREQUENCY N0TATION...................................................

1-3 1.17 DttETED.................................

1-3 l

1.18 I D E N T I F I E D L E A KAG E...................................................

1-4 1.19 ISOLATION SYSTEM RESPONSE TIME.......................................

1-4 1.20 LIMITING CONTROL R03 PATTERN.........................................

1-4 1.21 LINEAR HEAT GENERATION RATE.......'..................................

1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST.........................................

1-4 1.23 MAXIMUM FRAC'.. N OF LIMITING POWER DENSITY...........................

1-4 1.24 MEMBER (S) 0F IHE PUBLIC..............................................

1-4 1.25 MINIMUM CRITICAL POWER RATI0.........................................

1-5 1.26 0FFSITE DOSE CALCULATION MANUAL (00CM)...............................

1-5 1.27 OPERABLE - OPERABILITY...............................................

1-5 GRAND GULF-UNIT 1 i

Amendment No. 87

9 INDEX DEFINITIONS SECTION DEFINITIONS (Continued)

PAGE 1.28 OPERATIONAL CONDITION - CONDITION....................................

1-5 1.29 PHYSICS TESTS........................................................

1-5

1. 30 PRE S SURE BOUNDARY L E AKAC E............................................1-5 1.31 PRIMARY CONTAINMENT INTEGRITY........................................

1-6 1.32 PROCESS CONTROL PROGRAM (PCP)........................................

1-6 1.33 PURGE - PURGING......................................................

1-6 1.34 RATED THERMAL P0WER..................................................

1-6 1.35 REACTOR PROTECTION SYSTEM RESPONSE TIME..............................

1-7 1.36 REPORTABLE EVENT.....................................................

1-7 1.37 ROD DFNSITY..........................................................

1-7 1.38 SECONDARY CONTAINMENT INTEGRITY......................................

1-7 1,39 SHUTDOWN MARGIN......................................................

1-8 e

1.40 SITE 80VNDARY.......................

1-8 1.41 DELETED...............................................................

1-8 1.42 DELETED..............................................................

1-8 1.43 STAGGERED TEST BASIS.................................................

1-8

1. 4 4 T ! i E RMA L P 0WE R........................................................

1-8 1.45 UNIDENTIFIED LEAKAGE.................................................

1-8 1.46 UNRESTRICTED AREA....................................................

1-8 1,47 DELETED..............................................................

1-8 1.48 VENTING..............................................................

1-9 TABLE 1.1.-SURVEILLANCE FRcQUENCY N0TATION................................

1-10 TABLE 1.2, OPERATIONAL CONDITIONS.........................................

1-11 GRAND GULF-UNIT 1 ii Amendmer.t No. 87 1

l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...................

3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.........................

3/4 3-9 3/4.3.3 "ERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.....

3/4 3-27 3/4.3.4 RECIRC'LATION PUMP TRIP ACTUATION INSTRUMENTATION J

ATWS Recirculation Pump Trip System Instrumentation.........

3/4 3-37 End of-Cycle Recirculation rump Trip System Instrumentation.............................................

3/4 3-41 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.............................................

3/4 3-47 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION..........................

3/4 3-52 3/4.3.7 HONITORING INSTRUMENTATION Radiation Monitoring Instrumentation........................

3/4 3-58 Seismic Monitoring Instrumentation..........................

3/4 3-63 Meteorological Monitoring Instrumentation...................

3/4 3-66 Remote Shutdown System Instrumentation and Controls.........

3/4 3-69 Accident Monitoring Instrumentation.........................

3/4 3-73 Source Range Monitors.......................................

3/4 3-77 T raversi ng I n-Core Probe Sys tem.............................

3/4 3-79 Chlorine Detection System...

3/4 3-79 Fire Detection Instrument. ion (DELETED)....................

3/4 3-80 Loose-Part Detection Systa..................................

3/(I3-90 Main Condenser Offgas Treatment System - Explosive Gas Monitoring System Instrumentation...........................

3/4 3-91 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION......................

3/4 3-92 3/4.3.9 TURBINE OVERSPEED PROTECTION SYSTEM.........................

3/4 3 GRAND GULF-UNIT 1 v

Andment No. 67, 82,B7 i

.--------------A

<9=. Ls...

4 16, ix LIMITING CONDITIONS FOR OPERAT10N AND SURVEILLANCE REQUIRE SECTION pag 3/4;,4 REACTOR COOLANT SYSTEM 3/4.4.3 RECIRCULATION SYSTEM t

i Recirculation Loops.............................

3/4 4 1 Jet Pumps.......................................

3/4 4 2 R e c i rc ul a t i o n L t op F 1 ow....................................

3/443 Idle Recirculation Loop 5tartup............................

3/4 4 4 3/4.4.2 SAFETY VALVES Safety / Relief Valves.......................................

3/4 4-5 Saf ety/ Relief Valves Low-Low set Function..................

3/4 4 7 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection 5ystems..................................

3/4 4 8 Operaticnal Leakage........................................

3/4 4-9 3/4.4.4 CHEMISTRY..................................................

3/4 4 13 3/4.4.5 SPECIFIC ACTIVITY..........................................

3/4 4-16 3/a.4.6 PRESSURE /TEMPERATURC LIMITS Recetor Coolant 5ystem.....................................

3/4 4-19 R e a c t o r S t e a m 00me......................................

3/4 4 23 3/4.4.7 MAIN STEAM LINE ISOLATION VALVE5...........................

3/4 4 24 3// 4.8 STRUCTURA'. INTEGR1TY.......................................

3/4 4 28 3/4.4.9 RESIDUAL HEAT REMOVAL Hot 5hutdown...............................................

3/4 4*26 Cold 5hutdown..............................................

3/4'4-27 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 E C C S - 0 P E RA T I N G..........................................

3/4 5 1 3/4.5.2 ECCS - 5HUTDOWN,...........................................

3/4 5-6 3/4.5.3 SUPPRESSION P00L...........................................

3/4 5-8 GRAND GULF-UNIT 1 vi

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION

,PAGE l

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS I

Liquid Holdup Tanks..........................................

3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mixture........................................

3/4 11-2 Main Condenser...............................................

3/4 11-3 i

I l

l GRAND GULF-UNIT 1 xi Amendment No. 87

m, INDEX BASES SECTION PAGE 3/4.0 APPLICABillTY.......................;.......................

B 3/4 0-1

,3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.....................................

B 3/4 1-1 3/4.1.2 REACTIVITY AN0MAllES................................

B 3/4 1-1 3/4.1.3 CONTROL R0DS........................................

B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTROL5........................

B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.......................

