Text

Resubmitted Proposed Application to Amend License NPF-11 Changing Tech Specs Reflecting Unit 1 Changes Incorporated Into Unit 2 Tech Specs
	ML20087N316
Person / Time
Site:	LaSalle
Issue date:	03/22/1984
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20087N310	List: ML20087N309; ML20087N316;
References
	NUDOCS 8404030331
	Download: ML20087N316 (200)
	v • d • e

- . . - _- .

(

INDEX

, s DEFINITIONS SECTION-

~

OEFINITIONS (Continued) , PAGE 1.25 OPERA 8LE - OPERA 8ILITY............................................ 1-4 1.26 OPERATIONAL CONDITION - CONDITION................................. 1-4 1.27 PHYSICS TEST 5..................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE......................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY..................................... 1-5

1. 30 PROCESS CONTRO L PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.31 PURGE - PURGING........................'..........................

. 1-5 1.32 RATED THERNAL P0WER......................:........................ 1-5 1.33 REACTOR PROTECTION SYSTEM RESPONSE TIME........................... 1-5

\g 1.34 REPORTABLE OCCURRENCE.............................................

1-6 1.35 R00 0ENSITY....................................................... 1-6 1.36 SECONDARY CONTAINMENT INTEGRITY................................... 1-6

. 1.37 SHUTDOWN MARGIN................................................... 1-6 1.38 SOLIDIFICATION...........................,........................ 1-6 l

' i i

1.39 SOURCE CHECX...................................................... 1-7 1.40 STAGGERED TEST 8 ASIS.............................................. 1-7

. 1.41 THERNAL PCWER..................................................... 1-7

, , 1.42 TUR8INE BYPASS RESPONSE TIME.......................:..............

1-7 l

Ii 1.43 UNIDENTIFIED LEAKAGE..............................................

1-7

,: 1.44 VENTILATICN EXHAUST TREATMENT SYSTEF.............................. 1-7

. 1.45 VENTING........................................................... 1-7

. r1=, e - ,

w-u: e u - :ceyen: v .:,- - .

. aw -n

. .. . .,,r.oess-rnum. r...-.r,

....e

, ., n e , nue i..-

f LA SALLE - UNIT 1 II 8404030331 e40322 hDRADOCK 05000373

. m______. . - - -- - - -

INDEX f

'._,. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 CORE STANCBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Service Water System................... 3/4 7-1 Diesel Generator Cooling Water System. . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-2 Ul ti mate He a t S i n k. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-3 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM -

EMERGENCY FILTRATION SYSTEM.................................. 3/4 7-4 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM............I.......... 3/4 7-7 3/4.7.4 SEALED SOURCE CONTAMINATION.................................. 3/4 7-9 .

3/4.7.5 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System................................ 3/4 7-11 Del uge and/or Sprinkl e r Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-14 CO 2

Systems.................................................. 3/4 7-17 Fire Hose Stations...........'................................ 3/4 7-18 3/4.7.6 FIRE RATED ASSEMBLIES........................................ 3/4 7-22

3/4.7.7 AREA TEMPERATURE MONITORING.................................. 3/4 7-24 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS I STRUCTURES................... 3/4 7-25 3/4.7.9 SNUBBERS...................................................... 3/4 7-27 '

I

3/4.7.10 MAIN TURBINE BYPASS SYSTEM.................................... 3/47g-i 33 t

I l

l LA SALLE - UNIT 1 -

VIII 1

1 1

--- -e --

-. ---- ,, , . - - - - - , - , , . - - - - , - - - , - - , , , - - , ---,,,,,.,,----,,,,,w- - - - ,

B

\

i m

s

.L R

(/ BASES SECTION PAGE 3/4.7 PLANT $YSTEMS ,

a

' 3/4.7.1 CORE STAN08Y COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS......................................... 8 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM................ 8 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM................... B 3/4 7-1 3/4.7.4 SEALED SOURCE CONTAMINATION........'..................... B 3/4 7-2 3/4.7.5 FIRE SUPPRESSION SYSTEMS................................ 8 3/4 7-2 3/4.1.6 FIRE RATED,ASSEMSLIES...................................

B 3/4 7-3 3/4.7.7 AREA TEMPERATURE MONITORING............................. B 3/4 7-3 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS I STRUCTURES. . . . . . . . . . . . . . 8 3/4 7-3 3/4.7.9 SNU88ERS.....'..........................................

1 83/47j(5l

~

3/4. 7.10 MAIN TUR8INE BYPASS SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 . 3/4 7-5 f ; 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE PCWER i

OISTRISUTION SYSTEMS.................................... 8 3/4 8-1 .

3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES. . . . . . . . . . . . . . . . . 8 3/4 8-2 -

3/4.9 REFUELING CPERATICNS 3/4.9.1 REACTOR MODE SWITCH..................................... 8 3/4 9-1 3/4.9.2 INSTRUMEN TA TION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 9-1 l ; 3/4.9.3 CONTROL R00 POSITI0N....................................

8 3/4 9-1

} 3/4.9.4 DECAY TIME..............................................

8 3/4 9-1 ll 3/4.9.5 COMMUNICATIONS.......................................... 8 3/4 9-1 l 3/4.9.6 CRANE AND H0!ST.........................................

8 3/4 9-1 i 3/4. 9. 7 CRANE TRAVEL............................................

1 8 3/4 9-2

.i 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL L!

i4 and WATER LEVEL - SPENT FUEL STORAGE P00L............... 8 3/4 9-2 3/4.9.10 CONTROL R00 REM 0 VAL.....................................

8 3/4 9-2 l

l 3/4.9.11 RESIDUAL HEAT REMOVAL AND CCOLANT CIRCULATION. . . . . . . . . . .8 3/4 9-2

%m/

LA SALLE - UNIT 1 XV t

I

_w. _ __ _

LIST Or p7guag3

_ FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS ........................ 3/4 1-21

, 3.1.5-2 SODIUM PENTABORATE (Na:B i o0 3 . 10 H2O)

VOLUME / CONCENTRATION REQUIREMENTS . . . . . . . . . . . . . . . . 3/4 . 1-22 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CR183 8CR233 AN 8CR711 ........................,........,...D ......... 3/4 2-2 3.2.3-1 MINIMUM CRITICAL POWI? RATIO (MCPR) VERSUS t AT R ATED F LOW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 3.2.3-2 Kf FACTOR ......................................... 3/4 2-6 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE pg VS. REACTOR VESSEL PRESSURE .'...................... 3/4 4-)F

.7-1 SAMPLE PLAN 2) FOR SNUEEER FUNCTIONAL TEST .. . . .. .. 3/4 7-)( 35L B 3/* 3-1 REACTOR VESSEL WATER LEVEL ........................ B 3/4 3-7 B 3/4.s.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE ..................... B 3/4 4-7 5.1.1-1 EXCLUSION AREA ANC SITE BOUNDARY FOR GASEGUS .

AND LIQUID EFFLUENTS .............................. 5-2 5.1.2-1 LOW POPULATION ZONE .... .................... ..... 5-3 1

6.1-1 CORPORATE MANAGEMENT .............................. 6-11 l 6.1-2

, UNIT OR3ANIZATION ........"......................... 6-12 i

l 6.1-3 MINIMUM SHIFT CREW COMPOSITION .................... 6-13 l

l

(

l l

t i

l LASALLE-UNITgl ,

XIX OCT 4 :.g; 1 .

LIST 0: YASLES -

PRD0F & ILT# CDPY '

-: : p a-e.

._ x. 2 1.1 SURVEILLANCE FREQUENCY NOTATION ................... 1-3

1. 2 CPERATIONAL CONDIT!CNS .,.......................... 1-9

2. 2.1-1 '

Sc4Pw.O.NTS ...............

a REACT R PROTECTION

........................... 2-4 SYSTEM E2.1.2 *. ' UNCERTAINC ES US C IN THE DETERMINATION OF THE FUEL CLA*0!NG SAFETY LIMIT . . . . . . . . . . . . . . . . . . . . B 2-4 E2.1.2-2 NCMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLA00ING INTEGR! Y SAFETf LIMIT ...................................... B 2-5

=.,..,.2.* :" .

: ' n* . *. V . ..NN

. n. : PCs' ':9 0 .*S i nU .c *.*.N

. U.C.D *..N *. n'.:

G:~. a' S S e a" s. *." *4.*a* L a* N,* LYSt*.: ........................

= .' :

.n . . - . .n . : n. .:.n- ..v R ... n,e.a-.n...eN ..e.:n. .

... . w. .N Gc.n.a.

.e.nn.S.~ 4 w-.L nu Ah...v.S

. S .............................. .c 2 .

3.3.1-1 REA* TOR PROTIOT!ON SYSTEM INS RUMENTATION ......... 3/4 3-2 3.3.1-2 REAC CR PROTICT:CN SYS EM RESPONSE T!MES .......... 3/4 3-6

. . . . . . . .. n s..

- - . , n . p.c a . ;..-. . ..r.y gv. .e. ru. .i.g.e go.w..ry i .-T*

_ . a i ..N SURVE*LLAN0! REQUIREuENTS ......................... 3/4 3-7 2... .... ,..m e s i ,. . .w n N a. C . . . . - ..e N . N.e . s, M.eh.

wmj a.i..

-- N . . . . . . . . . . . . . . . ./. ,... . ..

3.3.2-2 ISOLATION ACTJAT!ON INSTRUMENTAT*0N SETPOINTS . . . . . 3/4 3 *.5 3.3.0-3 , ISO' AT:0N SYSTEM INS RMENTAT . 0N RESPONSE T*ME . . . . 3/4 3-13 4.2.2.1-1 ISOLATION SYSTEM !NSTRUMENTATION SURVEILLANCE -

REQUIREMENTS ...................................... 3/A 3-20 j 3.3.3-1 EMERGENCY CORE C*0L!NG SYS EM ACTUATICN t

!NS ROMENTATION ................................... 3/4 3-24 3.3.3-2 EWERGEN"Y CORE COOL *NG SYS'*EM AC UATION i

i

INS RUMENTATION SETPOIN S ......................... 3/4 3-2S j

3.3.3-3 EMERGENCY CORE C*0 LING SYS EM RE3PONSE TIME 3 ...... 3/4 3-31 1

4. 3. 3.1 ' . EMERGEN"Y CORE COCL*NG SYSTEM ACTJATION l

INSTRUMENTATICH SURVE!LLANCE RECUIR.C.uENTS ......... 3/4 3-32 l 3. 3. 4.1- 1 A7d5 RECIR*ULAT*0N PUMP TRIP SYSTEW INS RUMENTAT!ON ...................................

3/a 3-3S i

l

LA SALLE - UNIT /l ,

XX i

i 1

I

LIST 0~ T4?LES (Centinuedi

_ TAELE oACE 3.3.4.1-2 AIVS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS ......................... 3/4 3-27 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-38

, 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION .........'.........................

3/4 3-41 3.3.4.2-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM S ET PO I NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-42 3.3.4.2-3 END-OF-CYCLE RECIRCULATION' PUMP TRIP SYSTEM RESPONSE TIME ..................................... 3/4 3-43 4.3.4.2.1-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS .........................

3/4 3-44 3.3.E-2 REACTOR CORE ISCLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ............... ................... 3/4 3-46 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ......................... 3/4 3-48 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREME'TS N ......... 3/4 3-49 3.3.6-1 CONTROL ROO WITHCRAWAL BLCCK INSTRUMENTATION ...... 3/4 3-El 3.3.5-2 CONTROL RCD WITHCRAWAL ELOCK INSTRUMENTATION SETPOINTS ............................. ..... ..... 3/4 3-E3 4.3.5-1 CONTROL RCD WITHORAWAL BLOCK INSTRUMENTATION 1

. SURVEILLANCE REQUIREMENTS * . . . . . . . . . . . . . . . . . . . . . . . . . 3/43,is2f4 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION .............. 3/4 3-5(57 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .........................

i 3/4 3-59 I

3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION ................ 3/4 3-61 i

i

. 4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-62 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION ......... 3/4 3-64 4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . 3/4 3 ~E

l .

l LA SALLE - UNIT,71 ,

XXI ";7 .; :,[-

i l .

