ML20206T140

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Proposed Tech Specs Section 3/4.4.4, Chemistry, Relocating to UFSAR
ML20206T140
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 05/19/1999
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20206T138 List:
References
NUDOCS 9905240080
Download: ML20206T140 (18)
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Text

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4 ATTACHMENT B MARKED-UP PAGES FOR PROPOSED CHANGES REVISED PAGES NPF-11 NPF-18 VI VI XIll Xill XXII MII 3/4 4-10 3/4 4-11 3/4 4-11 3/4 4-12 3/4 4-12 3/4 4-13 8 3/4 4-2 B 3/4 4-2 8 3/4 4-2a Page 1 of1

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9905240080 990519 PDR ADOCK 05000373 P PDR _'  !

. . INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0'UIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.................................i ......... 3/4 4-1 Jet Pumps.................................................... 3/4 4-2 Recirculation Loop F1ow...................................... 3/4 4-3 Idle Recirculation Loop Startup.............................. 3/4 4-4 Thermal Hydraulic Stability.................................. 3/4 4-4a 3/4.4.2 SAFETY / RELIEF VALVES......................................... 3/4 4-5 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................... 3/4 4-6 Operational Le a -

....................................... 3/4 4-7 3/4.4.4 CHEMISTR El.ETED

................................................... 3/4 4-10 3/4.4.5 SPECIFIC ACTIVITY............................................ 3/4 4-13 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System....................................... 3/4 4-16 l

Reactor Steam Dome........................................... 3/4 4-20 1 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................. 3/4 4-21 i 3/4.4.8 STRUCTURAL INTEGRITY......................................... 3/4 4-22 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown................................................. 3/4 4-23 Cold Shutdown................................................ 3/4/4-24 I 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS-0PERATING............................................... 3/4 5-1 3/4.5.2 ECCS-SHUTDOWN........................................'........ 3/4 5-6 l

3/4.5.3 SUPPRESSION CHAM 8ER......................................... 3/4 5-8 LA SALLE - UNIT 1 VI Amendment No. 60

rT~ .

1 i

INDEX BASES l

SECTION PAGE INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Meteorological Monitoring Instrumentation. . . . . . . . . . B 3/4 3-4a Remote Shutdown Monitoring Instrumentation.. . . . . . . B 3/4 3-4a e

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 Explosive Gas Monitoring Instrumentation.. .. . . g B 3/4 3-6 Loose-Part Detection System.. .... ......... . . . B 3/4 3-6 3/4.3.B FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.................................... B 3/4 3-6 l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM.... . .... .. ...... .. ... .. B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.... .. ................ ... . B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.......................... . B 3/4 4-2 pgf_g,; RED Operational Leakage... ............................. B 3/4 4-2 3/4.4.4 hST@.I. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B.3/4 .

. . 4-2 3/4.4.5 SPECIFIC ACTIVITY.............. .. .... ....... .. B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS......... ..... .... B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.. .......... .. . B 3/4 4-5 3/4.4.B STRUCTURAL INTEGRITY. ..... .. . ... B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REMOVAL. . ... .... . . . .... B 3/4 4-5 LA SALLE - UNIT 1 XIII Amendment No. 127

4 m

1.Tst or TABfPs fcentinued1

. 2ARLE IAGE 4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEII. LANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-65 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 3.3.7.11-1 EEPLOSIVE GAS MONITORING INSTRUMENTATION........'........ 3/4 3-83 1 4.3.7.11-1 EXPLOSIVE CAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................... 3/4 3-84 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................... 3/4 3-87 1

3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS..................... 3/4 3-88 4.3.8.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-89 l

3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION j VALVES.................................................. 3/4 4-9 j 4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS................. 3/4 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM........................................ 3/4 4-15 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE..................................... 3/4 4-19 i

1 l

l LA SALLE - UNIT 1 XXII Amendment No.127

. 1 REhCTOR COOLANT SYSTEM 3/OM MEN DNLE LEFIODIN/(

\ AGES 3/4 4-// ANo 3/W-/2 ARE Dsts 3/4.A 4 CHEMISTRY

\ CENTER

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ON THIS PAGE

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LIMITINK CONDITION FOR OPERATION 3.4.4 Th chemistry of the reactor coolant system shall be maintained wit in i the limits pecified in Table 3.4.4-1.

]

APPLICABILITY. At all times.

ACTION:

I

a. In OPERATION L CONDITION 1:
1. With the nductivity, chloride concentration or p exceeding the I limit spec fied in Table 3.4.4-1 for less than 7 ours during one '

continuous ime interval and, fnr conductivity nd chloride -

concentratio , for less than 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> per ye , but with the l conductivity ess than 10 gmho/cm at 25'C a with the chloride concentration ss than 0.5 ppm, this nee not be reported to the ' '

Commission.

