Text

Amends 85 & 69 to Licenses NPF-11 & NPF-18,respectively, Changing TS by Removing TS on Radioactive Effluents & Radiological Environ Monitoring & Adding Controls to Include Them in ODCM
	ML20105C825
Person / Time
Site:	LaSalle
Issue date:	09/01/1992
From:	Barrett R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20105C826	List: ML20105C825; ML20105C840;
References
	NUDOCS 9209220478
	Download: ML20105C825 (124)
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UNITED STATES Y

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).i NUCLEAR REGULATORY COMMISSION f

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASALLE COUNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 85 License No. NPF-ll 1.

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

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee), dated May 22, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the 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 in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-ll is hereby amended to read as follows:

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9209220478 920901 PDR ADOCK 03000373 P

PDR L

, (2)

Technical Specifications and Environmental Prgtection Plan The Tecl.nical Specifications contained in Appendix A, as revised through Amendment No. 85, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accorda1:e with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective upon date of issuance, to be implemented within 30 days.

l FORTHENUCLEARRECLl0RYCOMMISSION li l

us' f)>N

Rich r

. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

September 1, 1992 t

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ATTA(HMENT TO LICENSE AMENDMENT N0. 85 FAClllTY OPERATING LICENSE NO, NPf-11 QQCKET NO 50-373 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages, The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

Pages indicated with an asterisk are provided for convenience.

REMOVE INSERT I

i 11 11

111

111 1

1V

1V V

V

VI

VI 2

VII

VII

Vill

VIII

1X

!X X

X XI XI XII XII XIII Xill

XIV

XIV

XV

XV XVI XVI

XVil

XVil XVill XVill

XIX

XIX XX XX

XXI

XXI XXII XXll XXill XX11]

l-4 1-4 1-5 1-5 1-6 1-6 1-7 1-7 1-8 1-8 1-9 l-9 1-10 3/4 3-81 3/4 3-81 3/4 3-82 3/4 3-82 3/4 3-83 3/4 3-83 3/4 3-84 3/4 3-84 3/4 3-85 3/4 3-85 3/4 3-86 3/4 3-86 3/4 3-87 3/4 3-87 3/4 3-88 3/4 3-88 3/4 3-89 3/4 3-89

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3/4 3-93 3/4 3-94 l

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3/4 11-17 3/4 11-18 3/4 11-19 3/4 11-20 3/4 11-21 3/4 11-22 3/4 12-1

-3/4 12-2 3/4 12-3 3/4 12 --

3/4 12-5 3/4 12-6 3/4 12-7 3/4 12-8 3/4 12-9 3/4 12-10 B 3/4 3-5 B 3/4 3-5 B 3/4 3-6 B 3/4 3-6 B 3/4 11-1 B 3/4 11-1 B 3/4 11-2 B 3/4-11 1--

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INDEX DEFINITIONS 1

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

1-1 1.2 AVERAGE PLANAR EXPOSURE..........................................

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

1-1 1.4 CHANNEL CALIBRAT10N...............................................

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

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

1-1 1.7 CORE ALTERAT10N.....................................

1-2 1.8 CORE OPERATING LIMITS REPORT..............

1-2 1.9 CRITICAL POWER RAT 10..............................................

1-2 1.10 DOSE EQUIVALENT l-131..................

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

1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME................

1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.........

1-2 1.14 FRACTION OF LIMITING POWER.0ENSITY................................

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

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

1-3 1.17 GASEOUS RADWASTE TREATMENT SYSTEM.................................

1-3 1.18 I D E NT I F I E D L E A KAG E................................................

1-3 1.19 ISOLATION SYSTEM RESPONSE TIME....................................

1-3 1.20 LIMITING CONTROL ROD PATTERN......................................

1-3 1.21 LINEAR HEAT GENERATION RATE.......................................

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

1-4 1.23 MAXIMUM FRACTION OF LIMITING POWER 0ENSITY........................

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

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

1-4 1.26 0FFSITE DOSE CALCULATION MANUAL...................................

1-4

. LA SALLE - UNIT 1 1

Amendment No. 85

INDEX 3

DEFINITIONS l

SECTION DEFINITIONS (Continued)

PAGE 1.27 OPERABLE - OPERABILITY............................................

1-5 1.28 OPERATIONAL CONDITION -

CONDIT10N.................................

1-5 3.29 PHYSICS TEST 5..........................

1-5 1.30 PRESSURE BOUNDARY LEAKAGE.........................................

1-5 1.31 PRIMARY CONTAINMENT INTEGRITY...

1-5 t

1.32 PROCESS CONTROL PR0 GRAM...........................................

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

1-6 1.34 RATED THERMAL POWER........

1-6 n

1.35 REACTOR PROTECTION SYSTEM RESPONSE TIME...........

1-6 1.36 REPORTABLE EVENT...............

1-6 1.37 R0D DENSITY.....

1-6 1.38 SECONDARY CONTAINMENT INTEGRITY.......

1-7 1.39 SHUTDOWN MARGIN...............

1-7 1.40 SITE BOUNDARY........

1-7 1.41 SOURCE CHECK.....................

1-8 1.42 STAGGERED TEST BA515............................................

1-8

1. 4 3 T H E R MA L P OW E R..................................................

1-8 1.44 TURBINE BYPASS RESPONSE TIME......................................

1 2 1.45 UNIDENT! fled LEAKAGE..............................................

1-8 1.46 VENTILATION EXHAUST-TREATMENT SY5 TEM..............................

1-8 1.47 VENTING................

1-8 LA SALLE - UNIT 1 11 Amendment No. 85

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J INDEx t

SAFETY LIMITS AND L1HITING SAFETY SYSTEM SETTINGS i

i SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pre s sure or Low F10w...........................

2-1 l

THERMAL POWER, High Pressure and High F1ow........................

2-1 Reactor Coolant System Pressure...................................

2-1 Reactor Vessel Water Leve1........................................

2-2

2. 2 LIMITING SAFETY SYSTEM SETT4NGS Reactor Protectior, System Instrumentation Setpoints..............

2-3 BASES 1

2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F1ow...........................

B 2-1 THERMAL POWER, High Pressure aad High F1ow........................

B 2-2 Reactor Coolant System Pressure...................................

B 2-8 Reactor Vessel Water Level.......................................

B 2-8 2._ _2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpo'nts...............

B 2-9 LA SALLE - UNIT 1 III

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY..................................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUT 00WN MARGIN..............................................

3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.........................................

3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability......................................

3/4 1-3 Control Rod Maximum Scram Insertion Times....................

3/4 1-6 Control Rod Average Scram Insertion Times....................

3/4 1-7 Four Control Rod Group Scram Insertion Times.................

3/4 1-8 Control Rod Scram Accumulators.................

3/4 1-9 Control Rod Drive Coupling...................................

3/4 1-11 Control Rod Position Indication..............................

3/4 1-13 Control Rod Drive Housing Support............................

3/4 1-15 3/4.1.4 CONTROL R00 PROGRAM CONTROLS Rod Wortt. Minimizer..........................................

3/4 1-16 Rod Sequence Control 5ystem..................................

3/4 1-17 Rod Block Monitor............................................

3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM...............................

3/4 1-19 3/4.1,6 ECONOMIC GENERATION CONTROL SYSTEM................

3/4 1-23 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................

3/4 2-1 3/4.2.2 APRM SETP0lNTS...............................................

3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0.................................

3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE..................................

3/4 2-5 LA SALLE - UNIT 1 IV Amendment No. 70 l

i INDEX LIMITING CONDITIONS FOR OPERAT!0N AND SURVEILLANCE REQUIREMENTS i

.SECTION PAGE 3/4.? 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 3-23 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation..........

3/4 3-35 End-of-Cycle Recirculation Pump Trip System Instrumentation...........................................

3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION............................................

3/4 3-45 3/4.3.0 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION.................

3/4 3-50 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.........................

3/4 3-56 Seismic Monitoring Instrumentation...........................

3/4 3-60 Meteorological Monitoring instrumentation....................

3/4 3-63 Remote Shutdown Monitoring Instrumentation...................

3/4 3-66 Accident Monitoring Instrumentation..........................

3/4 3-69 Source Range Monitors........................................

3/4 3-72 Traversing In-core Probe System..............................

3/4 3-73 Deleted......................................................

3/4 3-74

-Fire Detection Instrumentation...............................

3/4 3-75 t

Deleted......................................................

3/4 3-81 Explosive Gas Monitoring Instrumentation..............-......

3/4 3-82 Loose-Part Detection System..................................

3/4 3-85

-- 3 /4. 3. 8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.............................................

3/4 3-86 I

LA SALLE - UNIT 1 V

Amendment No. 85

-.- - -._.~-..

.I.NDEX I

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3

t JECTION PAGE

_3/4.4 REACDR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops..................

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

3/4 4-2 Recirculation Loop Flow.

3/4 4-3 Idl e Rec i rc ul ation Loop Startup.............................

3/4 4 4 1 h e rma l Hy d r a ul i c S t a b i l i ty.................................

3/4 4-4a 3/4.4.2 SAFETY / RELIEF VALVES.........

3/4 4-5 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leatage Detection Systems.................

3/4 4-6 Operational Leakage..

3/4 4-7 3/4.4.4 CHEMISTRY.,

3/4 4-10 3/4.4.5 SPECIFIC ACTIVITY.......

3/4 4-13 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Cools System.....................................

3/4 4-16 1

Reactor Steam Dome.........

3/4 4-20 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............

3/4 4-21 3/4.4.8 STRUCTURAL INTEGRITY........................................

3/4 4-22 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown.................................................

3/4 4-23 Cold Shutdown...............................................

3/4/4-24 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS-OPERATING................................

3/4 5-1 3/4.5.2 ECCS-SHUT 00WN................................................

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

3/4 5-8 LA SALLE - UNIT 1 VI Amendment No. 60

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT P r i ma ry C o nt a i nme nt I n t e g ri ty...............................

3/4 6-1 Primary Containment Leakage.................................

3/4 6-2 Prima ry Containment Ai r Loc ks...............................

3/4 6-5 MSIV Leakage Control System.................................

3/4 6-7 Primary Containment Structural Integrity....................

3/4 6-8 Drywell and Suppression Chamber Internal Pressure...........

3/4 6-13 D rywell Average Ai r Tempe rature.............................

3/4 6-14 Drywell and Suppression Chamber Purge System................

3/4 6-15 3/4.6.2 DEPRESSURIZATION SYSTEMS Suppression Chamber.........................................

3/4 6-16 Suppression Pool Spray......................................

3/4 6-20 Suppression Pool Cooling....................................

3/4 6-21 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES........................

3/4 6-22 3/4,6.4 VACUUM RELIEF..............................................,

3/4 6-35 1

3/4.6.5 SECONDARY CONTAINMENT Seconda ry Contai nment Integri ty.............................

3/4 6-37 Secondary Containment Automatic Isolation Dampers...........

3/4 6-38 Standby Gas Treatment System................................

3/4 6-40 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Drywell and Suppression Chamber Hydrogen Recombiner Systems...................................................

3/4 6-43 Drywell and Suppression Chamber Oxygen Concentration........

3/4 6-44 LA SALLE - UNIT 1 VII 4

INDEX LIMITING _ CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Service Water System...................

3/4 7-1 Diesel Generator Cooling Water System........................

3/4 7-2 Ultimate Heat Sink...........................................

3/4 7-3 3/4.7.2 CONTROL ROOM Ahs AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGtNCY FILTRATION SYSlEM..................................

3/4 7-4 3/4.7.3 REACTOR CORE-ISOLATION COOLING SYSTEM..............

3/4 7-7 3/4.7.4 SEALED SOURCE CONTAMINATION..................................

3/4 7-9.

3/4.7.5 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System...............................

3/4 7-11 Deluge and/or Sprinkler Systems..............................

3/4 7-14 C0 Systems.....................

3/4 7-17 p

Fire Hose Stations..........................................

3/4 7-18 3/4.7.6 FIRE RATED ASSEMBLIES..................

3/4 7-22 3/4.7.7 AREA TEMPERATURE MONITORING................................

3/4 7-24 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS I STRUCTURES...................

3/4 7-26 3/4.7.9 SNUBBERS..............

3/4 7-27 3/4.7.10 MAIN TURBINE BYPASS SYSTEM...................................

3/4 7-33 i-l l

LA SALLE - UNIT 1 VIII Amendment No. 18 l-

c INDEX L_lMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION

.P_A_G E 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8,1 A.C. SOURCES A.C. Sources-Operating.......................................

3/4 8-1 A.C. Sources-Shutdown........................................

3/4 8-8 3/4.8.2 ONSITE POWER 015TRIBUTION SYSTEMS A.C. Distribution - Operating................................

3/4 8-10 A.C. Distribution Shutdown.................................

3/4 8-12

0. C. Di s tri t>uti on - Ope rati ng................................

3/4 8-14 D.C. Distribution - Shutdown.................................

3/4 8-19 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. Circuits Insice Primary Containment.....................

3/4 8-21 l

Primary Containment Penetration Conductor Overcurrent Protective Devices.........................................

3/4 8-22 Motor Operated Valves Thermal Overload Protection....

