Text

Proposed Tech Specs Re Radioactive Effluent/Radiological Environ Monitoring Program & Solid Radwaste Program
	ML20024H308
Person / Time
Site:	LaSalle
Issue date:	05/22/1991
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20024H306	List: ML20024H305; ML20024H308;
References
	NUDOCS 9105310173
	Download: ML20024H308 (242)
	v • d • e

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INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTI0N.,.........................................................,

1-1 1.2 AVERAGE PLANAR EXP05URE............................... ........... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................ 1-1 1.4 CHANNEL CALIBRATION............................................... 1-1 1.5 CHANNEL CHECK.... ............................................... 1-1 1.6 CHANNEL FUNCTIONAL TE51........... ........ . .. . .. .. .. . . 1-1 1.7 CORE ALTERATION. . . ........... ...... . .. . .... ..... . .

1-2

1. 5 CORE OPERATING LIMITS REPORT. .......................... .... .. 1-2 1.9 CRITICAL POWER RATIO.................. ................. ...... . 1-2 1.10 DOSE EQUIVALENT I-131.. .., , ..................... ............. 1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY... ... .......... ....... ..... . 1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME.... ........ 1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME........ 1-2 1.14 FRACTION OF LIMITING POWER DENSITY.. . .. .. ... ... .... . . 1-3 1.15 FRACTION OF RATED THERMAL POWER. ..... .. .. ..... . .. .. . 1-3 1.16 FREQUENCY NOTATION......... .. ........ .. . . .. .. ... ...... 1- 3 1.17 GASEOUS RADWASTE TREATMENT SYSTEM......... ....... ...... , ...... 1- 3 1.18 I D E NT I F I E D L E A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 ...

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

sw 1.20 LIMITING CONTROL ROD PATTERN..................................... 1-3 jj@l@ 1.21 LINEAR HEAT GENERATION RATE....................................... 1-4

! r.gM M.

e.o 1.22 LOGIC SYSTEM FUNCTIONAL TE5T....................... ... .... ... 1-4 enz

, su 1.23 MAXIMJM FRACTION OF LIMITING POWER DENSITY.. . ... . .. . .. 1-4 I 53 I ~q 1.

24 MINIMUM CRITICAL POWER RATIO............... ... . .. . . 1-4 l0'

, t>

1.J6 0FFSITE DOSE CALCULATION MANUAL. ...... . ... . 1-4 l OG -

- ga '

LA SALLE - UNIT.1 -_m I Amendment No. 70

/. 2'l p1Er8ER(5) ofTHEPany --

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INDEX DEFINITIONS SECTION gt NuM6(A fAGES A 5 AliROP M I 6 DEFINITIONS (Continued) PAGE 21 1.J6' n OPERABLE - OPERABILITY. .......................................... 1-4 1.,27 0PERATIONAL CONDITION -

29 CONDITION.................... ............ 3-4 1..Trt! PHYSICS TEST 5.............. . ................................... 1-4 ao 1.Jf38 PRES 5URE BOUNDARY LEAKAGE. ............ . ..... .......... .. ... 1-5 1.)0 PRIMARY CONTAINMENT INTEGRITY... ... . ....... . .. 1-5 32 1.Jf PRDCE55 CONTROL PROGRAM.. .. .. .. .. . . .. ..... . .. 1-5 33 1.Jf PURGE - PURGING . . .... ....... ........... . ............ . .... 1-5 1.Jf' RATED THERMAL POWER..... . . . .... ......... . . . .. 1-5 26 1.)4' REACTOR PROTECTION SYSTEM RESPONSE TIME.. . . ..... .... .. ... 1-5 3b 1.)f'REPORTABLEEVENT. .. .. . .. .. ... .... . .... .. .. . 1-6 37 1.Js' ROD DENSITY.. ... .. .. . . .. .......... . 1-6 31 1 71'5ECONDARY CONTAINMENT INTEGRITY. . .... . . .... .. .. 1-6 39 1.Jf'5HUTDOWN MARGIN. .. ... . . .. .... .. ... .. . , 1-6 '"

1.39 SOLIO!FICATION. . . ... . ,

. . , , ... . .. .. 1-6 41 1.AF SOURCE CHECK. ... . ....

......... . ....................... ... 1-7 41 1.Af STAGGERED TEST BASI

3 5....................................... ...... 1-7

1. if T H E RMA L P 0WE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..................... 1-7 44 1.35' TURBINE BYPASS RESPONSE vs TIME..................... ................ 1-7

1. f( UN I D E N T I F I E D L E A KA G E . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... ..... 1-7 46 1.SY VENTILATION EXHAUST TREATMENT SY5 TEM.................... ..... ... 1-7 47

{ 1.if VENTING., . . ..., ... .. . ...... . .. . .. 1-7 1.40 site BOUNDARY LA SALLE - UNIT 1 II Amendment No.70

i INDEX I l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.......... ......... 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-9 3/4.3,3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...... 3/4 3-23 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATVS Recirculation Pump Trip System Instrumentation.......... 3/4 3-35 t

End-of-Cycle Recirculation Pump Trip System Instrumentation............................................ 3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION............................................ 3/4 3-45 l

3/4.3.6 CONTROL ROD WITHORAWAL BLOCK INSTRUMENTATION........ ........ 3/4 3-50 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation... ........... . .... . 3/4 3-56 ,

! Seismic Monitoring Instrumentation..... .... ........... . .. 3/4 3-60 l Meteorological Monitoring Instrumentation..... ......... .... 3/4 3-63 i Remote Shutdown Monitoring Instrumentation................... 3/4 3-66 Accident Monitoring Instrumentation. . . . . . ................... 3/4 3-69 Source Range Monitors............. .................. ....... 3/4 3-72 i

Traversing In-core Probe System.............................. 3/4 3-73 Fire Detection Instrumentation............. ................. 3/4 3-75 >

( Radioactive Liquid Effluent Monitoring Instrumentation....... 3/4 3-81 )

exnouve ses R:dit::tFec C :cca; Effluent Monitoring Instrumentation......

3/4 3-)6'1l l Loose-Part Detection System.... ....................... ....

3/4 3-)(?4 3/4.3.8 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION..... .... .. ...... ...... ... . .... 3/4 3-pr 257 l

l LA SALLE - UNIT 1 V Amendment No. 61 1

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration. .. .. . .. . ....... .. ......... ... .. 3/4 11-1 b Dose... ... .. ..... ...... ... ...................... .. 3/4 11-6 Liquid Waste Treatment System. . .......... .............. 3/4 11-7 L

/$. 3/411-/l

- --( Li q u i d H o l d up T a n k s . . . . . . . . . . . . . .. .. .................

fY d3/411.2 GASEOUS CFFLUENTS 3/4 11-9

' Dose Rate. ......... .... .. ............. ..........

f 3/4 11-13 l C se-Noble Gases.. .. . . .... ......... .......... .. ..

Dose-Radiciodines, Radioactive Material in Particulate l Form, and Radionuclides Other than Noble Gases. .. .... .. 3/4 11-14 Gaseous Waste Treu.... ant System.... . ........ . ...... ... 3/4 11-15 (Ventilation Exhaust Treatment System. . . .. .. ... .. ...... 3/4 11-16j i

I Explosive Gas Mixture., ..... .... .. ....... . ...... ... . 3/4 11-}f2 3/4 11- 43

( Main Condenser. ... .... .. ..... .. . ...... ........ . .

{ Venting or Purging. . .. . .. . . . . . . . . 3/4 11-20)

I 3/4 11.3 SOLID RADI0 ACTIVE WASTE. ............... ... ... ... . .. ..

3/4 11-21 [

3/4 11.4 TOTAL 00SE... .... . ... . . .... .. . .. . .... 3/4 11-22 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING ,

3/4 12.1 MONITORING PROGRAM. ... . . .. .. . .. ..... . 3/4 12-1 i

3/4 12.2 LAND USE CENSUS.. . . . . . ... .. .. ... . 3/4 12-9 t

3/4 12.3 INTERLABORATORY COMPARISON PROGRAM... .. . ..... . . 3/4 12-10, O

r 1

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LA SALLE - UNIT 1 XI

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INDEX BASES

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SECTION PAGE INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Heteorological Monitoring Instrumentation.. .. .... .... B 3/4 3-4 Remote Shutdown Monitoring Instrumentation.............. B 3/4 3-4 Accident Monitoring Instrumentation............... .... B 3/4 3-5 Source Range Monitors....................... .......... B 3/4 3-5 Traversing In-core Probe System....................... . B 3/4 3-5 Chlorine and Ammonia Detection System.... ..... ... .... B 3/4 3-5 Fire Detection Instrumentation...... ....... ... .... B 3/4 3-5 p-(RadioactiveLiquidEffluentMonitoringInstrumentation. B3/43-6)

ExPLostv[ GAS

":di :: tic: C :: u Ef'lucnt Monitoring Instrumentation.. B 3/4 3-6 Loose-Part Detection System. ............ ............... B 3/4 3-6 3/4.3.8 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.............................. ......... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..... ............ . ..... .. .. B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.. . ... ... ......... .. ..... . B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.. ................... . .. B 3/4 4-2 Operational Leakage. .... .. ................. ....... B 3/4 4-2 l 3/4.4.4 CHEMISTRY.... ..... .. ............ ......... ... . ... B 3/4 4-2

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

2 /4. 4. 6 PRESSURE / TEMPERATURE LIMITS................. . . B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES................. . . B 3/4 4-5 3/4.4.8 STRUCTURAL INTEGRITY... . . ..... . ... B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REMOVAL.. . . . B 3/4 4-5 LA SALLE - UNIT 1 XIII l

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INDEX l l

BASES s

SECTION PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY, . ......... . ..... .. B 3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM., . ... ........ . .... .. B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS., .. ........... ..... B 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS........ . ... ............ ...... B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION........ ..... .. . ....... .... . B 3/4 10-1 3/4.10.6 TRAINING STARTUPS.. . . .. . ... ... .. . .. .. B 3/4 10-1 3/4.10.7 CONFIRMA10RY FLOW INDUCED VIBRATION TEST. .... ...... .. B 3/4 10-1 3/4.10.8 SUPPRESSION CHAMBER WATER TEMPERATURE., ... .. . B 3/4 10-2 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS [I Concentration.. . . .... .... . ...... .. B 3/411-1 L

i Dose.......... . . .......... ... ......... ........ B 3/4 11-1 Liquid Waste Treatment System.......................... B 3/4 11-2;

-(LiquidHoldupTanks. . .. ...... ................. ..

B.3/411-/

-( 3/4.11. 2 GASEOUS EFFLUENTS f[

L Dose Rate...... . ... ... ...... ..... .. .............. B 3/4 11-2' Dose - Noble Gases.... . .. .... ...... ....... .... B 3/4 11-3

[ Dose - Radioiodines, Radioactive Materials in Particulate Form and Radionuclides Other than Noble Gases... ...... .. . ... ..... . , ,, B 3/4 11-3 Gaseous Radwaste Treatment System and Ventilation

( Exhaust Treatment System... ... .... . . .... B 3/4 11-4)

Explosive Gas Mixture. ....... .. ........ . ....., ,

B 3/4 11-/g Main Condenser.. .. .................. ... .... . .... B 3/4 11-gs)V Venting or Purging... ..... . . .......... ......

B3/411-5) 3/4.11.3 SOLID RADI0t.CTIVE WASTE.. .... . .. ... ...... .. B 3/4 11-5 3/4.11.4 TOTAL DOSE. . ..... .. ... . .. . .. . B 3/4 11-5 3/4.12 RADIO ^.CTIVE ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM.. . ..... .... . . . B 3/4 12-1 3/4 12.2 LAND USE CENSUS. . . . .. . . . B 3/4 12-1 3/4 12.3 INTERLABORATORY COMPARISON PROGRAM. . ..

B3/412-2)

LA SALLE - UNIT 1 XVI

INDEX A0t'INISTRATIVE CONTROLS 1 SE: TION PAGE 6.1 ORGANIZATION, REVIEW, INVESTIGATION, AND AUDIT.................... 6-1 6.1.1 High Radiation Areas....................................... 6-15 6.2 )

PLANT OPERATING PROCEDURES AND PR0 GRAMS........................... 6 16

6. 3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE IN PLANT 0PERATION..................................... 6 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED........ 6- 19

6. 5 PLANT OPERATING REC 0RDS........................................... 6- 19 6.6 REPORTING REQUIREMENTS............................................ 6- 21 6"7 PROCESS CONTROL PR0 GRAM........................................... 6- 28 6.8 0FFSITE DOSE CALCULATION MANUAL........................ ....... 6- 28

6. 9 MAJOR CHANGES TO RADIOACTIVE 'fASTE TREATMENT SYSTEMS.............. 6g1

/y6'4 S gt\

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LA SALLE - UNIT 1 XV7II Amendment No. 66 i

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INDEX LIST OF TABLES (Continued)

TABLE PAGE 4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .............................. 3/4 3-65 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION .......... .. 3/4 3-67 4.3.7.4-1 REMOTE SHUTOOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ................. ......... . 3/4 3-68 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION ............ ...... 3/4 3-70 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ....... ... ............ ... 3/4 3-71 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION ................... ..... 3/4 3-76 RADI0 ACTIVE LIQUID EF.LUENT MONITORING (3.3.7.10-1 INSTRUMENTATION ....... ........ .. ...... ... ..... 3/4 3-82 j[i

[4.3.7.10-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING

( INSTRUMENTATION SURVEILLANCE REQUIREMENTS .............. 3/4 3-84 j 3.3.7.11-1 ->( RADIDACTIVE GASEOUS EFFLUENTMORING INSTRUMENTATION . ..........

p .................. 3/4 3-pfgg 4.3.7,11-1 --*(RADI0 ACTIVE GASEOUS EFFLUENT) MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .. ....... . . 3/4 3-jiHfg3 3.3.8-1 g FEEDWATER/ MAIN TURBINE TRIP SYSTEM g ACTUATION INSTRUMENTATION ......... .......... . ... . , 3/4 3-pfgg, 3.3.8-2 W FEEDWATER/ MAIN TURBINE TRIP SYSTEM

$a

,o ACTUATION INSTRUMENTATION SETPOINTS ...... . ....... 3/4 3-)(g7 4.3.8.1-1J d < FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ... ........ . 3/4 3-pfg 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES ... ., 3/4 4-9 i

3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS . ..... ..... . 3/4 4-12 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ............... .. ...... . . . ... .. 3/4 4-15 4.4.6.1.3-1 REACTOR \ESSEL MATERIAL SURVEILLANCE PROGRAM--

VITHDRAWAL SCHEDULE . . . . . ... . .. . .. 3/4 4-19 I 4.6.1.5-1 TENDON SURVEILLANCE ... .. . . . . . . 3/4 6-11 LA SALLE - UNIT 1 XXII Amendment No. 18

INDEX LIST OF TABLES (Continued)

TABLE PAGE 4.6.1.5-2 TENDON LIFT-OFF FORCE .................................. 3/4 C-12 l

3.6.3-1 PRIMARY CONTAINMEN ISOLATION VALVES .................... 3/4 6-24 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES .......... . ........ .. .... 3/4 6-39 3.7.5.2-1 DELUGE AND SPRINKLER SYSTEMS ................. ..... .. 3/4 7-16 3.7.5.4-1 I FIRE HOSE STATIONS .............................. ...... 3/4 7-19 l 3.7.7-1 AREA TEMPERATURE MONITORING ......... ............... .. 3/4 7-25 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE ........ . . ........... 3/4 8-7 8.4.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS .. .. .... . 3/4 8-18 3.8.3.2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ............ ............ 3/4 8-24 3.8.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION ................................... ....... . 3/4 8-27 J

[

' 3.11.1-1 MAXIMUM PERMISSIBLE CONCENTRATION OF DISSOLVED OR ENTRAINED NOBLE GASES RELEASED FROM THE SITE

( TO UNRESTRICTED AREAS IN LIQUID WASTE .................. 3/4 11-2

[4.11.1-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS

( PROGRAM ........... ............................. ...... 3/4 11-3

'4.11.2-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING, AND u ANALYSIS PROGRAM . 3/4 11-10 3.12.1-1 RADI0 LOGICAL ENVIRONMENTAL MONITORING PROGRAM ...... ... 3/4 12-3 3.12.1-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES ........... ........ . . .... 3/4 12-6 t

4.12.1-1 MINIMUM VALUES FOR THE LOWER LIMITS OF DETECTION . . . . 3/4 12-7 B3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS ...... .. . .... ... B 3/4 2-2 ) i 03/4.4.6-1 REACTOR VCSSEL TOUGHNESS .. ....... ... ......... . .. B 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS .. ... .. ......... 5-6 ;

LA SALLE - UNIT 1 XXIII Amendment No. 1B

i DEFINITIONS LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit l 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, l 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 l performed by any series of sequential, overlapping or total system steps i such that the entire logic system is tested. '

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

MINIMUM CRITICAL POWER RATIO 1.J# The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

%s exists in the core.

'0FFSITE DOSE CALCULATION MANUAL

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p 1.25 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methocology l f and parameters used in the calculation of offsite doses due to radioactive g, ( gaseousandliquiaeffluentsandinthecalculationofgaseousandliquid) effluent monitoring alarm / trip setpoints.

OPERABLE - OPERASILITY 1.)(T A system, subsystem, train, component or device snall be OPERABLE or nave

{

21 OPERABILITY wnen it is capable of performing its specified function (s),

and when all necessary attencant instrumentation, controls, a normal anc an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION - CONDITION 1.ff An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive l 23 combination of mode switen position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1.JE PHYSICS TESTS shall be those tests performed to measure the fundamental l 29 nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise opproved by the Commission.

LA SALLE UNIT 1 1-4 Amendment No. 70

DEFINITIONS PRESSURE BOUNDARY LEAKAGE 1.J# PRES 5URE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault l

30 in a reactor coolant system component body, p pe wall or v'essel wall.

i PRIMARY CONTAINMENT INTEGR]TY 2.pfPRIMARYCONTAINMENTINTEGRITYshallexistwhen:

M l

a. All primary containment penetrations required to be closed during accident conditions are either:

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

2. Closed by at least one manual valve, blind flange, or ceactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification, 3.6.3.

b. All primary containment equipment hatenes are closec and sealed.

c. Each primary containment air lock is OPERABLE pursuant to specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.e.l.2.

e. The suppression chamber is OPERASLE pursuant to Soecification 3.6.2.1.

i (d f. The sealing mechanism associated with eacn primary containment penetration; e.g., weles, bellows or 0-rings, is OPERAELE.

l PROCESS CONTROL PROGRAM l 1.31 The PROCESS CONTROL PROGRAM (PCP) shall contain the sampling, analysis, '

anc formulation cetermination by which SOLIDIFICATION of racioactive wastes f rom licuic systems is assured. 3 PURGE - PURGING 1.X PURGE or PURGING shall be the controlled process of discharging air or l l 3'3 gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacem-ent air or gas is required to purify the confinement.

RATED THERMAL POWER 1.JW RATED THERMAL POWER shall be a total reactor core heat transfer rate to l 4 the reactor coolant of 3323 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME

1. K REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from l M when the monitored parameter exceeds its trip setpoint at the channel sensor until de energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

1 LA SALLE UNIT 1 1-5 Amencment No.70 l

l l

DEFINITIONS REPORTABLE EVENT 1.pfAREPORTABLEEVENTshallbeanyofthoseconditionsspecifiedin 36 Section 50.73 to 10 CFR Part 50. l ROD DENSITY 1.jT ROD DENSITY shall be the number of control rod notches inserted as a l 77 fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.)f SECONDARY CONTAINMENT INTEGRITY shall exist when:

N a. All secondary containment penetrations required to be closed l

during accident conditions are either:

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

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

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

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

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

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

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

SHUTDOWN MARGIN 1.J8'5HUTDOWN MARGIN shall be the amount of reactivity by which the reactor is l 39 subcritical or would be subcritical assuming all control rods are fully

,C 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 I condition; cold, i.e. 68"F; and xenon free.

SOLIDIFICATION 1.39 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid l systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and ihape, bounded by a stable surface of j distinct outline on all sides (free-standing). 1 LA SALLE UNIT 1 1-6 Amendment No. 70

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INSERT Al t!EBEERIS) 0F THE_EVEllC 1.24 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

INSERT A2 OEESHE_QQSE_CALCULMIOUANUAL 1,26 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

INSERT B 1.32 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

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

ZNLD/862/12

DEFINITIONS SOURCE CHECK 1.jd A SOURCE CHECK shall be the qualitative assessment of channel response 1

4) when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS ,

1.ff A STAGGERED TEST BASIS shall consist of: l O a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.

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

THERMAL POWER 1./l THERMAL POWER shall be the total reactor core heat transfer rate to the l 43 reactor coolar.t.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.PJI The TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when {

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

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

UNIDENT]FIED LEAKAGE

1. UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE. I VENTILATION EXHAUST TREATMENT SYSTEM 1.y!fAVENTILATIONEXHAUSTTREATMENTSYSTEMshallbeanysystemdesignedand l 6 installed to reduce gaseous radiciodine or radioactive material in particu-late form in effluents by passing ventilation or vent exhaust gases through-charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.f4IVENTINGshallbethecontrolledprocessofdischargingairorgasfroma l 47 confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

LA SALLE UNIT 1 1-7 Amendment No. 70

-~

W[ f.A'I/W M6E'

\

QiSTRUMENTATION RA 0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITIN CONDITION FOR OPERATION x <

3.3.7.10 The radioactive liquid effluent monitoring instrumentation ch nels shown in Table .3.7.10-1 shall be OPERABLE with their alarm / trip setp nts set to ensure t t the limits of Specification 3.11.1.1 are not excee ed. The alarm trip setpoi tQ these channels shall be determined in accor nce with the Offsite Dose C c n Manual (ODCM).

APPLICABILITY: At al .

ACTION:

a. With a radioactiv lig d effluent monitoring i trumentation channel alarm / trip setpoint sess onservative than re ired, immediately suspend the release ioactive liquid ef uents monitored by the affected channel or de the channel ino erable.

b. With less than the minim number of ra cactive liquid effluent monitoring instrumentation hannels OP ABLE, take the ACTION shown in Table 3.3.7.10-1. Resto the ino erable instrumentation to OPERABLE status within the ti e spe fied in the ACTION or, in lieu of a Licensee Event Repor e plain in the next Semiannual Radioactive Effluent Release Rep y this inoperability was not corrected within the time s e .

1

c. The provisions of Specificat'ons 3.0. g 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Nm

/ \\% \

l i 4.3.7.10 mac .adioactive 1 uid effluent monitoring in trumentation channel shall be demonstrated OPERA E by performance of the CHAN' W CK, SOURCE CHECK, CHANNEL FUNCTIONAL EST and CHANNEL CALIBRATION oper t at the frequencies shown in Tab e 4.3.7.10-1.

l l

SALLE - UNIT 1 3/4 3-81 Amendment No. 18

5 TABLE 3.3.7.10-1

'O RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

{

I E MINIMUM q CHANNELS g INSTRUMENT OPERABLE ACTION

1. GAMMA SCINTILLATI , ONITOR PROVIDING ALARM AND AUTOMATIC

'A TERMINATION OF REL E p

a. Liquid Radwaste Efflu t Line 100 l

2. GAhMA SCINTILLATION MONITORS PR DING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE R a. Service Water System Effluent Line nit 1) 1 101

b. RHR Service Water (Line A) Effluent ne 1 101 y c. RHR Service Water (Line B) Effluent Li 1 101 g d. Service Water System Effluent Line (Unit 2 1 101

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 102

b. River Discharge - Blowdown Pi 1 102 9

8 wd hi

! a i /

l'

- n k

. y.

m Ih

MtETE" ENrt&~ A4GC

\STRUMENTATION

, TABLE 3.3.7.10-1 (Continued)

TABLE NOTATION -

ACTION 100 -

With the number of OPERABLE channels less than. req red d by the Minimum Channels OPERABLE requirement, eff ent #

eleases may continue for up to 14 days provide that l pt or to initiating a release: )

least two independent samples are a lyzed in ordance with Specification 4.11.1. 3, and

b. At A st two technically qualified members of the a Staff independently ver y the release rate c Ic lat'ons and discharge line alving; Otherwise suspend release of rad active effluents via this pathw .

ACTION 101 -

With the numb of channels O RABLE less than required by the Minimum hannels OPE BLE requirement, effluent releases via thi pathway ay continue for up to 30 days i provided that, at east ce per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analy d a limit of detection of at least 10-7 microcurie /ml o .amma spectrometric analysis, j ACTION 102 -

With the number of han RABLE less than required by the Minimum Ch nels 0 E requirement, effluent releases via thi pathway tinue for up to 30 days provided the f w rate is es ima at least once per_

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months durin actual release Pu curves for Instru-ment 3a, or or known valve pos on for Instrument 3b, may be use to estimate flow.

m 7

G5 m

l LA SALLE - UNIT 1 3/4 3-83 Amendment No.18

5 TABLE 4.3.7.10-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

{

c CHANNEL h CHANNEL SOURCE FUNCT A CHANNEL g INSTRUMENT CHECK CHECK S CALIBRATION

1. GAMMA SClNTILLATION M ITOR PROVIDING ALARM

AND AUTOMATIC TERMINA N OF RELEASE

a. Liquid Radtraste Effluent ine D P Q(1) R(3)

2. GAMMA SCINTILLATION MONITORS PROVI G ALARM BUT NOT PROVIDING AUTOMATIC TERMINA ON OF RELEASE R

a. Service Water System Effluent Line (Unit M Q(2) R(3) l y b. RHR Service Water (Line A) Effluent Line O M Q(2) R(3) g c. RHR Service Water (Line B) Effluent Line D M Q(2) R(3)

d. Service Water Systen: Effluent Line (Unit M Q(2) R(3) l

3. FLOW RATE MEASUREMENT DEVICES a.

b.

Liquid Radwaste Effluent Line River Discharge - Blowdown y f D(4)

D(4)

N.A.

Q Q

R R

- .]

N l

- ,d e

. e la d bx m-

/ k' N

5 b

'1 5 / -

3m D

NLETEENTr&%C INSTRUMENTATION TABLE 4.3.7.10-1 (Continued)

~-

TABLE NOTATION isolatic,

-(1) T CHANNEL FUNCTIONAL TEST shall also demonstrate that autornati of is pathway and control alarm annunciation occurs if any of the foll ing conditions exist:

1. In indicates measured levels above the alarm / ip setpoint.

2. Loss p er.

3. Instrume ms on dt.wnscale f ailure.,

4. Instrument n- s not set in Operate or H h Voltage mode.

(2) The CHANNEL FUNCTIO T shall also demons ate that control room alarm annunciation o urs if any of the foll wing conditions exist:

1. Instrument indicat measured level above the alarm setpoint.

2. Loss of power.

3; Instrument alarms on dow cal 'tailure.