B 3/4 1-4 3/4.2 POWERDISTRIBUTIONLIMITj 3/4.2.1 AVERAGE PLANAR c.!NEAR HEAT GENERATION RATE..........

B 3/4 2-1 3/4.2.2 0ELETED.............................................

B 3/4 2-2 l

3/4.2.3 MINIMUM CRITICAL POWER RAT 10........................

B 3/4 2-4 3/4.2.4 LI N:AR HE AT GENIMATIC'N RAT E.........................

B 3/4 2-7 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...........

B 3/4 3-1 3/4.3.2 ISOLATION A'.TUATION INSTRUMENTATION.................

B 3/4 3-1 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION......................................

B 3/4 3-2 3/4.3.4 RECIRCU! ATION FUMP TRIP ACTUATION INSTRUMEN E 0N.....................................

B 3/4 3-2 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.......,...................

B 3 '4 3-3 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION...................

C 3/4 3-3 GRAND GULF-UNIT 1 xii Amendment No.16

ecicw s

INDEX l

BASES SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...............

B 3/4 3-4 Seismic Monitoring Instrumenktion.................

B 3/4 3-4 Meteorological Monitoring Instrumentation..........

B 3/4 3-4 Remote Shutdown System Instrumentation and Controls.........................................

B 3/4 3-4 Acci m.t Monitoring Instrumentation................

B 3/4 3-4 Source kange Monitors..............................

B 3/4 3-5 Traversing In-Core Probe System....................

B 3/4 3-5 Chlorine Detection System..........................

B 3/4 3-5 Fire Detection Instrumentation (DELETED)...........

t1 3/4 3-5 Loose-Part Detection System........................

B 3/4 3-6 Main Condenser Offgas Treatment System - Explosive Gas Monitoring System Instrumentation.............

B 3/4 3-6 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION............

B 3/4 3-6 3/t 3.9 TURBINE OVERSPEED PROTECTION.......................

B 3/4 3-7 3/4 J 10 NEUTRON FLUX MONITORING INSTRUMENTATION............

B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4 i RECIRCULATION SYSTEM...............................

B 3/4 4-1 j

3/4.4.2 SAFETY / RELIEF VALVES...............................

B 3/4 4-2 l

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE l

Leakage Detection Systems..................

B 3/4 4-2 Operational Leakage.................

B 3/4 4-2 1

3/4.4.4 CHEMISTRY..........................................

B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY..................................

B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS........................

B 3/4 4-4 l~

I 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...................

B 3/4 4-5 3/4.4.8 STRUCTURAL INTEGRITY...............................

B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REM 0 VAL..............................

B 3/4 4-5 GRAND GULF-UNIT 1 xiii Amendment No. 16, 82, 87

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1 t

i l$U 8Ast5 i

$ttTION L%l 1/4.5 pit 0EV CORf CO, CLING SY$7tM 3/4.5.1/2 ZCC5 OPERATING and SNUTD0WN......................

l 8 3/4 6 1 3/4.5.3

$UPPAl$$10N P00L...................................

8 3/4 5 2 2,/4.6 CONTA!NMENT $YSTEMS 3/4.6.1 PA! MARY CONTA!NMENT Prima ry Containeent Integri ty......................

B 3/4 6-1 4

' Containment Leakage................................

8 3/4 6 1 l

Containeer.i Air Locks..............................

8 3/4 6 1 M51V Leakage Cen' trol 5ystes........................

B 3/4 6 1 Feedwater Leakage Control 5ystes...................

B 3/4 6-2 Containeent Structural Integrity...................

8 3/4 6 2 Containment Internal Pressure......................

8 3/4 6 2 Containrent Average Air Temperature................

8 3/4 6 2 Containment Purge Systes...........................

8 3/4 6 2 6

3/4.3.2 DRYWELL D rywe l l 2 n t e g r i ty..................................

8 3/4 6-3 Dryweli 8, Tass Leakage.............................

B 3/4 6-3 O rywe l l A i r Loc ks..................................

8 3/4 6 3

-- Drywell Structural Integrity.......................

8 3/4 6 4 Drywell Internal Pressure..........................

'8 3/4 6 4 Orywell Average Air Temp'erature....................

8 3/4 6-4 Crywell Vent and Purge.............................

B 3/4 6 4 3/4.6.3 DEPRES$URIZATION SYSTEMS........................

8 3/4 6 4 3/4.6.4 CONTAINMENT AND OFWELL ISOLATION VALVES...........

8 3/4 6-7, 3/4.6.5 DRYWELLTVACUUM RELIEF..............................

83/46-((

1 3/4.6.6 SECONDARY CONTAINMENT...............;......a.......

8 3/4 6 8 3/4.6.7 ~ ATMOSPHERE CONTR0L.................................

8 3/4 6-9 2.

4

)

GRAND GULF UNIT 1 xiv Amendment No. 21

~1.

Effective Date:

October 20. 1986 4

..u....

.. _.w,,_,.

.m.m.-

INDEX SASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS..............................

B 3/4 7-1 l

3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM...........

B 3/4 7-1 l

l 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM..............

B 3/4 7-1 3/4.7.4 SNUBBERS...........................................

B 3/4 7-2 3/4.7.5 SEALED SOURCE CONTAMINATION........................

B 3/4 7-3 3/4.7.6 FIRE SUPPRESSION SYSTEMS (DELETED).................

B 3/4 7-4 3/4.7.7 FIRE RATED ASSEMBLIES (DELETED)....................

B 3/4 7-4 3/4.7.8 AREA TEMPERATURE W. 7 RING........................

B 3/4 7-4 3/4.7.9 SPENT FUE L STORAGE PWL TEMPERATURE................

B 3/4 7-5 3/4.7.10 FLOOD PROTECTION...................................

B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, A.C. SOURCES, D.C. SOURCES and ONSITE POWER and 0:STRIBUTION SYSTEMS...............................

B 3/4 6-1 3/4.8.3 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES............

B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH................................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION....................................

B 3/4 9-1 3/4.9.3 CONTROL ROD POSITION..........

4 B 3/4 9-1 3/4.9.4 DECAY TIME.........................................

B 3/4 9-1 3/4.9.5 COMMUNICATIONS.....................................

B 3/4 9-1 3/4.9.6 REFUELING EQUIPMENT................................

B S/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE P00LS...............................

B 3/4 9-2 l

GRAND GULF-UNIT 1 xv Amendment No. 82 l

i

.NDEX BASES SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.B and 3/4.9.9 WATER LEVEL - REACTOR VES:.EL and WATER LEVEL - SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE P00LS.....................

B 3/4 9-2 3/4.9.10 CONTROL ROD REMOVAL...............................

B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION......

