ML20082M888
| ML20082M888 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/23/1991 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082M892 | List: |
| References | |
| GL-88-12, NUDOCS 9109060083 | |
| Download: ML20082M888 (16) | |
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UNITED STATES i-8 NUCLEAR REGULATORY COMMISSION g
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u ENTERGY OPERATIONS, INC.
'I SYSTEM ENERGY RESOURCES, INC.
SOUTH HISSISSIPPI-ELECTRIC POWER ASSOCIATION MISSISSIPPI POWER AND LIGHT COMPANY DOCKET NO. 50-416 i
GRAND GULF NUCLEAR STATION, UNIT I AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 82 License No. NPF-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated May 19, 1987, as revised-August 22, 1990, complies with the st::r.dsrds and requirements of the Atomic Energy Act of 1954, as amended (the ict), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformi?y with the application, the provisions of the Act, and the rules an:4 regulations of the L
Commission; C.
There is reasonable assurance (i) 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'
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in compliance with the Commission's regulations; i
i D.
The issuance-of this amendment will not be inimical to the comm 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-L of the Commission's regulations and all applicable requirements have been satisfied.
l l
9109060003 910023 ADOCK 050 g 6 PDR P
2.
Accordingly, delete Operating License Condition (OLC) 2.C.(23) of Facility Operating License No. NPF-29 and replace it with OLC 2.C.(41) to read as follows:
(41) Fire Protectio,n Program Entergy Operations, Inc. shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in Revisior, 5 to the Updated Final Safety Analysis Report, and as approved in the Safety Evaluation dated August 23, 1991, subject to the following provisions:
The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
3.
Also, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read
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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. 82, are hereby incorporated into this license.
Entergy Operations, Inc. shall operate the facility in actordance with the Technical Specifications and the Environmental Protection Plan.
4.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION tid
,', Theodore R. Quay, Director Project Directorate IV-1 Division of Reactor Projects III, IV, and V Office of Nuclear Reactor Regulation Attachments:
1.
Pages 9 and 16 of the license 2.
Changes to the Technical Specifications Date of Issuance: August 23, 1991
4 ATTACHMENT TO LICENSE AMENDMENT NO. gp FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following pages of the facility Operating License and 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 License page 9 License page 9 License page 16 License page 16 y
v viii iii ix r
xiii xiii xv xv 3/4 3-80 through 3/4 3-89 3/4 3-80 3/4 7-17 through 3/4 7-31 3/4 7-17 B 3/4 3-5 B 3/4 3-5 B 3/4 7-4 B 3/4 7-4 6-2 6-2 6-8 6-8
(19) IE Information Notice 79-22, Qualification of control S.ystem (Section 7.8.C, SER,_SSER #2)
Prior to startup following the first refueling outage, MP&L shall complete any design changes found necessary as a result of this review.
(20) Standby Service Water System (Section 9.2.1 SER, SSER #2)
(Deleted)
(21) Spent Fuel Pool Ventilation System (Section 9.4.2 SER, SSER #2)
If spent irradiated fuel is placed in the spent fuel pool prior to installation and operability of the safety related backup fuel pool cooling pump room coolers, the plant shall be placed in shutdown condition and remain shut down with the RHR system dedicated to the fuel pool cooling mode.
(22) Remote Shutdown Panel (Section 9.5.4.1, SER, SSER #2)
Prior to startup following the first refueling outage, MP&L shall install electrical isolation switches between the control room and the Division 3 remote shutdown panel.
(23) Fire Protection Program (Section 9.5.9, SER)
Replaced by Paragraph 2.C.(41).
i l
l l
9 Amendment No. 65, 82
(41) Fh e Protection Program l
Entergy Operations, Inc. shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in Revision 5 to the Updated Final Safety Analysis Report, and as approved in the Safety Evaluation dated August 23, 1991, subject to the following provisions:
The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
D.
The facility required exemptions from certain requirements of Appendices A and J to 10 CFR Part 50 and from certain requirements of 10 CFR Part 100.