3/4 8-26 Reactor Protection Sistem Electrical Power Monitoring....................................

3/4 8-31 1

LA SALLE - UNIT 1 IX l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE _RERUIREMENTS SECTION PAGE 3/4.9 REFUELING OPIRA110NS 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 R00 POSITION......

3/4 9-5 3/4.9.4 DECAY TIME.....

3/4 9-6 3/4.9.5 COMMUNICATIONS..............................................

3/4 9-7 3/4.9.6 CRANE AND HOIST.........

3/4 9-8 3/4.9.7 CRANE TRAVEL................................................

3/4 9-9 3/4.9.8 WATER LEVEL - REACTOR VESSEL.................................

3/4 9-10 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L.........

3/4 9-11 3/4.9.10 CONTROL R0D REMOVAL Single Control Rod Removal.

3/4 9-12 Multiple Control Rod Removal..........................

3/4 9-14 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.......

3/4 9-16 Low Water Level.....

3/4 9-17 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY................................

3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM..................................

3/4 10-2 3/4.10.3 5HUTDOWN MARGIN DEMONSTRATIONS...............................

3/4 10-3 3/4.10.4 DELETED......................................................

3/4 10-4 3/4.10.5 0XYGEN CONCENTRATION.........................................

3/4 10-5 3/4.10.6 TRAINING STARTUPS............................................

3/4 10-6 3/4.10.7 DELETE0..................................

3/4 10-7 3/4.10.8 SUPPRESSION CHAMBER WATER TEMPERATURE........................

3/4 10-8 LA SALLE - UNIT 1 X

Amendment No. 85

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION

. PAG E.

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tank5..........................................

3/4 11-1 3/4 11.2 GASEOUS EFFLUENTS Explosive Gas Mixture........................................

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

3/4 11-3 LA SALLE - UNIT 1 XI Amendment No. 85

INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY..................................................

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 AN0MALIES....................................

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

B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS............................

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

B 3/4 1-4 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM......................

B 3/4 1 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..............

B 3/4 2-1 3/4.2.2 APRM SETP0lNTS.........................................

B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................

B 3/4 2-2 3/4.2.4 LINEAR HEAT GENERATION RATE.......

B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...............

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

B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION I N ST RUME NT AT I ON.........................................

B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.......

B 3/4 3-3 3/4.3.5 REAC10R CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.........................................

B 3/4 3-4 3/4.3.6 CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION............

B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...................

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

B 3/4 3-4 LA SALLE - UNIT 1 XII Amendment No. 85

INDEX BASES SECTION PAGE INSTRUMENTATION (Continued)

MONITORINGINSTRUMENTATION(Continued) i Meteorological Monitoring Instrumentation...............

B 3/4 3-4 Remote Shutdown Monitoring Instrumentation..............

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

B 3/4 3-5 Source Range Monitors...................................

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

B 3/4 3-5 Deleted...............

B 3/4 3-5 Fire Detection Instrumentation..........................

B 3/4 3-5

- Deleted.................................................

B 3/4 3 Explosive Gas Monitoring Instrumentation...............

B 3/4 3-6 Loose-Part Detection System..............................

B 3/4 3-6 i

3/4. 3. 8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION........................................

B 3/4 3-6 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 4

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

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

B 3/4 4-2 3/4.4.4 CHEMISTRY................................................

B 2/4 4-2 3/4.4.5 SPECIFIC ACTIVITY........................................

B 3/4 4,

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 i

i LA SALLE - UNIT 1 XIII Amendment No. 85

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INDEX BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS-OPERATING and SHUT 00VN.................

B 3/4 5-1 3/4.5.3 SUPPRESSION CHAMBER.....................................

B 3/4 5-2 3,/4. 6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT P ri ma ry Cont a i nme n t I nte g ri ty...........................

B 3/4 6-1 Prima ry Contai nment Leakage.............................

B 3/4 6-1 P rima ry Contai nment Ai r Loc ks...........................

B 3/4 6-1 MSIV Leakage Control System.............................

B 3/4 6-1 Primary Containment Structural Integrity................

B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure.......

B 3/4 6-2 Drywell Average Ai r Temperature.........................

B 3/4 6-2 Drywell and Suppression Chamber Purge System............

B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS................................

B 3/4 6-3 3/4.6.3 PRIMARY CONT AINMENT ISOLATION VALVES....................

B 3/4 6-4 3/4.6.4 VACUUM RELIEF.........................................

B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT...................................

B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTR0L..................

B 3/4 6-5 LA SALLE - UNIT 1 XIV I

INDEX

___SBA ES

)

SECTION PAGE 3

3/4.7 PLANT SYSTEMS 3/4.7.1 CORE STAND 3Y COOLING SYSTEM - EQUIPMENT COOLING WATER S YST EMS........

B 3/4 7-1 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EME RGENCY F I LT RAT ION SYSTEM.........................

B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM...................

B 3/4 7-1 3/4.7.4 SEALED SOURCE CONTAMlNAT10N.............................

B 3/4 7-2

~

3/4.7.5 FIRE SUPPRESSION SYSTEMS................................

B 3/4 7-2 3/4.7.6 FIRE RATED ASSEMBLIES........................

B 3/4 7-3 3/4.7.7 AREA TEMPERATURE MONITORING.............................

B 3/4 7-3 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS 1 ST RUCTURES..............

B 3/4 7 3 3/4.7.9 SNUBBERS.............................................

B 3/4 7-3 3/4.7.10 MAIN TURBlHE BYPASS SYSTEM..............................

B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTENS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE P0!R DISTRIBUTION SYSTEMS...................................

B 3/4 8-1 3/4.8.3 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 R0D POSITION...............

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 CRANE AND H0lST.........................................

B 3/4 9-1 3/4.9.7 CRANE TRAVEL............................................

B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L...............

B 3/4 9-2 3/4.9.10 CONTROL R0D REM 0 VAL.....................................

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

B 3/4 9-2 LA SALLE - UNIT 1 XV Amendment No. 18

.. - -. -.. = _ _ _ _ _ - _ -.

INDEX BASES SECTION PAGE 3/4.10 SP,ECIAL TETT EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.............

B 3/4 10-1 1

3/4.10.2 ROD SEQUENCE CONTROL SYSTEM.............................

B 3/4 10-1 4

3/4.10.3 SHUTDOWN KARGIN DEMONSTRATIONS........................

8 3/4 10-1 3/4.10.4 RECIRCULATION L00PS.....................................

B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION.......................

B 3/4 10-1 3/4.10.6 TRAINING STARTUPS.......................................

B 3/4 10-1 3/4.10.7 CONflRMATORY FLOW INDUCED V!BRATION TEST................

B 3/4 10-1 3/4 10.8 $0PPRES$10N CHAMBER WATER TEMPERATURE...................

B 3/4 10-2 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks.....................................

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

B 3/4 11-1 Main Condenser..........................................

B 3/4 11-1 l

l l

LA SALLE - UNIT 1 XVI Amendment No. 85

INDEX i

DESIGN FFATURES r

SECTION PAGE 5.1 SITE Exclusion Area...............

5-1 Low Population Zone...............................................

5-1 Site Boundary for Gaseous Effluents......................

5-1 Site Bour.dary for Liquid Effluents................................

5-1 5.2 CONTAINMENT Configuration...........................................

51 i

Design Temperature and Pressure...................................

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

5-1

5. 3 REACTOR CO_RE Fuel Assemblies.................................................

5-4 Control Rod Assemblies............................................

5-4 5.4 REACTOR C00LANT SYSTEM Design Pressure and Temperature............................

5-4 Volume..

5-4 5.5 METEOROLOGICAL TOWER L0 CATION.....................................

5 5. 6 FUEL STORAGE Criticality.......................................................

5-5 Drainage..........................................................

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

5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...............................

5-5 LA SALLE - UNIT 1 XVII

_.,._,__.,m.m_.._,..

..o...,

m..

..m m

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 ORGANIZATION, REVIEW, INVESTIGATION, AND AUDIT...................

6-1 6.1.1 High Radiation Areas.......................................

6 15 6.2 PLANT OPERATING PROCEDURES AND PR0 CRAMS...........................

6-16

6. 3 ACTION TO BE TAKEN IN THE EVENT OF A kEPORTABLE EVENT

_IN PLANT OPERAT10N................................................

6-20 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT 15 EXCEEDED........

6-20 6_. 5 PLANT OPERATING RECOR0S...........................................

6-21 6.6 REPORTING REQUIREMENTS.............................................

6-22 6.7 PROCESS CONTROL PR0 GRAM...........................................

6-26 6.8 0FFSITE DOSE CALCULATION MANUAL...................................

6-27

6. 9 MAJOR CHANGES 10 RADI0 ACTIVE WASTE TREATMENT SYSTEMS..............

6-27 LA SALLE - UNIT 1 XVill Amendment No. 85-

b i

INDL LIST OF FIGURES FIGURE PAGE 3.1.5-1 LODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS.........................

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na:B o038 10 M 0) i 2

VOLUME / CONCENTRATION REQUIREMENTS..................

3/4 1-22 3.4.1.5-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATE 0).............................

3/4 4-4c 3.4.6.1 1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE........................

3/4 4-18 3.'.6.1-la MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. RFACTOR VESSEL PRESSURE....................................

3/4 4-1Ba 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..........

3/4 7-32 i

B 3/4 3-1 xEACTOR VESSEL WATER LEVEL.........................

B 3/4 3-7 8 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS...................

B 3/4 6-3a 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS...............................

5-2 5.1.2-1 LOW POPULATION ZONE................................

5-3 6.1-1 DELETED............................................

6-11 6.1-2 DELETED............................................

6-12 6,1-3 MINIMUM SHIFT CREW COMPOSITION.....................

6-13 I

LA SALLE - UNIT 1 XIX Amendment No. 71

INDEX j

LIST OF TABLES TABLE PAGE 1.1 SURVEILLANCEFREQUENCYNOTATION........................

1-9 1.2 OPERATIONAL CONDITIONS.................................

1-10 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS.....

2-4 B2.1.2-1 DELETED................................................

B 2-4 B2.1.2-2 DELETED...............................................

B 2-5 B2.1.2-3 DELETED...............................................

B 2-6 82.1.2-4 DELETED................................................

B.2-7 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION..............

3/4 3-2 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES...............

3/4 3-6 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCEREQUIREMENTS.............................

3/4 3-7 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION....................

3/4 3-11 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS..........

3/4 3-15 i

3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME.........

3/4 3-18 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE I

REQUIREMENTS...........................................

3/4 3-20 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION........................................

3/4 3-24 3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS..............................

3/4 3 -t 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES...........

3/4 3-31 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............

3/4 3-32 3.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION....................................... -3/4 3-36 LA SALLE - UNIT 1 XX Amendment No. 85 e

e s*-

4evm no

~_.,.y

-m..

...m

INDEX LIST OF TABLES (Continued)

TABLE PAGE 4

3.3.4 1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS.........................

3/4 3-37 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-38 3.3.4.2-1 END-Of-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...................................

3/4 3-41 3.3.4.2-2 END*0r CYCLE RECIRCULATION PUMP TRIP SYSTEM SETPOINTS........................................

3/4 3-42 3.3.4.2-3 END-Of-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.....................................

3/4 3-43 4.3.4.2.1-1 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS........................

3/4 3-44 3.3.5-1 EACTOR CORE ISOLATION COOLING SYSTEM ACTUATION IPSTRUMENTATION...................................

3/4 3-46 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.........................

3/4 3-48 4.a.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-49 3.3.6-1 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION......

3/4 3-51 3.3.6-2 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS.........................................

3/4 3-53 4.3.6-2 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................

3/4 3-54 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION..............

3/4 3-57 4.3.7.1-1 RADIATION HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3 59 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION................

3/4 3-61 4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3-62 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION.........

3/4 3-64 LA SALLE - UNIT 1 XXI Amendment No. 18

..l

. E

INDEX ll5T Of TABLES (Continued)

SLE PAGE 4.3.7.3-1 METEOROLOGICAL HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................

3/4 3-65 3.3.7.4-1 REM 01E SHUTDOWN HONITORING INSTRUMENTATION.......

3/4 3-67 4.3.7.4-1 REMOTE SHUTDL.-N MON 110 RING INSTRUMENTATION SURVEllLANCE 'JQUIREMEN15........................

3/4 3-68 3.3.7.5-1 ACCIDENT Po'..' RING INSTRUMENTATION 3/4 3-70 4.3.7.5-1 ACCIDENT MON 110 RING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........

3/4 3-71 3.3.7.9-1 flRE DETECTION INSTRUMEN1ATION.................

3/4 3-76 3.3.7.11-1 EXPLO51VE GAS MONITORING

'.1 INSTRUMENTATION.......

3/4 3-83 4.3.7.11-1 EXPLOSIVE GAS MONITORING a

INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3/4 3-84 3.3.8-1 fEEDWATER/KAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION 3/4 3-87 3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION ]N51RUMENT AT10N SETPolN15............

3/4 3-88 4.3.8.1-1 FEEDWATER/MA!N TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS........

3/4 3-89 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES..