B 3/4 9-2 3/4.9.12 HORIZONTAL FUEL TRANSFER SYSTEM....................

D 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /0RYWELL INTEGRITY.,

B 3/4 10-1 3/4.10.2 R00 PATTERN CONTROL S' STEM........................

B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.....................

B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS................................

B 3/4 10-1 3/4.10.5 TRAINING STARTUPS..............

B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 L1 QUID EFFLUENTS....................................

B 3/4 11-1 Liquid Holdup Tanks.................................

B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS...................................

B 3/4 11-2 GRAND GULF-UNIT 1 xvi Amendment No. 87

A aw...

INDEX BASES

~'

SECTION P (>E RADIOACTIVE EFFLUENTS (Continued)

Explosive Gas Hixture...............................

B 3/4 11-2 Main Condenser......................................

B 3/4 11-2 i

GRAND GULF-UNIT 1 xvii Amendment No. 87 i

l

y i

i INDEX.

DESIGN FEATURES SECTION

-.GE PA --

5.1 SITE Exclusion Area.....................................................

6-1 tow Population Zone................................................

51 4

Unrestricted Area ard Site Boundary for Gaseous Effluents and for Liquid Effluents............................................

5 '1 5.2 CONTAINMENT Configuration......................................................

5-1 Design Temperature and Pressure.................................,..

5-1 Secondary Containment..............................................

5-1

5. 3 REACTOR CORE fuel Assemblies....................................................

55 Control Rod Assemblies.............................................

5-5 d

5. 4 REACTOR COOLANT SYSTEM

' sign Pressure and Temperature....................................

5-5 i

o1ume.............................................................

5-5 b.S METEOROLOGICAL TOWER LOCATION......................................

56 5.6 FUEL STORAGE h

Criticality........................................................

56 4

0rainage...........................................................

5-6 Capacity...........................................................

5-6 5.7 COMP 0hjyT CYCLIC OR TRANSIENT LIMIT................................

56

~

I GRAND GULF-UNIT 1 xviii

= _.

M.*

e

_INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 61 RESPONSl!L,ITY....................................................

6-1 n

6.2 ORGANIZATION 6.2.1 0FF$1TE AND ONSITE ORGANIZATIONS..............................

61 l 6.2.2 UNIT STAFF....................................................

6-1 6.2.3' INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function......................................................

6-2 Composition...................................................

62 Responsibilities............................

66 Authority.....................................................

66 6.2.4 SHIFT TECHHICAL ADVIS0R.......................................

6-6 6.3 UNIT STAFF QUALIFICATIONS..........................................

6-6 6.4 TRAINING...........................................................

66 6.5 REVIEW AND AUDIT

6. 5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)

Fu,1ction......................................................

6-6 Composition...................................................

6-7 i

Alternates....................................................

67 l

Meeting Frequency.............................................

67 l'

Quorum.....................,...................................

6-7 Responsibilities..............................................

6-7 l

Authority............................

6-8 p

Records.......................................................

6-8 6.5.2 SAFETY REVIEW COMITTEE (SRC)

Function......................................................

69 Composition...................................................

6-9 Alternates....................................................

6-9 Consultants...................................................

6-10 Meeting Frequency.............................................

6-10 l

Quorum........................................................

6-10 1

GRAND GULF-UNIT 1 xix Amentiment No. 45 l

l a

a.-.---..

INDEX ADMINISTRATIVE CONTROLS SEC'10N F;JE SAFETY REVIEW COMMITTEE (SRC) (Continued)

Review.......................................................

6-10 Audits.......................................................

6-11 Authority....................................................

6-12 Records......................................................

6 12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...................................................

6-12

6. 6 REPORTABLE EVE.NT_ ACTION...........................................

6-13 6.7 SAFETY LIMIT VIOLATION............................................

6-13 6.8 PROCEDURES AND PR0 GRAMS...........................................

6-14 6.9 REPORTING REQUIREMENTS Routine Reports..............................................

6-15 Startup Reports..............................................

6-15 Annual Reports...............................................

6-16 Annual Radiological Environmental Operating Report...........

6-17 Semiannual Radioactive Effluent Release Report...............

6-17 Monthly Operating Reports....................................

6-19 Special Reports..............................................

6-19 6 10 RECORD RETENTION.................................................

6-19 6.11 RADIATION PROTECTION PR0 GRAM.....................................

6-21 6.12 HIGH RADIATION AREA..............................................

6-21 6,._13 PROCESS PROGRAM CONTROL (PCP)...................................

6-22 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)...........................

6-22 t

I l

GRAND GULF-UNIT 1 xx Amendment No. 87 l

I

DEFINITIONS I AVERAGE DIS 1NTEGRATION ENERGY 1.11 I shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and ganma energies per disintegration, in Hey, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-lodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.12 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e.

the valves travel to their required positions, pump discharge pressures re,ach their required values, etc.

Times shall include diesel generator startin0 and sequence loading delays where applicable.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response tin,e is measured.

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a.

Turbine stop valve., and b.

Turbine control valves.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured, except for the breaker arc suppression time which is not measured but is validated by sur-veillance tests to conform to the manufac;urer's design value.

FRACTION OF LIMITING POWER DENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the limiting LHGR for that bundle type.

FRACTION OF RATED THERMAL POWER l

1.15 Tne FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER diviced by the RATED ThiRMAL POWER.

e I

FREQUENCY,N_0TATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

1.17 DELETED l

l GRAND GULF-UNIT 1 1-3 Amendment No. XB, 87

m

~. - - -. -..

.. =

.a c,

OEFINITIONS 10ENTIFIEDLEAKAGE 1.18 10EN11FIED LEAKAGE shall be:

Leakage into collection systees, such as pump seal or valve packing a.

leaks, tt.at is captured and conducted to a sump or collecting tank, or l.eakage into the dryvell atmosphere free sources.that are both b.

specific. ally located and known either not to interfere with the opera-tion of the leakage detection systems or not to be PRES $URE BOUNDARY LEAKAGE.

150LAT10N SYSTEM RESPONSE TIME 1.19 The 150LAT10N SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

Times shall include diesel generator starting and sequence loading delays where applicable. The response time may bt measured-by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL ROD PATTERN 1.20 A LlH111NG CONTROL ROD PATTERN shall be a core being on a thermal hydraulic limit, i.e., pattern which results in theoperating on a limit g

for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LNGR) shall be the heat generation per unit length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1,22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all lenic components, i.e., all relays and contacts, all tri etc.,ofaIcgiccircuit,fromsensor.punits,solidstateloS.: elements, through and including ihs actuated devir.e. to verify DPERABILITY. The LOGIC SYSTEM FUNCT10NAL TEST may be perforsed by any series of sequential, overlapping or total systen steps such that the entire logic system is tested.

max 1 MUM FitACTION OF LIMITING POWER DENSITY 1.23 The MAXIMUM FRACTION OF LIMITING POWER DFNSITY (MFLPD) shall be the highest value of the FLPD which exists in the core.