4. Instrument controls not se in Operate or High Voltage mode.

(3)- The initial CHANNEL CALIBRAT N sha rformed using one or more of the reference radioactive andards ed by the National Bureau of Standards or using stada s that have btained from suppliers that participate in measurem t assurance ac vit with NBS. These standards shall permit calibrati g the system over s in nded range of energy and measurement range. .r subsequent CHANNEL TION, the initial reference radioacti e standards or radioacti e ces that have been related to the in tal calibration shall be u (4) CHANNEL CHECK all consist of' verifying indicat n of flow during periods of release HANNEL CHECK shall be made at least nce per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on

~~~~*

~ , days g whi continuous, periodic, or batch releas g ade.

GS v /

LA SALLE - UNIT 1 3/4 3-85

DELETE ~ ENitK6~ t*MC MPLACE~ 5PEclFtc4Tto^/

N5TRUMENTATION # b R 0 ACTIVE G E005 EFrLUENT MONITORING IN51RUMENTA110N LIMITIN CONDITION FOR OPERATION ,/

3.3.7.11 The adio.ctive gaseous effluent monitoring instrumenta on channels shown in lable . ./ 11-1 shall be OPERABLE with their alarm /tri setpoints set to ensure th- limits of Specification 3.11.2.1 are no exceeded. The alarm / trip setpoin these channels shall be determined i. accordance with the ODCM.

APPLICABILITY: As show n ble 3.3.7.11-1 ACTION: y

a. With a radioactive seous effluent moni oring instrumentation channel alarm / trip se .oint less conse ative than required, immediately suspend the release of ra 10 active gaseous effluents monitored by the affecte channel o declare the channel inoperable,

b. With less than the minimum r mbe of radioactive gaseous effluent monitoring instrumentation ch els OPERABLE, take the ACTION shown in Table 3.3.7.11-1.

c. The provisions of Specific ions b nd 3.0.4 are not applicable.

~C %/

SURVEILLANCE REQUIREMENTS et

, sw b

4.3.7.11 Each radioactive aseous effluent monitoring nstrumentation channel shall be demonstrated OP ,ABLE by performance of the CHA NEL CHECK, SOURCE CHECK, CHANNEL FUNCTION TEST and CHANNEL CAllBRATION op a g at the frequencies shown in T ble 4.3.7.11-1, a

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DELETC EA/T/<C MW RERACE wtm N6N INSTRUMENTATION Mdl E TABLE 3.3.7.11.1 (Continued)

TABLE NOTATION A all times.

Duri 1 main condenser offgas treatment system operation

During n on of the main condenser air ejector.

1. During ope i on of the SBGTS.

AC110N 110 -

Wi i

- number of channels OPERABLE less than required by the Mini'um nnels OPERABLE requirement, efflu t r(leases via this ithwa' may continue for up to 30 days rovided grab sample- 1 ken at least once per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are anal ' d or noble gas gamma emitter within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 111 -

With the nu 5er of channels OPERABLE .ss than required by the Minimum Chant is OPERABLE requireme , operation of main condenser offg treatment system y continue for up to 30 days provide 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 analy. d within the ollowing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the recombiner(s) tempe ature rema'ns constant and THERMAL POWER has not changed, the rab sa le collection frequency may be changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 112 -

With the number of channy s RABLE less than required by the v Minimum Channels OPERAEt.E -

i ement, suspend release of radioactive effluentsj ia th's""7,st.hway.

4

/ .A ACTION 113 -

With the number of hannels OP BJ.f. less than required by the Minimum Channels ERABLE requir e p effluent releases via this pathway ma continue for up t ays provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the channel ha ee declared inoperable samples are ntinuously collected wi h auxiliary sampling equipment a required in Table 4.11.2-ACTION 114 -

With the umber of channels OPERABLE less h equi.*ed by the minimur Channels OPERABLE requirement, effl e eleases via this 'athway may continue for up to 30 days rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ogdtheflow ACTION 115 - W~ th the number of channels OPERABLE less than re utred by the inimum Channels OPERABLE requirement, the output om the charcoal adsorber vessels may be released to the env'ronment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

a, The offgas treatment system is not bypassed, and

b. The offgas treatment delay system noble gas activity t

effluent downstream monitor is OPERABLE; Otherwise, be in at least STARTUP with the main steam isolatio

valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , l

LA SALLE - UNIT 1 3/4 3-88 l

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AC/%AC5 wtTH NE}V TABLE 4.3.L 11-1 (Cantinteo) 774 81.E TARLE NOTAi!CN !

At all times. /

,4 ring main concenser offgas treatment system operation.

The coecified IS-month interval may De waived for Cycle 1 provided .he surve'llance is cerformed during Refuel 1. }

Ouring 9eration of the main condenser air ejector, ,

1. Ouring op ation of the SDGTS.

(1) The CHANNEL 'N L TEST shall also cemonstrate the aut;datic isolation caDaDility of occurs if any o t thway,andthatcontrolroomalarm[nunciation following conditions exists: (eac, channel will be tested indeoenden 1, s not to initiate automatic 1 olation during operation).

1. Instrument indic es asured levels above th alarm / trip setDoint.

2. Loss of power 3.

Instrument elarms on - ownscale f ailure.

4 Instrument controls not *et in Operate High Voltage mode. (Auto-matic isolation snall be emonstrated uring the CHANNEL CALIBRATION.)

(2) The CHANNEL FUNCTIONAL TEST for he log ale monitor shall also cemonstrate that control room alarm annunciat1'n oc Jr5 if any of the following conditions exists:

1. Int.trument indicates measured 1 1s above the alarm setpoint.

Loss of power.

2.

p

3. Instrument alarms on downs le failur g 4 Instrument con als not t in Operate c .ahqh Voltage mode.

(3) The initial CHANNEL CALIBRA .CN shall be perfor F ing one or more of the reference radioactive tandards certified by t e Jational Bureau of Standards (NBS) or using tandards that have been ec from suppliers that participate in mea arement assurance activitie with'NBS. These standards shall permi calibrating the system over it- intended range of enercy and measureme ranne. For suoseQuent CHANNEL u LIBRATION, the i itial related been referenceto r dioactive standards or radioactive s PNtt>that have initial calibration shall be used.

(4) The CHANNEL CAL ,, RATION shall include the use of standard ga r es containing a minal:

1. One vol me percent hydrogen, balance nitrogen, and

2. "our olume percent hydrogen, balance nitrogen.

l (5) The CHAT EL FUNCTIONAL TEST shall also demonstrate that control room alarm nnunciation occurs if any of the following conditions exists:

1. .nstrument indicates measured levels above the alarm setpoint.

2. - Circuit failure.

. Instrument controls not set in the Operate mode.

LA ALLE UNIT-1 3/4 3-90 Amendment No. 23 l

l f INSIRUMENIA110N f I

EXP_LO51VLGASliON130 RING INSTRUMENIA110N J

LIMll1NG COND1110tLf0R_0fCRAIION _- .

3.3.7.11 The explosive gas monitoring instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their Alarm / Trip setpoints set to ensure that the limits of specification 3.11.2.6 are not exceeded.

APPLICABillIY; During operation of the main condenser air ejector.

AC110N1

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

l

b. With less than the M nimum number of explosive gas monitoring .

Instrumentation channels OPERABLE, take the ACTION shown in Table 3.347.11-1. Restore the inoperable instrumentation channels to an OPERABLE status within 30 days, or prepare and submit a Special Report to the Commission pursuant to ,

Specification 6.6.C within the next 10 days outlining the !

cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not appilcable.

SURVEILLANCL REQUIREMENIS _ ___._ _ _ _ . _ . . . _ _ . _

4.3.7.11 Each explosive gas monitoring instrumentation channel shall be i demonstrated OPERABLE by performance of a CHANNEL CHECK, CHANNEL 4 FUNCTIONAL TEST and CHANNEL CAllBRATION at the frequencies shown in Table 4.3.7.11-1.

l l

l .

1 LA SALLE - UNIT 1 3/4 3-81 PROPOSED AMENDMENT ZNLD/862-13 -

L,..__.._,,,~,. . _ , , _ _ . . ~ . _ . _ _ _ _ _ _ . . _ _ . _ _

INSTRUMENTATION 1ADLE 343,7.11-1 EXPE0SIVE GAS HON 110 RING.lNSTRUMENTA110N MINIMUM CHANNELS INSTRUMENI . _ . _0PERADLL .- ACTION

1. MAIN CONDENSER OffGAS TREATMENT SYSTEM EXPLOSlVE GAS MONiiORING SYST[M '

(for systems designed to withstand the effects of a hydrogen explosion)

a. Hydrogen Monitor 1/ train 110 .

TADLE NOTATION 1

ACTION 110 - Hith the number of channels OPLRABLE less than required by the Minimum Channels OPERABLE requirement, operation of the main condenser offgas treatment system may continue for up to 30 days provided grab samples are co11ected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the fellowing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the recombiner(s) temperature remains constant and THERMAL POWER has not changed, the grab sample collection frequency may be changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

I i.

I r

LASALLE - UNIT 1 3/4 3-82 PROPOSED AMENDMENT ZNLD/862/14

1 INS 1RUMENJATION IABLL413,M1-J EXELOSlVE GASl!ON110 RING _IN51RUMENTATION OPERA 110NAL CWANNEL- COND1110NS FOR CHANNEL FUNCTIONAL CHAkNEL HHICH SURVEll- ,

INSIRUMENT CHECK _.J ES L_ CAllBRATIONt LANCE _RE0VIRED

, 1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MON!10 RING SYSTEM Hydrogen Monitor **

a. D M Q l

l TADLLN01A110N The CHANNEL CALIBRA110N shall include the use of standard gas samples containing a nominal:

i

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen, i

- During operation of the main condenser air ejector.

l i LASALLE - UNIT 1 3/4 3-83 PROPOSED nMENDMENT f

ZNLD/862-15

_ _ . _ , . . _ _- , _ . _ . . - - - -..__._ ,_ _ _ . _ ._ _ . ___._ _ _._. _.-._.._ -.~,___ -

INSTRUMINTATION LOOSE-PART DETECTION SYST[M LIMil!NG CONDITION FOR_0PERATION 3.3.7,12 ihe loose part detection system shall be OPERABLE.

APPLICABILITY: OPERA 110NAL CONDITIONS 1 and 2.-

AC110N:

a. With one or more loose part detection system channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.c within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, ,

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

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

b. CHANNEL IUNCTIONAL TEST at least once per 31 days, and

c. CHANNEL CALIBRATION at least once per 18 months.

M LA sAttt - uni 1 1 3/43-g

INSTRUMENTATICN 3/4.3.8 FEEDWATER/ MAIN TURBINE TR]P SYSTEM ACTUATICN INSTRUMENTATICN LIMITING CCNDITICN r0R OPERATICN 3.3.3 The f eecwater/ main turbine trip system actuation instrumentation enannels shown in Table 3.3.8-1 snall be OPERABLE with their trip setpoints set consistent with the values snown in the Trip Setpoint column of Tacle 3.3.3-C.

APPL!CABIL 'Y: OPERATIONAL CCNDIT!CN 1.

ACTION:

a. With a feecwater/ main turbine trip system actuation instrumentation channel-trip setpoint less conservative than the value shown in the Allowable values column of Table 3.3.8-2, declare the channel inoperable and either place the inoperable channel in the tripped condition until the channel _is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trio Setpoint value, or declare the associated system inoperaDie,

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

c. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels per Trip System requirement, restore at least one of the inoperable channels to OPERABLE status withir.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS 4.3.3.1 Eacn feecwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL. FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.8.1-1.

4.3.8.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all enannels shall be parformed at least once per 18 months. " )

r l

The specified 18 month interval may be waived for Cycle 1 provided the surveillance is performed during Refuel 1. I LASALLE UNIT-1 3/4 3;)(' Amendment No. 24 ST

l l

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W LA SALLE - UNil 1 3/4 3-A f

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

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i 5 TABLE 4.3.8.1-1.

p FEEDWATER/ MAIN TURBINE TRIP SYSTtM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

c- CHANNEL l 5 CHANNEL FUNCTIONAL CHANNEL

TRIP FUNCTION CHECK TEST CALIBRATION j ~

a. Reactor Vessel Water Level-High, 5 M R level 8 i

i i

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

ORETC ENTfft 84G[

3/4.11 RADI0 ACTIVE EffLV[NTS 3/4.11.1 LIQUID EFFLUENTS C CENTRATION LIMITI ~ CONDITION FOR OPERAT10N

\

3.11.1.1 Th ntration of radioactive material released from t e site (see figu.*e b. I shall be limited to the concentrations speci ed in 10 CFR Part 20, Ap n B, Table 11, Column 2 for radionuclides ot er than dissolved or entr i noble gases. For dissolved or entraine the concentration i e limited to the concentrations spec'p fied noblein gases, Table 3.11.1-1. )

APPLICABILITY: At all i ACTION:

With the concentration of ra active material rel sed from the site exceeding the above limits, immediately ' store the concent ation to within the above limits.

SURVE1LtANCE REQUIREMEN15

\ /

4.11.1.1.1 The radioactivity conten of ea 5 .- h of radioactive liquid waste shall be determined prior to lease b - p ing and analysis in accord-ance with Table 4.11.1-1. The res its of pre- .l g analyses shall be used with the calculational methods i the ODCM to as ure hat the concentration at the point of release is maintai ed within the lim o Specification 3.11.1.1.

4.11.1.1.2 Post-release an yses of samples composi h m batch releases shall be performed in acco ance with Table 4.11.1-1 e 'sults of the previous post-release an' yses shall be used with the ct i ional methods in the ODCM to assure tt ; t th concentre.tions at the poi of release were maintained within the imits of Specification 3.11.1.1.

4.11.1.1.3 The ra 'oactivity concentration of liquids disch th9 continuous release points sh IbedeterminedbycollectionandanalysisO(s les in accordance with oble 4.11.1-1. The results of the analyses sharg b l ed with the calculatio 1 methods in the ODCM to assure that the concentrati the point of rele se are maintained within the limits of Specification i 1 .

t SAI L L - UNil 1 3/4 11-1 I

AMTEw7?4C PAG 6" iABtE 3.11.1-1 MAXIMUM PERMISSIBLE CONCLNTRATION OF v

~IflssoLvf 0 DYTNTRITREDNIIITTiT5TF RC.LE ARTTROM THE SilE TO UNRESTRicT[FARl A5 gllQUIDWASTI' d

U h. MPC(pCl/ml)*

Kr 5 2L-4 85 SE-4 87 4f b 88 9L-5 Ar 41 7E-5 Xe 131 m 7E-4 133 m SE-4 133 6E-4 v

135 m 20 4 135 2E-4 7A -

~

rComputed from Equ< inn 20 of ICRP Put>11 cation 2 (195.1, a< sted for infinite cloud s >mersion in water, and R = 0.01 rem /w e , = 1.0 gm/cm3 ,

and P /P : 1.

9 1 A SAllt - UN)I 1 3/4 11-2

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

h[/ Cf( ENT/RCMGE TABLE 4.11,1-1 RADIDACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR_0 GRAM /

/

Minimum Type of LodrLimit Liquid Relea Analysis Type g { Sampling requency Activity 9(Detection Frequency Analysis ., (LLD)

DN / (pCi/ml)a A. Batch Waste

\

'(g P Principal Ga a/ 5x10*7 Release Ea h Each Batch Emitters #

d Tanks

-6 I-13[ 1x10 P H D olved and ~6 1x10 One Batch /M trained Gases

\ j Gamma emitters) ,

P M H-3 1x10

-5 D

Each Batch Co- osi e Gross Alpha ~7 1x10 ,

P d \ tr-89, Sr-90 5x10

~0 Each Batch C mposite ~7 (ek 1x10

-6

8. Continuous W Pr M Gamma 5x10

~7 Releases Continuo a Composite c Emit rg

-5 1-131 \ 1x10 M Dissolved Me Ix10 -5 G b Sample Entrained Ga e 7 (Gamma Emitte H-3 i -5 Continuous C

Composite M

c p\h10 CN Gross Alpha \ 1x10~7 ,

4 Q Sr-89, Sr-90 -8

\x10 Continuous C Composite c Fe-55 l -

1xk6 l

LA SALLE - UNIT I 3/4 11-3 Amendment No 18 l

05/E75 &f//d"ft16[

TABLE 4.11.1-1 (Continued) j TABLE NOTATIL.

a.

pe LLD is the smallest concentration of radioactive material a sample th t will be cetected with 95% probability with 5% probabilit of falsely con luding that a blank observation represents a "real" sig 1.

For a rticular measurement system (which may include r iochemical separat n)g

- 4.66 s LLD =

2.22x10' Y exp (-Aat)

Where:

LLD is the "a ri microcutie per i lower limit of detection as defined above (as s or volume),

s b is the standard eviation of thef ackground counting rate or of -

tne minute), counting rate o a blank samp 4 as appropriate (as counts per E is the countiag effici cy s counts per transformation),

V is the sample size (in of mass or volume),

2.22x106 is the number tran o ns per minute per microcurie, Y is the fractional r diochemical i g when applicable),

A is the radioacti e decay constant particular radionuclide and for composite sa les, and '

l At is the elap ed time between midpoint o collection and time of cou 'ing (for plant effluents, not nmental samples).

For batch $ mples taken and analyzed prior t release, at is taken to bezero./ l

/ %

The va)he of bs used in the calculation of the L f detection syst shall be based on the actual observed vari c he back-gr d counting rate or of the counting rate of the 1 amples (a, appropriate) rather than on an unverified theore ca redicted 6riance. Typical values of E, V, Y, and At shall be e in the calculation.

b A composite sample is one in which the quantity of liquid mpled is proportional to the quantity of liquid waste discharged and ' which the method of sample employed results in a specimen which is representative of the liquids released.

LA SALLE - UNIT 1 3/4 11-4 Amendment No. 26

OE&T ENT/dC $6C ,

TABLE 4.11.1-1 (continued)

TABLE NOTATION 4

c. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow o the effluent stream. Prior to analyses, all samples taken to the ompositt shall oe thoroughly mixed in order for the co osite i 5 mple to be representative of the etfluent release.

d. A ba c ease is the discharge of liquid waste of a discrete volum. r to sampling for analyses, each bat i shall be isolate , nd then thoroughly mixed, by a metho' described in the 00CM, to s re representative sampling,

e. A continuou r se is the discharge of li id wastes of a nondiscrete lume e.g., from a volume of system that has an input flow during th c 'nuous release.

f. The principal ga ters for whic the LLD specification applies exclusively are the following radio clides: Mn-54, Fe-59, C0-58, Co-60, Zn-65, Mo-99, s-134, Cs-13 , Ce-141, and Ce-144. This 1ist does not mean that on these nu ides are to be detected and reported. Other peaks ich ar measurable and identifiable, at the 95% confidence level, to ther with the above nuclices, shall also be identified and reported.

7-A-

P m

7 I

l l

LA SALLE - UNIT 1 3/4 11-5 l

DELErg &pRg / AGE

. ,'D10 ACTIVE EFFLUENTS DOS LIMIT 1h' CONDITION FOR OPERATION s i 3.11.1.2 The dos- dose commitment to an individual from radios ive materials in 1 ui ffluents released, f rom each reactor unit, f om the site (see figure 5.1.'-1 1 be limited.

a. During a r idar quarter to less than or equal 1.$ mrem to the total body a to less than or equal to 5 mrem to any organ, and

b. During any ca e..; ear to less than or equaly o 3 mrem to the total body and o le.s than or equal to 10 mr m to any organ.

M ICABILITY: At all time N

ACTION:

a. With the calculated dos from the re 'ase of radioactive rnaterials in liquid effluents esce ing any of the above limits, in lieu of any other report required Specfication 6.6. A, prepare and submit l to the Commission within 30 ays pursuant to Specification 6.6.C, a special Report which identifi s the cause(s) for exceeding the limit (s) and defines the corr . ive actions to be taken to reduce the releases of radioactive mate ial in liquid effluents during the remainder of the current c endar var r and during the subsequent three calendar quarters, o that thA t. , ative dose or dose commit-ment to an individual fr m these reib is within 3 mrem to the total body and 10 mrem o any organ, etial Report shall also include the radiologi al impact on fini ed inking water supplies at the nearest down . ream drinking water our .

b. The provisions o Specifications 3.0.3 and h 0 % re not applicable.

SURVEILLANCE REOUIREP N15

'S 4.11.1.2 Dose C 1ations. Cumulative dose contributions from id effluents shall 31 days. e aetermined in accordance with the ODCM at least c @ er l

LA SALLE - UNIT 1 3/4 11-6 Amenoment No. 23

del.c TG~ EM/Kc R1GE

. RAD 10ACT]vE EF FlUENTS UlD WASTE 1REATMENT SYSTEM LIM 111 CONDil10N FOR OPERATION 3.11.1.3 T ht liguji rad aste treatment system shall be OP[RABLE. he appropriate po 'iow of the system shall be used to reduce the ra-ioactive materials in lic'id es prior to their discharge when the pr etted doses due to the liquid f lue t from each reactor unit, from the si e (see Figure 5.1.1-1), wh n eraged over 31 days, would exceed O. s rttem to the total body or 0.2 mr . organ.

APPLICABILITY: At all . mes.

ACTION:

a. With the liquid rad. Ste treatment syste- inoperable for more than 31 days or with radio tive liquid was being discharged without treatment and in excess of the above imits, in lieu of any other report required by Speci ication 6. .A, prepare and submit to the l Commission within 30 days ursuant .o Specification 6.6.0 a Special Report which includes the f lon* g information:

1. Identification of the in rable equipment or subsystems and the reason for inoperab' it -

2.

~

b Action (s) taken to r store the i E rable equiprent to Op[RABLE status, and A-

3. Summary descrip on of' action (s) ta ei prevent a recurrence.

b. The provisions of .,pecifications 3.0.3 and -

are not applicable.

4 SURVEILLANCE REOUlREMEiTS ,

]

4.11.1.3.1 Dos >duetoliquidreleasesshallbeprojectedakle. ce per 31 days, in at ordance with the ODCM.

4.11.1.3.2 he liquid rad *aste treattrent sy!. tem shall be demonstratec bPERABLE by operat ig the liquid radwaste treatment system equipment for at leas, i

30 trinut 5 at least once per 92 days unless the liquid rad aste system ha. been utiliz to process radioactive liquid effluents during the previous 92 day l

LA SALLE - UN11 1 3/4 11-7 Anenm ent No. 23

RADIDACTIVE EFFLUENT $

LIQUID HOLOUP TANKS ,,

LlHITING CONDITION FOR OPERATION I

3.11.1.4 The quantity of radioactive material contained in any outside temporary tanks shall be limited to less than or equal to the limits calculated in the ODCH.

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

, s_

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above 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.

1 m

LA SAL LE - UNIT 1 3/4 11,F';

W W W CN77Ab' lk((( ;

1 RA 10 ACTIVE EFFLUENTS ,

3/4 .2 GASEOUS EFFLUENTS 7 DOSE RA LIMITING DITION FOR OPERATION 3.11.2.1 The s due to radioactive materials released in gaseous effluents from t e see Figure 5.1.1-1) shall be limited to'the following:

a. For noble ga . Less than or equal to 500 mrem /yr'to the total body and I s or equal to 3000 mrem /yr to the skin, and

b. For all radio e and for all radioactive ma crials in particulate form and radiot cl (other than noble gases) with half lives greater than 8 s: ss than or equal to 1500 f mrems/yr to any organ via the inh a athway. l APPLICABILITY: At all times. -

/

ACTION:

With the dose rate (s) exceeding the hove 1)inits, immediately decrease the release rate to within the above limit 5).

SURVEILLANCE REQUIREMENTS s -

4.11.2.1.1 The dose rate due to ble gases ous effluents shall be determined to be within the abov limits in acc d @ with the methods and procedt!res of the ODCM.

4.11.2.1.2 The dose rate du to radioactive materi s o ther thar noble gases, in gaseous effluent shall be determined to be 4 the above limits in accr.rdance with the me) ods and procedures of the 0 ottaining represen-tative samples and perf (ming analyses in accordance wit - Aling and analysis program specified in T le 4.11.2-1.

N

~P

m LA SA'LE - UNIT 1 3/4 11-9 Amendment No.18 l

.m , _ , .8 =

TABLE 4.11.2-1

, RADI0 ACTIVE GASEQUS WASTE SAMPLING AND ANALYSIS PPDGRAM c- ' /_ .-

E , ; Minimum Lower .ait (M

, l Sampling Analysis { Type of Dete- ing (LLD)

E: , Gaseous ReleasAType l frequency Frequency ! Activity Analysis '

.;Ci/ml)'

~#

A. Containment Vent Eac Purge Eac Purge _9 Principal Gamma Emi +rg x10 and Durge System rab ,

/ -

a le , 3 H-3 / A ' k' 1x 10_ 6 B. Main Vent Stick 9 M l Principal mmaEmifErs ,

1. 10

!Sample C

I C

!H-3 [ M '

1x10

-6 t C. Standby Gas i D l W ,

~#

{' Treatment System { Grab ; jPrincipal Gama Emitters 9 ,

Ix10

! Sample I i ,

lD. All Release Types ! Continuous 1 -12 1-131 l 1x10 I as listed in A and l Cha oal l mple h -10

$ {'

B above, at the

{ II-133 f 1x10

$ ~II C ted n ,

' Continuous ;, rincipal Gamma Emitters 9! 1x10 i at the SBGTS ticulate ; 31, Others) !

,i hample

! whenever there 3 i # " '# 55

is flow. Contin us W l Composite

. i 1x10'

' l Particulate D:

i Sample I k{

~

ntinuous I ; Q !5r-89, Si-90 1x10 i Composite i

Q() s '

l Particulate Sample g

tContinuous Noble Gas . Noble Gases 0-6 (Xe-133 i ,

l Monitor Gross Beta & Gama equ. Tent)

N i

i

< ( i N'

MlET6 Eiv7t#E M6E TABLE 4.11.2-1 (Continued)

TABLE NOTATION

a. Te LLD is the smallest concentration of r'.dioactive material in a sam e th t will be detected with 95% probability with St. probability of f a ely con uding that a blank observation represents a "real" signal.

For a 1rticular measurement system (which may include radiochem' cal separati n):

4,66 s LLD = m b

' f/ 2. 22x10 Y exp (-AAt) .

Where:

LLD is the "a ri h lower limit of detection as defined above (as microcurie t'r mass or volume),

h is the standard eviation of the backgr und counting rate or of s

tne counting rate o a blank sample as a ropriate (as counts per minute), .

E is the counting effici icy (as co ts per transformation),

V is the sample size (in un 5 of mass or volume),

2.22x10' 0 is the number of tra ormations per minute per microcurie, Y is the fractional radioc mical @ when applicable),

A is the radioactive de y constant particular radionuclide, and at is the elapsed tir between midpoint f le collection and time of counting (f r plant effluents, n ei ronmental samples).

The value of 5 b

sed in the calculation of e LLD for a detection systemshallbIbasedontheactualobserved riance of the back-ground counti g rate or of the counting rate o the nk samples ily predicted (as appropr' te) rather than on an unverified th rt variance. ypical values of E, V, Y, and At shal be, sed in the calculat' n.

b LA SALLE - UNIT 1 3/4 11-11

WL6 TE ENR$E'~ h 66'~

TABLE 4.11.2-1 (Continued)

TABLE NOTATION ,

b. n s sha.1 also be performed following shutdown, startu' , or a E OWER change exceeding 15% of the RATED THERMAL P0 ER within l a r eriod,

c. Whene er - is flow through the SBGTS.

d. Samples al changed at least once per 7 days d analyses shall be comple d w; n 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing or af er removal from sampler. p hall also be performed at I st once per 24 l hours for at east 7 days following each shutd n, startup or THERMAL POWER change e eeding 15 percent of RATED T RMAL POWER in I hour I ,

and analyses to leted within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of c anging. When samples collected for 24 ours are analyzed, the tresponding LLD's may be increased by a fac r of 10.

This requirement does et apply if (3 analysis shows that the DOSE EQUIVALENT I-131 concen ration in t primary coolant has not increased more than a fa or of 3; 2) the noble gas monitor shows that effluent activity has ot i reased more than a factor of 3.

e. Tritium grab samples shall be aken at least once per 7 days from the plant vent to determine eleases in the ventilation e,.haust from the spent fuel poo whenever spent fuel is in the spent fuel pool.

f. The ratio of the sampi flow rate to pied stream flow rate shall be known for t time period cos y each dose or dose rate calculation made in ccordance with Spe ions 3.11.2.1, 3.11.2.2 and 3.11.2.3.

g. The principal a emitters for which the L sp ification applies include the f

~

lowing radionuclides: Kr-87, Xe-133, Xe-133m, i Xe-135, and e-138 for gaseous emissions and Mn Co-60, Zn- , Mo-99, Cs-134, Cs-137, Ce-141 and y for e-59, Co-58, parti-culate em ssions. This list does not mean that on e nuclides are to detected and reported. Other peaks which asurable and id ntifiable, at the 95% confidence levtl, togeth r wi h the abov nuclides, shall also be identified and reported.

l l

i l

L LA SALLE - UNIT 1 3/4 11-12 Amendment No. 18

DDE W &7?fE PMG' t'D10 ACTIVE EFFLUENTS /

00 - NDBLE GASES LIM] TING ONDIT]ON FOR OPERATION 'I o

3.11.2.2 The a' due to noble gases released in gaseous effl ents, from each reactor uni m the site (see figure 5.1.1-1) shall be l' aited to the follcaing: g 7

a. During any al n ar quarter: Less than or equal t[5 mr.ad f or gamma radiation an 1. than or equal to 10 mrad for hita radiation, and

/

b. During any cale ar year: tessthanorequal/o10mradforgamma rad'ation and les than or equal to 20 mra'J tr beta radiation.

APPLICABIL]TY: At all times.

ACTION:

a. With the calculated air dose from adioactive noble gases in gaseous effluents exceeding any of th a ve limits, in lieu of any other report required by Specificatio 6.6 prepare and submit to the l

Commission within 30 days, pu .u tt pecification 6.6.C. a Special Report which identifies the ause (r exceeding the limit (s) and defines the corrective act' ns to b takt to reduce the releases and the proposed corrective ions to be la.e to assure that subsequent releases will be in com' iance with th a limits.

b. The provisions of Sp ificat' ions 3.0.3 an 3 h are not applicable.

SURVEILLANCE REQUIREMENTS

\

4

- -7 4.11.2.2 Dose Calcu ations Cumulative dose contributions for current calendar quarter a current calendar year shall be determined 1 accordance with the ODCM at east once per 31 days. -

LA 5ALLE - UNIT 1 3/4 11-13 Amenment No. 23

0[LETE ENT/M fAGE R 910 ACTIVE EFFLUENTS DOS - RAD]O10 DINES RADIDACTIVE 2 MATER] ALS IN PAR 11CUL Alf FORM, AND RAD 10tWCLlDES OlHEkNHAN NDBLE CA5E5 ,/

LIM] TING CMDil10N FOR OPERATION

\

l' D

3.11.2.3 lhe de e ,

individual from radiciodines and radisactive rtaterials in particulate fo ,, an radionuclides, other than noble gas [, with half-lives greater than 8 day ir aseous effluents released, from eat reactor unit, from the site (see figure 5. 1) shall be limited to the fol wing: .

a. During any ca endaKquarter: Less than or e al to 7.5 mrems to any organ, and

b. During any talen. r yeaY: Less than or ual to 15 mrems to any organ.

APPLICABillTY: At all times.

ACTION:

a. With the calculated dose f r al he release of radiciodines, radioactive materials in particulate for or radionuclides (other than noble gases) with half lives gre er th 8 days, in gaseous effluents exceeding any of the abov limi 3, lieu of any other report required by Specificatio 6.6.A, r e and submit to the Commission within 30 days, pursua to Specif 6.6.C, a Special Report l which identifies the ause(s) for ex e the limit and detines the corrective actions 4at have been tak educe the releases and the proposed corre tive actions to be Le i o assure inat subsequent releases will be n compliance with the 6 % imits.

b. The provisione of Specifications 3.0.3 and .0.4 are not applicable.

~

SURVEILLANCE REQU EMENTS NT@ s . . .

4.11.2.3 Do e Calculations Cumulative dose contributions for t calendar q rter ano current calendar year shall be determined in\ current ccordance with the CM at least once per 31 days.

/

l LA SALLE - UNIT 1 3/4 11-14 Amendment No. 23

- - A =-A 4- . J - ,L--, .KA- 1J - ------.6 M.I, s- A -

4 2-"+-~~ --a rs,..-----s A ---'--s. 4 Ai4 - a -a .__m_,_.,J W E EA'Tf W f/16 E n D10 ACTIVE EFFLUENTS GA US RADWASTE TREATMENT SYSTEM LIMITING ONDITION FOR OPERATION D

3.11.2.4 The %5hRADWASTETREATMENTSYSTEMshallbeinoper ion.

APPLICABILITY: 6 kver the main condenser air ejector syste i s in operation. I ACTION: .

1

a. With the GAS -'J5 " 9 WASTE TREATMENT SYSTEM in erable far rnore than 7 days, in lie o >

other report require by Specification 6.6.A, l prepare and subm't to the Commission withi 30 days, pursuant to Specification 6.6. , a Special Report wh' h includes the following information:

1. Identification o the inoperabl equipment or subsystems and the reason for ino irability,

2. Action (s) taken to re.. ore he inoperable equipment to OPERABLE status, and

3. Summary description of at 'on(s taken to prevent a recurrence.

b. The provisions of Specif cations 0. nd 3.0 4 are not applicable.

A~

SURVEILLANCE REOUIREMENTS e

s Dqss 4.11.2.4 The GASEOUS ADWASTE TREATMENT SYSTEM shall b verified to be in operation at least o ce per 7 days. ,

G' LA SALLE - UNIT 1 3/4 13-15 Amer.3 ent No.?3

Daew EN7?2e- paw ADIDACTIVE EF.tLUENTS V ILATION EXHAUSl 1RE ATMENT SYSTEM

~

LIMITIN CONDIT]ON FOR OPERATION

\ /

/

l 3.11.2.5 The , g riate portions of the VENTILATION EXHAUST TREATP NT SYSTEM shall be OPERA . n be used to reduce radioactive materials in gpeous waste prior to their 's aaroe when the projected doses due to gaseous pf fluent releases from eac r spr unit, f rom the site (see Figure 5.1.VI), when averaged over 31 da,s d exceed 0.3 mrem to any organ.

APPLICABIllTY: At al t ACTION: /

a. With the VENTILAT N EXHAUST TREATMENT SYSTf/1 inoperable f or more than 31 days, or wi gaseouswastebeingfischargedwithout treatment and in ext s of the above limTs, in lieu of any other report required by Spe ification 6.6.A, repare and submit to the l Com9ission within 30 day , pursuant t Specification 6.6.C, a Special Report which includes the following iformation:

1. Identification of the i pe le equipment cr subsy:tems and the reason for inoperabi 't ,

2. Action (s) taken to resto e e inoperable equipment to OPERABLE status, and

3. Summary description f action (s) to prevent a recurrence.

b. The provi. cions of Spe (fications 3.0.3 a- .4 are not applicable.

h\

SURVEILLANCE REQUIREMEMS ,

x l 4 4.11.2.5.1 Dosesdu.togaseousreleasesfromthesiteshall.'T e jetted at least once per 31 ys in accordance with the ODCM.

4.11.2.5.2 The ENTILATION EXHAUST TREATMENT SYSTEM shall be demo strated OPERABLE by op f[atin;, the VENTILATION TREATMENT SSSTEM EXHAUST equip,ent for at least 30 m nutes, at least once per 92 days unless the appropriate ystem has been L.t'lized to process radioactive gaseous ef fluents during the evious 92 days.

)

LA SALLE - UNIT 1 3/4 11-16 Amenc ent No. 23

l l

l RADIOACTIVE EFFLUENTS !

EXPLOSIVE GAS MIXTURE LIMITING CONDITION F09 OPERATION 3.11.2.6 The concentration of hydrogen in the asin condenser offgas treatment system shall be limited to less than or equal to 4% by voir.me.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTION:

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVE!'f.ANCEREQUIREMENTS 4.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limits as required by Table 3.3.7.11-1 of Specification 3.3.7.11.

i LA SALLE - UNIT 1 3/411,Ff[

RADI0 ACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.7 The release rate of the sem of the activities from the noble gases measured prior to the holdup line shall be limited to less than or equal to 3.4 x 105 microcuries/second.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

AClION:

With the release rate of the sum of the activities of the noble gases prior to the holdup line exceeding 3.4 x 105 microcuries/second restore the release rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP with the main steam isolation valves closed within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

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

a. At least once per 31 days,

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

LA SALLE - UNIT 1 3/4 11-pfg l

MLETE &v7/dC / AGE DI0 ACTIVE EFFLUENTS VE ING OR PURGING.

LIMITIN CONDITION FOR OPERATION

/

3.11.2.8 VEN N URGING of the containment drywell shall be rough the Primary Contain.e nt and Purge System or the Standby Gas Tr tment System.

APPLICABILITY: Wh n the drywell is vented or purged.

ACTION:

a. With the requi e of the above specifica on not satisfied, suspend all VEN PURGING of the dryw 1.

b. The provisions of ecifications 3.0.3 a 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

, s ,

4.11.2.8.1 The contaireent drywell sh

/ be determined to be aligned for VENTING or PURGING thcough the Prima Co ainment Vent and Purge System or the Standby Gas Treatment System wi in 4 urs 'or to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING o PURGING rywell.

4.11.8.2 Prior to use of the P rge System thro h q Standby Gas Treatment System in OPERATIONAL CONDITI 1, 2 or 3 assure hatA-

a. Both Standby Gas reatment System trains e

/ ABLE, and

b. Only one of t Standby Gas Treatment System r is used for PURGING.

A T

t LA SALLE - UNIT 1 3/4 11-19 Amendment No.18

l. . . . _ . . ...__._ .- __

'010 ACTIVE EFFLUENTS 3/4.'1.3 SOLID RADIDACTIVE WASTE

~

LIMITING CONDITION FOR OPERATION e

3.11.3 The so id ru aste system shall be OPERABLE and used, as a alicable in accordance with F. 5 CONTROL PROGRAM, for the 50L101FICAT10F and packaging of radioactive w- tes . ensure meeting the requirements of 10 .R Part 20 and of 10 CFR Part 71 rior *, shipment of radioactive wastes fror the site.

APPLICABILITY: At al time ACTION:

a. With the packagi requirements of 10 CFR P rt 20 and/or 10 CFR Part 71 not satisi d, suspend shipments defectively packaged solid radioactive w .tes from the site.

b. With the solid radwasti. system inoper le for more than 31 days, in lieu of any other report required b Specification 6.6.A, prepare and submit to the Commission ithin 30 days, pursuant to Specifi- l cation 6.6.C. a Special Re rt wt ch includes the following information:

1. Identification of the i op'rable equipment or subsystems and the reason for inoper ilit 2.

.b Action (s) taken to estore the ina;u.rable equipment to OPERABLE status, b

3. A description. the alternative u .d .f-cr SOLIDIFICATION and packaging of adioactive wastes, an y

, - 4. Summary d scription of action (s) taken ohvent a recurrence.

c. The provisi s of Specifications 3.0.3 and 3.0.- are not applicabic.

SURVEILLANCE REQU .EMENTS T l O' I 4.11.3.1 T solid radwaste system shall be demonstrated OPERABL a 1>ast once per 9 days by:

a. Operating the solid radwaste system at least once in the pr 'ious 92 days in accordance with the Process Control Program, or

b. Verification of the existence of a valid contract for 50LIDIFICA'10N to be performed by a contractor in accordance with a PROCESS CONT 7L PROGRAM.