These include:
(a) exemption from General Design Criterion 17 of Appendix A until startup following the first refueling outage, for (1) the emergency override of the test mode for the Division 3 diesel engine, (2) the second level undervoltage protection for the Division 3 diesel engine, and (3) the generator ground over current trip function for the Division 1 and 2 diesel generators (Section 8.3.1 of SSER #7) and (b) exemption from the requirements of Paragraph III.D.2(b)(ii) of Appendix J for the containment airlock testing following normal door opening when containment integrity is not required (Section 6.2.6 of SSER #7).
These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.
In addition, by exemption dated December 20, 1986, the Commission exempted licensees from 10 CFR 100.11(a)(1), insofar as it incorporates the definition of exclusion area in 10 CFR 100.3(a), until April 30, 1987 regarding demonstration of authority to control all activities within the exclusion area (safety evaluation accompanying Amendment No. 27 to License (NPF-29).
This exemption is authorized by law, and will not present an undue risk to the public health and safety, and is consistent with the common defense and security.
In addition, special circumstances have been found justifying the exemption.
Therefore, these exemptions are hereby granted pursuant to 10 CFR 50.12.
With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the previsions of the Act and the rules and regulations of the Commission.
E.
The licensee shall fully implement and maintain in effect all provision of the Commission-approved physical security, guard training and qualifi-cation, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
" Grand Gulf Nuclear Station Physical Security Plan," with revisions submitted through September 1,1987; " Grand Gulf Nuclear Station Guard Training and Qualification Plan," with revisions submitted through March 10, 1987; anu
" Grand Gulf Nuclear Station Safeguards Contingency Plan," with revisions submitted through August 25, 1987.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
16 Amendment No. 65,82
INDEX JINITINGCONDITIONSFOROPERATIONANDSURVEILLANCEREQUIREMENTS 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 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.....
3/4 3-27 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION AWS Recirculation Pump Trip System Instrumentation.........
3/4 3-37 End-of-Cycle Recirculation Pump 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 R0D BLOCK INSTRUMENTATION...........................
3/4 3-52 3/4.3.7 MONITORING 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 Traversing In-Core Probe System.............................
3/4 3-78 Chlorine Detection System...........
3/4 < 79 Fire Detection Instrumentation (DELETE 0)....................
3/4 3-80
-Loose-Part Detection System.................................
3/4 3-90 Radioactive Liquid Effluent Monitoring Instrumentation......
3/4 3-91 Radioactive Gaseous' Effluent Monitoring Instrumentation.....
3/4 3-96 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION....................
3/4 3-105
-3/4.3.9 TURBINE.0VERSPEED PROTECTION SYSTEM.......................
3/4 3-110 GRAND GULF-UNIT 1 v
Amendment No. 82
.m.
1
_INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops........................................
3/4 4-1 Jet Pumps.............
3/4 4-2 Recirculation Loop F1ow....................................
3/4 4-3 Idle Recirculation Loop Startup............................
3/4 4-4 3/4.4.2 SAFETY VALVES Safety / Relief Valves.....................................
3/4 4-5 Safety / Relief Valves Low-Low Set Function.............
3/4 4-7 3/a 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..................................
3/4 4-8 Operational Leakage........................................
3/4 4-9 3/4.4.4 CHEMISTRY..................................................
3/4 4-13 3/4.4.5 SPECIFIC ACTIVITY.......................................
3/4 4-16 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System........................
3/4 4-19 Reactor Steam 00me.........................................
3/4 4-23 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...........................
3/4 4-24 3/I.4.8 STRUCTURAL INTEGRITY.......................................
3/4 4-25 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown...............................................
3/4 4-26 Cold Shutdown..............................................
3/4 4-27 3/4.5 EMERGENCY CDP.E COOLING SYSTEMS 3/4.5.1 ECCS - 0PERATING.......................
3/4 5-1 3/4.5.2 ECCS - SHUTD0WN............................................
