Text

Amend 40 to License NPF-62,relocating Existing Procedural Details or Specific Requirements in Current Tech Specs Involving Radioactive Effluent Monitoring to ODCM & Solid Gold Waste to Process Control Program
	ML20055G650
Person / Time
Site:	Clinton
Issue date:	07/18/1990
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20055G651	List: ML20055G650;
References
	NUDOCS 9007240031
	Download: ML20055G650 (68)
	v • d • e

,

1 UNITED STATES

4 g

NUCLEAR REGULATORY COMMISSION l

'j WASHINGTON, D. C. 20555 i

^ %,...../

i i

i ILLIN0IS POWER COMPANY, ET AL.

-l DOCKET NO. 50-461 C_LINTON POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 40 -

'I License No. NPF-62

1. ;

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

A.

The. application for amendment by(the licensees) dated January 26, IllinoisPowerCompany*;(IP),and Soyland Power Cooperative,'Inc.

1990 complies with the. standards and req)uirements'of'the Atomic Energy Act of 1954, :as amended (the-Act, and the Comission's

' rules and-regulations set forth in 10 CFR Chapter-I; B.

The facility will operate in conformity with the application,Lthe j

-provisions of the Act, and the rules and regulations of the' Commission; C.

There istreasonable 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 Comission's' regulations; a

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 10lCFR Part 51 of the' Commission's regulations and allfapplicable requirements have:

been satisfied.

2.

Accordingly, the license is amended by changes to-lthe; Technical Specifi--

cations as indicated in the1 attachment to this license amendment, and!

j paragraphL2.C.(2)=of Facility Operating License No.- NPF-62Lis.hereby) i amended to read as follows:

1

Illinois Power Company is. authorized to act as agent'for-Soyland Power; Cooperative, Inc. and has. exclusive. responsibility andl control over the-physical construction, operation and maintenance _of the facility.

4

k..

J

)

3-

.P

-l v

a

l.

2.

t (2) Technical Specifications and Enironmental Protection Plan L

The Technical Specifications' contained in Appendix A and the Environmental Protection Plan' contained in Appendix B, as revised through Amendment No.

, are hereby incorporated into this license.

Illinois Power company'shall operate the-facility-in accorjance with the Technical Specifications and the Environmental l

Protection Plan.

3.

This-license amendrent is effective as of'1ts date of issuance and shall hc implemented within 90 days after its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION.

l ohn N. Hannon, Director Project Directorate 111 Division.of Reactor Projects'- III, IV, Y and Special Projects Office of Nuclear Reactor: Regulation' l

l

Attachment:

Changes to the Technical t

l Specifications t

i.

Date of issuance: July 18. 1990 i

(

I i

i l-L i

i i

F i

l 1

a i

b g

4

++v

.m

j

' et ATTACHMENT TO LICENSE AMENDMENT NO. 40 FACILITY OPERATING LICENSE NO. NPF-62 DOCKET NO. 50-461 Replace the following pages of the Appendix "A" Technical Specifications with 4

the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf i

- pages are provided to maintain document completeness.

Remove Insert

.i i

i ii.

ii iii iii l

viii viii ix ix xii.

x.i xiii-xiii f

xiv xiv xy xy xviii xviii 4

xx xx xxi Xxi xxii xxii xxvi xxvi

'1-4 through 1-9 1

1-4 through 1-9 3/4 3-70 through 3/4 3-74

'3/4i3-70 through 3/4.3-74 (including 3/43-72a)-

3/4'3-95 through 3/4 3-115-3/4.3-95.thrcugh 3/4 3-103 3/4 7-18 j

3/4'7-19.

t 3/4 7-20 l

i 3/4 11-1 through 3/4 11-20 l

3/4 12-1 through 3/4 12-15 B 3/4.3-8 B 3/4 3-8 1

B 3/4:7-3 B 3/4 7-3 il-3/4 7-4.-

.i B 3/4 11-1.through L 3/4 11-6 l

B 3/4.12-1 B 3/4 12-2 6-16a

- 16b 6-18 through 6-21

.6-18 throughi6-21 6-21a 6-23 through 6-26 6-23'through'6-26 i

i I

s

i INDEX

~l'

1. 0 DEFINITIONS SECTION PAGE 1.1 ACTI0N.........................................................

t 1-1.

1.2 AVERAGE PLANAR'EXP05URE.........................................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATIONRATE.....................

1-1

1. 4 CHANNEL CALIBRATION............................................

1-1 i

l

1. 5 CHANNEL CHECK..................................................

1-1

1. 6 CHANNEL FUNCTIONAL TEST........................................

1-1

1. 7 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM 1

RESPONSE TIME..................................................

1-2

1. 8 CORE ALTERATION.........,..............

1-2 7

CORE OPE' RATING LIMITS REPORT.........

1. 9 1-2 1.10 CRITICAL POWER RATI0............................................

1-2 1.11 00SE EQUIVALENT I-131...

1-2 1.12 DRYWELL'INTEGR11Y.......................................'c......

1-2 1.13 E - AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.14 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME.............

1-3 1.15 END-0F-CYCLE. RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME......

1-3 1

a

~1.16 DELETE0........................................................

1-3 1.17 FREQUENCY N0TATION..............................................

1-4 i

1.18 DELETE0........................................................

1-4 l

1.19 IDENTIFIED LEAKAGE.............................................

1-4 1.20 LIMITING CONTROL R00 PATTERN...................................

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

1-4 1

1.22 LOGIC SYSTEM FUNCTIONAL TEST...................................

1-4 o

1.23 DELETE0.........................................-...............

1-4 1

CLINTON - UNIT 1.

i Amendment No.28,40

INDEX

~

1 DEFINITIONS SECTION

_PAGE 1.24 MEMBER (S) 0F THE PUBLIC.........................................

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

1-5 l

1.26 0FFSITE 00SE CALCULATION MANUAL (0DCM).........................

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

1-5 1.28 OPERATIONAL CONDITION - C0NDITION..............................

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

1-5 1.30 PRESSURE 80VHDARY LEAKAGE.......................................

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

1-5 1.32 PROCESS CONTROL PROGRAM (PCP)...................................

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

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

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

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

1-6 r

1.37 R00 0ENSITY.................................................... 7 1.38 SECONDARY CONTAINMENT INTEGRITY................................

1-7.

1.39 SELF TEST SYSTEM...............................................

1-7 1.40 SHUTDOWN MARGIN....................................

1-8 j

1.41 SITE B0VNDARY................

1-8 1,42 DELETE0..........~..............................................

1-8 1.43 0ELETED.........................................................

1-8

-1.44-STAGGERED TEST BASIS...........

1-8~

1.45 THERMAL P0WER....................................................

1-8:

1.~46= TURBINE BYPASS SYSTEM RESPONSE TIME............................

1-8 1.47-UNIDENTIFIED LEAKAGE...........................................

1-8

(

i i

CLINTON -1UNITL1 11' AmendmentNo. 40 j

1 INDEX i

.i OEFINITIONS i

SECTION

\\

PAGE 1.48 UNRESTRICTED AREA...............................................

i 1-9 1.49 0ELETED..........................................................

1-9 l

1.50 VENTING........................................................

1-9 Table 1.1 Surveillance Frequency Notation..........................

1-10 Table 1.2 Operational Conditions....................................

1-11 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

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

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

2-1 i

Reactor Coolant System Pressure................................

2-1 Reactor Vessel Water level.....................................

2-1

2. 2 LIMITING SAFETY SYSTEM SETTINGS

?

Reactor Protection System Instrumentation "

.aints............ 2 i

Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints..

2-3:

BASES i'

2.1 SAFETY LIMITS I

i l

Introduction...................................................

~B 2-1; I

THERMAL POWER, Low Pres sure or -Low Flow........................

8-2-1 THERMAL POWER,-High Pressure.and High Flow.....................

.B 2-2~

(

Reactor Coolant System Pressure................................

B 2.

DELETE0.~......................................................

18 2-3' i

CLINTON - UNIT.1 iii-Amendment No. 18,'40

INDEX BASES SECTION-PAGE SAFETY LIMITS (Continued)

DELETE0.........................................................

8 2-4 Reactor Vessel Water Level.....................................

B 2-5 2.2 LIMITING SAFETY SY' STEM SETTINGS i

Reactor Protection System. Instrumentation Setpoints............