l LA SALLE - UNIT 1 3/4 11-20 Arenaient No. 23 l

PELETE EN7tKt PAGE RADI0 ACTIVE EFFLUENTS

/

Suk EILLANCE REQUIREMENTS (Continued)

/

4.11.3.2 THEPROCESSCONTROLPROGRAMshallbeusedtoverifythe/

SOL 101 FICA ON of t least one representative test specimen f ront/ at least every tenth at o ~ each type of wet radioactive waste (e.g. ,/ filter sludges, spent resins, va tor bottoms, and sodium sulfate soluti ') .

a. If any <:e(specimen f ails to verify SOLIDIFICAT 'ON, the !

SOLIDIFf ATi f the batch under test shall b suspended until such time as a di 1 test specimens can be obt ned, alternative SOLIDIFICA 0 pm meters can be determined in accordance with the PROCESS CON 1 L R RAM, and a subsequent est verifies SOLIDIFICA-T10N. SOLIDI CA of the batch may t en be resumed using the alternative 50L 1 ION parameters termined by the PROCESS CONTROL PROGRAM.

b. If the initial test pecimen from batch of waste fails to verify SOLIDIFICATION, the P CESS CONT L PROGRAM shall provide for the i collection and testing f repre entative test specimens from each consecutive batch of the ame ype of wet waste until at least 3 consecutive initial tes s ecimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM sh I be modified as required, as provided in Specification 6.7, to ss 'e IDIFICATION of subsequent batches of waste.

1 A.-

i l

3 G

l TN l

i l

l l LA SALLE - UNIT 1 3/4 11-21

METE ~ EA/T/R6~ fAGE

, RADIDACTIVE EFFLUENT 5

.31.4 10TAL 005E LIMIT G CONDITION FOR OPERATION _

r ,

3.11.4 The ose or dose commitment to any member of the public, du to releases of radioactiv'ty Aradiation, from uranium fuel cycle sources sigil be limited to les the thyroid, wh1 thw h r equal be to 25 mrem limited to the to less thantotal body to or equal or 7afn organover mrem) (except 12 consecutive mo th .

APPLICABILITY: At a 1 i.

ACTION: ,

a, With the calcula ed c . s from the releasep f radioactive materials in liquid or gase s uents exceeding wice the limits of Specifica-tions 3.11.1.2.a, 11.1.2.b 3.11.2.2. , 3.11.2.2.b, 3.11.2.3.a. or 3.11.2.3.b in lieu f any other repor required by Specification 6.6.A, l prepare and submit, p suant to Spec'tication 6.6.C, a Special Report to the Director, Nuclea ' Reactor R , ulation, U.S. Nuclear Regulatory Commission, Washington, C. 2055 , within 30 days, which defines the corrective action to be ta en t reduce subsequent releases to prevent recurrence of excee in the limits of Specification 3.11.4.

This Special Report shall in ude an analysis which estimates the radiation exposure (dose) t a member of the public from uranium fuel

cycle sources (including di ef uents athways and direct radiation) l for a 12 consecutive monM perio th -ludes the release (s) l covered by this report / If the es 'ma . dose (s) exceeds the limits

! of Specification 3.11/4, and if the el condition resulting in violation of 40 CFR 90 has not alrea ' be corrected, the Special Report shall inci e a request for a va 1 ie in accordance with the provisions of 4 FR 190 and including t. rpecified information of K 190.11. Subm' tal of the report is cons' der . a timely request, and a varianc is granted until staff actio or e request is complete. T e variance only relates to the 'm1 f 40 CFR 190, and

! does not a ly in any way to the requirements r dose limitation of 10 CFR P t 20, as addressed in other sections this technical specifi tion. ,

b. The ovisionsofSpecifications3.0.3and3.0.4arenMpplicable.

D l

SURVEILLAN* REQUIREMENTS

- s l

4.11 Dose Calculations Cumulative dose contributions f rom liquid an ga ous effluents shall be determined in accordance with Specifications 4 11.1.2, 4 1.2.2, and 4.11.2.3, and in accordance with the ODCM.

LA SALLE - UNIT 1 3/4 11-22 Amenc ent No. 23

00 &~ F.A/7/fC /%6E 3] g ' RADIOLOGICAL ENVIRDNMENTAL MDN110 RING

~ ~ ~

3/4.2A1 MONITORING PROGRAM LIMITING 'NDITION FOR OPERATION 1 /

3.12.1 The diological environmental monitoring program shall b conducted as specified 1 Tab 3.12.1-1.

APPLICABILITY: agimes.

3T10N: .

t

a. With the rad log i environmental monitorin program not being conducted as s ecifi d in Table 3.12.1-1, i lieu of any other report required y ification 6.6.A, pr pare and submit to the l Commission, in th nnual Radiological Op rating Report, a description of the reasons for not conducting the p gram as required and the plans for preventing a recurrence.

b. With the level of radi ctivity in n environmental sampling medium exceeding the reporting .vels in able 3.12.1-2 when averaged over any calendar quarter, in 'eu of any other report required by Specifi-cation 6.6.A, prepare and s m' to the Commission within 30 days i from the end of the affected alendar quarter a Special Report i pursuant to Specification 6 . .13. When more than one of the radionuclides in Table 3.1.1-2 re detected in the sampling medium, this report shall be sub tied i concentration (1

concentrati ) +

. Q 1. 0 limit level (1) limit level 2 When radionuclide other than those in T I I 2.1-2 are detected and are the res .t of plant effluents, th1 rt shall be submitted if the potenti annual dose to an individu 1 . qual to or greater than the cal dar year limits of Specificati s 3.11.1.2, 3.11,2.2 and 3.11.2. . This report is not required if e measured level of radioactiv ty was not the result of plant efflu ts-Awever, in such an ent, the condition shall be reported an de bed in the Annual ,adiological Environmental Operating Report.

c. Wit milk o, fresh leafy vegetable samples unavailabl G f

', one or mo - of the sample locations required by Table 3.12.1- ieu of a y other report required by Specification 6.6. A, prepar and submit

.o the Commission within 30 days, pursuant to Specificati 6.6.C, a Special Report which identifies the cause of the unavailabi ity of samples and identifies locations for obtaining replacement s ples.

The locations from which samples were unavailable may then be leted from those required by Table 3.12.1-1, provided the locations f m which the replacement samples were obtained are added to the envi on-mental monitoring program as replacement locations.

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

LA SALLE - UNIT I 3/4 12-1 Amendment No.23

W WTc GhT/Ke~ $ AGE

.DIOLOGICAL ENVIRONMENTAL HONITORING SURV LLANCE REQUIREMENTS s , -

4.12.1 The radiological environmental monitoring samples shall be c lected pursuant to ble 3.12.1-1 from the locations given in the table a figure in the ODCM and s all be analyzed pursuant to the requirement tf Tab s 3.12.1-1 and 4.12.1-1.

'CD GN C

A ex

'x z_

\ .A e

eX m

V P

UN LA SALLE - UNIT 1 3/4 12-2 l

i

TABLE 3.12.1-1 g RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM P

m i

Number of Samples g Exposure Pathway and Sampling and Type and equency g and/or Sample Sample locations

Collection frequency of 4nalysis

1. AIRBORNE Radioiodine and - ocations Continuous operation of f iodine canister.

Particulates sampler with sample col l lection as required

/ lyze at least once

per 7 days for I-131.

dust loading but < 1eas once per 7 days Particulate sampler.

inalyze for gross beta radioactivity 3 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months R following filter change.

Perform gamma isotopic Z analysis on each sample 0 when gross beta activity is > 10 times the yearly mean of control samples.

Perform gamma isotopic p hh analysis on composite (by location) sample Mg at least once per 92 days.

D h

2. DIRECT RADIATION '

ocat'ons At least once per 31 days. - Gamma dose. At least m s

dosimeters or 1 1 or ce per 31 days. lq rument for con- or g

E nuously measuring At least once per 92 days. Gamma ose. At least g 3 and recording dose (Read out frequencies are once per ? days. t

& rate at each determined by type of dosim-g location. M eters selected.) A

=

o S a

S

'4 TABLE 3.12.1-1 (Continued) 5 m RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

?

E . .

, Minimum c Number of Samples 5 Exposure Pathway and Sampling and H and/or Sample ype and Frequency Sample Locations

Collection Frequency of Analysis O 3. WATERBORNE

a. Surface 21 tions Composite sample le Gamma isotopic analysis over a period < ys. of each composite sample.

Tritium analysis of com-posite sample at least

.; once per 92 days.

I R

b. 5 locations Ground t least once per 92 days. Gamma isotopic and j g tritium analyses of 4 -

each sample.

c. Sediment from 1 location At least on per 184 days. Gamma isotopic analysis i Shoreline of each sample.

1

[ S 4;'

r 4b m i

. B 4' s

r+

F S 1:

a D

i-

\

TABLE 3.12.1-1 (Continued) .

! E '

l w RADIOLOGICAL'ENVIROIO4 ENTAL MONITORING PROGRAM i 5 ;

j- E

! . Minimum c Number of' Samples 5 Exposure Pathway and ,

Sampling and T Frequency

-* and/or Sample le Locations

Collection Frequency -

f Analysis i i

4. INGESTION 1

l a. Milk 3 locations At least o e per 15 ys Gamma isotopic and:

when an~ is are on pasture; I-131 analysis-at I t once per 31 days of each sample.

'. a ther times.

j t 5:' b. Fish 2 locations sample in season, or at Gamma isotopic analysis-1

lea once'per 184 days if on edible portions.

C not se .onal. l

' I j

Sample locations are descri in the ODCM. ki c ko, i l 9)N 1>

.S 1

I

TABLE 3.12.1-2 5~

REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting levels E

C Wate Airborne Particulate Fish Milk ood Products

[ Analysis (PCi/1 .. or Gases (pCi/m 3) (pCi/Kg, wet) (pCi/1 (pCi/Kg, wet) 9h H-3 2x 10 4I )

3 4 Mn-54 1 x 10 3 x 10 Fe-59 4 x 10 2 1x 3 4 Co-58 1 x 10 3 10 Co-60 3 x 10 2 1 10 4

7 a-65 3 x 10 2 \ 2x 10 4

2 G Z r-Nb -95 4 x 10 S

I-131 2 0.9 t's - 134 2

30

& 1 x 10 3 3

60 1x 10 1x 10 3

3 Cs '37 50 ZD 2 x 10 70 2 x 10 3 B a- L a- 14') 2 x 10 2 Qh 3 x 10 2 k 99 'N s 4

a)for d nking water samples. This is 40 CFR Part 141 value. \N k d

'x @'

3 m

( ( n

TABLE 4.12.1-1 5~

y, MAXIMUM VALUES FOR THE LOWER LIMIls 0F DETECTION (LLD)a,c

?

R

/

E h er Airborne Particulate Fish Milk Food Pr acts Sediment (pCi/ , wet) (pCi/kg, dry)

Z Analysis (PCi ) or Gases (pCi/m3 ) (pCi/Kg, wet) (pCi/1)

-2 NA 2000 gross beta +5 1 x 10 1000 NA

- NA NA NA H-3 200 NA Mn-54 NA Fe-59 NA Co-58,60 NA

$ Zn-65

NA

~

N ** * ** **

7 Z r-95 NA s

Nb-95 NA I-131 NA x

- 0.5 30 10

-2 100 Cs-134 10 1 x 10 -2 Cs-13/ 10 100 10 Di Ba-140

NA La-140 NA N

C ..ma isotopic analysis provides LLD of s 20pCi/l per nuclide. q Ganca isotopic analysis provides LLD of s20 pCi/1 per nuclice.

e

_ . ._ __ _ _.. _ _ _ _ _. _. _._ _ _. _ ~_ ._.___m _

NW7f &V7/fT $ AGE TABLE 4.12.1-1 (Continued)

TABLE NOTATION

/

a. -The LLD is the smallest concehtration of radioactive material i a sample.

' that will be detected with 95% probability with 5% probability of falsely cor.Cluding that a Diank observation represents a "real" sign .

For-a particular measurement system (which may include ra iochemical separation):

4.66 s b E.V 2.22 Y . exp (-Aat)

Where:--

LLO 1 t a priori" lower limit of d ection as defined above (as picoeur per unit mass or volume),

si si the tandard deviation of t background counting rate or of tne countin rate of a blank sam e as appropriate (as counts per minute),.

E is the countin efficienc- (as counts per transformation),

V is the sample siz (in nits of mass or volume),

2.22 is the number of ransformations.per minut3 per picocurie, Y is the fractiona radio hemical yield (when applicable),

s the idioa ive decay c.n t for the particular radionuclide, at is the apsed time between s llection (or end of the sample co ection period) and time o nting (for environment I

l.

samples not plant effluents.

i The value fs bused in the calculation of e LLD for a detection system shall be ased on the actual observed varian 4 e background counting rate o of the counting rate of the blank samp e appropriate) ratner than n an unverified theoretically predicted v i In calculating the LD for a'radionuclide determined by gamma-ra ometry, the b kground sh ll include the typical contributions f r racionuclides rmally present in the samples (e.g. potassium-40 i milk samples).

ypical values of E, V, Y, and at shall be used in th calculations.

l

b. LLD for drinking water.

c. Other peaks which are measurable and identifiable, toget er with tne radionuclides in Table 4.12-1, shall be identified and re rted.

LA SALLE - UNIT 1 3/4 12-8 I ~ , - .- ,,. , _ - , . , . ~ , , , . , . - , -, - , . - ,

06LCTE~ EA./77g[ fAGt i

I i

RAL OLOGICAL ENVIRONMENTAL MDNITORIN3 3/4. 2 LAND USE CENSUS !

LIMITING NDITION FOR OPERATION

/

3.12.2 A lan usecensusshallbeconductedandshallidentify,dhelocation of the nearest . '1k animal and the nearest residence in each gf the 16 teteor-ological sectors ithin a distance of five miles. (for elevpted releases as defined in Regulat y tXe land use census shall also identify he deations 1.111,ofRevision all milk 1, July 1977, animals in 6th of the 16 meteoro-logical sectors withi d tance of three miles.)

APPLICABILITY: At all t M A

ACTJON: g

a. With a land use cens identifying a lbcation(s) which yields a calculated dose or dos commitment gr' eater than the values currently being calculated in Spe ification t/11.2.3, in lieu of any other report required by Speci ' cation V,6. A. , prepare and submit to the l Commission within 30 days, nurstdnt to Specification 6.6.C. , a Special Report which identi 'e 'the new location (s),

b. f> 'ng a location (s) which yields a With a land calculated doseuse census or dose ccym iden)itmeht (via the same exposure pathway) 20 percent greater than K a locahon m which samples are currently being obtained in accor ante with she ation 3.12.1, in lieu of any other report requi ed by Specifi6 4t .6.A., prepare and l

submit to the Commis onwithin30dayh cant to Specifica-tion 6.6.C. , a Spec'al Repo'rt which ide i ps the new location.

The new location all be added to the ra io @ cal environmental monitoring progr .., within 30 days. The sa. 1fr location, excluding

! culated dose or

- the dosecontrol stg (ion commitmerit vialocation, the same having thepathway exposure low A cmay be deleted f om this monitori'ng program after (October 31) of t year in which this d AO landusecfsuswasconducted.

c. Theprov[sionsofSpecifications3.0.3and3.0.4areno ?plicable.

SURVEILLANCE RE REMENTS b 4.12.2 Th land use census shall be conducted at least once per 12 m ths

. between t e dates of (June 1 and October 1) using that information whi will l provide he best results, such as by a door-to-door survey, aerial surve or by con iting local agriculture authorities.

1 LA SALLE - UNIT 1 3/4 12-9 Amendment No.23

MLETE &TMC/ AGE RADIOLOGICAL ENVIRONMENIAL MONITORING 3/4.12.3- INTERLABORATORY-COMPARISON PROGRAM LIM TING CONDITION FOR OPERATION X /

1 3.12.3 -A 1 shall be performed on radioactive materials sup ied as part of an Inter ory Comparison Program which has been approv by the ,

Commission, '

q{"

APPLICABILITY: t al times.

ACTION:

a. With analyse not being performed as requi ed above, report the

_ corrective ac 'ons taken to prevent a re .rrence to tne Commission in the Annual diological Environment Operating Report,

b. .The provisions of ecifications 3. 3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS x <

4.12.3 A summary of the results ob ain d art of the above required Inter-laboratory Comparison Program an n acc with the 00CM (or participants.

in the EPA crosscheck program s 1 provi t % EPA program code designation for the unit) shall be include in the~Annu m (ological Environmental Operating-Report. Aa-- >

Q 7

C

\m LA SALLE - UNIT 1 3/4 12-10

l l

1NSTRUMENTATION BASES 3/3.3.7.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to

' monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensuts that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria t 60, 63, and 64 of Appendix A to 10 CFR Part 50. )

3/4. 3.7.11 { RADI0 ACTIVE GASEOUS EFFLUENT,

_ TORING INSTRUMENTATION I S The radioactive gaseous ef fluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The )

alarm / trip setpoints for these instruments shall be calculated in accordance

( with the_ procedures in the ODCM to ensure that the alarm / trip will occur prior g

to exceeding the limits of 10 CFR Part 20./ This instrumentation :1:: S:ld:gred, o R .;. %,n for monitoring (and controlling) the concentrations of ootentially explosive gas mixtures in the waste gas holdup system.T The OPERABILITY and ruse of this instrumentation is consistent with the requirements of General i Design Criteria'60, 63 and 64 of Appendix A to 10 CFR Part 50. f 3/4.3.7.12 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that suf ficient I capability is available to detect loose metallic parts in the primary system '

and avoid or mitigate damage to primary system components. The allowable out- H of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1.133, " Loose-Part Detection Program-for the Primary System of Light-Water-Cooled Reactors."

3/4.3.8 FEEDWATER/?4AIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate the feedwater system / main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 setpoint associated with a feedwater controller failure, to prevent overfilling the reactor vessel which may result in high pressure liquid discharge through the safety / relief valve discharge lines.

LA SALLE - UNIT 1 8 3/4 3-6 Amendment No. 61

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

96LETE ENf/G Ph 66 4.11~ RADI0 ACTIVE EFFLUENTS BA 3/4.11.1 IQUID EFFLUENTS 3/4.11.1.1 NCENTRATION This speci cation is provided to ensure that the concentrq ion of radio-active materials el in liquid waste effluents from the s te will be less than the concentra o els specified in 10 CFR Part 20, A endix B, Table II, Column 2. This lim a provides additional assurance th the levels of-radioactive exposure materials within (1) th Set in W ies of water outside the site ill result in n II.A design objectives o Appendix I, 10 CFR 50, to an individual, and (2 t imits of 10 CFR 20.106(e to the population.

The concentration limits r di solved or entrained 9 le gases were determined by converting their MPC's i ar o an equivalent to centration in water using the methods described in Int n nal Commission Radiological Protection (ICRP) Publication 2.

3/4.11.1.2 00SE This specification is provided t im ement the requirements of Sections II.A. III.A and IV.A of Appen i I, 10 CFR Part 50. The Limiting Condition for Operation implements to es set forth in Section II.A of Appendix 1. The ACTION statements pr ide the required operating flexibility and at the same time implement the ides s forth in Section IV.A of Appendix I to assure that the rele es of ra 'o e material in liquid-effluents will-be kept "as low as is reasonabl vable." Also, for fresh water sites with drinking water, upplies which n @ otentially affected by plant operations, there is re s'onable assurance t tt operation of the facility will not result in dionuclide concentrat the finished drinking water that are in excess of he requirements of 40 C 14-1. The dose calcula-tions in the ODCM impleme the requirements in Secti of Appendix I that conformance with th guides of Appendix I be shown culational procedures based on mo is and data, such that the actua e re of an individual through ap opriate pathways is unlikely to be u )s ntially under-estimated. The equa ions specified in the ODCM for calcula ng the doses due to the actual.reley'se rates of radioactive materials in liqui effluents are consistent with the methodology provided in Regulatory Guide 1. QCalculation ofAnnualDoses/oManfromRoutineReleasesofReactorEffluent fVthe Purpose of Ev Tuating Compliante with 10 CFR Part 50, Appendix I, R ion 1, October 1977 nd Regulatory Guide 1.113 " Estimating Aquatic Disper i Effluents f om Accidental and Routine Reactor Releases for the Purpo f Implement g Appendix I," April 1977.

T s specification applies to the release of radioactive materials 1 )

liqui effluents from each reactor at the site. For units with shared rad ste tre ent systems, the liquid effluents from the shared system are proporti ed am g the units sharing that system.

/

LA SALLE - UNIT 1 B 3/4 11-1 Amendment No. 1 8

p. _ . . . _ _ _ _ . _ _ _ _ . . _ _ . . . . . . ~ . _ .

i RADICACTIVE EFFLUENTS s.

BASES g,it, t uwib ErrutEt4TS 7

3/4.11.1.3 LIQUID WASTE TREATMENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid ef fluents will be kept "as low as is reasonably achievable." During extended shutdown or low power operation, i.e., > 92 days, when steam is not available to the concentrators, Surveillance Requirement 4.11.1.3.2 may be extended to 180 days. This specification imple-ments the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of JR f

( Appendix 1,10 CFR Part 50, for liauid ef fluents.

f3/4.11.1.4 LIQUID HOLDUP TANKS 4 Restricting the quantity of radioactive material contained in the specified

(

tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of s 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water (supply and the rearest surface water supply in an unrestricted area.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specificaticn is provided to ensure that the dose at any time at the site boundary from gaseous ef fluents from all units on the site will be within the annual dose limitt of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20,

[ Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous ef fluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

l for individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equel to 500 mrem / year to the total body or to less than or eaual to 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate aoove background to an -

~

LA SALLE - UNIT 1 B3/411-[

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DI0 ACTIVE EFFLUENTS BASE 1

DOSE RATE ontinued) infant via the cow-milk-infant pathway to less than or equal to 00 mrem /

year for the ne est cow to the plant.

This specific D 1

lies to the release of radioactiv effluents in gaseous effluents ft m 1 eactors at the site. For unit within shared radwaste treatment sy he gaseous effluents from th shared system are proportioned among the n aring that system.

b 3/4.11.2.2 DOSE - NOBLE GA i

This specification is pro ided to implement ne requirements of Sections II.B. III.A and IV.A o Appendix I, 10 FR Part 50. The Limiting Conditions for Operation are the ides set f th in Section II.B of Appendix I.

The ACTION statements provide the quired o rating flexibility and at the same time implement the guides set th i ection IV.A of Appendix I to assure that the releases of radioacti m' erial in gaseous effluents will be kept "as low as is reasonably achievabl The Surveillance Requirements implement the requirements in Section A of Appendix I that conformance with the guides of Appendix I be show by 1 tional procedures based on models and data such that the actual exposu a individual through appro-priate pathways is unlikely to be bstantial W estimated. The dose calculations established in the O M for calcu tin e doses due to the actual release rates of radioac ve noble gases 'n ous effluent, are consistent with the methodolo provided in Regul o" de 1.109, " Calculation of Annual Doses to Man from utine Releases of Rea to luents for the Purpose of Evaluating Compl' nce with 10 CFR Part 50, A ix I," Revision 1, October 1977 and Regulator Guide 1.111, " Methods for timating Atmospheric Transport and Dispersion f Gaseous Effluents in Routin Re s from Light-Water Cooled Reactors," evision 1, July 1977. The ODCM qu s provided for determining the a' doses at the site boundary are bas n the nistori-cal average atmosphe c conditions. p 3/4.11.2.3 DOS - RADIOI0 DINES, RADIDACTIVE MATERIALS IN PARTIC ATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES The sp ification is provided to implement the requirements of ctions II.C.

III.A and .A of Appendix I, 10 CFR Part 50. The Limiting Conditions or Operatio are the guides set. forth in Section II.C of Appendix I. The TION stateme s provide the required operating flexibility and at the same tim.

imple nt the guides set forth in Section IV. A of Appendix I to assure that the > leases of radioactive materials in gaseous ef fluents will be kept "as low as is easonably achievable." The ODCM calculational methods specified in t . Surve1;'ance Requirements implement the requirements in Section III.A of LA SALLE - UNIT 1 B 3/4 11-3 Amendment No.18

__ ~ - _

RADI0 ACTIVE EFFLUENTS BASES %

DOSE-RADIOI0 DINE [,RADI0ACTIVEMATERIALSINPARTICULATEFORMANDRADIONUCLIOES OTHER THAN NOBLE GASE5 (Continued)

Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestilrated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compiiance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Reguiatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determin-ing the actual doses based upon the historical average atmospheric conditions.

The release rate specifications for radiciodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consump-tion of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. -

3/4.11.2.4 AND 3/4.11.2.5 GASEOUS RADWASTE TREATMENT SYSTEM AND VENTILATION EXHAUST ~ TREATMENT SYSTEM l

The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of }

radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This cpecification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B QndII.CofAppendixI,10CFRPart50,forgaseouseffluents.

L 3/4.11.2.6 EXPLOSIVE GAS MIXTURE The specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is -

maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. f' LA SALLE - UNIT 1 8 3/4 11-p 9 9N#' N , ^P m wcomktt b v

RADI0 ACTIVE EFFLUENTS BASES ,

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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 j 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 Aj to 10 CFR Part 50, 3/4.11.2.8 VENTING OR PURGING This specification provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas.

3/4.11.3 SOLIO RADI0 ACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid /

solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190.

The specification requires the preparation and submittal of a Special Report j

whenever the calculated doses from plant radioactive effluents exceed twice l

the design objective doses of Appendix 1. For sites containing up to l

4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dene limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe I

i a course of action which should result in the limitation of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For

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TOTAL OSE (Continued) the purp es of the Special Report, it may be assumed that the se commitment to the mem er of the public from other uranium fuel cycle sourdes is negligible, with the ex ption that dose contributions from other nuclea fuel cycle facilities at the same site or within a radius of 5 miles st be considered, If the dose to ae.ber of the public is estimated to e eed the requirements of 40 CFR 190, t e -

ial Report with a request for a v riance (provided the release condition ing in violation of 40 CFR 19 have not already been corrected), in acc da with the provisions of 40 R 190,11, is considered to be a timely reque t nd fulfills the requirement of 40 CFR 190 until NRC staff action is compi e n individual is not nsidered a member of the public during any perio tich he/she is enga d in carrying out any opera-tion which is part of th n ar fuel cycle.

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3/4. 2 RADIOLOGICAL ENVIRONMENTAL MONITORING ~ /

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3/4.12.1 MON ORING PROGRAM The radiolo cal monitoring program required by this specificgtion provides measurements of ra iatto and of radioactive materials in those eyposure path-ways and for those di lides which lead to the highest pote . 'ial-radiation exposures of individ is lting from the. station operation. This monitoring program thereby supple n e radiological effluent monitor' g program by

. verifying that the meas ab7concentrationsofradioactive aterials and levels of radiation are n ? .r than expected on the ba s of the effluent measurements and modeling nvironmental exposure p thways. The ini-tially specified monitoring o will be effective f at least the first 3 years of commercial operation, as ined in the ODCH.

The detection capabilities quir d by Table 4 2-1 are state of-the-art for routine environmental measurem nts in industri laboratories. It should

- be recognized that the LLD is-defin as an "a p ori" (before the fact) limit representing the capability of a meas rement sy tem and not as "a posteriori" (after the fact) limit for a particula measur ment. Analyses shall be per-formed in such a manner that the stated LDs ill be achieved under routine conditions. Occasionally background flue tions, unavoidably small sample sizes, the presence of interfering nuclid or other uncontrollable circum-stances may render these LLDs unachievab e. n such cases, the contributing factors will be identified and describ int. 1 Radiological Environ-mental Operating Report.

,T 3/4.12.2 LAND USE CENSUS This specification is pro ded to ensure that cha e i the use of unrestricted areas are identi ied and that modifications t monitoring program are made if require by the results of this censu The best survey information from the door- o-door survey, aerial survey or onsulting with local agricultural autho ties shall be used. This census s l requirements of Section V.B.3 of Appendix I to 10 CFR Part 5 is g the 7

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3/4.12.3 NTERLABORATORY COMPARISON PROGRAM The req irement for participation in an Interlaboratory C parison Program is provided t ensu that independent checks on the precisio and accuracy of the measurement dioactive material in environmental s ple matrices are performed as par o quality assurance program for env ronmental monitoring in order to demons ra the results are reasonably lid.

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l AC'"N! Sij A,i!VE CONTROLS FL A'ii C4ERM 4NG PROCEDURES AND PROGRAMS (Continueo)

I. The following programs shdll be established, implemented, and maiatained:

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

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

a.

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

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

2, In-Plant Radiation Monitorino A program which will ensure the capability to accurately determine the airborne iodine concentration in vital creas under accident conditions. This program shall include the following:

a. Training of personnel,

b. Procedures for monitoring, ancf

c. Provisionr for maintenance of sampling and analysis equipment.

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

~ shall include the following:

j a. Training of personnel.

b. Procedures for sampling and analysis, i c. Provisions for maintenance of sampling and analysis equipment.

( &

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

a. The Commission shall be notified and a Licensee Event Report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

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

f6NuM 6W j LA SALLE UNIT 1 6-)/ pgc As Amendment No. 66 6tPRoMLATE

INSUIT D

4. Radioactive _ Ef fluent. Contro1LProgram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS Of THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating prgcedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. 1he program shall include the following elements:

a. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

c. Monitoring, samp1tng, and analysis of radioactive 11guld and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

d. Limitations on the annual and quarterly doses or dose commitment to a MEMBER Of THE PUBLIC from radioactive materials in 11guld effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix ! to 10 CFR Part 50,

e. Determination of cumulative and projected dose contributions from radioactiva 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,

f. Limitations on the operability and use of the liquid and gaseous effluent

. treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 3i-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,

g. Limitations on the dose rate resulting from radioactive material released in gaseous effiuents to areas beyond the Si1E BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B Table II, Column 1,

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix 1 to 10 CFR Part 50,

i. Limitations on the annual and quarterly dases to a MEMBER Of THE PUBLIC from lodine-131, lodine-133, tritium, and all radionutildes in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SIT [ BOUNDARY conforming to Appendix 1 to 10 CFR Part 50, ZNLD/862/19 m

I 1

AC#1N!STRtTIVE CONTROLS f.*

A~~!CN TO BE TAKEN IN THE EVENT A SAFETY LIMIT ]$ EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pursuant to Specification 2.1.1, 2.1.2 and 2.1.3, and critical reactor operation shall not be resumec until authorized by the NRC. The conditions of shutdown shall be promptly reported to the Vice President BWR Operations or his designated alternate. The incident shall be reviewed pursuant to Specifications 6.1.G.I.a and 6.1TG.2.a and a separate License Event Report for each occurrerce shall be preparce in accordance with Section 50.73 to 10 CFR Part 50. The NRC Operations l Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President BWR Operations and the Offsite Review and i l

Investigative Function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6. 5 PLANT OPERATING RECORDS A.

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

1.

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

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

3. Records and reports of reportable events; 4

Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirerhents (see Section 4 of these specifications) are being met. All equipment failing to meet surveil-lance requirements and the corrective action taken shall be recorded;

~

R cords of changes to operating procedures;

6. Shift engineers' logs; and

7. Byproduct material inventory records and source leak test results.

LA JALLE UNIT 1 6-)/ h- gwed f6 Amendment No. 66 genn#g ryo-y--7 .---v-- .w, .-w, y,r, , ,mm,.m- . , ,

ADMIN]STRATIVE CONTROLS PLANT OPERATIM RECORDS (Cont #nuedT B. Records and/or logs relative to the following items shall be recorded in a manner convenient for review and shall be retained for the life of the plant:

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

2. Changes inade to the plant as it is described in the SAR;

3. Records of new and spent f uel ir.ventory and assembly histories;

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

5. Records of plant radiation and contamination surveys;

6. Records of of fsite environmental monitoring surveys;

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

8. Records of radioactivity in liquid and gaseous wastes released to the environment;

9. Records of transient or operational cycling for those components that have been designed to operate safety for a limited number of transient or operational cycles (identified in Table 5.7.1-1);

10. Records of individual staff members indicating qualifications, experience, training, and retraining;

11. Inservice inspections of the reactor coolant system;

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

13. Records of reactor tests and experiments; i

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

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

16. Records of the service lives of all hydraulic and mechanical snubbers

! required by specification 3.7.9 including the date at which the service life commences and associated installation and maintenance records.

17. Records of analyses required by the radiological environmental monitoring program.

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LA SALLE UNIT 1 fcNum8CR 6-J# AGE

/ As Amendment No. 66 A f#oPRIME L

P':N15i ni N CONTROLS 6.^ CECORTING CEOUIREMENTS In accitien to the @plicable reporting reovirements of Title 10, Code of Feoeral Regulations, the following identified reports shall be submitted to the oirector of the apropriate Regicnal Office of Inspection and Enforce-ment unless otherwise noteo.

A. Routine Reports

1. Startup Report A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall in general include a description of the measured values of the operating conditions or characteristics obtaineo during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resuestion or commencement of commercici power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the startup report does not cover all three events (i.e.,

initial criticality completion of startup test program, and resumption or commen, cement of commercial power operation), supple-mentary rescrts : hall be submitted at least every 3 months until all three events have been completed.

2. Annual Report l A tabulation shall be submitted on an annual basis prior to March 1 of each year of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions (Note: this tabulation supplements the requirements of Section 20.407 of 10 CFR 20), e.g., reactor operations and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole l bocy dose received from external sources shall be assigned to specific major work functions, LA SALLE UNIT 1 g 6@

69f c Amendment No. 66 9

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ADu!N!!TRATIVE CONTROLS !

Annual Reoert (continued)

The results of specific activity analysis in which the primary coolant ,=ceeded the limits of Specification 3.4.5 shall be included in the Arnual Report along with the following information: (1) Reac-ter 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 l which the limit was exceeded; (2) Results of the last isotopic analy-sis for radiciocine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concen-tration in microcuries per gram as a function of time for the dura-tion of the specific activity above the steady state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

pSEgT 'E' i

LA SALLE UNIT I 6-g e# g Amendment No. 66 as

'r i

INSERT E ;

i

3. ANNUAL _ RADIOLOGICAL _ ENVIRONMENTAL _0PERATING_REPORif l 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 l summaries, interpretations, and analysis of trends of the results of i the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCH and (2) Sections IV B.2, IV.B.3, i and IV.C of Appendix 1 to 10 CFR Part 50.

4. SCHIANNUAL_RADICACTIVE.CffLVENT RELEASE REPORT **