LIET O' T;ELES fCentinued)

, TABLE PAGE 3.3.7.4-1 REMCTE SHUIDCWN MONITORING INSTRUMENTATION ........ 3/4 3-67

~

4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......................... 3/4 3-68 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION ............... 3/4 3-70 4.3.7.5-1 ACCIDENT MONITORING INSTR'UMENTATION SURVEILLANCE REQUIREMENTS ......................... 3/4 3-71 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION . . . . . . . . . . . . . . . . . . 3/4

. . 3-76 3.3.7.10-1 RADICACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ................................... 3/4 3-82 4.3.7.10-1 RADICACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE , REQUIREMENTS ......... 3/4 3-84 3.3.7.11-1 RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ................................... 3/4 3-57 4.3.7.11-1 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-29 3.3.E-1 FEEDWATER/ PAIN TUREINE TRIP SYSTEM ACTUATION INSTRUMENTATION ......................... 3/4 3-93 3.3.8-2 FEEDWATER/ PAIN TUREINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPCINTS ....... ....... 3/4 3-92 4.3.5.1-1 FEEDWATER/ MAIN TUREINE TRIP SYSTEM ACTUATICN

. INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-9E 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES ..3/44,269 3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ........... 3/4 4,k(17-4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND i

ANALYSIS PROGRAM .................................. 3/4 4,)(is,,

4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

WITHORAWAL SCHEDULE ............................... 3/4 4-) (19 4.6.1.5-1 TENCON SURVEILLANCE ............................... 3/4 6-11 4.6.1.5-2 TENDON LIFT-OFF FORCE ............................. 3/4 6-12 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES .............. 3/4 6,2f24

. I LA SALLE - UNIT jt f XXII OCT 4 ~553

m. ** *

...- f m w..s ===..:.=

. e. . r. ..

a a i.e- ee a .en

. ;sb a

. I F.: '. . p 4 g,.

.r a a

......t.9- 5grah.m

..w .nn . =Y C0e- a . .m.p.,e.;h

. a. . V.:a . .7 L....tv. e ,1 .(.e i ;p.

_. nu4vnn 4C :.e.,n...Ch. .v. e : k,n,V.:5

. . ........................ ,/*, :.

4 - F. M 3.7.5.2-1 DELU3E AND SPRINKLER SYSTEMS ...................... 3/4 7-15

4. / . .c . 4. ., P.g.:

-t p.o.e . gtal.n.NS i ................................ 4'/4 7 '..c 3.7.7-1 AREA TEMPERATURE MONITORING ....................... 3/4 7-2f;NT 4.e.a.. 3.4.,

. . n;:5:L C.re.:Rr.*0p.

... .a . i;5 -

S .w..e w s gie .. .................... 3/,. .: -<

.*.2.2.2

.n i a .:.:.V. 5". n'V .. :' L'-.a.a" F :.. n':. ' ; In' ..v..: a" i .c . . . . . . . . . . . . . . . . . '

./4 . '.:.

3.c.2.2-1 PRIMARY CONTAINMENT PENETRATICN C"NDUCTOR C k.:-.n .n. >.4:a sn r- r :..- 4 .: w i . v.: n..

. rV.t. C.:.C .................... .e

.c.e.

..S....:.,. l- .. L:,.v:::.m

. . . . . -:. v. n. .. : \r 5 : h:.:..v.n.. . i CV:.:. . 0AD

.e...

:r--

.&....La r ........................................ .

.e.4/

.t . 9. *. .. .- 9 U..:.'

A.. . ... -* . aa

..n

? '.'.i v. .: ..: :.. s ? .t.3. . .:. L . h. ,a a e :.n* is . an. w n-

s. tem

. . rm.J'.:.0.: .w Un La......p... nn.a a aw , : und:0 .. L:n.g V r n. .p .. . . g... g. .as:

p.. . ..... .. .,

..S. in

. p . . . q 4 *. . 4 0. mn ....

3.g ..m.... -

iU .L.tn:. .. .

. . . -f, 1

als L.%.4W ...g4: an. ............. J/- .. *4

.: .L ,aggan . ., . 4 . A ..*C1:

. 4 . 4 ,. ....p,L4.y.

P.nsawn m. .Cn* "t'.r .l 4' .Pig* e* h'u" n* a P A ' V.. .C .T .C

=1 .. .

. . . e . . o'. ...........................................

r is e n n, as - .4"O

.--.4"a a n. 4 L e.. . . . ; un::. 5 ...qi:-

nn. g , g, . .

.r... . ,:aa,g ., nh,s. .

m'an.e ' . v. .t . .C . ... -. . . :mV. ..................................

. - *i.*e.

. .. . . .....,...e.; c 4..*. n .J4V..'J.w... : a \, ;. .". . . ...a ..'.C

. h.. e. r.

g 9*..........e f. r.6an.. 0 n = .. y .....

0. . .. :t a l . na aJ O/ = .4"4

.q..4..

$ ..).......

r.:. s. ...a . t :* *

..... r.: -.r - . nm..

a-*- *-

-

F * .a an . . : .s . y . . na.:s aa

.e p s *-- r.m* u- Cs p:*

a

.t................m...

a . . n . . .. . . a an.r,:5...........................

. .f.

, 4. ' 2 . '. '

. MIN.'.M".M . VA' . U.F.C F0 F. TF.: . .. ' n'a':~R L'.M7T.C . OF.D: . E"TTON .. 2/4 "..'-7

....s..-

4 , . .Cin.4t.:

.e . A Cnar . h Pn. i : Ann.u.r..RS e-ano iO .h:. - -

L- s .. .a s n w . . s a n. P-

. 0.c .e . n. .: --- n a n .v. .e .t $ . . . . . . . . . . . . . . . . . e .,i- . 2 4 s .e ,r . :. :. -- ..n . s v m:

. .::. . , u. .r .e .C .: t. i..a .n. ~.a. . h. r .e 5 . . . . . . . . . . . . . . . . . . . . . . . . . .

./. 4 CCMFCh.r. w.. .i i .yr :.. uOn- RAn,....h..

a2: i L.h..-.a tia .............. . .

I i

.. 3. ., 5.-::.3 7

I I

\.y...

l .n :n..: ..... fl . .. 1:

r I

4 . .

hg- - TABLE 1.2

~

. OPERATIONAL CONDITIONS ss MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDQWN ShutdownY*** > 200*F I 4.. COLD SHUTDOWN Shutdown Y*** 1 200*F l

5. REFUELING" ShutdownorRefuel**N $ 140'F l

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

NThe reactor mode switch may be placed in the Refuel position while a single i control rod drive is being removed from the reactor pressure vessel per ,

l Specification 3.9.10.1.

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

- See Special Test Exception 3.10.3

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being n: r;134 provided that the one-rod-out inter 1cck is j OPERABLE. g o e d, .

LA SALLE - UNIT 1 1-9

2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

~

2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or low Flow -

2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPt.ICABILITY: OPERATIONAL CONDITIONS 1 and 2. .

! ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTOOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the

requirements of Specification 6.4.

THERFALPOWER.MichkessureandNichFlow 2.1.2 The MINIMUM CRIT AL POWER RATIO (MCPR) shall not be less than 1.06 .

with the reactor vessel s eam dome pressure greater than 785 psig and core flow greater than 10% of r ed flow.

l APPLICABILITY: OPERATIONAL NOITIONS 1 and 2. , jm l-ACTION:

I With MCPR less than 1.06 and the actor vessel steam dome pressure greater Fo*

l than 785 psig and core flow greater han 10% of rated flow, be in at least HOT

SHUTD0hN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply w the requirements of Specification 6.4.

REACTOR COOLANT SYSTEM PRESSURE l 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel

steam dome, shall not exceed 1325 psig. .

APPLICABIL!TY: OPERATIONAL CONDITIONS 1, 2, 3 and 4 ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.4.

l l

.-~

LA SALLE - UNIT 1 -

2-1

iw + toc cap D~l THEDMAL POWER, Hien pressure and Mich Flow 2 1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall net be less than ' 05 wtntworecirculationloopoperationandshallnotbelessthan['07 single retirculation looD operation with the reactor vessel staan d >

pressure greater than 785 psig and core flow greater than 10% of ra flow. 1 A8:LICAE:LITv: CPERATIONAL CONDITIONS 1 anc 2.

ACTION: -

[ith MCFR less tnan 1.06 with two rehirculation loop operation or les- 'han

. e. w w. n s;ng,.e re:;rculation locs operation anc the reactor vessel stea-dCme pressure greater than 785 psig ano core flow greater *han 10% o#. rate.

. oa, .,e in a- leas- HOT cuu"' -ar w t~-4 4..

urs and comply with the recuire-rnents of Spe:1ficaticn6.g-[

y r-I i

l l

~

l l

~

l s

e 4

~% \ g k

4

e-

u. -

20

], TABLE 2.2.1-1 .

. REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS c FUNCTIONAL I2itT g ALLOWABLE

" TRIP SETPOINT*

4

1. -

VALUES Intermediate Range Monitor, Neutron Flux-High

}

5 120 divisions of

2. Average Power Range Monttor: full scale $ 122 divisions i

a.

of full scale

' Neutron Flux-High, Setdoun i i 15% of NATED THERMAL POWER $ 20E of itATED p

4 pt t.

7 - ;;;. .; 3 :-1,i.e T-.-i 7

=1) 7:- :;;.. e - . . rm..i. - THERMAL POWER

$ ." : 51: .n:, .

i ON*d'd[ 2) ;;;,;. r! C; _. : - -_ .: 5 G.x - 5;% .hn a-i e. ~J- '

-+-103.5% e7 ;;7.7;;;; - -; _ .;

c. i=._ a H,.ek ii5.5E vi snico- -

Fixed Neutron Flux-High Tia!...; 7.~i ia j < 118% of RATED THEllMAL POWER

~

~< 120% of RATED

{ 3.

Reactor Vessel Steam Dome Pressure - High THERMAL POWER 3 1043 psig

4. 5 1063 psig

< Reactor Vessel Water Level - Low, Level 3 3 1 12.5 zero*inches above instrument -),,11.0 inches i

above instrument '

5. zero"

' Main Steam Line Isolation Valve - Closure * $ 8% closed 6.

Main Steam Line Radiation - High $ 12% closed

$ 3.0 x full power background j

5 3.6 x full

7. power background Primary Containment Pressure - High 3 1.69 psig

8. i 1.89 psig Scram Olscharge Volume Water Level - Higli 9.

1 767' % "

< r+

Turbine Stop Valve - Closure $ 767' %"

j p/ 1 5% closed

10. $ 7% closed Turbine Controi Valve Fast Closure, j

Trip 011 Pressure - Low ,

11 1 500 psig Reactor Mode Switch Shutdown Position 1 414 psig I h, 12. NA Manual Scra:4 NA F- . ,

NA j Sci liases Figure 6 3/4 3-1. .

NA e

inw+ &c gay 9-4

. . =

e' C 3 C w w 3 >= E M -E tw %%<a M E . OE3 wh C o mCh' *

+ g O

e E *C4 '

4 E

- -MSEo.. gg - 3exaE =W<

c c e gg .

x so. e .e =w e.Eew -=

cE-- oeae vi vi vi vt a

a 2 3 C

- w w 3 r=. E M >M Cw NhCJ M. h M3 . C t. 3 ca Q c C aCh a e{wL C -C e g a + E a 2 E-.Mn- S 2xa. -MS C o- .M C@ .=

cxma

.Cm= e. E a==e m e=

cEm~ omm~

, vi vi vi vt C

U f

5 c ,

C a =

C ma L C c Q= b 3J C C C O L 6 Q

= C. O MC C E C 6 C.

O C T T C C @*

C C .

r= C C.

== c F - a e .* e GO = 4 m

- e *C U eT U M +8 0 =W

-9 m 3 Um 3 C == m C 6 C E =-- - = - . - .C

=

= U sl w UC w C

M .b 2 E 3 f To '= U 3 o .C C. 3 c .C=

m ."E 6 Z e, 6. 2 C w.

-O C g3 ) m a wm m

>= c 4 W4 .C 3

C'

~m n

b. P8 t'%d

E i

2.1 SAFETY LIMITS-BASES 2.0 The fuct cladding, reactor pressure vessel and primary system piping ~

are the principal barriers to the relaas,e of radioactive materials to the environs. Safety Limits _are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel camage is calculated to occur if the limit is not violated. Because fuel damage is not directly l

observable, a step-back approach is used to establish a Safety Limit such that I the MCPR is not less than 1.06. MCPR greater than 1.06 rg r::: t: : :::er %

ethe margin relative to the conditions required to maintain fuel cladding 1 integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of -

the cladding, fission product migration from this source is incrementally cumulative and continuously measuranle. Fuel cladding perforations, however, l can result from thermal stresses which occur from reactor operation signif-

~

i icantly above design conditions and the Limiting Safety System Settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross l O rather than incremental cladcing deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signif-icant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power

i. calculations at pressures below 785 psig or core flows less than 10% of rated I flow. Therefore, the fuel cladding integrity Safety Limit is established by

! other means. This is done by establishing a limiting condition on core THERMAL

POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flowswillalwaysbegreaterthan4.5 psi. Analyses show that with a bundle

flow of 28 x 10 lbs/hr, bundle pressure drop'is nearly independent of bundle i power and has a value of 3.5 psi.3 Thus, the bundle flow with a 4.5 psi driving

! head will be greater than 28 x 10 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical

.: power at this flow is approximately 3.35 MWt. With the design peaking factors,

j this corresponds to a THERHAL POWER of more than 50% of RATED THERMAL POWER.

i Thus, a THERMAL POWER limit of 25% of RATED THERMAL PCWER for reactor pressure jl below 785 psig is conservative.