3/4 4-9 3 4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS.........

3/4 4-12 4.4.5-1 PRIMARY COOLANT SPEClfic ACTIVITY SAMPLE AND ANALYSIS PROGRAM....................

3/4 4-15 4.4.6.1.3-1 REACTOR VESSEL HATERIAL SURVEILLANCE PROGRAM l

WITHDRAWAL SCHEDULE..............................

3/4 4-19 4.6.1.5-1 TENDON SURVEILLANCE...............................

3/4 6-11 a

LA SALLE - UNIT 2 XX11 Amendment No. 85 E.

INDEX LIST OF TABLES (Cantinued)

TABLE PAGE 4.6.1.5-2 TENDON LIFT-OFF FORCE.............................

3/4 6-12 3.6.3-1 PRl!.RY CONTAINMENT ISOLATION VALVES..............

3/4 6-24 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS.......................

3/4 6-39 3.7.5.2-1 DELUGE AND SPRINKLdR SYSTEMS......................

3/4 7-16 1

3.7.5.4-1 FIRE HOSE STATIONS...............................

3/4 7-19 3.7.7-1 AREA TEMPERATURE MONITORING......................

3/4 7-25 4.8.1.1.2 1 DIESEL GENERATOR TEST SCHEDULE....................

3/4'8-7b 4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS.................

3/4 8-18 3.8.3.2-1 PRIMARY CONTAINMENT PENE'~il10N CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.................

3/4 8.?4 3.8.3.3-1 MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION.....................................

3/4 8-27 l

B3/4.4.6-1 REACTOR VESSEL TOUGHNESS..........................

B 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..............

5-6 f

LA SALLE-- UNIT 1 XX111 Amendment No. 85

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,--...--e-y

DEFINITIONS f

LINEAR HEAT GENERA 110N RATE 1.21 LINEAR HEAT GENERA 110N RATE (LHGR) 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 witf the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit, from sensor through and including the actuated device to verify OPERABILITY.

THE LOGIC SYSTEM FUNCTIONAL TEST may be parformed by any series of seqv ntial, overlapping or total system steps such that the entire logic system is tested.

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

MEMBER (S) 0F THE PUBLIC 1.24 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with ;he plant.

MINIMUM CRITICAL POWER RATIO 1.25 The MiklMUV RITICAL POWER RATIO (MCPR) shall be the smallest CPR which I

exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.26 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

The ODCH shall also contain (1) the Radioactive Effluent _ Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2) descrip-ions of the information that should be included in the Annual Radiological Environmental Operating and Semi-Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

1 l

i i

LA SALLE UNIT 1 1-4 Amendment No. 85 t

-. - -. -.-.~.. - -

DEFINITIONS i

OPERABLE - OPERABILITY-1.27 A system, subsystem, train, component or device shall be OPERABLE or have OPERABIL11Y..,en it-is capable of performing its specified function (s),

and when all necessary attendant instre9entation, controls, normal and an emergency electrical power source, cooling or seal water,

.brication 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 2./3 An UrERATIONAL CONDITION, i.e.

CONDITION, shall be any inclusive l

combination of mode switch position and average reactor >solant temperature as specified in Table 1.2.

PHYSICS TESTS 1.29 PHYSICS TESTS shall De those tests performed to measure the fundamental l

nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions 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 l

in i reactor coolant system component body, pipe wall or vessel wall, PRIMARY CONTAINMENT INTEGRITY 1.31 PRIMARY CONTAINMENT INTEGRITY shall exist when:

l a.

All primary containmant penetrations required to be closed durine ar-ident conditions are either:

1.

La.'M.e of being closed by an OPERABLE primary containment automatic isolation system, or 2.

Closed ';y et least one manual valve, blind flange, or-deactivated automatic valve serured in its closed position, excetpt as pr sided in Table 3.6.3-1 of Specification, 3.6.3.

l b.

All primary containment equ ' ment hatches are closed and sealed.

~

c.

Each primarv containment air lock is 0FERABLE pursuant to Spe ification 3.6.1.3.

l d.

The primary containment leakage _ rates are within the limits of Spec,fication 3.6.1.2.

LA SALLE UNIT 1 1-5 Amendment No. 85 l

4

--,.-.....-,--,----.-m

DEFINITIONS e.

The suppression chamber is OPERABLE pursuant to Specification 3.6.2.1.

f.

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

PROCESS CONTROL PROGRAM 1.32 The 40 CESS CONTROL PROGRAM (PCP) shall contain the current formuias, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes-based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 2C, 61, and 71, State regulations, burial ground recuirements, and other requirements governing the disposal of solid racioactive waste.

MGE-PURGING 1.33 PURGE or PURGING shall be the controlled process of discharging air or l

gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replace-ment air or gas is required to purify the confinement.

V"

.<ERMAL POWER 4 N 50 THEr'3L POWER shall be a total reactor core heat transfer rate to 1

reacto coolant of 3323 MWT.

dV "R 'ROTECTION SYSTEM RESPONSE TIME 1.35 TACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from l

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

R0D DENSITY 1.37 ROD DENSITY shall be the number of control rod notches inserted as a l

fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 100% R0D DENSITY.

l l

LA SALLE UNIT 1 1-6 Amendment No. 85

l DEFINITIONS SECONDARY CONTAINMENT INTEGRITY 1.38 SECONDARY CONTAINMENT INTEGRITY shall exist when; a.

A'il secondary containment penetrations required to be closed during accident conditions are either; 1.

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

Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed position, except a provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b.

All secondary containment hatches and blowout panels are closed and sealed.

c.

The standby gas treatment system is OPERABLE pursuant to Specification 3.6.5.3.

d.

At least one door in each access to the secondary containmeat is closed.

e.

The sealing mechanism associated with each secondary containment r7netration, e.g., wilds, bellows or 0-rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or equal to the value equired by Specification 4.6.5.1.a.

SHUTOOWN MARGIN 1.39 SHUTDOWN MARGIN shall be the amount of reactivity by.which the reactor is subtritical or would be subcritical 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 B0UNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

LA SALLE UNIT 1 1-7 Amendment No. 85 1

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~ _-

- - _ _ ~ -

DEFINITIONS l

SOURCE CHECK 1,41 A SOURCE CHECK shall be the qualitative assessment of channel response l

when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.42 A STAGGERED TEST BASIS shall consist of:

l a.

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

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

THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPAS; SYSTEM RESPONSE TIME 1.44 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when l

the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

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

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

VENTILATION EXHAUST TREATMENT SYSTEM 1.46 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l'

in nalled to reduce gaseous radioiodine or radioactive material in particu-late form in effluents by passi_ng ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmosphtrir cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT

-SYSTEM components.

VENTING 1.47 VENTING shall be the controlled process cf discharging air or gas from a l

confinement to maintain temperature, pressure, humidity, concentration ~or other operating condition, in such a manner that replacement air or gas is l

not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

LA SALLE UNIT 1 1-8 Amendment No. 85

)

TABLE 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY-5 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once por 184 days.

A At least once par 366 days.

R At least once per 18 months (550 days).

S/U Prior to each reactor startup.

P Prior to each radioactive release.

N.A.

Not applicable.

LA SALLE UNIT 1 1-9 Amendment No. 85

TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE 1.

POWER OPERATION Run Any temperature 2.

STARTUP Startup/ Hot Standby Any temperature 3.

HOT SHUTDOWN Shutdown # ***

> 200 F 4.

COLD SHUTDOWN Shutdown # ## ***

5 200 F 5.

REFUELING

Shutdown or Refuel ** #

5 140 F

ine reactor mode switch may be placed in the Run or Startua/ Hot Standby position to test the switch interlock functions provided t1at the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member cf the unit technical staff.

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being remo'..d from the reactor pressure vessel per Specification 3.9.10.1.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

l

- See Special Test Exception 3.10.3 h

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being moved provided that the one-rod-out interlock is OPERABLE.

LA SALLE UNIT 1 1-10 Amendment No. 85

L_

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_LASALLE UNIT-1 3/4 3-81 Amendment No. 85 i

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INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.11 The explosive gas raonitoring instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their Alarm / Trip setpoints set to ensure that the limits of specification 3.11.2.6 are not exceeded.

APPLICABILITY:

During operation of the main condenser air ejector.

ACTION:

a.

With an explosive gas monitoring instrumentation channel Alarm / Trip setpoint less conservative than required by the above specification, declare the channel inoperable, and take the ACTION shown in Table 3.3.7.11-1.

b.

With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.11-1.

Restore the inoperable instrumentation channels to an OPERABLE status wihtin 30 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.11 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies shown in Table 4.3.7.11-1.

LASALLE UNIT-1 3/4 3-82 Amendment No. 85

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INSTRUMENTATION

-TABLE 3.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1.

MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (for systems designed to withstand the effects of a hydrogen explosion) a.

Hydrogen Monitor 1/ train 110 TABLE NOTATION ACTION 110 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation 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 .

If the recombiner(s) temperature remains constant and THERMAL POWER has not changed, the grab sample collection frequency may be changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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l LASALLE UNIT-1 3/4 3-83 Amendment No. 85

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INSTRUMENTATION TABLE 4.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEIL-INSTRUMENT CHECK TEST CALIBRATION

LANCE REQUIRED 1.

MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM a.

Hydrogen Monitor D

M Q

TABLE NOTATION The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1.

One volume percent hydrogen, balance nitrogen, and 2.

Four volume percent hydrogen, balance nitrogen.

Duringoperationofthemaincondenserairejector.

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LA SALLE - UNIT 1 3/4 3-84 Amendment-No. 85

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INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.12 The loose part detection system shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With one or more loose-part detection system channels inoperable for a.

more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.c within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.12 Each channel of the loose part detection system shall be demonstrated OPERABLE by performance of:

a.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and c.

CHANNEL CALIBRATION at least once per 18 months.

LA SALLE - UNIT 1 3/4 3-85 Amendment No. 85

.i I

i INSTRUMENTATION 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.8 The feedwater/ main turbine trip system 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:

OPERATIONAL CONDITION 1.

ACTION:

a.

With a feedwater/ main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowatle Values column of Table 3.3.8-2, declare the channel inoperable and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable, b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel 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 .

c.

With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels per Trip System requirement, 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 .

SURVEILLANCE REQUIREMENTS 4.3.8.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations 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.*

"The specified 18 month interval may be waived for Cycle 1 provided the surveillance.is performed during Refuel 1.

LA SALLE - UNIT I 3/4 3-86 Amendment No. 85

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3/4.11 RADIOACTIVE EFFLUENTS 344.11.1 LIQUID EFFLUENTS LIQUID HOLCOP TANKS LIMITINC CONDITION FOR OPERATION 3.11.1.1 The quantity of radioactive material contained in any outside temporary canks shall be limited to less than or equal to the limits calculated in the ODCH.

APPLIi'JILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all 6dditions of radioactive material to the tank 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, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.1 The quantity of radioactive material contained in each of the above l

listed tanks 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 radioactive materials are being added to the tank.

LA SALLE - UNIT 1 3/4 11-1 Amendment No. 85

RADIOACTIVE EFFLUENTS 3/4 11.2 GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.1 The concentration of hydrogen in the main condenser offgas treatment

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system shall be limited to less than or equal to 4% by volume.

APPLICABILITY:

Whenever the main condenser air ejector system is in operation.

ACTION:

a.

With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b.

The provis'ons of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.1 The concentration of hydrogen in the main condenser offgas treatment l

system shall be determined to be within the above limits as required by Table 3.3.7.11-1 of Specification 3.3.7.11.

LA SALLE - UNIT 1 3/4 11-2 Amendment No. 85

=.

1 RADIDACTIVE EFFLUENTS MAIN CONDENSER LIMITING CCNDITION FOR OPERATION 3.13.2.2 The release rate of the sum of the activities from the noble gases measured prior to the holdup line shali be limited to less than or equal to 3.4 x 105 microcuries/second.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

Wich the release rate of the sum of the activities of the noble gases prior to the holdup line exceeding 3.4 x 105 microcuries/second restore the 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 STARTUP with the main steam isolation valves closed within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS d

4.11.2.7.1 The radioactivity rate of noble gases prior to the holdup line shall be continuously monitored in accordance with Specification 3.3.7.11.

4.11.2.7.2 The release rate of the sum of the activities from r,oble gases prior to the holdup lino shall be determined to be within the limits of Specification 3.11.2.7 at tne following frequencies by performing an isotopic analysis of a representative sample of gases taken prior to the holdup line.

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 off gas pre-treatment Noble Gas Activity 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 primary coolant.

LA SALLE - UNIT 1 3/4 11-3 Amendment No. 85

i INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.5 ACCIDEN' MONITORING INSTRUMENTATION The OPERABIL' " of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.

This capability is con-sistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

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 made 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.

3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe (TIP) system with the specified mininum complement of equipment ensures that the measurements obtained from use cf this equipment accurately represent the spatial neutron flux distribution of the reactor core.

The specification allows use of substituted TIP data froto symmetric channels if the control rod pattern is symmetric since the TIP data is adjusted by the plant computer to remove machine dependent and power level dependent bias.