The Semiannual Radioactive Effluent Release Report covering the f operation of the Unit during the previous 6 months of operation shall -

be submitted withia 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, r The material provided shall be (1) consistent with the objectives !

- outlined in the ODCM and PCP and (2) in conformance with 10 CfR 50.36a and Section IV.B.1 of Appendix 1 to 10 CFR Part 50.

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

A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

ZNLD/862-21

/)E?E76 ENTlW PAS e AM'MISTRAilVE CONTROLS

/

Annu ' [nvironmental Radiolooical Operatino Report (Continued) doses from primary effluent pathways and direct radia 'on) for the previous 12 consecutive months to show conforman with 40 CFR 190, Environmental Radiation Protection Stany rds for Nuclear Power Operation.

The assessment of radiatron doses shall be performed in accordance with the ODCM.

4 Semian -

dioactive Ei 'uent Release ReportE v

a. Routi ioactive vent release repor covering the operat the unit .sring the previous months of operation shall be u each year. $+ted within 60 days after anuary 1 and July 1 of We period of the first re rt shall begin with the date of ini al criticality.

b. The racioacti effluent release re orts shall include a summary of the quantiti of radioactive quid and gaseous effluents and solid waste leased from th unit as outlined in Regulatory Guide 1.21,"Measuing,Evaluat)ngandReportingRadioactivity in Solid Wastes and eleases 9f 3sdioactive Materials in Liquid and Gaseous Effluents rom L (ght-Water

Cooled Nuclear Power Plants," Revision 1, J e .4, with data summarized on a quarterly basis followin he format of Appendix B thereof.

+

b M Asingle submittal may e made for a multiple unit stat' n. The submittal should combine however, those ections that are common to all uni at the station; for units the releases of ra cactive th separate radwaste systems, the sub ittal shall specify material from each unit.

s SALLE UNIT 1 6-23 Amendment No. 66

ADMINISTRATIVE CONTROLS Semiannual Radioactive Effluent Release Report (Continued)

{

The radioactive effluent release report shall include the 'following information for each type of solid waste shipped offsite during the report period;

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

e.

Type of container (e.g., LSA, Type A, Type B, Larre Quantity),

and

f. Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

i The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

[

5. Monthly Operating Report M

h Routine reports of operating statistics and shutdown experience, A( incluoing documentation of all challenges to safety / relief valves, U shall be submitted on a monthly basis to the Director, Office of

$ Nuclear Reactor Regulation, Mail Station P1-137, US Nuclear Regulatory Commission, Washington, DC 20555, with a copy of the appropriate Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report.

( Any changes to the OFFSITE DOSE CALCULATION MANUAL shall bt semitteo with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the ocriod in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

6. CORE OPERATING LIMITS REPORT '

a. Core operating limits shall be established and documented in the CORE OPERATING LIMIT 5' REPORT before each reload cycle or any remaining part of a reload cycle for the following:

GEM M66R LA SALLE UN]T 1 6-[g $8cPEtA16 Amendment No. 70

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LA SALLE UNIT 1 6-26 Amendment No. 66

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

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i LA SALLE UNIT 1 6-27 Amendment No. 66 l

ADMINISTRATIVE CONTROLS C. Unique Reporting Requirements 1.

Special Reports shall be submitted to the Director of the Office of Inspection and Enforcement (Region 111) within the time period specified for each report.

6. 7 PROCESS CONTROL PROGRAM (PCP)*

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

6.7.2 Licensee initiated changes to the PCP:

a.

Shall be submitted to the Commission in the semi annual Radioactiv Effluent Release Report for the period in which the change (s) was made.

This submittal shall contain:

1.

k A Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; f I

i 2.

A determination that the change did not reduce the overall

.p conformance of the for solid wastes; andsolidified waste product to existing criteria

!gf d 3.

Documentation of the fact that the change has been reviewed and

( found acceptable by the Onsite Review and Investigative Function.)

b.

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

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

LA SALLE UNIT 1 gGtw%E P.

6-[g Nff tPRiATE Amendment No. 66

ADMINISTRATIVE CONTROLS

6. 8 0FFSITE DOSE CALCULATION MANUAL (ODCM)*

6. 8.1 The ODCM shall be approved by the Commission prior to implementatiun.

6.8.2 Licensee initiated changes to the ODCH:

a. Shall be submitted to the Commission within 90 days of the date the change (s) was made effective.

This submittal shall contain:

1. Sufficiently detailed informatic., to totally support the

(> -

rationale for the change without benefit of additional or suppitmental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, gMq 'Cs togetherwithappropriateanalysesorevaluationsjustifying the change (s);

2.

A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

3. Documentation of the fact that the change has been reviewed and found acceptable by the Onsite Review and Investigative Function.

b. Shall become effective upon review and acceptance by the Onsite Review '

l

( and Investigative function.

J l

l

6. 9 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATEMENT SYSTEMS l 6.9.1 Licensee initiated major changes to the radioactive waste systems t

(liquid, gaseous and solid):

a.

Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the Onsite Review and Investigative Function. The discussion of each change shall contain:

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

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

3. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

^1he La Salle OFFSITE Unit005E

2. CALCULATION MANUAL (ODCM) is common to La Salle Unit 1 and g(HuS6Ck V%C LA SALLE UNIT 1 6-[MWMNC Amendment No. 66

INSfRT f

a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.1L18. This documentation l shall contain: l

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 appItcable regulations.

INSIRI G

a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.D.18. This documentation shall contain:

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

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CfR part 190, 10 CfR 50.30a, and Appendix 1 to 10 CFR part 50 and not adversely impact the accut s's or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after review and acceptance by the On-Site Review and Investigative function and the approval of the plant Manager on the date specified by the On-Site Review and Investigative function.

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made effective. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

ZNLD/862/22

ADMINISTRATIVE CONTROLS

4. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous ef fluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendme" thereto;

$. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated 4 in the license application and amendments thereto;

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

7. An estimate of the exposure to plant operating personnel as a result of the change; and

9. Documentation of the fact that the change was reviewed and found acceptable by the Onsite Rcview and Investigative Function.

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

yh LA SALLE UNIT 1 6 yf cb h endment No. 66 Ft0Fpt

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

INDEK OEFINITIONS SECTION

1. 0 DEFINITIONS PAGE 1.1 ACT10N......................................................... .. 1-1 1.2 AVERAGE PLANAR EXP05VRE........................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................ 1-1 1.4 CHANNEL CALIBRAT10N............................................... 1-1 1.5 CHANNEL CHECK..................................................... 1-1 1.6 CHANNEL FUNCTIONAL TEST..... .... ................................ 1-1

1. 7 CORE ALTERATION....... ........................................... 1-2 1.8 CORE OPERATING LIMITS REPORT... ..................... ............ 1-2 1.9 CRITICAL POWER RAT 10...................................... ....... 1-2 i

1.10 DOSE EQUIVALENT l-131............................................. 1-2 1.11 E- AVER 4 GE D151NT EGRAT10N ENE RGY. . . . . . . . . . . . . . . . . . . . 1-2 ............

1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME. .............. 12 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME... 1-2 .....

1.14 FRACTION OF 61HITING POWER DEN 51TY................................ 1-3 1.15 F RACTION OF RATED THERMAL POWER. . . . . . . . . . . . . . . . . ........

....... 1-3 1.16 FREQUENCY NOTATION... ... ..

........... ................ ...... 1-3 1.17 GASEOUS RADWASTE TREATMENT 5YSTEM................................. 1-3 A.18 IDENTIFIED LEAKAGE............................................... 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME................................... 1-3

1. 20 LIMI TING CONT R01. ROD PATTE RN. . . . . . . . . . . . . . . . . . . . . . . . . . . .1-3 1.21 LINEAR HEAT GENERATION RATE....................................... 1-4 1.22 LOGIC SYSTEM FUNCTIONAL TE5T......................................

1-4 1.23 MAXIMUM' FRACTION OF LIMITING POWER DENSITY.............. 1-4

>~ te ........

1.)<' MINIMUM CRITICAL POWER RAT 26 10........................ ............ 1-4 1.J5" 0 F F 51 T E DO S E CA L C U L A T I O N M AN U A L . . . . . . . . . . . . . .1-4 ............

_ _ _ _ - -LA SALLE - UNIT 2 1 Amendment No. 54 t, zt/ n%A18ER(s) _ _ .

of THE f98LIC

INDEX DEFINITIONS gn --

SECTION O As AHRel%TC DEFINITIONS (Continued) PAGE t1 1.)#' O P E R A B L E - O P E RA B I L I T Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14 ............

1.Jf 0PERATIONAL CONDITION -

t?

CONDITION................................. 1-4

1. g PHYSICS So' TESTS..................................................... 1-4 1.pfPRE55UREBOUNDARYLEAKAGE........................................ 15 si i

1.Jef PRIMARY CONTAINMENT INTEGRITY.................... ............ ... 1-5 l St i 1.X PROCESS CONTROL PROGRAM. . . . . . . . . . . . ................... ... .... 1-5 33 1.,T PURGE -

vt PURGING................................................... 1-5 1.; rRATED THERMAL

,e P0WER.............................................. b5 1.X 2<.

RE ACTOR PROTECTION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . .1-5 .........

1.,PJ REPORTABLE EVENT................... ..... .................. ... 16 rf 1.fr5' ROD DEN 31 51TY.................................................... .. 1-6 1.Jr 5ECONDARY CONTAINMENT 39 INTEGRITY................................... 1-6

/

1.Je'5HUTDOWN MARGIN.... ...... ............................... .... . 1-6 1.39 SOLIDIFICATION...... .................................. ...... ..

1-6)

I 1.hSOURCECHECK......................................................

vt. 1-7

1. v3 f# S T AG G E R E D T E S T B A S I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 ............

1. W THERMAL W P0WER..................................................... 1-7 l 1.ft TURBINE BYPASS RESPONSE TIME...................................... 1-7 i

6 1./4' UNIDENTIFIED LEAKAGE.............................................. 1-7 1.f# VENTILATION EXHAUST TREATMENT 47 5YSTEM............................. 1-7 j

r 1.J(VENTING......................................................... 1-7 l

t i40 5ITE SouHtMT LA SALLE - UNIT 2 II- Amendment No. 54 t

l l

INDEX LlHITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.................... 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.......................... 3/4 3 9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...... 3/4 3-23 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATVS Recirculation Pump Trip System Instrumentation.......... 3/4 3-35 End-of-Cycle Recirculation Pump Trip System Instrumentation................................. .......... 3/4 3 39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION............................................ 3/4 3-45 3/4.3.6 CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION................. 3/4 3-50 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation......................... 3/4 3-57 Seismic Monitoring Instrumentation........................... 3/4 3 60 Meteorological Monitoring Instrurt 'ation.................... 3/4 3-63 Remote Shutdown Monitoring Instrumentation................... 3/4 3-66 Accident Monitoring Instrumentation.......................... 3/4 3-69 Source Range Monitors........................................ 3/4 3-72 Traversing In-core Probe System.............................. 3/4 3-73 I

fire Detection Instrumentation...............................

( Radioactive Liquid Ef fluent Monitoring Instrumentation.. . . . . .

3/4 3-75 3/4 3-81

[

t 6v 6AL

)

l R:d i e r:ttp t.v: 5/VECn:: . E f'h:nt Monitoring Instrumentation. . . . . .

3/4 3-pffg; l Loose-Part Detection System..................................

1 3/43-f%

3/4.3.8 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.,........................................... 3/4 3-)( g LA SALLE - UNIT 2 V Amendment No. 4?

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

l 1

l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ,

SECTION PAGE 3/4.11 RADI0 ACTIVE EFFLUENTS I

3/4.11.1 LIQUID EFFLUENTS Concentration............. .... ................... .... .... 3/4 11-1 Dose.................... ................... .............. 3/4 11 6 L

t Liquid Waste Treatment System................... ............ 3/4 11-7

-(LiquidHoldupTanks....................................... 3/411-gl

g. d3/411.2 GASf0VS EFFLUENis p/.b' ' Dose Rate.................................... ............... 3/4 11-9 '

Dose Noble Gases.......................... .................. 3/4 11-13 Dose-Radiciodines, Radioactive Material in Particulate Form, and Radionuclides Other than Noble Gases....... ..... 3/4 11-14 I

Gaseous Waste Treatment System............................... 3/4 11-15 L

Ventilation Exhaust Treatment System.... .... . . .. .. 3/4 11 16J

/ Explosive Gas Mixture........................................ 3/4 11,V 2-M L Ma i n Co nd e n s e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ 3/4 11,HT 3 (VentingorPurging......................... ........ .... ... 3/411-19]

3/4 11.3 SOLID RADIOACTIVE WASTE.... ................................. 3/4 11-20 I 3/4 11.4 TOTAL 00SE................................................... 3/4 11-22 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4 12.1 MONITORING PR0 GRAM........................................... 3/4 12-1 1/4 12.2 LAND USE CENSUS.............................................. 3/4 12-9 3/4 12.3 INTERLABORATORY COMPARISON PR0 GRAM..........................,

1 3/412-10) l l

LA SALLE - UNIT 2 XI Amendment No. 53

i INDEX i

BASES i SECTION PAGE INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Meteorological Monitoring Instrumentation.. . . .. .... B 3/4 3-4 Remote Shutdown Monitoring Instrumentation...... . ..... B 3/4 3-4 Accident Monitoring Instrumentation.... ....... ....... B 3/4 3-5 Source Range Monitors.... ....... .. .. .. .. ... . ... B 3/4 3-6 Traversing In-core Probe System....... .. ... . . . B 3/4 3-5 Chlorine and Ammonia Detection System. ............ ... B 3/4 3-5 Fire Detection Instrumentation.. ......... . ... .......

B3/43-5[

(RadioactiveLiquidEffluentMonitoringInstrumentation. B3/43-6)

(vnosWC GAS

" m" = t i . Ce n.; E m : Monitoring Instrumentation.. B 3/4 3-6 Loose-Part Detection System....... .. ... .. . .. . . B 3/4 3-6 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SY5h 4 TION INSTRUMENTATION...... ............ ................ .. B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM., ... ................. ...... ... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVE 5..................................... B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE i

Leakage Detection Systems., ... ..... ... .. ... . .. .. B 3/4 4-2 Operational Leakage........... ........ . ., ... .. ... . B 3/4 4-2 3/4.4.4 CHEMISTRY......... ............... .., . ................ B 3/4 4-2 3/4.4.5 SPECNICACTIVITY................................... B 3/4 4-3 3/4.4.6 PRES 5URE/ TEMPERATURE LIMITS.. . ... . .... ... .. . B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES. . . .. . . . ... B 3/4 4-5 l

3/4.4.8 STRUCTURAL INTEGRITY... .. ... ... .. .. .. ... B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REMOVAL..... . .... . . .. . . . . . B 3/4 4-5 LA SALLE - UNIT 2 X111

INDEX BASES SECTION PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.... ..... ... ..... . .. B 3/4 10-1 3/4.10.2 R00 SEQUENCE CONTROL SYSTEM.......... ...... .. .. .. B 3/4 10 1 3/4.10.3 SHUTOOWN MARGIN DEMONSTRATIONS...... ...... .. , .. B 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS........ ................ . ... ... B 3/4 10 1 3/4.10.5 OXYGEN CONCENTRATION..... .................. . .. .. .. B 3/4 10-1 3/4.10.6 TRAINING STARTUPS......... ..... .. ... ... .. . ... B 3/4 10-1 3/4.10.7 CONFIRMATORY FLOW INDUCE 0 VIBRATION TEST. . .. ...... B 3/4 10-1 3

3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration... . ...................

, f

...... ........ . B 3/4 11-1 b

a Dose......................... ..... .... .. .. .... B 3/4 11-1 t Liquid Waste Treatment System....... ....... ...... .... B 3/4 11-2,

-(LiquidHoldupTanks.................................. 03/411-//

( 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................. .................... ........

[

B 3/4 11-2 I h

i Dose - Noble Gases....... ........ ........ ............ B 3/4 11-3 Dose -Radiciodines, Radioactive Materials in Particulate form and Radionuclides Other than Noble Gases........................................... B 3/4 11 3 Gaseous Radwaste Treatment 3ystem and Ventilation 1 Exhaust Treatment System................... .... ..... B 3/4 11 / Explosive Gas Mixture...................... ............

B3/411-/l 5 Main Condenser................................... ......

~

B 3/4 11-fi Venting or T

Purging................................ ... B 3/4 11-5 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................. B 3/4 11-5 3/4.11.4 10TAL 00$E................... ................. ........ B 3/4 11-5 3/4.12 RADI0 ACTIVE ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM............................ ......... 3/4 12-1 3/4 12.2 LAND USE CENSUS.............................. ..... . .. 3/4 12-1 3/4 12.3 INTERLABORATORY COMPARISON PROGRAM.... ..... .... ... 3/4 12-1 LA SALLE - UNIT 2 a XVI Amendment No. 53

1 l

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 ORGANIZATION, REVIEW, INVESTIGATION, AND AUDIT ... ..... . .. ... 6-1 6.1.1 High Radiation Areas.......................... ........ .. 6-15

6. 2 PLANT OPERATING PROCEDURES AND PROGRAMS...... ...... ........... 6-16

6. 3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE IN PLANT OPERATION......... ......... ..... .. ........ 6 18 -

6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDE0....... 6- 19

6. 5 PLANT O'ERATING RECORDS...... ..................... ... .... .. . 6- 19 I

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

6. 7 PROCESS CONTROL PROGRAM.............. . .......................... 6- 28 6.8 0FFSITE DOSE CALCULATION MANUAL... .............. ................ 6- 29

6. 9 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS.............. 29 6O

/

Eh gc' NudpfPep #

p /

l LA SALLE - UNIT 2 XVIII

LIST OF TABLES (Con +1nued)

TABLE PAGE 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION ........ 3/4 3-67 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ........................ 3/4 3-68 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION ............... 3/4 3-70 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .......... .. ........... 3/4 3-71 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION .................... 3/4 3-76

[3.3.7.10-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING

( INSTRUMENTATION ....... .............. .. .. ..... 3/4 3-82 4.3.7.10-1 RADIOACTIVE LlQUID EFFLUENT MONITORING '

( INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,.. .....

3/43-84J 3.3.7.11-1 H RADI0 ACTIVE GASEOUS EFFLUE T[ MONITORING

.n-INSTRUMENTATION .............. .... ............... 3/4 3-pfg2.

4.3.7.11-1 ->4 RADIOACTIVE GASEOUS EFFLUENTDiONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-f/f g3 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION ........... ............. 3/4 3-)# gG, 3.3.8-2 r FEEDWATER/ MAIN TURBINE TRIP SYSTEM

g ACTUATION INSTRUMENTATION SETPOINTS ............... 3/4 3-p4'97

> ?

4.3.8.1-1 J FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION p jINSTRUMENTATIONSURVEILLANCEREQUIREMENTS ........ 3/4 3./ ?,3 3.4.3.2-1 O REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES .. 3/4 4-10 3.4.4-1 2

xtuREACTOR COOLANT SYSTEM CHEMISTRY u!MITS . . ....... 3/4 4-13 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ..................... .. ..... ... 3/4 4-16 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

, WITHDRAWAL SCHEDULE ......................... ..... 3/4 4-20 4.6.1.5-1 TENDON SURVEILLANCE .............................. 3/4 6-11 4.6.1.5-2 TENDON LIFT-OFF FORCE ...... ...................... 3/4 6-12 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES . .- .. ... 3/4 6-27 LA SALLE - UNIT 2 XXII

LIST OF TABLES (CONTINUED)

TABLE PAGE 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES ............... .. . .. 3/4 6-42 3.7.5.2-1 DELUGE AND SPRINKLER SYSTEMS . ............... .. . 3/4 7-16 3.7.5.4-1 FIRE HOSE STATIONS ................... . .. ..... 3/4 7-19 3.7.7-1 AREA TEMPERATURE MONITORING ,........... .... ..... 3/4 7-26 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE ............... .... 3/4 8-7 4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS ................. 3/4 8-18 3.8.3.2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ....... . .. ..... 3/4 8-24 3.8.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION ...................... ............. .. 3/4 8-27 i

3.11.1-1 MAXIMUM PERMISSIBLE CONCENTRATION OF DISSOLVFD OR ENTRAINED NOBLE GASES RELEASED FROM THE SITE TO UNRESTRICTED AREAS IN LIQUID WASTE .. ........ . 3/411-2f J4.11.1-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS

( PROGRAM ....................

l j

4.11.2-1 RADIOACTIVE GASEOUS WASTE SAMPLING, AND ANALYSIS PROGRAM ...................... ......... .

3/4 11-3J

<[

3/411-10J

' 3.12.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING P OGRAM ..... 3/4 12-3' 3.12.1-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES ..................... .... 3/4 12-6 (4.12.1-1 MINIMUM VALUES FOR THE LOWER LIMITS CF DETECTION .. 3/4 12-7) 83.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS . . .. . ....

B3/42-2)

B3/4.4.6-1 REACTOR VESSEL TOUGHNESS . ............... . ... . B 3/4 4-6 w 5.7.1-1 COMPONENT CYCLIC CR TRANSIENT LIMITS ... ....... .. 5-6 f LA SALLE - UNIT 2 XXIII

DEFINITICNS 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 l 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, l 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 performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

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

,/~-Y MINIMUM CRITICAL POWER RATIO 1.)( The MINIMUM d exists CR]TICAL POWER RATIO (MCPR) shall be the smallest CPR in the core. I w OFFSITE DOSE CALCULATION MANUAL

k. 1.25 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology I

and parameters used in the calculation of offsite doses due to radioactive

( gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoripg alarm / trip setpoints.

J

[ OPERABLE - OPERABIL}TY 1.pf A system, subsystem, train, component or device shall be OPERABLE or have l

'rti ODERABILITY and when it is capable of performing its specified function (s),

when all necessary attendant instrumentction, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION - CCNDITION 1.JrAnOPERATIONALCONDITION,i.e., CONDITION,shallbeanyoneinclusive l g combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1.jfPHYSICSTESTSshallbethosetestsperformedtomeasurethefundamental l

29 nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

LA SALLE - UNIT 2 1-4 Amendment No. 54

OEFINITIONS PRESSURE EDUNDARY LEAKAGE 1.;M in a reactor coolant system component bodywall.

, pipe wall or vessel 5 PRESSURE l PRIMARY CONTAINMENT INTEGRITY 1.)6 31 PRIMARY CONTAINMENT INTEGRITY shall exist when: l a.

All primary contaihment penetrations required to be closed during accident conditions are either:

1.

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

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification.

3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

c. Each primary containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d.

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

e.

The suppression chamber is OPERABLE pursuant to Specification 3.6.2.1.

Ngd f. The sealing mechanism associated with each primary containtrent I

penetration; e.g., welds, bellows or 0 rings, is OPERABLE.

_ PROCESS CONTROL PROGRAM

'I.31 The PROCESS CCNTROL PROGRAM (PCP) shall contain the sampling, analysis, l and formulation determination by which SOLIDIFICATICN of radioactive

( wastes from liquid systems is assured.

l PURGE-PURGI_Q 1.)r PURGE or PURGING shall be the controlled process of discharging air or

! l i

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

l RATED THERMAL POWER 1.)r RATED THERMAL POWER shall be a total reactor core heat transfer rate to l t 3V the reactor coolant of 3323 %'T.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.)4 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from l 35 when the monitored parameter exceeds its trip setpoint at the channel sensor until de energization of the scram pilot valve solenoids. The response time tray be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

I LA SALLE - UNIT 2 1- 5 Amendment No, 54

DEFINITIONS REPORTABLE EVENT 1.yf A REPORTABLE EVENT shall be any of those conditions specified in 36 Section 50.73 to 10 CFR Part 50. l ROD DENSITY 1.Jef ROD DENSITY shall be the number of control rod notches inserted as a l 37 fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.yf SECONDARY CONTAINMENT INTEGRITY shall exist when:

l

a. All secondary containment penetrations required to be closed during accident conditions are either:

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

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

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

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

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

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

f. The pressure within the second.ery containment is less than or equal to the value required by Spccification 4.6.5.1.a.

SHUT 00WN MARGIN 1.)6 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is l

'c; 39 subtritical or would be subcritical assuming all control rods are fully inserted except for the t. ingle control rod of highest reactivity worth j which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

L SOLIDIFICATION 1.39 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid l systems to a homogen'eous (uniformly distributed), monolithic, immobili:ed

( solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). f j

LA SALLE - UNIT 2 1-6 Amendment No.54

INSERT Al tiBSERIS) 0F THE PUBuf 1.24 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does inalude persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

INSERT A2 01ESULDOSE_fALCULAHON MANU&L 1.26 The OFFSITE DOSE CALCULATION MANUAL (ODCM) sha'l contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid ef'Inents, in the calculation of gaseous and liquid effluent monite ing Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (') the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

INSERT B 1.32 The PROCESS CONfROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

INSERT C SLIE_BRURDARY l 1.40 The SITE BOUNDARY shall be that line beyond which the land is !

neither owned, nor leased, nor otherwise controlled by the licensee. l

)

1 ZNLD/862/12

{

9 4

DEFINITIONS SOURCE CHECK

1. M A SOURCE CHECK shall be the-qualitative assessment of channel response l 4// when the channel sensor is exposed to a radioactive source.

$rAGGERED TEST BASIS i

1.fr A STAGGERED TEST BASIS shall consist of: (

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

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

-THERMAL POWER 1.S( THERMAL POWER shall be the total reactor core heat transfer rate to the l 9 reactor coolant.

i TURBINE BYPASS SYSTEM RESPONSE TIME j

1. K The_ TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when l

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

The response time _may be measured by any serie's of sequential, o'eerlapping or total steps such that the entire response time is measured.

UNIDENTIFIED LEAKAGE

- 1.)6(

' UNIDENTIFIED LEA r. hall

% be all leakage which is not IDENTIFIED LEAKAGE. l 4 VFNTILATION EXHAUST j f .NT SYSTEM 1.prAVENTILATIONEXHAUSTTREATMENTSYSTEMshallbeanysystemdesignedand l 6 installed to reduce gaseous radiciodine or radioactive material ia particu-late form in effluents by passing ventilation or vent exhaust gases through :harcoal adsorbers and/or HEPA filters for the purpose of removing iodines or-particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric-cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1 5 VENTING : Nil be the controlled process of discharging air or gas from a l 47 confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING, Vent, used in system names, does not imply a VENTING process.

LA SALLE - UNIT 2 1-7 Amendment No.54

DCLETE ENTIRE PAGE INSTRUMENIATION DI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMI G CONDITION FOR OPERATION x

/

3.3.7.10 e radioactive liquid effluent monitoring instrumentat'on channels showr in Tab e 3.3.7.10-1 shall be OPERABLE with their alarm /tr setpoints set to ensure . hat e limits of Specification 3.11.1.1 are no exceeded. The alarm trip setp f these channels shall be determined i accordance with the Offsite Dose a ation Manual (00CM).

APPLICABILITY:

, At ges.

ACTION: A

a. With a radioacti quid effluent monit ing instrumentation channel alarm / trip setpoi less conservative t an required, immediately suspend the release of radioactive lir id effluents monitored by the affected channel or clare the chan il inoperable.

b. With less than the mini, m number f radioactive liquid effluent monitoring instrumentatio chann s OPERABLE, take the ACTION shown in Table 3.3.7.10-1. Rest e e inoperable instrumentation to OPERABLE status within the t e specified in the ACTION or, in lieu of a Licensee Event Re r , explain in the next Semiannual Radioactive Effluent Relea Re rt why this inopersbi;ity was not corrected within the ime spe ified,

c. The provisions of Spec fications 3. 40.4,and6.6.8.2. bare not applicable, e

SURVEILLANCE REQUIREMENTS k,

sy 4.3.7.10 Each radica ive liquid effluent monitoring ins u ation channel shall be demonstrat OPERABLE by performance of the CHANN. K, SOURCE CHECK, CHANNEL FUN IONAL TEST, and CHANNEL CALIBRATION oper< s at the frequencies show in Table 4.3.7.10-1. @

D l

LA SALLE - UNIT 2 3/4 3-81

r TABLE 3.3.7.10-1 2-p RADI0 ACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION E

MINIMUM c CHANNELS

(

It RUMENT OPERABLE ACTION

1. GAMMA SCINTILLATION TOR PROVIDiNG ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent e 100

2. GAMMA SCINTILLATION MONITORS PROVID AUTOMATIC TERMINATION OF RELEASE ALARM BUT NOT PROVIDING &/

a. Service Water System Effluent Line (Uni 1) 1 101

b. RHR Service Water (Line A) Effluent Lin 1 101 M c. RHR Service Water (Line B) Effluent Line 1 101

d. Service Water System Effluent Line (Unit 2) -

1 101 2 3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 102

b. River Discharge - Blowdown Pipe f 1 102 Y E a

a ~

1 9

kcs N

. _ _ _ _ _ . . _ _ . _ ._--._ _ __ _ _._. . _ _ _ _ ..._._____.m.___._.

CEu TE E^tTIKC $h GC e

TABLE 3.3.7.10-1 (Continued)

.1 ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels less than equired by the Minimum Channels OPERABLE requirement, effluent C@ -releases may continue for up to 14 days pro ded that prior to initiating a release:

At least two independent samples a e analyzed in '

accordance with Specification 4. .1.1.3, and

b. % 1 east two technically qual ied members of the Taci.lity Staff independently verify -the release rate alculations and discharge ine valving; Otherwi , suspend release o radioactive effluents via this path y.

ACTION 101 -

With the num .r of chan s OPERABLE less than required by the Minimu Channel OPERABLE requirement, effluent releases via th e pat way may continue for-up to 30 days provided that, at 1 st once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and anal 'ed at a limit of detection of at least 10 7 microcurie / o gamma spectrometric analysis.

ACTION 102 -

With the numb of chan e ERABLE less than required by the Minim Channels E requirement, effluent releases v this pathway- y 7tinue for up to 30 days provided e flow rate is e im A d at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months uring actual release Jump curves for Instru-ment 3 , or for known valve po 't. for Instrument 3b, may used to estimate flow.

A T

L G@S l-t l

I l- LA SALLE - UNIT 2 3/4 3-83 l

I

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

r

> TABLE 4.3.7.10-1 y RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS U

CHANNE E CHANNEL SOURCE FUNCT JAL CHANNEL

~_; INSTRUMENT CHECK CHECK ST CALIBRATION

1. GAMMA SCINTILLATION M TOR PROVIDING ALARM AND AUTOMATIC TERMINAT 4 0F RELEASE

a. Liquid Radwaste Effluents # A ine D P Q(1) R(3)

2. GAMMA SCINTILLATION MONITORS PROVID ALARM

>3 adT NOT PROVIDING AUTOMATIC TERMINA N OF RELEASE m a. Service Water System Effluent Line (Unit 0 M Q(2) R(3)

} b. RHR Service Water (Line A) Effluent Line M Q(2) R(3) m c. RHR Service Water (Line B) Effluent Line M R(3) d.

Q(2) g Service Water System Effluent Line (Unit D M Q(2) R(3)

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line d D(4) N.A. Q R

b. River Discharge - Blowdown P' e \ D(4) A. Q R 9 -

a S

M S 9 %

=

it, n

9

DELETE ~ EA/TlRE l%GE~ ^

l TABLE-4.3.7.10-1 (Continued)

TABLE NOTATION j j

(1) The 1ANNEL FUNCTIONAL TEST shall also demonstrate that automa ic isolation of thi pathway and control alarm annunciation occurs if any f the followl conditions exist:

1. Instr N ndicates measured levels above the alar / trip setpoint.

2. Loss of .

3. Instrument M on downscale failure.

A

4. Instrument con o ot set in Operate or gh Voltage mode.

(2) _The CHANNEL FUNCTIONAL TEST shall also demone- rate that control room alarm annunciation.occu if any of the foi owing conditions ex.ist:

1. Instrument indicates asured level above the alarm setpoint.

2. Loss of power.

3. Instrument alarms on downsc- failure.

4. Instrument controls not se in erate or High Voltage mode.

(3)' The initial CHANNEL CALIBRAT N shall >

~ rmed using nne or more of the reference radioactive s andards cert y-the National Bureau of Standards or using standa s that have bet 'ned from suppliers that participate in measureme t assurance activi i th NBS. These standards shall permit calibrati the system over its n d range of. energy and measurement range. F r subsequent CHANNEL CAL s N, the initial reference radioacti - standards or radioactive he that have been related to the ini al calibration shall be used.

(4) CHANNEL CHECK s 11 consist of verifying indication t G during periods of release. C ANNEL CHECK shall be made at least onc p 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months on days in whic continuous, periodic, or batch releases a @e .

i' LA SALLE - UNIT 2 3/4 3-85

DELETE' ENTittE PR6E i?dtAcE SPEctFt ckTroH WITU FouowtHG PAGE l

I INSTRUMENTATION

. R D10 ACTIVE GASE0US EFFLUEWT MONITORING INSTRUMENTATION LIMIT.4G CONDITION FOR OPERATION

\ i 3.3.7.11 e radioactive gaseous effluent monitoring instrume tation channels shown in Tab 3.3.7.11-1 shall be OPERABLE with their alarm rip setpoints set to ensure hat the limits of Specification 3.11.2.1 ar ot exceeded. The alarm / trip setp ints of these channels shall be determine in accordance with the ODCM.

d APPLICABILITY: As i 3.3.7.11-1

? n Table ACTION: h h

a. With a radioact1 e seous effluent m nitoring instrumentation channel alarm /tri petpoint s less to ervative than required, immediately suspen the release of adioactive gaseous effluents monitored by the af cted channe or declare the channel inoperable, b, With less than the mini um nu, er of radioactive gaseous ef fluent monitoring instrumentati c nnels OPERABLE, take the ACTION shown in Table 3.3.7.11-1.

c. The provisions of Specif cati s 3.0.3 and 3.0.4 are not applicable.

h SURVEILLANCE REQUIREMENTS

/

k

\ /\

/

4.3.7.11 Each radioacti e gaseous effluent moni instrumentation channel shall be demonstrated ERABLE by performance of t C ANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTI 4AL TEST and CHANNEL CALIBRAT N operations at the frequencies shown i Table 4.3.7.11-1.

6N

. DELETE' fAGE fcPth c5 sfEaFicMic h/

wuTH Fouowulc7 PAGE LA SALLE - UNIT 2 3/4 3-86

g TABLE 3.3.7.11-1 p"

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS

[

z INSTRUMEN OPERABLE APPLICABIL ACTION Z 1.

N MAIN CONDENSER OFFGA TREATMENT SYSTEM EFFLUENT MONITORING TEM

a. Noble Gas Activity Mon r - Providing A Alarm and Automatic Termi tion of Release 1

110 2.

MAIN CONDENSER OFFGAS TREATMENT SYS EXPLOSIVE GAS MONITORING SYSTEM (for' systems de 'gned to

@>p I withstand the effects of a hydrogen exp ion)

a. Hydrogen Monitor

/ train **

111

{ 3. MAIN STACK MONITORING SYSTEM y a. Noble Gas Activity Monitor co *

" b. Iodine Sampler 110 [

c. 1

Particulate Sampler 113

d. Effluent System Flow Rate Monitor Q 1

113

e. Sampler Flow Rate Monitor y

y/ 1

114 1

114

4. A g CONDENSER AIR EJECTOR RADIDACTIVI OR Ch g (Prior to Input to Holdup Sy em) S-

a. Noble Gas Activity Mon' r 1

@h g

5.

