Text

Proposed Tech Spec Changes to Conform to 10CFR50.73 Requirements & to Delete Channel Check Requirements from Certain Instruments
	ML20107M575
Person / Time
Site:	LaSalle
Issue date:	02/21/1985
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20107M570	List: ML20107M560; ML20107M575;
References
	NUDOCS 8503010347
	Download: ML20107M575 (84)
	v • d • e

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ATTACHMENT A I

Marked Up Technical Specification Pages Unit 1 (NPF-ll) 10 CFR 50.73' CHANGES l-Page II Page 1-6

.Pages 3/4 3-60, 3/4 3-63, 3/4 3-75 Pages 3/4 4-13 Pages 3/4 7-11, 3/4 7-14, 3/4 7-17, 3/4 7-18, 3/4 7-22, 3/4 7-24 Page 3/4 8-7 Pages 3/4 11-6, 3/4 11-7, 3/4 11-13, 3/4 11-14, 3/4 11-15, 3/4 11-16, 3/4 11-20, 3/4 11-22 Page 3/4 12-1, 3/4 12-9 Pages 6-9, 6-18, 6-19, 6-24 through 6-28 CHANNEL CHECK CHANGES Pages 3/4 3-7, 3/4 3-20, 3/4-3-22, 3/4 3-32, 3/4 3-33, 3/4 3-34, 3/4 3-49 9785N s

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ATTACHMENT A

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Marked Up Technical Specification Pages Unit 2 (NPF-18) 10 CFR 50.73 CHANGES Page II Page 1-6 Pages 3/4 3-60, 3/4 3-63, 3/4 3-75 Pages 3/4 4-14 Pages 3/4 7-11, 3/4 7-14, 3/4 7-17, 3/4 7-18, 3/4 7-23, 3/4 7-25 Page 3/4 8-7 Pages 3/4 11-6,.3/4 11-7, 3/4 11-13, 3/4 11-14, 3/4 11-15, 3/4 11-16, 3/4 11-20, 3/4 11-22 Page- 3/4 12-1, 3/4 12-9 Pages 6-9, 6-18, 6-19, 6-24 through 6-28 CHANNEL CHECK CHANGES Pages 3/4 3-7, 3/4 3-20, 3/4 3-22, 3/4 3-32, 3/4 3-33, 3/4 3-34, 3/4 3-49 l 9785N t

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E' ATTACHMENT B

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e-f LASALLE COUNTY STATION UNITS 1 and 2 TECHNICAL SPECIFICATION CHANGE REQUEST 10 CFR 50.73 CHANGES BACKGROUND:

The final rule for Reference (a) was published in the Federal Register on July 27, 1983. This rule became ef fective on

-January 1, 1984. Paragraph 50.73(g) states that the reporting requirements contained in reference (a) replace the reporting requirements in all Nuclear Power Plant Technical Specifications that.are typically associated with Reportable Occurances (reference (c)).

DISCUSSION:

Reference (b) outlines the Technical Specification changes that'are needed to implement the regulation changes. As of January 1, 1984 LaSalle County Station has complied with Reference (3) in lieu of Section 6.6.8 of the Technical Specifications as required by Section 10 CFR 50.73 paragraph (g). The enclosed marked up pages Attachment C for LaSalle County Station Unit 1 (NPF-11) and Unit 2 (NPF-18) are believed to be in accordance with these regulations and as outlined in Reference (b).

CONCLUSION:

Commonwealth Edison believes that these changes are in accordance'with NRC regulations and that no unreviewed safety

-questions-exist.

9785N

A S M ws.,T c.

INDEX DEFINITIONS v

SECTION DEFINITIONS (Continued) PAGE 1.25 OPERABLE - OPERABILITY............................................ 1-4 1.26 OPERATIONAL CONDITION - CONDITION................................. 1-4 1.27 PHYSICS TESTS..................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE......................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY..................................... 1-5 1.30 PROCESS CONTROL PR0 GRAM........................................... 1-5 1.31 PURGE - PURGING................................................... 1-5 1.32 RATED THERMAL P0WER............................................... 1-5 1.33 REACTOR PROTECTION SYSTEM RESPONSE TIME........................... 1-5

. OEr4 r 1.34 REPORTABLE 96664AEMCE............................................. 1-6 9

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1.35 ROD DENSITY....................................................... 1-6 1.36 SECONDARY CONTAINMENT INTEGRITY................................... 1-6 ,

1.37 SHUTDOWN MARGIN................................................... 1-6 1.38 SOLIDIFICATION.................................................... 1-6 1.39 SOURCE CHECK...................................................... 1-7 1.40 STAGGERED TEST BASIS.............................................. 1-7 1.41 THERMAL P0WER..................................................... 1-7 1.42 TURBINE BYPASS RESPONSE IIME...................................... 1-7

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1.43 UNIDENTIFIED LEAKAGE.............................................. 1-7 1.44 VENTILATION EXHAUST TREATMENT SYSTEM.............................. 1-7 1.45 VENTING........................................................... 1-7 TABLE 1.1, SURVEILLANCE FREQUENCY NOTATION............ ............ 1-8 TABLE 1.2, OPERATIONAL CONDITIONS..................... g ............ 1-9

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LA SALLE - UNIT 1 ,

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8 DEFINITIONS E44 '

REPORTA8LE 4GGUAREMCf Ev'GMT -

1.34 A REPORTABLE 96GUAAENGE shall be any of those conditions specified in Sp="hth . 0. 0. 0.1 end 0. 0. 0. 2. <gcT /cd 60.15 Tb o c.FR F3.2r s ,

ROD DENSITY 1.35 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% R00 DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.36 SECONDARY CONTAINMENT INTEGRITY shall' exist when:

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

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

2. Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed 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.

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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.F.5.1.a.

SHUTDOWN MARGIN 1.37 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth .

which is assumed to be fully withdrawn and the reactor is in the shutdown

- conditlon; cold, i.e. 68'F; and xenon free.

SOLIDIFICATION 1.38 SOLIDIFICATION 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 I distinct outline on all sides (free-standing).

w LA SALLE - UNIT 1 1-6

INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION

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l LIMITING CONDITION FOR OPERATION i 3.3.7.2 The seismic monitoring instrumentation shown in Table 3.3.7.2-1 shall be OPERA 8LE.** l APPLICA8ILITY: At all times.

ACTION:

a. With one or more seismic monitoring instruments inoperable for more i than 30 days, fr. 1;.. 7 :n; cr.:- 7:;:M : ;t: ' by !;::t't::tiaa ;

-C.;.;, prepara 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 instrument (s) to !

OPERA 8LE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. 1 SURVEILLANCE REQUIREMENTS 4.3.7.2.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL

' FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in l Table 4.3.7.2-1.

4.3.7.2.2 Each of the above required seismic monitoring instruments actuated during'a seismic event greater than or equal to 0.02g shall be restored to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within

, 5 days following the seismic event. Data shall be retrieved from actuated round instrumentsandanalyzedtodeterminethemagnitudeofthevibratoryg%Special motion. " I f = d : ;, .tM r :;:d 7 4 f . .J L, $y. i iis.U: ; 5. 5. 53 Report shall be prepared and submitted to the Commission pursuant to Specifica-

. tion 6.6.C within 10 days describing the magnitude, frequency spectrum and l resultant effect upon unit features important to safety.

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The Seismic Monitoring Instrumentation System is shared between La Salle Unit 1 and La Salle Unit 2. ,

l **The normal or emergency power source may be inoperable in OPERATIONAL l l

CONDITION 4 or 5 or when defueled.

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LA SALLE - UNIT 1 3/4 3-60 Amendment Nc.18 l

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INSTRUMENTATION METEOROLOGICAL MONITORING INSTRUMENTATION

l LIMITING CONDITION FOR OPERATION l

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1 3.3.7.3 The m'eteorological monitoring instrumentation channels shown in Table 3.3.7.3-1 shall be OPERA 8LE.** l APPLICA8ILITY: At all times.

ACTION:

a. With one or more meteorological monitoring instrumentation channels

. inoperable for more than 7 days, ' er :? : ; _^.:... 4;rt ---M -*

., ^.,...";;.M ;. C.C.0, 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 instrumentation to OPERA 8LE status.

a

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

4 SURVEILLANCE REQUIREMENTS i

^i 4.3.7.3 Each of the above required meteorological monitoring instrumentation channels shall be demonstrated OPERA 8LE by the performance of the CHANNEL

, CHECK and. CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.3-1. ,

The Noteorological Monitoring Instrumentation System is shared between La Salle Unit 1 and La Salle Unit 2.

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- The normal or emergency power source may be inoperable in OPERATIONAL l CONDITION 4 or 5 or when defueled.

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! LA SALLE - UNIT 1 3/4 3-63 Amendment N'o.18

O INSTRUMENTATION FIRE DETECTION INSTRUMENTATION l

LINITING CONDITION FOR OPERATION 3.3.7.9 As a minimum, the fire detection instrumentation for each fire detection zone of Unit 1 and Unit 2 shown in Tahle 3.3.7.9-1 shall be OPERABLE.*

APPLICA81LITY:. Whenever equipment protected 'by the fire detection instrument is required to be OPERA 8LE.

ACTION:

With the number of OPERA 8LE fire detection instruments less than the Minimum Instruments OPERABLE requirement of Table 3.3.7.9-1:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the primary containment, then inspect the primary containment at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.7.

b. Restore the minimum number of instruments to OPERA 8LE status within 14 days or, # " -" :? ;.., ww .. . .y . . ..y.' .4 ;, Se;if t:; tin 6,+.4, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERA 8LE status.

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

SURVEILLANCE REQUIREMENTS 4.3.7.9.1 Each of the above required fire detection instruments which are ,

accessible'during unit operation shall be demonstrated OPERA 8LE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST. Fire detectors which are not accessible during unit operation shall be demonstrated OPERA 8LE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

4.3.7.9.2 The NFPA Standard 720 supervised circuits supervision associated with the detector alarms of each of the above required fire detection

- instruments shall be demonstrated OPERA 8LE at least once per 6 months. Super-

vised circuits with detectors which are inassessible during unit operation shall be demonstrated OPERA 8LE during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

"The normal or emergency power source may be inoperable 6 '*ERATIONAL

, CON 0!T!0N 4 or 5 or when defueled. 4 LA SALLE - UNIT 1 3/4 3-75

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REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to: ,.

a. Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT l r I-131, and

b. Less than or equal to 100 d microcuries per gram.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of the primary coolant;

1. Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that the cumulative operat-ing time under these circumstances does not exceed 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months in ,

any consecutive 12 month period. With the total cumulative operating time at a primary coolant specific activity greater than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131 l exceeding 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in any consecutive six month period, prepare and submit a special report to the Commission pursuant to Specification 6.6.C within 30 days indicating the number of hours of operation above this limit. The provisions of Specification 3.0.4 are not applicable.

2. Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or for l

r more than 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months cumulative operating time in a consecutive

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12-month period, or greater than or equal to 4 microcuries per l l ,

gram, be in at least HOT SHUTDOWN with the main steam line i

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

3. Greater than 100/E microcuries per gram, be in at least HOT l SHUTDOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i b. In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0: 2 microcurie per gram DOSE l

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l EQUIVALENT I-131 or greater than 100 E microcuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to y within the limit. A ZZT" E ^"'JZ Z7shall be preparea and gg i

submitted to the Commission pursuant to Specification 6.6.K.* This

! report shall contain the results of the specific activity analyses and the time duration when the specific activity of the coolant l exceeded 0.2 microcurie per gram 00$E EQUIVALENT I-131 together

l with the following additional infomation.

LA SALLE - UNIT 1 3/4 4-13 Amendment No.18

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PLANT SYSTEMS'

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=3/4.7.5 FIRE SUPPRESSION $YSTEMS .

FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.5.1 The fire suppression water system shall be OPERABLF* with:

a. Two OPERABLE fire suppression diesel driven fire pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header,

b. An OPERABLE flow path capable of taking suction from the CSCS water tunnel and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler, hose standpipe, and the last valve ahead of the deluge valve on each deluge or spray system required to be OPERABLE per Specifications 3.7.5.2 and 3.7.5.4.

APPLICABILITY: At all times. ,

ACTION:

a. With one fire pump and/or one water supply inoperable, restore the inoperable eouipment to OPERABLE status within 7 days or, k lie. er

q tMc .. pert r;; ir;d h Sp::i'h:thr 6.5.S, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the plans and procedures to be

. used to restore the inoperable equipment to OPERABLE status or to provide an alternate backup pump or supply. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b. With the fire suppression water system otherwise inoperable:

1. Establish a backup fire suppression water system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. h " eu ef r., et..% r rep e t r:; b ed O Spec 4 " c?t h r C.C.0 ,

pgnypu pro submit a Special Report in accordance with Specification 6.6.C; a) By telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

. b) Confirmed by telegraph, mailgram or facsimile transmission ,

- - no later than the first working day following the event,

. . and c) In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

"The normal or emergency power source may be inoperable in OPERATIONAL CONDITION 4 or 5 or when defueled.

LA SALLE - UNIT 1 ,

3/4 7-11

PLANT SYSTEMS t

DELUGE AND/OR SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.5.2 The deluge and sprinkler systems of Unit 1 and Unit 2 shown in -

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Table 3.7.5.2-1 shall be OPERABLE.*

APPLICA8ILITY_: Whenever equipment protected by the deluge / sprinkler systems are re;. aired to be OPERABLE.