The source of data for the substitution may also be a 3-dimensional BWR core simulator calculated data set which is norma zed to available real data.

Since uncertainty could be introduced by the simulatin and normalization process, an evaluation of the specific control rod par. rn and core operating state must be performed to ensure that adequate margin to core operating limits is maintained.

3/4.3.7.8 DELETED 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.

This capa-bility is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to safety-related eouipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, increasing-the frequency of fire watch patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

LA SALLE - UNIT 1 B 3/4 3-5 inndment No. 85

1 i

INSTRUMENTATION BASES 3/4.3.7.10 DELETED 3/4.3.7.11 EXPLOSIVE GAS MONITORING INSTRUMENTATION This instrumentation provides for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

3/4.3.7.12 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 primary system and avoid or mitigate damage to primary system components.

The allowable out-of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors."

3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMGJATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate the feedwater system / main turbine trip syster in the event of reactor vessel water level equal to or greater than the level 8 setpoint associated with a feedwater controller failure, to prevent overfilling the reactor vessel which may result in high pressure liquid discharge through the safety / relief valve discharge lines.

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LA SALLE - UNIT 1 B 3/4 3-6 Amendment No. 85

3/4.11 RADI0 ACTIVE EFFLUENTS l

BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks'provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting concentrations would be less then 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.

3/4.11.2 GASE0US EFFLUENTS 3/4.11.2.1 EXPLO5IVE GAS MIXTURE The specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrcgen and oxygen.

Maintcining the concentration of hydrogen and oxygen below their flammability li:aits provides assurance that the releases of radioattive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.2 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area bcundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently 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.

LA SALLE - UNIT 1 B 3/4 11-1 Amendment No. 85

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5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.

SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1.1-1.

SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 The site boundary for liquid effluents shall be as shown in Figure 5.1.1-1.

5. 2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel lined post-tensioned concrete structure consisting of a drywell and suppression chamber.

The drywell is a steel-lined post-stressed concrete vessel in the shape of a truncated cone closed by a steel dome.

The drywell is above a cylindrical steel-lined post-stressed concrete suppression chamber and is attached to the suppression chamber through a series of downcomer vents.

The drywell has a minimum free air volume of 229,538 cubic feet.

The suppression chamber has an air region of 164,800 to I

168,100 cubic feet and a water region of 128,800 to 131,900 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a.

Maximum internal pressure 45 psig, b.

Maximum internal temperature:

drywell 340 F.

suppression chamber 275 F.

c.

Maximum external pressure 5 psig.

d.

Maximum floor differential pressure:

25 psid, downward.

5 psid, upward.

SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the Reactor Building, the equipment access structure and a portion of the main steam tunnel and has a minimum free volume of 2,875,000 cubic feet.

LA SALLE - UNIT 1 5-1 Amendment No.18

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u-LOW POPULATION %ONE Figure 5.1.2-1 LA SALLE - UNIT 1 5-3 Amendment No. 85

ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) f.

The following programs shall be established, implemented, and maintained:

1.

Primary Coolant Sources Outside Primary Containment A program to reduce leakage from those portions of systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include LPCS, HPCS, RHR/LPCI, RCIC, hydrogen recombiner, process sampling, containment monitoring, and standby gas treatment systems.

The program shall include the following:

a.

Preventive maintenance and periodic visual inspection require-ments, and b.

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

2.

Jn-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:

P a.

Training of personnel, b,

Procedures for monitoring, and c.

Provisions for maintenance of sampling and analysis equipment.

3, 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 program shall include the following:

a.

Training of personnel, b.

Procedures for sampling and analysis, c.

Provisions for maintenance of sampling and analysis equipment.

4.

Radioactive Effluent Controls Program A program shall be provided conforming 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 (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include-remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

LA SALLE UNIT 1 6-18 Amendment No. 85

8 ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGS.S (Continued) a.

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation includir.g surveillance tests and set-point determination in accordance with the methodology in the

ODCM, b.

Limitations on the concentrations of radioactive raterial re-leased in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2, c.

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and w'th the methodology and parameters in the ODCM, d.

Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conform-ing to Appendix I to 10 CFR Part 50, e.

Determination of cumulative and projected dose contri' :tions from radioactive effluents for the turrent calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, f.

Limitations on the operability 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 projected doses in a 31-day period would exceed 2 per-cent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, q.

Limitations or, the dose rate resulting i'om radioactive material

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released in gaseous effluents to areas beyond the SITE B0UNDARY conforming to the doses associated with 10 CFR Part 20, Appen-dix B, Table II, Column 1, h.

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

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from lodine-131, Iodine-133, tritium. and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BUUND-ARY conforming to Appendix I to 10 CFR Part 50, j.

Limitations on venting and purging of the containment through the Primary Containment Vent and Purge System or Standby Gas Treatment System to maintain releases as low as reasonably achievable, LA SALLE UNIT 1 6-19 Amendment No. 85

.o ADMINISTRATIVE CONTROLS

[LANT OPERATING PROCEDURES AND PROGRAMS (Continued) k.

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

5.

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

The program shall provide (1) represen-tative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitor-ing program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM, (2) conform to the guid-ance of Appendix I to 10 CFR Part 50, and (3) include the following:

Monitoring, sampling, analysis, and reporting of radiation and a.

radionuclides in the environment in accordance with the method-ology and parameters in the ODCM, b.

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE B0UNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and Participation in a Interlaboratory Comparison Program to ensure c.

that independent checks on the precision and accuracy of the measurements t f radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN PLANT OPERATION The following actions shall be taken for REPORTABLE EVENTS:

The Commission shall be notified and a Licensee Event Report a.

submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed pursuant to Specifi-cation 6.1.G.2.c(1).

6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pur-suant to Specification 2.1.1, 2.1.2 and 2.1.3, and critical reactor operation shall not be resumed until authorized by the NRC.

The conditions of shutdown shall be promptly reported to the Vice President BWR Operations or his designated alternate.

The incident shall be reviewed pursuant to Specifications 6.1.G.I.a and 6.1.G.2.a and a separate Licensee Event Report for each occurrence shall be prepared in accordance with Section 50.73 to 10 CFR Part 50.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.

The Vice President BWR Operations and the Manager of Offsite Review and Investigative Function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

LA SALLE UNIT 1 6-20 Amendment No, 85

_=

.f ADMINISTRATIVE CONTROLS 6.5 PLANT OPERATING RECORDS A.

Records and/or logs relative to the following items shall be kept in a manner convenient for review and shall be retained for at least 5 years:

1.

Records of normal plant operation, including power levels and periods of operation at each power level; 2.

Records of principal maintenance and activities, including inspection and repair, regarding principal items of equipment pertaining to nuclear safety; 3.

Records and reports of reportable events; 4.

Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (see Section 4 of these specifications) are being met, All equipment failing to meet surveil-

'ince requirements and the corrective action taken shall be recorded; 5.

Records of changes to operating procedures; 6.

Shift engineers' logs; and 7.

Byproduct material inventory records and source leak test results.

B.

Records and/or logs relative to the following items shall be recorded in a manner convenient for review and shall be retained for the life of the plant:

1.

Substitution or replacement of principal items of equipment pertain-ing to nuclear safety; 2.

Changes made to the plant as it is described in the SAR; 3.

Records of new and spent fuel inventory and assembly histories; 4.

Updated, corrected, and as-built drawings of the plant; 5.

Records of plant radiation and contaminecion surveys; 6.

Records of offsite environmental monitoring surveys; 7.

Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant, in accordance with 10 CFR Part 20; 8.

Records of radioactivity in liquid and gaseous wastes released to the environment; LA SALLE UNIT 1 6-21 Amendment No. 85 i

N

4 ADMINISTRATIVE CONTROLS l

PLANT OPERATING RECORDS (Continued) 9.

Records of trar. eent or operational cycling for those components that have beec designed to opa" ate safety for a limited number of transient or operational cyclo (identified in Table 5.7.1-3);

10.

Records of individual staff members indicating qualifications.

l experience, training, and retraining; 11.

Inservice inspettions of the reactor coolant system; 12.

Minutes of meetings and results of reviews and audits performed by the offsite and onsite review and audit functions; 13.

Records of reactor tests and experiments; 14.

Records of Quality Assurance activities required by the QA Manual, except for those items specified in Section 6.5.A; 15.

Records of reviews performed for changes made to procedures on equip-ment or reviews of tests and experiments pursuant to 10 CFR 50.59; l

16.

Records of the service lives of all hydraulic and mechanical snubbers 1

required by specification 3.7.9 including the r% e at which the ser-vice life commences and associated installation and maintenance records; l

17.

Records of aralyses required by the radiological environmental monitoring program; and I

18.

Records of reviews performed fo? changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

.6 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the director of the appropriate Regiora1 Office c' Inspection and Enforce-ment unless otherwise noted.

A.

Routine Reports 1.

Startup Report A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amend-ment to the license involving a plannee increase in power level, (3) installation of fuel that has a different design or has been manufac-tured by a different fuel supplier, and (4) modifications that may have significantly a'tered the nuclear, thermal, or hydraulic perform-ance of the plant.

lne report shall in general include a description LA SALLE UNIT 1 6-22 Amendment No. 85

ADMINI5TRAllVE CONTROL 5 6.6 REPORTING REQUIRIMENTS (Continued) of the measured values of the operating conditions or characteristics cbtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfact(ry operation shall also be described.

Any additional specific details required in license con-ditions based on other commitments shall be included in this report.

Startup reports shall be. submitted within (1) 90 days following com-pletion of the startup test program, (2) 30 days following resumption er commencement of commercial power operation, or (3) 9 months follow-ing initial critical'ty, whichever is earliest, if the startup report does not ct er all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

2.

Annual Report

. tabulation shall be submitted on an annual basis prior to March 1 of each year of the number of station, utility, and other personnel (inc11 ding contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions (Note:

this t3bulation supplements the requirements of Section 20.407 of 10 CFR 20), e.g.,

reactor operations and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste p ocessing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling less than 20% of the individual totsal dose need not be accounted for.

In the aggregate, at least 80't of the total whole body dose received f rom external sources shall be assigned to specific major work functions.

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5 shull be included in the Annual Report along with the following information:

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

Each result should include date and time of sampling and the radioiodine concentrations; (3)

Clean up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 con-centration and one other radioiodine isotope concentration in micro-curies p?r gram as a function of time for the durat -n of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiJdine limit.

LA SALLE UNIT 1 6-23 Amendment No. 85

ADMINISTRATIVE CONTROL 5 s-3.

Annual Radiological Environmental Operating Report

The Annual Radiological Ervironmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the,'eporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.P.3, and IV.C of Appendix 1 to 10 CFR Part 50.

4.

Semiannual Radioactive Effluent Release Report **

The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be >ubmitted within 60 days after January 1 and July 1 of each year The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be (1) consistent with the objectives outlined in he ODCM rnd PCP and (2) in conf ormance with 10 Cf R 50.36a and dection IV.B.1 of Appendix 1 to 10 CFR Part 50.

5.

Monthly Operating Repor:

Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Nuclear Reactor Regulation, Mail Station P1-137 US Nuclear Regulatory Commission, Washington, DC 20555, with a copy of the appropriate Regional Office, to arrive no later than the 15th of each month following the caleadar month covered by the report.

Any changes to the OffSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative function.

6.

Core Operating Limits Report

{

a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

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

The submittal should combine those sections that are common to all units at the station; however, for uhits with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

LA SALLE UNIT 1 6-24 Amendment No.

85

.t ADMINISTRATIVE CONTROLS C.

Unique Reporting Requirements 1.

Special Reports shall be submitted to the Director of the Office of Inspection and Enforcement (Region Ill) within the time period specified for each report.

6.7 PROCESS CONTROL PROGRAM (PCP)*

6.7.1 The PCP shall be approved by the Commission prior to implementativf.

6.7.2 Licensee initiated changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.b.18.

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 con-formance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.

Shall ricome (fiettive upon review and acceptance by the Onsite k.' view and Inv-tigative Function.

The Process Control Program (PCP) is common to La Salle Unit I and La Salle Unit 2.

LA SALLE UNIl 1 6-26 Amendment No.85

(

{

ADMINISTRATIVE CONTROLS 6.8 OffSITE DOSE CALCULATION MANUAL (ODCM)*

6.8.3 The ODCM shall be approved by the Commission prior to implementation.

6.8.2 Licensee initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.18.

This documentation shall contain:

1)

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

A determinatie that the change will maintain the level of radio-active 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 accurac dose, or setpoint calculations. y or reliability of effluent, b.

Shall become effective after review and acceptance by the On-Site Re-view and Investigative Function and the approval of the Plant Manager on the date specified by the On-Site Review and Investigative function.

Shall be submitted to the Commission in the form of a complete, leg-c.

ible copy of the entire ODCM as a part of or concurrent with the Semiannual Radios.ctive Effluent Release Report for the period of the report in which any change to the ODCM was made effective.

Each I

change shall be identified 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.

6.9 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATHENT SYSTEMS 6.9.1 Licensee initiated major changes to the radioactive waste treatment systems (liquid, gaseous and solid):

Shall be reported to the Commission in the Monthly Operating Report a.