' & kao etre)JoaleeU O'*'

j k,, gje.rwn..w.sL..g,hoeerus 7 a**

ca"5'a'* *

. LA SALLE - UNIT 1 8 2-1 e t

.- i 3 .. Bases Table 82.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION i 0F THE FUEL CLADDING SAFETY LIMIT *

. ~

Standard Deviation Quantity (% of Point)

Feeawater Flow .

1.76 Feedwater Temperature 0.76 9

Reactor Pressure 0.5 Core Inlet Temperature 0.2 .

aw, 2. 5 Core Total Fi g . ,,g g ,s y ,i s.uw G.0 -

Channel T%Iss Flow Area n e t w :ea LeeI 3.0 Friction Factor Multiplier '10.0 .

Channel Friction Factor

-m Multiplier 5.0 l

TIPReaggs ,w, yAu 6.3 R FactoE Y * * " "f *" q ' f# I.*!

Critical Power 3.6 i

i l

i i -

.i .

'l "

ine uncertainty analysis used to establish the core wide Safety Limit MCPR j is based on the assumption of quadrant power symetry for the reactor core.

t e. A lu w kam ;a agly 4e beA 4wo ver.irsvleb leep of*Abiv Ad eg ainutau we .pnm~, unet n uM i

4

=

LA SALLE - UNIT 1 B 2-4

^

l 3/4.1 REACTIVITY CONTROL SYSTEMS

~

3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:

a. 0.38% delta k/k with the highest worth rod analytically determined, or

b. 0.28% delta k/k with the highest worth rod determined by test.

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

ACTION:

With the SHUTDOWN MARGIN less than specified:

.. a . In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that t ,

could reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4, F' establish SECONDARY CONTAINMENT INTEGRITV within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l )

c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS

and other activities that could reduce the SHUTDOWN MARGIN, and insert l

all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Establish SECDNDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l SURVEILLANCE REQUIREMENTS 1

4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater j than specified at any time during the fuel cycle:

i

a. By measurement, prior to or during the first startup after each refueling.

b. By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.

c. Within sh r hoursafter detection of a withdrawn control rod that is immovable, as a result of excessive friction or mechanical inter- l ference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod.

i "Except movement of IRMs, SRMs or special movable detectors.

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

l s

, REACTIVITY CONTROL SYSTEM E 3/4.1.3 CONTROL R005 CONTROL R00 OPERASILITY .

i LIMITING CONDITION FOR OPERATION s

3.1.3.1 All control rods shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS T and 2.

ACTION:

a. With one control rod inoperable due to being immovable, as a result

, of excessive friction or mechanical interference, or known to be untrippable l

1. Within hour:

l a) Verify that the inoperable control rod, if withdrawn, is separated from all other inoperable control rods by at least two control cells in all directions. .

b) Ofsarm the associated directional control valves either: l l ; 1) Electrically, or i

2) Hydraulically by closing the drive water and exhaust water isolation valves. .

', , c) Comply with Surveillance Requirement 4.1.1.c.

l I A 2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I

3. Restore the inoperable control rod to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. .With one or more control rods trippable but inoperible for causes

other than addressed in ACTION a, above

l 1. If the inoperable control rod (s) is withdrawn:

a) Immediately verify:

} ' 3

1) That the inoperable withdrawn control _ rod (i) is /

separated from all other inoperable

control red (s) ey l l ,

at least two control cells in all directions, and

2) The insertion capability of the inoperable withdrawn l control rod (s) by in'serting the control rod (s) at least j one notch by drive water pressure within the normal operating range".* ,

I b) Otherwise, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valvesWeither: 1

1) Electrically, or Hydraulically by closing the drive water and exhaust 2)

.j water isolation valves 3 f *"ina teccers:14 ::ctes! red ay tren de with:rawn to a position no far:re-

, i i witncrawn than its position wnen founo to De inoperaola.

' "May be rearmed intermittently, under administrative control, to permit testing .:

( associated with restoring the control rod to OPERABLE status.

I LA SALLE - UNIT 1 -

3/4 1-3 I

~

, REACTIVITY CONTROL SYSTEM

(-

, LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) ~

2. If the inoperable control rod (s) is inserted:

a) Within hour disarm the associated directional control I valves either:

1) Electrically, or 2)' Hydraulically by closing'the drive water and exhaust water isolation valves.

b) Otherwise, be in at least HOT SHUTDOWN within the next l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . *

c. With more 'than 8 control rods inoperable, be in at leas't HOT SHUTDChN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS

'3 4.1.3.1.1 The scran discharge volume drain and vent valves shall be demonstrated CPERABLE by: ,

a. At least once per 31 days verifying each valve to be open*? and

[

b. At least once per 92 days cycling each valve through at least one complete cycle of full travel.

( : 4.1.3.1.2 When above the low power setpoint of the RWM and RSCS, all withdrawn l control rods not required to have their directional control valves disarmed l

,- electrically or hydraulically shall be demonstratad CPERABLE by moving each

! control rod at least one notch:

a. At least once per 7 days, and l

l b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when any control rod is immovable as a result of excessive friction or mechanical intarfarence.

l 4.1.3.1.3 All control rods shall be demonstrated CPERABLE by perfomance of l Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.5 and 4.1.3.7.

\ .

r"*"These va;ves may be c1csed inteMttently for testing urcer administratite f I control.

l M May be rearmed intermittently, under administrative control, to permit testing l associated with restoring the control rod to OPERABLE status.

l f LA SALLE - UNIT 1 3/4 1-4 1 :

. _ . REACTIVITY CONTROL SYSTEM .

2 -

SURVEILLANCE REQUIREMENTS (Continued)

- ~~

W 1?I.T 4 The scram discharge volume shall be determined OPERA 8LE by demonstrating: ,

i a. The scram discharge volume drain and vent valves OPERABLE, when

control rods are scram tasted from a normal control rod configura-tion of Ue s than or equal to 50 R00 DENSITY at least once per

18 month eby verifying that the drain and vent valves: g I

1. Close within 30 seconds after receipt of a signal for control rods to scram, and

2. Open after the scram signal is reset.

~

b. Proper float response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod bicek level instrumentation after each scram from a pressurized condition.

.R i

i i

I l .

I

'*The provisions of Specification 4.0.4 are not applicable for entry into

~j OPERATIONAL CONDITION 2 provided the surveillance is performed within g 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving less than or equal to 50% ROD DENSITY.

J' i LA SALLE - UNIT 1 3/4 1-5 1 . .

i

, . I

~ REACTIVITY CONTROL SYSTEM .

t -

.'- CONTROL RCD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERAT 0N 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 05, . based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

g. With the maximum scram insertion time of one or more control rods exceeding l

7.0 seconds

f.l. Declare the control rod (s) with the slow insertion time inoperable, 1 and ,

,g.1 Perform the Surveillance Requirements of Specification 4.1.3.2.c at I least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of -

t .

7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .

', b. The prosslaus of Spec 4:utou 3. o.4 = w + A g 1:u ble , {

l SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be i demonstrated through measurement with reactor coolant pressure greater than or l equal to 950 psig and, during single control rod scram time tests, the control l rod drive pumps isolated from the accumulators:

a. For all control rods prior to THERMAL POWER exceeding 40% of RATED

! ; THERMAL POWER following CORE ALTERATIONS

or after a reactor l8 shutdown that is greater than 120 days,

l. .

l b. For specifically affectad individual control rods following maintenance on or modification to the control red or control red drive system which could affect the scram insertion time of thos' specific control rods, and

?

! c. For at least 10% of the control rods, on a rotating basis, at least once per 120 days of operation.

'Exca n :.e a.:. of SR!), IR:t or s::ecia' movaela detect: s c: ner a' :: . trol roc :o.ement.

o LA SALLE - UNIT 1 3/4 1-6

REACTIVITY TONTROL SYSTEM N -

FOUR CONTROL R00 GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion tim'e, from the fully withdrawn position,

, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve solenof us as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.45 39 0.92 -

25 2.05 05 3.70 -

APPLICABILITY: OPERATIONAL CONDITIONS 1 anf 2.

ACTION:

d. With the average scram insertion times of control rods exceeding the above (

limits:

pt L Declare the control rods with the slewer than average scram insertion l times inoperable until an analysis is performed to determine that l required scram reactivity remains for the slow four control rod group, and

, b.Q. Perform the Survef11ance Requirements of Specification 4.1.3.2.c at [

1 east .once per 60 days when aparation is continued with an average scram insertion time (s) in excess of the average scram insertion time Itait.

l Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Th f rovis,cus o4' ' S ac:Ziea400 f 3. D.4 a ec. m + nff;cale.. I

}

l . SURVEILLoNCE RE0VIREMENTS I.

4.1.3.4 All control rods shall be demonstratad OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.

l

!- LA SALLE - UNIT 1 3/4 1-8 t

emenese sem me ,

i i

_. REACTIVITY t0NTROL SYSTEM '

CONTROL R00 SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERA 8LE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a. In OPEitATIONAL CONDITION 1 or 2.:

L With one control rod scram accumulator inoperable:

a) Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

1) Restore the inoperable accumulator to OPERABLE status, or

2) Declare the control rod associated with the inoperable accumulator inoperable. ,

b) Otherwise, be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . -

2. With more than one control rod scram accumulator inoperable, declare the associated control rod inoperable and:

~

a) If the control rod associated with any inoperable scram

^

accumulator is withdrawn, immediately verify that at least one CAD pump is operating by inserting at least one with-drawn control rod at least one notch by drive water pressure within the normal operating range or place the reactor mode switch in the Shutdown position.

b) Insert the inoperable control rods and disarm the associated directional control valves either:

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

Othendse, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. In OPEitATIONAL CONDITION 5 with:

1. One withdrawn control rod with its associated scram accumlator fooperable, insert the affected control rod and disam the

, associated directional control valves wthin 3pe hour, either: l a) Electrically, or .

L

,i b) Hydraulically by closing the drive water and exnaust water T

isolation valves.

l ! 2. More than one withdrawn control rod with the associated scram accumulator incperable or with no control rod drive pump crerating, imeciately plac: the react:r ::e A tc% S the 5 N t:: e ;csitien.

,' . C. ~fl e pfw.s.ous of Speed:ca boa 2, 0. ' Are wt Applicable, f "At least tne abulator associated with each witherswn control fod. Not applicable to control rods removed per specification 3.9.10.1 or 3.9.10.2.

LA SALLE - UNIT 1 3/4 1-9

- - , . . , - - - . - - - - - . - -.,._...g _ - . . ,--- ,,,.w- ,w_. -+. , ,. . ~ . , --._,,e_,..w.- -y --, . - , , - - . - , - .-

REACTIVITY CONTROL SYSTEM CONTROL ROD DRIVE COUPLING LIMITING CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,and 5*. I ACTION:

a. In OPERATIONAL CONDITION 1 and 2'with one control. rod not coupled to its associated drive mechanism:

l 1. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either: __ _

a) If permitted by the RWM and RSCS, insert the control rod drive mechanism to accomplish recoupling and verify recoupling by withdrawing the control rod, and:

1) Observing any indicated response of the nuclear instrumentation, and

2) Demonstrating that the control rod will not go to the overtravel position.

b) If recoupling is not accomplished on the first attempt or, if not permitted by the RWM or RSCS then until oermitted by the RWM and RSC5, declare the control rod inoperable

. and insert the control rod and disarm the associated

~

directional control valves ** either:

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either:

1. Insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel pasition, or

2. If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves ** either:

a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.

"At least each withdrawn control rod. Not applicable to control rods removed

~

...per Specification 3.9.10.1 or 3.9.10.2.

- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status. ,

LA SALLE - UNIT 1 3/4 1-11

l l .

REACTIVITY CONTROL SYSTEM t

LIMITING CONDITION FOR'0PERATION (Continued)

ACTION (Continued)

2. With one or more control rod " Full-in" and " Full-out" position 1

indicators inoperable: ~

l .

a) Either: ,

1) When THERMAL POWER is within the low power setpoint i of the RSCS:

(a) Within one hour:

(1) Determine the position of the control rod (s) by:

(a) Moving thd control rod, by single notch movement, to a position with an OPERABLE position indicator,

' (b) Returning the control rod, by single notch

, movement, to its original position, and (c) Verifying no control rod drift alarm at least per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or (2) Move the c"ontrol red to a position with an .

OPERABLE position indicator, or (3) Declare the control rod inoperable.

O (b) Verify the position and bypassing of control rods with inoperable " Full-in" and/or " Full-out" position indica-tors by a second licensed operator or other technically qualified member of the unit technical staff.