3/4 5-6 3/4.5.3 SUPPRESSION P00L............................
3/4 5-8 GRAND GULF-UNIT 1 vi
INDEX L1MITIN3 CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.6 CON'tAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT t
P rima ry Conta i nment I ntegri ty..............................
3/4 6-1 i
Containment Leakage.e......................................
3/4 6-2 Containment Air Locks......................................
3/4 6-5 t
MSIV Leakage Control System................................
3/4 6-7 Feed ater Leakage Control System...........................
3/4 6-6 Containment Structural Integrity...........................
3/4 6-9 Containment Internal Pressure.............................
3/4 6-10 t
Containment Average Air Temperature........................
3/4 6-11 i
Containment Purge Syster............................
3/4 6-12 3/4.6.2 DRYWELL i
Drywell Integrity..........................................
3/4 6-13 Drywell Bypass Leakage.....................................
3/4 6-14 Drywell Air Locks..........................................
3/4 6-15 s
Orywell Structural Integrity...............................
3/4 6-17 D rywe l l I nt e rna l P re s s u re..................................
3/4 6-18
+
Drywell Average Air Temperature............................
3/4 6-19 7
D rywe l l V e n t a nd P u rg e.....................................
3/4 6-20 3/4.6.3 DEPRESSURIZATION SYSTEMS Suppression Poo1...........................................
3/4 6-21
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Containment Spray.......................................
3/4 6-25 l
Suppression Pool Cooling...............................,,..
3/4 6-26 Suppression Pool Hakeup System.............................
3/4 6-27 l
l 3/4.6.4 CONTAINMENT AND DRYWELL ISOLATION VALVES..................
3/4 6-28 GRAND GULF-UNIT 1 vii t
l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE CONTAINMENT SYSTEMS (Contir.aed) 3/4.6.5 DRYWELL VACUUM RELIEF......................................
3/4 6-46 3/4.6.6 SECONDARY CONTAINMENT Secondary Contai nment Integri ty............................
3/4 6-48 Secondary Containment Automatic Isolation Dam Valves.......................................pers/
3/4 6-49 Standby Gas Treatment System...............................
3/4 6-55 3/4.6.7 ATMOSPHERE CONTROL Containment Hydrogen Recombiner Systems....................
3/4 6-58 Containment and Drywell Hydrogen Ignition System...........
3/4 6-59 Combustible Gas Control Purge System.......................
3/4 6-66 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS Standby Service Water System...............................
3/4 7-1 High Pressure Core Spray Service Water System..............
3/4 7-3 Ultimate Heat Sink.........................................
3/4 7-4 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM...................
3/4 7-5 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM......................
3/4 7-7 3/4.7.4 SNUBBERS...............
3/4 7-9 3/4.7.5 SEALED SOURCE CONTAMINATION..............................
3/4 7-15 3/4.7.6 FIRE SUPPRESSION SYSTEMS (DELETED).........................
3/4 7-17 GRAND GULF-UNIT 1 viii Amendment No. 82 l
i INDEX I
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS r
SECTION PAGE PLANT SYSTEMS (Continued) i 3/4.7.7 FIRE RATED ASSEMBLIES (DELETED)...........................
i 3/4.7.8 AREA TEMPERATURE MONITORING................................
3/4 7-32 e
3/4.7.9 SPENT FUEL STORAGE POOL TEMPERATURE.......................
3/4-7-34 4
3/4.7.10 FLOOO PR0TECTION...........................................
3/4 7-35 i
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 5
A.C. Sources-Operating.....................................
3/4 8-1 A.C. Sources-Shutdown......................................
3/4 8-9 3/4.8.2 D.C. SOURCES D.C. Sources - Operating...................................
3/4 8-10 t
D.C. Sources - Shutdown....................................
3/4 8-14 l
3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Di stri buti on - Operati ng...................................
3/4 8-15 Di stributi on - Shutdown....................................