2. With the conducti ity, chloride conce ration or pH exceeding the j limit specified in Table 3.4.4-1 for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous time int rval or, for c ductivity and chloride '

concentration, for m e than 336 ours per year, be in at least STARTUP within the ne 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

3. With the conductivity ex eed g 10 pmho/cm at 25'C or chloride concentration exceeding 0.5 pm, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTD N as rapidly as the cooldown rate limit permits.
b. In OPERATIONAL CONDITIONS 2 and 3 w'th the conductivity, chloride  !

concentration or pH excee ng the li it specified in Table 3.4.4-1 for , l more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> durin one continu s time interval, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> d in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

, c. At all other times:

1. With the co ductivity or pH exceeding t e limit specified in Table 3.4 -1, restore the conductivity d pH to within the limit within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
2. With he chloride concentration exceeding t e limit specified in Tabl 3.4.4-1 for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, perfo an engineering ev uation to determine the effects of the ou -of-limit condition on e structural integrity of the reactor coolan system. Determine hat the structural integrity of the reactor to acceptableforcontinuedoperationpriortoproc(antsystemremains e ding to OPERATIONAL CONDITION 3.

The provisions of_ Specification 3.0.3 are not applica le.

NE)(T hGEIS*3}l4~lh LA SALLE - UNIT 1 3/4 4-10 Amendment No. 94

p- . . . _ . - . . . . . . . . . .... ....._.- . u ..-... ;--. w.

REACTOR COOLANT SYSTEM - '

SUR ILLANCE REOUIREMENTS ,

^

4.4.4 e reactor coolant stia11 be determined to be within the specif f e chemist ifmit by:

a. , Measurement prior to pressurizing the reactor during each s rtup, if not performed within the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. /
b. Anal ing a sample of the reactor coolant:
1. Ch orides at least once per:

a) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and l b) 8 ours wt.enever conductivity is grea r than 1.0 pmho/cm at 5'C. -

2. Conductivi at least once per 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> . .-
3. pH at least a ce per:

a) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, d b) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> whene er conducti it'y is greater than 1.0 paho/cm l at 25'C.

c. Continuously recording the con etivity of the reactor coolant, or, when the continuous recordi conductivity monitor is inoperable for up to 31 days, obtainin dip or flow through conductivity measurement at least once er:

! 1. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in OPERATI NAL COND'TIONS 1, 2 and 3, and

2. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at al other times.
d. Performance of HANNEL CHECK of the continuous conductivity monitor with a flow ce at least once per:

1

1. 7 days and
2. 24 ours whenever conductivity is gre ter' than 1.0 pmho/cm 25'C.
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REA'CTOR COOLANT SYSTEM =

8ASES-3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. Analysis has shown that with the safety function of one of the eighteen safety / relief valves in9perable the reactor pressure is limited to within ASME III allowable values for the worst case upset transient. Therefore, operation with any 17 SRV's capable of opening.is allowable, although all . installed SRV's must be closed and have position indication to ensure that integrity of the primary coolant boundary is known to exist at all times.

Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are pro-vided to monitor and detect leakage from the reactor coolant pressure boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"

May 1973.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The ,

normally expected background leakage due to equipment design and the detection  !

capability of the instrumentation for determining systes leakage was also 1 considered. The evidence obtained from experiments suggests- that for leakage  ;

somewhat greater than that specified for unidentified leakage the probability '

is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor i will be shutdown to allow further investigation and corrective action. i The Surveillance Requirements for RCS pressure isolation valves provide  ;

added assurance of valve integrity thereby reducing the probability of gross I valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

'3/4.4.4 (CHEMlIT c-D EL E754 E The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor saterials in contact with the coolant. Chlori limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the i coolant is low, thus the higher _ limit on chlorides is permitted during POWER I OPERATION. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so high concentrations of chlorides re not considered harmful during these periods.

LA SALLE-UNIT 1- 8 3/4 4-2 Amendment No. 60

REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued)

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the con-ductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With'the conductivity j meter inoperable, additional samples must be analyzed to ensure that the  !

chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take corrective action.

, - C

, )ELC 7a 5 pgG&

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LA SALLE-UNIT 1 B 3/4 4-2a Amendment No. 60 J

e INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.................................i ......... 3/4 4-1 Jet Pumps.................................................... 3/4 4-3 Recirculation Loop F1ow...................................... 3/4 4-4 Idle Recirculation Loop Startup.............................. 3/4 4-5 Thermal Hydraulic Stability.................................. 3/4 4-Sa 3/4.4.2 SAFETY / RELIEF VALVES......................................... 3/4 4-6 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leckage Detection Systems.................................... 3/4 4-7 Operational Leak e.......................................... 3/4 4-8 3/4.4.4 (CREMISTRY...... . . . . u.....................................