1 for the period in which the evaluation was reviewed by the Onsite Review and Investigative Function.

The discussion of each change shall contain:

1.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; 2.

Sufficient detailed information to totally support the reason for the change without benefit or additional or supplemental information; 3.

A detailec description of the equipment, components and processes involved and the interfaces with other plant systems; "The OFFSITE DOSE CALCULATION MANUAL (0DCM) is common to La Salle Unit I and La Salle Unit 2.

LA SALLE UNIT 1 6-27 Amendment No.

85 i

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ADMINISTRAllVE CONTROLS MAJOR CHANGES TO RAD 10AC11VE WASTE TREATMENT SYSTEMS (Continued) 4.

An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseoLs effluents and/or quantf'y of solid waste that differ from those previously predicud in the license application and amendments thereto; 5.

An evaluation of the change which shows the expected maximum s

exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; 6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period to when the changes are to be made; 7.

An estimate of the epante operating personnel as a result of the changs Of 8.

Documentation of the fact 1. hat the change was reviewed and found acceptable by the On.ite Review and Investigative function, b.

Shall become effective upon review and acceptance by the Onsite Review and Investigative function.

LA SALLE UNIT 1 6-28 Amendment No. 85

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3

'M UNITED STATES i

NUCLEAR REGULATORY COMMISSION

d s.... +,/

W AtiHINGioN. D C X6M 1

LOMbiO!LWIAWLED1 SON COMPANt E0H ET NO. 50-374 1&ALil_C.0JNTY STAT 10N. UNIT 2 MilfDMTNT 10 FA(JLITY OPERATIfl0_Ll(ItiSI Amendment No. 69 License No. NPF-18 1.

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

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee), dated May 22, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regu1ations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

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

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

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

2.

Accordingly, the licen:e is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2,0.(2) of the facility Operating License No, NPF-18 is hereby amended to read as follows:

i oa

. (2)

Itchnical_ Specifications _ and Environment ti Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 69, and the Environmental Protection Plan contained in Appendix 0, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective upon date of issuance, to be implemented within 30 days.

FOR 1HE NUCLEAR REC L ORY COMMISSION OM$

Richard'J. Barrett Director Project Directorate 111-2 Division of Reactor Projects - !!!/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the lechnical Specifications Date of Issuance: September 1, 1992 I

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ATTACHMENT T0 llCENSE AMENAMENT N0. 69_

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[htl(ITY OPERATING llCENSE NO. NPF-lf DOCKE1 NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

Pages indicated with an asterisk are provided for convenience.

J REMOVE 1H11El I

1 l

11

111

111 i

1V

1V V

V VI V1

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Vil!

1X

1X X

X XI XI Xil XII X111 Xill XIV

XIV XV XV 4

XVI XVI XVil

XVil XVill XVill

XIX

XIX XX XX XXI

XXI-XXll XX11 XXill XXill 1-4 1-4 i

1-5 1-5 1-6 1-6 1-7 1-7

l-8

l-8 3/4 3-81 3/4 3-81 3/4 3-82 3/4 3-82 3/4 3-83 3/4 3-83 3/4 3-84 3/4 3-84 3/4 3-85 3/4 3-85 3/4.3-86 3/4 3-86 3/4 3-87 3/4 3-87 3/4 3-88 3/4 3-88 3/4 3-89 3/4 3-89 3/4 3-90 2

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3/4 11-12 3/4 11-13 3/4 11-14 3/4 11-15 3/4 11-16 3/4 11-17 3/4 11-18 3/4 11-19 3/4 11-20 3/4 11-21 3/4 11-22 3/4 12-1 3/4 12-2 3/4 12-3 3/4 12-4 3/4 12-5 3/4 12-6 3/4 12-7 3/4 12-8 3/4 12-9 3/4 12-10 B 3/4 3-5 8 3/4 3-5 B 3/4 3-6 8 3/4 3-6 o

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5-1 5-2 5-2 5-3 5-3 6-18 6-18 6-19 6-19 6-20 6-20

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INDEX l

DEFINITIONS l

SECTION.

1.0 OEFINITIONS PAGE 1.1 ACTION..

1-1 1.2 AVERAGE PLANAR EXPOSURE..........................................

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

1-1 1.4 CHANNEL CAllBRATION.

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

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

1-1 1.7 CORE ALTERAT]ON.

1-2

1. 8 CORE OPERATING LIMITS REPORT.,....................

1-2 i

1.9 CRITICAL POWER RATIO......

1-2 1.10 DOSE EQUIVALENT I-131.....

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

1-2 1,12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............

1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.........

1-2 1.14 FRACTION OF LIMITING POWER DENSITY....

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

1 1.16 FREQUENCY NOTATION...........

1-3 1.17 GASE0US RADWASTE TREATMENT SY5 TEM..............................

1-3 1.18 IDENTIFIED LEAKAGE...............................................

1-3 P

1.19 ISOLATION SYSTEM RESPONSE TIME....................................

1-3 1.20 LIMITING CONTROL R0D PATTERN.....................

1-3 1.21 LINEAR HEAT GENERATION RATE.......................................

_1-4 1.22 LOGIC SYSTEM FUNCTIONAL TE5T......................................

1-4 1.23 MAXIMUM FRACTION OF LIMITING POWER DEN 51TY,.......................

1-4 1.24 MEMBER (s) 0F THE PUBLIC.................................

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

1-4 3

1. 26 0FFSITE 005E CALCULATION MANUAL..................................

1-4 LA SALLE - UNIT 2 1

Amendment No. 69

..u--

u u

INDEX DEFINITIONS SECTION DEFINITIONS (Continued)

.P A.G.E.

1.27 OPERABLE - OPERABILITY...........................................

1-4 1.28 OPERATIONAL CONDITION - CONDITION................

1-5 1.29 PHYSICS TE515......

1-5 1.30 PRESSURE BOUNDARY LEAKAGE............

1-5 1.31 PRIMARY CONTAINMENT INTEGRITY.....................................

1-5 1.32 PROCESS CONTROL PROGRAM....

1-5 1.33 PURGE - PURGING.........

...................................... 5 1.34 RATED THERMAL POWER........

1-6 1 35 REACTOR PROTECTION SYSTEM RESPONSE TIME.....

1-6 1.36 REPORTABLE EVENT............

1-6 1.37 ROD DEN 51TY.............

1-6 4........

1.38 SECONDARY CONTAINMENT INTEGRITY................

1-6 1.39 SHUTDOWN MARGIN....

1-6 1.40 51TE BOUNDARY.......

1-7 1.41 SOURCE CHECK.....................................................

1-7 1.42 STAGGERED TEST BA515..............................................

1-7 1.43 THERMAL POWER....................................................

1-7 1.44 TURBINE BYPASS RESPONSE TIME......................................

1-7 1.45 UNIDENTIFIED LEAKAGE..............................................

1-7 1.46 VENTILATION EXHAUST TREATMENT SY5 TEM..............................

1-7 1.47 VENTING................................................

1-7 i

LA SALLE - UNIT 2 11 Amendment No. 69

d INDEX t

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F10w...........................

2-1 THERMAL POWER, High Pressure and High F10w........................

2-1 Reactor Coolant System Pressure...................................

2-1 Reactor Vessel Water Leve1........................................

2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation-Setpoints..............

2-3 BASES 2.1 SAFETY LIMITS i

THERMAL POWER, Low Pressure or Low Flow...........................

B 2-1 THERMAL POWER, High Pressure and High F1ow........................

B 2-2 Reactor Coolant System Pressure..........................

B 2-8 Reactor Vessel Water Leve1........................................

B 2-8 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints...............

B 2-9 LA SALLE - UNIT 2 III r-..--n.-,.

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY..................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN..............................................

3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.........................................

3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability......................................

3/4 1-3 Control Rod Maximum Scram Insertion Times....................

3/4 1-6 Control Rod Average Scram Insertion Times....................

3/4 1-7 Four Control Rod Group 3 cram Insertion Times.................

3/4 1-8 Control Rod Scram Accumulators...............................

3/4 1-9 Control Rod Drive Coupling...................................

3/4 1-11 Control Rod Position Indication..............................

3/4 1-13 Control Rod Drive Housing Support............................

3/4 1-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer..........................................

3/4 1-16 Rod Sequence Control System..................................

3/4 1-17 Rod Block Monitor............................................

3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................

3/4 1-19 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM...........................

3/4 1-23 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...................

3/4 2-1 3/4.2.2 APRM SETP0lNTS...............................................

3/4 2-2 3/4.2.3 MINIMUM CRI.ICAL POWER RATI0.................................

3/4 2-J 3/4.2.4 LINEAR HEAT GENERATION RATE..................................

3/4 2 5 LA SALLE - UNIT 2 IV Amendment No. 54

.e INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE i

3 /4. 3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION....................

3/4 3-1 3/4.3.2 ISOLATION ACTUATION INST RUMENTATION..........................

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

3/4 3-23 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATVS Recirculation Pump Trip System Instrumentation..........

3/4 3-35 End-of-Cycle Recirculation Pump Trip System Instrumentation............................................

3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...........................................

3/4 3-45 3/4.3.6 CONTROL ROD WITHDRAWAL BLOCK IN$1RUMENTATION.................

3/4 3-50 3/4.3,7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation................

3/4 3-57 Seismic Monitoring Instrumentation...........................

3/4 3-60 Meteorological Monitoring Instrumentation....................

3/4 3-53 Remote Shutdown Monitoring Instrumentation...................

3/4 3-66 Accident Monitoring Instrumentation..........................

3/4 3-69 Source Range Monitors........................................

3/4 3-72 Traversing In-core Probe System..............................

3/4 3-73 Deleted......................................................

3/4 3-74 Fire Detection Instrumentation...............................

3/4 3-75 Deleted......................................................

3/4 3-81 l

Exploxive Gas Monitoring Instrumentation.....................

3/4 3-82 Loose-Part Detection System..................................

3/4 3-85 3/4.3.8 FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTUATION

^

INSTRUMENTATION.........................

3/4 3-86 IA SALLE - UNIT 2 V

Amendment No. 69

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM i

Recirculation Loops..........................................

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

3/4 4-3 Recirculation loop Flow.........

3/4 4 4 i

Idle Recirculation Loop Startup..............................

3/4 4 5 Thermal Hydraulic Stability..................................

3/4 4-Sa 3/4.4.2 SAFETY / RELIEF VALVES.........................................

3/4 4-6 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....................................

3/4 4-7 Operational Leakage..........................................

3/4 4-8

)

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

3/4 4-11 3/4.4.5 SPECIFIC ACTIVITY.,.........................................

3/4 4-14 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.......................................

3/4 4-17 R e a c t o r S t e am Dome........................................... - 3 /4 4 - 21 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.............................

3/4 4-22 3/4.4.8 STRUCTURAL INTEGRITY.........................................

3/4 4-23 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown................................................

3/4 4-24 Cold Shutdown................................................

3/4 4-25 3/4.5 EMERGENCY CORE 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 CHAMBER.........................................

3/4 5-8 LA SALLE - UNIT 2 VI Amendment No. 69 F

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l e-INDEX 1

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity...............................

3/4 6-1 P ri ma ry Cont ai nment Lea kage.................................

3/4 6-2 Primary Containment Air Locks...............................

3/4 6-5 MSIV Leakage Control System.................................

3/4 6-7 Primary Containment Structural Integrity....................

3/4 6-8 Drywell and Suppression Chambei Internal Pressure...........

3/4 6-16 Drywell Average Air Temperature...............

3/4 6-17 Drywell and Suppression Chamber Purge System................

3/4 6 3/4.6.2 DEPRESSURIZATION SYSTEMS Suppression Chamber.............

3/4 6-19 Suppression Pool Spray.......

3/4 6-23 Suppression Pool Cooling....................................

3/4 6-2^

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES........................

3/4 6-25 3/4.6.4 VACUUM RELIEF...............................................

3/4 6-38 3/4.6.5 SECONDARY CONTAINMENT Secondary Containment Integrity.............................

3/4 6-40 Secondary Containment Automatic Isolation Dampers...........

3/4 6 41 Standby Gas Treatment System................................

3/4 6-43 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Drywell and Suppression Chamber Hydrogen Recombiner Systems...................................................

3/4 6-46 Drywell and Suppression Chamber Oxygen Concentration........

3/4 6-47 LA SALLE - UNIT 2 VII

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 COPE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Service Water System...................

3/4 7-1 Diesel Generator Cooling Water System........................

3/4 7-2 Ultimate Heat Sink...........................................

3/4 7-3 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM.................................

3/47-4 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM........................

3/4 7-7 3/4.7.4 SEALED SOURCE CONTAMINATION..................................

3/4 7-9 3/4.7.5 FIRE SUDPRESbl0N SYSTEMS Fire Suppression Water System................................

3/4 7-11 Deluge and/or Sprinkler Systems..............................

3/4 7-14 C0 Systems..................................................

3/4 7-17 7

Fire Hose Stations...........................................

3/4 7-18 3/4.7.6 FIRE RATED ASSEMBLIES.......................................

3/4 7-23 3/4.7.7 AREA TEMPERATURE MONITORING..................................