2) When THERMAL POWER is greater than the low power setpoint of the RSCS, determine the position of the control rod (s) per ACTION a.2.a) 1)(a)(1), above.

b) Othenvise, be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITION 5" with a withe awn control rod position

, indicator inoperabl.e, move the control rod to a position with an

! , OPERABLE position indicator or insert the control rod.

} SURVEILLANCE REQUIREMENTS C< %s. Proi.s;ous .o1 a.o.4 c e mt app lita kle. . /

3 . .

l 4.1.3.7 The control rod position indication system shall be determined OPERABLE

by verifying: -

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the position of aach control rod is

! indicated,

b. That the indicated control rod position changes during the movement of the control rod drive when performing Surveillance Requirement j 4.1.3.1.2, and

c. That the ::ntr:1 red posit'on indicator corresoc,:s to the icn .rci I /

roc posit!:n S i:stad by tne " Full out" cosi tice indicator ..s, perfor:ning Sur.wiliance Requirement 4.1.3.g. l "At least eacn withdrawn control rod not applicable to control rods removed per Specifications 2.9.10.1 or 3.9.10.2.

LA SALLE - UNIT 1 3/4 1-14 w . 4

-- --,----~,n- m ---<m-w--,,n -,-,y , , , - - , - - ~r- -, , - n, , - - - - _ , - - ~ _ - - _ , - - - - - . - - _ , - - . - - - - - - - - - -

i REACTIVITY TCNTROL SYSTEM 3/4.1.5 STAN08Y LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE. '

APPLICABILITY: OPERATIONAL CONDITIONS f. 2, and 5".

l ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With one motor operated suction valve, one pump and/or one explosive valve incperable, restore the inoperable suction valve, pump and/or explosive valve to OPERA 8LE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. With the standby ifquid control system inoperabla, restore the

system to CPERABLE status within a hours or b.e in at least HOT l SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l

b. In OPERATIONAL CONDITION 5*: .

1. With one motor operated suction valve, one pump and/or one explosive valve inoperable, restore the inoperable suction valve, pump and/or explosive valve to CPERA8LE status within 30 days or '

insert all insertable control rods within the next hour.

2. With the standby licuid control system inoperable, insert all insertable control rods withinjwis hour.

1 l SUREILLANCE REQUIREMENTS

4.1.5 The standby ' quid control system shall be demonstrated OPERABLE

' a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that; l}

1. The available volume and temperature of the sodium pentaborate solution are within the limits of Figures 3.1.5-1 and 3.1.5-2, and

2. The heat tracing circuit is CPERABLE-by verifying the indicated

, temperature to be > 6C'F on the local indicator.

l l .

l ; . "With any control rod withdrawn. Not applicable to control rods removed per

, e Soeci fication 3.9.10.1 or 3.9.10. 2.

,e 4

l 5 '

LA SALLE - UNIT 1 3/4 1-19

, - - . - - - -, ,.,.r.--- - ,- ,.,ea--,,,.w. , , s, . - - , - , , . - , - , - - , - - - - , - - - - - . _ . < - - - - - - - -

4

- . i 3/4.2 POWER OISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE .

LIMITING CONDITION FOR OPERATION l 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type 4

of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3 2.1-1. W (;.4.+s sj sgow, s.o.l-1 skans be md cad 4ea. vahae.

( 0.25 &mes 41a iwo weirckin+ient leeg #peNeau I; .*+ A L r$ le vain.atSuu (,,p ope.Aou ;

APPLICABILIN: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 257. of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25" of RATED THERML POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . .

q_

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits ..

determined from Figure 3.2.1-1:

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

! b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL 80WER increase of at i least 15". of RATED THERMAL POWER, and I

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL RCD PATTERN for APLHGR.

i 1

l!

i l ; .

i(

, LA SALLE - UNIT 1 3/.1 2-1 l

- .e e

, , . _ , ,_ ,--,v_, --

--+.-w -+y,. ,,y, -

I

. POWER DISTRIBUTION LIMITS -

3/4.2.2 APRM SETPOINTS

' LIMITING CONDITION FOR OPER/YION 3.2.2 The APRM ficw biased simulated thermal power-upscale scram trip setpoint (5) and flow biased simulatad thermal power-upscale control rod block trip setpoint (SR8) shall be established according to the following relationships:

t pr Y ? '" "

2 0 01 te C .07' ^ Sl*' E

.p _

T _e , u_._._ ....--..3 ...c in ecu . A 6,.

where: 5 and 5 are in percent of RATED THERMAL PCWER, W=Lochgrecirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million Ibs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL PCWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. T .

is always less than or equal to 1.0.

A_PLICABILITY:

P OPERATIONAL CONDITION 1, when THERMAL POWER-is greater than or

, equal to 2fJ. of RATED THERMAL POWER. -

ACTICN: .

With the APRM flow biased simulated thermal power-upscale scram trip satpoint

~

and/or the flow biased simulated thermal power-upscale control rod block trip setpoint set less conservatively than 5 or S as above determined,-initiate correctiveactionwithin15minutesandrestNe,5and/ ors g to within the required limits

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to fess than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE0VIREMENTS 4.2.2 The FRTP and the MFLPD for each class of fuel shall be detarmined, the

value of T calculated, and the most recent actual APRM flow biased simulated thermal power upscale scram and control rod block trip setpoint verified to be t

within the above limits or adjusted, as required:

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL oCWER increase of at least 15% of RATED THERMAL POWER, and

~

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when, the reactor is operating with MFLPD greater than or equal to FRTP.

4 "Wita MFLPD greater than the FRTP during power ascension up to 90% of RATED

! THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be I adjusted such that APRM readings are greater than or equal to 100*. times MFLPD,

provided that the adjusted APRM reading does not exceed 100". of RATED THERMAL POWER, the required gain adjust.?ent incremen* does rot exceed 10" of RATED T!.EE"AL ?C..ER. and a r.otice of na ujust. ar: is :cna :.- . .a asc:s :Or.:-::

, panal.

o i

[ LA SALLE - UNIT 1 3/4 2-3 ,

7pSeG PA&E y/4/ dl -3

a. TA Recirculation Loop Operation 5 less than or equal to (0.66W + 5 3)T S 1ess than or equal to (0.66W + 4 3)T RB

b. Single Recirculation Loop Operation S less than or equal to (0.66W + 45.A)T S less than or equal to (0.66W + 36.7%)T RB l

I l

1 l

1

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

. . I POWER DISTRIBUTION LIMITS POWER DISTRIBUTION LIMITS 3

~/4.2.3 MINIMUM CRITICAL POPER RATIO

~

LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM ITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit d ermined from Figure 3.2.3-1 times the K datar=ined from Figure 3.2.3-2. 7 APPLICABILITY:

dprb OPERATIONAL CONDITION 1, when RMAL POWER is greater than or equal to l 25% of RATED THERMAL POWER.

ACTION With MCPR less than the'appitcable MCPR 1 it determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action wit. n 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or r uce THERMAL POWER to less than

, 25% of RATED THERMAL POWER within the next 4 ho s.

. SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

a.

t*** = 0.86 prior to perfomance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or

b. I,y, determined within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2:

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

, b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after conoletion of a THERMAL. POWER increase of at least 15% of RATED THERMAL POWER, and i :'

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating i

with a LIMITING CONTROL R00 PATTERN for MCPR.

O t

LA SALLE - UNIT 1 3/4 2-4 .

INSetT fMr5 3/4 2~4 3.2.3 than the MCPR limit detarained from Figure 3.2.3-1 times t to the K or greater than the MCPR limit determined from Figurefrom 3.2.3-1 + 0.01 times provic,ed that the end-of cycle recirculation pump trip (EOC OPERABLE per Specification 3.3.4.2.

APPLICABILITY:

OPERATIONAL 25% of RATED THERMAL POWER.

CONDITION 1, when THERMAL POWER is greater than or equal t ACTION a.

With the end-of cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of MCPR is determined to be equal to or greater than the shown in Figure in Figure 3.2.3-2. 3.2.3-1 EOC-RPT inoperable curve, times the K y shown b.

t With MCPR less than the applicable MCPR limit' determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within

15 minutes and restore MCPR to within the required limit within 2POWER hourswithin or reduce the next 4THERMAL hours. POWER to less than 25% of RATED l

l l

l 1

1 i

e

< , - , , - - -----..,,,,,.e., , - - , , ,,-,..n--,, - , . , , , ,,--,---,,,n.

\

%n -

r{ tgy

,pm:o ,

~v Be5gz s C3 t 1

/

6 8

. .i- ls -

qi'

\. :J_ 0

'V l l.! . I

.- :: 5 8

- I r, .

0 W

O L

4 F f

8 DE

- 0 T

, A R

T

t. .

. 8 3

T A

.l r-0 t J,

S U

- 2 S

, 8 R

, i E

0 V

, )

- R P

- 1 C 8 M i

(

0

, O I

T

, A 0 R 8

, l R

i 0 E W

' O P

- 9 L

' 7 A

- 1 C

0 I T

, I R

C 8

M 1 7 U

0 M

. I N

I

[.

7 M 7

1 0 -

3 2

6 7 3 0 e r

u g

i 5 F

. i 7 0

4 7

, I

. 0

'0 5 0 5 2 3 2 1

- 1 1 1 1 1 ' O abM c 1

9 v,> r~li . ezy g wsA 7u

- 1; :,i il

,:, {' l

30Em3 Euy3ec2oz im 0 3 0 5 0 4 3 3 2 2 1 1 . . 1 1 1

- 0 0

i 5

0 W

/

O

/

. L F

4 D 8 E T

A H .

T 3 A 8 r

/ p S

TE U PL S RB A 2 R 0 E C- ER V s

/

O P )

- EO R P

1 C 0 M

(

I O

T 0 A 0 R R

n .

E'

/ '

9 7

W P

L A

E O

T L C

- P B I T

RA 0 l

- R 7 n CE C OP M EO U NI M 7 I 7 N I

M 1

0 3 7

2 3

, g r

5 u 7 l o

F y 4 7

0 0 5 0 5 0 3 3 2 2 7 4 3 1 1 1 1 1

,; i .< ' :l I : Ii

L

^

l .

). 3/4.3 INSTRUMENTATIO' 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION .

3.3.1 As a minimum, the reactor protaci. ion system instrumentatien channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM l

RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

i ACTION: .

a. With the number of OPERABLE channels less than required by the Minimum CPERA8LE Channels per Trip System requirement for one trip system, place that trip system in the tripped condition" within_ par hour, The provisions l of Specification 3.0.4 are not applicable. 1

b. With the the r$ umber of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one. trip system"* in the tripped condition within Sne hour l 1 and take the ACTION required by Table 3.3.1-1. 1 '

SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated GPERABLE by the performance of the CHANNEL CHECX, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITICNS and at the frequencies shcwn in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic oceration of

! all channels shall be performed at least once per 18 mo'nths.

li

! ; 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip 1 functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its

limit at least once per 18 months. Each testishall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system. .

i "With a design providing only one channel per trip systas, an incperable channel f need not be placed in the tripped condition where this would cause the Trip j Function to occur. In these cases, the inocerable channel shall be restored to OPERA 3LE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.1-1 l for that Trip Function shall be taken. -

l' :

22 If =cre cnar e's n e:ersale in :ne tr5 system than 5 the otne see:t that trip system to place in tne trippea canaition, except wnen snis would

-U cause the Trip Function to occur. .

LA SALLE - UNIT 1 3/4 3-1

-,e.-r. -~

s .

s ..

./ REACTOR PROTECTION SYSTEM INSTRUMENTATION l ACTION

~

ACTIONi- - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 -

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within

,, . one hour.

! ACTION 3 -

Suspend all operations involving CCRE ALTERATIONS

and insert all insertable control rods within one hour.

, ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Be in STARTUP with tne main steam line isolation valves closed '

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 6 -

Initiate a reduction in THERMAL POWER withi 15 minutes and i reduce turbine first stage pressure to < 14 psig, equivalent l to THERMAL PC'nER less than 30% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

A ACTION 7 -

Verify all insertable control rods to be inserted within,pinf l hour.

ACTION 8 -

Lock the reactor mode switch in the Shutdown position within'

, ,one~ hour. I

! 1 I i ACTION 9 -

Suspend all operations involving CORE ALTERATIONS,* and insert .

l1 all insertable control rods and lock the reactor mode switen in

)l the SHUTDOWN position within,gne hour. l l I t

d. .

"Except movement of IRM, SRM'or special movab!e detecters, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.

i E$niti;isetpeint. Tinei:epeintt;b;d:t:h-'.eddur'n;:tariwwie=Lv.;;.;..-

! e uiree en:,;: te ts<s ::t;: int :n 3, se sce-stted to tn. :.._.ssica mitntn 90 day; cf-test cv giiirtfo% .

1 t

(- -

J T

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

f m

f ,

.) TA8LE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION .

! TABLE NOTATIONS -

e

. I 4

(a) A channel any be placed in an in.cerable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for I

required surveillance without placing the trip system in the tripped condition provided at least one OPERA 8LE channel in the same trip system

.f is monitoring that parameter.

(b) The " shorting ifnks" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn" and during shutdown margin demonstrations performed per Specification 3.10.3.

(c) An APRN channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel. .

i (d) This function is not required to be OPERA 8LE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

t (e) This function shall be automatically bypassed when the reactor mode ~

l switch is not in the Run position.

(f) This function is not required to be OPERA 8LE when PRIMARY CONTAINMENT INTEGRITY is not required.

i :

(g) Also actuates the standby gas treatment system.

! (h) With any control rod withdrawn. Not applicable to control rods removed '

i I

per Specification 3.9.10.1 or 3.9.10.2.

! (f) Thisfunctionshagbeautomaticallybypassedwhenturbinefirststage i

pressure is < 140 psig, equivalent to THERMAL PC'nER less than 30*.' of I RATED THERMAE POWER.

(j) Also actuates the EOC-RPT system.

"Not requirea for control rods removed per Specifications 3.9.10.1 or 3.9.10.2.

tial-setcof-t.

. . . : __ a w.___

e5? cet;ef-t t- 5e

.ws, 9,,t--

ed t-'a;..ete-*u; +--* ;- ; t-

.. . ....:.. .w. w. ...w_,.... .w. _ ,, . .. .<.u.

' ' - ' - '"' ~~ ' ' '" ~ ' '"

a__ _,..n' a..'.Ti n. .._.__T.~ , 7_ C. . . Ei ' ' ' '" ' ~ ~ ~ ~' ' " '

9 ml l s. v l I. LA SALLE - UNIT 1 3/4 3-5 i

l I

- * ..me

.,-g.,- .r-- .-es v -- -r.,-i--- . . - ,. . - - - 3-.- ,----c-c. ,-~.,,..g

)

g t' 'O ,

.) -

.)

TA8tE 3.3.2-1 .

{ ISOLATION ACTUATION INSTRtJMENTATION '

e VALVE GROUPS MINIMUM OPERABLE APPLICABLE E OPERATED BY CHANNELS PER OPERATIONAL M TRIP FilNCTION SIGNAL (a) TRIP SYSTEM (b) CONolTION ACTION ,

A. Al_l!0MATIC INITIATION

1. l'HINARY CONTAINMENT ISOLATI0fi ~

Neactor Vessel Water Level (1) Low, Level 3 7 2 1, 2, 3 20 7 j

(2) Low Low, Level 2 1, 2, 3 2 1, 2, 3 20 l

b. Drywell Pressure - High 2, 7 2 1,2,3 20 -

. c. Main Steam Line i I

1) Radiation - liigh 1 2 1,2,3 21 3 2 1,2,3 22 w 2) Pressure - Low

' 1 2 23 t 3) Flow - liigh 1 2/line(d) 1 1, 2, 3 21

't* d. .

- Main Steam Line Tunnel

Temperature - High 1 2 1, 2, 3 21

e. Main Steam Line Tinuiel di A Temperature - High 1 2 If k, 34*) 21 l f. Condenser. Vacuum - Low 1 2 1, 2* , 3* 21 I 2. }LCONDARY CONTAINHLNT 150 TAT 10N

! .I

a. Neactor Rullding Vent Exhaust Plenum Radiation - Hloh 4 ICII*) 2 1, 2,, 3,and **

' i 24 le. Drywell Pressure - High 4(c)(e) 2 1,2,3 24 l c. Reactor Vessel Water

j Level - low Low, level 2 ICII*)

4 2 1, 2, 3, and # 24

d. Fuel Pool Vent Exhaust

}

. Radiation - liigh 4(c)(e) 2 1, 2, 3, and ** 24 .

i e

l i

I

l. TA8LE 3.3.2-1 (Continued)

. ISOLATICN ACTUATION INSTRUMENTATION

[ ACTION ACTION 20 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN with the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 21 -

Be in at least STARTUP with the associated isolation valves

closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDCWN within i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 22 -

Close the affected system isolation valves within p t hour and l declare the affected system inoperable. '[.

ACTION 23 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 -

Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating withindone hour. 4 I ACTION 25 -

Lock the affected system isolation valves closed within pat hour ,

and declare the affected system inoperable.

ACTION 26 - Provided that the manual initiation function is OPERABLE for

- each other group valve, inboard or outboard, as applicable, in

. each line, restore the manual. initiation function to CPERABLE

status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, restore the manual initiation

! function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise:

.; a. Se in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in

. COLD SHUTD0kN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

q' b. Close the affected system isolation valves within the next t

, hour and declare the affected system in operable.

NOTES ,

May be bypassed with reactor steam pressure < 1043 psig and all tureine stop valves closed.

- When handling irradiated fuel in the secondary containment and during CORE i ALTERATIONS and operations with a potential for draining the reactor vessel.

During CORE ALTERATIONS and operations with a potential for draining the 4 reactor vessel.

(a) See Specification 3.6.3, Table 3.6.3-1 for valves in each valve grouo.

. ; (b) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped j- condition provided at least one other OPERABLE channel in the same trip

system is monitoring that parameter.

i (c) Also actuates the stancby gas treatment system. ,

j (d) A cnannel is OPERABLE if 2 of 4 instruments in that channel are OPERA 8LE.

' ' (e) Also actuatas secondarj containment ventilation isciation dampers per Tante 3.6.5.2-1.

(f) Closes only RWCU system inlet outboard valve.

(g) Requires RCIC steam supply pressure-low coincident with drywell .

,, pressure-high. ,,,

(h) Manual initiation isolates 1EJ1-F008 only and only with a coincident reac cr vessel water level-Icw, level 3. F ral. .

l~/ ,

([,) h4k CkCNnLlt o4 eack -bvl Sys4er< y Ay he. lnce A o'u on s'avoprable SMs for- op * 'f hcbes 4cf re

?

W',4o# {p;ig S. 4r.g jp 4 sys4entF8. 6 Add dA y r Cych Ac m

f. d,J. -Ad A. 4d.cu t hprabf.

3/4 3-14 LA SALLE - UNIT 1 dauMs ,9 4k S Ame bi f 4M M- '

d A NG- .

~

I

I O ")

, .)

t YABLE 3.3.2-2

l l E '

l un ISOLATION ACTUATION INSTRUMENTATION SETPOINTS i E ALLOWA8tE l E. TRil' ilHICTION TRIP SETPOINT VALUE

+

c A. AUT0HAllt INITIATION 5_, 1. .P.HIMARY CONTAINMINT ISOLATION j a.

Reactor Vessel Water Level

1) low, level 3 1 12.5 inches

1 11.0 inches

l , 2) Low tow Level 2 > -50 inches

1 -57 inches

i b. Drywell Pren.sure - High 1 1.69 psig i 1.89 psig .-

, c. Main Steam Line

1) Radiation - High 1 3.0 x full power background 1 3.6 x full background *

2) Pressure - Low 1 854 psig > 834 psig ,

Flow - liigh t

3) i III psid i 116 psid

' d. Main Steam Line Tunnel .

Temperature - High 1 140*F 8 i., e. Main Steam Line Tunnel 1 146*F F 1 a Temperature - High <24*FY $ 30* d .

y f. Condenser Vacuum - Low > 7 inches Hg vacuum > 5.5 inches Hg vacuun ,

.l 2. SECONDARY CONTAINHENT IS0tATION -

a. Reactor Building Vent Exhaust i ,

h.

Plenum Radiation - High Drywell Pressure - liigh 1 10 mr/h 8 1 15 mr/h 8 .

i 1.69 psig i 1.89 psig I

c. Reactor Vessel Water level - Low Low, Level 2 1 -50 inches *

d. Fuel Pool Vent Exhaust 1 -57 inches" Radiation - liigh 1 10 mr/h 8

, i i 15 mr/h Y

3. Hil ACTOR WATER CLEANUP SYSTEM IS0tATION

a. Aflow - High i 70 gpm h., i 87.5 gym Heat Exchanger Area Temperature 13/ 187

{ - High 1 Jah*F@

c. lleat Exchanger Area Ventilation $ JSi*FF at - Hiuh 3~ le qg -

5 5 Jota 12 '7-

d. Pump Area lemperature - HI0h < y.0 8'

e. < J66 F

1.

Pump Area Ventilation AT - High SLCS Initiation 5 38"f8 5 36*F@

NA 13 , NA L g g i 9 Reactor Vessel Water Level -

tow low, I,evel 2 > -50 inches

l > -57 inches

t

. l l ..

.t s

/

g; TABLE 3.3.2-2 (Continued) i g 150LATION ACTUATION INSTRteqENTATION SETPOINTS

. F r"

ALLOWABLE 1'

TRIP FUNCTION TRIP SETP0lNT VALUE

,' =

l q 4. HIACIOR CORE ISOLATION C00LillG SYSTEM IS0tATION

a. RCic Steam Line Flow - High 1 290% of rated flow, 178" H O

h. RCIC Steam Supply Pressure - Low 2 1 2 M of ra W H w , W H 2O '

1 57 psig 1 53 psig

c. RCic Turbine Exhaust Diaphrage Pressure - liigh 1 10.0 psig -

1 20.0 psig

' d. RCIC Equipment Room Temperature - High 1 200* M

e. RCIC Steam Line Tunnel i 206* @ .

i Temperature - High 1 200*

f. RCIC Steam Line Tunnel 1 206*F M y g.

A Teagecrature - High Drywell Pressure - liigh 1 117* #

i 1.69 psig 1 123* 8

i 1.89 psig j y 5. #llN SYSTEM SifAN COMOENSING N00E 150tATION '

e i og 1 a. RilR Equipment Area- .

l A Temperature - High 1 50*F @

< 56* W

! b. RilR Area Cooler Temperature -

liigh l

1 200* M i 206* Y

, r. . RilR lleat Exchanger Steam Sigsply Flow - liigh 1 123" H20 ,

i 128" H 2O i

l I

I t

y G

rM O -

TABLE 3.3.2-2 (Continued) ;

E t' y,

' ISOLATION ACTUATION INSTRUNENTATION SETPolNTS

?.

, i ALLOWA8LE j . TRIP FullCT10N TRIP SETPOINT '

VALUE c-

-e

6. Rild SYSTEN SiluT00 W COOLING MODE IS0tA110N

~

a. Reactor Vessel Water Level -

Low, Level 3 1 12.5 inches

1 11.0 faches*

b. Reactor Vessel ' '

s (RilR Cut-in Permissive)

Pressure - High i135psig**[ i145psig*d i

c. RilR Pump Suction flow - High i 180" H 2O 1 186" If20 -

d. RilR Area Cooler Temperature -

!:" High

+ 1 200*f @ 1 206*F &

Y e. RilR E uipment Area AT - liigh i 50*F Y t's $ 56*F W

8. HAlillAL INIIIAIION Not Appilcable Not Applicable

1. Inboard Valves .

2. Outboard Valves

3. Inhoard Valves

4. Outboard Valves .

5. Inboard Valves

6. Outhoard Valves . .

/. Outboard Valve 7 d 'Hases figure B 3/4 3-1.

. "Im ii.ii wipu;i t.

-th"_ tgeint-thall 5:

I!a:! ;;ipint 5 S det:=i..e4 d. 0.g stest.+ tesi ri+-. .ty i;&!s;4 d- . ito-

/**Correa Led for cold water head with reactor vessel flo oe.d d::6:lited to the Caissien eithin ^0 day; cf t; t :=;det!r l

e&* . e th

!i

~

j t. TA8LE 3.3.2-3

, ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (SecondsM '

A AUTWRTIC INITIATION

1. MtIMARY CONTAINMENT ISOLATION ,

s. Reactor Vessel Water Level

1) Low, Level 3

NA

2) Low Low,l.evel 2 < 1. < 13(*)..

k. Drywell Pressure - High 713g-

c. Main Steam Line

1) Radiation - High(b) < 1.0*/< 13 I *)..

I 1.0"/7 13 I

2) Pressure - Low

3) Flow - High 7 0.5"/7 13I **).. )..

d. Main Steam Line Tunnel Temperature - High RA

~ '

e. Condenser Vacuus - Low NA

f. Main Steam Line Tunnel a Temperature - High -44A

2. SECCNOARY CONTAINMENT ISCLATION a.

ReactorBuildinhgntExhaustPlenum Radiation - Hig < 13(a) ,

b. Orywell Pressure - High 7 13I ")

i c. I

d. ReactorVesselWaterLevel-Low,Levelg) i713 "")

Fuel Pool Vent Exhaust Radiation - HighI 13I)

3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. A Flow - High < 13(a)H

b. Heat Exchanger Area Temperature - High NA

c. Heat Exchanger Area Ventilation AT-High NA l d. Pump Area Temperature - High NA

e. Pump Area Ventilation AT - High NA

f. SLCS Initiation NA

g. Reactor Vessel Water Level - Low Low, Level 2 < 13(a)

4. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam Line Flow - High < 13(a) I N

b. RCIC Steam supply Pressure - Low '

7 13 *)

, c. RCIC Turbine Exhaust Diaphragm Pressure - High NA

. d. RCIC Equipment Room Temperature - High NA lI

e. RCIC Steam Line Tunnel Temperature - High NA

f. RCIC Steam Line Tunnel a Temperature - High NA

g. Drywell Pressure - High NA

5. DiR SYSTE'4 3 TEAM CONDE'IS MG C2E ISOLATION l 4. .tiiR Iquipment Area a 74.pdrnure - Hign NA l b. RHR Area Cooler Temperature - Hign NA c.

l %

RHR Heat Ex: hanger Steam Supply Flow High NA LA SALLE - UNIT 1 3/4 3-16 I

~

i

k TA8LE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME .

TRIP FUNCTION RESPONSE TIME (Seconds)#

6. RHR SYSTEM SHUTDOWN C0OLING MODE ISOLATION

a. Reactor Vessel Water Level - Low, Level 3 -

< 13I ")

b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High , Jik N. A.

c. RHR Pump Suction Flow - High NA d.A.

d. RHR Area Cooler Temperature High NA A/.4.

e. RHR Equipment Area AT High NA v. A

8. MANUAL INITIATION e t ".pp;. W N.A.

1. Inboard Valves

2. Outboard Valves

~

3. Inboard Valves *

4. Outboard Valves

5. Inboard Valves .

6. Outboard Valves

! 7. Outboard Valve (a) The isolation system instrumentation resconse time shall be measured and recorded as a part of the ISCLATION SYSTEM RESPONSE TIME. Isolation system instrumentation response tiae specified includes the delay for diesel generator starting assumed in the accident analysis.

i (b) Radiation detectors are exempt from response time testing. Responsa time li shall be measured from detector output or the input of the first electronic component in the channel.

Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed.

i Isolation system instrumentation response time for associated valves except MSIVs.

i

Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time i snown in Table 3.6.3-1 and 3.6.5.2-1 for valves in each valve group

to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

!, ## Without 45+1 second time delay.

.rsg w,sa 6 s suml we 4ky.

N.A . []$ lo'cOlL LA SALLE - UNIT 1 3/4 3-19

--.m . . - . , _ . - ,, . . ,- __m _ _ _ -_ _ , . _ _ _ - - - - _ _

I INSTRUMENTATICN ,

END-0F-CYCLE RECTRCULdTION PL'MP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITICN FOR OPERA ION 3.3.4.2 The end of-cycle recirculation pump trip (ECC-RPT) system instrumenta- l tion channels shown in Table 3.3.4.2-1 shall be OPERA 8LE with their trio setpoints set consistent with the values shown in the Trip 5etpoint column of Table shown in3.3.a.2-2andwiththe(RECIRCULATIONPUMPTRIPSYSTEMRE Table 3.3.4.2-3. (

EM D - o r- c y c.a.E APPLICABILITY: OPERATICNAL CONDITION 1, when THERMAL POWER is greater than or equal to JG of RATED THERMAL POWEP..

ACTION:

a. With an end-of-cycle recirculation pump trip system instrumentation

cnannel trip setpoint less* conservative than the value shewn in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip 5etpoint value.

b. With the number of OPERABLE channels one less than required by the Minimum CPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within.pme hour.

A. l

c. With the number of OPERA 8LE channels two or more less than required '

by the Minimum OPERABLE Channels per Trip System requirement (s) for l one trip system and: -

l l

1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition withinjae hour. l

.i.

2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d. With one trip system inoperable, restore the inoperable trip systes

! to OPERASLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or r:t: 5:""."L Z .; ^.e '

a.a w -r nar n ue "u_ m ro s t M n th. - e 5 e m . + ,hr.

rugvirrJ :n o cACTION t

Spac b + ion 3.2. 3 L e.

With both t ip systems ingperable, restore at least one trip system to OPERASLE status within s on, hour or r;=: TliC.1"A .:s ER w ine <

t.= = J "a!C Win."'.L ."C'in " W- th: mt 5 5:r;. ++tte + k mil 0A/

gv'aed $ ca +M 3.2.3.

l .

4 LA SALLE - UNIT 1 3/4 3-39

~

) 4 .

~

TABLE 3.3.4.2-1 .

g ENO-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION N

f. MININUM -

?

c TRIP filNCTION OPENAfstE CilANNEg) ,'

PER TRIP SYSTEl4 t

1. innbine Stop Valve - Closure 2(b)

2. Tuibfne Control Valve - Fast Closure 2(b) r .

R w (a)A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required survel11ance i

H provided that the other trip system is SPERA3LE. *

(b)this function shall be automatically bypassed when turbine.first stage pressure is less than or equal to

! 14tDpsig,equivalenttoTilERHALPOWERlessthan30%ofRATEDTHERMALPOWER. k Mkhi <ataaint_ J bet .atpot - te 5: *:te:wir d d--tm -* r* - test : q - * ; :;r?-co :t- ;: Os '

. tit 15TEYp61nt:7hatt%ubaltted to theTosentssinn wtthirr-907tays of-test compkM-- j i

""** ,e

1 p _T_ A8tE 3. 3. 0-2_ '

b* TRIP FUNCTION -

_C_ONTROL R0D WITWRAWAL BLOCK INSTRUMENTATION SETP

1. TRIP SETPOINT ROD SLOCK MONITOR ALLOWABLE VAllE

.E.

a. Upscale

b. Inoperative < 9 M u ^ '""

htA c. Downscale JiA. W.A.

0. ',0 ;; : " 3".

StA.U. A .

2. APRM 15% of RATED THEllMAL POWER ]

> 35 of RATED THEllMAL POWER a.

Flow Blased Steulated .

gg37 InoperativeThermal Power-Upscale s

?. 55 Y ^ 'ya .

c. 24 U.A. a 9jg u . aw a Downscale

d. Neutron Flux-High JiA N. A .

i > 5% of RATED THEAMAL POWER l

3. 312%ofRATEDTHERMALPOWER > 31 of RATED THEIDIAL POWER SOCRCE RANGE MONITORS R a.

514% of RATED THERMAL POWER

b. Detector not full in-Upscale Y c. Inoperative MAN

< 2 x 4105 cps .NA#.A.

i 0 d. Downscale Jih M. A. < a s 10 cps l

4. 1 0.7 cps .JiA-s.4. I INTElWEDIATE RANGE MONITORS '

1 0.5 cps a.

b. Detector Upscale not full in NA M.A.

c. Inoperative 414M.A.

j d. Downscale M$ 108/125 M .A.

of full scale

$ 110/125 of full scale l

} 5. 1 5/1250f full scale -NA N.A.

SCRAM DISCHARGE VOLUME 3 3/125 of full scale g

a. Water Level-High

b. Scram Discharge Volume -< 765' %"

l i & Switch in 8ypass '< 765' %"

g 6. JIA d. A .

-NA N 4

! @ REACTOR

a. Upscale COOLANT SYSTEM RECIRCULATION FLOW i

k b. Inoperative < 108/125 of full scale

{ c.

, Comparator JiA A). A. < 111/125 of full scale

< 10% flow deviation JIA-N.A .

5 11% flow deviation '

j

1he (W). Average Power Range Monitor rod block function is varied as a function The trip setting of this function must be maintained in of recirculation loop flow I

accordance with Specification 3.2.2.

\

l

JM5eETS FOR PAGE 3/g 3-53

" inset-T A

1) Two Recirculation < 0.66 W + 43%

Loop Operation < 0.66 W + 40%

2) Single Recirculation < 0.66W + 37.7%

Loop Operation < 0.66W + 34.7%

l JNSecT G

1) Two Recirculation

. Loop Operation < 0.66 W + 42%*

< 0.66 W + 45%*

2) Single Recirculation Loop Operation < 0.66W + 36.7%* < 0.66W + 39.7%*

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

~

I I ,

('t

. 5 TABLE 4.3.6-1 I

g '

y,

' CONTROL ROD Will10RAWAL BLOCK INSTRilMENTATION SURVEILLANCE REQUIREMENTS N

l;; C11ANNEL OPERATIONAL l

, , CHANNEL Il#4CT10NAL CllANNEL CONDITIONS FOR WilICH ..

1Ril' IUllC110N g

. c _ClllCK lESI CAllRRATION ,) SilRViltlANCE REQUIRED ,'

l h 1. R0ll,ulDCK MONITOR !

! " a. Upscale NA S/U(b)(c) d, (c) 9 ya I

h. Inoperative NA S/U NA- N o. la I

c. Downscale NA S/U d,* Q 1* .

l 2. Aflui

a. fl0W 0lased Situlated .

lisensal Power-Upscale NA S/U(b) y yy '

le. Inoperative S/U(b),M y

u_

NA , .JiA.N.A 1, 2, 5

e. . Ilownscale NA S/U .M N SA 1 it. Ileutron flux-liiuh NA S/U ,M A- SA 2, 5

)" 3. *filll:CF RANGt HUNilDRS

, <;' a. lietector not full in MA ,W -NA.N .A . 2,'5 I-

g h. Upscale NA S/Ufb S/U w Q 2, 5

c. Inoperative NA S/U(b)'W

.NA-d . A . 2, 5 ,

it. Ilownscale NA S/Ug),W , , Q 2, 5 8

4. Ilill HHilll ATE RANGE MONITORS -

. . . lietector not full in NA S/U ,W JiA-A1.A . 2, 5 I

h. Upscale NA S/U ,W 2, 5 Inoperative

Q

a. .

d. Ilown!.cale NA S/U ,W 44A-#.A. 2, 5 l NA S/U ,W , Q 2, 5

5. stl, RAM DISCilARGE VUluME

a. Water Level-liigh NA q R' 1, 2, 5**

1,. Scross Discharge Volume Switch in flypass NA M -ilA.A1. A . 5** l

6. N1,_AC10R COOLANT SYSILM RECIRCULATION FluW . ;

a. Upscale NA S/U(b) g ,

9 y

h. Inoperative NA S/U ,M -NA #, A. I !

c. Comparator HA S/U ,M l Q g I

TA8LE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION .

ACTION ~

, ACTION 70 -

a. With one of the required monitors inoperable, place the inoperable channel in the downscale tripped condition l within aneihour; restore the inoperable channel to I l OPERABLE status within 7 days, or, within the next 5 l hours, initiate and maintain operation of the control room emergency filtration system in the pressurization i

mode of operation.

b. With both of the required monitors inoperable, initiate .

and maintain operation of the control room emergency filtration system in the pressuri:ation mode of operation within Amer hour.

l

.1 8

l l

l .

l I

C.'.

LA SALLE - UNIT 1. 3/4 3-58 l -

l guy .ee w e- ene e me .6-e em 7 _. r .-,n.-e -.--*e --

,-,,r

. - -. =.

l INSTRUrtENTATION .

() SEISMIC MONITORDIGTINS' RUMENTATION #

LIMITING CONDITION FOR OPERATION 3.3.7.2 The seismic monitoring instrume'ntation shown in Table 3.3.7.2-1 shall be OPERABLE."#

l l

[ APPt.ICABILITY: At all times.

ACTION:

a. With one or more seismic monitoring instruments inoperable for more than 30 days, in lieu of any other report required by Specification 6.6.8, prepare and submit a Special Report to the Comission pursuant i

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 OPERABLE status. -

I

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

SURVEILLANCE REQUIREMENTS v

4.3.7.2.1 Each of the above required seismic monitoring instruments shall be .

demonstrated CPERABLE by the performance of the CHANNEL CHECX, CHANNEL Table 4.3.7.2-1). FUNCTIONAL TEST and CHANNEL CALIBRATION operations at( the freq t

l 4.3.7.2.2 Each of the above required seismic monitoring instruments actuatad i during a seismic event greater than or equal to 0.02g shall be restored to

OPERA 2LE 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 saismic. event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground

} zotion. In lieu of any other report required.by Specification 6.6.B, a Special l .

Report shall be prepared and submitted to the Commission pursuant to Specifica-l tion 6.6.C within 10 days describing the magnitude, frequency spectrum and I

resultant affect upon unit features important to safety.,

e t

i i

! *"The nor al or emergency power source may be inoperable in OPERATIONAL I

l CONDITION 4 or 5 or when defueled.

% Se% o Mco.'4o r: .L s k w t ph x 5'rs W .s ska(d.

O% l-ASalle. 00 $. cad LaSalle. Un.+ 9, LA SALLE - UNIT 1 3/4 3-60 dt--- , -r--- , - - ,y, . . - . - , , , - . , ,-,-.-r----,. - - - , - - - - . - - . .w--- . ,,_.-..-...------w-------.- - - - _ - - - - - - - - , - -

i.

s.