. .D L EETKe 3/4 4-11 3/4.4.5 SPECIFIC ACTIVITY............................................ 3/4 4-14 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System....................................... 3/4 4-17 Reactor Steam Dome........................................... 3/4 4-21 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................. 3/4 4-22 3/4.4.8 STRUCTURAL INTEGRITY......................................... 3/4 4-23 ;

3/4.4.9 RESIDUAL HEAT REMOVAL i

Hot Shutdown................................................. 3/4 4-24 Cold Shutdown................................................ 3/4 4-25 i 3/4.5 EMERGENCY CORE COOLING SYSTEMS i 3/4.5.1 ECCS-0PERATING............................................... 3/4 5-1 3/4.5.2 ECCS-SHUTD0WN................................................ 3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER......................................... 3/4 5-8 LA SALLE - UNIT 2 VI Amendment No. 69

INDEX BASES SECTION A EA,Gg INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Meteorological Monitoring instrumentation .............................. B 3/4 3-4a Remote Shutdown Monitoring instrumentation....L................... B 3/4 3 4a

. _ Accident Mt.nitoring instrumentation......................................... B 3/4 3-5 Source Range Monitors....... ........ ... ...... ............... ............. ........ B 3/4 3-5 Deleted.................................................................................... B 3/4 3-5 Explosive Gas Monitoring instrumentation................................ I B 3/4 3-6 Loose-Part Detection System................................................... B 3/4 3 6 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION

, I NSTRUM ENTATI ON. ............. .... ...... ...... ............................... B 3/4 S-6 3/4 4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..................................................... B 3/4 4-1 i

3/4.4.2 SAFETY / RELIEF VALVES....................................................... B 3/4 4-1a 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................. .................. B 3/4 4-2 06lET60 Operational Lea kage...... ..... ........ ........ .............. ....................... B 3/4 4-2 3/4.4.4 B 3/4 4-2 QST_Rd....................................................................

3/4.4.5 S P ECI FIC ACTIVITY.............. . . ... ......... ........... . ....... ...... .... .. ..... B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS............... ...................... B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES................................ B 3/4 4-5 1 3/4.4.8 STRUCTURAL INTEG RITY........................... .......... .... ......... B 3/4 4-5 3/4.4.g RESIDUAL HEAT REMOVAL...... .......................................... B 3/4 4-5 l

l l

l LA SALLE- UNIT 2 Xill Amendment No.112

INDEX LIST OF TABLES IContinued)

TJAj,LE f.M.E 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION.............. 3/4 3-67 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQ UIREMENTS................................................... 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 J 3.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION................ ...... 3/4 3-83 4.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................. 3/4 3-84 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.................................................... 3/4 3-87 3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.......................... .... 3/4 3-88 4.3.8.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS................. 3/4 3-89 l

3.4.3.2-1' REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES... 3/4 4-10

[3)4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS.................. .... 3/44-1j j 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROG RAM......... . . .. ....... .. ... . ....... ....... .... . .. .. . . .... ............. . 3/4 4-16 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM l WITH DRAWAL SCH EDU LE.... .. .. . ... .. . ...... .. ... .. .... ... . . .. . . . . . ... ......... . ... . 3/4 4-20 l l

I

)

LA SALLE - UNIT 2 XXil Amendment No.112 f

  1. 4,4.4 INTE.NTl0NALLY 4EF7 8LA4/K REACTOR COOLANT SYSTEM pas 55 W4 4-1). AND sh 4-13 AAEDELETE 3/\4 4.4 CHEMISTRY )

5.NTEA 0/v THis PMS LIMIT G CONDITION FOR OPERATION .}

3.4.4 T chemistry of the reactor coolant system shall be maintained wit n I the limit specified in Table 3.4.4-1. I APPLICABILI Y: At all times.

FACTION: /

a. In OPERAT NAL CONDITION 1:
1. With t e conductivity, chloride concentration or p exceeding the limit ecified in Table 3.4.4-1 for less than 72 ours during on, '

continu us time interval and, for conductivity d chloride concentrgtion, for less than 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> per ye , but with the conductiv ty less than 10 gnho/cm at 25*C an with the chloride concentra ion less than 0.5 ppm, this need at be reported to the Commission. k

2. With the con uctivity, chloride concen ration or pH exceeding the limit specifik in Table 3.4.4-1 for ore than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous tim interval or, for co ductivity and chloride concentration, more than 336 urs per year, be in at least STARTUP within th next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3. With the conductivit exceedi 10 pnho/cm at 25'C or chloride concentration exceedin 0.5 pm, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHU N as rapidly as the cooldown rate limit permits.
b. In OPERATIONAL CONDITIONS 2 nd 3 ith the conductivity, chloride concentrationorpHexceedfngthe1 it specified in Table 3.4.4-1 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> durin /one contin aus time interval, be in at least HOT SHUTDOWN within the ext 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

t

c. At all other times:
1. With the co uctivity or pH exceeding e limit specified in Table 3.4. -1, restore the conductivity nd pH to within the limit within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
2. With t e chloride concentration exceeding th limit specified in Tabl 3.4.4-1 for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, perform an engineering eva uation to determine the effects of the out- f-limit condition on t structural integrity of the reactor coolant stem. Determine at the structural integrity of the reactor coola t system remains i acceptable for continued operation prior to proceed g to l OPERATIONAL CONDITION 3.