B'2-6 3.0/4.0 LIMITINGCONDITIONSFOROPERATIONANDSURVEILLANCEREQUIREMENTS

)

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 ROOS Control Rod 0perability....................................

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

3/4 1-6.

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'l-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS' Control Rod Withdrawal.....................................

3/4 1-16 Rod Pattern Contrel Sy v4m.................................

3/4'l-17 l

.CLINTON - UNIT ~1 iv

. Amendment No.-18

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...........................................

3/4 3-58 Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instrumentation...........................................

3/4 3-59 Table 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation Setpoints.................................

3/4 3-61 Table 4.3.5.1-1 Reactor Core Isolation Cooling System Actuation Instrumentation Surveillance Requirements...........*:....

-3/4 3-62 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION.........................

3/4 3-63 Table 3.3.6-1 Control Rod Block Instrumentation....................

3/4 3-64 t

Table 3.3.6-2 Control Rod Block Instrumentation Setpoints..........

3/4 3-66 Table 4.3.6-1 Control Rod Block Instrumentation Surveillance Requirements..............................................

3/4 3-68 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation......................

3/4 3-70 Table 3.3.7.1-1 Radiation Monitoring Instrumentation...............

3/4 3-71 Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance R e q u i r eme n t s................. '............................. ~ 3 /4 3-7 3 Seismic Monitoring Instrumentation......................

3/4 3-75 Table 3.3.7.2-1 Seismic Monitoring Instrumentation.................

3/4 3-76 Table 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance Requirements..............................................

3/4 3-77 Meteorological Monitoring Instrumentation.................

3/4 3-78 Table 3.3.7.3-1 Meteorological Monitoring Instrumentation..........

3/4 3-79 Table 4.3.7.3-1 Meteorological Monitoring Instrumentation Surveillance Requirements.................................

3/4 3-80 Remote Shutdown Monitoring Instrumentation................

3/4 3-81 CLINTON - UNIT 1 vii

r INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

Table 3.3.7.4-1 Remote Shutdown Monitoring Instrumentation.........

3/4 3-82 Table 3.3.7.4-2 Remote Shutdown System Controls....................

3/4 3-83 Table 4.3.7.4-1 Remote Shutdown Monitoring Instrumentation Surveillance Requirements.................................

3/4 3-85 Accident Monitoring Instrumentation.......................

3/4 3-86 Table 3.3.7'.5-1 Accident Moni toring Ins trumentation................

3/4 3-87 Table 4.3.7.5-1 Accident Monitoring Instrumentation Surveillan Requirements.........................................ce 3/4 3-89 Source Range Mcnitors.....................................

3/4 3-90 Traversing In-Core Probe System...........................

3/4 3-91 Chlo ri ne De tect i on Sys tem.................................

3/4 3-92 Fire Detection Instrumentation (deleted)..................

3/4 3-93 Loose-Part Detection System...............................

3/4 3-94 Main Condenser Offgas Treatment S Monitoring Instrumentation.......ystem Explosive Gas 3/4 3-95 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM........................

3/4 3-96 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION....................

3/4 3-97 l

P CLINTON - UNIT 1 viii Amendment No. 40

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

Table 3.3.9-1 Plant Systems Actuation Instrumentation..............

3/4 3-99 Table 3.3.9-2 Plant Systems Actuation Instrumentation Setroints....

3/4 3-101 Table 4.3.9.1-1 Plant Systems Actuation Instrumentation Surveillance Requirements.................................

3/4 3-102 3/4.3.10 NUCLEAR SYSTEM PROTECTION SYSTEM - SELF TEST SYSTEM........

3/4 3-103 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.......................................

3/4 4-1 Figure 3,4.1.1-1 Thermal Power Versus Core Flow....................

3/4 4-5 Jet Pumps.................................................

3/4 4-6 Recirculation loop Flow...................................

3/4 4-7 Idle Recirculation Loop Startup........................-...

3/4 4-8 3/4.4.2 SAFETY VALVES Safety / Relief Valves......................................

3/4 4-9 Safety / Relief Valves low-Low Set Function................,

3/4 4-11 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................

3/4 4-12 Operational Leakage...............-.......................

3/4 4 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves...

3/4 4-15 3/4.4.4 CHEMISTRY.................................................

3/4 4-16 Table 3.4.4-1 Reactor Coolant Sys tem Chemi s try Limi ts..............

3/4 4-18 3/4.4.5 SPECIFIC ACTIVITY.........................................

3/4 4-19 Table 4.4.5-1 Primary Coolant Specific Activit Analysis Program....................y Sample and 3/4 4-21 CLINTON - UNIT 1 ix Amendment No. 40

INDEX l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.6 PRESSURE / TEMPERATURE LIMITS...............................

3/4 4-22 Figure 3.4.6.1-1 Minimum Reactor Vessel Metal Temp'erature Versus Reactor Vessel Pressure...................'................

3/4 4-24 Table 4.4.6.1-1 Reactor Vessel Material Surveillance Program-Withdrawal Schedule.......................................

3/4 4-25 Reactor Steam 00me........................................

3/4 4-26 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES..........................

3/4 4-27 3/4.4.8 STRUCTURAL INTEGRITY......................................

3/4 4-28 3/4.4.9 RESIOVAL HEAT REMOVAL Ho't Shutdown..............................................

3/4 4-29 Cold Shutdown.............................................

3/4 4-31 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - 0PERAlING..........................................

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

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

3/4 5-7 l

l 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity.............................

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

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

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

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

3/4 6-9 Containment Internal Pressure.............................

3/4 6-10 l

l CLINTON - UNIT 1 x

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRE SECTION PAGE CONTAINMENT SYSTEMS (Continued) l P r ima ry Conta inment Ave rage Ai r Tempe ra tu re...............

3/4 6-11 Containment Building Ventilation and Purge Systems........

l 3/4 6-12 3/4.6.2 DRYWELL Drywell Integrity.........................................

3/4 6-14 Drywell Bypass Leakage..,..................

3/4 6-15 Drywell Air Locks.........................................

3/4 6-16 Drywell Structural Integrity..............................

3/4 6-18 Drywell Internal Pressure.................................

3/4 6-19 Drywell Average Air Temperature...........................

3/4 6-20 Drywell Vent and Purge System.............................

3/4 6-21 3/4.6.3 DEPRESSURIZATION SYSTEMS Suppression Poo1..........................................

3/4 6-23 Containment Spray.........................................

3/4 6-26 Suppression Pool Cooling..................................

3/4 6-27 Suppression Pool Makeup System............................

3/4 6-28 3/4.6.4 CONTAINMENT ISOLATION VALVES..............................

3/4 6-29 Table 3.6.4-1 Containment Isolation Va1ves.........................

3/4 6-31 Table 3.6.4-2 Containment Isolation Trip Signals...................

3/4 6-62 3/4.6.5 DRYWELL POST-LOCA VACUUM RELIEFVALVES....................

3/4 6-63 3/4.6.6 SECONDARY CONTAINMENT Secondary Containment Integrity...........................

3/4 6-65 Figure 4.6.6.1-1 Secondary Containment Drawdown Tim for 1500 CFM Boundary Leakage.............e................

3/4 6-66a Secondary Containment Automatic Isolation Dampers........

3/4 6-67 CLINTON - UNIT 1 xi Amendment No. 21

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRDtENTS SECTION PAGE CONTAINMENT SYSTEMS (Continued)

Table 3.6.6.2-1 Secondary Containment Ventilation System Automatic Isolation Dampers...............................

3/4 6-69 Standby Gas Treatment System..............................

3/4 6-70 3/4.6.7 ATMOSPHERE CONTROL Containment Hydrogen Recombiner Systems...................

3/4 6-73 Containment /Drywell Hydrogen Mixing System................

3/4 6-75 Primary Containment /Drywell Hydrogen Ignition System......

3/4 6-76 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS Shutdown Service Water System (Loops A, B, C).............

3/4 7-1 Ultimate Heat Sink........................................

3/4 7-2 3/4.7.2 CONTROL ROOM VENTILATION SYSTEM...........................

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

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

3/4 7-9 Figure 4.7.4-1 Sample Plan for Snubber Functional Test.............

3/4 7-14 3/4.7.5 SEALE0 SOURCE CONTAMINATION...............................

3/4 7-15 3/4.7.6 MAIN TURBINE BYPASS SYSTEM................................