ACTION:

a. With one or more of the deluge and/or sprinkler systems shown in Table 3.7.5.2-1 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous I fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system tc OPERABLE status within 14 days or, ' 'i : :' x , , C . ..p.c; c a .ir d t, Ce nt' 5 'f:- 5.0." prepare and submit a Special Report to the Commission pursuant to Sp,ecification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERA 8LE status.

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

SURVEILLANCE REQUIREMENTS _

4.7.5.2 Each of the above required deluge and sprinkler systems shown in Table 3.7.5.2-1 shall be demonstrated, OPERA 8LE:

a. At least once per 31 days by verifying that each valve, (manual, power operated or automatic), in the flow path is in its correct position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

c. At least once per 18 months: ;

. . 1. By performing a system functional test which includes simulated automatic actuation of the system, and:

a) Verifying that the automatic valves in the flow path actuate to their correct positions on a test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

"The normal or emergency power source may be inoperable in OPERATIONAL CONDITION 4 or 5 or when defueled.

LA SALLE - UNIT 1 3/4 7 14 Amendment No. 18

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c i PLANT SYSTEMS i

C0. SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.5.3 The following low pressure C0 systems of Unit 1 and Unit 2 shall be,.

'l '

OPERA 8LE.*

b

a. Division 1 diesel generator 0 room.

b. Division 2 diesel generator IA room.

c. Division 3 diesel generator 18 room.

d. Unit 2 Division 2 diesel generator 2A room.

j, 4

APPLICABILITY: Whenever equipment protected by the low pressure C02 systems is required to be OPERA 8LE.

, s l ACTION:

a. With one or more of the above rwquired low pressure CO systems '

i inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with lid backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERA 8LE status within 14 days or, t, M :: c' r; :P :7 .;;;c; ._ ,_ fred 4

!;::f'i :ti:: 5.5Aprepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the.inoperability and the

< plans and schedule for restoring the system to 0PERA8LE status.

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

SURVEILLANCE REQUIREMENTS s

4 4.7.5.3 Each of the above required low pressure C0 systems shall be demonstrated OPERA 8LE:

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' a. At least once per 7 days by verifying C0 storage tank level to be greater than 505 full and pressure to be greater than 290 psig, and l 13 y

b. At least once per 31 days by verifying that each valve (manual, power 4 operated, or automatic) in the flow path Is in the correct position.

i

c. At least once per 18 months by verifying:

1. The system valves and associated motor operated ventilation

dampers actuate, manually and automatically, upon receipt of a simulated actuation signal, and

2. Flow from each n'ozzle during a " Puff Test."

t "The normal or emergency power source may be inoperable in OPERATIONAL ;

, CONDITION 4 or.5 or when defueled.

LA SALLE - UNIT 1 3/4 7-17 Amendment No.18

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PLANT SYSTEMS FIRE HOSE STATI0h5 .

LIMITING CONDITION FOR OPERATION

, 3.7.5.4 The fire hose stations of Unit 1 and Unit 2 shown in Table 3.7.5.4-1 shall be OPERA 8LE.

APPLICA81LITY: Whenever equipment in the areas protected by the fire hose ,

stations is required to be OPERA 8LE.

ACTION:

a. With one or more of the fire hose stations shown in Table 3.7.5.4-1 inoperable, route an additional fire hose of equal or greater diameter to the unprotected area (s)/ zone (s) from an OPERA 8LE hose station 4

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable fire hose is the primary means of fire suppression; otherwise, route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Restore the inoperable fire hose station (s) to OPERA 8LE status within 14 days or , ' " r ' r; et': ca;rt . .,. . . .. ;,, ' r f ::t ? - f E ",

prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applic.able.

SURVEILLANCE REQUIREMENTS 1

4.7.5.4 Each of the above required fire hose stations shown in Table 3.7.5.4-1 .

shall be demonstrated OPERA 8LE:

a. At least once per 31 days by a visual inspection of the fire hose stations accessible during plant operation to assure all require (.

. equipment is at the station.

b. At least once per 18 months by:

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1. Visual inspection of the fire hose stations not accessible during plant opsration to assure all required equipment is at the station.

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Removing the hose for inspection and reracking, and l 2.

3. Inspecting all gaskets and replacing any degraded gaskets in the couplings.

I

c. At least once per 3 years by partially opening each hose station valve to verify valve OPERA 8ILITY and no flow blockage.

d. Within 5 years and between 5 and 8 years after purchase date and at least every 2 years thereafter by conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above the maximum fire main i 6

operating pressure, whichever is greater.

LA SALLE - UNIT 1 3/4 7-18 Amendment ilo.18 l .

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' PLANT SYSTEMS 3/4.7.6 FIRE RATED ASSEM8 LIES LIMITING CON 0! TION FOR OPERATION 3.7.6 All fire rated assemblies, including walls, floor / ceilings, cable tray I enclosures and other fire barriers separating safety related fire areas or I separating portions of redundant systems important to safe shutdown within a fire area, and all sealing devices in fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable and piping penetration seals and ventilation seals) shall be OPERA 8LE.

APPLICA8ILITY: At all times.

ACTION:

a. With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within one hour either establish a con-tinuous fire watch on at least one side of the af facted assembly (s) l and/or device (s) or verify the OPERA 81LITY of fire detectors on at least one side of the inoperable assembly (s) and/or sealing device (s) and establish an hourly fire watch patrol. Restore the inoperable fire rated assembly (s) and/or sealing device (s) to 0PERABLE status within 7 days or, 'r. ";; 7 ::, ;th ..,~,;. 7:;i:f O W444een

--ti:: 5.5.0, prepare and submit a Special Report to the Commission pursuant the actiontotaken, Specification the cause 6.6.C of thewithin the next inoperable 30rated fire daysassembly outlining (s) and/or sealing device (s) and plans and schedule for restoring the fire rated assembly (s) and/or sealing device (s) to OPERA 8LE status,

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

$URVEILLANCE REQUIREMENTS 4.7.6.1 Each of the above required fire rated assemblies and sealing devices -

shall be verified to be OPERA 8LE at least once per 18 months by performing a visual inspection of: -

a. The ex;:or.d surfaces of each fire rated assemblies.

b. Each fire window / fire damper and associated hardware.

c. At least 10 percent of each type of sealed penetration. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10 percent of each type of saaled penetration shall be made. This inspection process shall continue until a 10 percent sample with no apparent changes in appearance

( or abnormal degradation is found.

LA SALLE - UNIT 1 3/4 7-22 Amendment Ifo. B

r PLANT SYSTEMS 3/4.7.7 AREA TEMPERATURE MONITORING LIMITINJ CONDITION FOR OPERATION 3.7.7 The temperature of each area of Unit 1 and Unit 2 shown in Table 3.7.7-1 l shall be maintained within the limits indicated in Table 3.7.7-1.

APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE.

ACTION:

With one or more areas exceeding the temperature limit (s) shown in Table 3.7.7-1:

a. For more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , in lieu of any 7:;r-+

LW s.euep., Gu p::ifj.

;r;d by$.nt Epegl

ti:: S.S.", prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days providing a record of the amount by which and the cumulative time the temperature '

in the affected area exceeded its limit and an analysis to demonstrate the continued OPERABILITY of the affected equipment.

b. By more than 30*F, in addition to the Special Report required above, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area to within its temperature limit or declare the equipment in the affected area inoperable.

SURVEILLANCE REQUIREMENTS 4.7.7 The temperature in each of the above required areas shown in Table 3.7.7-1 shall be determined to be within its limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e I .

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3/4 7-24 Amendment No.18 LA SALLE - UNIT 1

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting diesel gener-ators 0, lA and IB simultaneously, during shutdown, and verifying that all three diesel generators accelerate to 900 rpm + 5. -2% in less than or equal to 13 seconds.

f. At least once per 10 years by:

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and

2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND, of the ASME Code in accordance with ASME Code Section 11, Article IWD-5000.

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall /wa(;g ]

be reported to the Commission pursuant to Specification 6.6JBG'gKeports of diesel generator failures shall include the information recommended in Regula-[_so dagQ tory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the l number of failures in the last 100 valid tests, on a per nuclear unit basis, l is greater than or equal to 7 -the report shall be supplemented to include the additional information recommended in Regulatory Position c.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

TABLE 4.8.1.1.2-1 l DIESEL GENERATOR TEST SCHEDULE Number of Failures in Last 100 Valid Tests

Test Frequency i1 At least once per 31 days 2 _

At least once per 14 days 3 At least once psr 7 days

>4 At least once per 3 days

" Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the last 100 tests are determined on a per nuclear unit basis. With the exception of the semi-annual fast start, no starting time re-quirements are required to meet the valid test requirements of

( Regulatory Guide 1.108.

LA SALLE - UNIT 1 3/4 8-7 Amendment No. 18

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RADI0 ACTIVE EFFLUENTS 8 l

l DOSE ,.

9 LIMITING CONDITION FOR OPERATION l 3.11.1.2 The dose or dose commitment to an individual from radioactive l materials in liquid effluents released, from eacn reactor unit, from the site l (see Figure 5.1.1-1) shall be limited:

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 or equal to 3 mrem to the total body and to less than or-equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With 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 Specification 6.6.A er S.S.B, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 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 calendar quarter and '- '

during the subsequent three calendar quarters, so that the cumulativa 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 impact on finished drinking water supplies at the nearest downstream drinking water source.

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

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid .

. effluents sheil be determined in accordance w.ith the 00CM at least once per

, . 31 days.

LA SALLE - UNIT 1 3/4 11-6 e

RADIOACTIVE EFFLUENTS LIQUID WASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION

/

3.11.1.3 The liquid radwaste treatment system shall be OPERAB'E. L 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 (see Figure 5.1.1-1), when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.

APPLICABILITY: At all times.

~

ACTION:

a. With the liquid 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 Specification S.6. A e.4 0.0.0, prepare and submit -

to the Commission within 30 days pursuant to 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.

o \

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

F .

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due. to liquid releases shal'1 be projected at least once per 31 days, in accordance with the ODCM. -

4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 30 minutes at least cnce cer 92 days unless the liquid radwaste system has been utili ec ;a precar.; . :. , -::' .e ;icu'c .f ficerts cerirg :n1 ;ct icas 52 ca:. .

LA SALLE - UNIT 1 3/4 11-7

RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES.

LIMITING CONDITION FOR OPERATION l

l 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site (see Figure 5.1.1-1) shall be limited to the following:

a. During any calendar 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. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.6. A 4w=4h4W4F, prepare and submit to the Commission within 30 days, pursuant to 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 and the proposed corrective actions to be taken to assurs that subsequent releases will be in compliance with the above limits.

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

SURVEILLANCE REQUIREMENTS

.4.11.2.2 Dose Calculations Cumulative dose contributions for the current l calendar quarter and current calendar year shall be determined in accordance l with the 00CM at least once per 31 days.

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LA SALLE - UNIT 1 3/4 11-13 Amendment No. 18 I.. .

RADI0 ACTIVE EFFLUENTS DOSE - RADIOI0 DINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiciodines and radioactive materials in particulate form, and radionuclides, other than noble gases, with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site (see Figure 5.1.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to any l organ, and

b. During any calendar year: Less than or equal to 15 areas to any I organ.

l APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides (other than noble gases) with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.6. A .. C.:.0,- prepare and submit to the Commission within 30 days, pursuant to Specification 6.6.C a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been 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.

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

l SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once per 31 days.

IA SALLE - UNIT 1 3/4 11-14 Amendment No. 18

6 RADI0 ACTIVE EFFLUENTS ,

GASEOUS RA0 WASTE TREATMENT SYSTEM .

LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEQUS RADWASTE TREATMENT SYSTEM shall be in operation.

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

ACTION:

a. With the GASEOUS RADWASTE TREATMENT SYSTEM inoperable for more than 7 days, in lieu of any other report required by Specification 6.6. A

-er 0.C.0, prepare and submit to the Commission within 30 days, pursuant to 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 CPERABLE status, and

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

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

SURVEILLANCE REOUIREMENTS 4.11.2.4 The GASEOUS RA0 WASTE TREATMEhi SYSTEM shall be verified to be in operation at least once per-7 days.

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LA SALLE - UNIT 1 3/4 11-15

+

i l

, l RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM .

LIMITING CONDITION FOR OPERATION 4

3.11.2.5 The appropriate portions of the VENTILATION EXHAUST TREATMENT 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 (see Figure 5.1.1-1), when averaged over.31 days, would exceed 0.3 mrem to any organ. .

APPLICABILITY: At all times.

ACTION: .

a. With the VENTILATION EXHAUST TREATMENT 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 Specification 6.6. A r C.C.", prepare and submit I

~

to the Commission within 30 days, pursuant to-Scecification 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.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i

~

SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with the 00CM.

4.11.2.5.2 The VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated l . 0PERABLE by cperating the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 30 minutes, at least once per 92 days unless the appropriate system has been utili:ed to Maecess radica:'.i"o ' 'tr:s eu?--- du '*- t" Oce irue 92 c3'.3.

i LA SALLE - UNIT 1 3/4 11-16

' . . . . . ... . .: :. .-- . . - a--- . . .. :-. ~

RADI0 ACTIVE EFFLUENTS 3/4.11:3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be OPERABLE and used, as applicable in l accordance with a PROCESS CONTROL PROGRAM, for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactiv,e wastes from the site.