3/4 8-17 l
3/4.8.4
' ELECTRICAL EOJIPMENT PROTECTIVE DEVICES Primary Containment Penetration Conductor Overcurrent Protective Devices.......................................
3/4 8-19 l
t Motor Operated Valve Thermal Overload Protection...........
3/4 8-46 Reactor Protection System Electric Power Monitoring........
3/.4 8-54' l
3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH..........................................
3/4 9-1 3/4.9.2 INSTRUMENTATION..............................................
3/4 9-3 3/4.9.3 CONTROL R0D POSITION...............
3/4 9-5 1
GRAND GULF-UNIT 1 ix Amendment No. 82 i
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION
(
PAGE REFUELING OPERATIONS (Continued) 3/4.9.4 DECAY TIME............
3/4 9-6 3/4.9.5 COMMUNICATIONS..........
3/4 9-7 3/4.9.6 REFUELING EQUIPMENT.........
3/4 9-8 Refueling Platform..
3/4 9-8 Auxiliary Platform...
3/4 9-9 Fuel Handling Platform...
3/4 9-10 3/4.9.7 CRANE TRAVEL - SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE POOLS...
3/4 9-11 3/4.9.8 WATER LEVEL - REACTOR VESSEL...
3/4 9-12 3/4.9.9 WATER LEVEL - SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE POOLS...
3/4 9-13 3/4.9.10 CONTROL ROD REMOVAL Single Cont ol Rod Removal..
3/4 9-14 Multiple Control Rod Removal.
3/4 9-16 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level..
3/4 9-18 Low Water Level...
3/4 9-19 3/4.9.12 HORIZONTAL FUEL TRANSFER SYSTEM.....
3/4 9-20 3/4.10 SPECIAL TEST EXCEPTIONS l
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3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /DRYWELL INTEGRITY...
3/4 10-1 3/4.10.2 ROD PATTERN CONTROL SYSTEM..
3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.
3/4 10-3 3/4.10.4 RECIRCULATION LOOPS.
3/4 10-4 3/4.10.5 TRAINING STARTUPS.
3/4 10-5 GRAND GULF-UNIT 1 x
INDEX BASES
~
SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...............
B 3/4 3-4 Seismic Monitoring Instrumentation.................
B 3/4 3-4 Meteorological Monitoring Instrumentation..........
B 3/4 3-4 Remote Shetdown System Instrumentation and i
Controls.........................................
B 3/4 3-4 Accident Monitoring Instrumentation................
B 3/4 3-4
(
Source Range. Monitors..............................
B 3/4 3-5 Traversing In-Core Probe System....................
B 3/4 3-5 Chlorine Detection System..........................
B 3/4 3-5 i
Fire Detection Instrumentation (DELETED)...........
B 3/4 3-5 l
Loose-Part Detection System........................
B 3/4 3-6 Radioactive Liquid Effluent Monitoring Instrumentation..................................
B 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instrumentation..................................
B 3/4 3-6 L
3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION............
B 3/4 3-6 3/4.3.9 TURBINE OVERSPEED PROTECTION.......................
B 3/4 3-7 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION............
B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM...............................
B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES...............................
B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE f
Leakage Detection Systems..........................
E 3/4 4-2 Operational Leakage................................
B 3/4 4-2 3/4.4.4 CHEMISTRY..........................................
B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY..................................
B 3/4 4-3 f
3/4.4.6 PRESSURE / TEMPERATURE LIMITS........................
B 3/4 4-4 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.82
INDEX 1
BASES I-SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEM 3/4.5.1/2 ECCS - OPERATING and SHUTD0WN......................
B 3/4 5-1 3/4.5.3 S U P P RE S S I ON P00 L...................................
B 3/4 5-2 i --
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT I
Prima ry Contai nment Integri ty......................
B 3/4 6-1 Containment Leakage.s..............................
B 3/4 6-1 Containment Air Locks..............................
B 3/4 6-1 MSIV Leakage Centrol System........................