3/4 7-17 3/4.7.7 LIQUID STORAGE TANXS......................................

3/4 7-18 3/4.7.8 MAIN CONDENSER OFFGAS MONITORING Offgas - Explosive Gas Mixture............................

3/4 7-19 Offgas - Noble Gas Radioactivity Rate.....................

3/4 7-20 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 AC SOURCES AC Sources - Operating....................................

3/4 8-1 fstm7m mR21 w

o el INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS (Continued)

Table 4.8.1.1.2-1 Diesel Generator Test Schedule..........

3/4 8-10 AC Sources - Shutdown.....................................

3/4 8-11 3/4.8.2 OC SOURCEr DC Sources - Operating....................................

3/4 8-12 Table 4.8.2.1-1 Battery Surveillance Requirements..................

3/4 8-15 OC Sources - Shutdown.....................................

3/4 8-16 3/4.8.3 ONSITE POWER O!STRIBUTION SYSTEMS Distribution - Operating..................................

3/4 8-17 D i stri buti on - Shutdown...................................3/4 8-21 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices.....................

3/4 8-24 Table 3.8.4.1-1 Containment Penetration Conductor O l

Protective 0evices........................vercurrent 3/4 8-26 Motor Operated Valves Thermal Overload Protection.........

3/4 8-45 Table 3.8.4.2-1 Motor Operated Valves Thermal Overload Protection..

3/4 8-46 Reactor Protection System Electric Power Monitoring.......

3/4 8-53 l

3/4.9 REFUELING OPERATIONS 3/*.9.1 REACTOR MODE SWITCH.......................................

3/4 9-1 1

3/4.9.2 INSTRUMENTATION...........................................

3/4 9-3 3/4.9.3 CONTROL R0D P0SITION......................................

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

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

3/4 9-7 CLINTON - UNIT 1 xiii Amendment No. 40

h 4

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS (Continued) 3/4.9.6 FUEL HANDLING EQUIPMENT R e f u e l i n g P l a t f o rm........................................

3/4 9-8 Auxiliary Platform........................................

3/4 9-9 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL, UPPER CONTAINMENT FUEL POOL, AND NEW FUEL STORAGE VAULT.....................

3/4 9-10 3/4.9.8 WATER LEVEL - REACTOR VESSEL..............................

3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE AND UPPER CONTAINMENT FUEL P00LS................................................

3/4 9-12 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Remova1................................

3/4 9-13 Multiple Control Rod Removal..............................

3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.............

3/4 9-17 Low Water Level...........................................

3/4 9-18 t

l 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM.............................

3/4 9-19 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /ORYWELL INTEGRITY...........-

3/4 10-1 3/4.10.2 R00 PATTERN CONTROL SYSTEM................................

3/4 10-2 3/4.10.3 SHUTOOWN MARGIN DEMONSTRATIONS............................

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

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

3/4 10-5 3/4.10.6 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING............

3/4 10-6 CLINTON - UNIT 1 xiv Amendment No. 40

i l

8.

INDEX l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 3/411-1 thru 20 DELETED 3/4.12 3/4 12-1 thru 15 DELETED 9

l CLINTON - UNIT 1 xv Amendment No. 40 1

\\

INDEX

(

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 1

BASES 3/4.0 APPLICABILITY................................................

B 3/4 0-;

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 R005..............................................

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

B 3/4 2-1 3/4.2.2 APRM SETP0!NTS............................................

B 3/4 2-2 Bases Table B 3.2.1-1 Significant Input Parameters to the Loss-of-Coolant Accident Analysis..............................

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

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

B 3/4 2-5 Bases Figure B 3/4.2.3-1 Reactor Operating Ma Reci rculation Loop Opera tion............. p for Two B 3/4 2-7 l

l Bases Figure B 3/4.2.3-2 Reactor Operating Map for Sin Recirculation Loop 0peration......................gle B 3/4 2-8 l

CLINTON - UNIT 1 xvi Amendment flo 18

INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.................

B 3/4 3-1 3/4.3.2 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM...

B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION... ~B 3/4 3-3 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......................................<.3..

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

B 3/4 3-5 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.....................

B 3/4 3-5 Seismic Monitoring Instrumentation........................

B 3/4 3-5 Meteorological Monitoring Instrumentation.................

B 3/4 3-5 Remote Shutdown Monitoring Instrumentation................

B 3/4 3-6 Accident Monitoring Instrumentation.......................

B 3/4 3-6 Source Range Monitors.....................................

B 3/4 3-6 Traversing In-Core Probe System...........................

B 3/4 3-6 Chlorine Detection System.................................

B 3/4 3-7 Fire Detection Instrumentation (deleted)..................

B 3/4 3-7 Loose-Part Detection System...............................

B 3/4 3-7 CLINTON - UNIT 1 xvii

+

t INDEX BASES t

SECTION PAGE INSTRUMENTATION (Continued)

Main Condenser Offgas Treatment System Explosive Gas Monitoring Instrumentation............................

B 3/4 3-8 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM.......................

B 3/4 3-8 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION...................

B 3/4 3-8 3/4 3.10 NUCLEAR SYSTEM PROTECTION SYSTEM - SELF TEST SYSTEM.......

B 3/4 3-9 Bases Figure B 3/4.3-1 Reactor Vessel Water Level..................

B 3/4 3-10 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RiCiRCULATIONSYSTEM......................................

B 3/4 4-1 3/4.4.2 3AFETY/ RELIEF VALVES.......................

B 3/4 4-3 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................

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

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

B 3/4 4-4 3/4.4,5 SPECIFIC ACTIVITY.........................................

B 3/4 4-4 1

3/4.4.6 PRESSURE / TEMPERATURE LIMITS...............................

B 3/4 4-5 3/4.4.7 MAIN STEAM ' THE ISOLATION VALVES..........................B 3/4 4-6 3/4.4.8 STRUCTURAL I, IGRITY......................................

B 3/4 4-7 3/4.4.9 RESIDUAL HEAT REM 0 VAL.....................................

B 3/4 4-7 Bases Table B 3/4.4.6-1 Reactor Vessel Toughness Values............

B 3/4 4-8 Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>l Mev at k T-as a Function of Service Life....................)............

B 3/4 4-10 Figure B 3/4.4.6-2 Predicted Adjustment of Reference Temperature, "A", as a Function of Fluence-and Copper Content per NRC l

Regulatory Guide 1.99.....................................

B 3/4 4-11 l

l CLINTON - UNIT 1 xviii Amendment No. 40'

INDEX L

BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS l

l 3/4.5.1 l

and ECCS - OPERATING AND SHUT 00WN.............................

B 3/4 5-1 l

3/4.5.2 l

3/4.5.3 SUPPRESSION P00L..........................................

B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity..............................

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

B 3/4 6-1 Primary Containment Air Locks.............................

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

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

B 3/4 6-2 Containment Internal Pressure........................~

B 3/4 6-2 i

Primary Containment Average Air Temperature...............

B 3/4 6-2 Containment Building Ventilation and Purge Systems........

B 3/4 6-3 3/4.6.2 ORYWELL Drywell Integrity.........................................

B 3/4 6-4 l

Drywell Bypass Leakage....................................

B 3/4 6-4 Drywell Air Locks.........................................

B 3/4 6-4 Drywell Structural Integrity..............................

B 3/4 6-5 l

D rywel l I n te rna l - P re s s u re.................................

B 3/4 6 l Drywell Average A'ir Temperature...........................

B 3/4 6-5 l

Drywell Vent and Purge....................................

B'3/4 6-5 3/4.6.3 DEPRESSURIZATION SYSTEMS..................................

B 3/4 6-6 3/4.6.4 CONTAINMENT ISOLATION VALVES..............................

B 3/4 6-7 3/4.6.5 DRYWELL POST-LOCA VACUUM RELIEF VALVES....................

B 3/4 6-8 CLINTON - UNIT 1 xix

INDEX BASES SECTION g

CONTAINMENT SYSTEMS (Continued) 3/4.6.6 SECONDARY CONTAINMENT.....................................'

B 3/4 6-8 3/4.6.7 ATMOSPHERE CONTR0L........................................

3 3/4 6-8 3/4.7 PLANT SYSTEMS 3/4.7.1 SHbT00WN SERVICE WATER SYSTEM.............................