APPLICABILITY: At all times.

ACTION:

a. With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Part 71 not satisifed, suspend shipments of defectively packaged solid radioactive wastes from the site.

b. With the solid radwaste system inoperable for more than 31 days, in

- lieu of any other report required by Specification 6.6.A .c C.C.0, .(-

prepare and submit to the Cor. mission within 30 days, pursuant to Specification 6.5.C, a Special Report which includes the following '

information:

1. Identification of the inoperable aquipment or subsystems and the reason for inoperaDility, ,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, .

3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and

4. Summary description of action (s) taken to prevent a recurrence,

c. The provisions of Specifications 3.0.3 and 3.0'.4 are not apolicable.

. SURVEILLANCE REQUIREMENTS 4.11.3.1 The solid radwaste system shall be demonstrated OPERABLE at least once per 92 days by:

a. Cee-atir.c the ic'id -:dwaste :ystem at least o. ce in the pre.'cus

, '22 da;. s -

;;: 3 :e vi .- .:a ?r:cesa Centroi Tr::-27., cc

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b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

LA SALLE - UNIT 1 3/4 11-20

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RADIOACTIVE EFFLUENTS

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3/4.11.4 TOTAL DOSE .

LIMITING CONDITION FOR OPERATION 3.11.4 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 orga.n (except l.

the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica-ti ons 3.11.1. 2. a , 3.11.1. 2. b , 3.11. 2. 2. a , 3.11. 2. 2. b , 3.11. 2. 3. a , or .

3.11.2.3.b, in lieu of any other report required by Specification 6.6. A er S.C.", prepare and submit, pursuant to Specification 6.6.C, a [

Special Report to the Director, Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, Washington, D.C. 20555, within 30' days,'which defines the corrective action to be takea to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle scurces (including all effluents pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the esti-mated dose (s) exceeds the limits of Specification 3.11.4, and if the o 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 f 190.11. Submittal of the report is considered a. timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 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.

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

SURVEILLANCE REOUIREMENTS 3.11. : :~ ;e C. ' :i a:' CumuTa:' e 1:- - -

.:' - '-~~N v:

gaseous ef fluents snali ce determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance witn the ODCM.

LA SALLE - UNIT 1 3/4 11-22

f," '

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORI%G

'3/4.12.1 MONITORING PROGRAM .

LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12.1-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, in lieu of any other (,.G A report required oy Specification 6 & i,~ prepare and submit to the -

Commission, in the Annual Radiological Operating Report, a descrip-tion 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 l

l exceeding the reporting levels in Table 3.12.1-2 when averaged over l any calendar quarter, in lieu of any other report required by Specification ,6&i, prepare ind submit to the Commission within 30

%M.A;- days from the end of the affected calendar quarter a Report pursuant ,

l to Specification ' ^ ' " cted When more than one of the radionuclides l

l in Table 3.12.1-2 are in the sampling medium, this report shall be submitted if. 4,g,c l ,

concentration (1) , concentration (2) + * * * > 1.0 limit level (1) limit level (2) -

When radionuclides other than those in Table 3.12.1-2 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 Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This report is not required if the measured level of radioactivity 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 3.12.1-1, in lieu of G any other report required by Specification 4:9-t; prepare and suDmit b i to the Commission within 30 days, pursuant to Specification 4.".2, a .G.G.C Special Report which identifies the cause of the unavailability of samples .and identifies locations for obtaining replacement samples.

The locations from which samples were unavailable may then be deleted from those required by Table 3.12.1-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

~

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

LA SALLE - UNIT 1 3/4 12-1 u

- - -.-___-____m __ m___.____ _.

a RADIOLOGICAL ENVIRONMENTAL MONITORING .

O 3/4.12.2 LAND USE CENSUS (V , ,

I LIMITING CONDITION FOR OPERATION 3.12.2 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 meteor-ological sectors within a distance of five miles. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify the locations of all milk animals in each of tne 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 calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of any other report required by Specification 6.6. A. e 0.0.0. , prepare and I submit to the Commission within 30 days, prusuant to Specifica-tion 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)

(,)

's 20 percent greater than at a location front which samples are currently bein etMxd in accordance with Specification 3.12.1, in lieu of any other report required by Specification 6.6. A. w C.0.2;, prepare [

g g and submit to the Commission within 30 days, pursuant to Specifica-tion 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.

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

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be con' ducted 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, aerial survey, or by consulting local agriculture authorities.

O LA SALLE - UNIT 1 3/4 12-9 1

+

,,O ADMINISTRATIVE CONTROLS .

Ohsite Review and Investigative Function (Continued)

3) ReviewofallproposedchangestotheTechnicalSdeci-fications.

4) Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

5) Investigation of all noncompliance with NRC requirements and shall prepare and forward a report covering evalua-tion and recommendations to prevent recurrence to the t.

Division Vice President-Nuclear Stations and to the j Supervisor of the Offsite Nuclear and Investigative Function.

6) Review of facility operations to detect potential safety hazards.

7) Performance of special reviews and investigations and reports thereon as requested by the Supervisor of the Offsite Review and Investigative Function.

8) Review of the Station Security Plan and shall submit recommended changes to the Division Vice President-Nuclear Stations.

9) Review of the Emergency Plan and station implementing l

procedures and shall submit recommended changes to the Division Vice President-Nuclear Stations.

I

10) Review of reportable ::IE#C"*T3' u- r ree and actions taken to

prevent recurrence.

11) Review of every unplanned onsite release of radioactive material to the environs including the preparation ar.d forwarding of reports covering evaluation, recommendations

' and disposition of tne corrective action to prevent recur-rence to the Division Vice President-Nuclear Stations and to the Supervisor of the Offsite Nuclear Review and Investigative Function.

Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE 12)

DOSE CALCULATION MANUAL, and radwaste treatment systems.

b. Authority The Technical Staff Supervisor is responsible to the Station Superintentent and shall make recommendations in a timely manner in all areas of review, investigation, and quality control phases of plant maintenance, operation, and administra- -

tive procedures relating to facility operations and shall have

- ' the authority to request the action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such action is necessary. The Station Superintendent shall follow such recommendations or select a course of action that isAll more conservative regarding safe operation of the facility. such disagreements shall be reported immediately to the Division Vice President-Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function.

6-9

- LA SALLE - UNIT 1

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ADMINISTRATIVE CONTROLS _

PLANT OPERATING PROCEDURES AND PROGRAMS (Continued)

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

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

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

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

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

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

a. Training of personnel,

b. Procedures for monitoring, and ,

c. Provisi~ons for maintenance of sampling and analysis equipment.

3. Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor cpolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

f

a. Training of personnel,

! b. . Procedures for sampling and analysis, c . .

c. Provisions for maintenance of sampling and analysis equipment.

i

. gy 6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE """"""'"" IN PLANT l OPERATION ,

Any reportable occurrence shall be promptly reported to the 7' vision Vice ,g President - Nuclear Stations or his designat d alternate.

be promptly reviewed pursuant to Specificati 6.1.G.2.a(E

- incident shall ad a separate MG p

pp,l:8 l

report for each reportable occurrence shall b prepared ir # :ordance m with the requirements of Specification 6.6.8.

LA SALLE - llNIT 1 6-18 L

y 9AN b~U ,

TAl58 FCE

,. r' S The following actions shall be taken for REPORTABLE EVENT :

a. The Conunission shill 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 specification 6.1.C.2.a(5).

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ADMINISTRATIVE CONTROLS l

6.4 ACTION T0-BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pursuant to Specification shall not be resumed until authorized by the NRC. The conditions of shutdown shall be promptly reported to the Division Vice President-Nuclear Stations or The incident shall be reviewed pursuant to Specifica- bc N 6 a his designated alternate.tions 6.1.G.I.a and 6.I.G.2.a TheandNRCa Operations separate Center repe ,tifor ELAs eacn T occu i

aga3ar

.fien,sy,bepreparedinaccordancewithySpacificater6.6.8.

c.73 rfhour. shall be notified by telephone as soon as possible and in all cas

([f[ , Safety 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/cr logs relative to the following items shall be kept in a manner convenient for review and shall be retained for at least 5 years:

1.

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

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

3. Records and reports of reportable 3rd refety '4-ft :::erren:::;

4. Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (see Section 4 of theseA specifications) are being met. lance requirements and the corrective ac

5. Records of changes to operating procedures;

6. Shift engineers' logs; and.

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

6-19 LA SALLE - UNIT 1

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J 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 measurementorestimate),

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

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

I and l - 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. v The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

5. Monthly Operating Report Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, ,

shall be submitted on a monthly basis to the Director, Office of Management Information and Program Control, US Nuclear Regulatory Commission, Washington, DC 20555, with a copy of the appropriate Regional Office, to arrive no later than the Qt_h of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition,.a report of any major changes to *

,- the radioactive waste treatment systems shall be submitted with the

- Monthly Operating Report for the period 'in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

B. ";;;nt m ^ r_'erances 'DELETG.D Reportable occurrences, i ciuding corrective act; : and measures to

- prevent recurrence, shall e reported to the NR' V importaitee of an occurrence with respect 'gnificance to safr* .r. general, the determines the immediacy of reporting quired. In some ca. . however, the LA SALLE -' UNIT 1 .

6-24

,[ fb %M, We#wA N LT LGPr % LAM

. s

. ADMhISTRATIVECONTROLS Report le Occurrences (Continued) sign'ficance of an event may not be obvious at the time of its occurrence.

In su cases, the NRC shall be informed promptly of an increased signi-ficanc in the licensee's assessment of the event. In addition, supple-mental ports may be required to fully describe final resolution of the occurren .

In case of corrected or supplemental reports, a licensee event repo t shall be completed and reference shall be made to the original report date 4 1. Prompt tification with Written Followup The types f events listed below shall be reported as expeditiously 2 as possibl but within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 'by telephone and confirmed by telegraph, ilgram, or facsimile transmission to the director of the appropri e regional office or his designate no later than the first working ay following the event, with a written followup report within weeks. The written followup report shall include, as a minimum, a ompleted copy of a licensee event report form.

Information prov ed on the licensee event repcrt form shall be supplemented as n ded by additional narrative material to provide complete explanati of the circumstances surrounding the event.

a. Failure of the eactor protection system or other system subject to limiting safe y system settings to initiate the required .

protective functi n by the time a monitored parameter reaches '

the setpoint spec 1 ied as the limiting safety system setting i'n the technical spec) ications, or failure to complete the required protective function.

Note: Instrument dri t discovered as a result of testing need not be reported under is item but may be reportable under Items B.1.e, B.1.f., or .2.a below.

b. Operation of the unit or ffected systems when any parameter or operation subject to a lim ting condition is less conservative ,

than the least conservative aspect of the limiting condition for operation established in the technical specifications.

I Note: If specified action is aken when a system is found to be l

operating between the most cons rvative and the least corser-vative and the least conservativ aspects of a limiting condi-

)

tion for operation listed in the echnical specifications, the

' - limiting condition for operation i not considered to have been

~ violated and need not be reported u er this item, but it may 6e reportable under Item B.2.b. below.

c. Abnormal degradation discovered in fue cladding, reactor coolant pressure boundary, or primary c ,tainment.

Note: Leakage of valve packing or ga! . witnin the limits for l

identified leakage set forth in technical ecifications need not be reported under this item.

i 6-25 i

LA SALLE - UNIT 1

- , r e .----,-------% ----._w---r,~-cy -,-cw-ev-+v.nm-w,,w --y.... .---w.-.....,e r - - - - -- - , , . . - - , - - ,

M %6 Ir4Ter3T\craAu.d LG; T Ehd v

ADMINISTRATIVE CONTROLS Reportabi currences (Continued)

d. eactivity anomalies, involving disagreement with the predicted lue of reactivity balance under steady-state conditions du ing power operation, greater than or equal to 1% Ak/k; a cal lated reactivity balance indicating a shutdown margin less conse vative than specified in the technical specifications; short- erm reactivity increases that correspond to a reactor period f less than 5 seconds or, if subcritical, an unplanned reactivi insertion of more than 0.5% ak/k or occurrence of any unplan ed criticality.

e. Failure or m function of one or more components which prevents or could preve t, by itself, the fulfillment of the functional requirements o system (s) used to cope with accidents analyzed in the SAR.

f. Personnel error pr edural inadequacy which prevents or could prevent, by itself, he fulfillment of the functional require-ments of systems used to cope with accidents analyzed in the SAR.

Note: for Items B.l.e. d B.l.f., reduced redundancy that

does not result in a loss f system. function need not be reported under this sectio but may be reportable under Items B.2.b. and B.2.c. belo .

g. Conditions arising from natur or manmade events that, as a direct result of the event, re ire plant shutdown, operation t

of safety systems, or other prot ctive measures required by technical specifications.

h. Errors discovered in the transient r accident analyses or in the methods used for such analyses a described in the SAR or the bases for the technical specifica ions that have or could have permitted reactor operation in a nner less conservative than assumed in the analyses.

i. Performance of structures, systems, or co onents that requires remedial action or corrective measures to revent operation in e accident analyses a manner less conservative than assumed in in the SAR or technical specifications bases, or discovery

~

during plant life of condit. ions not specifica ly considered in ,

the SAR or technical specifications.that requi remedial action or corrective measures to prevent the ext tence or

development of an unsafe condition.