3/4 7-25 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS 1 STRUCTURES...................

3/4 7-27 3/4.7.9 SNUBBERS....................................................

3/4 7-28 3/4.7.10 MAIN TURBINE BYPASS SYSTEM...................................

3/4 7-34 l

i I

LA SALLE - UNIT 2 VIII I

--.-,-.,...-.n..

~

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A. C. S o u rc e s - Ope ra t i n g.......................................

3/4 8-1 i

A.C. Sources-Shutdown......................................

3/4 8-8 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating................................

3/4 8 A. C. Di s t ribut i on - Shutdown................................

3/4 8-12 D.C. Distribution - Operating................................

3/4 8 14 D. C. Di s t ri buti on - Shutdown.................................

3/4 8-19 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. Circuits Inside Primary Containment.....................

3/4 8-21 Primary Containment Fenetration Conductor Overcurrent Protective Devices.........................................

3/4 8-22 Motor Operated Valves Thermal Overload Protection............

3/4 8-26 Reactor Protection System Electric

?

Power Monitoring...........................................

3/4 8-31 i

e I

LA SALLE - UNIT 2 IX e

w 2..,

r-+..,..-.-,.,.

,-..,,,.,.,..,,y_.,,_,-....,-

n.

v

,....,.,r.,

,.,,,p.,..

y-,

_..-,-.-,e-.w,y,-,m..y, erw,---.-,3-3y-_,_-e,--

cy.,.,

... = _- -._..- -._-- -.._-- - -. -

INDEX 4

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

SECTION

.PAGE 1

3/4.9 REFUELING OPERATION 3 i

3/4.9.1 REACTOR MODE SWITCH..........................................

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

3/4 9-3 i

3/4.9.3 CONTROL R0D P0SITION.........................................

3/4 9-5 3/4.9.4 DECAY TIME...................................................

3/4 9-6 3/4.9.S COMMUNICATIONS...............................................

3/4 9-7 3/4.9.6 CRANE AND H0lST..............................................

3/4 9-8 3/4.9.7 CRANE 1 RAVEL.................................................

3/4 9-9 3/4.9.8 WATER LEVEL - REACTOR VESSEL.................................

3/4 9-10 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L........................

3/4 9-11 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal.........

3/4 9-12 Multiple Control Rod Remova1.................................

3/4 9-14 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1.............................................

3/4 9-16 Low Water Leve1......................................

3/4 9-17 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY................................

3/4 10-1 3/4.10.2 ROD SEQUENCE CONTR01. SYSTEM...................................

3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS...............................

3/4 10-3 3/4.10.4 DELETED......................................................

3/4 10-4 l

3/4.10.5 OXYGEN CONCENTRATION.........................................

3/4 10-5 3/4,10.6 TRAINING STARTUPS............................

3/4 10-6 3/4.10.7 DELETED............................................

3/4 10-7 LA SALLE - UNIT 2 X

Amendment No. 69

~e-,

,,,.,....--ev

,..ym,

.,,,--.,,,#.y.

_,,,__,,.m.,._

,_w-- -., -. -

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVE]LLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks.........................................

3/4 11-1 3/4 11.2 GASEOUS EFFLUENTS E x-p l o s i v e G a s M i x t u r e........................................

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

3/4 11-3 LA SALLE - UNIT 2 XI Amendraent No. 69 1

i

INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY...................................................

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 AN0MALIES....................................

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

B 3/4 1-2 l

3/4.1.4 CONTROL R0D PROGRAM CONTR0LS............................

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

B 3/4 1-4 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM......................

B 3/4 1-5 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..............

B 3/4 2-1 3/4.2.2 APRM SETP0lNTS.............

B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................

B 3/4 2-2 3/4.2.4 LINEAR HEAT GENERATION RATE.......

B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...............

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

B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.........................................

B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.......

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

B 3/4 3-4

/4.3.6 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION............

B 3/4 3-4 3

3/4.3.7 MONITORING INSTRUMENTATION t

1 Radiation Monitoring Instrumentation....................

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

B 3/4 3-4

\\

1-LA SALLE - UNIT 2 XII Amendment No. 69

i l

INDEX BASES SECTION PAGE INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Meteorological Monitoring Instrumentation...............

B 3/4 3-4 Remote Shutdown Monitoring Instrumentation..............

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

B 3/4 3-5 Source Range Monitors...................................

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

B 3/4 3-5 Deleted.................................................

B 3/4 3-5 Fire Detection Instrumentation..........................

B 3/4 3-5 Deleted................................

B 3/4-3-5 Explosive Gas Monitoring Instrumentation.................

B 3/4 3-6 Loose-Part Detection 5ystem..............................

B 3/4 3-6 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION........................................

B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM....................................

B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVE 5.....................................

B 3/4 4-la 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...

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

B 3/4 4-2 3/4.

CHEMISTRY................................................

B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY........................................

B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMIT5..............................

B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVE 5.........................

B-3/4 4-5 j.

3/4.4.8 STRUCTURAL INTEGRITY..................................

B 3/4 4-5 1

3/4.4.9 RESIDUAL HEAT REM 0 VAL....................................

B 3/4 4-5 l

l LA SALLE - UNIT 2-XIII Amendment No. 69

.. _.... ~.

-..~.

INDEX BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS-OPERATING and SHUTDOWN.................

B 3/4 5 1 3/4.5.3 SUPPRESSION CHAMBER.....................................

B 3/4 5-2 3/4.6 CONTAINME:

SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity...........................

B 3/4 6-1 Primary Containment Leakage.............................

B 3/4 6 Primary Containment Air Locks....................

B 3/4 6-1 MSIV Leakage Control System.............................

B 3/4 6-1 Primary Containment Structural Integrity................

B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure.......

B 3/4 6-2 Drywell Average Air Temperature.........................

B 3/4 6-2 Drywell and Suppression Chamber Purge System............

B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS................................

B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES....................

8 3/4 6-4 3/4.6.4 VACUUM RELIEF...........................................

B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT...................................

B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTR0L..................

B 3/4 6-5 5

LA SALLE - UNIT 2 XIV

.-.+.-.-.----.,...--,....__.n.-

...., - - - _, -,, -,. - - -,,,, - - - -.,. ~, - - - - - -

,------,..,,n

, - ~., - -,

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS.........................................

B 3/4 7-1 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM...........................

B 3/4 7-1 I

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

B 3/4 7-1

4.7.4 SEALED SOURCE CONTAMINATION.............................

B 3/4 7-2 3/4.7.5 FIRE SUPPRESSION SYSTEMS..............................

B 3/4 7-2 3/4.7.6 FIRE RATED ASSEMBLIES.................................

B 3/4 7-3 3/4.7.7 AREA TEMPERATURE MONITORING.............................

B 3/4 7-3 3/4.7.8 STRUClVRAL INTEGRITY OF CLASS 1 STRJCTURES..............

B 3/4 7-3 3/4.7.9 SNUBBERS................

B 3/4 7-3 3/4.7.10 MAIN TV'RBINE BYPASS SYSTEM.............................

B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS.................................

B 3/4 8-1 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.................

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

B 3/4 9-1 3/4.9.^

INSTRUMENTATION.........................................

B 3/4 9-1 3/4.9.3 CONTROL ROD POSIT 10N..............

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 CRANE AND H0lST.........................................

B 3/4 9-1 3/4.9.7 CPANE TRAVEL............................................

B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L...............

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

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

B 3/4 9-2 LA SALLE - UNIT 2 XV Amendment No. 69

~

INDEX BASES SECTION PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY...........................

B 3/4 10-1 3/4.10.2 P90 SEQUENCE CONTROL SYSTEM.............................

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

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

B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION....................................

B 3/4 10-1 3/4.10.6 TRAINING STARTUPS.......................................

B 3/4 10-1 3/4.10.7 CONFIRMATORY FLOW INDUCEO VIBRATION TEST................

B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks..................

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

Explosive Gas Mixture.

B 3/4 11-1 Main Condenser..

B 3/4 11-1 i

I i

LA SALLE - UNIT 2 XVI Amendment No. 69

- ~..

e

.t INDEX l

DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area....................................................

5-1 Low Population Zone.......................................

5,

Site Boundary for Gaseous Ef fl uents...............................

5-1 Site Boundary for Liquid Effluents................................

5-1 5.2 CONTAINMENT Configuratio_n.....................................................

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

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

5-1

5. 3 REACTOR CORE Fuel Assemblies...................................................

5-4 Control Rod Assemblies...........................................

5-4 5.4 REACTOR COOLANT SYSTEM Des ign P re s s t.re and Tempe ra ture...................................

5-4 Vo1tme............................................................

5-4 5.5 METEOROLOGICAL TOWER LOCATION.....................................

5-4 5.6 FUEL STORAGE Criticality.......................................................

5-5 Drainage............

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

5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...............................

5-5 l

[

I I

LA SALLE - UNIT 2 XVII L

i I

l

INDtX ADMINISTRATIVE CONTROLS SECTION PAGE 6j ORGANIZAlION, REVIEW, INVESTIGATION. AND AUDIT....................

6-1 6.1.1 High Radiation Areas.........

6-15 6.2 PLANT OPERATING PROCEDURES AND PR0 GRAMS...........................

6-16 6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE-EVENT IN PLANT 0PERATION..........................................

6-20 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT 15 EXCEEDED........

6-20 6.5 PLANT OPERATING REC 0RDS...........................................

6-21

[.6_REPORTINGREQUIREMENTS.........................................

6-22 6

6.7 PROCESS CONTROL PROGRAM.........

6-26 6.8 0FFSITE DOSE CALCULATION MANUAL...................................

6-27 6.9 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS..............

6-27 I

i l-l l

l l

l I

(

i l

l LA SALLE - UNIT 2 XVIII Amendment No. 69

-c,--,----

LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS..

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na 0 0 10 H O) 2 10 16 2

VOLUME /CONCENTRAT10N REQUIREMENT 5 3/4 1-22 3.4.1.5-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATED)..

3/4 4-Sc 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE 3/4 4-19 3.4.6.1-la MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE 3/4 4-19a 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST 3/4 7-33 C 3/4 3-1 REACTOR VESSEL WATER LEVEL,

B 3/4 3-7 l

B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS B 3/4 6-3a 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LlQUID EFFLUENTS 5-2 5.1.2-1 LOW PO?..ATION ZONE 5-3 6.1-1 DELETED 6-11 6.1-2 DELETED 6-12 6.1-3 MINIMJM SHIFT CREW COMPOSITION 6-13 LA SALLE - UNIT 2 XIX Amendment No. 5t

LIST OF TABLES TABLE PAGE 1.1 SURVEILLANCE FREQUENCY NOTATION..................

1-8 1.2 OPERATIONAL CONDITIONS............................

1-9 2.2,1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS.........................................

2-4 B2.1.2-1 DELETED.......................................

B 2-4 B2.1.2-2 DELETED...

B 2-5 B2.1.2-3 DELETED....

B 2-6 5

B2.1.2-4 DELETED B 2-7 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 3/4 3-2 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES........

3/4 3-6 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................

37, 3-7 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION 3/4 3-11 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS....

3/4 3-15 3,3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME....

3/4 3-18 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3/4 3-20 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..............................

3/4 3-24 3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS......................

3/4 3-28 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES......

3/4 3-31 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-32 3.3.4.1-1 ATVS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION........

3/4 3-36 LA SALLE - UNIT 2 XX Amendment No. 69

l LIST OF TABLES (Continued)

TABLE PAGE 3.3.4.1-2 ATWS RECIRCv!ATION PUMP TRIP SYSTEM INSTPUMENTATION SETPOINTS......................

3/4 3-37 4.3.4.1-1 ATVS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-38 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION................................

3/4 3-41 3.3.4.2-2 END-0F-CYCLE RECIRCULAT'IN PUMP TRIP SYSTEM SETPOINTS............

3/4 3-42 3.3.4.2-3 END-OF-CYCLE RECIRCULATION PUMF. TRIP SYSTEM RESPONSE TIME.......

3/4 3-43 4.3.4.2.1-1 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS.............

3/4 3-44 3.3.5-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION,

3/4 3-46 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 3/4 3-48 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......

3/4 3-49 3.3.6-1 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION 3/4 3-51 3.3.6-2 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS...

3/4 3-53 4.3.6-1 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............

3/4 3-55 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION.

3/4 3-58 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............

3/4 3-59 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION.

3/4 3-61 4.3.7.2-1 SEISMIC HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................

3/4 3-62 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION 3/4 3-64 4.3.7.3-1 METEOR 0!001 CAL MONITORING INSTRUMENTATION SURVEI!.U.hCE REQUIREMENTS.

3/4 3-65 1

i LA SALLE - UNIT 2 XXI

LIST OF TABLES (Continued)

TABLE PAGE 3.3.7.4-1 REMOTE SHUTDOWN HONITORING INSTRUMENTATION........

3/4 3-67 4.3.7.4-1 REMOTE SHUTDOWN H3NITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................

3/4 3-68 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION...............

3/4 3-70 4.3.7.5-1 ACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................

3/4 3-71 3.3.7.9-1 FIRE DFTECTION INSTRUMENTATION....................