B 3/4 7-1 3/4.7.2 CONTROL ROOM VENTILATION SYSTEM............................

B 3/4 7-1 i

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

B 3/4.7-1 3/4.7.4 SNUBBERS..................................................

B 3/4 7-2 3/4.7.5 SEALED SOURCE C0NTA.MINATION...............................

B 3/4 7-3 3/4.7.6 MAIN TURBINE BYPASS SYSTEM................................

B 3/4 7-3 i

3/4.7.7 LIQUID STORAGE TANKS......................................

B 3/4 7-3 I

3/4.7.8 MAIN CONDENSER OFFGAS MONITORING l

Offgas - Explosive Gas Mixture............................

B 3/4 7-3 Offgas - Noble Gas Radioactivity Rate.....................

B 3/4 7-4 l

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4 8.2, AC SOURCES, DC SOURCES, AND ONSITE POWER I

and DISTRIBUTION SYSTEMS......................................

B 3/4 8-1 l

3/4.8.3 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES...................

B 3/4 8-2 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 R00 P0SITION......................................

B 3/4 9-1 CLINTON - UNIT 1 xx Amendment No. 40

INDEX BASES i

SECTION PAGE REFUELING OPERATIONS (Continued) 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 FUEL HANDLING EQUIPMENT...................................

B 3/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT AND NEW FUEL STORAGE AND UPP CONTAINMtNT FUEL P00LS...........................ER B 3/4 9-2 3/4.9.8 WATER LEVEL - REACTOR VESSEL and WATER and LEVEL - SPENT FUEL STORAGE AND UPP 3/4.9.9 FUEL P00LS........................ER CONTAINMENT 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 AP? COOLANT CIRCULATION.............

B 3/4 9-2 l

l 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM.............................

B 3/4 9-3 1

l 3/4.10 SPECIAL TEST EXCEPTICNS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /DRYWELL INTEGRITY...........

B 3/4 10-1 3/4.10.2 R00 PATTERN CONTROL SYSTEM................................

B 3/4 10-1 i

1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS............................

B 3/4 10-1 3/4.10.4 RECIRCULATION l00PS.......................................

B 3/4 10-1 l

3/4.10.5 TRAINING STARTUPS.........................................

B 3/4 10-1 3/4.10.6 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING............B 3/4 10-1 3/4.11 Deleted CLINTON - UNIT 1 xxi Amendment No. 40

INDEX BASES SECTION PAGE J/4.12 Deleted i

5.0 OESIGN FEATURES

_~

~

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

5-1 i

Low Population Zone............................................

5-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents.......................

5-1 5.2 CONTAINMENT Configur'ation..................................................

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

5-1 Figure 5.1.1-1 Exclusion Area......................................

5-2 Figure 5.1.2-1 Low Population Zone.................................

5-3 Figure 5.1.3-1 Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents..................

5-4 i

l Secondary Containment..........................................

5-5 t

k CLINTON - UNIT 1 xxii Amendment No. 40

1 o

l INDEX ADMINISTRATIVE CONTROLS l

_SECTION PAGE DEVIEW AND AUDIT (Continued)

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

6-10 Quorum.......................................................

6-11 l

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

6-11 Audits.......................................................

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

6-13 Records......................................................

6-13 6.5.3 TECHN] CAL REVIEW AND CONTR0L.................................

6-13 Activities...................................................

6-13

6. 6 REPORTABLE EVENT ACTI0N........................................

6-14

6. 7 SAFETY LIMIT VIOLATION.........................................

6-14 6.8 PROCEDURES AND PR0 GRAMS........................................

6-15 6.8.1 PROCEDURES.................................

6-15 REVIEWANDAPPRbVAL..........................................

6.8.2 5-15 6.8.3 TEMPORARY CHANGES............................................

6-15

6. 8. 4 PR0 GRAMS.....................................................

6-16 l

p

6. 9 REPORTING REQUIREMENTS.........................................

6-17 l

6. S.1 ROUTINE REP 0RTS..............................................

6-17 Startup Report...............................................

6-17 Annual Reports...............................................

6-17 Annual Radiological Environmental Operating Report...........

6-18 Semiannual Radioactive Effluent Release Report...............

6-19 Monthly Operating Reports....................................

6-21 Core Operating Limits Report.................................

6-21 l

CLINTON - UNIT 1 xxv Amendment Mq. 28

0 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE REPORTING REQUIREMENTS (Continued)

'PECIAL REPORTS.............................................

6-21 6.9.2 S

6.10 RECORD RETENTION..............................................

6-21 6.11 RADIATION PROTECTION PR0 GRAM..................................

6-23 6.12 HIGH RADIATION AREA...........................................

6-23 6.13 PROCESS CONTROL PROGRAM (PCP).................................

6-24 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM)........................

6-25 i

i l

CLINTON - UNIT 1 xxvi Amendment No. 40

DEFINITIONS ORYWELL INTEGRITY (Continued) t d.

The drywell leakage rates are within the limits of Specification 3.6.2.2.

The suppression pool is OPERABLE pursuant to Specification 3.6.3.1.

e.

f.

The sealing mechanism associated with each drywell penetration, e.g.,

i welds, bellows or 0 rings, is OPERABLE.

E - AVERAGE DISINTEGRATION ENERGY 1.13 I shall be the average, w6ighted in proportion to the concentration of l

each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.14 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME I

interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function; i.e., the valves travel to their required positions, pump dis-charge pressures reach their required values, etc.

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

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

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.15 The END-OF-CYCLE RECIRCULATION PUMP TRIP ~ SYSTEM RESPON l

that time interval to complete suppression of the electric are between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a.

Turbine stop valves and b.

Turbine control valves.

The response time may be measured by any series of sequential, overlapping or:

total steps such that the entire response time is measured.

1.16 (DELETED)

CLINTON - UNIT 1 1-3 Amendment No.18, 26

DEFI_NITIONS i

FREQUENCY NOTATION 1.17 The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined in-Table 1.1.

1.18 DELETED l

IDENTIFIED LEAKAGE 1.19 IDENTIFIED LEAKAGE shall be:

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

that is captured and conducted to a sump or collecting tank or b.

Leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

LIMITING CONTROL R00 PATTERN 1.20 A LIMITING CONTROL ROD PATTERN'shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

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 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 per-formed by any series of sequential, overlapping er iotal system steps such that the entire logic system is tested.

s 1.23 [0ELETE0]

MEMBER (S) 0F THE PUBLIC 1.24 MEMBER (S) 0F THE PUBLIC shall include all persons who-are tu t occupationally associated with the-plant.

This category does not include employees of the licensee, its contractors or vendors. -Also excluded from this category are

?

i l

l l

l CLINTON - UNIT 1 1-4 Amendment No.JB,40

'l

s DEFINITIONS 1

i MEMBER (S) 0F THE PUBLIC (Ccntinued) 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 rot associated with the plant.

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

OFFSITE DOSE CALCULATION MANUAL (0DCM) 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 l

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 Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

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

OPERATIONAL CONDITION - CONDITION 1.28 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive com-bination 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 nuclear characteristics of the reactor core and related instrumentation as 1) described in Chapter 14 of the FSAR, 2) as authorized under the provisions of.

10 CFR 50.59, or 3) as otherwise approved by the Commission.

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

CLINTON - UNIT 1 1-5 Amendment No. 40

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

All containment penetrations required to be closed during accident condi-a.

tions are either:

1.

Capable of being closed by an OPERABLE containment automatic isolation I

system or 2.

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

All containment equipment hatches are closed and sealed.

Each containment air lock is in compliance with the requirements _of c.

Specification 3.6.1,3.

d.

The containment leakage rates are within the limits of Specification 3,6.1.2.

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

I Specification 3.6.3.1.

f.

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

PROCESS CONTROL PROGRAM (PCP) 1.32 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the 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 Part 20, 10 CFR Part 61, 10 CFR Part 71 and Federal and State regulations, burial ground requirements and other require-ments governing the disposal of the radioactive waste.

PURGE - PURGING

^

1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.34 RATED THERMAL POWER shall be a total reactor core heat transfer rate to I

the reactor coolant of 2894 MWt.

CLINTON - UNIT 1 1-6 Amendment No. 40

OEFINITIONS l

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

The response time may be measured by any series of sequential, overlapping or total steps such l

that the entire response time is measured, l

REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in t

Section 50.73 to 10 CFR Part 50.