INSTRUMENTATION V

k METEOROLOGICAL HONITORING I.1STRUMENTATION LIMITING CON 0! TION FOR OPERATION 4

3.3.7.3 The meteorological monitoring instrumentation channels shown in Table l 3.3.7.3-1 shall be OPERA 8LE.* * -

1 APPL,ICABILITY: At all times.

ACTION:

a. With one or more meteorological monitoring instrumentation channels inoperable for more than 7 days, in lieu of any other report required by Specification 6.6.8, 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 OPERABLE status. -

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

SURVEILLANCE REQUIREMENTS 4.3.7.3 Each of the above required meteorological acnf tering instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECX and CHANNEL CALIBRATICN operations at the frequencies shown in Tacle 4.3.7.3-1. .

i i

4"The normal or emergency power source may be inoperable in OPERATICNAL

CONDITICN 4 or 5 or when defueled.

o rd lot e'cA I IM cN'.h .~ngt A sfr a N M N h10m is Sho rad l Nh% La% :. Un.k $ ok1.Ahlb. Us.'4 0.

4 LA SALLE - UNIT 1 3/4 3-63

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

\

l, j

I (} }

a I ,

4 r TABLE 3.3.7.5-1 5

ACCIDENT MDNITORINC INSTRt24ENTAT1011

^

! W

[ REQUIRED MINiltti 4 ; , HulBER Of CilANNELS i

e g CilANNELS OPERA 8tE .

I 4 1. Heactor Vessel Pressure 2 1 e -

l l 2. Reactor Vessel hier Level 2 1

3. Suppression Chasher hier Level . 2 1 '

4. Suppression Chamber hter Temperature 7,1/wel 7,1/wel I i 5. Suppression Chamber Air Temperature 2 1

]

firywell Pressure

6. 2 1 i

j M* 7. urywell Air Temperature i 2 1 y

8. iisyaell Oxygen Concentration

2 1

'j, 9. Drywell Hydrogen Concentration Analyzer

and Monitor -

2 1

,.l l 10. Primary Containment Gross Gamma Radiation 2 1 l

11. Safety / Relief Valve Position Indicators 1/ valve 1/ valve I ll 1 t

12. Hohle Gas Monitor, Main Stack

i ,

i 1 1 e

13. Ilotele Gas Monitor, Standby Gas Treatmant System Stack 1 1 '

1

.i f.

(! Actuated after LOCA.

i R L: requirement. -Final requirement.to be determined aftee-demonstratten-of-correlattepf-peel-buiti te . ,. . : tur: :: =::: :d by rd d!:!:!:: t ; r! 5:!$

cer:*::: :: =rered-by-both-divistens.-Results (

of skumnstraLion-and.necessary chavs to--this-spectiIsation-shalI-be-subelL4ed-to-the-Gesels !:: ".:!t!;!n- -

90.al.apwf-demonsteaMem b f

O

e f

INSTRUMENTATION I I

SOURCE RANGE MONITORS LINITING CONDITION FOR OPERATION 3.3.7.6 At least three source range monitor channels shall be OPERA 8LE.

APPLICA8ILITY: I OPERATIONALCON0!TIONS2*,3)and4.

ACTION:

a. In OPERATIONAL CONDITION 2* with one of the above required sources 4(,,e, g range monitor channels inoperable, restore at leastJ ~ source range '

monitor channels to 0PERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HDT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l b. In 0PERATIONAL CONDITION 3 or 4 with two or more of the above required

! source range monitor channels inoperable, verify all insertable control l-rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within,,ons hour. ,

l 1

SURVEILLANCE REQUIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERA 8LE by:

I a.

Performance of a:

(

, 1. CHANNEL CHECK at least once per:

a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in CONDITION 2", and b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in CONDITION 3 or 4.

1

2. CHANNEL CALIBRATION ** at least once per 18 months.

b. Performance of a CHANNEL FUNCTIONAL TEST:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to moving the reactor mode switch from the Shutdown position, if not performed within the previous 7 days, and l 2. At least once per 31 days.

c. Verifying, prior to withdrawal of control rods, that the SRN count rate is at least 0.7 cps % ith the detector fully inserted.

l "With IRM's on range 2 or below.

- Neutron detectors may be excluded from CHANNEL CALIBRATION.

b'tuded. Sgod coisc rate. s 2 2..

04Aeusd. , 3 eps. l LA SALLE - UNIT 1 3/4 3-72 Amendment No. 2 l

l .- .

. INSTRUMENT # TION l

8ADIDACTIVE LIQUID EFFi.UENT MONITORING INSTRUMENTATION

( . .

l LIMITING CONDITION FOR OPERATION i

3.3.7.10 The radioactive liquid effluen.t monitoring instrumentation channels shown in Table 3.3.7.10-1 shall be OPERA 8LE with their alars/ trip setpoints l set to ensure that the limits of Specification 3.11.1.1 are not exceedec. The alars trip setpoints of these channels shall be determined in accordance with the Offsite Oose Calculation Manual (00CN).

l APPLICA8ILITY: At all times.

! ACTION:

l

a. With a radioactive liquid effluent monitoring instrumentation channel -

alam/ trip setpoint less conservative than required, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable. .

b. With less than the minimum number of radioactive ifquid affluent .

monitoring instrumentation channels OPERA 8LE, take the ACTION shown

. in Table 3.3.7.10-1.

c. The provisions of Specifications 3.0.3 and 3.0.4-are.not appitcable.

SURVEILLANCE REQUIREMENTS l 4.3.7.10 Each radioactive liquid effluent monitoring instrumentation channel

! shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.10-1.

. Restore the inoperacle instrumentation to

' ! OPERA 8LE status within the time specified in the ACTICN or, in lieu of a Licensee Event Report, explain in the next Semiannual l ! I Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

j

. LA SALLE - UNIT 1 3/4 3-81 m -

--wa-- -.-e-+- ,.,--,.-,r- .,r * - -- -- --,--m. --e--,-e-e-+--w-w -e -ese *esw ww.se----m -- - - - -w- e v- =- mea--m +-ars - *- rw - --- - - --

g TABLE 3.3.7.10-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION e

e MININUM 5

CHANNELS INSTRUNENT w OPERABLE ACTION

1. GAMA SCINTILLATION MONITOR PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line (001" "^^0) I 1 100

2. GAM 4 SCINTILLATION MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE 4

, s. Service Water System Effluent Line-(101" "0^0) (Unit O 1 101 l

} b. RHR Service Water (Line A) Effluent Line (1010 %500) 1 101 m c. RHR Service Water (Line B) Effluent Line-(1013-;665}- 1 101 i

a m

d. Service Water System Effluent' Line -fe-916-M608)-- (va;+ g) 1 . 101 l l i

3. FLOW RATE MEASUREMENT DEVICES -

a. Liquid Radwaste Effluent Line -(OTIT '.' 017 ...J 010) 1 102

b. River Discharge - 810wdown Pipe (Ori '4001) 1 102 l

^

E

3

, n

On

}

i i

l . = . . .

' ~

~

l fS INSTRLMEWfATION ,

~'

i TABM 3.3.7.10-1 (tllontinued)

~

/

TA8 G NOTATION

.. ~

ACTION 100 -

With the number of QPERABG channels less than requimd by the Minfeum Channels OPERAB M requirement, effluent releases may continue for up to 14 days provided that prior to initiating a release:

a. At least two independent saspies are analyzed in accordance with Specification 4.11.1.1.3, and

b. At least two technically qualified members of 'the i '

Facility Staff independently verify the release rate l calculations and discharge line vstving; Otherwise, suspend release of radioactive affluents via this pathway. ,_

ACTION 101 -

With the number of channels CPERA82 less than required by the Minimus Channels OPERA 8 2 requirement, effluent

_ releases via this pathway may continue for up to 30 aays '

provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are

. collected and analyzed at a limit of detection of at least 3

(/ 10-7 microcurief7a1 or gamma spectrometric analysis. l ACTION 102 -

With the maber of channels OPERABLE less than required I

by the Minimus Channels OPERA 8 2 requirement, effluent l releases via this pathway any continue for up to 30 days j provided the flow rate is estimated at least once per j 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves for Instrt:-

i sent 3a, or for known valve positions for Instrument 3b, i may be used to estimate flow.

e O

t

.R 1

u -

0 l

1 LA SALLE - UNIT 1 3/4 3-83 .

,,---p.- w ,-w,,,-- ,y-,w- y - - - , g 7 gw,--w----4-*..y --w-- --- y -.---e g. - --e* - *e--- _ _ - -

g TABLE 4.3.7.10-1

% RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

,E ,

c CHANNEL 5

-4 CHANNEL SOURCE FUNCTIONAL CHANNEL INSTRUMENT CHECK CHECK TEST CALIBRATION s

1. GAMA SCINTILLATION MONITOR PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluents Line D P Q(1) R(3)

2. GAMA SCINTILLATION MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE

$ a. Service Water System Effluent Line (Uvif f) .

, 4 1919 8599)- D M Q(2) l,

b. R(3)

I RHR Service Water (Line A) Effluent Line O M .Q(2) R(3)

c. RHR Service Water (Line B) Effluent Line D M

d. Q(2) R(3)

Service Water System Effluent Line (Ugif 2)

(2010 K666T D

[

M Q(2) R(3)

3. FLOW RATE MEASUREMENT DEVICES

)

a. Liquid Radwaste Effluent Line O(4) N.A.

b. River Discharge - Blowdown Pipe Q R D(4) N.A. Q R 2

3 0%

~ ~ ~ ~ ~ ~ ~ '

INSTRLDIENTATION .

, TABLE 4.3.7.u-1 (Continued) a TABLE NOTATION

~

At all times.

During main condenser offgas treatment system operation.

During operation of the main condenser air ejector.

N Ouring operation of the 58GT5.

4kt.

(1) The CHANNEL FUNCTIONAL TEST shall also denonstrate Det automatic isoTation cpl 4g of this pathway,and76ntrol room alars annunciation occurs if any'"~ "of "the L

'*d'd %'".*E*"' % ' #,it ^*;1:M is. " " *f*" ' " #

Instrument i catesmeasuredNevelsabovethealars/tripsetpoint. -

L Loss of power. ,

3. Instrument alares on downscale failure.

I": $' O $ $ $ dY $ u % \*$a

W U Os\ " $ s b 5 * * **

(2) The CHANNEL FUNCTIONAL TEST for tN1og scale monitor shall also >-

i demonstrate that control room alars annunciation occurs if any of the

. following conditions axists:

1. Instrument indicates measured levels above the alars setpoint.

2. Loss of power. '

3. Instrument alarms on downscale failure.

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

l .

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of i

the reference radioactive standa'rds certified by the National Bureau of g3 standard 4cr using standards that have been obtained from suppliers that [

', participate in measurement assurance activities with N85. These standards shall permit calibrating the system over its intended range of energy and

, measurement range. For subsequent CHANNEL CALIBRATION, the initial reference radioactive standards or radioactive sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: '

1. One volume percent hydrogen, balance nitrogen, and

! 2. Four volume percent hydrogen, balance nitrogen.

I (5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control roon l alarm annunciation occurs if any of the following conditions exitts:

1. Instrument indicates . measured levels above the alar.. setpoint.

2. Circuit failure. '

3. Instrument controls not set in the Operate mode.

i LA SALLE - UNIT 1 3/4 3-90 "

.._.e.~ . .

. _ . , _ _ _ _._y..,y_m.. . _ . _ y ,_._m_ _ . - _ , _ , , . ,__,,._,_ ,__. _ . . _ _ _ _ _ _ _ . ,

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RE RCULATION SYSTEM 'JNSEM 4

( g4 4-l, YYY~j~

RECIECULATION PS LIMITING CONDITION OR OPERATION 3.4.1.1 Two reactor co lant system recirculation loops shall be in operation.

l APPLICABILITY: OPERATION CONDITIONS 1* and 2*.

ACTION:

a. With one reactor coolant stem recirculation loop not in operation, be in at least HOT SHUTD0 ithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With no reactor coolant system ecirculation loops in operation, place the reactor acde switch i the Shutdown position.

SURVEILLANCE REQUIREMENTS i.

x

, 4.4.1.1 Each reactor coolant system recirculation loop flow control valve j shall be demonstrated OPERABLE at least once per 18 nths by:

a. Verifying that the control valve fails "as i " on loss of hydraulic pressure at the hydraulic power unit, and

, b. Verifying that the average rate of control valve ovement is:

1. Less than or equal to 11% of stroke per secon opening, and l 2. Less than or equal to 11% of stroke per second c sing.