Note: This item is intended to provide for -oorting o otential generic problems

~

- o w LA SALLE - UNIT 1 6-26

TWS % E. 19T6moah-( upr h4 e

ADMIN TRATIVE CONTROLS Reporta\.'

b Occurrences (Continued)

j. Offsite release of radioactive materials in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1.

k. ceeding the limits in Specification 3.11.1.4~or 3.11.2.6 for t storage of radioactive materials in the listed tanks. The wr1 ten follow-up report shall include a schuedule and a descrip-tion of activities planned and/or taken to reduce the contents to wi hin the specified limits.

2. Thirty-Day ritten Reports The reportab occurrences discussed below have lesser immediate

! importance th those described under B.1 above. Such events shall be the subject f written reports to the director of the appropriate regional office ithin 30 days of occurrence of the event. The written report sh 11 include, as a minimum, a completed copy of a licensee event rep rt form. Information provided on the licensee event report form s all be supplemented, as needed, by additional narrative material t provide complete explanation of the circumstances surroun ing the event.

a. Reactor protectic system or engineered safety feature instrument v - settings which are ound to be less conservative than those ,

established by the chnical specifications but which do not prevent the fulfills t of the functional requirements of affected systems.

b. Conditions leading to o eration in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition f r operation.

Note: Routine surveillance testing, instrument calibration, or preventative maintenance wh1 h require system configurations as described in Items B.2.a. an B.2.b. need not be reported except where test results the elves reveal a degraded mode as described above.

c. Observed inadequacies in the imp ementation of administrative or procedural controls which thre ten to cause reduction of degree of redundancy provided in r actor protection systems or engineered safety feature systems. .

- d. Abnormal degradation of systems othe than those specified in Item B.1.c above designed to contain dioactive material resulting from the fission process.

- Note: Sealed sources or calibration so ces are not included under this item. Leakage of valve packa ing or gaskets within the limits for identified leakage set for in technical

. specifications need not be reported under his item.

LA SALLE - UNIT 1 6-27 ,

j.

ADMINIS TIVE CONTRCLS N x Thirty-Day Witten Reports (Continued)

e. unplanned offsite release of 1) more than 1 curie of radio-act,ive material in liquid effluents, 2) more than 150 curies of 1

gas in gaseous effluents, or 3) more than 0.05 curies of noble,fodine radio in gaseous effluents.The report of an unplanned offsiteN elease of radioactive material shall include the following\information: \ \

1. A descr on of the event a'nd equipment involved.

2. Cause(s) f the unplanned release.

3. Actions taken' o prevent recurrence.

4. Consequences of he unplanned release.

f. Measured levels of radi' ctivity in an envi ental sampling medium determined to exce the reporting lev values of Tabla 3.12-2 when averaged ver any calendar qu ter sampling period.

C. Unique Reporting Requirements

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

6.7 PROCESS CONTROL PROGRAM (PCP)* l 6.7.1 The PCP shall ue 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 Radioactive I Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:

1. Sufficiently detailed infomation to totally support the rationale for the change without benefit of additional or supplemental information;

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. 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. .

(PCP)* Common to LaSalle Unit 1 and LaSalle Unit 2 l

LA SALLE - UNIT 1 6-28 Amendment No. 18

__._=____.__.____..___.-_

, f"'

s

/ ,.

INDEX DEFINITIONS .

SECTION DEFINITIONS (Continued) PAGE 1.25 OPERA 8LE - OPERABILITY............................................ 1-4 1.26 OPERATIONAL CONDITION - CONDITION................................. 1-4 1.27 PHYSICS TESTS..................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE......................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY..................................... 1-5

1. 30 P ROCESS CONTRO L P R0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5

~ 1. 31 PU RG E - P U RG I NG . . . . . . . . . . . . . . . . . . .'.'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 l

1.32 RATED THERMAL P0WER............................................... 1-5 1.33 REACTOR PROTECTION SYSTEM RESPONSE TIME........................... 1 l 646MT i 1.34 REPORTABLE te99RREMef............................................. 1-6 1.35 R00 DENSITY...l................................................... 1-6 1.36 SECONDARY CONTAINMENT INTEGRITY................................... 1-6 1.37 SHUTDOWN MARGIN................................................... 1-6

1. 38 SO LI D I F I CATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.39 SOURCE CHECK...................................................... 1-7 1.40 STAGGERED' TEST 8 ASIS.............................................. 1-7 1.41 THERMAL P0WER..................................................... 1-7

.' . 1.42 TUR8INE BYPASS RESPONSE TIME...................................... 1-7 1.43 UNIDENTIFIED LEAKAGE.............................................. 1-7 1.44 VENTILATION EXHAUST TREATMENT SYSTEM.............................. 1-7 1.45 VENTING........................................................... 1-7 LA SALLE - UNIT 2 . II

DEFINITIONS

s. e r REPORTABLE GGGUAAENGE '

s esT 1.34 A REPORTA8LE 4GGUARENCE shall be any of those conditions specified in 0:cf*ie:ti::: 5.5.S.1 : d 5.5.8.2. m:tson so,g To s oc g a, Ecn Go .

R00 DENSITY .

1.35 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.36 SECONDARY CONTAINMENT INTEGRITY shall exist when:

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

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

2. Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed 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., walds, bellows or 0-rings, is OPERABLE.

f. The pressure within the secondary containment is less than or equal to the value required by Sp~:ification 4.6.5.1.a.

SHUTOOWN MARGIN f

1.37 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be suberitical assuming all control rods are fully inserted except for the single control red of highest reactivity worth ,

which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

SOLIDIFICATION 1.38 SOLIDIFICATION shall be the conversion of radioact b wastes from liquid systems to a homogeneous (uniformly distributed), - lithic, immobilized solid with definite volume and shape, bounded by a ~~.>v-----.,--,,,-,*,.m,.-.----,,,.-%~,-.-,-~-3.--,,-,rg-.,,-w,,.-n., ,--+-----,,------.g--y ,v,e-~.,--r.,--w.-v-~w-- -r-.~~

~~PLANT SYSTEMS CO, SYSTEMS LIMITING CONDITION FOR OPERATION~~

~~. l. . -~~

~~3.7.5.3 The following low pressure CD systems of Unit 1 &nd Unit 2 shall be OPERA 8LE."-~~

~~a. Division 1 diesel generator 0 room.~~
~~b. Division 2 diesel generator 2A room.~~
~~c. Division 3 diesel generator 28 room.~~
~~d. Unit 1 Division 2 diesel generator 1A room.~~
APPLICABILITY: Whenever equipment protected by the low pressure CD systems ,
is required to be OPERA 8LE.
ACTION:
a. 'With one or more of the above required low pressure CD systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas. establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or, h if: :' rj ethe- -:;:-t r f ;d by 0;;;f *f::th.. 0.C.F. prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the .inoperability and the plans and schedule for restoring the system to OPERABLE status.
~~b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.~~
~~SURVEILLANCE REQUIREMENTS 4.7.5.3- Each of the above required low pressure CDs systems shall be demonstrated OPERABLE:~~
~~a. At least once per 7 days by verifying CO2 storage tank level to be greater than 50% full and pressure to be greater than 290 psig, and~~
~~b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in the correct position.~~
~~c. At least once per 18 months by verifying:~~
~~1. -The system valves and associated motor operated ventilation darpers actuate, manually and automatically upon receipt of a simulated actuation signal, and~~
2. Flow from each nozzle during a " Puff Test "The normal or opergency power source may be inoperable JPERATIONAL CONDITION 4 or 5 or when defueled.
LA SALLE - UNIT 2 3/4 7-17
,~,- .
s PLANT SYSTEMS FIRE HOSE STATIONS .
LIMITING CONDITION FOR OPERATION 3.7.5.4 The fire hose stations of Unit 1 and Unit 2 shown in Table 3.7.5.4-1 shall be OPERA 8LE.
'APPLICA8ILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERA 8LE.
ACTION: -
a. With one or more of the fire hose stations shown in Table 3.7.5.4-1 inoperable, route an additional fire hose of equal or greater diameter to the unprotected area (s)/ zone (s) from an OPERABLE hose station within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoparable fire hose is the primary means of fire suppression; otherwise, route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Restore the inoparable fire hose station (s) to OPERABLE status within 14 days or,3- 11.. vi eny :the- -epeat -equir;d by Speci*ic:ti n S.S.S, prepara and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERA 8LE status.
~~b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
~~SURVEILLANCE REQUIREMENTS 4.7.5.4 Each of the above required fire hose stations shown in Table 3.7.5.4-1 shall be demonstrated OPERABLE:~~
~~a. At least once per 31 days by a visual inspection of the fire hose stations accessible during plant operation to assure all required equipment is at the station.~~
~~b. At least once per 18 months by:~~
~~1. Visual inspection of the fire hose stations not accessible during .~~
plant operation to assure all required equipment is at the station.
.' 2. Removing the hose for inspection and reracking, and
. - 3. Insjecting all gaskets and replacing any degraded gaskets 'in the couplings. -
~~c. At least once per 3 years by partially opening each hose station valve to verify valve OPERA 8ILITY and no flow blockage.~~
. d. Within 5 years and between 5 and 8 years after purchase date and at least every 2 years thereafter by conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above the maximum fire main operating pressure, whichever is greater.
LA SALLE - UNIT 2 , 3/4 7-18
_ _ - - _ - _ _ - - _ _ - _ _ _ l
PLANT SYSTEMS 3/4.7.6 FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.6 All . fire rated assemblies, including walls, floor / ceilings, cable tray enclosures and other fire barriers separating safety-related fire areas or separating portions of redundant syttems important to safe shutdown within a fire area, and all sealing devices in fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable and piping penetration seals and
. ventilation seals) shall be OPERABLE.
APPLICABILITY: At all times.
ACTION:
a. With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either establish a con-tinuous fire watch on.at least one side of the affected assembly (s) and/or device (s) or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly (s) and/or sealing device (s) and establish an hourly fire watch patrol. Restore the inoperable -
fire rated assembly (s) and/or sealing device (s) to OPERABLE status within 7 days or, 4; li:;.; Of :q cth:r h;;;rt -r;;.; ired b Sp.6ifisa-ti:r 8 A B,- prepara and submit a Special Reprit to the Commission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperable fire rated assembly (s) and/or sealing device (s) and plans and schedule for restoring the fire rated assembly (s) and/or sealing device (s) to OPERABLE status.
~~b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
SURVEILLANCE REQUIREMENTS 4.7.6.1- Each of the above required fire rated assemblies and sealing devices shall be verified to be OPERABLE at least once per 18 months by performing a visual inspection of:
. , a. The exposed surfaces of each fire rated assemblies. ,
. ,- b. Each fire window / fire damper and associated hardware.
c. At least 10% of each. type of sealed penetration. If apparent changes in appearance or abnormal degradations are founc. a visual inspection of an additional 10% of each type of maled pe .< . ration shall be made.
.This inspection process shall continue until 3 sample with no apparent changes in appearance or abnormal de.o ation is found.
LA SALLE - UNIT 2 ,
3/4 7-23
, f*'
s PLANT SYSTEMS 3/4.7.7 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.7 The temperature of each area of Unit 1 and Unit 2 shown in Table 3.7.7-1 shall be maintained within the limits indicated in Table 3.7.7-1.
APPLICABILITY: Whenever the equipment in an affected area is required to be OPERA 8LE.
ACTION:
With one or more areas exceeding the temperature limit (s) shown in Table 3.7.7-1:
a. For more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , * " e ef ry n;:-t nwired :;y %;rt-csH aa K 5.B, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 30 days providing a record of the amount by which-and the cumulative time the temperature in the affected area exceeded its limit and an analysis to demonstrate the continued OPEPA8ILITY of the affected equipment.
~~b. By more than 30*F, in addition to the Special Report ret:;uired above, ,~~
within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area to within its tempsrature '
limit or declare the equipment in the affected area inoperable.
SURVEILLANCE REQUIREMENTS 4.7.7 The temperature in each of the above required areas shown in Table 3.7.7-1 shall be determined to be within its limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
~
e LA SALLE - UNIT 2 3/4 7-25 e
, f~
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
e. At least once per 10 years or after any modifications which could affect diesel generator interdependence by' starting diesel gener-ators 0, 2A, and 28 simultaneously, during shutdown, and verifying that all three diesel generators accelerate to 900 rps + 5 -2% in less than or equal to 13 seconds,
~~f. _ At least once per 10 years by:~~
~~1. Draining each fuel oil storage tank, removing ths accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and~~
~~2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection NO, of the ASME Code ir, accordance with ASME Code Section 11, Article IWD-5000.~~
4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall , g be reported to the Commission pursuant to. Specification 6.6.J C 9 eports of diesel generator failures shall include the information recommended in Regula
% 4*
tory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests, on a per nuclear unit basis, is greater than or equal to 7, the report shall be supplemented to include the ,
additional information recommended in Regulatory Position c.3.b of Regulatory -
Guide 1.108, Revision 1, August 1977.
TABLE 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE .
Number of Failures in Last 100 Valid Tests
Test Frequency 5,1 At least once per 31 days 2 At least once per 14 days l
3 At least once per 7 days ,
>4 At least once per 3 days
" Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the last 100 tests are determined on a per nuclear unit basis. With the I exception of the semiannual fast start, no starting time require-ments are ren,uired to meet the valid test requirements of Regulatory Guide 1.108.
LA SALLE - UNIT 2 , , 3/4 8-7 l -- .
l I
RADI0 ACTIVE EFFLUENTS
~
! DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 . The dose or dose commitment to an- individual from radloactive materials in liquid effluents released, from each reactor unit, from the site (see Figure 5.1.1-1) shall be limited:
~~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 mrom to any organ, and~~
~~b. During any calendar year to less than or equal to 3 arem to the total body and to less than or equal to 10 ares to any organ.~~
APPLICABILITY: At all times.
4 ACTION:
a. With 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 Specification 6.6. A, r 5.5.S, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.6.C, a Special Report which iden'tifies 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 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 ares to the total body and 10 arem to any organ. This Special Report shall also include the radiological impact on finished
~~- drinking water supplies at the nearest downstream drinking water source.~~
~~b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
SURVEILLANCE REQUIREMENTS -4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid
^ '
~~effluents shall ne determined in accordance with the ODCM at least once per~~
-31 days.
4 i
LA SALLE - UNIT 2 3/4 11-6
~~.s_.__. : _ ._... ..~~
l l
RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION
-3.11.1.3 ~ 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 affluent from each reactor unit, from the site (see Figure 5.1.1-1), when averaged over 31 days, would exceed 0.06 mres to the total body or 0.2 stem to any organ.
APPLICABILITY: At all times.
ACTION:
a. With the liquid 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 Specification 6.6. A e,~ ".o.6, prepare and submit l to the Commission within 30 days pursuant to 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. SummaryLdescription af action (s) taken to prevent a recurrence.~~
,b . The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per
-31 days, in accordance with the.00CM. '
4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radweste 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.
e 4
O
'LA SALLE'- UNIT 2 ', 3/4 11-7
__ __ ._ .__ .+._ ..
RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site (see Figure 5.1.1-1) shall be limited to the following:
~~a. During any calendar 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.~~
~~! APPLICA8ILITY: At all times.~~
~~ACTION~~