ROD DENSITY 1.37 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches.

All rods fully inserted are equivalent to 100% R00 DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.38 SECONDARY CONTAINMENT INTEGRITY shall exist when:

All secondary containment penetrations required to be closed during a.

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 valve or damper as applicable secured in its closed posi--

tion, except as provided in Table 3.6.6.2-1 of Specification 3.6.6.2.

1 L

b.

All secondary containment equ;pment hatches are closed and sealed.

The standby gas treatment system is in compliance with the requirements c.

of Speci fication 3.6.6.3.

d.

At least one door in each access to the secondary containment is closed, except for normal entry and exit.

The sealing mechanism associated with each secondary containment penetra-e.

tion, e.g., welds, bellows or 0 rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or equal to the j

value required by Specification 4.6.6.1.a.

SELF TEST SYSTEM 1.39 The SELF TEST SYSTEM (STS) shall be that automatic test system designed to continually monitor the solid state nuclear system protection system (NSPS) functional circuitry by injecting short-duration pulses into circuits and verifying proper circuit response to various input combinations.

The SELF TEST CLINTON - UNIT 1 1-7 Amendment No. 40

l F

l DEFINITIONS I

j SELF TEST SYSTEM (Continued)

SYSTEM is designed to maintain surveillance over all NSPS cabinet circuitry essential to the Reactor Protection System, Emergency Core Cooling Systems, Reactor Core Isolation Cooling System, and the Nuclear Steam Supply Shutoff System on a continuous cyclic basis.

l The SELF TEST SYSTEM may be used to perform various surveillance testing functions to satisfy technical specifications requirements for those components I

it is designed to monitor.

The STS may be used to augment conventional testing methods to perform CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, CHANNEL CALIBRA-TIONS, RESPONSE TIME TESTS AND LOGIC SYSTEM FUNCTIONAL TESTS provided that OPERABILITY of the STS has first been verified.

SHUT 00WN MARGIN 1.40 SHUT 00WN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical assuming all control rods are fully inserted, except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn, and the reactor is in the shutdown condition; cold, i.e.,

68*F; and xenon-free.

_ SITE BOUNDARY 1.41 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, 'or otherwise controlled by the licensee.

1.42 DELETED 1.43 DELETED I

STAGGERED TEST BASIS 1.44 A STAGGERED TEST BASIS shall consist of:

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

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

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

THERMAL POWER 1.45 THERMAL POWER shall be the total reactor core heat transfer rate to the l

reactor coolant.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.46 The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two separate time intervals:

a) time from initial movement of the main turbine stop valve or control valve until-80% of turbine bypass capacity is established, and b) the time from initial movement of the main turbine stop valve or control valve CLINTON - UNIT 1 1-8 Amendment No. 40 I

y

-- m

O 1

OEFINITIONS TURBINE BYPASS SYSTEM RESPONSE TIME (Continued) until initial movement of the turbine bypass valve.

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

UNIDENTIFIED LEAKAGE 1,47 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

UNRESTRICTED AREA 1

1.48 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY i

access to which is not controlled by the licensee for purpo es of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

1.49 DELETED VENTING 1.50 VENTING is the controlled process of discharging air or gas from a con-finement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING.

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

l l

L CLINTON - UNIT 1 1-9 Amendment No. 40

TABLE 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FRE0t'ENCY S

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 per 184 days.

A At least once per 366 daysi-R At least once per 18 months (550 days).

P Prior to each release.

S/V Prior to each reactor startup.

NA Not applicable.

l 4

4 9

I CLINTON - UNIT 1 1-10

~'

l TABLE 4.3.6-1 (Continued)

CONTROL R00 BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS l

TABLE NOTATIONS

_(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

(c) Within one hour prior to control rod movement, unless performed within ti e previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and as each power range above the RPCS low power setpoint is entered.for the first time during any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period during power increase or decrease.

(d) At least once per 31 days while operation continues within a given power range above the RPCS low power setpoint.

(e)

Includes reactor. manual control multiplexing system input.

(f) Calibrate the analog trip module at least once per 31 days.

With any control rod withdrawn.

Not applicable to control rods removed per. Specification 3.9.10.1 or 3.9.10.2.

L l

l CLINTON - UNIT 1 3/4 3-69

-0 a.

INSTRUMENTATION 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION

)

LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE, with their alarm / trip setpoints within the specified limits.

APPLICABILITY:

As shown in Table 3.3.7.1-1.

ACT:0N:

With a radiation monitoring instrumentation channel alarm / trip setpoint a.

exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the-limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b.

With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.7.1-1, The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIkEMENTS 4.3.7.1.Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the l

conditions and at the frequencies shown in Table 4.3.7.1-1.

e CLINTON - UNIT 1 3/4 3-70 Amendment' No.40 i

n TABLE 3.3.7.1-1

.C RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM / TRIP g

INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION 1.

Main Control Room 2/ intake *)

1, 2, 3, 5, and

5 10 mR/hr 70 I

Air Intake Radiation Monitor 2.

Area Monitors a.

New Fuel 1

1 2.5 mR/hr**

71 Storage Vault b.

Spent Fuel 1

< 2.5 mR/hr**

71 Storage Pool c.

Control Room 1

At all times 5 2.5 mR/hr**

71 Direct Radiation w,',

Monitor w

B*

O 8

o 4

f*

TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS When irradiated fuel is being handled in the secondary containment.

Alarm only.

I

With fuel in the new fuel storage vault.

1. With irradiated fuel in the spent fuel storage pool.

1 I

(a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for re-i quired surveillance without placing the trip system in the tripped condi-tion provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

l ACTION ACTION 70 -

a.

With one of the required monitors inoperable, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months initiate and maintain operation of the control room emerge,ncy filtration system in the high radiation mode of operation.

b.

With both of the required monitors inoperable, initiate and maintain operation of the control room emergency filtration system in the high radiation mode of operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 71 -

With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i CLINTON - UNIT 1 3/4 3-72 Amendment NoJ6,40 i

o TABLE 4.3.7.1-1 RADIATION M9NITORING INSTPUMENTATION SURVEILLANCE REQUIREMENTS c

CHANNEL 3

CHANNEL FUNCTIONAL CHANNEL

[ INSTRUMENTATION CHECK TEST i;ALIBRATION APPLICABILITY 1.

Main Control Room Air S

M R

1, 2, 3, 5, and

Intake Radiation Monitor 2.

Area Monitors a.

New Fuel Storage 5

M R

Vault y

b.

Spent Fuel Storage 5

M R

Pool O

c.

Control Room Direct S

M R

At all times Radiation Monitor 3n

O sa TABLE 4.3.7.1-1(Continued]

RADIATION HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION

When irradiated fuel is being handled in the secondary containment.

I

With fuel in the new fuel storage vault.

1. With irradiated fuel in the spent fuel storage pool.

h t

CLINTON - UNIT 1 3/4 3-74 Amendment No)D 40

i

\\

INSTRUMENTATION MAIN CONDENSER OFFGAS TREATHENT SYSTEM EXPLOSIVE GAS MONITORING IN l

l LIMITING CONDITION FOR OPERATION 3.3.7.11 At least one main condenser offgas treatment system explosive gas monitoring instrumentation channel shall be OPERABLE with its alarm / trip setpoint set to ensure that the limits of Specification 3.7.8.1 are not exceeded.

APPLICABILITY:

Duringoperationofthemaincondenserairejector.

{

ACTION:

With the explosive gas monitoring instrumentation channel alarm / trip a.

setpoint less conservative than required by Specification 3.7.8.1, declare the channel inoperable and take the ACTION required below, b.

With less than 1 explosive gas monitoring instrumentation channel OPERABLE, operation of the main condenser offgas treatment system may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Restore the inoperable channel to OPERABLE status within 30 days and, if unsuc-cessful, prepare and submit a Special Report to the Commission l

pursuant to Specification 6.9.2 to ex was not corrected in a timely manner, plain why this inoperability The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.3.7.11 The explosive gas monitoring instrumentation channel (s) 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 , a CHANNEL CALIBRATION

at least once per 92 days and a CHANNEL FUNCTIONAL TEST at least once per 31 days.

The CHANNEL CALIBRATION shall include the use of standard samples containing a nominal.

1.

1.0 vol. % hydrogen, balance nitrogen, and 2.