B 3/4 6-1 Feedwater Leakage Control System...................
B 3/4 6-2 Containment Structural Integrity...................
B 3/4 6-2 Containment Internal Pressure......................
B 3/4 6-2
}
Containment. Average Air' Temperature................
B 3/4 6-2 j
Containment Purge System.......................
B 3/4 6-2 i
3/4.6.2 DRYWELL
=
u Drywell Integrity..................................
B 3/4 6-3 j
Drywell Bypass Leakage.............................
B 3/4 6-3 O rywe l l A i r Loc ks..................................
B 3/4 6-3 Drywell Structural Integrity.......................
B 3/4 6-4 Drywell Internal Pressure..........................
B 3/4 6-4 1
Drywell Average Air Temperature....................
B 3/4 6-4 1
,[
D rywe l l Ve n t - and Pu rg e.............................
B 3/4 6-4 i:
3/4.6.3 DEPRESSURIZATION SYSTEMS...........................
B 3/4 6-4 3/4.6.4 CONTAINMENT AND DRYWELL ISO LATION VALVES...........
B 3/4 6-7 4
3/4.6.5 DRYWELL VACUUM RELIEF..............................
B 3/4 6-7 l 1
3/4.6.6 SECONDARY CONTAINMENT..............................
B 3/4 6-8 3/4.6.7 ATHOSPHERE CONTR0L.................................
B 3/4 6-9 GRAND GULF-UNIT 1 xiv Amendment No. 21 Effective Da te:
Octobor ?0. 1986
1 j
INDEF BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS..............................
B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM...........
B 3/4 7-1 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 MONITORING.......................,
B 3/4 7-4 3/4.7.9 SPENT FUEL STORAGE POOL 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 DISTRIBUTION SYSTEMS...............................
B 3/4 8-1 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 P0SITION...............................
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 3/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT FUEL AND UPPER CONTAINMENT F U E L STO RAG E P00 L S.............................
B 3/4 9-2 GRAND GULF-UNIT 1 xv Amendment No. 82 1
ti
.r i
INDEX
_ BASES
~;
i i
SECTION PAGE REFUELING OPERATIONS (Continued)
I 3/c 9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL i
i and WATER LEVEL - SPENT FUEL-AND UPPER CONTAINMENT FUEL STORAGE P00LS...............
B 3/4 9-2 3/4.9.10 CONTROL R0D 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.......
B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /ORYWELL INTEGRl'Y..
E 3/4 10-1 l
3/4.10.2 ROD PATTERN CONTROL SYSTEM.................
B 3/4 10-1 3/4.10.3 SHUTDOWN MARC.IN DEMONSTRATIONS.,..........
B 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS........
B 3/4 10-1 3/4.10.5 TRAINING STARTUPS.......
B 3/4 10-1 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.....
B 3/4 11-1 Concentration....................
B 3/4 11-1 00se......................
B 3/4 11-1 Liquid Waste Treatment.........................
B 3/4 11-2 Liquid Holdup Tanks...................
B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS.,
B 3/4 11-3 Dose Rate....
B 3/4 11-3 Dose - Noble Gases......
B 3/4 11-3 Dose - Iodine ';31, Iodine-133, Tritium and Radionuclides in Particulate Form.........
B 3/4 11-4 Gaseous Radwaste Treatment and Ventilation Exhaust System.
B 3/4 11-5 GRAND GULF-UNIT 1 xvi
9 5
0 INSTRUMENTATION l
CHLORINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8 DELETED SURVEILLANCE REQUIREMENTS 4.3.7.8 DELETED 8
O 4
e e
e e
O e
GRAND GULF-UNIT 1 3/4 3-79 Amendment No. 25 l
_M
,_,%MMN O@
O N %EO
INSTRUMENTATION i
FIRE DETECTION INSTRUMENTATION
-(Deleted)
Pages 3/4 3-80 through 3/4 3-89 have been deleted.