I a. With the calculated air dose from radioactive noble gases in gaseous affluents report required exceeding by Specification any of the C.C. aboveA limits, in#, lieu of any otherprepare [and submit er 6.6.,R to the Commission within 30 days, pursuant to Specification 6.6.C, a Special Report which identifies the cause(s) for exceeding the I- limit (s) and defines the corrective actions to 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,
~~b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once per 31 days.
e e
~
LA SALLE - UNIT 2 3/4 11-13
~~-- --- :e.~~
RADI0 ACTIVE EFFLUENTS DOSE - RADI0 IODINES RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES ,
LIMITING CONDITION FOR OPERATION l
3.11.2.3 The dose to an individual from radiciodines and radioactive materials in particulate form, and radionuclides, other than noble gases, with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site (see Figure 5.1.1-1) shall be limited to the following: .
a. During any calendar quarter: Less than or equal to 7.5 mress to any organ, and .
~~b. During any calendar year: Less than or equal to 15 areas to any organ.~~
APPLICA8ILITY: At all times.
ACTION:
f- .
I a. ..With the calculated dose from the release of radiciodines, radioactive i " materials in particulate form, or radionuclides (other than noble I gases) with half-lives greater than 8 days, in gaseous affluents exceeding any of the above limits, in lieu of any other report required by Specification 6.6.A 4, 0.C.;, prepare and submit to the .l Commission within 30 days, pursuant to Specification 6.6.C. a special Report which identifies the cause(s) for exceeding the Ifmit and defines the corrective' actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure o that subsequent releases will be in compliance with the above limits.
~~b. The provisions of Specifications 3.0.3 and 3.0.4 are not' applicable.~~
SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once per 31 days.
LA SALLE - UNIT 2 3/4 11-14
l t
RADI0 ACTIVE EFFLUENTS GASEOUS RA0 WASTE TREATMENT SYSTEM m
LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATHENT SYSTEM s'ha11 be in operation.
APPLICABILITY: Whenever the main condenser air ejector system is in operation.
ACTION:
a.- With the GASEOUS RA0 WASTE TRF.ATMENT SYSTEM inoperable for more than 7 days, in lieu of any other report required by Specification 6.6. A l a 0.0.0, prepare and submit to the Commission within 30 days, pursuant to 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,~~
~~b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
SURVEILLANCE REQUIREMENTS 4.11.2.4 The GASEQUS RADWASTE TREATMENT SYSTEM shall be verified to be in operatio'n at least once per 7 days.
d 4
~~e l~~
3/4 11-15 -
.LA SALLE - UNIT 2 -
RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM LIMITING CD'?.ITION FOR OPERATION 3.11.2.5 The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERA 8LE 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 (see Figure 5.1.1-1), when averaged over 31 days, would exceed 0.3 mrom to any organ. ,
APPLICABILITY: At all times.
ACTION:
a. With the VENTILATION EXHAUST TREATMENT 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 requirec by Specification 6.6. A r 5.0.0, prepare and submit l to the Commissicn within 30 days, pursuant to Specification 6.6.C, a Special Report wnich includes the following information: .
' "1. Identification of the inoperable equipment or subsystems and .
the reason for inoperability, L ~
~~2. Action (s) taken to restore the inoperable equipment tc OPERABLE status, and~~
~~3. Summary description of action (s) taken to prevent a recurrence.~~
f
. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
'4.11.2.5.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with.the 00CM.
4.11.2.5.2 The VE'NTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the VENTILATION EXHAUST TREATMENT 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.
s G
LA SALLE'- UNIT 2 ', 3/4 11-16 .
RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLIO RADIOACTIVE WASTE LIMITINGCONDITIONFOROPERATIO(
3.11.3 The solid radwaste system shall be OPERABLE and used, as applicable :-
accordance with a PROCESS CONTROL PROGRAM, for the SOLI 0IFICATION and packaging
- of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and
- of 10 CFR Part 71 prior to shipment of radioactive wastes from the site. .
APPLICABILITY: At all times.
ACTION: .
~~a. With the packr.ging requirements of 10 CFR Part 20 and/or 10 CFR Part 71 not satis 1 fed, suspend shipments of defectively packaged solid radioactive wastes from the site.~~
b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other report required by Specification 6.6. A ec C.0.0, l prepare and submit to the Commission within 30 days, pursuant to Specification 6.6.C, a Special Report which includes the followi.ng information:
~~1. Identification of the inoperable equipment or subsystems and the reason for inopersoility,~~
~~2. Action (s)takentoresteretheinoperableequipmenttoOPERABi.E status,~~
~~3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and~~
~~4. Summarydescriptionofaction(s)takentopreventarecurrence.~~
~~c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
i- ,
SURVEILLANCE REQUIREMENTS 1
4.11.3.1 The solid radwaste system shall be demonstrated OPERA 8LE at least
~~once per 92 days by: .~~
a. Operating the. solid radwaste system at least once in the previous 92'. days in accordance with the Process Control Program, or b' . VeritIcation of the existence of a valid contract.for SOLIDIFICATION i to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.
LA SALLE - UNIT 2' 3/4 11-20 -
-w- -y - ,.w- -c , ,,--,.m9a.,~v-,.~y,-w..,ww-ww_ _we,a.r---
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RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERAT70N
. 3.11.4 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 than or equal to 75 mrem) over 12consecutivemonths. - .
APPLICABILITY: At all times.
ACTION:
a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica-tions 3.11.1.2.a 3 3 3.11. 2. 3. b , i n l l e u.11.1. 2. b , 3.11. 2. 2. a . 3.11. 2.6.6.
ification 2. bof A any othe r 5.0.0, prepare and submit, pursuant to Specification 6.6.C, a l Special Report to the Director, Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, Washington D
~30 days,whichdefinesthecorrectiveaction.C.20555,within to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a membs.a of the public from uranium fuel cycle sources (inclu' ding all effluents pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the esti-mated dose (s) exceeds the limits of Specification 3.11.4, 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 $ 190.11. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only and does not apply in any way relates to the limits offor to the requirements 40dose CFRlimitat 190, ion of 10 CFR Part 20, as .
~~- addressed in other sections of this technical specification.~~
~~b. The prov,isions of Specifications 3.0.3 and 3.0.4 are not appifcable.~~
SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents snail ne determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the 00CM.
LA SALLE - UNIf 2 '. 3/4 11-22
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM
,,,_ , _ LIMITING CONDITION FOR OPERATION .- -
3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12.1-1.
APPLICABILITY: At all times.
ACTION:
a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, in lieu of any other report required by Specification 6.6.A,e 5.54, 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 prep.1 ting a recurrence.
b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels in Table 3.12.1-2 when averaged over any calendar quarter, in lieu of any cther report required by Specification 6.6.A p 5.5. , prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a(Report k C& calk pursuant to Specification 4.0...13. When more than one of the racio- '
nuclidesinTable3.12.1-2are}'detectedinthesamplingmedium,this report shall be submitted if: g,gc, ,} .
concentration (1) , concentration (2) + * * * > 1. 0 -
limit level (1) limit level (2) -
When racionuclides other than those in Table 3.12.1-2 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 Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant affluents; however, in such an event, the ccndition shall be reported and described in the Annual Radiological Ervironmental Operating Report.
i
c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12.1-1, in lieu of '
any other report required by Specification 6.6. A sw 0.0.5, prepare
- and submit to the Commission within 30 days, purduant to Specifica-tion 6.6.C a Special Report which identifies the cause of the un-availability of samples and identifies locations for obtaining replacement samples. The locations from which -amples were unavail-able may then be deleted from those required b 'able 3.12.1-1 provided the' locations from which the replacee-- samples were
- obtained are added to the environmental monite , g program as a replacement locations.
~~d. The provisions of Specifications 3.0.3 and 3.0.- are rot applicable.~~
LA SALLE - UNIT 2 3/4 12-1
'.J- . . . .,[ . . .1 .
, . . .: -n. -
5=
i RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS ,
LIMITING CONDITION FOR OPERATION
-3.12.2 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 meteor-4 -
ological sectors within a distance of five miles. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the 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 calculated dose or dose commitment greater than the values currently 4~~
being calculated in Specification 4. H.2.3, in lieu of any other l
report required by Specification 6.6. A. r 5.5.i, prepare and 1 i
submit to the Commission within 30 days, prusuant to Specifica- '
tion 6.6.C., a Special Report which identifies the new location (s).
b. Withalandusecensusidentifyingalocation(s)whichyieldsa calculated dose or dose commitment (via the same exposure pathway) 20 percent armater than at a location from which samples are currently being(obtianed)in accordance with Specification 3.12.1, in lieu of
- f any otner report required by Specification 6.6. A. r 0.0.1, prepare pursuant to Specifica- -
g ,n { and submit to the Commission within 30 daystion 6.6.C., a Special Report wh The new location shall be added to the radiological environmental i 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 j land use census was conducted.
~~c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.~~
i SURVEILLANCE REQUIREMENTS - 4.12.2 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, aerial survey, or by consulting local agriculture authorities. ,
i LA SALLE - UNIT 2 3/4 12-9
ADMINISTRATIVE CONTROLS Onsite Review and Investicative Function (Continued)
~~3) Review of all proposed changes to the Technical Spaci-fications.~~
~~4) Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.~~
5) Investigation of all noncompliance with NRC r4quirements and shall prepara and forward a report cove'.ing evalua-tion and recommendations to prevent recurvence to the Division Vice President-Nuclear Statior.s and to the Supervisor of the Offsite huclear and Investigative Function.
~~6) Review of facility operations to detect potential safety hazards.~~
~~7) Perfomance of special reviews and investigations and -~~
~~reports thereon as requested by the Superviste of the Offsite Review and Investigative Function.~~
~~8) Review of the Station Secarity Plan and shall submit recommended changes to the Division Vice President-Nuclear Stations.~~
~~9) Review of the Emergency Plan and station implementing procedures and shall submit recommended changes to the Division Vice President-Nuclear Stations.~~
~~10) Review of reportable di d ..... and actions taken to prevent recurrence.~~
11) Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recur-
- rence to the Division Vice President-Nuclear Stations and ~
to the Supervisor of the Offsite Nuclear Review and Investigative Function.