MEMBER (5) 0F THE PUMIC 1.24 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include esiployees of the ptility, its contractors or vendors. Also excluded from this categor eries. y are persons who enter the site to service equipment or to make deliv-This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

GRAND GULF-UNIT 1 1-4 l

m.

DEFINITIONS MINIMUM CRITICAL POWER RATIO 1.25 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE CALCULATION MAMML (ODCM) 1.26 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents,ints, and in the conduct of the Environ-in effluent nonitoring Alarm / Trip Setpo mental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.3 and (2) descriptions of the information that should be included in De Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6, 6.9.1.7, 6.9.1.8 and 6.9.1.9.

OPERABLE - OPERARILITY 1.27 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION - CONDITION 1.28 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1,29 PHYSICS TESTS shall be those tests performed to measure the fundamental nu: lear claracte-is'.' : of the u actor core and related instrumentation and

1) descrioed in Cnapter 14 of the FSAR, 2) authorized under the prov ons of-10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.30 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.

GRAND GULF-UNIT 1 1-5 Amendment No. 87

~ -.- -_

DEFINITIONS PRIMARf CONTAINMENT INTEGRITY 1.31 PRIMARY CONTAINMENT INTEGRITY shall exist when:

All cont:inment penetrations required to be closed during accident a.

conditions are either:

1.

Capable of being closed by an OPERABLE containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification 3.6.4.

b.

The containment equipment hatch is closed and sealed, Each containment air lock 1 in compliance with the requirements of c.

Specification 3.6.1.3.

d.

The containment leakage rates are within the limits of Specification 3.6.1.2.

The suppression pool is in compliance with the requirements of e.

Specification 3.6.3.1.

f.

The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM (PCP) 1.32 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that proces-sing and packaging of solid radioactive wastes based on demonstrated processing of sctual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste, PURGE - PURGING 1.33 PURGE or PuttGING is the controlled process of discharging air or gas from a confinement to maintain temperature, aressure, humidity, concentration or other operating condition, in r,uch a manner t1at replacement air or gas is required

~to purify the confinement.

RATED THERMAL POWER 1.34 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the rc:: tor esolant of 3833 MW.

c \\ND GULF-UNIT 1 1-6 Amendment No. 87 m

m l

DEFINITIONS REACTOR PROTECTION SYSTEM RESPONSE TIME 1.35 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REPORTABLE EVEkT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

ROD DENSITY 1.37 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.38 SECONDARY CONIAINMENT INTEGRITY shall exist when:

a.

All Auxiliary Building and Enclosure Building penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Cics e:: by at least ene manual valve, bli-d flange, reture disc or deactivated automatic valve or damper, as applicable, secured in its closed position, except as provided in Table' 3.6.6.2-1 of Specification 3.6.6.2.

b.

1,11 Auxiliary Building and Enclosure Building equipment hatches and blowout panels are closed and sealed.

c.

The standby gas treatment system is in compliance with the require-ments of Specification 3.6.6.3.

l d.

The door in each access to the Auxiliary Building and Enclosure Building is closed, except for normal entry and exit.

e.

The sealing mechanism associated with each Auxiliary Building arfd Enclosure Bui.1 ding penetration, e.g., welds, bellows or 0 rings, is OPERABLE.

l GRAND GULF-UNIT 1 1-7 AmendmentNo.52 l

DECitIF

W.kma w a.-

DEFINITIONS SHUTDOWN MARGIN 1.39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subtritical or would be suberitical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e.

68'f; and xenon free.

SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

1.41 DELETED 1.42 DELETED STAGGERED TEST BASIS 1.43 A STACCERED TEST BASIS shall consist of:

A test schedule for n systems, subsystems, trains or other designated a.

components obtained by dividing the specified test interval into n equal subintervals.

b.

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

THERMAL POWER 3.44 THERMAL POWER shall be the total reacto^ core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.45 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTlflED LEAKAGE.

UNRESTRICTFD AREA 1.46 An UNRESTRICTED AREA shall be any area at or beycnd the SITE BOUNDARY ac-cess to which is not controlled by the licensee for r,urposes of protection of MEMBERS OF THE PUBLIC from exposure to radiation and r adioactive materials, or any area within the SITE BOUNDARY used for resider,tial quarters or for indus-trial, commercial, institutional, and/or recreational purposes.

1.47 DELETED GRAND GULF-UNIT 1 1-8 Amendment No. 87

l INSTRUMENTATION i

MAIN CONDENSER OFFGAS TREATMENT - EXPLOSIVE GAS MONITORING SYSTEM TETRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.12 At least one explosive gas monitoring system instrumentation channel of the main condenser offgas treatment system shall be OPERABLE with its alarm setpoint set to ensure that the limit of Specification 3.11.2.6 is not exceeded.

APPLICABILITY:

During main condenser offgas treatment system operation ACTION:

a.

With the explosive gas monitoring system instrumentation channel

- alarm setpoint of the main condenser offgas treatment system less conservative that required by Specification 3.11.2.6, declare the channel inoperable and take the ACTION required below.

b.

With lets than one explosive gas monitoring system instrumentation channel of the main condenser offgas treatment system OPERABLE, oper-ation of the main condenser offgas treatment system may continue for up to 30 days provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Restore the in-operable-instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission t

pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner, The provisions of Specifications 3.0.3 and 3.0.4 are not applic'able.

~

c.

SURVEILLANCE REQUIREMENTS 4.3.7.12-Each explosive gas monitoring. system instrumentation channel'of the main condenser offgas treatment system shall be demonstrated OPERABLE by

- perf

nance of a CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a CHANNEL CALIBRA-TION" at least once per 92 days and a CHANNEL FU,wIION TEST at least once per 31 days..

"The CHANNEL CALIBRATION shall include the use of standard gas samples containing a_ nominal:

1.-

One volume percent hydrogen, balance nitrogen, and j

2.

Four volume percent hydrogen, balance nitrogen.

l l

GRAh'1 GULF-UNIT 1 3/4 3-91 Amendment No. 87 4m--m,~,--

--- n r e - r-er r-s 1

.v-me..,,w,mw-

,m

,,-,mn--m.-m,m-e.,rm-:vm.

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, w ere v -v e mm ent,m,,m-ww,-a -,m

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INSTRUMENTATION 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION INSTRUMENTATION 3/4.3.8 PLANT SYSTEMS ACTUA110N INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.8 The plant systems actuation instrumentation channels shown in Table 3.3.8-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.8-2.