3/4 3-76 3.3.7.11-1 EXPLOSIVE GAS HONITORING INSTRUMENTATION..........

3/4 3-83 4.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...................

3/4 3-84 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.........................

3/4 3-87 3.3.8-2 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS...............

3/4 3-88 4.3.8.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-89 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.

3/4 4-10 3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...........

3/4 4-13 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.................................

3/4 4-16 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE...............................

3/4 4-20 4.6.1.5-1 TENDON SURVEILLANCE..............................

3/4 6-11 4.6.1.5-2 TENDON LIFT-0FF FORCE...........................

3/4 6-12 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES.............

3/4 6-27 LA SALLE - UNIT 2 XXII Amendment No. 69

.~

LIST OF TABLES (Continued)

TABLE PAGE 3.6.5.2-1 SECONDARY CONTAINHENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS....................

3/4 6-42 3.7.5.2-1 DELUGE AND SPRINKLER SYSTEMS.....................

3/4 7-16 3.7.5.4-1 FIRE HOSE STATIONS...............................

3/4 7-19 3.7.7-1 AREA TEMPERATURE MONITORING.......................

3/4 7-26 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE....................

3/4 8-7b 4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS.................

3/4 8-18

? 8.3.2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 3/4 8-24 3.8.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION.........

3/4 8-27 B3/4.4.6-1 REACTOR VESSEL TOUGHNESS....................

8 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS 5-6 LA 3ALLE - UNIT 2 AXIII Amendment No. 69

f DEFINITIONS LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LHGR) 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 logic components, i.e., all relays and contacts, all tri) units, solid state logic elements, etc. of a logic circuit, from sensor tirough and including the actuated device to verify OPERABILITY.

THE LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

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

MEMBER (5) 0F THE PUBLIC 1.24 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupation-ally associated with the plant.

This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.25 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.26 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2) descriptions of the information that should be i

included in the Annual Radiological Environmental Operating and l

Semi-Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

OPERAELE - OPERABILITY l

1.27 A system, subsystem, train, component or device shall be OPERABLE or have l

l OPERABIllTY when it is capable of performing its specified-function (s),

l and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, 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).

LA SALLE - UNIT 2 1-4 Amendment No. 69 l

DEFINITIONS OPERATIONAL CONDITION - CONDITION 1.28 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive l

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 I

nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Connission.

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

PRIMARY CONTAINMENT INTEGRITY 1.31 PRIMARY CONTAINMENT INTEGRITY shall exist when; a.

All primary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE primary 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 3.6.3-1 of Specification, 3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

c.

Each primary containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d.

The primary containment leakage rates are within the limits cf Specification 3.6.1.2.

e.

The suppression chamber is OPERABLE pursuant to Specification 3.6.2.1.

f.

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

PROCESS CONTROL PROGRAM 1,32 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demon-strated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial grouno requirements, and other requirements governing the disposal of solid radioactive waste.

l PURGE - PURGING 1.33 PURGE or PURGING shall be the controlled process of discharging air or f

gas from a confinement to maintain temperature, pressure, humidity, l

concentration or other operating condition, in such a manner that replace-ment air or gas is required to purify the confinement.

LA SALLE - UNIT 2 1-5 Amendment No. 69 t

5 DEFINITIONS RATED THERMAL POWER 1.34 RATED THERMAL POWER shall be a total reactor core heat transfer rate to l

the reactor coolant of 3323 H4T.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.35 REACTOR PROTECTION SYSTEM RE5PONSE 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 EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in i-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 l

fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 1001 ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.38 SECONDARY CONTAINMENT INTEGRITY shall exist when:

a.

All secondary containment 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.

Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed i

position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b.

All secondary containment hatches and blowout panels are closed and sealed, c.

The standby gas treatment system is OPERABLE pursuant to Specification 3.6.5.3.

d.

At least one door in each access to the secondary containment is closed.

e.

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

l f.

The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

l SHUTDOWN MARGIN 1,39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully t

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 l

condition; cold, i.e. 68 F; and xenon free.

LA SALLE - UNIT 2 1-6 Amendment No. 69

DEFINITIONS 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.

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel _ response l

when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.42 A STAGGERED TEST BASIS shall consist-of:

l a.

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

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

THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the l

reactor coolant.

TURB]NE BYPASS SYSTEM RESPONSE TIME 1.44 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when l

the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

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

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

I VENTILATION EXHAUST TREATMENT SYSTEM 1.46 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and I

installed to reduce gaseous radioiodine or radioactive material in particu-late _ form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the. purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.47 VENTING shall be the controlled process of discharging air or gas from a l

confinement to maintain temperature, pressure, humidity, concentration or-other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

LA SALLE - UNIT 2 1-7 Amendment No. 69

~, -,

.. _ -... ~ =_

TABLE 1.1 SURVEILLANCE FREQUENCY NOTATION NOTAT10N' FREQUENCY.

5 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days, Q

At least once per 92 days.

SA At least on'ce per 184 days.

A At least once per 366 days.

R At least once per 18 months (550 days).

5/U Prior to each reactor startup.

P Prior to each radioactive release.

N. A.

Not applicable.

Y LA SALLE - UNIT 2 1-8

.-=

4 3.3.7.10 deleted PAGE 3/4 3-81 INTENTIONALLY LEFT BLANK i

i LASALLE - UNIT 2 3/4 3-81 Amendment No. 69

INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.11 The explosive gas monitoring instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their Alarm / Trip setpoints set to ensure that the limits of specification 3.11.2.6 are not exceeded.

APPLICABILITY:

Duringoperationofthemaincondenserairejector.

ACTION:

a.

With an explosive gas monitoring instrumentation channel Alarm / Trip setpoint less conservative than required by the above specification, declare the channel inoperable, and take the ACTION shown in Table 3.3.7.11-1.

b.

With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.11-1.

Restore the inoserable instrumentation channels to an OPERABLE status wit 1in 30 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C. within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status, c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.11 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by aerformance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at-the frequencies shown in Table 4.3.7.11-1.

LASALLE - UNIT 2 3/4 3-82 Amendment No. 69

INSTRUMENTATION TABLE 3.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1.

MAIN CONDENSER OFFGAS TREATHENT SYSTEM EXPLOSIVE GAS HONITORING SYSTEM (for systems designed to withstand the effects of a hydrogen explosion) a.

Hydrogen Monitor 1/ train 110 TABLE NOTATION ACTION 110 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation 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 .

If the recombiner(s) temperature remains constant and THERMAL POWER has not changed, the grab sample collection frequency may be-changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

LASALLE - UNIT 2 3/4 3-83 Amendment No. 69

4 INSTRUMENTATION TABLE 4.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEIL-INSTRUMENT CHECK TEST CALIBRATION

LANCE REQUIRED 1.

MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM a.

Hydrogen Monitor D

M Q

TABLE NOTATION The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1.

One volume percent hydrogen, balance nitrogen, and 2.

Four volume percent hydrogen, balance nitrogen.

Duringoperationofthemaincondenserairejector.

LASAli.E - UNIT 2 3/4 3-84 Amendment No. 69

INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.12 The loos part detection system shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a.

With one or more loose part detection system channels inoperable for more'than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.c within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status, b.

The provisions of Specifications 3.0,3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.12 Each channel of the loose part detection system shall be demonstrated OPERABLE by performance of:

a.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

CHANNEL FUNCTIONAL TES1 at least once per 31 days, and c.

CHANNEL CALIBRATION at least once per 18 months.

LASALLE - UNIT 2 3/4 3-85 Amendment No. 69

INSTRUMENTATION 3/4.3.8 FEE 0 WATER /MplNTURBINETRIPSYSTEMACTUATIONINSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.8 The feedwater/ main turbine trip system actuation instrumentation channels shown in Table 3.3.8-1 shall be OPERABLE with their tri) setpoints set consistent with the values shown in the Trip Setpoint column of Ta)1e 3.3.8-2.

APPLICABILITY:

OPERATIONAL CONDITION 1.

ACTION:

With a feedwater/ main turbine trip system actuation instrumentation a.

channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.8-2, declare the channel inoperable and either place the inoperable channel in the trip]ed condition until the channel is rostored to OPERABLE status witi its trip setpoint adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable, b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel 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 .

With the number of OPERABLE channels two less than required by the c.

Minimum OPERABLE Channels per Trip System requirement, 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 .

SURVEILLANCE REQUIREMENTS 4.3.8.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations 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.

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

69 1

i 9

s P

TABLE 3.3.8 m FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION I

E a

n MINIMUM OPERABLE CHANNELS TRIP FUNCTION PER TRIP SYSTEM i

i a.

Reactor Vessel Water Level-High. Level 8 3

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TABLE 3.3.8-2 E

Z FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS i

ro ALLOWA8LE TRIP FUNCTION TRIP SETPOINT VALUE s.

Reactor Vessel Water Level-High, Level 8

< 55.5 inches *

< 56.0 inches

l l

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P TABLE 4.3.8.1-1 1

E FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUR NTATION SURVEILLANCE REQUIREMENTS

-e to CHANNEL CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION CHECK TEST CALIBRATION a.

Reactor Vessel Water Level-High, S M

R Level 8 Y

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3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.1 The quantity of radioactive material contained in any outside temporary l-tankt shall be limited to less than or equal to the limits calculated in the ODCM.

APPLICABILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive aaterial to the tank 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.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.1 The quantity of radioactive material contained in each of the above l

listed tanks 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 radioactive materials are being added to the tank.

LA SALLE - UNIT 2 3/4 11-1 Amendment No. 69

4 RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASE0US EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.1 The concentration of hydrogen in the main condenser offgas treatment l

system shall be limited to less than or equal to 4% by volume.

APPLICABILITY:

Wheneverthemaincondenserairejectorsystemisinoperation.

ACTION:

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

treatment system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , b.

The provisinns of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.1 The concentration of hydrogen in the main condenser offgas treatment l

system shall be determined to be within the above limits as required by-Table 3.3.7.11-1 of Specification 3.3.7.11.

I 1

LA SALLE - UNIT 2 3/4 11-2 Amendment No. 69 w

+

is--

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RADIOACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR CPERATION 3.11.2.2 The release rate of the sum of the activities f rom the noble gases measured prior to the holdup line shall be limited to less than or equal to 3.4 x 105 microcuries/second.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the release rate of the sum of the activities of the noble gases prior to the holdup line exceeding 3.4 x 105 microcuries/second restore the 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 STARTUP with the main steam isolation valves closed within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.11.2.7.1 The radioactivity rate of noble gases prior to the holdup line shall be continuously monitored in accordance with Specification 3.3.7.11.

4.11.2.7.2 The release rate of the sum of the activities from noble gases prior to the holdup line shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies by performing an isotepic analysis of a representative sample of gases taken prior to the holdup line, 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 off gas pre-treatment Noble Gas Activity Honitor, 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 primary coolant.

l L

LA SALLE - UNIT 2 3/4 11-3 Amendment No. 69

INSTRUMENTATION BASES MONITORINGINSTRUMENTATION(Continued) 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.

This capability is con-sistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Pl::nt Conditions During and Following an Accident," December 1975 and NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

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 made 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.

3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe (TIP) system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux dis-tribution of the reactor core.

The specification allows use of substituted TIP data from symmetric channels if the control rod pattern is symmetric since the TIP data is adjusted by the plant computer to remove machine dependent and power level dependent bias.

The source of data for the substitution may also be a 3-dimensional BWR core simulator calculated data set which 'is normalized to available real data.

Since uncertainty could be introduced by the simulation and normalization process, an evaluation of the specific control rod pattern and core operating state must be performed to ensure that adequate margin to core operating limits is maintained.

3/4.3.7.8 DELE h 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.

This capability is recuired in order to detect and locate fires in their early stages.

Prompt cetection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, increasing the frequency of fire watch patrols in the affected L

areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.7.10 DELETED LA SALLE - UNIT 2 B 3/4 3-5 Amendr.!ent No. 69

-INSTRUMENTATION BASES 3/4.3.7.11 EXPLOSIVE GAS MONITORING INSTRUMENTATION This instrumentation provides for monitoring (and contrelling) the con-centrations of potentially explosive gas mixtures in the wasta gas holdup system.

3/4.3.7.12 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part etection system ensures that sufficient capability is available to detect loose a,etallic parts in the primary system and avoid or mitigate damage to primary system components.

The allowable out-of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1.133, "Loosc-Part Detection Program for the Primary System of Light-' water-Cooled Reactors."

3/4.3.8 FEFDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate the feedwater system / main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 setpoint associated with a feedwater controller failure to prevent over-filling the reactor vessel which may result in high pressure liquid dis-charge through the safety / relief valve discharge lines, LA SALLE - UNIT 2 8 3/4 3-6 Amendment No. 69

~_-.

3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.

LIQUID EFFLUENTS 3/4.11.1.1 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled 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.

314.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 EXPLOSIVE GAS MIXTURE The specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen.

Maintaining the concentration of hydrogen and oxygen below their flammability limits 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.

3/4.11.2.2 MAIN CONDENSER, l

Restricting tha gross radioactivity rate.of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area-boundary will r.at exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment weshout treatment.