i GRAND GULF-UNIT 1 3/4 3-80 Amendment No. 82 (Next page is 3/4 3-90)
l e
e PLANT SYSTEMS 3/4.7.6 FIRE SUPPRESSION SYSTEMS (Deleted) 3/4.7.7FIRERATEDASSEMBLIEjs_
(Del eted)
Pages 3/4 7-17 through 3/4 7-31 have been deleted GRAND GULF-UNIl 1 3/4 7-17 Amendment No. 82 (Next page is 3/4 7-32)
PLANT SYSTEMS 3/4.7.8 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.8 The temperature of each area shown in Table 3.7.8-1 shall be maintained within the limits indicated in Table 3.7.8-1.
APPLICABILITY:
Whenever the equipment in an affected area is required to be OPERABLE.
AC110N:
With one or more areas exceeding the temperature limit (s) shown in Table 3.7.8-1:
For more than eight hours, prepare and submit a Special Report to a.
the Commission pursuant to Specifica+. ion 6.9.2 within the next 30 days providing a record of the amount by which and the cumulative time
/
the temperature in the affected area exceeded its limit and an analysis to demonstrate the continued OPERABILITY of the affected equipment.
b.
By more than 30 F, in addition to the Special Report required above, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area to within its temperature limit or declare the equipment in the affected area inoperable.
SURVEILLANCE REQUIREMENTS 4.7.8 The temperature in each of the areas shown in Table 3.7.8-1 shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i GRAND GULF-UNIT 1 3/4 7-32
l I
INSTRUMENTATION BASES 3/4.3.7.6 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 ciade 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-CODE
- JE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.
The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e., detectors) prior to performing an LPRM calibration function.
Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE.
The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output with data obtained during the previous LPRM calibrations.
3/4.3.7.8 CHLORINE DETECTION SYSTEM DELETED 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION DELETED GRAND GULF-UNIT 1 B 3/4 3-5 Amendment No.82 l
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.f INSTRUMENTATION i
1 BASES' y
3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient i
capability is available to detect loose metallic parts in the primary 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, to provide the input; to the 8 monitoring channels.
The remaining 8 sensors I
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 of Light-Water-Cooled Reactors,"' May 1981.
3/4.3.7.11 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to i
monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The alarm / trip setpoints for these instruments shall be calculated in accordance-e with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR 50.
3/4.3.7.12 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION j,
The radioactive gaseous effluent monitoring instrument'ation is provided to
~
monitor and control, as applicable, gaseous effluents during actual or poten-tial releases.
Those instruments that monitor the activity of gaseous efflu-ents being released to the environment shall have their alarm / trip setpoints calculated in accordance with the methods in the ODCM to ensure that the alarm /
trip will occur prior to exceeding the limits of 10 CFR Part 20.
Other instru-1 ments that monitor offgas processing, (i.e., the Explosive Gas Monitor, Offgas Pre-Treatment Monitor, and Offgas Post-Treatment Monitor) are calibrated accord-ing to plant procedures.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50, 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION The plant systems actuation instrumentation is provided to initiate action to mitigate the consequences of accidents that are beyond the ability of the operator to control.
The LPCI mode of the RHR system is automatically initiated 1
on a high drywell pressure signal and/or a low reactor water level, level 1, signal.
The containment spray system will then actuate automatically following t
high drywell and high containment pressure signals.
Negative barometric pres-sure fluctuations are accounted for in the trip setpoints and allowable values specified 'for drywell and containment pressure-high.
A 10 minute minimum, 13-minute maximum time delay exists between initiation of LPCI and containment 1-spray actuation.
A high reactor water level, level 8, signal will actuate the feedwater system / main turbine trip system.
The suppression 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 receipt of a LOCA signal.
The low low suppression pool water level Trip Setpoint and Allowable Value are relative to the surface floor of the suppression pool (93'0k" above mean sea level).
GRAND GULF-UNIT 1 B 3/4 3-6 Amendment No. 60 t
l PLANT SYSTEMS 4
BASES t
3/4.7.4 SNUBBERS (Continued)
The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers.