APPLICABILITY:

As shown in Table 3.0.8-1.

ACTION:

With a plant system actuation instrumentation channel trip setpoint a.

less conservative th6n the value shown in the Allowable Values column of Table 3.3.8-2, declare the channel inoperable and take the ACTION required by Table 3.3.8-1.

b.

With one or more plant systems actuation instrument channels in-operable, take the ACTION required by Table 3.3.8-1.

SURVE1LLANCE REQUIREMENTS 4.3.8.1 Each plant system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CAllBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.8.1 1.

4.3.8.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

GRAND GULF-UNIT 1 3/4 3-92 Amendment No. 87

4 m

TABLE 3.3.8-1 5'

p, PLANT SYSTif15 ACTUATION INSTRUffENTATION c3 c:

5 -

MINIMUM APPLICABLE OPERABLE CHANN OPERiJ IONAL PERTRIPSYSTEMg g

TRIP FUNCTION CONDITIONS ACTION

[

1.

CONTAINMENT SPRAY SYSTE7 l

a.

Drywell Pressure-High 2

1, 2, 3 130 b.

Containment Pressure-High 1

1,2,3 131 c.

Reactor Vessel Water Level-Low i

Low Loa, level 1 2

1,2,3 130 d.

Timers

1) System A 1

1,2,3 131

2) System B 1

1,2,3 131 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM a.

Reactor Vessel Water Level-High, level 8 3

1 132 3.

SUPPRESSION POOL MAKEUP SYSTEM a.

Drywell Pressure - High (ECCS) 2 1,2,3 135 b.

Drywell Pressure - High (RPS) 2 1,2,3 135 c.

Reactor Vessel Water Level - Low Low Low, Level 1 2

1,2,3 135 d.

Reactor Vessel Water Level - Low

[

Low, Level 2 2

1,2,3 135 e.

Suppression Pool Water level - Low low 1

1,2,3 133 5

f.

Suppression Pool Makeup Timer 1

1, 2, 3 133 I

g.

SPMU Manual Initiation 2

1,2,3 134 0

O (a)A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of required surveillance provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

we l

TABLE 3.3.8-1 (Continued)

PLANT SYSTEMS ACTUATION INSTRUMENTATION ACTION ACTION 130 -

a.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels aer Trip System requirement, place the inoperable channel in tie tripped condition within one hour; otherwise, declare the associated containment spray system inoperable and take the action required by Tech-nical Specification 3.6.3.2.

b.

With the number of OPERABLE channels two less than required by the Minimum OPERABLE channels per Trip System require-ment, declare the associated containment spray system inoperable and take the action required by lechnical Specification 3.6.3.2.

ACTION 131 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channels to OPERAB E status within one hour; other-wise, declare the associated containment spray system inoperable and take the action required by Technical Specification 3.6.3.2.

ACTION 132 -

For the feedwater system / main turbine trip system:

a.

With the numbei of OPERABLE channels one less than required by the Minimum roERABLE Channels requirement, restore the inoperable chan.iel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With the number of OPERABLE chant.els two less than reouired by the Minimum OPERABLE Channels per Trip System require-ment, restore at least one of the inoperable channels to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 133 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip _ System requirement, declare the associated suppression pool makeup system inoperable and take the action required by Specification 3.6.3.4.

ACTION 134 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, castore the inoperable channels to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; other-wise, declare the associated suppression pool makeup system inoperable and take the action required by Specification 3.6.3.4.

ACTION 135 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, place the inoperable channel in the tripped condition within one hour or declare the associated system (s) inoperable.

b.

With more than one channel inoperable, declare the associated system (s) inoperable.

GRAND GULF-UNIT 1 3/4 3-94 Amendment No. 69. 87

\\'

'l.l

$1 Ok

.y:

~

l e

r f

.p ri

~

i h

h TALE 3.3.8-2 5

PLA!'I SYSTEMS ACTUATICLINSTRUMEi:'" 104 SETPOINTS o

~~

C A([I

'*U g

TRIP FUNCTION TRIP SETNINT Va..

[

1.

CONTAINMENT ; PRAY SYSTEM a.

Dr>well Pressure-High

< 1.39 psig

< 1.44 psig b.

Containment Pressure-High

~ 7.44 psig 7 8.34 osig c.

Reactor Vessel Water Level-Low

~

1.cw Low, Level 1

> - 150.3 int.hes 1

57 : inches d.

Timers

1) System A 10.85 + 0.10 minutes 10.26 - < 4, + 1.18 minutes

2) System B 10.851c.10 minutes **

10.26 - 0.00, + 1.18 minutes l

2.

FEEDWATER SYSTtM/ MAIN TURBINE TRIP SYSTEM a.

Reactor Vessel Wa' r Level-High, Level 8

$ 53.5.sches*

5 54.1 inches 3.

SUPPRESSION POOL MAKEUP SYSTEM a.

Drywell Pressore - High (ECCST

$ 1.39 psig 5 1.44 psig b.

Drywell Pressure - High (R$.

$ 1.23 psig 5 1.43 psig c.

Reactor Vesse? Water Level

. a Lou Low, Leve 1 1 -150.3 inches

1 -152.5 inches d.

Reactor Vessel Water Level - Low Low, Level 2 1 -41.6 inches

1-43.8ii.an y

e.

Suppression Pool Water Level - Low law 1 17 ft 5 inches 3 17 ft 2 inches

[

f.

Suppression Pool Makeup Timer 5 29.0 minutes

$ 29.5 minutes g.

SPHU Manual Initiation NA NA E

i

- See Bases Figure B 3/4 3-1.

Setpoint for System B is the sum of E12-K0938 plus E12-K116.

E12-K116 is not to exceed 10.00 seconds.

o S

i 1

TABLE 4.3.8.1-1

=

[

PLANT Sf,i $ ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS C

'm-E:

i 5

CilANNEL OPERATIONAL C'IANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

[ TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIRED l

1.

CONTAINMiNT SPRAY SYSTEM Rfa a.

Drywell Pressure-High 5

K a

1, 2, 3 b.

Containment Pressure-High S

M R

1, 2, 3 c.

Reactor Vessel Water Level -

Low Low Low, level 1 S

M R(a) 1, 2, 3 d.

Timers NA M

Q 1, 2, 3 2

FEEDWATER SYSTEM / MAIN TURBINE TRIP SY5lLM w

a.

Reactor Vessel Water level-liigh, 5

M R

1 g

Level 8 3.

SUPPRESSION POOL MAKEUP SYSTEM a.

Drywell Pressure - High (ECCS)

S M

R(a) 1, 2, 3 b.