115 l MS SBGTS MONITORING SYSTE C) a.

b.

Noble Gas Act' sty Iodine Samp' r litor i ##

[)

5 A

c.

d.

Partical e Sampler Efflue System Flow Rate Monitor 1

1 1

10 113 g

(Dd

$ e. 1 %

Samp r Flow Rate Monitor ## 114 'h 5 1 ## 114 E 3 2

o 9 A m

u /

/ >9 LV F

ny

DELET[ EA(TIN /%GE i REPLACE WIT // AFCW i TASLE TABLE 3.3.7.~11.1 (Continued)

TABLE NOTAT10NS'

At all mes.

- During mai condenser offgas treatment system operation i

0uring operat n of the main condenser air ejector.

1. During operation SBGTS.

ACTION 110 -

With the

(&m ACTION STATEMENTS of channels OPERABLE ess than required by the Minimum Cha n OPERABLE requirem t, ef fluent releases via this pathway m ontinue for up 30 days provided grab samples are ta en at least once er 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed fo noble gas gam emitters within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 111 -

With the number.of hannels PERABLE less than required by the Minimum Channels OP ABLE >quirement, operation of main con-denser offgas treatme s stem may continue for up to 30 days

provided grab samples collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within th llowing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the recombiner(s) temperature remains nst t and THERMAL POWER has not changed, the grab sample col >ction equency may be changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 112 -

With the number channels OP less than required by the Minimum Channel -OPERABLE requi m suspend release of radioactive e luents via this pa ACTION 113 -

With the n Der of channels OPERABL 1 han required by the Minimum C nnels OPERABLE requirement, ent releases via this pa way may continue for up to 30 ys provided that within hours after the channel has bee declared inoperable sampi s are continucusly collected with au i sampling i equ ment as required in Table 4.11.2-1.

ACTION 114 -

th the number of channels OPERABLE less than ed by the inimum Channels OPERABLE-requirement, ef fluent a s via this pathway may continue for up to 30 dys provi 'd the flow rate is' estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 1 - With the number of channels OPERABLE less than require by the Minimum Channels OPERABLE requirement, the output from e charcoal adsorber vessels may be released to the environm nt for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

a. The offgas treatment system is not bypassed, and

b. The offgas treatment delay system noble gas activity l effluent downstream monitor is OPERABLE; Otherwise, be in at least STARTUP with the main steam isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

LA SALLE - UNIT 2 3/4 3-88

1 TABLE 4.3.7.11-1 E ,

m RADI0 ACTIVE GASE0US EFFLUENT MONIl0 RING INSfRUMENTATI('N SURVEILLANCE REQUIREMENTS p '

i g OPERAT10NAL

, CHANNEL CONDITIONS FOR t c CilANNEL SOURCE- FUNCTIONAL CHANNE WHICH SURVEIL--

5 INSTRUMENT CllECK CHECK TEST CALIBP . ION LANCE REQUIRED H

v 1. MAIN CONDENSER OFFGAS ATHENT SYSTEM EFFLUENT MONITORING SY .

,. a. Noble Gas Activity Monito - t Providing Alarm and Automati Termination of Release D D y R(3) *

2. MAIN CONDENSER OFFGAS TREATMENT SYSTEM

[

EXPLOSIVE GAS MONITORING SYSTEM ,

y a. ilydrogen Monitor N. A. H Q(4) **

a -

i

3. MAIN STACK MONITORING SYSTEM

a. Noble Gas Activity Monitor D H Q(5) R(3) * '

b. Iodine Sampler W N.A. N.A. N.A. * ,

c.

d.

Particulate Sampler Effluent System Flow Rate Moni hW D N.A.

N.A.

N.A. N.A.

R-

[

e. Sampler Flow Rate Monitor \ D N.A. Q R *

4. CONDENSER AIR-EJECTOR RADIO IVI NITOR gf

a. Noble Gas Activit o tor D M Q(2) R(3) # '> b,

5. SBG15 MONITORING S

a. ;4chle Act ty Monitor D M Q(5) R(3) ## $

b.

c.

Iodi Sampler iculate Sampler W N.A.

N.A.

N. A.

N.A.

Rmfg. ;

d.

P f fluent ' System Flow Rate Monitor W %N !

D N.A. Q R ##

Sampler Flow Rate Monitor D N.A. Q R % $ },

k g,

)3-a rat f

D "I !

I

MLC76 ENT/RE Ph E RULACE wITH NEN 7A6LC TABLE 4.3 7.11-1 (Continued)

TABLE NOTATIONS

At 1 times.

- 0uring ain condenser offgas treatment system operation.

0uring op ation of the main condenser air ejector.

1. During opera the SBGTS.

(1) The CHANNEL FUN I AL TEST shall also demonstrate he automatic isolation capability of thi y, and that control roo alarm annunciation occurs

-if any of the foll onditions exists: (ca channel will be tested independently so as ot t initiate automatic . solation during operation).

1. Instrument indica es sured levels a ve the alarm / trip setpoint.

2. Loss of power.

3. Instrument alarms on wnscale fai 're.

1. Instrument controls not et in 0 erate or High Voltage mode. (Auto-matic isolation shall be emon rated during the CHANNEL CALIBRATION.)

(2) The CHANNEL FUNCTIONAL TEST for log scale monitor shall also demonstrate that control room al annunciation occurs if any of the following conditions exists:

1. Instrument indicates me ured le is above the alarm setpoint.

2.

3.

Loss of power, Instrument alarms downscale failu .b h

4 Instrument contr s not set in Operate r i_h Voltage mode.

(3) The initial CHANNEL ALIBRATION shall be perfo 64 ing one or more of the reference radi active standards certified by tt tional Bureau of Standards (NBS) using standards that have been o d from suppliers that participat in measurement assurance activiti ith NBS. These standards shal permit calibrating the system over 1 s intended range of energy and m surement range. For subsequent CHANNEL ALIBRATION, the initial ref rence radioactive standards or radioactive o* that have been rela d to the initial calibration shall be used.

(4) The CHA EL CALIBRATION shall include the use of standard , ples contai ing a nominal:

1. One volume percent hydrogen, balance nitrogen, and 2 Four volume percent hydrogen, balance nitrogen.

(5) he CHANNEL FUNCTIONAL TEST shall also demonstrate that control ro alarm annunciation occurs-if any'of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint,

2. Circuit failure.

3. Instrument controls not set in the Operate mode.

LA SALLE - UNIT 2 3/4 3-90 i

INSIRUMERIAlIQN EXPLOSIVE _ GAS MONIIORING_INSIRUMENIATION l 1

LIM 111NG_CONQ]Il0N_LOL0EERAIIOR __. _

3.3.7.11 The explosive gas monitoring instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their Alarm / Trip setpoints set to i ensure that the limits of specification 3.11.2.6 are not exceeded.

AEELICABILIIYJ During operation of the main condenser air ejector. '

ACI!QN1

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

b. With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.11-1. Restore the inoperable instrumentation channels to an OPERABLE status within 30 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C. within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVflLLANCE_REQUlREMERLS . _

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

=

LA SALLE - UNIT 2 3/4 3-81 PROPOSED AMENDMENT ZNLD/862/16

INSIEVMENIA110N LABLL3,3Jdl-1 EELOSIVE_ GAS _MONIJORING_INSIRUMENIA110N MINIMUM CHANNELS INSIRUMENI ___DEERABLE___ ACJ10N

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

a. Hydrogen Monitor 1/ train 110 1ABLE_N01aIlON

-ACTION 110 - Hith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the main condenser offgas treatment system may continue for up to 30

-days provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the recombiner(s) temperature remains constant and THERMAL POWER has not changed, the grab sample collection frequency may be changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , l

l LASALLE - UNIT 2 3/4 3-82 PROPOSED AMENDMENT ZNLD/862/17

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

INSTRUMENTATION IABLE_4 aAll-1 EXPLOSIVLGAS_ MONITORING. INSTRUMENTATION OPERATIONAL CHAN N EL. CONDITIONS FOR CHANNEL FUNCTION *L CHANNEL HHICH SURVEll-INSIRUMENI _ CHECK _._] E SL _ CAllBRA110N' LANCE _RE0VIRED

1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM a, llydrogen Monitor M '*

D Q

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

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

During operation of the main condenser air ejector.

i LASALLE - UNIT 2 3/4 3-83 PROPOSED AMENDMENT ZNLD/862/18 l

INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7,12 The loose part detection system shall be OPERABLE.

i APPLICABILITY: OPERATIONAL COND TIONS 1 and 2.

ACTION:

a. With one or more loose part detection system channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.c within the next 10 days outlining the cause of the malfunction ind -the plans for restoring the channel (s) to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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

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

b. CHANNEL FUNCTIONAL TEST at least once per 31 days, rqd

c. CHANNEL CALIBRATION at least once per 18 months.

S4 LA SALLE - UNIT 2 3/4 3-jg

INSTRUMENTATION 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.8 The feedwater/ main turbine trip svitem actuation instrumentation channels shown in Table 3.3.8-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.8-2.

APPLICABILITY: OPERATIONAL CONDITION 1.

ACTION:

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

b. With the number of OPERABLE channels one less t1an required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

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

SURVEILLANCE RE0VIREMENTS 4.3,8.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.8.1-1.

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

LA SALLE - UNIT 2 3/43-pf S5

TABLE 3.3.8-1 5

m FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION

?

. m MINIM 11M c OPERABLE CilANNELS TRIP FUNCTION PER TRIP SYS1EM N a. Reactor Vessel Water Level-High, Level 8 3 ,

a w

e

- o

i TABLE 3.3.8-2 r-

[ FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS e

r-ALLOWABLE

[ IRIP FUNCTION TRIP SETPOINT VALUE z .

-4 a. Reactor Vessel Water Level-High, Level 8 < 55.5 inches * < 56.0 inches

R a

w ec N

See Bases Figure B 3/4 3-1.

l

[

TABLE 4.3.8.1-1 5

, FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS N f r;. :

, CilANNEL CliANNEL FUNCTIONAL CilANNEL

_E TRIP FUNCTION CitECK TEST' CALIBRATION a

ru a. React.or Vessel Water Level-liigh, S H R Level 8 i

9 w

i i

i L

i

D6LEW ENTtRE A4(,c-4.11 RADIOACTIVE-EFFLUENTS 3/ 1.1 LIQUID EFFLUENTS

-CONCE EATION LIMITING C DITION FOR OPERATION x <

3.11.1.1 The co htion of radioactive material rele ,ed from the site (see Figure 5.1.1 11 be limited to the concentra ons specified in 10 CFR Part 20, Append 1 ble II, Column 2 for radior clides other than dissolved or entraine gases. For dissolved o entrained noble gases,

-the concentration shal beMitedtotheconcentr ions specified in Table-3.11.1-1.

APPLICABILITY: At all time.

ACTION:

With the concentration of radioac ive ma rial released from the site exceeding the above limits, immediately rest et i concentration to within the above limits.

SURVEILLANCE REOUIREMENTS .

/ \Mx-A 4.11.1.1.1- The radioactivi content of each M % radioactive liquid waste shall be determined rior to release by sa' MI and analysis in accord-ance with Table 4.11.1-1 The results of pre-rel e alyses shall be used with the calculational ethods in the ODCM to assur 6 the concentration at the point of release i maintained within the limits Specification'3.11.1.1.

4.11.1.1.2 Post r ease analyses of samples composited rom batch releases

-shall be performe in accordance with Table 4.11.1-1. Th U4 31ts of the previous post re.eise analyses shall ce used with the calc nal methods in the 00CM to ssure that toe concentrations at the point o' -

se were maintained wi in the limits of Specification 3.11.1.1.

4.11.1.1.3 The radioactivity concentration of liquids discharge m continuous release p ints shall be determined by collection and analysis of mples in accorda e with Table 4.11.1-1. The results of the analyses shall used with the ca ulational methods in the 00CM to assure that the concentrati s at the poin of release are maintained within the limits of Specification 3.1.1.1.

LA SALLE - UNIT 2 3/4 11 l' l

DELETE En7/RE PAGC TABLE 3.11.1-1 MAXIMUM PFRMISSIBLE CONCENTRATION OF DISSOLVED OR ENTRAINED NOBLE GASES RELEASED FROM THE SITE TO UNRESTRICIED AREAS IN LlQUID WASTE NUCLIDE MPC(p ml)*

Kr 85 m b cE-4 85 SE-4 87 4E-5 88 9E-5 Ar 41 7E-5 Xe 131 m 7E-4 133 m SE-4 133

$ 6E-4 135 m 2E-4 135 A

/ 2E-4 e

o i

l- " Computed f m Equation 20 of ICRP Publication 2 (1959), adju r l infinite oud submersion in water, and R = 0.01 rem / week, p* = gm/cma ,

I and P /Pt = 1.0.

LA SALLE - UNIT 2 3/4 11-2

O EL E T*E' 9 / f/ W k s-i TABLE 4.11.1-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM TYPE OF OWER LIMIT LIQUID RE ASE SAMP' LING ANALYSIS. ACTIVITY OF DETECTION, TYPE FREQUENCY ANALYSIS FREQUENCY (LLD)

Q (pCi/ml)a A. Batch Waste P P Principal Gamma 5x10 7 Release patch Each Batch Emitter Tanks g

I-[1 1x10 8 P M issolved and 1x10 5 One Bate 'M Entrained Gases (Gamma emitters)

P M H-3 1x10 5 Each Batch omp site D Gross Alpha 1x10-7 P

Each Batch Composi Q gr-89,Sr-90 5x10 8 N .

5 1x10 8 7 B. Continuous W- iod 1 Gamma 5x10 7

1. Releases

Cont' uous C c Composite Emi te r[

/- I-131 1x10 5 M

Grab Sample M Dissolved n% 1x10 5 Entrained s (Gamma Emitt Continuous c

Composite M

c H-3 $\ 1x10J Gross Alpha 1x10 7 Q Sr-89, Sr-90 10 8 Continuous c Composite c Fe-55 1x1k8

\

LA SALLE - UNIT 2 3/4 11-3

__ - _ _ - _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ - - - _ - _ -

A fl.E W E N f/ M ' f M (

TABLE'4.11.1-1 (Continued) i TABLE NOTATION a.

The LLD is the smallest concentration of radioactive materi in a sample hat will be detected with 95% probability with 5% probab ity of falsely c cluding that a blank observation represents a "real" ignal.

For particular measurement system (which may includ radiochemical separ ion):

d4 4.66 s b E V 2.22x106 Y exp (-Mt)

Where:

b LLD is the " pr)(ki" lower limit of etection as defined above (as microcurie per uhi? mass or volume) sh is the standar deviation of le background counting rate or of -i the counting rate a blank s pie as appropriate (as counts per minute), '

E is the counting effic n (as counts per transformation),

V is the sample size (i u ts of mass or volume),

2.22x108 is the numbe of tra eformations per minute per microcurie, Y.is the fractiona radiochemica y (when applicable).

A is the-radioa ive decay-constant e particular radionuclide and for composite amples, and p G collection and At-is the e s 2d time between midpoint f sample o;

time of ce nting (for plant effluents, no environmental samples).

For bat samples taken and analyzed prior to A r se, at is taken to be zero.

Th value of s used'in the calculation of the LL hadetection s atem shall bN based on the actual observed varian e of the back-round counting rate or of the counting rate'of the ank samples (as appropriate) rather than on an unverified theoreti ally predicted variance. Typical values of E, V, Y, and at shall be u d in the

calculation.

j. b. A composite sample is ene in which the quantity of liquid sa led is !

l proportional to the quantity of liquid waste discharged and i which the method of sample employed results in a specimen which is representative of the liquids released.

LA SALLE - UNIT 2 3/4 11-4 Amendment No.14

WW 6/V7/fC fb6E

/

TABLE 4.11.1-1 (Continued) / i TABLE NOTATION / l t'

To be representative of the quantities and concentrations of' ,

radioactive materials in liquid effluents, samples shall be !

collected continuously in preportion to the rate of floy/of the ]

fflu t' stream. Prior to anar ses, all samples taken or the l c c to shall be thoroughly uxed in order for the omposite ;

san le be representativa of the effluent release

d. A batc (el se is the discharge of liquid waste f a discrete ,

volume, P to sampling for analyses,-each b tch shall be 1 isolated, d hen thoroughly mixed, by a met d described in the 00CM, to a t representative sampling,

e. A continuous '

is the discharge of Iquid wastes of a nondiscrete vol me; e.g., from a volume f system that has an input flow during the ntinuous release.

f. The principal gamma mitters for wh' h the LLD specification applies exclusively are the t lowing radi nuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs 134, Cs- 7, Ce-141, and Ce-144. This list does not mean that only ese n clides are to be detected and reported. Other peaks whi 1 e measurable and identifiable, at the 95% confidence level, toget' with the above nuclides, shall also be identified and reported.

-N~

m m

7 G'

LA SALLE - UNIT 2 3/4 11-5

MMT ENT/G MG E RADIDACTIVE EFFLUENTS

. 05E L ll]NG CONDITION FOR OPERAll0N $

I 3.11.1. The dose or dose commitment to an individual from radi itive materials 'n liquid effluents released, from each reactor unit, from the sita (see figure 1.1-1) shall be limited:

a. Duri ' calendar quarter to less than or equa to 1.5 mrem to the

.- nd to less than or equal to 5 mrem o any organ, and

~

total

b. During a - endar year to less than or eqt; to 3 mrem to the total body ( , o less than or equal to 10 fnrem to any organ.

APPLICABILITY: At all i /

ACTION:

a. With the calculate dose from the elease of radioactive materials in liquid effluents xceeding a of the above limits, in lieu of any other report requ ed by Se cification 6.6.A, prepare and submit l to the Commission with) 30 ys, pursuant to Specification 6.6.C, a Special Report which ide 'i es the cause(s) for exceeding the limit (s) and defines the rrective actions to be taken to reduce the releases of radioatt' re ,aterials in liquid ef fluents during the remainder of the curre cale dar quarter and during the subsequent three calendar quarte , so th cumulative dose or dose commit-ment to an individu from these ses is within 3 mrem to the total body and 10 em to any org @ is Special Report shall also include the radi ogical impact on In " ed drinking water supplies at the nearest 'awnstream drinking w er ource.

/

b. The provisio of Specifications 3.0.3 N 0.4 are not applicable.

SURVEILL ANCE RE0' REMENTS 4

.y 4.11.1.2 Dq/eCalculations. Cumulative dose contributions ' quid effluents shall be d ermined in accordance with the ODCM at least once n days.

1 LA SALLE - UNIT 2 3/4 11-6 Amendment No.ll

O6LEYC ENftK MfE RADIDACTIVE EFFLUENTS-

. . . 001D WASTE TREATMENT SYSTEM LIMI NG-CONDITION FOR OPERATION

)

3.11.1.3 e liquid radwaste _ treatment system shall be OPERA E. The 1 appropriate rt s of the system shall be used to reduce t radioactive l materials in astes prior _to their discharge when t projected doses due to the liqu luent from each reactor unit, from t site (see Figure 5.1.1-1), h averaged over 31 days, would exte 0.06 mr'em to the total body or 0.2 o-any organ.

APPLICABILITY: At al tig.

ACTION:

a. With the liquid r dwaste treatment s stem inoparable for more than 31 days or with ra oactive liquid aste being discharged without treatment and in ext ss of the ab e limits, in lieu of any other report required by Sp ificatio 6.6.A prepare and submit to the Commission within 30 da s purs nt to Specification 6.6.C a l ,

l Special Report which inc de the following information:

1. Identification of th operable equipment or subsystems and the_ reason for inop .abi ity,

~2. Action (s) taken restore- . operable equipment to OPERABLE status, and

3. Summary desc ption of action (s h to prevent a recurrence.

l b. -The provisions f Specifications 3.0.3 . .4 are not applicable.

l -

I

-SURVEILLANCE RE001P MENTS w

< s 2

'V l

l- 4.11.'1.3.1 D ses.due to liquid releases shall be projected a least once per 6N i

31 days, in ccordance with the ODCM.

-4.11.1.3. The liquid radwaste treatment system shall be demonstr ted OPERABLE by oper ing the liquid radwaste treatment system equipment for at ast 30 mi, tes 'at least once per 92 days unless the liquid radwaste syst -has been util' ed to process radioactive liquid effluents during the previous 9 days.

1 I

l LA SALLE - UNIT 2. 3/4 11-7 Amendment No.11

i RADI0 ACTIVE EFFLUENTS LIQUID HOLOUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outside temporary tanks shall be limited to less than or equal to the limits calculated in the ODCM.

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit,

b. -The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

[ SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a l representative sample of the tank's contents at least once per 7 days when l

radioactive materials are being added to the tank.

t l

l l

1 l

l l

l LA SALLE - UNIT 2 3/411s/g

.DELET ENTitt PAGE RADI0 ACTIVE EFFLUENTS 11.2 GASEOUS EFFLUENTS i 005 TE LIMITIN CONDITION FOR OPERATION

/

3.11.2.1 T dose rate due to radioactive materials released /in gaseous

/

effluents fro the site (see Figure 5.1.1-1) shall be limited to the follow'ng:

a. For no e C es: Less than or equal to 500 mreds/yr l to the total body an 1 than or equal- to 3000 mrems/yr the skin, and

b. -For all ra 'o p nes and for all radioacti materials in particulate form and rad nuc des (other than noble ases) with half lives greater than d Less than or equa to 1500 mrems/yr to any organ via the i . n pathway.

, APPLICABILITY: At all times.

ACTION:

With the dose rate (s) exceeding th abo limits, immediately decrease the release rate to within the above lim s).

SURVEILLANCE REQUIREMENTS

, s 4.11.2.1.-1 The dose rate due noble gases heous effluents shall be determined to be within the ovelimitsinacorgcewiththemethodsand procedures of the ODCM.

4.11.2.1.2 The dise rat due to radioactive mater b her than noble gases, in gaseous efflu nts shall be determined to in the above imits in accordance with th methods and procedures of: the C obtaining represen-tative samples-and p rforming analyses in accordance wi a.pling and analysis program specified Table 4.11.2-1.

7 9

'LA SALLE - UNIT 2 3/4 11-9

l TABLE 4.11.2-1 5 RAD 10ACTItE GASEOUS WASTE SAMPtING AND ANALYSIS .'ROGRAM

, l Minimum Lcwer ti t of 7 Sampling Analysis Type of Detec og (LLD) c Gaseous Release Type equency Frcquency Activity Analysis ( " / mil a N '

b P

b -4

n. Containment Vent Each Pu e Each ourge Print.ipal Gamma Emitte ^6Ix10 and Purge Syste;a Grab -6 Sample 11- 3 //\ 7 Ix10 b rs 9 lx10 B. Main Vent Stack MD M Prir.cipal Ga N

m Grab -6 Sample C C 11- 3 / 47' lx10 C. Standby Gas D W -4 Treatment System Grab Pr' cipal Gamma Emitters 9 lx10

Sample

-N

- D. All Release lypes Continuous # bd -131 1x10 T as listed in A and Charcoal O B above, at the Sample _ I-lh lx10

- 1"-

~ll 5

C ntinu us Principa Gammar mitters9 lx10 ted C ve- ar i ate (1-131,Ot. s) at the sects whenever there Eu' g

Gross Alpha -11 is flow.

9 lx10 g Continuous \[mposite g Particulate N Sample N t ' I M Con *duous Q Sr-89, S r-90 lx10 g Composite q%

Q Particulate Semple Q lx10-6 (L 133 N

.jontinuous# Noble Gas Noble Gases Monitor Gross Beta & Ga.ma equivalent)

N :

,~, .

DELEY EN7/RG" fhGE TABLE 4.11.2-1 (Continued)

TABLE NOTATION

a. he LLD is the smallest concentration of radioactive material n a sample ti t'will be detected with 95% probability with 5% probabil' y of falsely con uding that a blank observation represents a "real" si nai.

For a t rticular measurement system (which may include r diochemical separati ):

4.66 c,b D= 'V 2.?2x105 Y exp (-Aat)

Where: %

, t.LD is the "a p i r ' lower limit of d .ection as defined above (as microcurie p. -

mass or vnlu ),

b is the standard viation of th background counting rate or of s ,

tne counting rate of blank sam e as appropriate (as counts per minute),

E is the counting efficien as counts per transformation),

V is the sample size (in it of mass or volume),

2.22x106 is the number f transf mati ns per minute per microcurie, Y is the fractional adiochemical y - when applicable).

A is the radioac ve decay constant fo harticularradionuclide,and at is the ela ed time between midpoint o .a e collection and time of cou ing (for plant ef fluents, not t )v mental samples).

The value of s used in the calculation of the LD for a detection system

all bh based on the actual observed var 3 M f the back-groun counting rate or of the counting rate of tt k samples (as propriate) rather than on an unverified theor i predicted var ance. Typics: values of E. V, Y, and At shall be n the c .culation.

LA SALLE - UNIT 2 3/4 11-11

klETC l~Atrigg- MGC TABLE 4.11.2-1 (Continued)

TABLE NOTATION (Continued) {

i i

b Analyses shall diso be performed following shutdown, star up, or a THERMAL POWER change exceeding 15% of the RATED THERMAL '0WER l ithin a 1-hour period.

l

c. Wh bthereisflowthroughtheSBGTS.

d. Sampi shall be changed at least once per 7 day and analyses shall becompe\e ithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing or fter removal from sampler. 5 ng shall also be perfo'rmt.d at east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at lea t s following each shutdown startup or THERMAL POWER change exce i 5% of RATED THERMAL POW in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses completed wit n hours of changing, en samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are an the correspondin LLDs may be increased by a-factor of 10. T s requirement does t apply if (1) analysis shows that the DOSE EQU LENT I-131 conce ration in the primary coolant-has not increased m e than a fact of 3; and (2) the noble gas monitor shows that of uent activ' y has not increased more than a factor of 3.

e. Tritium grab samples shal b taken at least once per 7 days from the plant vent to determine ritium releases in the ventilation i exhaust from the spent fu -p 1 area whenever spent fuel is in the spent fuel pool.

f. The ratio of the samp flow rate o sampled stream flow rate shall be known for e time period v by each dose or dose rate calculation made i accordance with e cations 3.11.2.1, 3.11.2.2, and 3.11.2.3.

g. The principal amma emitters for which th Lhecificationapplies include the llowing iadionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, r i

Xe-135, an Xe-130 for gaseous emissions and n-54, Fe-59 Co-58, r Co-60, Zn- 5 Mo-99, Cs-134 Cs-137, Ce-141 an Ce-144 for particu-late emi sions, This list does not mean that on M e nuclides are to e detected and reported. Other peaks whi. measureble and i entifiable, at the 95% confidence level, toge th the abo nuclides, shall also be identified and reporte G\

LA SALLE - UNIT 2 3/4 11-12

DELETC EN7tsc fMg MD10ACTIVEtrFLUENTS /

Dose - NDBtf GASES L1 TING CONDITION FOR OPERATION / ,

s e 3.11.2.2 The air dose due to noble gases released in gaseo s ef fluents, f rom each react r unit, trom the site (see figure 5.1.1-1) 7,ha be limited to the follo'ing:

C

a. Duri okcalendar quarter: Less than or qual to b rerad for gamma radiat'o d less than or equal to 10 mr for beta radiation, and

b. During ar.Wtndar year: Less than or equal to 10 mrad for gamma radiation 'idAss than or equal to 2' mrad for beta radiation.

APPLICABILITY:

At all ti es.

ACTION:

a. With the calculated a'r dos from radioactive noble gases in gaseous offluents exceeding an o the above limits, in lieu of any other report required by Spec) ication 6.6.A, prepare and submit to the l Commission within 30 d 5, pursuant to Specification 6.6.C, a Special Report which 'denti ies the cause(s) for exceeding the limit (s) and define' the cor >ctive actions to be taken to reduce the releases and t .e proposed tive actions to be taken to assure that s sequent relea. ill be in compliance with the above limit . .

A

b. The provisio s of Specifications 3. 3 ft:$ 3.0.4 are not applicable.

b SURVEILLANCE REOU).EMENTS

, s 4

4.11.2.2 0 e Calculations Cumulctive 1ose contributio s the current calendar arter ano current calendar p ar shall be deter.' n accordance with the DCM at least once per 31 days.

LA SALLE - UNIT 2 3/4 11-13 Amendment No.11

1 OflE7E EN77RC P/lGE RAD 10 ACTIVE frrtutNT;

. .. DD.r . RAD 1010 DINES RADICACTIVE MATERI Als IN PARTICULATE FORM, AND RAD)JNUCt1CES DE THAW tGBLE CASE 5

_ LIMITING (DNDITIONFOROPERATION

+

3.11.2.3 The b

~5t an indiviaval from radiciodines and r ioactive naterials in particulate r ,. , d radionuclides, other than noble g es, with half-lives greater than 8 da o seous ef fluents released, f rom e ch reactor unit, f rom the site (see Figu -1) shall be limited to the f lowing:

a. During any le. a quarter: Less than or > qual to 7.5 mrem 5 to any organ, and

b. During any talen ar year: Less than o equal to 15 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose f the release of radiciodines, radioactive materials in particulate f . , or radionuclides (other than noble gases) with half-lives gr ate than 8 days, in gaseous ef fluents exceeding any of the ab <e limi s, i lieu of any other report required by Specificat' n 6.6.A, ..

e and submit to the l Commission within 30 ays, pursuan etification 6.6.C, a Special Report whic identifies the a s) for exceeding the limit and defines the c rective actions th i e been taken to reduce the releases and the reposed corrective a o be taken to assure that subsequen releases will be in comp i with the above limits,

b. The provisio s of Specifications 3.0.3 and , are not applicable.

/0 SURVE1LL ANCE REO' P.EME NT S

, S y

4.11.2.3 .se Calculations Cumulative dose contributions for th- current calendar varter anc current calendar year shall be determined in tcorcance with th ODCM at least once per 31 days.

LA SALLE - UNIT 2 3/4 I H 4 A endment No.11 i

NLETE EN7/K'E MQC RAD 10ACT!vE EFrLUENTS

, .. ASEOUS RADWASTE TREATMENT SYSTIM LIM !G COND1110N FOR OPERATION s

ip e

3.11.2.4 Se GASEOUS RADWASTE TRI A1 MENT SYSTEM shall be in peration.

APPLICABILITY. ver the main condenser air ejector s stem is in operation.

ACTION: (

a. With the RADWASTE TREATMENT SYSTE' inoperable for more than 7 days, in ie f any other report req red by Specification 6.6.A, l prepare and .,'b to the Commission w' hin 30 days. nursuant to Specificati '6.6.C, a Special Re .rt which i" ' J, the following information:

1. Identificatio of the inope able equiph at or subsystems and the reason for operabili y,

2. Action (s) taken to est re the inoperable equipment to OPERABLE status, aiid

3. Summary descriptio< of tion (s) taken to prevent a recurrence.

b. The provisions of Sp> ification. and 3.0.4 are not applicable.

A

. /

SURVE!LLANCE RE001REMENT.

i 1 -- 3 4.11.2.4 The GASE0' RADWASTE TREATMENT SYSTEM shall e g ified to be in operation at leas once per 7 days. j h

LA SALLE - UNIT 2 3/4 11-15 loendment No.11

naETC &YRPf 89GE i

RADIDACTIVE EFFLUENTS

'ENTIL ATION EXHAUST TREATMENT SYSTEM l

LIM] NG CONDITION FOR OPERATION t l

d ;I 3.11.2.5 h h4opriate portions of the VENTILATION EXHAUS 1REATHENT SYSTEM shall be OP AB nd be used to reduce radioactive trateri s in gaseous waste prior to thei c rge when the projected doses due to seous effluent releases from e et -tor unit, from the site (see Fig e 5.1.1-1), when averaged over 31 a> s uld exceed 0.3 mrem to any or n.

APPLICABILITY: At ps.

ACTION:

a. With the VENTIL ION EXHAUST TREATMr4T SYSTEM inoperable for more than 31 days, or ith gaseous was being discharged without treatment and in excess of the a ve limits, in lieu of any other report required by 5'pecificatio 6.6.A, prepare and submit l to the Commission with n 30 d 's, pursuant to Specification 6.6.C, a Special Report which in lude the following inforrtation:

1. Identification of th inoperable equipment or subsystems and the reason for ino >ra flity,

2. Action (s) taken o restor inoperable equipment to OPERABLE status, and

3. Summary desgription of action ( ken to prevent a recurrence.

b. The provisions /of Specifications 3.0.3P g0.4arenotapplicable.

/

SURVEILLANCE REOU)REMENTS x

e -- .

i / N i

/ &

4.11.2.5.1 700ses due to gaseous releases from the site shal -

ojected at least once/per 31 days in accordance with the ODCM.

/

4.11.2[3.2 OPERAp E by The VENTILATION operating the VENTILATIONEXHAUST TREATMENT EXHAUST TREATMcni SYSTEM equ SYSTEM ment for shall be de atJIast30 minutes,atleastonceper92daysunlesstheappropriat system hap been utilized to process radioactive gaseous effluents during the revious SI days.

l LA SALLE - UNIT 2 3/4 11-16 Amendment No.11

l RADI0 ACTIVE EFFLUENTS ;

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

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTION:

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limits as required by Table 3.3.7.11-1 of Specification 3.3.7.11.

l 1

LA SALLE - UNIT 2 3/4 11-)/2

1

\

RADI0 ACTIVE EFFLUENTS MAIN CONDENSER I

LIMITING CONDITION FOR OPERATION i

l 3.11.2.7 The release rate of the sum of the activities from the noble gases measured prior to the holdup line shall be limited to less than or equal to 3.4 x 105 microcuries/second.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the release rate of the sum of the activities of the noble gases prior to the holdup line exceeding 3.4 x 10% microcuries/second restore the release rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP with the main steam isolation valves closed within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS 4.11.2.7.1 The radioactivity rate of noble gases prior to the holdup line shall be continuously monitored in accordance with Specification 3.3.7.11.

4.11.2.7.2 The release rate of the sum of the activities from noble gases prior to the holdup line shall be determined to be within the limits of Specir ' tion 3.11.2.7 at the following frequencies by performing an isotopic analy! of a representative sample of gases taken prior to the holdup line.

a. At least once per 31 days.

l

b. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the off gas pre-treatment Noble Gas Activity Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in

, the nominal steady state fission gas release from the primary coolant.

l LA SALLE - UNIT 2 3/4 11-)6 3

AELE7E ENT/W /% GE' RADIOACTIVE EFFLUENTS NTING OR PURGING LIMIT F CONDITION FOR OPERATION x ,

3.11.2.8 VE PURGING of the containment drywell sha be through the Primary Contai e nt and Purge System or the Standby G s Treatment System.

APPLICABILITY: Wn he drywell is vented or purge .

ACTION:

K

a. With the requi'e s of the above epec'fication not satisfied, suspend all VEN 'NG and PURGING of th drywell.

b. The provisions of s ecifications 3 .3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

\/

4.11.2.8.1 The containment dryw 1 shall ermined to be aligned for VENTING or PURGING through the Primary ontainment .n d Purge System or the Standby Gas Treatment System within 4 ours prior to ta of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or P RGING of the dryw 1.f'O 4.11.2.8.2 Prior to use f the Purge System thro h Standby Gas Treatment System in OPERATIONAL C DITION 1, 2 or 3 assure th t.

l a. Both Stand Gas Treatment System trains are PERA E, and

b. Only on of the Standby Gas Treatment System tr nsAs used for PURGD .

l p@

l l

LA SALLE - UNIT 2 3/4 11-19

1 bELETE EN7tK( /HG['

l l

l RAD 10 ACTIVE EFFLUENTS

'4.11.3 50L1D RADIDACTIVE WASIE LIMll .(3 CONDITION FUR OPERATION 3.11.3 The olid rac.aste system shall be OPERABLE and used, s applicable in accordance wi h a P 3CE55 CONTROL PROGRAM, for the 50LIDif]C 10N and pac 6 aging of radioactive ae to ensure meeting the reauirements of 10 CFR Part 20 and of ID CFR Part pptoshipmentofradioactivewastes rom the site.

APPLICABILITY: At 11 i ACTION: &s.

a. With the packa 'ng uirements of 10 R Part 20 and/or 10 CFR Part 71 not sati if , suspend shipme 5 of defectively packaged solid radioactive 'astes from the s' e.

b. With the solid rada'a e system i perable for more than 31 days, in lieu of any other r nort req red by Specification 6.6.A, g prepare and submit to tt Comm' sion within 30 days, pursuant to Specification 6.6.C. a Sp ri Report which includes the following inf ormation:

1. Identification of tt in ,erable equipment or subsystems and the reason for ino*erabili '

2. Action (s) taken o restore th e able equipment to OPERABLE status,

3. A descripti ) of the alternative ed r 50LIDiflCAT10N and packaging f radioactive wastes, an p

4. Summary description of action (s) taken .o vent a recurrence.

c. The provi ons of Specifications 3.0.3 and 3.0. are not applicable.

A SURVEILLANCE RE. IREMENTS

, ,I, ,

4.11.3.1 e solid rad aste system shall be demonstrated OPERABL east once per .2 days by:

Operating the solid rada'aste system at least once in the pre ious 92 days in accordance witt the Process Control Program, or

b. Verification of the existence of a valid contract for SOLIDIFICA 'ON to L.e performed by a contractor in accordance with a PROCESS CONih L PROGRAM.

LA SALLE - UNIT 2 3/4 11-20 A endment No.li

DELETh~ ENTIRE PAGE RADI0 ACTIVE EFFLUENTS Sus 'EILLANCE 1

REQUIREMENTS (Continued)

I 4.11.3.2 THE PROCESS CONTROL PROGRAM shall be used to verify tb.

SOLIDIFICA '0N of at least one representative test specimen fr<a at least every tenth sitch M each type of wet radioactive waste (e.g , filter sludges, spent resins, vaQ@torbottoms,andsodiumsulfatesoluti s).

1

a. If an e ecimen fails to verifv SOLIDIFIC ION, the SOLIDfFI TN f the batch under t'est shall e suspended until such !

time as a i' test specimens can be ob ained, alternative SOLIDIFICAT N aymeters can be determin in accordance with the PROCESS CONT- PFA AM, and a subseque test verifies SOLIDIFICA-TION. SOLIDIF "AT f the batch may hen be resumed using the alternative 50L ION parameter determined by the PROCESS CONTROL PROGRAM.

b. If the initial test s ecimen fro a batch of waste fails to verify SOLIDIFICATION, the PR ESS C0 OL PROGRAM shall provide for the ,

collection and testing o rep >sentative test specimens from each consecutive batch of the . , type of wet waste until at least 3 consecutive initial test ecimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM e al be modified as required, as provided in Specification 6.7, t assur SOLIDIFICATION of subsequent batches of waste.

b tAe 6

o i

e@ ,

l l

LA SALLE - UNIT 2 3/4 11-21 l' )

l i . . . .- --

BEL.ETC ENTAE PAGE RADIDACTIVE EFFLUENTS

.. 3/4;11.4 TOTAL DOSE

_L MITING CONDITION FOR OPERATION x ,,

3.11.4 The e or dose commitment to any member of the pu ic, due to releases of radio t and radiation, from uranium fuel cycle so ces shall be limited to le an or equal to 25 mrem to the total bo or any organ (except the thyroid, ich shall be limited to less than or equ to 75 mrem) over 12 consecutiv nh .

APPLICABILITY: 11 imes.

ACTION:

a. With the cal lated doses from the ' lease of radioactive materials in liquid or g seous ef fluents ext eding twice the limits of Specifica-tions 3.11.1.2. 3.11.1.2.b 3. .2.2.a. 3.11.2.2.b, 3.11.2.3.a. or 3.11.2.3.b, in 1 of any othe report required by Specification 6.6.A, prepare and ubmit, pur uant to Specification 6.6.C. a Special Report to th Direct , Nuclear Reactor Regulation, U.S.