4.0 vol. % hydrogen, balance nitrogen.

CLINTON - UNIT 1 3/4 3-95 Amendment No. 40

J INSTRUMENTATION 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.8 At least one turbine overspeed protection system shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one turbine control valve, one turbine stop valve and/or with one

[

a.

turbine combined intermediate valve inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or close at least one valve in the affected steam lead or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

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

The provisions of Specification 3.0.4 are not applicable for entry into c.

OPERATIONAL CONDITION 2 provided the turbine is tripped and the turbine is isolated from the main steam supply by closed turbine stop and control valves.

SURVEILLANCE REQUIREMENTS 4.3.8.1 The provisions of Specification 4.0.4 are not applicable.

4.3.8.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE:

i At least once per 7 days by cycling each of the following valves from the I

a.

l running position and observing valve closure:

1.

Four low pressure turbine combined intermediate valves, and 2.

Four high pressure turbine stop valves.

b.

At least once per 31 days by cycling each of the four high pressure turbine control valves from the running position and observing valve closure, i

At least once per 18 months by performanca of a CHANNEL' CALIBRATION of c.

the turbine overspeed protection instrumentation.

d.

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

If unacceptable flaws or excessive corrosion are found, all l

other valves of that type shall be inspected.

r 1

1 CLINTON - UNIT 1 3/4 3-96 Amendment No.40 l

INSTRUMENTATION 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The plant systems actuation instrumentation channels shown in Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9 2.

APPLICABILITY:

As shown in Table 3.3.9 1.

ACTION:

With a plant system actuation instrumentation channel trip setpoint less a.

conservative than the value shown in the Allowable Value column of Table 3.3.9-2, 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 / loop inoperable.

b.

For the containment spray system:

1.

With the number of'0PERABLE channels less than required by the Minimum OPERABLE Channels per containment spray loop requirement for one containment spray loop, place at least one inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or declare the associated loop inoperable.

2.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per containment spray loop requirement for both loops, declare the associated loop inoperable, For the feedwater system / main turbine trip system:

c.

1.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels 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 .

2.

With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels 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 .

d.

For the suppression pool makeup system (SPMS) with the number of channels, one less than the minimum OPERABLE Channels Requirement of Table 3.3.9-1 take the ACTION required by Table 3.3.9-1.

I CLINTON - UNIT 1 3/4 3-97 Amendment No. 40 l

9 INSTRUMENTATION PLANT SYSTEMS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.9.1 Each plant system actuation instrumi' ction r,hannel shall be demon-strated OPERABLE by the performance of th1 c'iiNNEL ' HECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.9.1-1 4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS shall be performed at least once per 18 months.

Each trip system or division of the plant system actuation logic associated with the Nuclear System Protection System shall be manually tested independent of the SELF TEST SYSTEM during separate refueling outages such that all divisions and all trip functions are tested at least once every four fuel cycles.*

Manual testing for the purpose of satisfying Specification 4.3.9.2 is not required until af ter shutdown during the first regularly scheduled refueling outage.

CLINTON - UNIT 1 3/4 3-98 Amendment No. 40 l

n TABLE 3.3.9-1 c-E PLANT SYSTEMS ACTUATION INSTRUMENTATION a

MINIMUM OPERA 8LE CHANNELS PER APPLICA8LE E

CONTAINMENT OPERATIONAL Q TRIP FUNCTION SPRAY LOOP CONDITIONS

~

1.

CONTAINMENT SPRAY SYSTEM a.

Drywell Pressure-High 2*

1, 2, 3 b.

Containment Pressure-High 2*

1,2,3 Reactor Vessel Water Level-tow Low Low, Level 1 2*

1, 2, 3 c.

d.

Timers i

(1) Loop A, Loop B (10 minutes) 1 1,2,3 (2) Loop B only (90 seconds) 1 1, 2, 3 R

e.

Manual Initiation 1

1,2,3 i

y MINIMUM OPERABLE 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM a.

Reactor Vessel Water Level-High, Level 8 3*

1 MINIMUM OPERABLE 3.

SUPPRESSION POOL MAKEUP SYSTEM CHANNELS PER TRIP SYSTEM ACTION a.

Drywell Pressure-High 2*

1, 2, 3 50 b.

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

1, 2, 3 50 1

c.

Suppression Pool Water Level-Low Low 2*

1,2,3 51 d.

Suppression Pool Makeup Timer 1

1, 2, 3 51 y

e.

SPMS Manual Initiair. ion 2

1,2,3 51 g

f.

SPMS Mode Switch Permissive 1

1,2,3 51

&g 4

CA channel may be placed in an' inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without 2

- placing the trip system in the tripped condition provided at least one other OPERABLE channel in o

g the same trip system is monitoring that parameter.

m

+

TABLE 3.3.9-1 (Continued)

PLANT SYSTEMS ACTUATION INSTRUMENTATION ACTION ACTION 50 -

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

For 1 trip system, place the inoperable channel (s) and/or a.

that trip system in the tripped condition within one hour or declare the SPMS inoperable and take the action of Specification 3.6.3.4.

b.

For both triD systems, declare the SPMS inoperable and take the action of Specification 3.6.3.4.

ACTION 51 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip System requirement, declare the SPMS system inoperable and take tre action of Specification 3.6.3.4.

i l

1 CLINTON - UNIT 1 3/4 3-100-Amendment No. 40 l

4 p

TABLE 3.3.9-2 PLANT SYSTEMS ACTUATION INSTRUMENTATION SETPOINIS c:

ALLOWABLE

} TRIP FUNCTION TRIP SETPOINT VALUE 1.

CONTAINMENT SPRAY SYSTEM a.

Drywell Pressure-High 1 1.68 psig

$ 1.88 psig b.

Containment Pressure-High

< 22.3 psia

< 22.4 psia c.

Reactor Vessel Water Level-Low Low Low, Level 1 d.

Timers 145.5 in.*

i-147.7 in.

.1.

Loop A, Loop 8 10.17 min.

> 10.10 < 10.23 min.

2.

Loop B only 5 90 sec.

5 90.6 sec.

w 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM N

w a.

Reactur Vessel Water Level-High, Level 8

$ 52.0 in.*

$ 52.6 in.

h 3.

SUPPRESSION POOL MAKEUP SYSTEM

-a.

Drywell Pressure-High

$ 1.68 psig i 1.88 psig b.

Reactor Vessel Water Level-Low Low Low, Level 1 3 -145.5 in.*

1 -147.7 in.

c.

Suppression Pool Water Level-Low Low 3 37 9/16 in.**

1 29 in.**

d.

Suppression Pool Makeup Timer

$ 25 minutes i 30 minutes e.

SPMS Manual Initiation NA NA l

a o.

3 f.

SPMS Mode Switch Permissive NA NA E

F

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

~

8 " Instrument zero is 727'-0" as1.

=

l TABLE 4.3.9.1-1 n

r-5 PLANT SYSTEMS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

c CHANNEL OPERATIONAL i

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH H

TRIP FUNCTION NECK TEST CALIBRATION SURVEILLANCE REQUIRED I

w 1.

CONTAINMENT SPRAY SYSTEM a.

Drywell Pressure-High S

M R

1, 2, 3 b.

Containment Pressure-High S

M R

1,2,3 c.

Reactor Vessel Water Level-Low IR ')

Low Low, Level 1 5

M 1, 2, 3 d.

Timers MA M

R 1, 2, 3 e.

Manual Initiation NA R

NA 1, 2, 3 2.

FEEDWATER SYSTEM / MAIN 1DRBINE TRIP SYSTEM I T Reactor Vessel Water Level-High, a.

y Level 8 S

M R

1 3.

SUPPRESSION POOL MAKEUP IR *)

a.

Drywell Pressure-High S

M 1, 2, 3 b.

Reactor Vessel Water Level -

Low Low Low, Level 1 S

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

Suppression Pool Water Level-Low Low S M

R(b) 1, 2, 3

[

d.

Suppression Pool Makeup Timer NA M

Q 1,2,3 e.

SPMS Manual Initiation MA R

NA 1,2,3 f.

SPMS Mode Switch Permissive NA R

NA 1,2,3

$ (c) Calibrate the analog trip module at least once every 31 days.

(b) Calibrate the analog comparitor unit at least once every 31 days.