Drywell Pressure - liigh (RPS)

S M

R(a) 1, 2, 3 c.

Reactor Vessel Water Level - Low Low Low, Level 1 S

M R(a) 1, 2, 3 3"

d.

Reactor Vessel Water Level - Low

[

Low, Level 2 5

M R(a) 7, 7, 3 9

8 e.

Suppression Pool Water Level -

S M

R(a) 1, 2, 3 low LOW f.

Suppressiop Pool Makeup Timer HA M

Q 1, 2, 3 3

g.

SPMU Manual Initia+ ion NA R

NA 1,2,3 (a) Calibrate trip unit at least once per 31 days.

~

\\

INSTRUMENTATION l

3/4.3.9 TURBINE OVERSPEED PROTECTION SYSTEM i

LIMITING CONDITION FOR GPERATION 3.3.9 At least one turbine overspeed protection system shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

Uith one stop valve or one control valve per high pr?sr.re turbine a.

steam line inoperable and/or with one stop valve or one control valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or close at least one valve in the affected steam line.

Otherwise isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With the above required turbine overspeed protection system otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply.

SURVEILLANCE REQUIREMENTS 4.3.9.1 The provisions of Specification 4.0.4 are not applicable.

4.3.9.2 The above required overspeed protection system shall be demon-strated OPERABLE:

2 At least once per 14 days by cycling each of the following valves through at least one complete cycle from the running position using the manual test or Automatic Turbine Tester (ATT):

1)

Four high pressure turbine stop valves,

?)

Four high pressure turbine control valves, 3)

Six low pressure turbine stop valves, and 4)

Six low pressure turbine control valves, b.

At least once per 14 days '. testing of the two mechanical overspeed devices using the Automatic Turbine Tester or manual test, c.

At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws.

If unacceptable flaws are found, all other valves of that type shall be inspected.

l GRAND GULF-UNIT 1 3/4 b-97 Amendment No. 87

[

3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS L10VIDHOLDUPTANg LIMITING CONDITION FOR OPERATION s

i 3.11.1.4 The quantity of radioactive material contained in any outside tempo-rary tank, not including liners for shipping radwaste, shall be limited to

-less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above 4

specified tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tanks and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit and describe the eventsleadingtotheconditioninthenext$emiannualRadioactive Effluent Release Report.

b.

The provisions of Specification 3.0.3 are not applicable.

4 SURVEILLANCE REr"IIREMENTS 4.21.1.4 Tr: quantity of radioactive material contained in each of the above specified tar.ks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactite materials are being added to the tank.

r r

a

~

GRAND GULF-UNIT 1 3/4 11-1 Amendment No. W, 07

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

RADI0 ACTIVE EFTLUENTS 3/4.11.2 GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be limited to less than or equal to 4% by volume.

APPLICADILITY: Whenever the main condenser offgas treatment system is in operation.

ACTION:

With the concentration of hydrogen in the main condenser offgas a.

treatment system exceedin the limit, restore the concentration to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The oncentration of hydrogen in the main condenser offgas treatment system shall bs determined to be within the above limits by monitoring the waste gas in the main condenser off gas treatment system with_the hydrogen monitor OPERABLE as required by Specification 3.3.7.12.

l E

GRAND GULF-UNIT 1 3/4 11-2 Amendment No. 87

1 RADI0 ACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.7 The gross radioactivity (gamma) rate of the noble gases measured at the offgas recombiner effluent shall be limited to less than cr equal to 380 millicuries /second, after 30 minutes decay.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2,* and 3*

ACTION:

With the gross radioactivity rate of the noble gases at the offgas recombiner effluent exceeding 380 millicuries /second after 30 minutes decay, restore the gross radioactivity release rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HCT 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 4.11.2.7.1 The radioactivity release rate of noble gases near the ottlet of the main condenser air.ojector shall be continuously monitored in occordance with the ODCM.

[

4.11.2.7.2 The gross radioactivity release rate of the noble gases from the main condenser air ejector shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies ** by performing an isotopic analysis of a representative sample of gases taken at the offgas recombiner effluent:

a.

At least once per 31 days, b.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Offgas Pretreatment Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in the nominal steady state fission gas release from the prima., coolant.

When the main condenser air ejector is in operation.

- The provisions of Specification 4.0.4 are not applicable.

GRAND GULF-UNIT 1 3/4 11-3 Amendment No. 87 l

lab.h i

INSTRUMENTATION BASES 3/4.3.7.E SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions should not be side without this flux level information available to the operator. When the inter-mediate range monitors are on scale adequate information is available without the SRMs and they can be retracted.

The SRMs are required OPERABLE in OPERATIONAL CONDITION 2 to provide for rod block capability, and are required OPERABLE in OPERATIONAL CONDITIONS 3 and 4 to provide monitoring capability which provides diversity of protection to the mode switch interlocks.

3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEF The OPERABILITY of the traversing in-core probe syster. wit.5 the specified minimum complement of equipment ensures that the measurements cbiained from use of this equipment accurate'er t epresent the spatial neutron flux distribution of the reactor core.

The TIP system 0 9 ABILITY is demonstrated by normalizing all probes (i.e., detectors) priv v performing an LPRM calibration function. Monitoring core thermal limits may

.volve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be 08ir: E_E.

The OPERASILITY of individual detectors to be used for ronitoring is cemenstratec ty ccmparing the detector (s) c;t;ct with cate ottained curing the previous LFRM calibrati:ns.

3/4.3.7.8 CHLORINE DETECTION SYSTEM DELETED 3L4.3.7.9 FIRE DETECTION INSTRUMENTATION DELETED l

GRAND GULF-UNIT 1 B 3/4 3-5 Amendment No. 75,82

INSTRUMENTATION BASES 3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the primar.y system and avoid or mitigate damage to primary system components.

The system consists of 16 sensors of which only 8 are selected and need to be OPERABLE at a time, l

to provide the inputs to the 8 monitoring channels.

The remaining 8 sencers may be used as replacement sensor inputs for failed sensors or to provide a change in location of the area being monitored.

The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System i

of Light-Water-Cooled Reactors," May 1981, 3/4.3.7.12 MAIN CONDENSER OFFGAS TREATMENT SYSTEM - EXPLOSIVE GAS MONITORING SYSTEM INSTRUMENTATION The explost 2 gas monitoring system instrumentation of the main condenser off-gas treatment system is provided to monitor the concentrations of potentially explosive gas mixtures in the main condenser offgas treatment system.