Nuclear _ Regulatory Co issi n, Washington, D.C. 20555, within 30 days, which defines corrective action to be taken to reduce subsequent releases to p vent recurrence of exceeding the limits of Specification 3.11.4. hi Special Report shall include an analysis which estimates the r diatio e ute (dose) to a member of the public from uranium uel cyc1 su es (including all effluents pathways and direc radiation) consecutive month period that includes th release (s-) cov e this report. If the esti-mated dose (s) e eeds the limits o t fication-3.11.4, and if the release condi on resulting in viol se% f 40 CFR 190 has not already been orrected, the Special R hall include a request for a vari ce in accordance with the ovisions of 40 CFR 190 and including he specified information of 6 90.11. Submittal of the report i considered a timely request, an a p(Diance is granted until aff action on the request is com,,le , e variance only relat to the limits of 40 CFR 190, and does n ply in any way

to e requirements for dose limitation of 10 t 20, as ad essed in other sections of this technical sp tion,

b. he provisions of Specifications 3.0.3 and 3.0.4 ar not applicable.

SURVEI1 LANCE REQUIREMENTS

- s

/.11.4 Dose Calculations Cumulative dose contributions from liquid an gaseous ef fluents shall be determined in accordance with Specifications 11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCH.

LA SALLE - UNIT 2 3/4 11-22 Amendment No.11

DELE TE Et4TIFE P% E

/

3/4. 2 RADIOLOGICAL ENVIRONWNT AL MON 110 RING ,/

~~

3/4.1 MONITORING PROGRAM LIM 111NG NDITION FOR OPERATION 3.12.1 The r h ital environmental monitoring program shall [e' conducted as specified i T 3.12.1-1.

APPLICABILITY: A a mes.

ACTION:

A .

a. With the radi o 1 environmental monitor' g program not being conducted as sp fied in Table 3.12.1-1, n lieu of any other report required t > Specification 6.6. A. repare and submit to the l Commission, in the nnual Qadiological perating Report, a description of the reasons for 't conducting the rogram as required and the plans for preventing recurrence.

b. Withthelevelofradica+ivityi an environmental sampling medium exceeding the reporting le els t Table 3.12.1-2 when averaged over any calendar quarter, in li f any other report required by Specifi-cation 6.6.A; prepare and su' it to the Commission within 30 days from the end of the affect c endar quarter a Special Report pursuant to Specification / .6.C. When more than one of the radio-nuclides in Table 3.12. *'2 are de, ct n the sampling medium, this report shall be submit d if:

h concentration f ) , concentration (2k *> 1*0 limit level ) limit level ( ~ j -

When radionuti es other than those in Tab 4 1-2 are detected and are the r sult of plant effluents, this >irt shall be submitted 5 if the pote ial annual dose to an individual 's equal to or greater than the c endar year limits of Specifications .11.1.2, 3.11.2.2 and 3.11. 3. This report is not required if th m @ ed level of

, radioat Avity was not the result of plant effluent $ ver, in such event, the condition shall be reported and set d in the Annu Radiological Environmental Operating Report.

c. W h milk or fresh leafy vegetable samples unavailable f @oneor m ore of the sample locations required by Table 3.12.1-1, 1 lieu of any other report required by Specification 6.6.A, prepare a submit l to the Commission within 30 days, pursuant to Specification 6.C, a Special Report which identifies the cause of the unavailabilit) of samples and identifies locations for obtaining replacement samp s.

The locations from which samples were unavailable may then be del ed from those required by Table 3.12.1-1, provided the locations f rom which the replacement samplos were obtained are added to the environ mental monitoring program as replacement locations.

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, LA SALLE - UNIT 2 3/4 12-1 Amendment No. 11

{

J

OELETE CMif RE PMC RA OLOGICAL ENVIRONMENTAL MONITORING SURV NCE REQUIREMENT 5 s -

4.12.1 The r diological environmental monitoring samples shall t ollected pursuant to Te le 3.12.1-1 from the locations given in the tablf and figure in the ODCM and sh I be analyzed pursuant to the requirement of ables 3.12,1-1 and 4.12.1-1.

n

'7/

Y 4

/

+,

J 4

9 e

/

9

/

LA SALLE - UNIT 2 3/4 12-2

TABLE 3.12.1-1 y RADIOLOGICAL FNVIRONMENTAL HONITORING PROGRAM

, l-m MINIMUM NUMBER OF SAMPLES g EXPOSURE PATHWAY AND SAMPLING AND TYPE At FREQUENCY g AND/OR SAMPLE SAMPLE LOCATIONS

COLLECTION FREQUENCY . ANALYSIS ca

1. AIRBORNE O

! Radioiodine and 5 loca 'ons Continuous operation o Particulates sampler with saeple -

hadioiodinecanister.

Analyze at least once lection as requir b per 7 days for I-131.

dust loading b at 1 st once per 7 s. Particulate sampler.

Analyze for gross beta radioactivity > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months y

/ following filter change.

Perform gamma isotopic g analysis on each sample 4 when gross beta activity is > 10 times the yearly mean of control samples.

[ Perform gamma isotopic analysis oa composite (by location) sample at least once per g 92 days.

q

2. DIRECT RADIATION Locations At least once per 31 days. a dose. At least

_ 2 dosimeters or > 1 or onc er 31 days. I instrument for con- r tinuously measuring At least once per 92 days. Gamma doh At least and recording dose (Read-out frequencies aru once per 9 ays.

rate at each determined by type of dosim-location. eters selected.)

9 1,'

Wy

t

> TABLE 3.12.1-1 (Continued) g RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM P

rn MINIMUM g NUMBER OF SAMPLES q EXPOSURE PATHWAY AND SAMPLING AND TYPE FREQUENCY y AND/0R SAMPLE SAMPLE LOCATIONS

COLLECTION FREQUENCY F, ANALYSIS

3. WATERBORNE

> a. Surface 2 locat~ ns Composite sample coli e A9 Gama isotopic analysis over a period of < ds of each composite sample.

Tritita analysis of com-posite sample at least once per 92 days.

Y

b. Ground S locations east once per 92 days. Gama isotopic and y tritita analyses of a each sample.

c. Sediment from 1 location At least once p 184 days. Gamma isotopic analysis Shoreline s}h of each sample. g S

a a 9

n h

k a

> TABLE 3.12.1-1 (Continued)

'y RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM F

m MINIMUM g NUMBER OF SAMPLES

~; EXPOSURE PATifWAY AND SAMPLING AND TYPE AND dEQUENCY

, AND/OR SAMPLE SAMPLE LOCATIONS

0 NALYSIS

. COLLECTION FREQUENCY

4. INGESTION g b

a. Milk 3 locations At least once per 15 da bmmaisotopicand

, when animals are on urhI-131 analysis at least once per a ', of each sample.

at other times.

y b. Fish 2 locations One sa e in season, or at Gansna isotopic analysis ence per 184 days if on edible portions.

g %easonal.

J, sT NOTATION

/

Sample locations are shown and descri _d ig ,tt OCM.

o

\X m r-f x

U cn

' 2 i d.

h 70 41 '

\ D '

O 1

TABLE 3.12.1-2 5

y, REPORTING LEVELS FGR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES

?

E Reporting levels Water Airborne Particulate Fish Milk Fo Products m Analysis (PCi/l) r Gases (pCi/m3) (pCi/Kg,.tet) (pCi/1) Ci/Kg, wet) j H-3 2 x 10 (a)

O I

Mn-54 'l x 103 3 x 10* /

Fe-59 4 x 10 2 1 x 10 4 Co-58 1 x 10 3 3- 0 4

, Co-60 3 x 10 2 1 x 10 4

Zn-65 3 x 10 2 , , gg 4

., Z r-Nb-95 4 x 10 2 1-131 2 0.9 T 3 1 x 10 2 Cs-134 30 10 1 x 10 3 60 1 x 10 3 6

Cs-137 50 2 2 x 10 3 2 x 10 3 g k

Ba-La-140 2 x 10 2 3 x 10' y

ct, D

(* For drink' g water samples. This is 40 CFR Part 141 value. . D 9

}

m i

h

't TABLE 4.12.1-1 l

[ HAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a,c

-?

E E Water Airborne Particulate Fish Hilk Food Produ s Sediment

+

q' Analysis (PCi/1) or Gases (pCi/m3 ) (pCi/Kg, wet) (pCi/1) (pCi/K etg (pCi/kg, dry) 1 x 10 ~2 gross beta +5 1000 NA NA 2000 !

, H-3 200 NA **

NA o NA Mn-54 * **

NA * ** **

Fe-59

NA ** * ** **

3 Co-58,60

NA ** * ** **

R*

Zn-65

NA ** * ** **

i M. Zr-95

NA ** * ** **

i Nb-95

NA ** * ** **

I-131 NA 10 x I ~ **

0. 30 **

Cs-134 10 x 10 -2 100 10 ** **

Cs-137 10 ~2 N U i x 10 100 10 * ** b f qt Ba-140

NA ** * ** **

1 l la-140 *

/ NA ** * ** ** I I

j Gamma is opic analysis provides LLD of s 20pCi/1 per nuclide.

N I

Ga. isotopic analysis provides LLD of s20 pCi/1 per nuclide. l

. e f

DELE T6 GWAC PAGE

/

TABLE 4.12.1-1 (Continued)

TABLE NOTATION

a. The LD is the smallest concentration of radioactive material a sample that ill be detected with 95% probability with 5% probabilit f falsely conclu ing st a blank observation represents a "real" sign- .

For a par ic measurement system (which may include r..iochemical separation

(

LLD =

E

\

4.66 s b 22 . Y . exp (-Aat)

/

LLD is the "a pric 1" lower limit of e'tection as defined above (as picocurie per unit ss or volume),

s h is the standard dev ition of t > background counting rate or of tne counting rate of a t ank sa e as appropriate (as counts per minute),

E is the counting efficienc; as counts per transformation),

/

V is the sample size (i / units f tr or volume),

2.22 is the number o transformati< y minute per picocurie, Y is the fraction- radiochemical yie n applicable),

s the radio tive decay constant for Me icular radionuclide, at is the lapsed time between sample collec 'on (or end of the sample lection period) and time of counting for vironment sample not plant effluents.

The valu s used in the calculation of the LLD for @tionsystem

, shall b based bon the actual observed variance of the bac M ' counting l rate of the counting rate of the blank samples (as appr 'r < () rather tha on an unverified theoretically predicted variance. In 1.culating th LLO for a radionuclide determined by gamma-ray spectromett', the b 'ckground shall include the typical contributions of other ra anuclides iormally present in the samples (e.g. potassium-40 in milk sampit ).

Typical values of E, V, Y, and at shall be used in the calculation.

b. LLD for drinking water.

/ c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported.

LA SALLE - UNIT 2 3/4 12-8

DaETE em,ee MW l i

l RADIOLOGICAL ENVIRONMENTAL H3NITORING l

.32.2 LAND USE CENSUS

\ I .

,LIMllMG CONDITION FOR OPERATION I j

f 3.12.2 A hsecensusshallbeconductedandshallidentiythelocation of the near i animal and the nearest residence in eact of the 16 meteor-ologic.11 sect hin a distance of five miles. (For el ated releases as !

defined in Reg a ry Guide 1.111, Revision 1. July 1977 the land use census shall also ident locations of all milk animals i each of the 16 meteoro- i logicial sectors w distance of three miles.)

Appl?CABILITY: At a k.

ACTION:

a. With a land use nsus identifying location (s) which yields a calculated dose or ose commitmen greater than the values currently being calculated in pecificati 4.11.2.3, in lieu of any other report required by Sp ificati 6.6.A prepare and submit to the l

Commission within 30 d 's, pr suant to Specification 6.6.C., a ",

Special Report which ide ti les the new location (s).

b. With a land use census i en ifying a location (s) which yields a calculated dose or dos comm ment (via the same exposure pathway) 20 percent greater th at a 1 a from which samples are currently being obtained in a ordance wi- fication 3.12.1, in lieu of any other report r uired by Spec rdr fi the Commission w hin 30 days, purs an'.Jn 6.6A, prepare te Specification and, submit 6.6.C. a to l Special Report ich identifies the w shall be adde to the radiological ens'Ttte A ation.

new locationropmental monitoring within 30 da s. The sampling location, 'ng the control station location, t ving the lowest calculated do e ose commitment (via the same posure pathway) may be deleted s monitoring program fter (October 31) of the year in wh' is land use census was co ucted.

c. The rovisions of Specifications 3.0.3 and 3.0.4 applicable.

SURVEILLAN REQUIREMENTS @

4.12.

VT The--land use census shall be conducted at least once per 1 months betw n the dates of (June 1 and October 1) using that information ich will pr ide the best results, such as by a door-to-door survey, aer ial su vey, or b consulting local agriculture authorities.

ll LA SALLE - UNIT 2 3/4 12-9 Amendment No.11

OR&TE EN7/2C $ACC l

A010 LOGICAL ENVIRONMENTAL MONITORING L 12. 3 INTERLABORATORY COMPARISON PROGRAM LIMIT CONDITION FOR OPERATION 3.12.3 Analy. be performed on radioactive materials upplied as part of an Interlebo'a Comparison Program which has been ap oved by the Commission.

APPLICABILITY: At b1 es.

ACTION:

a. With analyses n being performed as rrquired above, report the corrective actio taken to prevent recurrence to the Commission in the Annual Radi logical Environm-.tal Operating Report,

b. The provisions of Spe ifications a.O.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

/ \

4.12.3 A summary of the results obtained s of the above required Inter-laboratory Comparison Program .d in accor n h the ODCM (or participants in the EPA crosscheck progra shall provide 5 program code designation for the unit) shall be incl ed in the Annual 'a gical Environmental Operating Report, f

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l l LA SALLE - UNIT 2 3/4 12-10

INSTRUMENTATION I

( BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to atonitor and assess important variables following an accident. This capability is con-sistent with the recommendations of Regulatory Guide 1.97. " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

3/4.3.7.6 SOURCE RANGE MONITORS j

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

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

( The specification allows use of substituted TIP data from symmetric channels if the control rod pattern is symmetric since the TIP data is adjusted by the plant computer to remove machine dependent and power level dependent bias. The source of data for the substitution may also be a 3-dimensional BWR core sies?ator calculated data set which is normalized to available real data. Sinc: uncertainty could be introduced by the simulation and normalization process, an evaluation of the specific control rod pattern and core operating state must be performed to ensure that adequate margin to core operating limits is maintained.

3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

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

I 3/3.3.7.10 RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to

( (monitorandcontrol,asapplicable,thereleasesofradioactivematerialsin J LA SALLE - UNIT 2 8 3/4 3-5 Amendment No. 42 l

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INSTRUMENTATION DASES RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION (Continued) liquid effluents during actual or potential releases of liquid effluents. The aMrm/ trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria h} -

(60,63,and64ofAppendixAto10CFRPart50j g 1 u i 3/4.3.7.11 RADIOACTIVEGASEOUSEFFLUENT]MONITORINGINSTRUMENTATION ;

dg Theradioactivegaseouseffluentmonitoringinstrumentationisprovidedtol f monitor and control, as applicable, the releases of radioactive materials in pig gaseous effluents during actual or potential releases of gaseous effluents.

alarm / trip setpoints for these instruments shall be calculated in accordance The

( to exceeding the limits of 10 CFR Part 20.(TMi~TnsTrumentatT6n m:with .

%:he:- f roi des

-p m ai: = for mon H~oring (and controlling) the concentrations _of_potentially explosive gas mixtures in the waste gas holdup system./ The OPERABILITY and fuse of this instrumentation is consistent with the requirements of General l Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. f 3/4.3.7.12 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out- t <

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

3/4.3.8 FEEDWATER/ MAIN TURBlNE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate the feedwater system / main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 l setpoint associated with a feedwater controller f ailure to prevent over-filling the reactor vessel which may result in high pressure liquid dis-charge through the safety / relief valve discharge lines.

LA SALLE - UNIT 2 8 3/4 3-6 Amendment No.42

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3 ( 11 RADIOACTIVE EFFLUENTS

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3/4.11.1 L VI)4FFLUENTS

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3/4.11.1.1 C TION This specif1 tknisprovidedtoensurethattheconcerationof radioactive materia $asedinliquidwasteeffluentsfrmthesitewillbe less than the concen aVo levels specified in 10 CFR P 20 Appendix 8 Table II, Column 2. T s tation provides additional ssurance that the levels of radioactive m er in bodies of water out de the site will result in exposure within 1 SectionII.Adesign/objectivesofAppendixI, 10 CFR 50, to an individus and (2) the limits of 0 CFR 20.106(e) to the population. The concentrati limits for dissolve or entrained noble gases were determined by converting heir MPC's in air o an equivalent concentration in water using the methods desc bed in Interna onal Commission on Radiological Protection (ICRP) Publication 2.

3/4.11.1.2 DOSE This specification is provided to lement the requirements of Sections II.A. III.A and IV.A of Appe dix 10 CFR Part 50. The Limiting Condition for Operation implements guid s th in Section II.A of Appendix I. The ACTION statements rovide th ed operating flexibility and at the same time implement t guides set r Section IV.A of Appendix I to assure that the r eases of radioa vpMh>terial in liquid effluents will be kept "as lo as is reasonably ac ie .

Also, for fresh water sites with drinking wa r supplies which can ially affected by plant operations, there is easonable assurance that h % tion of the facility will not result radionuclide concentration ist the finished drinking water that are in exces of the requirements of 40 CFR 1 1. The ose calcula-tions in the 00CM impi ent the requirements in Section I . pendix 1 that conformance wit the guides of Appendix I be shown by 1 onal procedures based on odels and data, such that the actual ex su an individual throug appropriate pathways is unlikely to be subs n underestimated. he equations specified'in the ODCM for calcul i doses due to the actu release rates of radioactive materials in liqui uents are consisten with the methodology provided in Regulatory Guide 1. 09,

" Calculation of Annual Doses to Man from Routine Releases of Reactor ffluents for the Pur ose of Evaluating Compliance with 10 CFR Part 50, Appendi I,"

Revision ,-October 1977 and Regulatoty Guide 1.113 " Estimating Aquati Dispers n of Effluents-from Accidental and Routine Reactor Releases for he Purpos of Implementing Appendix I," April 1977.

This specification applies to the release of radioactive materials in l li uid effluents from each reactor at the site. For units with shared rad- '

ste treatment systems, the liquid effluents from the shared system are j preportioned among the units sharing that system. l l

LA SALLE - UNIT 2 B 3/4 11-1 r

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l RADIOACTIVE EFFLUENTS BASES

'3/4. I1. I LlodlD EFFLUF3JTS I

3/4.11.1.3 LIQUID WASTE TREATMENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." During extended shutdown or low power operation, i.e.,>92 days,whensteamisnotavailabletotheconcentrators,Surveillancef Requirement 4.11.1.3.2 may be extended to 180 days. This specification imple-ments the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of b ( Appendix I, 10 CFR Part 50, for liquid effluents.

f3/4.11.1.4 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified h L tanks provides assurance that in the event of ar, uncontrolled release of the f tanks' contents, the resulting concentrations would be less than the limits of (supplyandthenearestsurfacewatersupplyinanunrestrictedarea.10 CFR Part 20 3/4.11.2 G/.SE005 EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose a' any time at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive raterial discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). t For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently lcw to compensate for any increase in the 5 atmospheric dif fusion f actor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to

[ less than or equal to 500 mrem / year to the total body or to less than or

! These release rate limits also restrict, l

( equal to 3000 mrem / year to the skin.at all times, the corresponding thyroid dose l

LA SALLE - UNIT 2 B3/411-[

? DELETs Ernes Phst i

i RA 0 ACTIVE EFFLUENTS BASES

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1 DOSE RATE (Co infant via the c w -infant pathway to less than or equal o 1500 mrem /

year for the near c to the plant.

This specificat n lies to the release of radio tive effluents in gaseous effluents-fro a eactors at the site, for its within shared ra(%aste treatment syst s, the gaseous effluents f ro the shared system are

, proportioned among the u ts sharing that system.

3/4.11.2.2 DOSE - NOBLE GAS This specification is prov ed to imple ent the requirements of Sections II.B.

III.A and IV.A of Appendix I, 10 R Part . The Limiting Conditions for Operation are the guides set forth n Sec on 11.8 of Appendix I. The ACTION statements provide the required oper ti flexibility and at the same time implement the guides set forth in Sec n IV.A of Appendix I to assure that the releases of radioactive material .n aseous effluents will be kept "as low as is reasonably achievable." The rve lance Requirements implement-the requirements in Section III.A of pendix t .. onformance with the guides of Appendix I be shown by calcul ional pro s based on models and data such that the actual exposure o an individua tbrifufh appropriate pathways is unlikely to be substantially derestimated, e Ar6 e calculations established in the 00CM for calculating le doses due to th a release rates of radioactive noble gases in aseous effluents are tw d lent with the methodology provided in Regulatory Gu e 1'109, " Calculation o Anf41 Doses to Man from Routine Releases of Rea or Effluents for the Purpos c' luating Compliance with 10 CFR Part 50, A 'endix 1," Revision 1, October d Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transp r and Dispersion of Gaseous Effluents i Routine Releases from Light-Water C led Reactors,"

Revision 1, July 7. The ODCM equations.provided for d. erminin the air doses at the sit boundary are based upon the historical av mospheric conditions.

3/4.11.2.3 DOSE - RADI0 IODINES, RADIOACTIVE MATERIALS IN PARTIC ATM AND RADIO)f0CLIDES OTHER THAN N0BLE GASES N '

T specification is provided to implement the requirements of 5 tions II.C.

III. and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions or Ope tion are the guides set forth in Section II.C of Appendix 1. The A TION s tements provide the required operating flexibility and at the same tim j piement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of LA SALLE - UNIT 2 B 3/4 11-3 l

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PADI0 ACTIVE EFFLUENTS B^SES DOSE-RADIOI0 DINES, RADIOACTIVE MATERI ALS IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES (Continued)

Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The 00CM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Reguiatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Ptrpose of Evaluating Corrpliance with 10 CFR iart 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atnospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determin-ing the actual doses based upon the historical average atmospheric conditions.

The release rate specifications for radiciodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area, The pathways which were examined in the development of these calculations were: 1) individual t inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vagetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consump-tion of the milk and meat by man, and 4) deposition on the ground with subtequent exposure of man.

3/4.11.2.4 AND 3/4.11.2.5 GASEOUS RADWASTE TREATMENT SYSTEM AND VENTILATION IXHAUST TREATMENT SYSTEM The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when spe::ified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievaale." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B d II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

3/4.11.2.6 EXPLOSIVE GAS MIXTURE h The specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in g3 conformance with the requirements of General Design Criterion 60 of Appendix A f4 to 10 CFR Part 50. 4d bM pset$

LA SALLE - UNIT 2 B3/411-/ A5 M

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l RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.7 MAIN CONDENSER {N t L 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 J 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.

I 3/4.11.2.8 VENTING OR PURGING This specification provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas.

3/4.11.3 SOLID RADIOACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whe'ever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid /

solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190.

The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a member (ofthepublicfor12consecutivemonthstowithinthe40CFR190 limits. For J ResuMEP-LA SALLE UNIT 2 B 3/411-)I AS APPRcPDATE

DREre' EWr/& /A6E RADIOACTIVE EFFLUENTS BAS

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TOTAL 005 (Cognued) the purposes f pecial Report, it may be assumed that he dose commitinent to the member i public from other uranium fuel cycle ources is negligible, with the except1 n dose contributions from other n ear fuel cycle facilities at the a 'te or within a radius of 5 mi s must be considered.

If the dose to any embe f the public is estimated o exceed the require-ments of 40 CFR 190, h cial Report with a reou ,t for 3 wr ence (provided the release on s resulting in viola ion of 10 4 19ft have not already been corrected), ordance with the p ovisions of 40 _ ' 190.11, is considered to be a ti ly request and fulfil 2 the requiremer.ts cf 40 CFR 190 until NRC staf f action is completed. An i dividual is nat crasidered a member of the public during ny period in wh h he/she is -n enad i, a cyings out any operation whicn is pa of the nucl ar fuel cycle.

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LA SALLE - UNIT 2 8 3/4 11-6

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3/4. 2 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES x

3/4.12.1 MON OR ROGRAM The radiolo 'c toring program required by this s cification provides measurements of ra a nd of radioactive materials in phose exposure path-ways and for those dio des which lead to the highegt potential radiation exposures of individu ting from the station ope tion. This monitoring program thereby supple n e radiological effluent onitoring program by verifying that the measu ab e concentrations of radi ctive materials and levels of radiation are not high than expected on the b is of the effluent measure-ments and modeling of the e vironmental exposure thways. The initially speci-fied monitoring program will e effective for at east the first 3 years of com-mercial operation, as defined the ODCM.

The detection capabilities t. uired by able 4.12-1 are state-of-the-art for routine environmental measurem s in dustrial laboratories. It should be recognized that the LLD is defin as n "a priori" (before the fact) limit representing the capability of a meas ent system and not as "a posteriori" (after the fact) limit for a particul measurement. Analyses shall be performed in such a manner that the at LLDs will be achieved under routine tenditions. Occasionally backgroun fluc%ations,unavoidablysmallsample sizes, the presence of inter 0.rin nuclides, or r uncontrollable circum-stances may render these LLDs un hievable.kn cases, the contributing factors will be identified and escribed in th 1 Radiological Environmental Operating Report.

3/4.12.2 LAND USE CENSUS This spe:ificatio is provided to ensure that cha unrestricted areas ar identified and that modifications to the monitoring

@@intheuseof es program are made if ' required by the results of this censu The best survey information from t e door-to-door survey, from aerial surve m consulting with local agric tural authorities shall be used. This cen s sfies the requirements of ection IV.B.3 of Appendix I'to 10 CFR Part 5.

3/4.12.3 I ERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparis n Program is pro ded to ensure that independent checks on the precision and act acy of the m asurements of radioactive material in environmental sample matrice are perf rmed as part of the quality assurance program for environmental moni > ring in order to demonstrate that the results are reasonably valid.

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WRI'TU,T!DN CONTROL 5 t:" 00E0:'!N3 to0CEDURES AND PROGRAMS (Continued)

F.

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

1. Primary Coolant Sources Outside Primary Containment

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

The systems include LPC5, HPCs, RHR/LPCI, RCIC, hydrogen recombiner, process systems.

sampling, containment monitoring, and standby gas treatment The program shall include the following:

a.

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

Integrated cycle leak or intervals testless.

requirements for each system at refueling

2. Ir-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas unde'r accident conditions. This program shall include the following:

a. Training of personnel, -

b. Procedures for monitoring, and c.

. Provisions for maintenance of sampling and analysis equipment.

3. Post-accident Samoling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous ef fluents, and containment atmosphere samples under accident conditions. The program shall include the following:

T 'D' a. Training of personnel,

b.

l Procedures for sampling and analysis, c.

Provisions for maintenance of sampling and analysis equipment.

/

6. 3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN PLANT OPERATIO The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified and a Licensee Event Report submitted pursuant to tne requirements of Section 50.73 to 10 CFR Part 50, and b.

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

LA SALLE - UNIT 2 j2ENuMPE f' 6- [ g g ' Amendment No. 47 M P W @hTE

INSERT D

4. RadinttiyLEf.flunLControls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCH,

b. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B. Table II, Column 2,

c. 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 00CM,

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

e. 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 00CM at least every 31 days,

! f. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the

, doses a:sociated with 10 CFR Part 20. Appendix B, Table II, Column 1, l

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

1. Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate fym with half-lives greater than 8 days in gaseuw effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, ZNLD/862/19

l INSERT D (continued)

j. Limitations on venting and purging of the containment through the Primary Containment Vent and Purge System or Standby Gas Treatment System to maintain releases as low as reasonably achievable,

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

5. Radiological Environmental Moaltoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be concained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the fol'owing:

a. Monitoring, sampling, ar.alysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

b. A Land Use Cansus 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

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

ZNLD/862-20

l ADMINISTRATION CONTROLS 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT 15 EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately oursuant to Specification 2.1.1, 2.1.2 and 2.1.3, and critical reactor operation

. hall not be resumed until authorized by the NRC. The conditions of shutdown iall be promptly reported to the Vice President BWR Operations or his designated

' ternate. The incident shall be reviewed pursuant to Specifications 6.1.G.I.a ad 6.1.G.2.a and a separate Licensee Event Report for each occurrence shall be prepared in accordance with Section 50.73 to 10 CFR Part 50. The NRC Operations l Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President BWR Operations and the Manager of Off-site Review and Investigative Function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6.5 PLANT OPERATING RECORDS A.

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

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

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

3. Records and reports of reportable events;

4. Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (see Section 4 of these specifications) are being met. All equipment failing to meet surveil-lance requirements and the corrective action taken shall be recorded; l 5. Records of changes to operating procedures;

6. Shift engineers' logs; and

7. Byproduct material inventory records and source leak test results.

gENdMME-LA SALLE - UNIT 2 Amendment No. 47 6-)#

l l

ADMINISTRATION CONTROLS PLANT OPERATING RECORDS (Continued)

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

1.

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

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

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

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

5. Records of plant radiation and contamination surveys;

6. Records of offsite environmental monitoring surveys;

7. Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant, in accordance with 10 CFR '

Part 20;

8. Records of radioactivity in liquid'and gaseo.us wastes released to the environment;

9. Records of transient or operational cycling for those components that have been designed to operate safety for a limited number of transient or operational cycles (identified in Table 5.7.1-1); ,

10. Records of individual staff members indicating qualifications, '

experience, training, and retraining;

11. Inservice inspections of the reactor coolent system;

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

13. Records of reactor tests and experin+nts:

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

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

16. Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.9 including the date at which the service life commences and associated installation and maintenance records; and

17. Records-of analyses required by the radiological environmental monitoring progra _

' g c b es muk % -h OFFSITC h'3 cAltNIcdcr 4ce,mMhHUAL cl -forwJ M CESS Ccm RCL f90GRMb.

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LA SALLE - UNIT 2 Amendment No. 29 6-[ RE/*1BE R PA6e A5 APpgo pgt ATE l l

O : F W 10N CONTROLS t i REPORIIN3 REOUIREMENTS in adcition to the applicable reporting requirements of Title 10 , Code of Feceral Regulations, the following identified reports shall be submitted tc the director of the appropriate Regional Office of Inspection and Enforce-ment unless otherwise noted.

A. Routine Reports

1. Startup Report A summary report of plant startup and power escalation testing soall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comnariscs of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial pcwer opration, or (3) 9 months following initial criticality, whichever is earliest.

If the startup report does not cover all three events (i.e.,

initial criticality completion of startup test program, and resumption or commen, cement of commercial power operation), supple-mentary reports shall be submitted at least every 3 months until all three events have been completed.

2. Annual Report A tabulation shall be submitted on an annual basis prior to March 1 of each year of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions (Note: this tabulation su Section 20.407 of 10 CFR 20), e.g., pplements the requirements reactor operations of and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions, g

LA SALLE - UNIT 2 6-J/ h rA Amendment No. 47 w

\

AP"4IST W IOh CONTROLS Annual Reoort (continued)

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5 shall be included in the Annual Report along with the following in'ormation: (1) Reac-tor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of tha last isotopic analy-sis for radiciodine performed prior to exceedirg the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope oncen-tration in microcuries per gram as a function of time for the dura-tion of the specific activity above the steady state level; and (5) The time duration when the specific activity of the p'rimary coolant exceeded the radiciodine limit.

,ggg7'E' ,

1 LA SALLE - UNIT 2 6-J,M(f f [pf>W P

4 Amendment No. 47

INSERT E l

i l

3. ABRUALBAD10LOGLCaLEtWIR0f!MENIALDEER&IING REEORI*

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 ODCH and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

4. SIBIAt1NUALBAD.10&CIIVLEfLLU ENLREL E ASLREEORI *

The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January I and July 1 of each year.

The report shall include a summary of the quantitles 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 ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Seetton IV.B.1 of Appendlx I to 10 CFR Part 50.

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

A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station.; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

ZNLD/862 '1

$CLETE Ep7tW tM Gt=-

(':':STRIT10N CONTROLS Annual Environmental Radiological Operatino ReportM a.

Routine environmental radiological operating reportp covering the operation of the unit during the previous calendar year 11 be submitted prior to May 1 of each year, jfhe initial i t shall be' submitted prior to May 1 of the/ year following 1 criticality.

b. Th inc -

al environmental radiological operp ing reports shall mmaries, interpretations, and an analysis of trends.

of th r its of the radiological envir nmental surveillance activi the report period, incl ding a comparison with preopera studies, as appropria , operational controls, as approp ate, and previous enviro ental surveillance reports and an asse sment of the observed mpacts of the plant operation on the envir ment. The reports hall also include the results of lano use ce suses required Specification 3,12.2.