~~12) Review of changes to the' PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems,~~
~~b. Authority The Technical Staff Supervisor is responsible to the Static'~~
,- Superintentent and shall make recommendations in a timely manner in all areas of revier, investigation, and quality .
control phases of plant maint4 nance,. operation, and administra-tive procedures relating to facility operations and shall have the authority to request the action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such action is necessary. The Station Scre-intendent shall follow such recommendations or select a course *' .action that is more facility. All such conservative regarding safe operation e' .
disagreements shall be reported immediat
~~to the Division ,~~
Vice President-Nuclear Stations and the .. ervisor of the Offsite Review and Investigative Function.
LA SALLE - UNIT 2 6-9
/"'
e ADMINISTRATIVE CCNTROLS Pt. ANT OPERATING PROCEDURES AND PROGRAMS (Continued)
F. The following programs shall be established, implemented, and maintained:
1. Primary Coolant Sources Outside primary Containment A program to reduce leakage from those portions of systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.
~
The systems include LPCS, HPCS, RHR/LPCI, RCIC, hydrogen recombiner, process sampling, containment monitoring, and standby gas treatment systems. The program shall include the following:
~~a. Preventive maintenance and periodic visual inspection require-ments, and~~
~~b. Integrated leak test requirements for each system at refueling cycle intervals or'less.~~
2. In-Plant Radiation Monitoring-A progrius which will ensure the capability to accurately determine the air)orne iodine concentration in vital areas under 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.~~
4 i 3. Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous affluents, and containment atmosphere samples under accident conditions. The program shall include the following:
i
~~a. Training of personnel,~~
! b. Procedures for sampling and analysis, .
1
, c. Provisions for maintenance of sampling and analysis equipment.
7 VCMT 6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE GGGUARENGE IN PLANT OPERATION '
[ Any reportable occurrence hall be promptly reported to the Division Vice President - Nuclear Station or his designated alternate. The incident
~~hall' be promptly reviewed pursuan to Specification 6.1.G.2.a(5), and a separate MuMI~~
, report for each reportable o urrence shall be prepared in accordance eq,mga j
with the requirements of Spec ication 6.6.8. T> g i .
LA SALLE - UNIT 2 6-18 L_
r- 902 9AM h-O ,
ptg p The following actions shall be taken for REPOETABLE EVENTS:
i
~~s. The Consnission shall be notified and a Licensee Event Report subaltted pursuant to the requirements of S-ection 50.73 to 10 CFR Part 50, and b.~~
Each REPORTABLE EVENT shall be reviewed pursuent to specification 6.1.C.2.a($).
l E
9 e
a O
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ADMINISTRATIVE CONTROLS .
6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pursuant to Specification 2.1.1, 2.1.2 and 2.1.3, and critical reactor operation
- shall not be resumed until authorized by the NRC. The conditions of shutdown shall be promptly reported to the Division Vice President-Nuclear Stations or his designated alternate. The incident shall be reviewed pursuant to Specifica-ib""" I tions 6.1.G.1.a and 6.I.G.2.a and a separate mpoetfor each occurrence sna t i
' M'N The NRC Operations Center y$, r' 50 % *k be prepared in accordance with,S::i'k:the S.S.S.}shallbenotifiedbyteleph ~
, sou1t. Air hour. The Division Vice President-Nuclear Stations and the Director of Nuclear Safety 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. 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; ,
ga art
~~3. Records and reports of reportableg: d ::.... .~~
~~. .. ~ .........,~~
4. Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (see Section 4 of these .
~~specifications) are being met. All equipment,failing to meet surveil-lance requirements and the corrective action taken shall be recorded;~~
~~5. Records of changes to operating procedures; .~~
~~6. Shift engineers' logs; and~~
~~7. Byproduct material inventory records and source leak test results.~~
~~9 LA SALLE - UNIT 2 6-19~~
~~fm l ,~~
l 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 8, Large Quantity),~~
~~and~~
~~f. Solidification agent (e.g. , cement, urea formaldehyde).~~
The radioactive effluent rele$se reports shall include unplanned releases freci the site to unrestricted areas of radioactive materials in gaseous and liquid affluents on a quarterly basis.
The radioactive effluent release reports shall include any changes -
to the PROCESS CONTROL PROGRAM (PCP) made during the reporting
. period.
5. Monthly Operating Report Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Management Information and Program Control, 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 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, g, n ,,n - ne n Reportable occurrences, including co ective actions and measures to prevent recurrence, shall be ccported to the NRC. In general, the importance of an occurrence with respe t to safety significance determines the immediacy of reporting required. I some cases, however, the t
LA SALLE - UNIT 2 6-24
i GIN RATIVE CONTROLS Reportabl Occurrences (Continued) signifi nce of an event may not be obvious at the time of its occurrence.
In such ses, the NRC shall be informed promptly of an increased signi-ficance i the licensee's assessment of the event. In addition, supple-mental rep may be required to ful.ly describe final hsolution of the occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
~~1. Prompt Not fication with Written Followup~~
\
The types of\eva..ts listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confimed by telegraph, mailgram, or facsimile transmission to the director of the appropriate \ regional office or his designate no later than the first working daysfollowing the event, with a written followup report within 2 weeks. The written followup report shall include, as a minimum, a compslated copy of a licensee event report form.
Information provided'on the licensee event report form shall be supplemented as needed\by additional narrative material to provide >
complete explanation of\the circumstances surrounding the event. -
a. Failure of the reach e protection system or other system subject to limiting safety system settings to initiate the required protective function by'the time a monitored parameter reaches thesetpointspecifieda\thelimitingsafetysystemsettingin ,
the technical specifications, or failure to complete the required protective function.
Note: Instrument drift disc .ered as a' result of testing need not be reported under this iten but s may be reportable under Items 8.1.e, 8.1.f., or C.2.a oe' low.
~~b. Operation of the unit or affected' systems when any parameter or operation subject to a limiting condition is less conservative than the least conservative aspect of,the limiting condition~~
- for operation established in the technical specifications.
Note: Ifspecifiedactionistakenwhebsasystemisfoundtobe operating between the most conservative and the least conser-vative and the least conservative aspects of a limiting condi-tion for operation listed in the technical s'pecifications, the limiting condition for operation is nnt considered to have been violated and need not be reported under this ites, but it may be
.,' reportable under Item 8.2.b. .below. ,
~~c. Abnormal degradation discovered in fuel cladding,\ actor e coolant pressure boundary, or primary containment.~~
Note: Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this ites. \
\
> c \
\ ,
LA SALLE - UNIT 2 6-25
__-n2m_=
. l l
, ,'T' ,
l l
e "T46 7Nas iMTG%e,4 Awr LGFr %L b.3 4-ADMINISTRATIVE CONTROLS . ,
1
.i Reportab1k Occurrences (Continued)
~~d. Reactivity anomalies, involving disagreement with the predicted value of reactivity balance ander steady-state conditions~~
~
during power operation, greater than or equal to 1% Ak/k; a alculated reactivity balance indicating a shutdown margin *1ess e servative than specified in the technical specifications; sh t-term reactivity increases that correspond to a reactor per d of less than 5 seconds or, if suberitical, an unplanned react vity insertion of more than 0.5% ak/k or occurrence of any un lanned criticality.
e. Failure malfunction of one or more components which prevents or could p event, by itself, the fulfillment of the functional requirement of system (s) used to cope with accidents analyzed in the SAR.
f. Personnel errer rocedural inadequacy which prevents or could p event, by itse , the fulfillment of the functional require- ,
ments of systems ed to cope with accidents analyzed in the SAR.
Note: for Items B.1. . and 8.1.f. , reduced redundancy that does not result in a 1 s of system function need not be reported under this sec ion but may be reportable under .
Items B.2.b. and 8.2.c. low.
'g . Conditions arising from na tal or manmade events that, as a
' direct result of the event, equire plant shutdown, operation of safety systems, or other p otective measures required by technical specifications.
i h. Errors discovered in the transi t or accident analyses or in the methods used for such analyse as described in the SAR or i,
the bases for the technical speci cations that have or could have permitted reactor operation in manner less conservative than assumed in the analyses.
i.
Performance of structures, systems, or components that requires
. remedial action or corrective measures prevent operation in
- a manner less conservative than assumed the accident analyses in the SAR or technical specifications ba s, or discovery during plant life of conditions not specif ally considered in 4
the SAR or technical specifications that req ire remedial .
l*-
~~action or corrective measures to prevent the istence or~~
-
~~development of an unsafe condition.~~
Nete: This item is intended to provide for reporting f potential ,
generic problems I
e e
LA SALLE - UNIT 2 6-26 e
- - - . - . . - , _ . - . ..,.,.--..-,,,.-,.~,-,,.,.,_-.-m.,m.-~,,,,. -,,,,,,-,,..,,,~m.-- ,y,-m ---,,--- - ~~- w, . , - . . - , . . . -
i T h 7, m e. W T w m osa.u w t Gr-y-pad ADMINISTRAMVE CONTROLS ReportableOc\ currences (Continued)
~~j. Off its release of radioactive materials in liquid and gaseous eff1 nts which exceed the limits of Specification 3.11.1.1 or 3.11.2. . .~~
k. Exceedin the limits in Specification 3.11.1.4 or 3.11.2.6 for the storag of radioactive materials in the listed tanks. The written foi up report shall include a schuedule and a descrip-tion of acti ties planned and/or taken to reduce the contents to within the ecified limits.
~~2. Thirty-Day Written Re rts The reportable occurren s discussed below have lesser immediate importarca thar. those de ribed under B.1 above. Such events shall~~
~~be the subject of written eports to the director of the appropriate regional office within 30 s of occurrence of the event. The~~
written report shall include as a minimum, a completed copy of a licensee event report form. formation provided on the licensee event report form shall be sup lesented, as needed, by additional narrative material to provide c late explanation of the circumstances surrounding the tv t.
a. Reactor protection system or ineered safety. feature instrument settings which are found to be ess conservative than those established by the technical sp ifications but which do not prevent the fulfillment of the f etional requirements of affected systems. ,
~~b. Conditions leading to operation in degraded mode permitted by a limiting condition for operation o plant shutdown required by a limiting condition for operation.~~
~~Note: Routine surveillance testing, i trument calibration, or preventative maintenance which require stem configurations as~~