This instrumentation is calibrated in accordance with plant procedures.

l 3/4.3.8 PLANTSYSTEMSACTUATIONINSTRUMENTATION The plant systems actuation instrumentation is provided to initiate action to mitigate the conse operator to control. quences of accidents that are beyond the ability of the The LPCI mode of the RHR system is automatically initiated on a high drywell pressure signal and/or a low reactor water level, level 1, signsl.

The containment spray system will then actuate automatically following high drywell and high containment pressure signals.

Negative barometric pr:s-sure fluctuations are accounted for in the trip setpoints and allowable values l

specified for drywell and containment pressure-high.

A 10-minute minimum, 13-minute maximum time delay exists between initiation of LPCI and containment spray actustion.

A high reactor water level, level 8, signal will actuate the feedwater system / main turbine trip system.

The supprusion pool makeup system is automatically initiated on a low low suppression pool water level signal with a concurrent LOCA signal or following a specified time delay after recelpt of a LOCA signal.

The low low suppression pool water level Trip Setpoint and.

Allowable Value are relative to the surface flour ef the suppression pool l

(93'Oh" above mean sea leve D.

l GRAND GULF-UNIT 1 B 3/4 3-6 Amendment No. 60, 87

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3/4.11 RADIDACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS I

3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those tanks containing radioactive material that are not surrounded by liners, dikes, or walls capable of holding the contents and that do not have overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncor. trolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

i GRAND GULF-UNIT 1 B 3/4 21-1 Amendment No. 87

RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2 GASE0US EFFLUENTS 3/4.11.2.6 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially Explosive gas mixtures contained in the offgas holdup pi)ing is maintained below the flammability limits of hydrogen.

Maintaining tie concentra-tion of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.7 MAIN CONDENSER Restricting the gross radioactivity rate of noble gaser from the main condenser provides reasonab'e assurance that the total body exposure to a MEMBER OF THE PUBLIC at and beyond the SITE B0UNDARY will not exceed a small fra: tion of the limits of 10 CFR Part 100 in the event this effluent is inad-vertently discharged directly to the environment without treatment.

This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

GRAND GULF-UNIT 1 B 3/4 11-2 Amendment No. 87

ADMINISTRATIVE CONTROLS FROCEDURES AND PROGRAMS (Con',inued) 6.

Feedwater leakage ontrol system.

7.

Post-accident sampling system.

8.

Suppression pool level detection portion of the suppression pool maceup system.

The program shall include the following:

1.

Preventive maintenance and periodic visual inspection requirements, and 2.

Integrated leak test requirements for each system at refueling cycle intervals or less, b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

Training of personnel,ing, and 1.

Procedures for monitor 2.

3.

Provisions for maintenance of sampling and analysis equipment.

c.

Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The progr u shall include the following.

1.

Training of personnel, 2.

Procedures for sampling and analysis, 3.

Provisions for maintenance of sampling and analysis equipment, d.

Radioactive Ef'luent Controls'Procram A program shall be provided :onforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable The program (2)shallbeimlementedbyope(1)shallbecontainedinthe ODCM rating 3rocedures,and(3) shallincluderemedia actions to be taken wienever the program limits are exceeded.

The program shall include the following i

elements:

l 1.

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation includina surveillance tests and set)oint determination in accordance wif.h the methodology in the 0DCM, I

GRAND GULF-UNIT 1 6-15 Amendment No. 80, 87

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ADMINISTRATIVE-CONTROLS PROCEDURESANDPROGRAMS(Continued) 2.

-Limitations on the concentrations of radioactive w'erial released in liquid effluents to UNRESTRICTED AREAS conformir:g to 10 CFR Part 20, Appendix B, Table II, Column 2, Monitoringhuents in,accordance with 10 CFR 20.106 and withand analy 3.

sampling gaseous ef the methodology and parameters in the ODCM, 4.

Limitations on the annual and quarterly doses or dose commitment to a MEMBER of the PUBLIC from radioactive materials in liquid effluents released from thu unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5.

Determination of cumulative and arojected dose contributions from radioactive effluents for tie current calends.r quarter and current calendar year in accordance with the.aethodology and parameters.in the ODCH at least every 31 oays, 6.

Limitat0ons on the operabiitty and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions

-of these systems are used to reduce releases-of radioactivity when the proiected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conform-ing to A.ppendix I to 10 CFR Part 50, 7.

Limitations on the dose rate resulting from radioactive material

. released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appen-dix B,; Table II, Column I,

-8.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the unit to areas beyond the SITE BOUNDARY-conforming to Appendix I to 10 CFR Part 50, 9.

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, lives gre,ater th*n 8 day tritium and all radio-nuclides in particulate form with half-in caseous effluents released from the unit to areas Leyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and.

10.

Limitations on the annual dose or dose commitment to any MEMBER OF.THE PUBi.IC due to releases of radioactivity and to radiation-from uranitm fuel cycle sources conforming to 40 CFR Part 190.

e.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environt of the plant.

The program sha11 provide (1) repre-sentative measurenents of radioactivity in the highest potential GRAND GULF-UNIT 1 045a Amendment No. 87

1 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) exposure pathways, and (2) verification of the accuracy of the ef-fluent monitoring program and modeling of environmental exposure pathways.

Th9 program shall (1) be contained in the ODCM, (2) con-form to the guidance of Appendix I to 10 CFR Part 50, and (3) include the foll0 wing:

1.

Monitoring, sampling, analysis, and reporting of rediation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, 2.

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identifieu and that modifica-tions to the monitoring program are made if required by the results of this census, and 3.

Participation in an Interlaboratory Comparison Program to ensure that indeper. dent checks on the precision and accurhey of the measurems.ts of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 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 Nuclear Regulatory Commission pursuant to Section 50.4 of 10 CFR Part 50.

STARTUP REPORTS.

6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) ameadment to the license involving a plar.ned increase in power level, (3) installation of fuel that has a different design or has,been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

l GRAND GUf.F-UNIT 1 6-15b Amendment No. 87

vu m ADMINISTRATIVE CONTROLS STARTUP REFORTS (Continued)

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6. 9.1. 2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the eensured values of the operating conditions or characteristics obtained during the test program and a compari-son of these values with design predictions and specifications.

Any correc-tive actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on otner commitmants shall be included in this report.

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

ANNUAL REP 3RTSM 6.9.1.4 Annual reports covering the activities of the unit as described below for the p evious calendar year shall be submitted prior to March I of each lo ir.g initial criticality.ine initial report shall be submitted prior to March 1 of the year.