The annual enviro ental ra ological operating reports shall include summarized nd tub ated results in the format of Regulatory Guide _4.8, De mber 1975,-of.all radiological-environ-mental samples taken ng the report period. In the event that some results~are available for inclusion with the report, the report s 11 e submitted noting and explaining-the reasons for the mis ing re 1 The missing data shall be submitted as soon s possib supplementary report.

The-reports sh I also include behlowing:

- a summary description o the environmental actioJgical 1 monitoring pro; ram; a map of al sampling locations ke &# U a table giving distances and direct ^ons.from one reactor; an i suits of licenset--

particip ion in the Interlaboratory a ison Program, required by Spec' ication-3.12.3.

The port shall include.an annual summar o da rly meteorological collected ov_er the previous year. Th1 al summary may b either in the form of an hour-by-hour lis wind speed, ind direction, and atmospheric stability, an itation (if measured) on magnetic tape, or in the form frequency distributions of wind speed, wind direc 1 n, and atmospheric stability. This same report shall inc de an assessment of the radiation-doses due to the radioac ive liquid and gaseous effluents released-from the unit or stati during the previous calendar year. The assessment of radiatio doses l

shall be performed in accordance with the OFFSITE 005E CALCULATION MANUAL (00CM).

The report shall also include an assessment of radiation dose to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including l LA SALLE - UNIT 2 6-22 Amendment No. 47

OEl W$ ggf(f(" (865

& :$T sG'T10N CONTROLS e

4~ - '

4 nmental Radiotogical Operating Report (Continued) s2s from primary effluent pathways and direct radi ion) for t.

40 L revious 12 consecutive months to show conform ce with 190, Environmental Radiation Protection St dards for

'uc ep Power Operation. The assessment of radi tion doses s

ll\be erformed in accordance with the ODCM

4. Semiannua Ra iotctive Effluent Release ReportE D

a. Routine d' ive effluent release re rts covering the operation f he unit during the previ s 6 months of operation shall be su ,itted within 60 days af r January 1 and July 1 of each year. e period of the first eport shall begin with the date of initia criticality,

b. The radioactive fluent releas reports shall include a summary of the quantitles f radioacti liquid and gaseous effluents and solid waste rel sed fro the unit as outlined in Regulatory Guide 1.21, "Measurin., Eva ating and Reporting Radioactivity in Solid Wastes and Re as s of Radioactive Materials in Liquid and Gaseous Effluents fr Light-Water- Cooled Nuclear Power Plants," Revision 1, J e 974, with data summarized on a quarterly basis follo ng t e form t of Appendix B thereof.

1 Y

A l

l ke U Asingle submitt may be made for a multiple unit station should combine ubmittal ose sections that are common to all units tation; however, for u its with separate radwaste systems, the submit a all specify the releases f radioactive material from each unit.

1 LA SALLE - UNIT 2 6-23 Amendment No. 47 i

ADMINISTRATION CONTROLS I

Semiannual Radioactive Effluent Release Report (Continued) l The radioactive effluent release report shall include the following '

information for each type of solid waste shipped offsite during the ,

report period:

a. Container volume, i

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal rt.dionuclides (specify whether determined by l measurement or estimate),

d.

Type of waste evaporator (e.g).,

bottoms , spent resin, compacted dry waste,

e. Type of contMner (e.g., LSA, Type A, Type B. Large Quantity),

and

f. Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

d 5. Monthly Operating Report h Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, h ( shall be submitted on a monthly basis to the Director, Office of Nuclear Reactor Regulation, Mail Station PI-137 US Nuclear Regulatory Comission. Washington, DC 20555, with a copy of the appropriate Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report.

Any changes to the 0FFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

6. CORE OPERATING LIMITS REPORT

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

REtJW86E-LA SALLE - UNIT'2 6-[eure A5 Amendment No. 54

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l LA SALLE - UNIT 2 6-26 Amendment No. 47

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LA SALLE - UNIT 2 6-27 Amendment No. 47

l ADMINISTRATION CONTROLS l C. Unique Reporting Requirements I i

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

6.7 PROCESS CONTROL PROGRAM (PCP)* l 6.7.] The PCP shall be approved by the Commission prior to implementation.

6.7.2 Licensee-initiated changes to the PCP:

)

r a.

Shall be submitted to the Commission in the semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:

1. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental I information; l l .

2. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and 3.

6d Documentation of the fact that the change has been reviewed and

( foundacceptablebytheOnsiteReviewandInvestigativefunction.)

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

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

i kwum 6ER LA SALLE - UNIT 2 6-[A5 AtPtt/nATEA mendment No. 47

AD1*It.I57 RATION CONTROLS f.e OrrSITE DOSE CALCULATION MANUAL (ODCM)*

6. 8.1 The ODCl1 shall be approved by the Commission prior to implementation.

6 E.2 Licensee-initiated changes to the ODCM:

I

a. ~ Shall be submitted to the Commission within 90 days of the date the 1 change (s) was made effective.

This submittal shall contain:

1.

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

. 2.

determination that the change will nct reduce the accuracy or g6 eliability of dose calculations or setpoint determinations; and

\V 3.

l Documentation of the fact that the chaoge has been reviewed and found acceptable by the Onsite Review and Investigative Function.

b

(.

ShallbecomeeffectiveuponreviewandacceptancebytheOnsite and Investigative Function. .

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

a.

Shall be reported to the Commission in the Monthly Operating Report i for the period in which the evaluation was reviewed by the Onsite Review and Investigative Function. The discussion of each change shall contain:

1.

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

! 2.

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

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; "The and La OFFSITE Salle Unit DOSE2. CALCULATION MANUAL (ODCM) is common to La Salle Unit g

LA SALLE - UNIT 2 6-[ g Amendment No. 47

INSERT F

a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.8.18. This documentation shall contain:

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

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

INSERT G

a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.S.B.18. This documentation shall contain:

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

2) A determination that the change will maintain the level of 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 accuracy or reliability of effluent, dese, or setpoint calculations.

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

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrer't with the Semiannual Radioactive Effluent Release Raport for the period of the report in which any change to tha ^^'"i was 3ade effective. Each change shall be Identified by markings in the margin of the affected pages, clearly indicatir, the area of the page tnat was changed, and shall indicate the da'< (e.g., month / year) the change was implemented.

ZHLD/862/22

.,. . _ _ __ _ _ _ _ _ __ .. . _ ~ _ _ _ _ . . . _ _

AD"IN!5TRDION CONTROLS t%J0E CHANCE 5 TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Continued) 4.

An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; C,

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

An estimate of the exposure to plant operating personnel as a result of the change; and E. Documentation of the fact that the change was reyiewed and found acceptable by the Onsite Review and Investigative Function,

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

l l

l I

h Q~

LA SALLE - UNIT 2 6- g'((b 4 Amendment No. 47

,3, ><

ATTACHMENf C EVALUATION Of SIGNIFICANf HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations.

According to 10 CFR 50.92 (c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different Lind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because:

The proposed changes only alter the format and location of procedural details and administrative controls of the radioactive effluents, radiological environmental monitoring and solid radioactive waste programs. The changes are administrative in nature and do not involve any change to the configuration or operation of plant equipment.

The Radiological Effluent Technical Specifications (RETS) procedural details are being moved to the Offsite Dose Calculation Manual (ODCM) or the Process Control Program (PCP). However, the contents of RETS are not being changed. The purpose of RETS is to assure that the proper controls for the radiological effluent system are in place. The programs containing the procedural details of RETS will continue to perform the same functions. Any future changes to the programs containing RETS' procedural details would require a 10 CFR 50.59 safety evaluation.

Therefore, RETS will continue to receive a high level of consideration even though they will no longer be part of the Technical Specifications.

All the current provisions of RETS are being retained. Therefore, all assumptions applicable to RETS used in the accident analysis remain valid. Based on this evaluation, Commonwealth Edison has determined that the changes do not affect the probability or consequentes of a previously evaluated accident.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because:

The procedural requirements of RETS will be maintained in the ODCM or PCP. Operation of the plant will not be altered by the changes proposed to the administration of RETS. This change will not place the plant in any new condition or introduce any mode of operation not previously analyzed. Therefore, this char.ge will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve a significant reduction in a margin of safety because:

ZNLD/862-23

2-The proposed changes relocate the procedural details and bases of RETS from the Technical Specifications to the ODCM or PCP. The RETS procedural details and bases will be maintained by these programs.

Changes to these programs will require a safety evaluation per 10 CFR 50.59 and an On-Site Review. In addition, new administrative controls have been added to the Technical Specifications which assure the proper control and maintenance of these documents and provide an equivalent level of assurance that activities involving radio Stive effluents, solid radioactive waste, anr1 radiological environmental monitoring are conducted in full compliance with regulatory requirements. Therefore, there is no reduction in the margin of safety.

CONCLUSION Guidance has been provided in 51 FR 7744 for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are not likely considered to involve significant hazards considerations.

This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations. This amendment can be classified as a purely administritive change to the Technical Specifications. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92 (e), the proposed change does not constitute a significant hazards consideration, i

i ZNLD/862/24

ATTACllHENT D LNVIRONHLNIAL ASSISSHLNT STAILHENT APPLICABill1Y REVILH Commonwealth Edison has evaluated the proposed amendment a;ainst 4 the criteria for the identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.20. It has been determined that the proposed changes meet the crl'eria for a categorical exclusion as provided under 10 CFR 51.22 (c)/t, and 10 CFR i 51.22 (c)(10). This conclusion has been determined for the following l reasons: :

(1) The amendment involves no significant hazards consideration. As denonstrated in Attachment C, this proposed amendment does not :

involve any significant hazards considerations.

(11) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

This amendment is administrative in nature and only involves the format and location of procedural details anu administrative controls related to radioactive effluents. No changes to the actual methods of controlling and monitoring radioactive effluents are made by this action. Continued compliance with egulatory requirements relative to radioactive effluents is assured by specific adminis'rative controls which are added to the Technical Specifications. Therefore, there will be no change in types or increase in the amounts of any effluents released offsite.

(111) There is no significant increase In Individual or cumulative occupational radiation exposure. This proposed change will not result in changes in the operation or configuration of the facility. In addition, there will be no change in the level of controls or methodology used for processing of radioactive effluents, handling of solid radioactive waste, or radiological environmental monitoring. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

The proposed changes for the RETS procedural detail relocation and redefining the ODCM and PCP are provided for in GL 89-01. They meet the categorical exclusion criteria as provided under 10 CFR 51.22(c)(ii)

(changes recordkeeping, reporting, or administrative procedures or requirements). The proposed change relocates the RETS to the ODCM or PCP. However, all applicable surveillances, limitations, and regulatory conditions are maintained.

This request does not involve a significant increase in Individual or cumulative occupational radiation exposure. Therefore, the Environmental Assessment Statement is not applicable for these changes.

ZNLD/862-25

ATTACHMENT E LA SALLE REVISION o.A APRIL 1993 FOR INFORMATION ON,.Y This page reserved for GENERIC LASALLE ANNEX INDEX CHAPTER 12.0 j-All pages in Chapter 12.0 are designated REVIS10il 0.A i

SECIALN01E ,

Until removal of the Radiological Effluent Technical Specifications has been 1 approved by the Nuclear Regulatory Commission, the requirements of the Technical Specifications shall take precedence over this chapter, should any-differences occur.

12-1

LA SALLE- REVISION 0.A CHAPTER 12 RADI0 ACTIVE EFFLUENT TECHNICAL STANDARDS (RETS)-

TABLE OF CONTENTS 12.1 DEFINITIONS 12-1 12.2 INSTRUMENTATION 12-5 12.2.1.A/B- Radioactive Liquid Effluent !

Monitoring Instrumentation 12-5 !~

12.2.2.A/B Radioactive Gaseous-Effluent Monitoring Instrumentation 12-10 12.2.C Liquid and Gaseous Instrumentation Bases 12-16 12.3 LIQUID EFFLUENT 5 12-17 12.3.~ 1. A/B Concentration 12-17 12.3.2.A/B Dose 12-22 ,

12.3.3.A/B Liquid Haste Treatment System 12-23 i 12.3,C Liquid Effluents Bases 12-24 l

12.4 GASE0US EFFLUENTS 12-26 12.4.1 A/B Dose Rete 12-26 12.4.2.A/B Oose - Noble Gases 12-30 12.4.3.A/B Dose - Iodine-131, Iodine-133, Tritium, !-

l

__ and Radionuclides in Particulate Form 12-31 j 12.4.4.A/B- Gaseous Radwaste Treatment System 12-32 12.4.5.A/B. Ventilation Exhaust Treatment System 12-33

-12.4.6.A/B- Venting or Purging 12-34 12.4.7.A/B- Total Dose 12 12.4.C Gaseous Effluents Bases 12 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING 12-39 12.5.1.A Monitoring Program 12-39 u 12.5.1.B Honitoring Program 12-40 12.5.2.A/B Land Use Census 12-47 l 12.5.3.A/B Interlaboratory Comparison Program 12 l 12.5.C Radiologi-cal Environmental i Monitoring Program Bases 12-49 12.6- REPORTING REQUIREMENTS 12-50

-12.5.1 Annual Environmental Radiological Operating Report 12-50 12.6.2 Semiannual Radioactive Effluent Release Report 12-51 12.6.3 Offsite Dose Caculation Manual (00CM) 12-53 12.6.4 Major Changes to Radioactive Haste Treatment Systems 12-54 l

12-11

t LA SALLE REVISION 0.A APRIL 1991

0R lNFOR R OL ON 3 CliAPTER 12 RADIDACTIVE ErrLUENT TECHNICAL STANDARDS (RETS)

LIST Of TABLES 12.1-1 Surveillance frequency Notation 12-4 12.2.1-1 kadioactive Liquid Effluent Monitoring Instrumentation 12-6 12.2.1-2 Radioattive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 12-8 12.2.2-3 Ratioactive Gaseous Effluent Monitoring Instrumentation 12-11 12.2.2-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 1*-13 12.3.1-1 Maximum Permissible Concentration of Dissolved or Entrained Noble GPses Release from the Site 12-18 to Unrestricted Areas in Liquid Waste 3.1-2 Radioactive Liquid Waste Sampling and Analysis Program 12-19 12.4.1-1 Radioactive Gaseous Waste Sampling and Analysis Program 12-27 12.5.1-1 itadiological Environmental Honitoring Program 12-41 1

12.5.1-2 Reporting Levels for Radioactivity Concentrations '

in Environmental Sampling 12-44 12.5.1-3 Maximum Values for the Lower Limits of Detection (LLD) 12-45 l

12-111

REVISIOli 0.A

'^'^'" ^ >>a n ""

FOR INFORMATION ONLY 12.1 DIfidill0NS 12.1.1 ACIl0N - ACTION shall be that part of a requirement which prescribes remedial measures required under designated conditions.

12.1.2 CHANNELCAL.1EA110N - A CHANNEL CALIDRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompsss the entire channel including the alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

12.1.3 CHAMELCHICK - A CHANNEL CHECK shall be the qualitative

. assessment of channel behavior during operation by observation.

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

12.1.4 CBANNELEUNCIl0NALIISI - A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practical to verify OPERABILITY including alarm and/or trip functions and channel failure trips,

b. B1 stable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is tested.

12.1.5 DOSE _10VLVALENT I-131 - DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14B44,

" Calculation of Distance factors for Power and Test Reactor Sites."

12.1.6 GASIOUS_MQHASIE TREATHENT SYSl[8 - A GASEOUS RADHASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing total radioactivity prior to release to the environment.

12-1

LA SALLE REVISION 0.A

- APR?L 1991 '

FO R IU D R M E, j u 12.1.7 ELMER(S) Of THE _ PUEllC - HEMBER(S) Of THE PUBLIC shall include all persons who are not occupationally associated with tne plant. This category does not include employees of the licensee, its contractors, or vendors. Also, excluded from this category are persons who enter the site to service equipment or to make deliveries. This rategory does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

12.1.8 Q.PELABLE - QEBARill1Y - 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, 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).

12.1.9 EROC[iS_CONTRQL PROGRAM - The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes t'ased on demonstrated processing of actual or simulated wet solid wastes shall be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground r' 4uirements, ano other requirements governing the disposal of solid radioactive waste.

12.1.10 EURGE - PURGLNS - PURGE or PURGING shall be 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.

12.1.11 R&TED THERHAL P0H G - RATED THERHAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3323 HWT.

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

12.1.13 S.0LIDIFICATION - SOLIDIflCATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

12-2

LA SALLE REVISION 0.A APRIL 1991 FOR INFORMAT DN Om-i 12.1.14 SERCLCHICK - A SOURCE CHECK shall be the qualitative

! assessment of channel response when the channel sensor is exposed to a radioactive source.

12.1.15 IBIPPdALEQHER - THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

12.1.16 Y1HIILATION EGAUST TBE&lt!EHLSYSJIB - A VEhTILATION EXHAUST TREATHENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in ef fluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust system prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atomspheric cleanup systems are not considered to be VENTILATION EXHAUST TREATHENT SYSTEM components.

12.1.17 VENTING - VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressura, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

4 12.1.18 Definitions Peculiar to Estimating Dose to Hembers of the Public Using the 00CH Computer Program,

a. ACTUAL - ACTUAL refers to using known release data to project the dose to members of the public for the previous j time period. This data is stored in the database and used to demonstrate compliance with the reporting requirements to Chapter 12.

b.- PROJECTED - PROJECTED refers to using known release data from the previous time period or estimated release data to forecast a future dose to members of the public. This data is not incorporated into the database.

l 12-3 J

LA SALLE nryIsIoN o.A

^"" '"'

FOR INFORMAT DN ON Y LADLE _. lad-l SUMEILLANCElRLOUENCLNQIA110N NQIAll0N ERE0VENCY 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 .

H At least once per 7 days.

H 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 days.

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

S/U Prior to each reactor startup.

P Prior to each radioactive release.

N.A. Not applicable.

12-4

m i LA SALLE REVISION 0.A FOR INFORMAuBN ONLY * = 1ee1 F

12.2 INSTR E NTATION RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION

OPERABILITY REQUIREMENTS 12.2.1.A The radioactive liquid-effluent monitoring instrumentation channels shown in-Table 12.2.1 1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Section 12.3.1.A are not exceeded. The alarm trip setpoints of these channels shall be determined in accordance with the ODCH 10.

6PEllCABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required, immediately suspend the release of radioactive 11guld effluents monitored by the affected channel or declare the channel 1

-_ inoperabl e,

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION l shown in Table 12.2.1 1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time spectfied.

SURVEILLANCE REQUIREHENTS l

l 12.2.1.B_Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK,_ CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 12.2.1 2.

12-5

i i

i-i' LA SALLE REVISION 0.A APRIL 1991 TABLE 12.2.1-1

RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION 4

71 l CD I MINIMUM N

! CHANNELS _

i INSTRUMENT OPERABLE ACTION z m

i 1. GAMMA SCINTILLATION MONITOR PROVIDING ALARM AND AUTOMATIC C3 TERMINATION OF RELEASE N I E l a. Liquid 'wwaste Effluent Line l' 100 M

2. GA M A SCINTILLATION MONITORS PROVIDING ALARM BUT NOT PROVIDING C j AUTOMATIC TERMINATION OF RELEASE E

a. Service Water System Effluent Line (Unit 1) 1 101 O b.

c.

RHR Service Mater (Line A) Effluent Line-RHR Service Mater (Line B) Effluent Line 1

1 101 101

d. Service Water System Effluent Line (Unit 2) 1 101

3. FLCH RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 102

b. River Discharge - Blowdown Pipe 1 102 i

i i

I l

i

!. 12-6 i

i

LA SALLE REVISION 0.A ;

^*"'

FOR INFORMATION ONLY INSTRU!iENTATION TABLE 12.2.1-1 (Continued)

IMLLNOTATION ACTION 100 - Hith the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 14 days provided that prior to initiating a release:

a. At-least two independent samples are analyzed in accordance with Section 12.3.1.B.3, and

b. At least two technically' qualified members of the facility Staff independently verify the release rate calculations and discharge line valving:

4 .

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 101 - With the number of channels OPERABLE less than required by -

the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days

- provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are ,

collected and analyzed at a limit of detection of at least 10-7 microcurie /ml or gamma spectrometric analysis.

ACTION 102 - Hith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves for-Instrument 3a, or for known valve positions for Instrument 3b, may be used to estimate flow.

l l 1 l

12-7 l;

s LA SALLE REVISION 0.A-l APRIL 1991 l I

l TABLE 12.2.1-2 !

j RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS I CHANNEL !

!' CHANNEL SOURCE FUNCTIONAL CHANNEL I INSTRUMENT _ CHECK CHECK TEST CALIBRATION g I

! 1. GAMMA SCINTILLATION MONITOR PROVIDING ALARM -

! AND AUTOMATIC TERMINATION OF RFLEASE -

l 2 l

! a. Liquid Radwaste Effluents Line D P Q(1) R(3) T g :

2. 2 l

GAMMA SCINTILLATION MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION g l p i 0F RELEASE w l i

a. Service Water System Effluent Line (Unit 1)

D M Q(2) R(3) 3 g

b. RHR Service Mater (Line A) Effluent Line D H Q(2) R(3) i c. RHR Service Mater (Line 8) Effluent Line D M Q(2) R(3) o

- d. Service Water System Effluent Line (Unit 2) D M Q(2) R(3) a ;

{ 3. FLON RATE MEASUREMENT DEVICES <

j a. Liquid Radwaste Effluent Line D(4) N.A. Q R

b. River Discharge - Blowdown Pipe' 0(4) N.A. Q R l

)

4 ,

i i i

2 12-8 l

I 4 :

LA SALLE REVISION 0.A

^ ~ ' " '

FOR INFORMATION ONLY ,

1 INSTRUMENTATIOR TABLE 12.2.1_2 (Continued)

TABLE NOTATION (1) The CHANNEL FUNCTIONAL. TEST shall also demonstrate that automatic isolation of this pathway and control alarm annunciation occurs if any of the following conditions exist: :

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Loss of power.

3. Instrument alarms on downscale failure.

4. Instrument controls not set in Operate or High Voltage mode. >

^

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm setpoint.

2. Loss of power.

3. Instrument alarms on downscale failure.

4. Instrument controls not set in Operate or High Voltage mode. *

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference radioactive standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance act!vities with NBS. These standards shall permit calibrating the system over ti: intended range of energy and measurement range. For-subsequent CHANNEL CALICRATION, the initial reference radioactive-standards or radioactive souices that have been related to the initial calibration shall be used.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days in which continuous, periodic, or batch releases are made. -

12-9

LA SALLE RI: VISION 0.A FOR INFORVION ON3 INSTRUMENTATIQN l RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTAi!QN l OPERABILITY REQUIREHENTS 12.2.2.A The radioactive gaseous effluent monitoring instrumentation channels I shown in Table 12.2.2 1 shill be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Section 12.4.1.A are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the 00CH.

APELICABILITY: As shown in Table 12.2.2-1.

ACTION:

a. Hith a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 12.2.2-1.

SURVEILLANCE REQUIREMENTS 12,2,2.B Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION l

operations at the frequencies shown in Table 12.2.2_2. l i

l t

l l

l l 12-10 l

LA SALLE REVISION O.A -l

! APRIL 1991 -

j TABLE 12.2.2-1 l i- i

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION i

3 N

f 4

MINIMUM CHANNELS '

l INSTRUMENT OPERABLE APPLICABILITY ACTION .-

t 2 j

j

1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EFFLUENT MONITORING SYSTEM g 3

.g j

a. Noble Gas Activity Monitor - Providing i Alarm and Automatic Termination of Release 1 110 g l l

i

2. MAIN STACK MONITORING SYSTEM

a. Noble Gas Activity Monitor

h t 1 110 e ;

i b. Iodine Sampler 1 111 g ;

c. Particulate Sampler 1 111 g- - -

{ d. Effluent System Flow Rate Monitor 1 112 4

e. Sampler Flow Rate Monitsr 1 112 i ,

! 3. CONDENSER AIR EJECTOR RADI0 ACTIVITY MONITOR !

l (Prior to Input to Holdup System)

a. Noble Gas Activity Monitor 1 # 113 ,

l 4. SBGTS HONITORING SYSTEM t i i r

a. Noble Gas Activity Monitor 1 ## 110 !

b. Iodine Sampler 1 ## 111 i

c. Particulate Sampler 1 ## 111 [

! d. Effluent System Flow Rate Monitor 1 ## 112 i

e. Sampler Flow Rate Monitor 1 ## 112 l

! f I

12-11

[

i

?

'^ s ^ ' " ^

0R IN:0RMAL0\ rNLY nF;"

INSTRUMENTATIQH ,

TABLE 12.2.2-1 (Continued) l IABLE NOTAT10h At all times.

- During main condenser offgas treatment system operation.

During operation of the main condenser air ejector.

1. During operation of the SBGTS.

ACTION 110 - Hith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for noble gas gamma emitters within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 111 - Hith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the channel has been declared inoperable, samples are continuously collected with auxiliary sampling equipment as required in Table 12.4.1-1.

ACTION 112 - Hith the number of channels OPERABLE less than required by the Minimum Channels OPERACLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 113 - Hith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the output from the charcoal adsorber vessels may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

a. The offgas treatment system is not bypassed, and

b. The offgas treatment delay system noble gas activity effluent downstream monitor is OPERABLE; Otherwise, be in at least STARTUP with the main steam isolation ,

valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

{

/

12-12

IA. SALLE REVISION O.A APRIL 1991 TABLE 12.2.2-2 RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION SURVEILLR CE RE0VIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL SOURCE FUNCTIONAL CHANNEL HHICH SURVEIL-INSTRUMENT _ CHECK _ CHECK TEST CALIBRATICN LANCE REOUIRED_

1. MAIN CONDENSER OFFGAS TREATHENT SYSTEM EFFLUENT HONITORING SYSTEM

a. Noble Gas Activity Honitor - T g

Providing Alarm and Automatic ;;;g;y Termination of Release D D Q(1) R(3)

~

2. MAIN STACK HONITORING SYSTEM E

c:::s

a. Noble Gas Activity Monitor D H Q(5) R(3) *

g:p b.

c.

Iodine Sampler Particulate Sampler H N.A. N.A. N.A.

N.A.

g H N.A. N.A. E:mi

d. Effluent System Flow Rate Monitor D N.A. O R W

e. Sampler Flow Rate Monitor D N.A. O R c

3. CONDENSER AIR EJECTOR RADI0A:TIVITY MONITOR c::3

a. Noble Gas Activity Monitor D H Q(2) R(3) # 2 m

! 4. SBGTS HONITORING SYSTEM

a. Noble Gas Activity Monitor D H Q(4) R(3) ##

b. Iodine Sampler H N.A. N.A. N.A. ##

c. Particulate Sampler H N.A. N.A. N.A. ##

d. Effluent System Flow Rate Monitor D N.A. Q R ##

e. Sampler Flow Rate Monitor D N.A. Q R ##

3 12-13

LA SALLE REVISION 0.A l l

APRIL 1991 TABLE 12.2.2-2 (Continued)

TABLE NOTATIQM At all times. _

i'

During operation of the main condenser air ejector.

1. During operattor. of the SBGTS.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate the automatic isolation capability of this pathway, and that control room alarm annunciation occurs if any of the following conditions exists: (each channel will be tested independently so as not to initiate automatic isolation during operation).

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Loss of power.

3. Instrument alarms on downscale failure.

4. Instrument controls not set in Operate or High Voltage mode.

(Automatic isolation shall be demonstrated during the CHANNEL CALIBRATION).

(2) The CHANNEL FUNCTIONAL TEST for the log scale monitor shall also demonstrate that control rocm alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Loss of power.

3. Instri' ment alarms on downscale failure.

4. Instrument controls not set in Operator or High Voltage mode.

l-(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference radioactive standards certified by the National Bureau of Standards (ribs) or using standards that have.been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, the ,:

initial reference radioactive standards or radioactive sources that have !.

been related to the initial calibration shall be used. '

1 12-14

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

"'^"'

20R INFORMAT ON 0iLY MIPf"ei^

(4) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument controls not set in the Operate mode.

l l

12-15

LA SALLE REVISZoN 0.A FOR INFORMATION ONLY xean an I

INSTRUMENTATION LIOUID AND GASEOUS INSTRUMENTATION BAS [S 12.2.1.C RADIOACTIVE LIOUID EFFLUENT HONI.TORING INSTRUMENTATION The radioactive liquid effluent mmitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the 00CH to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

12.2.2.C RADI0 ACTIVE GASEOUS EFFLUENT MQNITORIN_G INSTRUMENTATI.03 The radioactive gaseous effluent monitoring instrumentation is provided to !

monitor and conttr>1, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. l:

The alarm / trip setpoints for these instruments shall be calculated in i accordance with the procedures in the ODCH to ensure that the alarm / trip will I occur prior to' exceeding the limits of 10 CFR Part 20.

i.

1 12-16 o -

1 TR POR K 0N ULY u sa'u - o^

APRIL 1991

.121LL10VID_Lfl.l.UENIS 1 CONCBIRAIION OPERABILITY REQUIREMENTS 12.3.1.A The concentration of radioactive material released from the site shall be limited to the concentrations specified in 10 CFR Part 20 Appendia B Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to the concentrations specified in Table 12.3.1-1.

APPLICABILITY: At all times.

i ACHON:

Hith the concentration of radioactive material released from the site exceeding the above limits, immediately restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 12.3.1.B.1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 12.3.1-2. The results of pre-release analyses shall be used with the calculational methods in the ODCH to assure that the concentration at the point of release is maintained within the limits of Section 12.3.1.A.

t 12.3.1.B.2 Post. release analyses of samples composited from batch releases

! shall be performed in accordance with Table 12.3.1-2. The results l

of the previous post-release analyses shall be used with the

calculational methods in the 00CH to assure that the concentrations l at the point of release were maintained within the limits of Section 12.3.1.A.

12.3.1.B.3 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 12.3.1-2. The results l of the analyses shall be used with the calculational methods in the OOCH to assure that the concentrations at the point of release are maintained within the limits of Section 12.3.1.A.

l 12 17

FOR INFORMATION ONLY

'^ $^ut inPi";^ ,

TABLE 12,3.1-1 2 ,

MAXIMU.M PERMISilBLLCONCINIMIIOJLOL DISSQLV10_QfLERIR&lh1D_ NOB LLGASES RELEASED FROM THLSITE T0 UNRE11RIE1ED_. ARIAS IN LIOUID HASTI NUCLIDE BECluClimlM Kr 85 m 2E-4 85 5E-4

87 4E-5 88 9E-5 Ar 41 7E-5 Xe 131 m 7E-4 133 m SE-4 133 6E-4 135 m 2E-4 135 2E-4

Computed from Equation 20 of ICRP Publication 2 (1959) adjusted for infinite ,

cloud submersion in water and R - 0.01 rem / week, pw - 1.0 gm/cm3, and Pw/Pt - 1.0.

12-18 ,

++-e----e.w--o-e'w-a. v- .e ,,,,-+-+.p+-m e-m-y-.+ .e,e--,-m-rwe--v-w-,-,,---r-e.,---%-----..e-------.m-*p-+e--rg.gi-.+vww-w-w-<sm ,#w 1~+vi+---we e --- e 4r-v r- re v m' * < - - - +3 -

~a----.r++===+

LA SALLE RI:VIsloN 0.A

^"" '"'

FOR INFORMATION DNLY uBu u. u -2 MQlDACTIVE LIOUID HASTE SAMPLIfiG AND ANALYSIS PROGRAM Hinimum Type of Lower Limit Liquid Release Sampling Analysis Activity of Detection Type frequency frequency Analysis (LLD)

(uC1/ml)a A. Batch Haste P P Principal Gamma 5x10-7 Release Each Batch Each Batch Emitters I Tanksd I-131 1x10-6 P H Dissolved and One Batch /H Entrained Gases 1x10-6 (Gamma emitters)

P H H-3 1x10-5 Each Batch Composite b Gross Alpha 1x10-7 P Q Sr-89 Sr-90 5x10-8 .

Each Batch Composite b Fe-55 1x10-6 B. Continuous Principal Gamma 5x10-7 Releases' ContinuousC CompositeC Emittersf I-131 1x10-5 H H Dissolved and Grab Sample Entrained Gases 1x10-5 (Gamma emitters)

H H-3 1x10-5 Continuouse CompositeC Gross Alpha 1x10-7 0 Sr-89, Sr-90 5x10-8 ContinuousC Compositec Fe-55 1x10-6 l

l 12-19 i

FOR INFORMATION ONI.Y

'^S^uc ryg,e ,

l 4

IABLE_12.3.1-2 (Continued)

IABLE301A110N

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely ;

concluding that a blank observation represents a "real" signal. ,

For 6 particular measurement system (which may include radiochemical separation):

4.66 sb LLD - E . V . 2,22x100 , Y exp (-Aat)

Where:

LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a bl&nk sample as appropriate (as counts per minute).

E is the counting efficiency (as counts per transformation),

V is the sample size (in units of mass or volume),

2.22x106 is the number of transformations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide and for composite samples, and at is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples). For batch samples taken and analyzed prior to release, at is taken to be zero.

The value of sb used in the calculation of the LLO for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverifled_ theoretically predicted variance. _ Typical values of E, V, Y, and at shall be used in the calculation,

b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the l method of sample employed results in a specimen which is representative of l the liquids released.

12-20

FOR INFORMA"DN M u situ - - o.x APRIL 1991 i

IABLL121L1-2 (Continued) :

IABLE_E01AI103

c. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release,

d. A batch relea',e is the discharge of 11guld waste of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in the ODCM, to assure representative sampling,

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous release,

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65 Mo-99, Cs-134 Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which. arc measurable and identifiable, at the 951 confidence level, together with the above nuclides, shall also be identified and reported.

I: 1 f

12-21 l- ._ _ _ . . _ . ~ . _ . . _ _ . . _ . . . _ . _ _ . . . _ - - _ _ . . . . _ . _ _ _ _ _ _ . _ . . _ _ _ _

FOR INFORMATION ONLY '^ 5^t't

=E;"a ^

I I

LIDUID EFFLUERIS ;

DQSE OPERABILITY REQUIREMENTS 12.3.2.A The dose or dose commitment to an individual from radioactive i materials in liquid effluents released, from each reactor unit, from <

the site shall be limited: t

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than to equal to 3 mrem to the total body _and to less than or equal to 10 mrem to any organ.

APPLICABILIIY: At all' times.

ACIIDS

a. Hith the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any

. other report required by LaSalle Technical Specification 6.6.A.

prepare and submit to the Commission within 30 days, pursuant to

. LaSalle' Technical Specification 6.6.C a Special Report which ;

identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the-remainder of the current 1 calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment. to an individual from these releases is within 3 mrem to the total body and 10 mrem to any organ. This Special Report shall also include the radiological l impact on finished drinking water supplies at the nearest downstream drinking water source.

SURVEILLANCE REQUIREMENTS 1

12.3.2.B Dose Calculations . Cumulative dose contributions from liquid effluents shall be determined in_accordance with the ODCH at least

once per 31 days.

[ 12-22 B

-*wv*e-e e,=-.-ve-1 w,er<.--.a,v--+v..,-.,,, ...~,--,-_.,-..,,.-e,, ~ _.%-,.,,-.m,,w,-,.m,,r-r,-,w-. .m, , a.r w- r m- m y g ww.ge.,.-,cr ,, - y -r e e

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

f 11,1991

(

L10VJD_fELUIMIS L10VID_ HASTE TREATMMLS151W OPERABILITY REQUIREMENTS

......................_________...............__...........................-_.. i 12.3.3.A The liquid radwaste treatment system shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid _ wastes prior to their discharge when .

the projected doses due to the liquid effluent from each reactor unit, from the_ site, when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.

1 APELICABIL1H: At all times.

AC110N: ;

a. With the 11guld radwaste treatment system inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by LaSa11e' Technical Specification 6.6.A, prepare and submit to the Commission within 30 days pursuant to LaSalle Technical Specification 6.6.C. a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence.

SURVEILLANCE REQUIREMENTS ,

12.3.3.B.1 Doses due to liquid releases shall be projected at least once -

per 31 days, in accordance with the ODCH. '

12.3.3.B.2 The liquid radwaste treatment _ system shall be demonstrated OPERABLE by operating the 11guld radwaste treatment system equipment for at least 30 minutes at least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.

12 23

_ _ _ _ _ m ... _ _ _ _.

SA REVISION 0.A FOR INFORMATION ONLY = ' -

I fl L10WD EFFLUEH15 L10U10.IEELUENIi3 ASIS 12.3.1.C CONCENTRATION This requirement is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be

. less than the concentration levels specified in 10 CFR Part 20. Appendix B.

Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposure within (1) the Section II.A design objectives of Appendix I,10 CFR 50, to an individual, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limits for dissolved or entrained noble gases were determined by converting their MPC's in air to an equivalent concentration in water using the methods described in International Commission "

on Radiological Protection (ICRP) Publication 2.

12 3.2.C DOSE This requirement is provided to implement the requirements of Sections II.A. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements to guides set forth in Section II.A of Appendix I.

The ACTION statements provide' the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix ! to assure that the releases of radioactive material in liquid effluents will be

kept "as low-as is reasonably achievable." Also, for fresh water sites with drinking vater supplies which car be potentially affected by plant operations, there-is reasonable assurance that the operation of the facility will not ,

result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the 00CM implement the requirements in Section III.A of Appendix I that i

conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be.substantially underestimated.

The equations specified in the ODCH for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistant with the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for_the Purpose-of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision l', October -1977 and Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

This requirement applies.to the release of radioactive materials in liquid l effluents- from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared are proportioned among the units sharing that system.

12-24

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

hh LA SALLE REVISION 0.A APRIL 1991 LIMICLEIILULHIS ,

LI MID EFFLUENIS_. BASES f

ILLJJ_LI@lD_8ASTE TREAIENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." During extended shutdown or low power operation, i.e., > 92 days, when steam is not available to the concentrators, Surveillance Requirement 12.3.3.B.2 may be extended to 180 days. Thir specification implements the requirements of 10 CFR Part 50.36a. General Design Criterion 50 of Appendix A to 10 CFR Part 50 and-the design objective given in Section 11.0 of Appendix ! to 10 CFR Part 50. The specified limi:s governing the use of appropriate portions of the liquid radwiste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents.

l l

f 12-25 l

L . - - - - - - . - - - - - - . - . - - - . - - _

LA SALLE REVISION 0.A

""" 2 " 2 FOR INFORMATON DNLY ,

i l

12.4 GASEOUS EFFLUENTS DQSE RATI OPERABILITY REQUIREMENTS I

12.4.1.A The dose rate due to radioactive materials released in gaseous effluents from the site, to areas at and beyond the SITE BOUNDARY, shall be limited to the following: j

a. For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and

b. For iodine _131, for iodine _133, for tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any nrgan via the inhalation' pathway.

APPLICABILITY: At Til times.

ACTION:

With the dose rate (s) exceeding the above limits, immediately decrease the ,

release rate to within the above limit (s).

o SURVEILLANCE REQUIREMENTS-12.4.1.B.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCH.

12.4.1.B.2 The dose rate due to iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than eight days, in accordance with the methodology and parameters of the ODCM by obtaining representative samples and performing analyses in accordance with sampling and analysis program

- specified in Table 12.4.1 1.

12 26

- - . . - - -..=_.--a-_--=-.-.---,-,._ . - _ , -

i l

, LA SALLE !

REVISEON O.A ,

T_A_BLE 12.4.1-1 APRIL 1991 g i !

i - '

_RA_D.I0 ACTIVE GASEOUS HASTE SAMPLING AND ANALYSIS PROGRAM g

l Sampling Minimum Analysis Type of Lower Limit of Detection 3

m Gaseous Release Type ' Frequency Frequency Activity Analysis (LLD) g -

(uC1/ml)a p i C i l A. Containment Vent P P Principal Gamma Emitters 9 1x10-4 3 i

and Purge System Each Purge D Each Purge D l

Grab Sample CD H-3 1x10-6 2 r--

M

8. Main Vent Stack Mb ,e gb Principal Gama Emitters 9 1x10-4 ;

Grab Sample l H-3 1x10-6 i oc r

, C. Standby Gas Grab Sample Hc Principal Gamma Emitters 9 1x10-4

Treatment System I 4

Continuousf Hd I-131 1x10-12

D. All Release Types as Charcoal -

l' listed in A and B Sample above, at the Vent I-133 1x10-10 i Stack, and as listed

, in C above, at the H" g 3

SBGTS whenever there Continucasf Particulate Principal Gamm4 Emitters 9 1x10-II j' is flow. _ _t Samole (I-131. Others)

! M i 2 Continuousf Composite Gross Alpha 1x10-II

{ Particulate !

Sample [

l Q

1 Continuousf Composite Sr-89, Sr-90 1x10-II i

Particulate

Samole

^

Continuousf Noble Gas Noble Gases 1x1C ' (Xe-133)

, Monitor ' Gross Beta & Gamma ecub alent) i

. 12-27

4 LA SALLE REvgsfoN 0.A "a" 1" '

10R INFORMATION ONLY TABLE 12.4.1-1 (Continued) ,

IABLE NOTATION

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a-particular measurement system (which may include radiochemical separation):

4.66 sb LLD - E . V . 2.22x10b . Y exp (-lat)

Where:

LLD..is the "a priori" lower limit of betection as d6 fined above (as microcurie per unit mass or volume),

sb is tne standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per transformation),

V is the sample size (in units of mass or volume),

2.22x106 is the number of transformations per minute per microcurie,

-Y is the fractional radiochemical yleid ;.inen applicable)..

A is the radioactive decay constant for the particular radionuclide, and at is tha elapsed time between midpoint of sample collection-and time of.

counting (fc plant effluents, not environmental samples).

The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance-of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than' on an unverified theoretically-predicted variance. Typical values of E, V Y. and at shall be used in the. calculation,

b. Analyses shall also be performed following shutdown, startup, or a THERMAL P0HER change exceeding 15% of the RATED THERHAL POWER within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.

12-28

LA SALLE REVISION 0.A

^~"'

FOR INFORMATION ONLY TABLE 12.4.1-1 (Continued)

TABLE NOTATION

c. Whenever there is flow through the SBGTS.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing or after removal from sampler.

Sampling shall also be performed at least once per 2A hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are. analyzed, the corresponding LLP'e may be increased by a factor of 10.

'This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

e. Tritium grab samples shall be taken at least once per 7 days from the plant vent to determine tritium releases in the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool,

f. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time pertoa i.e'ered by each dose or dose rate calculation

-made in accordance with Sections 12.4.1.A. 12.4.2.A and 12.4.3.A.

g. The principal 3mma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Hn-54, Fe-59 Co-58, Co-60, Zn-65, Ho-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are meast'rable and identifiable, at the 957. confidence level, together with the above nuclides, shall also be identified and reported.

12-29

LA SALLE REVISION 0.A

-u 1=

FOR INFORMATION ONLY GASEOUS EFFLUENTS DOSE - NOBLE GASES e

OPERABILITY REQUIREMENTS 12.4.2.A The air dose due to noble gases released in gasecut effluents, from each reactor unit, from the site shall be limited to the following:

a. During any calenaar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTION:

a. Hith the calculated 61r dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report required by LaSalle Technical Specification 6.6.A prepare and submit to the Commission within 30 days, pursuant to LaSalle Technical Specification 6.6.C. a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions 'o be taken to reduce the. releases and the proposed corrective actions to be taken to assure that subsequent releases

! will be in compliance with the above limits.

SURVEILLANCE REQUIREMENTS 12.4.2.B Dose Calculations - Cumulative dose contributions for the current l calendar quarter and current calendar year shall be determined in i accordance with the ODCH at least once per 31 days, l

f t

i 12 30

FOR INFORMATION ONLY =!P;",;^ i i

GASE0VLEflLUEMIS D01E_ ._10 DINE-131dODINE-131_.IR111UL_AND_ RAD 10NUCL10ES_IH_PARILCULAILf.0RM OPERABILITY REQUIREMENTS 12.4.3. A The dose to a HEMBER 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 each reactor unit, to areas at and beyond the SITE BOUNDARY shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,

b. During any calendar year: Less than to equal to 15 mrems to any organ, i2pklCABILITY: At all times.

AC110E: .

a. With the calculated dose from the release of todine-131, todine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licenses Event Report, prepare and submit to the Commission within 30 days, pursuant to LaSalle Technical Specification 6.6.C, a Special Report that_ identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed correction actions to be taken to assure that subsequent releases will be in compliance

.with the above limits.

SURVEILLANCE REQUIREMENTS 12.4.3.8 Cumulative dose contributions for the current calendar quarter and -

current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCH at least once per 31 days, '

12-31

l

^ ^

REVZSION 0.A

"""'"2 FOR INFORMATION ONLY GASEOUS EFFLUE415 f

GASEOUS RADHASTE TREATMENT SYSTEM OPERABILITY REQUIREMENTS 12.4.4.A The GASEOUS RADHASTE (OFF-GAS) TREATMENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTION:

a. Hith the GASEOUS RADHASTE (OFF_ GAS) TREATMENT SYSTEM inoperable for more than 7 days, in lieu of any other report required by LaSalle Technical Specification 6.6.A, prepare and submit to the Commission within 30 days, pursuant to LaSalle Technical Specification 6.6.C, a l Special Report which includes the following information: !

-1. Identification of the inoperable equipment or subsystems and the reason for inoperability.

2. Action (s) taken-to restore the inoperable equipment to OPERABLE l status, and l

3. -Summary description of action (s) taken to prevent a recurrence.

SURVEILLANCE REQUIREHENTS 12.4.4.B The GASEOUS RADWASTE TREATMENT SYSTEM shall be verified to be in operation at least once per 7 days.

12-32

-h hh h LA SALLE REVISION 0.A APRIL 1991 GASEOUS EFFLUENIS

-VENTILATION EXHAUST TREATHENT SYSTEM OPERABILITY REQUIREMENTS 12.4.5.A The appropriate portions of the VENTILATION EXHAUST TREATHENT SYSTEM shall be OPERABLE and be used to reduce radioactive materials in gaseous waste. prior to their discharge when the projected doses due to gaseous effluent releases from each reactor unit, from the site, when averaged over 31 days, would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

Hith the VENTILATION EXHAUST TREATHENT SYSTEM inoperable for more than 31 days, or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by LaSalle Technical Specification 6.6.A, prepare and submit to the Commission within 30 days, pursuant to LaSalle Technical Specification 6.6.C, a Special Report which includes the following information:

.1. Identification of the inoperable equipment or subsystems and the reason for inoperability,.

2. Action (s) taken to restore the inoperable equipment.to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence.

SURVEILLANCE REQUIREMENTS 12.4.5.8.1 Doses due to gaseous releases from the' site shall be projected- .

l

-at least once per 31 days in accordance with the ODCH.

12.4.5.8.2 The VENTILATION EXHAUST TREATHENT SYSTEM shall be demonstrated !

OPERABLE by operating the VENTILATION EXHAUST TREATHENT SYSTEM equipment for at least 30 minutes, at least once per 92 days unless.the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

12-33 1

...a..~ - . . - . , . - ,.~ ~ - -.,-,- ,.,- ..+. . - ,.- . - _ , ,..,., e , n - , . ~ . . . , ~ - - - - , , ~ , - , ,

'^$"

FOR INFORMATION ONLY =ma ^ ,

GASEQUS EFFLUENTS VENTING OR PURGlHG OPERABILITY REQUIREMENTS 12.4.6.A VENTING or PURGING of the containment drywell shall be through the Primary Containment Vent and Purge System or the Standby Gas Treatment System.

AEELICABILITY: Whenevr.. the drywell is vented or purged.

ACTION:

a. With the requirements of the above specification not satisfied, suspend all VENTING and PURGING of the drywell.

SURVEILLANCE REQUIREMENTS 12.4.6.B.1 The containment drywell shall be determined to be aligned for VENTING or PURGING through the Primary Containment Vent and Purge System or the Standby Gas Treatment System within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or PURGING of the drywell.

12.4.6.B.2- Prior to use of the Purge System through the Standby Gas Treatment System in OPERATIONAL CONDITION 1, 2 or 3 assure that:

a. Both Standby Gas-Treatment System trains are OPERABLE, and

b. Only one of the Standby Gas Treatment System trains is used for PURGING.

l l

1 1

1 12 34

'^'^"'

FOR INFORMATION ONLY =Sa ^

GASEOUS EFFLUENTS TOTAL-DOSE OPERABILITY REQUIREMENTS 12.4.7.A The dose or dose commitment to any member of the public, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less or equal to 75 mrem) over 12 consecutive months.

APPLICABILITY: At all times.

ACl10.!i:

Hith the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limit of Sections '

12.3.2.A.a. 12.3.2.A.b, 12.4.2.A.a. 12.4.2.A b, 12.4.3.A.a or 12.4.3.A.b.

in lieu of any other report required by LaSalle Technical Specification i 6.6.A. prepare and submit, pursuant to LaSalle Technical Specification :

6.6.C a Special Report to the Director, Nuclear Reactor Regulation, U.S. i Nuclear Regulatory Commission, Washington, D.C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Section 12.4.7.A. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluents pathways and direct radiation) for i I

a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dose (s) exceeds the limits of Section 12.4.7.A.

and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified information of K 190.11. Subnittal of the report is considered a timely request, and a variance is granted until staff action on the I request is complete. The variance only relates to the limits of 40 CFR L 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this technical specification.

SURVEILLANCE REQUIREMENTS l l l

12.4.7.B Dose Calculations - Cumulative dose contributions from liquid and i i gaseous effluents shall be determined in accordance with Sections 1 12.3.2.B.12.4.2.B and 12.4.3.B. and in accordance with the 00CM.

12 35 L__ , _ ,

'^ = a ~ e.x FOR INFORMATION ONLY .

APRIL 1992 l l

l MSEQUS EFFLUEMIS MSIOUS EFFLUENTS BAS.ES OPERABILITY REQUIREMENTS 12.4.1.C DOSE RATE This specification is provided to ensure that the dose at any time at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B Table II, Column 1, These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an _ individual in an unrestricted area, either within or outside the-site boundary, to annual average concentrations exceeding the limits specified in Appendix B. Table II of.10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within the site boundary, the cccupancy of the individual will be sufficiently low to compen: ate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem / year.

This specification applies to the release of radioactive effluents in gaseous effluents ~ from all reactors at the site. For units within shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

12.4.2.C DOSE - NOBLE GASES This specification is provided to inplement the requirements of Sections II.8, III.A and IV.A of Appendix I, 10 CFR Part 50. The Operability Requirements are the guides set forth in Section II.B of Appendix I. The i

ACTION statements provide the required operating flexibility and at the same l time implement the guides set forth in Stction IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." Tht Surveillance Requirements implement the requirements in Section III.A of Apperdix I that conformance with the guides of Appendix I be shown by calculati:nal procedures based on models and data such that the actual exposure of an ildividual through appropriate pathways is unlikely to be substantially utderestimated. The dose calculations established in the ODCH for c3lculating the doses due to the actual release rates of radioactive noble gises in gaseous effluents are consistent with the methodology provided in legulatory Guide 1.109, 12-36

REVISZON 0.A

'^'^"' """ 2 " 2 FOR INFORMAIl0N ONLY GASEOUS EFFLUENTS GASEOUS EFFLUENTS BASES

-OPERABILITY REQUIREMENTS DOSE RATE (Continued)

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50,-Appendix I,

" Revision 1, October 1977 and Regulatory Glide 1.111, " Methods for Estimating Atmospheric Transport and Olspersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision I, July 1977. The OOCH equations provided for determining the air doses at the site boundary are based upon the historical average atmospheric conditions.

12.4.3.C DOSE _ IODINE _131. 10 DINE.133. TRITIUM. AND RADIONUCLIDES IN PARTICULATE FORM The specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Operability Requirements are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable '- The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be'substantially underestimated. The ODCH calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided _in Regulatory Guide 1.109, Lalculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,

" Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating _

Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Hater-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

12-37

_ _ . - ._ _ ~~

hh hh LA SALLE REVISION 0.A APRIL 1991 f@ l0 ACTIVE EFFLUENTS BASES 12.4.4.C and 12.4.5.C GASEOUS RA0 HASTE TREATMENT SYSTEM AND VENTILATION EXHAUST TREATMENT SYSTEM The OPERABILITY of the GASEOUS RADHASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYS1EM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems !

be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous. effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General C' sign Criterion-60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II,0 of Appendix I to 10 CFR Part

50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.3 and 11.0 of Appendix I, 10 CFR Part 50, for gaseous effluents.

12.4.6.C VENTING OR PURGING This specification provides reasonable assurance that releases from drywell purging operations will act exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas.

12.4.7 C TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it.is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will-describe a course of action which should result in the limitation of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 l limits. For the purpose of the Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance-(provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle.

12-38

FOR INFORMATION ONLY LA SALLE REVISION 0.A APRIL 1991 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORIEG HONITORING PROGRAM ,

Il OPERABILITY REQUIREMENTS

______ ________________________________________________________________________ p 12.5.1.A The radiological environmental monitoring program shall be conducted as specified in Table 12.5.1-1.

APPLICABILITY: At all times ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 12.5.1-1, in lieu of any other report required by LaSalle Technical Specification 6.6. A. prepare and submit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels in Table 12.5.1_2 when averaged over l any calendar quarter, in lieu of any other report required by LaSalle i Technical Specification 6.6.A. prepare and suomit to the Commission !'

within 30 days from the end of the affected calendar quarter a i Special Report pursuant to LaSalle Technical Specification 6.9.1_13.

When more than one of the radionuclides in Table 12.5.1-2 are detected in the sampling medium, this report shall be submitted if: l{

i-concentration (1) + concentration (2) + .. 2.1.0 limit level (1) limit level (2)

When radionuclides other than those in Table 12.5.1-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Sections 12.3.2.A, 12.4.2.A and 12.4.3.A.

This report is not- required if the measured level of radioactivity l l

was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report,

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 12.5.1-1, in lieu of any other report required by LaSalle Technical Specification 6.6.A, prepare and submit to the Commission within 30 days, pursuant to LaSalle Technical Specification 6.6.C a Special Report which 1

12 39

' ^ " ' "REVISION

" " ' " 0'A FOR INFORMAT ON DNI.Y 12.5 RADIOLOGICAL ENVIRONMENTAL HONITORING FONITORING PROGRAM OPERABILITY REQUIREMENTS (Cont'd) identifies the cause of the unavailability of samples and iden&ifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table 12.5.1-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

SURVEILLANCE REQUIREMENTS

_____-_-____---__---___----_____-_____--____ ---_____--_-____-___ _ -===--_

12.5.1.B The radiological environmental monitoring samples shall be collected pursuant to Table 12.5.1-1 from the locations given in the table and figure in the ODCH and shall be analyzed pursuant to the requirement of Tables 12.5.1-1 and 12.5.1-3.

I 12-40

LA SALLE RDTISION Oo A APRIL 1991-TABLE 12.5.1-1 3 "

.lElF RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Qen Number of Samples M Exposure Pahtway and Sampling and Type and Frequency E and/or Samole Samole locationJ' Collection Freauency of Analyjiis .. D -

d

1. AIRBORNE C

'E

, Radiolodine and 5 Locations Continuous operation of Radioiodine canister.

I Particulates sampler with sample col- Analyze at least once C lection as required by per 7 days for I-131. 2 dust loading but at least once per 7. days. Particulate sampler.

Analyze for gross beta j radioactivity 1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

following filter change.

Perform gamma isotopic analysis on each sample when 1 gross beta activity is > 10 times the yearly mean of control samples. Perform gamma isotopic analysis on composite (by location) sample at least once per 92 3 days.

2. DIRECT RADIATION 38 Locations At least once per 31 days. Gamma dose. At least 1 2 dosimeters or 1 1 or once per 31 days.

instrument for con- or tinuously measuring At least once per 92 days. Gamma dose. At least and recording dose (Read-out frequencies are once per 92 days.

rate at each location. determined by type of '

dosimeters selected). .

' Sample locations are described in the ODCH.

12-41 ,

_ m --__s__ ___. _m _ _-- -- _ _ _ _ - ___-__-m___ __ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ _ - . _ _ . . _ _

'LA SALLE REVISION O.A APRIL 1991 g j TABLE 12.5.1-1 (Continued)

RADI3 LOGICAL ENVIRONMENTAL MONITORING PROGRAM g E

Number of Samples

=

W Exposure Pahtway and Sampling and Type and frequency

_3nd/or Sample Samole Locations' Collection Freaug.ntcy of Analysis 5

3 l 3. HATERBORNE c:3 l

2 2 locations

a. Surface Composite sample collected Gama isotopic analysis r==--

over a period of I 31 days. of each composite sample. N Tritium analysis of composite sample at least once per 92

days.

b. Ground 5 locations At least once per 92 days. Gamma isotopic and tritium analyses of each sample.

c. Sediment i location At least once per 184 days. Gamma isotopic analysis of ;

from Shoreline each sample.

L

' Sample locations are described in the ODCH.

12-42

LA SALLE REVISION O.A APRIL 1991 h, '

"ECF TABLE 12.5.1-1 (Continued) [

RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM EEEE T

i

'C"">

"E3 Exposure Pahtway Number of Samples and Sampling and Type and Frequency

_g 33, and/or Samole Samole Locations

Collection Frecuency of Analysis ---4 4.

k!!5 INGESTION gggs ;

, a. Hilk 3 locations At least once per 15 days Gamma isotopic and I-131 C::3 when animals are-on pasture; analysis of each sample. E!!

at least once per 31 days P"""

at other times. ,

b. Fish 2 locations One sample in season, or at Gamma isotopic analysis least once per 184 days if on edible portions.

not seasonal.

Sample locations are described in the ODCM. ,

e i

b f

L 12-43 i

LA SALLE -

REVISION o.A TABLE 12.5.1-2 @

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL-SAMPLES Reporting Levels llllIO Analysis Hater (PCi/1)

Airborne Particulate-or Gases (pCi/m3)

Fish (pCi/Kg wet).

Hilk Food Products E (pCill) (pCi/Kg, wet) >

D C

H-3 2 X 104 (a) 3

! Mn-54' 1 X 103 3 X 104 C

, 2 Fe-59 4 X'102 1 X 104 Q

, Co-58 1 X 103 3 X 104 Co-60 3 X 102 1 X 104 Zn-65 3 X 102 2 X 104

=

Zr-Nb-95 4 X 102 1-131 2 0.9 3 1 X 102 Cs-134 30 10 1 X 103 60 1 X 103 Cs-137 50 20 2 X 103 70 2 X 103 Ba-La-140 2 X 102 3 X 102 I

(a)for drinking water samples. This is 40 CFR Part 141 value. f 12-44 I

E p ,

LA SALLE REVISION O.A "

APRIL 1991 TABLE 12.5.1-3 -

[

=

HAXIMUM VALUES FOR THE LOHER LIMITS OF DETECTION (LLD)a,c 3 _ l

E i m i Water Airborne Particulate. Fish Hilk Food Products Sediment C- !

Analysis (PCill or Gases (pC1/m3) .(pCi/Kg wet) (pCi/1)- pCi/Kg, wet) (pCi/kg, dry) N E l 3mme -'

gross beta +5 1 X.10-2 1000 NA NA 2000 - !

l O !

, H-3 200 NA **

NA NA NA 2 I Hn-54 '*

NA **' * ** ** O Z >

Fe-59

NA ** * ** **

'k :

t Co-58,60

NA ** * ** **

l Zn-65

NA ** * ** **

Zr-95 NA * ** **

1 Nb-95 NA I-131 NA 10 X 10-2 0.5 30 **

' Cs-134 10 1 X 10-2 ** **

100 10 I

Cs-137 10 1 X 10-2 100 10 ** " !

[

Ba-140 NA La-140 NA

t

,' Gamma isotopic analysis provides LLD of N 20pCi/l per nuclide.

Gamma isotopic analysis provides LLD of N 20pCi/1 per nuclide. ;

L 12-45 a

\

l LA SALLE REVISION 0.A FOR INFORMATION DNLY TABLE 12.5.1-3 (Continued)

TABLE NOTATION

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

4.66 sb LLD - E . V . 2.22 . Y . exp (-K E Where:

LLD is the "a priori" lower limit of detection as defined above (as picocurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank' sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per transformation),

V is the sample size (in units of mass or volume),

2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

1 is the radioactive decay constant for the particular radionuclide,-and at is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environment samples, not plant effluents).

The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g. potassium-40 in milk samples).

Typical values of E, V, Y, and at shall be used in the calculation,

b. LLD for drinking water. s

c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 12.5.1-3, shall be identified and reported. i 12-46

FOR INFORMATION ONLY '^ S^ m - "o^

APRIL 1991 RADIOLOGICAL ENVIRONMENTAL MQNITORING LAND USE CENSUS OPERABILITY REQUIREMENTS 12.5.2.A A land use census shall be conducted and shall identify the location of the nearest milk animal and the nearest residence in each of the 16 meteorological sectors within a distance of five miles. (for elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, then land use census shall also identify the locations of all milk' animals in each of the 16 meteorological sectors within a distance of three miles).

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) which yields a chlculated dose or dose commitment greater than the values currently being calculated in Section 12.4.3.B. in lieu of any other report required by LaSalle Technical Specification 6.6.A., prepare and submit to the Commission within 30 days, pursuant to LaSalle Technical Specification 6.6.C., a Special Report which identifies the new location (s).

b. With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Section 12.5.1.A, in lieu of any other report required by LaSalle Technical Specification 6.6.A., prepare and submit to the Commission within 30 days,

. pursuant to LaSalle Technical Specification 6.6,C., a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.

SURVEILLANCE REQUIREMENTS 12.5.2.B The land use census shall be conducted at least once per 12 months between the dates of (June 1 and October 1) using that information which will provide the best results, such as by a door-to-door survey, aer'al survey, or by consulting local agriculture authortties.

l i

12 47 l

'^ " = Fife F FOR INFORMAIl0N DNLY RADIOLOGICAL ENVIRONMENTAL HONITORING INTERLABORATORY COMPARISON PROGRAM OPERABILITY REQUIREMENTS 12.5.3.A Analyses shall be performed'on radioactive materials supplied as part of an Interikboratory Comparison Program which has been approved by the Commission.

APPLICABILITY: At all times.

ACTION:

Hith analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

SURVEILLANCE REQUIREMENTS 12.5.3.B A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the OOCH (or participants in the EPA crosscheck program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report.

i l

12-48

{g gn pr c n ~rm " ,

LA SALLE fp 9 RADIOLOGICAL ENVIRONMENTAL MONITORING BADIOLOGICAL ENVIRONMENTAL MONITQRING BASES 12.5.1.C HONITORING PROGRAM j The radiological monitoring program required by this specification l provides measurements of radiation and of radioactive materials in those ,

exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulthg from the station j operation. This monitoring program thereby supplements the radiological i effluent monitoring program by verifying that the measurable concentrations of !

radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation, as defined in the 00CH.

The detection capabilities required by Table 12.5.13 are state-of-the-art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as a "a posteriori" (after the fact) limit for a particular measurement. Analyses shr.Il be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

12.5.2.C LAND USE CENSUS This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from tha door-to-door survey, aerial survey or consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

12.5.3.C INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

12-49

LA SALLE REVISION 0.A

^ " '"'

FOR INFORIEW E 12.6 REPORTING REOUIREMENTS 12.6.1 Annual Environmental Radioloaical Ocetittina Remrt 3 l l:

a. Routine environmental radiological operating reports covering the i operation of the unit during the previous calendar year shall be l' submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality, i:

b. The annual environmental radiological operating reports shall include I

^

summaries, interpretations, and an analysis of trends of the results '_

of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, as appropriate, operational controls, as appropriate, and previous !!

environmental. surveillance reports and an assessment of the observed i impacts of the plant operation on the environment. The reports shall i.

also include the results of land use censuses required by Section !

12.5.2.

The annual environmental radiological operating reports shall include I summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975, of all radiological environmental samples taken <

during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. ;-

The missing data shall be submitted as soon as possible in a i supplementary report. -

The reports shall also include the following: a summary description .

of the environmental radiological monitoring program; a map of all !

sampling locations keyed to a table giv_ing distances and directions ;

from one reactor; and the results of licensee participation in the l Interlaboratory Comparison Program, required by Section 12.5.3. p

-The' report shall include an annual summary of hourly meteorological ,

data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured)

! on magnetic tape, or in the form of joint frequency distributions of l wind speed, wind direction, and atmospheric stability. This same i- report shall include an assessment of.the radiation dose due to the

radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. The assessment of radiation doses shall be performed in accordance with the 00CH.

3 A single submittal may be made for a multiple unit station. The submittal shculd combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

12-50 l

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

LA SALLE REVZSION 0.A

^ " " " ' '

FOR INFOREM OM REPORTING REQUIREMENTS l

Annual Environmental Radioloalcal Ooeratina Report (Continued)

The report shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. The assessment of radiation doses shall be performed in accordance.wlth the ODCH.

12.6.2 Semiannual-Radioactive Effluent Release Reoort3 -

a. Routine radioactive effluent release reports covering the operation F of the unit during the previous 6 months of operation shall be [

submitted within 60 days after January l_ and 5J1y 1 of each year. [

The period of the first report shall begin with the date of initial ,

criticality, p

b. The radioactive effluent release reports shall include a summary of I the quantities of radioactive liquid and gaseous effluents and solid !

waste released from the unit as outlined in Regulatory Guide 1.21, !-

"Heasuring, Evaluating and Reporting Radioactivity in Solid Hastes I and Releases of Radioactive Materials in Liquid and Gaseous Effluents [

from Light-Water-Ccoled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarti ly basis following the format of Appendix B thereof.

l L.

3 A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

I 1

12-51

l.

0** l At gr 91 hPhMhb .

IJ REPORTING REQUIREMENTS ,!

Semiannual Radioactive Effluant Release 8epar_t (Continued)

The radioactive effluent release report shall include the following i' information for each type of solid waste shipped offsite during the report period:

a. Container volume, ;

b. Total curie quantity (specify whether determined by measurement or !

estimate), !

c. Principal radionuclides (specify whether determined by measurement or estimate), ;

i

d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms), ,

e. Type of container (e.g., LSA, Type A, Type B Large Quantity), and 4 i

f. Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents en a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

l I

l l

t 12-52

FOR lhiFORMTK O!M

'^S^"' =P!",P REPORTING REQUIREMENTS

-12.6.3 Offsite Dose CAg u ation (ODCM)*

.12.6.3.1 The ODCH shall be approved by the Commission prior to implementation.

12.6.3.2 Licensee-initiated changes to the 00CH:

a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B. This documentation shall contain:

1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes (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 30 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective aft'er review and acceptance by the Onsite Review and Investigative Function and the approval of the Plant Manager on the date specified by the Onsite Review and Investigative Function.

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a-part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCH was made effective. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of-the page that was changed, and shall-indicate the date (e.g.,

month / year) the change was implemented.

12.6.4 MAJOR CHANGES TO RADI0 ACTIVE HASTE TREATHENT SYSTEMS 12.6.4.1 License initiated major changes to the radioactive waste treatment

. systems (liquid and gaseous):

a. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the Onsite Review and Investigative Function. The discussion of each change shall contain:

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

'The OFFSITE DOSE CALCULATION MANUAL (00CM) is common to LaSalle Unit I and LaSalle Unit 2.

12 53

k R M SION 0.A i -

~;'1 LA SALLE APRZL 1991

) ,;..

i REPORTING REQUIREMENTS l

12.6.4 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATHENT SYSTEMS (Cont'd)

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

3. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

4. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;

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

7. An estimate of the exposure to plant operating personnel as a result of the change; and

8. Documentation of the fact that the change was reviewed and found acceptable by the Onsite Review and Investigative Function,

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

12-54

ML20024H308

Text

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