~~described in Items 8.2.a. and 5.2.b. noe not be reported except where test results themselves reve 1 a degraded mode as described above,~~
c. Observed inadecuacies in the implementation f administrative or procedural controls which threaten to cau:h reduction of degree of redundancy provided in reactor prote tion systems or engineered safety feature systems.
~~d. Abnormal d9 gradation of systems other than those specified in~~
~~Item 8.1.c above designed to contain radioactive Asterial~~
- resulting from the fission process.-
Note: Sealed sources or calibration sources are not ?ncluded under this item. Leakage of valve packag'ng or gaskets within the limits for identified leakage set fo 'n in technic 1 '
specifications need not be reported unct* this item.
I- .
i 6-27 LA SALLE - UNIT 2 s ,. am o m --
'\
ADMINISTRATIVE CONTROLS Thirty-Day We ten Reports (Continuad)
e. unplanned offsite release of 1) more than 1 curie of radio-ac ive material in liquid effluents, 2) more than 150 curies of nob gas in gaseous effluents, or 3) more than 0.05 curie of radioiodine in gaseous effluents. The report of an unplanned offsiteNrelease of radioactive material shall include the followin information:
~~1. A des iption of the event and equipment involved.~~
~~2. Cause(s) for the unplanned release.~~
~~3. Actions ta n to prevsnt recurrence.~~
~
~~4. Consequences f the unplanned release.~~
~~f. Measured levels of raKoactivity in an environmental sampling medium detemined to exceed the reporting level values of Table 3.12-2whenaveragehoveranycalendarquartersampling period.~~
~
C. Unique Reporting Requirements
1. Special Reports shall be submitted to the Director of the Office of Inspection and Enforcement (Region III).within the time period specified for each report. .
6.7 PROCESS CONTROL PROGRAM (PCP P 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 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 infomation;~~
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. 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 1 and La Salle Unit 2. .
LA SALLE - UNIT 2 . 6-28 e
ATTACHMENT D SIGNIFICANT HAZARDS CONSIDERATION 10 CFR 50.73 CHANGES Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a signficant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFP 50.92, operation of LaSalle County Station Unit 1 or Unit 2 in accordance with the proposed AMENDMENT will not:
~~1) Involve a significant increase in the probability or consequences of an accident previously evaluated because these changes conform the Technical Specifications to existing Federal regulations.~~
~~2) Create the possibility of a new or different kind of accident from any accident previously evaluated because this change is administrative only and does not affect any accident analysis.~~
~~3) Involve a significant reduction in the margin of safety because the change has no effect on operation but on reporting of events.~~
Based on the preceding discussion, it is concluded that the proposed system change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased.
Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 "FR 50.92(e),
the proposed change does not constitute a significant hazards consideration.
9785N
~~4+ + IMAGE EVALUATION ((/jj// - [$!!f g,+y+>;,, <>~~
44 l.0 lf22 B ylll HL4 l'l E in EllM L2 1.25 1.4 1.6
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'~
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ATTACHMENT E ,
'}
3' TECHNICAL SPECIFICATION CHANGE REQUEST
. .LASALLE COUNTY STATION UNITS 1 and 2 n D:= CHANNEL-CHECK CHANGES BACKGROUND:
1LaSalle County Station was constructed with Barton differential pressure'. indicating switches to measure reactor vessel level and
~~various systems flows.~~
L
_ LThese switches.are presently installed in the Reactor Protection
' M/ System-(RPS), Primary Containment Isolation System-(PCIS), ECCS and RCIC Actuation Systems. (This attachment provides a
= listing.)
._g-
, These Barton differential pressure indicating switches have-not
~~met the environmental qualification requirements of 10CFR50.49.~~
Therefore to comply with 10CFR50.49 and LaSalle County licenses 4 NPF-ll and NPF-18'(References (e) and (f)), these switches must be replaced.
t
. DISCUSSION: ,
Commonwealth Edison has evaluated the . replacement of these instruments and will. install environmentaly qualified differential pressure switches manufactured by Static-0-Ring, LInc,J(SOR).' These new" switches are blind dP switches and do not have local indication. Technical' Specification Tablec
. l4.3.1.1-1, 4.3.2.1-1,L4.3.3.l'-1, and 4.3.5.1-1 now require shiftly channel' checks on these instrument' channels. During the
. preparation of;the LaSalle County Technical Specifications, required channel checks.were added where indication:was
=available for performing these' checks. 'Due to the requi' red EQ upgrade these channel checks are not possible and must be
~deletedifrom~the Technica1' Specifications.\
.-Itshouldbenotedthdtwh'lethesespecificinstrumentchannels i
lare deleted, in'all cases ~except one other_ instrumentation-from the samefreactor~ vessel reference and variable legs is'stilli required to have, channel checks (Tech Spec' Tables 4.3.4.1-1,
._4;3.7.4-1, 4.3.'7.5-1:and 4.3.8.1-1.)
~
~~These' instruments.will be. replaced on a schedule consistent with Commonwealth Edison's commitments to the.NRC with regards to 10CFR50.49'and the references listed above. These instrument~~
~
~~replacements will-be completed for both unit:1 and asby November 30,:1985. -The present requirement for 'ch'annel checks will- be performed-on allLinstrument channels where required until each~~
~~.instrumentiis replaced regardless of any1 Tech Spec requirements~~
~~when: amended.-~~
l.__r:
. m
.i
(
TA8LE 4.3.1.1-1 4
r- REACTOR P_ROTECTION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL p
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH l g CHANNEL CHECK TEST CALIBRATION (,) SURVEILLANCE REQUIRED
, FUNCTIONAL UNIT l
$ 1. Intermediate Range Monitors S/U(b) 5 S/U(c) ,W Neutron Flux - High R 2 H a. ,
3,4,5 i - S W k i
Inoperative NA W NA 2,3,4,5 /
b.
l II)
~~2. Average Power Range Monitor:~~
Neutron Flux - High, S/U(b) 5 S/U k') ,W
! a. 1, 2 SA Setdown ,
3, 5 i S W SA 2
R b. W(d)(e),SA,R(h) j Flow Biased Simulated ThermalI9) S/U IC) i
' Power-Upscale 5, D ,W Y c. Fixed Neutron Flux -
S/U IC) ,W W(d) , SA 1
1 N High 5 l NA 1, 2, 3, 5 >
i d. Inoperative NA W ,
i .
e
~~e i 3. Reactor Vessel Steam Dome 1, 2 5~~~
! Pressure - High NA M Q I
i
~~4. Reactor Vessel Water Level - 8/~~
~~1, 2 4"g Low, Level 3 M R~~
~~5. Main Steam Line Isolation 1 3~~
M R l;
Valve - Closure NA
! 6. Main Steam Line Radiation - 1, 2 M R 1 High S
7. Primary Containment Pressure - 1, 2 j High NA M Q*
l l
, 'N .
, 2 9 TABLE 4.3.2.1-1 ,
h ISOLATION ACTUATION INSTRUENTATION SURVEILLANCE REQUIREENTS E CHANNEL OPERATIONAL
' CHANNEL FUNCTIONAL- CHAIGGEL ColeITIONS FOR WHICH CALIBRATION SURVEILLANCE REQUIRED g TRIP FUNCTION CHECK TEST A. AUTOMATIC INITIATION
~~1. PRIMARY CONTAll#ENT ISOLATION~~
~~a. Reactor Vessel Water Level~~
~~1) Low, Level 3 fMA M R 1,2,3~~
~~2) Low Low, Level 2 ##A M R 1, 2, 3 1, 2, 3~~
~~b. Drywell Pressure - High - NA M Q~~
~~c. Main Steam Line~~
~~1) Radiation - High S M R 1,2,3~~
~
M 1
~~2) Pressure - Low NA Q 1, 2, 3 d.~~
3) Flow - High Main Steam Line Tunnel fMA M R l Temperature - High MA M R 1, 2 3
~~e. Condenser Vacuum - Low M 1, 2g 38 g NA Q ,~~
~
= f. Main Steam Line Tunnel 1, 2, 3 p A Temperature - High - NA M R E! 2. SECONDARY CONTAllelENT ISOLATION
a. Reactor Building Vent Exhaust Plenum Radiation - High S M R 1, 2, 3 and **
~~b. Drywell Pressure - High NA M Q 1, 2, 3~~
c. Reactor Vessel Water g 1, 2, 3, and Level - Low Low, Level 2 gNA M R l
d. Fuel Pool Vent Exhaust Radiation - High S M R 1, 2, 3 and **
~~3. REACTOR WATER CLEANUP SYSTEM ISOLATION ,~~
~~A Flow - High M R 1, 2, 3~~
~~a. S~~
~~b. Heat Exchanger Area Temperature - High NA M Q 1,2,3~~
~~c. Heat Exchanger Area Ventilation AT - High NA M Q 1, 2, 3~~
~~d. Pump Area Temperature - High NA M Q 1,2,3~~
~~e. Pump Area Ventilation - A Temperature - High NA M Q 1,2,3~~
~~f. SLCS Initiation MA R NA 1,2,3~~
~~g. Reactor Vessel Water Level - Low Low, Level 2 JNA M R 1,2,3 l k~~
i i .
l TABLE 4.3.2.1-1 (Continued)
! 9
! y. ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS i # OPERATIONAL l;; CHANNEL i
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH l TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED c
i 3 i 6. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION i e
' a. Reactor Vessel Water Level -
Low, Level 3 /A/A M R 1, 2, 3 l l b. Reactor Vessel j (RHR Cut-in Permissive) i Pressure - High NA M Q 1,2,3 1 c. RHR Pump Suction Flow - High NA M Q 1,2,3
d. RHR Area Temperature - High NA M Q 1,2,3 j e. RHR Equipment Area AT - High NA M Q 1,2,3 I
! R B. MANUAL INITIATION i Y 1. Inboard Valves MA R NA 1,2,3 y 2. Outboard Valves 1, 2, 3 1
! 3. Inboard Valves 1, 2, 3 and **,#
~~4. Outboard Valves 1, 2, 3 and **,#~~
~~5. Inboard Valves 1, 2, 3~~
~~6. Outboard Valves 1, 2, 3~~
~~7. Outboard Valve 1,2,3 i,~~
~~When reactor steam pressure > 1043 psig and/or any turbine stop valve is open.~~
When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations j with a potential for draining tha. reactor vessel.
l #0uring CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
!, ( ( .
(.
g.'
5 taste 4.3.3.1-1 C
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREE NTS
{
8 CHANNEL OPERATIONAL CH4NNEL FUNCTIONAL CHAIRGEL C00SITIONS FOR WHICH E CHECK TEST CALIBRATION SURVEILLAllCE REQUIRED y TRIP FUNCTIO 18 A. DIVISI0ll I TRIP SYSTEM
~~1. RHR-A (LPCI MDOE) AND LPCS SYSTEM~~
a. Reactor Vessel Water Level - 1, 2, 3, 4*, 58 gNA Low Low Low, Level 1 N R l M 1, 2, 3
b. Drywell Pressure - High NA Q 1, 2, 3, 4*, 5*
~~c. LPCS Pump Rischarge Flow-Low NA M Q~~
d. LPCE and LPCI A Injection Valve Injection Line Pressure low 1, 2, 3, 4* , 5*
~~M R Interlock NA~~
e. LCPS and LPCI A Injection Valve IIA M R 1, 2, 3, 4*, 5*
m Reactor Pressure Low Interlock
';;; f. LPCI Pump A Start Time Delay MA M Q 1, 2, 3, 4*, 5*
. Relay M Q 1, 2, 3, 4*, 5" 2, g. LPCI Pump A Flow-Low NA N h. Manual Initiation NA R . NA , 7, 2, 3, 4* , 5*
~~2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#~~
~~a. Reactor Vessel Water Level -~~
~~Low Low Low, Level 1 fNo M R 1, 2, 3 l Drywell Pressure-High NA M Q 1,2,3~~
~~b. 1, 2, 3~~
~~c. ADS Timer NA M Q~~
d. Reactor Vessel Water Level - 1, 2, 3 M R y Low, Level 3 fNA g e. LPCS Pump Discharge M 1,2,3 g Pressure-High NA Q g f. LPCI Pump A Discharge M 1, 2, 3
<+ Pressure-High NA Q NA 1, 2, 3 1 g. Manual Initiation MA R N
l 4
_ ________l
. .~
~~s.*-~~
l
. TABLE 4.3.3.1-1 (Continued)
}
G EMSGENCY CORE COOLING SYSTEM ACTUATION INSTRUIENTATION SURVEILLANCE REQUIREENTS -
g
, E CHAISIEL OPERATIONAL i CHANNEL FUNCTIONAL CHAIRGEL COISITIONS FOR 145tCH
! TRIP FUNCTION CHECK TEST CALIBRATICII SURVEILLANCE REQUIRED 1 E j 4 3. DIVISION 2 TRIP SYSTEM i
H 1 SHR S AIS C (LPCI MODE)
, a. Reactor Vessel Water Level -
i Low Low Low, Level 1 g#A M - R 1, 2, 3, 4*, 5*
l b. Drywell Pressure - High IIA M Q 1, 2, 3 1
c. LPCI B and C Injection Valve Injection Line Pressure s Low Interlock IIA M R 1, 2, 3, 48, 5* i
~~d. LPCI Pump B Start Tlee Delay~~
! Y* Relay IIA M Q 1, 2, 3, 48, 5*
l e. LPCI Pump Discharge Flow-Low NA M Q 1, 2, 3, 4* , 5*
j Y f. Manuel Initiation IIA R NA 1, 2, 3, 48, 5* !
U g. LPCI B and C Injection Valve -
Reactor Pressure Low Interlock NA M R . 1, 2, 3, 4*, 58
2. AUT0ftATIC DEPRESSURIZATIO10 SYSTEM TRIP SYSTEM B J
< a. Reactor Vessel Water Level -
Low Low Low, Level 1 fNA M R 1, 2, 3 .
b. Drywell Pressure-High 10 4 M Q 1, 2, 3 i y c. ADS Timer 80 4 M Q 1, 2, 3 j 5 d. Reactor Vessel Water Level - '
! Low, Level 3 JWA M R 1, 2, 3 ;
l e. LPCI Pump 8 and C Discharge j ?+ Pressure-High NA M Q 1, 2, 3 I , f. Manual, Initiation MA R NA 1, 2, 3 i t
i i
f e
i
TABLE 4.3.3.1-1 (Continued)
I r.
2 .
l E