6. 9.1. 5.1 cf statien, utility, and ether persont 1,Re;:rts shall it.clude a tabulation on a including contractors, receivin exposures greater than 100 nrem/yr and their associatec manrem exp:sure a:g-cordir; to w:rk and job functions,U

.g., reacter operations ard surveti-e len:e. itse vi:e insce: tion

(:es: rite main'.enan:e), was'.,e pec:ine maintenan:e, s;e:ial eaintenan:e reut essing, and refueling.

te various cuty functions eay be estimated based on pocket dnsimeter, TLD, orTh film tac;e etasurements.

the incivikal total d:se need not be accounted for.Small exposures totalling less tha In the aggragste, at least 60 per:ent of the total whole body dose re:eived from external sources should be assigned to specific major work functions.

6.9.1.5.2 relief valves.Reperts shall irclude do:umentation of all challenges to safety and 6.9.1.5.3 which the primary coolant exceeded tL limits of Specification 3.4.5 lowing informatier shall be included:

The fol-(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 limit, results of analysis while the limit was exceeded, and r analysis after the radiciodine activity was reduced to less than the limit 1/

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

The submittal U should combine those sections that are common to all units at the ' stati This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.

1 GRAND GULF-UNIT 1 6-16 AmendmentNo.28l m

'LC C ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6. 9.1. 6 The Annual Radiologiccl Environuntal Ope ating Report coverino the before May 1 of each year. g the previous calendar yer.r shall be submitted operation of the unit durin 6.9.1.7 The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectivesoutlinedin(1)theODCMand(2)SectionsIV.B.2,IV.B.3,andIV.C of Anpendix I to 10 CFR Part 50.

SEMIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT

6. !>.1. 8 The Semiar.nual Radioactive Efiluent Release Report covering the opera-tion of the unit durit.g the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

6. 9.1. 9 The report shall include a summary of the quantities of '.adioactive liquid and gaseous effluents and solic waste released from the unit.

The mate-rialprovidedshallbe(1)consistentwiththeobjectivesoutlinedintheODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appen-dix I to 10 CFR Part 50.

GRAND GULF-UNIT 1 6-17 Amendment No. 87

ADMINISTRATIVE CONTROLS SEMIANNUALRADI0ACTIVEEFFLUENTRELEASEREPORT(Continued) 4 Ptge intentionally left blank.

GRAND GULF-UNIT 1 6-18 Amendment No. 87

ers,e 3 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.10 Routine reports of operating statistics and shutdown experience valves,gdocumentationofallchallengestomainsteamsystemsafety/rellef includin shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Nuclear Regulatory Commii.sion pursuant to Section 50.4 of 10 CFR Part 50 within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10 Code of Federal Regulations thefollowingrecordsshallberetainedforatleast theminimumperiodindicated.

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

a.

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

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

c.

All REPORTABLE EVENTS.

l 1

l GRAND GULF-UNIT 1 6-19 Amendment No. 80,87

ADMINISTRATIVE CONTR3LS 6.10 RECORDRETENT10_N(Continued) l d.

Records of surveillance activities inspections and calibrations requiredbytheseTechnicalSpeciflcations, Records of changes made to the procedures required by Specification e.

6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak testo 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:

Records and drawing changes reflecting unit design modifications made a.

to systems and equipment described in the Final Safety Analysis Report, b.

Records of new and irradiated fuel invertory, fuel transfers and assembly burnup histories, Records of radiation exposure for all individuals entering radiation c.

control areas.

d.

Records of gaseous and liquid radioactive material released to the environs.

Records of transient or operational cycles for those unit components e.

identified in Table 5.7.1-1.

f.

Records of rea: tor tests anc m$riments.

g.

Records of training and qualification for current members of the unit

staff, Records of in-service inspections performed pursuant to these Technical n.

Specifications.

i.

Records of Quality Assurance activities required by the Operational Quality Assurance Manual not listed in Section 6.10.1.

j.

Records of reviews performed for changes made to arocedures or equipment

. or reviews of tests and experiments pursuant to D CFR 50.59.

k.

Records of mcetings of the PSRC and the SRC.

1.

Records of the service lives of all hydraulic and mechanical snuhbers including the date at which the service life commences and associated installation and maintenance records.

Records of analyses required by the radiological environmental

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monitoring program.

Records of reviews performed for changes made to the OFFSITE DOSE n.

CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

r GRAND GULF-UNIT 1 6-20 Amendment No. 2f, 87 i

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l ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, aaintained and adhered to for all operations involving personnel radiation

exposure, 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiction area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrom/hr shall be barri-caded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation W;rk Permit (RWP).*

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area, b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated,

dose is received.

Entry into such areas with this monitoring device may be made after the dose rate level in the area has'been established and personnel have been made knowledgeable of them, c.

A health physics qualified individual, i.e., qualified in radiation protection procedures, with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the unit Health Physicist in the Radiation l

Work Permit.

6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the bo*Jy r.ould receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the adminis-trative control of the Shift Superintendent on duty and/or t.. unit Radiation Control Supervisor.

Doors shall remain locked except during p riods of access by personnel under an approved RWP which shall specify the dose rate levels in the irmediate work area and the maximum allowable stay time for ir#viduals in that orea.

For indisidual areas accessible tc personnel with rMf a61on levels such that a major-portion of the body could receive in one hour ;, dose in i

" Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiatien protection duties, provided they are otherwise follcwing plant radiation protection procedures for entry into high radiation areas.

GRAND GULF-UNIT 1 6-21

ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA (Continued) excess of 1000 mrem

that are located within large areas, such as the contain-ment, where no enclosure exists for purposes of locking, and no enclosure can be reasonablv constructed around the individual areas, then that area shall be roped off, conspicuoi: sly posted and a flashing light shall be activated as a warning device.

In lieu of the stay time specification of the RWP, continuous surveillance, direct or remote, such as use of closed circuit TV cameras, may be made by personnel qualified in adiation protection procedures to provide positive exposure control over the activities within the area.

6.13 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2n.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.

Shall become effective after review and acceptance oy the PSRC and the approval of the General Manager, Plant Operations.

6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)

Changes to the ODCM:

a Shah be documented and records of reviews performed shall be retained as rec,uired by Specification 6.10.2n.

This occumentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluatiens justifying the change (s) and 2T

determination that the change will maintain the level of

$dioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint chaculations, b.

Shall become effective after review and acceptance by the PSRC and the approval of the General Panager, Plant Operations, Shall be submitted to the Commission in the form of a complete, c.

legible copy of the entire ODCH as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be identifioc. by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

Measurement made at 18" from source of radioactivity.

GRAND GULF-UNIT 1 6-22 Amendment No. 87 m

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Person / Time
ML20086U043
Site:	Grand Gulf (NPF-29-A-087, NPF-29-A-87)
Issue date:	12/18/1991
From:	Larkins J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20086U048	List: ML20086U043 ML20086U051
References
NUDOCS 9201070178
Download: ML20086U043 (48)
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ML20086U043

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