~

w

INSTRUMENTATION 3/4.3.10 NUCLEAR SYSTEM PROTECTION SYSTEM - SELF TEST SYSTEM LIMITING CONDITION FOR OPERATION 3.3.10 The SELF TEST SYSTEM (STS) of the Nuclear System Protection System shall be OPERABLE and operating in the fully automatic mode.*

APPt.*CABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.

ACTION:

With the STS not operating in the fully automatic mode initiate corrective a.

action within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to restore the STS to automatic operation for the maximum number of divisions available, b.

If the STS is not restored to fully automatic operation within 30 days, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , The provisions of Specification 3.0.4 cre not applicable.

c.

SURVEILLANCE REQUIREMENTS:

4.3.10 Status indications of the STS shall be obtained at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , whenever the STS is operating in the fully or partially automatic mode.

I 1

The STS may be periodically taken out of the fully automatic mode of operation for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing surveillance testing and pre-ventative or corrective maintenance to satisfy technical specification requirements for those components the STS is designed to monitor.

CLINTON - UNIT 1 3/4 3-103 Amendment No. 40 l

PLANT SYSTEMS l

_3/4.7.6 MAIN TURBINE BYPASS SYSTEM 11gITINGCONDITIONFOROPERATION 3.7.6 The main turbine bypass system shall be OPERA 8LE.

APPLICABILITY:

OPERATIONAL CON 0! TION 1 with THERMAL POWER greater than or equal to 25% of RATED THEMAL POWER.

ACTION:

With the main turbine bypass system inoperable, restore the system to OPERABLE POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or reduce SURVEILLANCE REQUIREMENTS 4.7.6 The main turbine bypass system shall be demonstrated OPERABLE:

At least once per 31 days by cycling each turbine bypass valve through at a.

ienst one complete cycle of full travel, and j

b.

At least once per 18 months by:

1.

Performing a system functional test which includes simulated auto-matic actuation and verifying that each automatic valve actuates to its correct position.

2.

Demonstrating TUR8INE BYPASS SYSTEM RESPONSE TIME meets the requirements when measured from the initial movement of the main turbine stop or control valve:

a) 80% of turbine bypass system capacity shall be established in 1 0.3 seconds.

b)

Bypass valve opening shall start in 1 0.1 seconds.

I e

CLINTON - UNIT 1 3/4 7-17

PLANT SYSTEMS 3/4.7.7 LIOUID STORAGE TANKS

LIMITING CONDITION FOR OPERATION 3.7.7 The quantity of radioactive material contained in each of the following l

unprotected outdoor tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases, a.

Cycled Condensate Storage Tank l

b.

RCIC Storage Tank c.

Outside temporary tank APPLICABILITY:

At all times.

ACTION:

With the quantity of radioactive material in any of the above listed tanks a.

exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.7 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.

" Tanks included in this specification are those cetdoor tanks that are not surrounded by liners, dikes, or walls capable of hciding the tank contents and that do not have tank overflows and surrounding aren drains connected to the liquid radwaste treatment system.

CLINTON - UNIT 1 3/4 7-18 Amendment No. 40 l

PLANT SYSTEMS 3/4.7.8 MAIN CONDENSER OFFGAS MONITORING OFFGAS-EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.7.8.1 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:

Whenever the main condenser air ejector is in operation.

ACTION:

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

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 provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.8.1 The concentration of hydro 0en in the main condenser offgas treatment l

system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system whenever the main condenser evacuation system is in operatio) with the hydrogen monitors required OPERABLE by Specification 3.3.7.12.

l l

{

1 l

CLINTON - UNIT 1 3/4 7-19 Amendment No. 40 l

PLANT SYSTEMS 0FFGAS-NOBLE GAS RAD 70 ACTIVITY RATE LIMITING CONDITION FOR OPERATION 3.7.8.2 The radioactivity rate of noble gases measured at the offgas recom-biner effluent shall be limited to less than or equal to 289 millicuries /sec after 30 minutes decay.

APPLICABILITY:

Duringoperationofthemaincondenserairejector.

ACTION With the radioactivity rate of noble gases at the offgas recombiner effluent exceeding 289 militcuries per second after 30 minutes decay, restore the gross radioactivity 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 HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I SURVEILLANCE REQUIREMENTS 4.7.8.2.1 The radioactivity rate of noble gates at the offgas recombiner l

effluent shall be continuously monitored by the Pretreatment Off-Gas process radiation monitor required to be OPERABLE or as otherwise provided by the ODCM.

l 4.7.8.2.2 The radioactivity rate of noble gases from the offgas recombiner effluent shall be determined to be within the limit of Specification 3.7.8.2 (above) at the following frequencies

by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the offgas recombiner:

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 Pretreatment Off-Gas process radiation monitor required to be OPERABLE or as otherwise provided by the ODCM 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

The provisions of Specification 4.0.4 are not applicable.

\\

CLINTON - UNIT 1 3/4 7-20 Amendment No. 40

l l

INSTRUMENTATION BASES 3/4.3.7.7 TRAVERSING IN CORE PROBE SYSTEM (Continued) by comparing the detector (s) output with data obtained during the previous LPRM calibrations.

3/4.3.7.8 CHLORINE DETECTION SYSTEM The OPERABILITY of the chlorine detection system ensures that an accidental chlorine release will be detected promptly and the necessary protective actions will be autom:'.lcally initiattd to provide protection for control room personnel.

Automatic transfer to the chlorine mode is not required when chlorine containers having a capacity of 150 pounds or less are stored 100 meters or more from the control room or its fresh air inlets.

Upon detection of a high concentration of chlorine, the control room ventilation system will automatically be placed in the chlorine. mode of operation to provide the required protect. ion.

The detection systems required by this specification are consistent with the recommendations of Regulatory Guide 1.95, "Pr6tection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorina Release," January, 1977.*

1 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION Deleted 3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capa-i bility 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 recommenda-System of Light Water Cooled Reactors," May 1981.tions of Regul "This specification is not applicable after all chlorine containers having a capacity of 100 pounds or greater are removed from the site including the chlorine containers located at the site sewage treatment plant.

CLINTON - UNIT 1 B 3/4 3-7 Amendment No.12

INSTRUMENTATION BASES 3/4.3.7.11 MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING IN5TRUMENTATION The main condenser offgas treatment system explosive gas monitoring instrumen-tation is provided to monitor and control the concentrations of potentially_

explosive gas mixtures in the main condenser offgas treatment system.

The intent of the

note attached to the CHANNEL CALIBRATION requirement is to specify that the CHANNEL CALIBRATION is to be performed using at least two separate gas samples of different specific hydrogen concentrations appropriate for the sensor range., The balance of the sample gas mixture (normally nitrogen) is not necessarily restricted purely to nitrogen but must be in accordance with the requirements or recommendations provided by the manufacturer of the explosive gas monitoring instrumentation.

3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.

Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.

3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION The plant systems actuation instrumentation is provided to initiate action to mitigate the consequences of accidents that are beyond the ability of the opera-tor to control.

The LPCI mode of the RHR system is automatically initiated on a high drywell pressure signal and/or a low reactor water level, level 1, signal.

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

Negative barometric pressure l

l l

l CLINT021 -JffdUL TUWu.A Amendment No.40

ee PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS (Continued) etc....

The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

These records will provide statistical bases for future consideration of snubber service life.

The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.

3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with surveil-lance requirements commensurate with the probability of damage to a source in that group.

Those sources which are frequently handled are required to be tested more often than those which are not.

Sealed sources which are continu-ously enclosed within a shielded mechanism, i.e., sealed sources within radia-tion monitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.6 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system is required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis in FSAR Chapter 15, 3/4.7.7 LIQUID STORAGE TANKS The tanks listed in this Specification include all those outdoor storage tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in each of the specified tanks provides assurance that in the event of an uncontrolled release of the contents from any of these tanks, 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.

CLINTON - UNIT 1 B 3/4 7-3 Amendment No.40

PLANT SYSTEMS BASES 3/4.7.8 MAIN CONDENSER OFFGAS HONITORING 3/4.7.8.1 0FFGAS - EXPLOSIVE GAS MIXTURE Although there should normally be more than sufficient steam flow to the steam jet air ejectors to ensure adequate dilution of hydrogen (and thus prevent the offgas from attaining hydrogen levels in excess of the flammability limit),

this specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas holdup system is monitored and maintained below the flammability limit of hydrogen.