. The provisions of Specification 3.0.3 are not applicabl NEWPAGE%5/4'll-14 LA SALLE - UNIT 2 3/4 4-11 Amendment No. 78

NEACTOR C001. ANT SYSTEM Subflu.ANCE REQUIREMENTS -

/ \

~ '

fied 4.4.4 chemis The limitreactor by: coolant shall be determined to be 'withi7 the's a Measurement prior to pressurizing the reactor during ch startup, f l if not performed within the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. _

b. Analyzing a sample of the reactor coolant: .

I

\

L Chlorides at least once per:

a) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,and f

. ) 8 hourt whenever conductivity is ter than L0 paho/cm

. 'at 2P C. ~

2. ivity at least once per 72 urs.

\

pH at least once per:

3.

a) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />., and .

% \.

b) hourt whenever vity is greater than L0 paho/cm 25*C. .

c. Continuously ing the ivity of the reactor coo.lant, or, I I

when the cent puous recording conductivi*,y monitor is inoperable for up to 31 ddys, obtaining'a dip or flow through conductivity measurement at 1 t once r:

L 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in TI C0fCITIONS 1, 2 and 3, ar.:t L 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at al r times.

, d. Performance of a CHECK of the continuous conductivity monitor . . .

'with a fTow col at 1 ones per:

L 7 days -

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..g..Un uns aise considered. ' The erfdence obtained free egeriennts suggests that for-b unidentf ffet leakage the f ..

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ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Comed has evaluated this proposed amendment and has determined that it does not represent a significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated; q

Create the possibility of a new or different kind of accident from any previously analyzed; or Involve a significant reduction in a margin of safety.

Comed proposes to relocate, to administrative controls, the chemistry limits provided in Technical Specifications (TS) Section 3/4.4.4, Reactor Coolant System Chemistry.

The determination that the criteria set forth in 10 CFR 50.92 (c) is met for this amendment request is indicated below:

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes simplify the TS, meet regulatory requirements for relocated TS, and implement the recommendations of the NRC's Final Policy Statement on TS improvements. The Chemistry requirements will be relocated to the Updated Final Safety Analysis Report (UFSAR) and Administrative Technical Requirement that has been incorporated into the UFSAR by reference.

Future changes to these requirements will be controlled by 10 CFR 50.59. The proposed changes are administrative in nature and do not involve any modification to any plant equipment or affect plant operation. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any previously evaluated accident.

Consequently, this proposed amendment does not involve a significant increase l in the probability or consequences of any accident previously evaluated. l i

l l

Page 1 of 2 i

l

4 ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Does the change create the possibility of a new or different kind of accident

- from any accident previously evaluated?

The proposed changes are administrative in nature, do not involve any physical alterations to any plant equipment, and cause no change in the method by which l any safety related system performs its function. Therefore, this proposed TS amendment would not create the possibility of a new or different kind of accident

. from any accident previously evaluated.

Does the change involve a significant reduction in a margin of safety?

The proposed amendment represents the relocation of current requirements that are based on generic guidance or previously approved provisions for other stations.~ The proposed changes are administrative in nature and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. The proposed changes have been evaluated and found to be acceptable for use at Duane Arnold Energy Center and Quad Cities Nuclear Power Station. Since the proposed changes are administrative in nature, and are based on NRC accepted provisions which have been adopted at other nuclear facilities, and maintain the necessary levels of system reliability, the proposed changes do not involve a significant reduction in the margin of safety.

Therefore, based upon the above evaluation, Comed has concluded that these changes do not constitute a significant hazards consideration.

Page 2 of 2

ATTACHMENT D ENVIRONMENTAL ASSESSMENT Comed has evaluated this proposed operating license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) The amendment involves no significant hazards consideration.

As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed change is administrative in nature. There will be no change in the types or significant increase in the amounts of any effluents released offsite.

(iii) There is no significant increase in individual or cumulative occupational I radiation exposure. l The proposed changes will not result in changes in the operation or ,

configuration of the facility. There will be no change in the level of controls {

cr methodology used for processing of radioactive effluents or handling of l solid radioactive waste, nor will the proposal result in any change in the j normal radiation levels within the plant. Therefore, there will be no increase l in individual or cumulative occupational radiation exposure resulting from I this change.

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