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS r-CHANNEL OPERATIONAL
[ CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH
$ CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 1
-8 TRIP FUNCTION l C. DIVISION 3 TRIP SYSTEM j 1. HPCS SYSTEM
a. Reactor Vessel Water Level - 1, 2, 3, 4 *, 5*
i l
Low Low, Level 2 MO M R 1, 2, 3
~~b. Drywell Pressure-High NA M Q j.~~
c. Reactor Vessel Water Level-High h M R 1, 2, 3, 4*, 5*
Level 8 JVA/A i d. Condensate Storage Tank Level - 1, 2, 3, 4*, 5*
Low MA M Q 4 u e. Suppression Pool Water 1, 2, 3, 4*, 5*
A Level - High NA M Q 1, 2, 3, 4* , 5*
w Pump Discharge Pressure-High NA M Q i f.
M Q 1, 2, 3, 4*, 5*
I O
g. HPCS System Flow Rate-Low NA NA 1, 2, 3, 4*, 5*
Manual Initiation NA R h.
I D. LOSS OF POWER NA R 1, 2, 3, 4**, 5**
~~1. 4.16 kw Emergency Bus Under- NA I~~
voltage (Loss of Voltage)
! # Not required to be OPERABLE when reactor steam done pressure is less than or equal to 122 psig.
~~When the system is required to be OPERA 8tE after being manually realigned, as applicable, per Specification 3.5.2.~~
~~Required when ESF equipment is required to be OPERA 8LE.~~
i i
( ( (
l
(
TABLE 4.3.5.1-1 E
y, REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUNENTATION SURVEIL E CHANNEL E CHANNEL FUNCTIONAL CHANNEL
. TEST CALIBRATION FUNCTIONAL UNITS CHECK c
3
~~a. Reactor Vessel Water Level - R~~
~~" Low Low, Level 2 /NA M~~
~~b. Reactor Vessel Water /NA M R Level - High, Level 8 MA R NA~~
~~c. Manual Initiation R.~~
Y
TABLE 4.3.1.1-1 ,
REACTOR PROTECTION SYSTEM INSTRteENTATION SURVEILLANCE REWIREBENTS .
E OPERATIONAL !
E CHNSIEL FUNCTIONAL UIIIT CHMAIEL CIECK FUNCTIONAL-TEST CHANNE'.
CALIRRATI00t*I COISITIONS FOR tRGICH I SURVEILLANCE REQUIRE 0 l, -
,i 3 l U 1. Intermodlate Range Monitors
re a. Neutron Flux - High SM(b} ,5 S/uIC) W R 2 5 W R 3,4,5 i h. Insperative IIR W 10 4 2,3,4,5 ,
~~2. Average Power Range leoniter: .~~
~~a. Iteutron Flux - High, ' '~~
setdown s/u(b) s s/u(*), v st 1, 2
. s v -
u
~~3, 5 wid )(*),~~
~~b. SA, R(h)~~
Flow Riesed Steulated Thermal s, o (s) s/u(*),w j l- w i c.
Power-upscale Flued Blowtren Flux -
i i w High s- s/u(*),v w(d), SA 1 .
Insperative IIA W IIA 1. ,2, 3, 5 i
~~d. ,~~
~~i 1~~
~~3. Reacter Vessel Steam Dome Pressure - High IIA M q 1, 2 l~~
I j 4. . Reacter Wessel Water Level - .
1, 2 l ,
1 Low, Level 3 f)/O M R ,
I
5. Main Steam Line Isolatten 1 Valve - Closure IIA M R t
6. Main Steam Line Radiatten - 1, 2 High 5 M R l 7. Primary Centainment Pressure -
High IIA N q 1, 2 s
i 4
i
~
~~TABLE 4.3.2.1-1 g ISOLATION ACTUATION INSTRISENTATION SURVEILLANCE REQUIRE 8ENTS OPERATIONAL CHAlelEL~~
~~g CHAIRIEL FUNCTIONAL CHAIRIEL ColelTI00tS FOR tRtICH ,~~
~~l r-CHECK TEST CALIRRATI001 SURVEILLABICE. REQUIRED j m TRIP FulICTICII~~
~~A i E . AUTOIIATIC 1.~~
~~INITIAT10It PRIIIARY CGIITAlteqENT ISOLATION l q "~~
~~e. a. Reacter Wessel niater Level i . 1) Law. Level 3 .~~
JNA . M R I . 2. 3 *
~~1. 2, 3~~
~~2) Law Law. Level 2 JNA M R .~~
i
~~b. Drywell Pressure - High .~~
~~IIA ' M q 1. 2. 3'~~
~~c. IInin Steam Line~~
1) Radiatten - High 5 M R 1. 2. 3 l N 1.
~~2) Pressure - Law IIA Q J~~
~~3) Flow - High JA/O M R 1. 2. 3 g~~
~~d. IInin Steam Line Tunnel Temperature - High NA N R 1. 2 3 l~~
' HA M Q 1. 2g 3" w e. Camdenser Vacuum - Law j i f. Main Steam Line T' wel 1, 2. 3 w A Temperature - Mgh IIA ,M R E$ 2. SEC008ARY CGIITAIISIENT ISOLATIcIl J * -
j a. Reacter Rullding Vent Exhaust 1; 2. 3 and **
Plenum Radiation - High 5 N A i
1 b. Dryuell Pressure - High IIA M Q 1. 2. 3
~~c. Reacter Wessel tfater .~~
g I Level - Law Law. Level 2 gNA N R 1. 2. 3. and (
I d. Fuel Peel Vent Exhaust
! Radiatten - High 5 N R ,
1. 2. 3 and **
~~j 3. REACTOR tmATER CLEAIIUP SYSTEM ISOLATI006 i a. A Flow - High 5 M R 1. 2. '3~~
~~h. Heat Exchanger Area .~~
Temperature - High NA M Q 1. 2. 3
]
~~c. Heat Exchanger Area . ,~~
~~Ventilatten AT - High IIA ' M . Q 1. 2. 3 l~~
~~d. Pump Area Temperature - High NA ' M- Q 1. 2. 3 i~~
~~e. Pump Area Ventilatten - A l Temperature - High NA M Q 1. 2. 3 3~~
IIA R IIA 1. 2.; 3
} f. SLCS Initiatten ,
i g. Reacter Vessel niater Level - Lew Law. Level 2 //h} M R 1. '2. 3 ' l l
)
TABLE 4.3.2.1-1 (Continued) .
1 5 IS0 TAT 1001 ACTUATICII INSTRIIENTATICII SURVEILLANCE REWIREIIENTS C OPERATIONAL i E CNAIGICL CHAleIEL COISITICIIS FOR 1911CN CHAISIEL FUNCTIGIIAL
~~SURVEILLA14CE RE0tlIRE0 !~~
TRIP FMIICTICII CHECK TEST CAtlRRAT1011 '
i ,
E -
M 6. RNR SYSTEM SatiThetal C00tilIG IWSE ISOLATICII , ,
l n -
~~a. Seacter Wessel niater Level - .~~
~~Law. Level 3 f#A N- E I. 2. 3 l .~~
~~b. Reacter Wessel (RIIR Cut-in Permissive) -~~
Pressure - High IIA . . M q 1.23 I' c. RIIR Pump Section Flow - Nigh IIA M q . 1. 2. 3,
~~d. NR Area Temperature - High IIA M Q l. 2. 3 M 1, 2. 3~~
~~e. NIR Equipment Area AT - Nigh NA Q l~~
w 5. NAltlAL IIIITIAT1011 j! 1 -
IIA R M4 1. 2. 3 1! w 1. Inboard Valves lj e'. 2. Outboard Valves 1. 2. 3
~~3. Inbeard Valves .~~
~~- 1. 2. 3.and **.#~~
~~l. 2. 3 and~~
~~4. Outboard Valves 1. 2, 3 I~~
i
~~5. Inbeard Valves~~
~~6. Dutboard Valves~~
~~1. 2. 3 ;~~
~~7. Outboard Valve - 1. 2. 3 l l~~
l heten reacter steam pressure > 1943 psig and/or any turbine step valve is open. .
l
~~nAten hand 11ag irradiated fuel la the secondary contalanent and during CORE ALTERATI0lIS and operations~~
.J with a potential for draining the reacter vessel.
) S0uring C2RE ALTERAT10115 and operations with a potential for draining the reactor vessel.
4 l
4 I
i . .
) -
~~al '~~
3 ss s s sss gEl .
c c c c ccc ce -
.I ] e., i, -
aaa a a aaa a a aaa aaa a a aa l l]=l l .:.:.: .: .: .:. .: .:.:.: : .: .:.:
i
! ls!s .
~
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lla
" = = z.
~
a ss. ===
d
< g* === s g .
. s..
y ,
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l ,
s a -
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g li 18 3 83 ! l L1 1 1 L . .! ! 'l' l I a i sir #g N! !
s ,si!3 3: =
31,-
s i l
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~~Ly ci a 3 j s' s y 3-Iss=!nj1.lf= 3= <ai 3 3~~
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LA SALLC - UNIT 2 3/4 3-32
- - _ - - - - _ . . _ . .__.__._._,_.._.._,,-...__..._r.,._,._,_.m_, , . . . _ ,___. _ __
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s 1l. .:c. c ccc c l
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nas a E Kas1.na
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n 3 -
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5 3a 3-3 - - .
5 g [
s.I {.-.nj s -
.,s{_ i !
I 35")a"is s
~~a. :-~~
s
~~3 o,I. .m.o~~
. g .E 3
31 l 35I- se E E
- lel1 3 3
3=}j3gjl3
~~g. 3.j3}.-}gigjg a g--g a jlj;gj . ,~~
s i
.,. x , ,a ..a t s ia. ,c E s; a LA SAll.E - UNIT 2 3/4-33
-..,,----,----.,_,,.-_-,.-,.-,,w----ween-_w, , _ , - - _,_ - ,-,,_ _ e.wna.,.,,e.,.,wc., ..,,mm.,,-w,,mwa,__ m-en-ar-e--ww.-w . , , < -
e
~.
TABLE 4.3.3.1-1 (Continued)
~
g EBERGENCY COM COOL *IS SYSTEM ACTUATIGII INSTAleENTATICII SISWEILNF AEGUIMIElff5 k ~
CaudeIEL OPERATleIIAL CCISITIGIIS Fee teIICII
~~CNAleIEL FIMCTIonAL CNAleEL SURWILLAIICE GEGUIRES g TRIP FWICTIGII CatCK TEST CAtleAATICII "e~~
C. BIVISIGII 3 TRIP SYSTEN
. 1.* IWC5 SYSTERI .
~~a. Reacter Wessel tinter Level -~~
1, 2, 3, 4*, 5*
Law Law, Level 2 y# II -
R g s
~~b. Bryuell Pressure-Ni$ N4 N 4 1, 2, 3 .~~
~~c. Reacter Wessel % Inter Level-Iti$ - !~~
~~Level 8 fa/A N R - 1, 2. 3. 4*, 5*- l~~
~~d. Camdensate Storage Tank Level -~~
Law M4 . N 4 1, 2, 3, 48, 5* l 8
t:a e. Segpression Peel % Inter .
~~Level - Nigh 11 4 N q 1, 2, 3, 48, 5" :~~
f. Pimp Bischarge Pressure-Illgh IIA 'N 4 1, 2, 3, 48, 5*
~~y g. NPCS System Flow Rate-Law NA . N 4 1, 2, 3, 4*, 58~~
~~h. 31onnel Initiatten IIA . R IIA 1, 2, 3, 48, 5" g I~~
~~a. toss oF PCIER . . ,~~
I
~~1. 4.16 kV b .;: ry Bus Neder- IIA IIA R 1, 2, 3,M**, 5** -~~
weltage (Less of Vs,1tage)
.m
2. 4.16 kV i- :;:- y aus-under- na NA R 1, 2, 3, 4**, 5**
weitage (segraded Weltage) -
(Blwisten 3) .
TABLE ISTATICIIS mint required to be OPERAALE when reacter steam dame pressure is less than er equal to 122 psig.
~~tfhen the system is required to be SPERASLE after helag mammally reallgood, as appilcable, per Specificatten 3.5.2.~~
~~Seguired uten ESF equipment is required to be SPERAGLE.~~
~~l . .~~
, 1, - -
I ~
3 .
TAtti 4.3.5.1-1
~~E - SEacies CaeE 150tATEGII C80 LING SYSTEM ACTIIATIGIl INSIBISENTATIGIS SMOVEltLAIICE REGIIISE8ENTS~~
% CalAfgEL E .
DIAleEL FuutileIIAL CIIAleEL e
FMIICTlellhL IAIITS CBE M TEST Call 8AAil8II i
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O ATTACHMENT G SIGNIFICANT HAZARDS CONSIDERATION CHANNEL CHECK CHANGES Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the critoria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units 1 and 2 in accordance with the proposed amendment will not
~~1) Involve a significant increase in the probability or consequences of an accident previously evaluated because this change only removes the channel check requirements.~~
~~Channel functional testing and calibrations are still periodically required to ensure system availability as necessary. Single failure critoria is not affected by this revision.~~
2) Create the possibility of a new or different kind of accident from any accldent previously evaluated because failure of these instruments is evaluated and no now accident is postulated from removing the channel check requiremont.
~~3) Involvo a significant reduction in the margin of safety because the availability of safety related systems is not significantly affected.~~
Dased on the preceding discussion, it is concluded that the proposed system change clearly falls within all acceptable critoria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased. Therefore, based on the guidance provided in the Federal Register and the critoria established in 10 CFR 50.92(e), the proposed chango does not constituto a significant hazards consideration.
9785N u
T ATTACtNENT E
- Pagc 2 7 AFFECTED INSTRUMENTS CHANNEL CFECK CHANGES INSTRlNENT NLNBER SYSTEM (s) TSCH SPEC AFFECTED TABLES 1(2)B21-N026A,0,C,D PCIS (Groups 1 through 5) 4.3.2.1.-l item 1.a.2) level 2 1(2)B21-N024A,0,C,0 RPS,PCIS (Groups 6 and 7) 4.3.1.1-1 item 4 level 3 4.3.2.1-1 item 1.a.1) 1(2)E31-N008A,0,C,0 PCIS (Group 1) Main Steam 4.3.2.1-1 item 1.c.3)
Line Flow 1(2)E31-N009A,0,C,0 PCIS (Group 1) Main Steam 4.3.2.1-1 item 1.c.3)
Line Flow 1(2)E31-N010A,B,C,0 PCIS (Group 1) Main Steam 4.3.2.1-1 item 1.c.3)
Line Flow '
1(2)E31-POllA,0,C,0 PCIS (Group 1) Main Steam 4.3.2.1-1 item 1.c.3)
Line Flow 1(2)D21-f037A,C ADS (Div 1) (level 1) 4.3.3.1-1 item A.2.0 LPCS, LPCI A (level 1) 4.3.3.1-1 item A.1.0 RCIC (level 2) 4.3.5.1-1 item a 1(2)D21-f0370,C ADS (Div 2) (level 1) 4.3.3.1-1 item 0.2.a U)CI B & C (level 1) 4.3.3.1-1 item 0.1.a RCIC (lovel 2) 4.3.5.1-1 item a 1(2)B21-N101A,0 RCIC (level 8) 4.3.5.1-1 item b 1(2)D21-f030A ADS (lovel 3) 4.3.3.1-1 item A.2.d 1(2)021-t0300 ADS (lcycl 3) 4.3.3.1-1 item 0.2.d 1(2)D21-N100A,0 HPCS (lovel 8) 4.3.3.1-1 item C.l.c 1(2)021-NO31A,0,C,0 HPCS (level 2) 4.3.3.1-1 item C.l.a 9765N
p Commonwealth Edison O one First National Ptara. Chicago. litinois Address Reply to. Post Othee Box 767 Chicago, lihr.ois 60690 February 14, 1985 Director of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555
~~Subject:~~
~~Dresden Nuclear Power Station Use of the Exxon Nuclear Company Inc. RA-3 Fuel Shipping Containers NRC Docket Nos. 50-010/237 and 249~~
~~Dear Sirs:~~
Pursuant to the requirements of 10 CFR 71.12, Dresden Station, as holder of License Nos. OPR-2, 19, and 25, wishes to register as user of the Exxon Nuclear Co., Inc. fuel shipping containers. Package identification number for this container is USA /4986/AF, NRC Docket Number 71-4986. Your certificate of Compliance Number is 4986, Rev. 16, which expires March 31, 1987.
Very truly yours,
- gh D. Ryba Nuclear Licensing A ministrator im cc: R. Gilbert - NR [
NRC Resident Inspector - Dresden 9764N

ML20107M575

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Subject:

Dear Sirs:

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