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

3/4.7.8.2 0FFGAS - NOBLE GAS RADI0 ACTIVITY RATE 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 boundary 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, l

l i

i i

i k

l 7

0 k

l CLINTON - UNIT 1 B 3/4 7-4 Amendment No, 40

4 1

ADMINISTRATIVE CONTROLS PROGRAMS f.

Radioactive Effluent controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive offluents 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 00CM, (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 l

following elements:

1.

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the 00CM; 2.

Limitations on the concentrations of radioactive material released

)

in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix 8. Table II, Column 2; 3.

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM; 4.

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

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days; 6.

Limitations on the operability and use of the liquid and gaseous l

effluent treatment systems to ensure that the appropriate portions of these systems are used to redse 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 Appendix I to 10 CFR Part 50; 7.

Limitations on the dose rate resulting from radicartive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, l

Table II, Column 1; CLINTON - UNIT 1 6-16a Amendment No.40

I

^

ADMINISTRATIVE CONTROLS PROGRAMS Radioactive Effluent Controls Program (Continued) 8.

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 I to 10 CFR Part 50; 9.

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

Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from l

uranium fuel cycle sources conforming to 40 CFR Part 190.

g.

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

The program shall provide (1) representative i

measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the affluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM 10CFRPart50,and(3)Inc(2)conformtotheguidanceofAppendixIto lude the following:

1.

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

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census; and t

3.

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

h i

4 i

CLINTON - UNIT 1 6-16b Amendment No.40

l l

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10. Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License, (2) 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 unit.

6. 9.1. 2 The startup report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-tions.

Any corrective actions that were required to obtain satisfactory opera-tion shall also be described.

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

6. 9.1. 3 Startup reports shall be submitted within (1) 90 days following com-pletion of the startup test program, (2) 90 days following resumption or com-mencement 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., ini,tial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of station, utility, and a.

other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure accordin job functions * (e.g., reactor operations and surveillance,g to work andinservice in-spection, routine maintenance, special maintenance, waste processing, and refuelings).

The dose assignments to various duty functions may be

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

CLINTON - UNIT 1 6-17 W

e ADMIN _ISTRATIVE CONTROLS ANNUAL M PORTS (Continued) estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD),

or film badge measurements.

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

In the aggregate, at least 80% of the total whole-body oose received from external sources should be assigned to specific major work functions:

b.

Documentation of all challenges to safety valves or safety / relief valves, and c.

Any other unit unique reports required on an annual basis.

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5 should include 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 iso-topic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radiciodine cer.cen-trations.; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 con-centration and one other radiolodine isotope concentration in microcuries per graW as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6. 9.1. 6 The Annual Radiological Environmental 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 reporting period.

The material provided shall be consistent with the objectives outlined in (1) the 00CM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix ! to 10 CFR Part 50.

CLINT0" - l e 1 6-18 Amendment No.40

a ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT

6. 9.1. 7 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 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 the 00CM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

1 l

CLINTON - UNIT 1 6-19 Amendment No.40 f

THIS PAGE INTENTIONALLY DELETED.

l l

l i

CLINTON - UNIT 1 6-20 Amendment No. 40 l

\\

o ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challennes to the main steam system safety /

relief valves, shall be submitted on a monthly. basis to the Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, D.C. 2']555, with a copy to the Regional Acministrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING L.!MITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before e;ch reload cycle or any remaining part of a reload cycle.

The analytical methods used to determine the core operating limits shall be thost! previously reviewed and approved by the NRC in General Electric Standard Application for Reactor Fuel (GESTAR), NEDE-24011-P-A-8, as amended (latest approved version) and Maximum Extended Operating Domain and Feedwater Heater Out-of-Service Analysis for Clinton Power Station.

NEDC-31546P, August 1988.

The core operating limits shall be determined so that all application limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

THE CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supplements thereto, for each reload cycle, shall be submitted upon issuance to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL rep 0RTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

l 6.10 RECORD RETENTION r

I 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The.A lowing records shall be retained for at least 5 years:

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

level, b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal item; of equipment related to nu;; lear safety, c.

All REPORTABLE EVENTS.

l d.

Records of surveillance activities, inspections, a.d o librations required by these Technical Specifications and the Fire P Ntsction Program.

CLINTON - UNIT 1 6-21 Amendment No.28, 40 l

ADMINISTRATIVE CONTROLS 9

RECORD RETENTION (Continued)

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

e.

f.

Records of radioactiva shipments, Records of sealed source and fission detector leak tests and results.

g.

h.

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

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

Records and drawing changes reflecting unit design modifications made to a.

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

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

control areas.

d.

Records of gaseous and liquid radioactive material released to the

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

identified in Table 5.7.1-1.

f.

Records of reactor tests and experiments, Records of training and qualification for current members of the unit staff.

g.

h.

Records of inservice inspections performed pursuant to these Technical Specifications, i.

Records of quality assurance activities required by the Operational Quality Assurance Manual not listed in Section 6.10.2, which are classi-fied as permanent records by applicable regulations, codes, and standards.

j.

Records of reviews performed for changes.made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k.

Records of meetings of the FRG and the NRAG.

i CLINTON - UNIT 1 6-22

I N

ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) 1.

Records of the service lives of all snubbers including the date at which the service life commences and associated installation and maintenance

records, Records of analyses required by the radiological environmental monitoring m.

program that would permit evaluation of the accuracy of the analysis at a later date.

This should include procedures effective at specified times and QA records showing that these procedures were followed.

Records of reviews performed for changes made to the OFFSITE 00SE n.

CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consic-tent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH PADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20,203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit d

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

g; A radiation monitoring device which continuously indicates the radiation a.

dose rate in the area.

nA b.

A radiation monitoring device which continuously integrates the radiation N$

dose rate in the area and alarms when a preset integrated dose is received.

E Entry into such areas with this monitoring device may be made after the 4

dose rate levels in the area have been established and personnel have i

been made knowledgeable of them.

A health physics qualified individual (i.e., qualified in radiation pro-c.

tection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Protection Supervisor in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas access-ible to personnel with radiation levels such that a ma couldreceivein1houradosegreaterthan1000 mrem *{orportionofthebody shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained

Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the RWP issuance requirement for fields of less than 3000 mrem per hour during the performance of their assigned radiation pro-tection duties, provided they are otherwise following plant radiation protec-tion procedures for entry into high radiation areas.

- Measurements made at 18 inches from sources of radioactivity.

CLINTON - UNIT 1 6-23 Amendment No. 40

1

.~

t ADMINISTRATIVE CONTROLS under the administrative control of the Shift Supervisor on duty and/or the E

Radiation Protection Supervisor.

Doors shall remain locked except during

(

periods of access by personnel under an approved RWP.* For individual areas accessible to personnel with radiation levels such that a major portion of X

the bedv could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a dose in excess of 1000 mrem ** that are C

locatti within large areas, such as the containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted, and a flashing light shall be activated as a warning device.

6.12.3 In addition to the requirements of Specifications 6.12.1 and 6.12.2, 1

for individual areas accessible to personnel such that a major portion of the body could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a dose in excess of 3000 mrtm,* entry shall require an approved RWP which will specify dose rate levels in the immediate

{

work area and require that stay times shall be established, i

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

l 6.13 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3.n.

This documentation shall contain:

1.

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

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

Shall become effective after review and acceptance by the FRG and the approval of the Manager - Clinton Power Station.

" Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the RWP issuance requirement for fields of less than 3000 mrem per hour during the performance of their assigned radia-tion protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.

- Measurements made at 18 inches from sources of radioactivity.

l CLINTON - UNIT 1 6-24 Amendment No. 40

r c.

i a

ADMINISTRATIVE CONTROLS 6.14 0FFSITE 00SE CALCULATION MANUAL (00CM)

Changes to the 00CM:

Shall be documented and records of reviews performed shall be a.

retained as re shall contain: quired by Specification 6.10.3.n.

This documentation 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 radioactive 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 FRG and the approval of the Manager - Clinton Power Station.

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

legible copy of the entire 00CM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be 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.

CLINTON - UNIT 1 6-25 Amendment No.40

s

.a ADMINISTRATIVE CONTROLS THIS PAGE INTENTIONALLY DELETED.

l l

l t

I l

l

' CLINTON - UNIT 1 6-26 Amendment No. 40

-

ML20055G650

Text

Attachment:

Navigation menu

Search