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

LA SALLE - UNIT 2 8 3/4 11-1 Amendment No. 69

5- 0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.

SITE BOUNDARY FOR GASEOUS EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1.1-1.

SITE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 The site boundary for liquid ef fluents shall be as shown in Figure 5.1.1-1.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel linea post-tensioned concrete structure consisting of a drywell and suppression chamber.

The drywell is a steel-lined post-stressed concrete vessel in the shape of a truncated cone closed by a steel dome.

The drywell is above a cylindrical steel-lined post-stressed concrete suppression chamber and is attached to the suppression chamber through a series of downcomer vents.

The drywell has a minimum free air volume of 229,538 cubic feet.

The suppression chamber has an air region of 164,800 to 168,100 cubic feet and a water region of 128,800 to 131,900 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is_ designed and shall be maintained for; a.

Maximum internal pressure:

45 psig.

b.

Maximum internal temperature:

drywell 340*F.

suppression chamber 275'F.

c.

Maximum external pressure:

5 psig.

d.

Maximum floor differential pressure:

25 psid, downward.

5 psid, upward.

SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the Reactor Building, the equipment access structure and a portion of the main steam tunnel and has a minimum free volume of 2,875,000 cubic feet.

LA SALLE - UNIT 2 5-1

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~ _ _ _.

ADMIN!5TRATION CONTROLS PLANT OPERATING PROCEOURES AND PROGRAMS (Continued)

F.

The following programs shall be established, imple.mented, and maintained:

i 1.

Primary Coolant Sources Outside Primary Containment A program to reduce leakage from those portions of systems outside primary containment that could contain highly radioactive fluids-during a serious transient or accident to as low as practical levels.

The systems include LPCS, HPCS, RHR/LPCI, RCIC, hydrogen recombiner, process sampling, containment monitoring, and standby gas treatment systems.

The program shall include the following:

Preventive maintenance and periodic visual insoectio'i require-a.

ments, and b.

Integrated leak test requirements for each system et refueling cycle intervals or less.

2.

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 progr'9 shall include the folinwing:

a.

Training of personnel, b.

Procedures for monitoring, and Provisions for maintenance of sampling and analysis equipment.

c.

3.

Post-accident Sampi,in_g li A program which will (<nsure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following:

a.

Training of personnel, b.

Procedures for sampling and analysis, c.

Provisions for maintenance of sampling and analysis equipment.

4.

Radioactive Effluent-Controls Program A program shall be provided confenning with 10 CFR 50.36a for_ the control of radioactive effluents and for maintai_ning the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievablei The program (1) shall be contained in the 0DCM, (2) shall be implemented by operating procedures, and (3)- shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

LA AALLE - UNIT 2 6-18 Amendment No. 69

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.-..-. m m.

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4 ADMINISTRATION CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) a.

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, b.

Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2, Monitoring, sampling, and analysis of radioactive liquid and gaseous c.

etfluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM, i

d.

Limitations on the annual and quarterly doses or dose commitment to a 4

MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix !

to 10 CFR Part 50, i

c.

Determination of cumulative and projected dose contributions from radioactive ef fluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, f.

Limitations on the operability and use of the liquid and gaseous ef-fluent treatment systems to ensure that the appropriate po. ions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the-guidelines for the annual dose or dose commitment conforming to Appen-dix ! to 10 CFR Part 50, g.

Limitations on the dose rate resulting from radioactive material re-i leased in gaseous effluents to areas beyond the SITE BOUNDARY conform-ing to the doses associated with-10 CFR Part 20, Appendix B, Table 11.

4 Column 1, h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix 1 to 10 CFR Part 50, i.

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in parti-culate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, J.

Limitations on venting and_ purging of the containment through the Primary Containment Vent and Purge System or Standby Gas Treatment System to maintain releases as low as reasonably achievable, k.

Limitations on the annual dose or dose commitment to any MEMBE9 0F THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel. cycle sources ronforming to 40 CFR Part 190.

LA SALLE - UNIT 2 6-19 Amendment No. 69

a.

i 9

ADMINISTRATION CONTROLS i

?

PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) 5.

Radiological r~dronmental Monitoring Prograr' A program s'

W provided to monitor the radiation and radionuclides in I

the envirora the plant.

The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, i

and (2) verification cf the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part $0, and (3) include the following:

Monitoring, sampling, analysis, and reporting of radiation and radiu--

a.

nuclides in the environment in accordance with the methodology and parameters in the ODCM, b.

A Land Use Census to ensure that c h nges in the use of areas at and beyond the SITE BOUNEt.RY cre identified ar.d that modifications to the monitoring program are made if required by the results of this census, and c.

Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are per-formed as part of the quality assurance program for environmental monitoring.

i 6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN PLANT OPERATION The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified and a Licensee Event Report submitted pursuant to the requirements of Section 50,73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed pursuant to Specifica-tion 6.1.G.2.c(1).

L 6.4 ACTION-10 BE TAKEN IN THE EVENT A SAFETY LIMli IS EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pursuant to Specification 2.1.1, 2.1.2 and 2.1.3, and critical reactor operation shall not be resumed until authorized by the NRC.

The conditions of shutdown shall be promptly reported to the Vice President BWR Operations or his designated alternate.

The incident shall be reviewed pursuant to Specifications 6.1.G.1.a and 6.1 G,2.a and a separate Licensee Event Report for each occurrence shall be prepared in accordance with Section 50.73 to 10 CFR Part 50.

The NRC Operations Center shall be notified by telephone as soon as possible and in all-cases within I

one hour.

The Vice President BWR Operations and the Manager of Off-site Review and Investigative function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l L

LA SALLE - UNIT 2 6-20 Amendment No. 69 l

=

ADMIN 1STRAT10' CONTROLS 6.5 PLANT OPERATING RECORDS A.

Records and/or logs relative to the following items shall be kept in a manner convenient for review and shall be retained for at least 5 years:

1.

Records of normal plant operation, including power levels and periods of operation at each power level; 2.

Records of principal maintenance and activities, including inspection and repair, regarding principal iteme of equipment pertaining to nuclear safety; 3.

Records and reports of reportable esents; 4.

Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (see Section 4 of these specifications) are being met.

All equipment failing to meet surveil-lance requirements and the corrective action taken shall be recorded; 5.

Records of changes to operating procedures; 6.

Shift engineers' logs; and 7.

Byproduct material inventory records and source leak test results.

B.

Records and/or logs relative to the following items shall be recorord in a manner convenient for review and shall be cetained for the life of the plant:

1.

Substitution or replacement of principal items of equipment pert.ain-ing to nuclear safety; 2.

Changes made to the plant as it is described in the SAR; 3.

Records of new and spent fuel inventory and assembly histories; 4,

Updated, corrected, and as-built drawings of the plant; 5.

Records of plant radiation and contamination surveys; 6.

Records of offsite environmental monitoring surveys; 7.

Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant, in accordance with 10 Cfh Part 20; 8.

Records of radioactivity in liquid and gaseous wastes released to the environment; 9.

Records of transient or operational cycling for those components that have been designed to operate safety for a limited number of transient or operational cycles (identified in Table 5.7.1-1);

LA SALLE - UNIT 2 6-21 Amendment No. 69

ADMINISTRATION CONTROLS PLANT OPERAllNG RECORDS (Continued) 10.

Records of individual staff members indicating qualifications, experience, training, and retraining; IL Inservice inspections of the reactor coolant sy< tem; 12.

Minutes of meetings and results of reviews and audits performed by the off,ite and onsite review and audit functions; 13.

Records of reactor tests and experiments:

14.

Records of Quality Assurance activities required by the QA Manual, except for those items specified in Section 6.5.A; 15.

Records of reviews performed for changes made to procedures on equip-ment or reviews of tests and experiments pursuant to 10 CFR 50.59; 16.

Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.9 including the date at which the service life commences and associated installation and maintenance recotds; l

17.

Records of analyses required by the radiological environmental monitoring program; and 18.

Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

6.6 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of federal Regulations, the following identified reports shall be submitted to the director of the appropriate Regional Office of Inspection and Enforce-ment unless otherwise noted.

A.

Routine Reports 1.

Startup Report A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license.

(2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a 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 plant, The report shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications, Any corrective actions that were required to obtain satisfactory operation shall also be described.

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

LA SALLE - UNIT 2 6-22 Amendment No. 69

i ADMINISTRATION CONTROLS

6. 6 REPORTING REQUIREMENTS (Continued)

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the startup report does not cover all three events (i.e.,

initial criticality completion of startup test program, and resumption or commen, cement of commercial power Operation), supple-mentary reports shall be submitted at least every 3 months until all three events have been completed.

2.

Annual Report A tabulation shall be sutimitted on an annual basis prior to March 1 of each year of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions (Note:

this tabulation supplements the requirements of Section 20.407 of 10 CFR 20), e.g., reactor operations and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling less than 20% of the individual total dose need not be accounted for.

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

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5 shall be included in the Annual Report along with the following information:

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

Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other ridioiodine isotope concen-tration in microcuries per gram as a function of time for the dura-tion of the specific activity above the steady-state level and (5)ThetimedurationwhenthespecificactivityoftheprImary coolant exceeded the radiciodine limit.

LA SALLE - UNIT 2 6-23 Amendment No. 69

ADMINISTRATION CONTR0l5 3.

Annual Radiological Environmental Operating Reporta The Annual Radiological Environmentel Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May I of each year.

The report shall include summaries, interpretations, and analysis of trends of the resalts of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix 1 to 10 CfR Part 50.

t 4.

Semiannual Radioactive Effluent Release Report **

The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January I and July I of each year.

The report shall include a su try of the quantities of radioactive liquid and gaseous effluents ar.d solid waste released from the unit.

The material provided shall be (1) consistel.t with the objectives outlined in the ODCM and PCP and (2) in conform 4 rate with 10 CFR 50.36a and Section IV.B.1 of Appendix ! to 10 Cid Part 50.

5.

Monthly Operating Report Routine *eparts of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Nuclear Reactor Regulation, Mail Station-P1-137, US Nuclear Regulatory Commission, Washington, DC 20555, with a copy of the appropriate Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report.

Any changes to the Off511E DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative function.

6.

Core Operating Limits Report a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

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

The submittal-should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

LA SALLE - UNIT 2 6-24 Amendment No. 69

1 *.

ADMINISTRATION CONTROLS C.

Unique Reporting Requirements 1.

Special Reports shall be submitted to the Director of the Office of Inspection and Enforcement (Region III) within the time period specified for each report.

J 6.7 PROCESS CONTROL PROGRAM (PCP)*

6.7.1 The PCP shall be approved by the Commission prior to implementation.

6.7.2 Licensee initiated changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.18.

This documentation shall contain:

1)

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

and 1

2)

A determination that the change will maintain the overall con-formance of the solidified waste product to existin of federal, State, or other applicable regulations.g requirements-b.

Shall become effective upon review and acceptance by the Onsite Review and Investigative Function.

l 4

1 "The Process Control Program (PCP) is common to La Salle Unit I and La Salle Unit 2.

LA SALLE - UNIT 2 6-26 Amendment No. 69 i

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ADMINISTRATION CONTROLS i

6.8 0FFSITE DOSE CALCULATION MANUAL (ODCM)*

6.8.1 The ODCM shall be approved by the Commission prior to implementation.

6.8.2 Licensee-initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.18.

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 level of radi-oactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix 1 to 10 CFR Part-50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after review and acceptance by the On-Site Re-view and Inve.cigative Function and the approval of the Plant Manager on the date specified by the On-Site Review and Investigative function.

Shall be submitted to the Commission in the form of a complete, leg

c.

ible copy of the entire ODCM 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 effective.

Each change shall be identified 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.

6.9 MAJOR CHANGES TO RADI0ACT]VE WASTE TREATMENT SYSTEMS 6.9.1 Licensee initiated major changes to the radioactive waste treatment systems (liquid, gaseous, and solid):

a.

Shall be reported to the Commission-in the Monthly Operating Report for the period in which the evaluation was_ reviewed by the Onsite Review and Investigative function.

The discussion of each change shall contain:

1.

A summary of the evaluation that led to_the determination that the change could be made in accordance with 10 CFR 50.59; 2.

Sufficient detailed information to totilly support the reason-for the_ change without benefit or additional or supplemental information; 3.

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; i

The OFF5ITE DOSE CA(CULATION MANUAL (0DCM) is common to La Salle Unit I and La Salto tinit 2.

LA SALLE - UNIT 2 6-27 Amendment No. 69

1 ADMINISTRATION CONTROLS MAJOR CHAP,GES TO RAD 10 ACTIVE WASTE TREATMENT SYSTEMS (Continued) 4.

An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5.

An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; 6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period to when the changes are to be made; 7.

An estimate of the exposure to plant operating personnel as a result of the change; and 8.

Documentation of the fact that the change was reviewed and found accepteble by the Onsite Review and Investigative Function, b.

Shall become effective upon review and acceptance by the Onsite Review and Investigative function.

LA SALLE - UNIT 2 6-28 Amendment No. 69

-

ML20105C825

Text

Attachment:

Attachment:

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