For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing.
To provide assurance of snubber functional reliability one of two l
functional testing methods is used with the stated acceptance criteria:
1.
Functionally test 10% of a type of snubber with an additional 5%
tested for each functional testing failure, or 2.
Functionally test a sample size and determine sample acceptance using Figure 4.7.4-1.
Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio Plan" described in " Quality Control and Industrial Statistics" by Acheson J. Duncan and the Revision 2 draft (9/86) of the ANSI /ASME OM4 document (Examination and Performance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers).
Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the completion of their fabrication or at a subseovent date.
Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.
The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and asso-ciated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.).
The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.
These records will provide statistical bases for future consideration of snubber service life.
3/4.7.5 SEALED SOURCE CONTAMINATION The limitation on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.
Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group.
Those sources which are frequently handled are required to be tested more often than those which are not.
Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
GRAND GULF-UNIT 1 B 3/4 7-3 Amendment No. AE, 56
PLANT SYSTEMS BASES 3/4.7.6 FIRE SUPPRESSION SYSTEMS DELETED 3/4.7.7 FIRE RATED ASSEMBLIES DELETED 3/4.7.8 AREA TEMPERATURE MONITORING The area temperature limitaticas ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures.
Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY.
The temperature limits include allowance for instrument error, i
GRAND GULF-UNIT 1 B 3/4 7-4 Amendment No. 82
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I e.
a ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)
A health physics technician
- shall be onsite when fuel is in the c.
reactor.
d.
All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator i
Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
e.
(DELETED) f.
Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; t
e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.
Adequate shift coverage shall be maintained without routine heavy use of overtime.
However, in the event that unforeseen problems require substantial amounts of overtime to be used, the following guidelines shall be followed:
1.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.
2.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven-day period, all excluding shift turnover time.
3.
A break of at least eight hours should be allowed between work periods, including shift turnover time.
t 4.
Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
i Any deviation from the above guidelines shall be authorized by the General Manager, Plant perations or his designee, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.
Controls shall be included in the procedures such that individual overtime shall be
- The number of health physics technicians may be less than the minimum require-l ments for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected. absence provided immediate action is taken to fill the required positions.
GRAND GULF-UNIT 1 6-2 Amendment No. 82 l
= -..
~ -
' ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) f.-
Reports of_ violations of codes, regulations, orders, Technical Specifications, or_ Operating License requireeents having nuclear safety significance or reports.of abnormal degradation of systems designed to contain radioactive material.
g.
Reports of significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
h.
Review of all REPORTABLE EVENTS.
i.
All recognized indications of an unanticipated deficiency in some-aspect of design or operation of safety-related structures, systems, or components, j.
The plant Security Plan and changes thereto.
k.
The Emergency Plan and changes thereto.
1.
Items which may constitute a potential nuclear safety hazard as identified during review of facility operations, Investigations or analyses of special subjects as requested by the m.
Chairman of the Safety Review Committee, Changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, n.
and radwaste systems.
o.
_ Fire Protection Program and changes thereto.
l AUTHORITY 6.5.1.7 The PSRC shall:
Recommend in writing to the General Manager, Plant Operations approval L
a.
or disapproval of items considered under 6.5.1.6.a. c, d, e, j, and k, above.
L b.
Render determinations in writing.o the General Manager, Plant Opera-tions with regard to whether or-not each item considered under-6.5.1.6.a, e and_d, above, constitutes an unreviewed safety question.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the SRC of disagreement c.
between the PSRC and the General Manager, Plant Operations; however, the General Manager, Plant Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS l
- 6. 5.1. 8 The PSRC shall maintain written minutes of each PSRC meeting that, at a minimum, document the results of all PSRC activities performed under the responsibility and authority provisions of these Technical Specifications.
Copies shall be provided to the SRC.
GRAND GULF-UNIT 1 6-8 Amendment No. 82 e
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