"See Special Test Exception 3.10.4.

i t

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

l

[

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM l RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION l 3.4.1.1 Two reactor coolant system rect'rculation loops shall be in operation. i i

APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2*. l l

l ACTION:

l

a. With one reactor coolant system recirculation loop not in operation: l l 1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

r l l a) Place the recirculation flow control system in the Master Manual mode, and l b) Reduce THERMAL POWER to < 50% of RATED THERMAL POWER, and,

< c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety l- Limit by 0.01 to 1.07 per Specification 2.1.2, and, d) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation by 0.01 per Specification 3.2.3, and, e) Reduce the MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) limit to a value of 0.85 times the two recirculation a loop operation limit per Specification 3.2.1, and, f) Reduce the Average Power Range Monitor ( PRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single loop recirculation l loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6.

2. At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

l a) Verify that the APRM flux noise averaged over 30 minutes i does not exceed 5% peak to peak; otherwise, reduce the recirculation loop flow until the APRM flux noise is less than the 5% peak to peak limit, and, b) Verify that the core plate .iP noise does not exceed 1 psi peak to peak; otherwise, reduce the recirculation loop flow until the AP noise is less than the 1 psi limit.

I See Special Test Exception 3.10.4. l l l l LASALLE-UNIT /j, 3/4 A-1 i l

y,, , ,. . , , . .p,- _,,.,~._p.-._. _ , , . . -

__.s-.r, ,,v- , . , ,, ,,, , , . ,-y,,,,n. ,..

, . o f.- REACTOR COOLANT SYSTEN s .. -

LINITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

3. The provisions of Specification 3.0.4 are not applicable.

4. Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With no reactor coolant system recirculation loops in operation, immediately initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERA 8LE at least once per 18 months by: ,

a. Verifying that the control valve fails "as is" on loss of hydraulic pressure at the hydraulic power unit, and b

b. Verifying that the averagIn rata of control valve movement is:

1. Less than or equal to 11% of stroke per second opening, and .

2. Less than or equal to 11% of stroke per second closing.

l l

e 4

L t

5 LA SALLE - LNITf3, 3/44-fla.,

- --.-~ - --

.~ REACTOR COOLANT $YSTEM

.[ 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be Ifaited to: ~

a. Less than or equal to 0.2 afcrocurie/ per gram DOSE EQUIVALBT l I-131, and

b. Less than or equal to 100/E aferocuries per gram.

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

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of the priman 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 ,

operatingtimeataprimarycoolantspecificactivitygreater than or equal to 0.2 aicrocuri p per gram DOSE EQUIVAlt.NT 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 repcrt to the Commission pursuant to Specification 6.6.C within 30 days indicating the number of hours of operation above this Ifait. The provisions of

! Specification 3.0.4 are not appifcable.

i

2. Greater than 0.2 aferocuries 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 acre than 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months cumulative operating time in a consecutive 12-month period or greater than or equal to 4A microcuries per l i gram,beinatleastHOTSHUTDOWNwiththemainsteamline isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c -

l 3. Greater than 100/E aicrocuries per gram, be in at least HOT t

SHUTDOWN with the asin steaaline isolation valves closed within l

j 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity

of the primary coolant greater than 0.2 microcurief per graa DOSE l l EQUIVALB T I-131 or greater than 100 E aferocuries per gram, perfors the sampling and analysis requirements of Itaa 4a of Table 4.4.5-1 f until the specific activity of the primary coolant is restored to

! I within the Ifait. A REPORTABLE OCCURRENCE shall be prepared and t submitted to the Commission pursuant to Specification 6.6.8. This report shall contain the results of the specific activity analyses and t.he time dur3 tion when tha soecific activity of the coc'a .t A,-

exceeded C.: sicrocurief :er gram DOSE EQUIVALEhr I-131 toga .'.ar )

with the fc;i: wing adaitional infor:natien. ,

i

! LA SALLE - UNIT 1 3/4 4-13

_ , _ _ , - - - _ , ,, .,.._-.. _ ,. __y,_,...,-_.,w, -,.y- .e.w, , ., , , - _- e_ ,,_.__.__,___.,,m..

~

q _. - t -

+ . .. ,. w __-. -

.a ._,,

i, REACTOR COOLANT SYSTD -

LIMITING CONDITION POR OPERATION (Continued)

ACTION (Continued)

c. In OPERATIONAL CONDITION 1 or 2, with:

1. THERMAL POWER changed by more than 15% of RATED THERMAL POWER in ;

houra , or l l 2. The off gas level, prior to the holdg ifne, increased by more l

than 25,000 microcuries per second in one hour during steady state operation at release rates itss than 100,000 microcuries per secsad, or

3. The off gasi level, prior to the holdup line, increased by more than 15% in pe hour during steady state operation at release I rates greater than 100,000 microcuries per second, perform the sampling and analysis requirements of Ites 4b of l Table 4.4.5-1 until the specific activity of the primary coolant is :

. restored to within its limit. Prepare and submit to the Commission a Special Report pursuant to Specification 6.6.C at least once per -

92 days containing the results of the specific activity analysis

. together with the below additional information for each occurrence.

Additional Information

1. Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to: -

a) The first sample'in which the limit was exceeded, and/or -

b) The THERMAL POWER or off gas level change.

l

2. Fuel burnup by core region. .

i

3. Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to:

j, a) The first sample in which the Itait was exceeded, and/or l b) The THERMAL POWER or off gas ie, vel change.

j 4

4. Off gas level starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to:

a) The first sample in which the Ifnit was exceeded, and/or i

b) The THERMAL POWER or off gas level change.

1 SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to

}

be within the limits :y cerformance of the sampling and analysis progru: of Table 4.4.5-1.

Not applicable during the Startup Test Progras.

LA SALLE - UNIT 1 3/4 4-14 l . .

i 1 _ _ _ _ _ _ _ - . _ . . _ _ _ _ _.. _ . _ . _ . _ _ _ _ __ _ , _ . _ . . _ _ _ _ _ _ _ _

REACTOR COOLANT SYSTEM JET PUMPS ,

(' LIMITING CONDITION FOR OPERATION

3. 4.1. 2 All jet pumps shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS-

4. 4.1. 2 EachoftheaboverequkedjetpumpsshallbedemonstratedOPERABLE prior to THERMAL POWER exceeding 5% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by measuring and recording each of the below specified parameters v'

\ and verifying that no two of the fb31owing conditions occur when the recircula-l tion loops are operating at the same flow control valve position.

i l a. Theindicatedrecirculatio\ n\ loop flow differs by more than 10% from theestablishedflowcontrol\valveposition-loopflowcharacteristics.

g b. The indicated total core flow \ differs by more than 10% from the established total core flow value derived from either the:

W5arD goJ9 1. Established THERMAL POWER c e flow relationship, or fT 2. Established core plate diffe ntial pressure-core flow relationship.

l c. The indicated diffuser-to-lower plen differential pressure of any l individual jet pump differs from estab ished patterns by more than 10%.

l l

l f

l LA SALLE - UNIT 1 3/4 4-2 L

4.4.1.2.1 Each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least cnce per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by measuring and recording each of the below specified parameters end verifying that no two of the following conditions occur when both recircula-tion loops are operating at the same flow control valve position. .

a. The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characteristics for two recirculation loop operation.

b. The indicated total core flow differs by more than 10% from the established total core fis value derived from either the:

1. Established THERMAL POWER-core flow relationship, or

2. Established core plate differential pressure-core flow relationship for two recirculation loop operation.

c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs frow established two recirculation loop operation patterns by more than 10%.

4.4.1.2.2 During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that no two of the following conditions occur:

a. The indicated recirculation loop flow in the operating loop differs by more that 10% from the established single recirculation flow control valve position-loop flow characteristics.

i

b. The indicated total core flow differs by more than 10% from the established total core flow value from single recirculation loop flow measurements.

c. The indicatedNZheR. ..": ;.ee to-lower plenum differential pressure of any individual jet pump differs from established single recirculation loop y more than 10%.

Oftdans p Heras zuseer eae ea6a

44 1

1 I

r

,,, . . - , , , _ _ , . _ _ , , .,_.,,,,,..e,.,.e_.%, . . , , , , , - _ _ . _ . . , , . , , . . - ,

e -.

--r. m, _y , .- , .

-~

_.... p. . . . . . . .

. i '

REACTOR CDOLANT SYSTEM '

S ..

. ,f s . -

RECIRCULATION LOOP FLOW LDtITING CONDITION FOR OPERA ~ ION i

3.4.L 3 Racirculation loop flow mismatdh shall be maintained within:

a. 55 of rated recirculation flow with core flow greater than or equal to 7GE of rated core flow.

b. 105 of rated recirculation flow with core flow less than 7CE of rated core flow.

APPLICA8ILITY: OPERATIONAL CONDITIONS I and 2. Jur.ag Ne recirculdi'*8" loop ope raison. . i ACTION: . .

With recirculation loop flows different by more than the sp_ecified Itaits, either: *

a. Restore the recirculation loop flows to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or ~

b. Declare the recirculation loop with the lower flow not in operation and taka the ACTION require by Specification 3.4.1.1.

I t

i SURVEILLANCE REQUIREMENTS i

i 4.4.1.3 Recirculation loop flows shall be verified to be within the limits at l 1 east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I i

l I '

I.

l lI

(

l .

s) l .

1 l 'A SALLE.- UNIT 1 3/4 4-3 '

l . .

f

... , , , . - . , r-, ,- ., , y_,.,, c-.,._ _ ,.,,~

g REACTOR CDOLANT SYSTEM i

3/4.4.2 SAFETY / RELIEF VALVES .

IntITIIC CONDITION FOR OPERArt0N __

i .

~

' 3.4.2 The safety valve function of eigtiteen reactor coolant systes safety /

relief valves s a11 be OPERABLE with the specified code safety valve function lift settings.j

a. 4 safat lief valves 91205 psig 13

b. 4 safe Ifef valves 9 1195 psig 2 3

c. 4 safe ifef valves 91185 psig 2 3

d. 4 safet / lief valves 9 1175 psig t 1%

. e. 2 safe Ifef valves 9 1146 psig 2 3 l _ ~ APPLICA8ILITY: 0 . RATIONAL CONDITIONS 1, 2 and 3.

ACTION: '

7

a. With the safety valve function of one or more of the above required I

safety /reifef valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more safety / relief valves stuck open, provided that ~

suppression pool average water taaperature is less than 110*F, close the stuck open relief valve (s); if unable to close the open valve (s) within 2 minutes or if suppression pool average water tamperature is 110*F or greater, place the reactor mode switch in the Shutdown position.

. c. With one or more safety / relief valve stes position indicato_rp inoperable, restore the inoperable stem position indicatio#to OPERA 8LE status within 7 days or be in at least HOT SHUTDOWN within l l

j the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l } SURVETLLANCE REQUIREMENTS i

l 4.4.2.1 The safety / relief valve stem position indicators of each safety / relief

, valve shall be demonstrated OPERABLE by perfomance of a:

.j a. CHANNEL CHECK at least once per 31 days, and a

b. CHANNEL CALIBRATION at least once per 18 sonths.** ,

4.4.2.2 The low low set function shall be demonstrated not to interfere with

) ! the OPERA 8ILITY of the safety relief valves or the ADS by performance of a

i CHANNEL CALIBRATION at ,least once per 18 months I

9 l ; "The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temoeratures anc pressures.

fUp t.) two inocerable valves may be solaced with spare JPERABLE valves with

, lower setpoints until the next' refueling outage.

**The provisions of Specification 4.0.4 are not applicable provided the surveil-

{ 1ance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate j to perfore the test.

LA SALLE - UNIT 1 3/4 4-5 .

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

REACTOR COOLANT SYSTEM ,

g ,

OPERATIONAL LEAKAGE

~

. LIMITING CONDITION FOR OPERATION

3.4.3.2 Reactor coolant system leakage shall be limited to:

,,' a. No PRESSURE SOUNDARY LEAKAGE.

b. 5 gpa UNIDENTIFIED LEAKAGE.

c. 25 gpa total leakage averaged over an'y 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. g

. d. 1 gpa leakage at a reactor coolant system pressure at 1000 i)6 psig \

from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1. _

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION: *

a. With ~any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With any reactor coolant system leakage greater than the limits in b and/or c. above, reduce the leakage rate to within the limits within l 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With any reactor coolant system pressure isolation valve leakage g greater than the above limit, isolate the high pressure portion of

. N. the affected system frow the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed valves, or be in at least HOT SHUTDOWN I within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With one or more high/ low pressure interface valve leakage pressure monitors inoperable, restore the inoperable monitor (s) to OPERABLE

! status within 7 days or verify the pressure to be less than the alare setpoint at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local indication; restore the inoperable monitor (s) to OPERABLE status within 30 days l or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD l SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

a. Monitoring the primary containment atmospheric particulate and

. gaseous radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. Monitoring the primary containment sump flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

c. Monitoring the primary containment air coolers condensate flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

LA SALLE - UNIT 1 3/4 4-7

el

. i,

~

REACTOR CDOLANT SYSTEM

\, <

SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit ifne of Figure 3.4.6.1-1 curves C and C' within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality.

4.4.6.1.3 The reactor vesset material specimens shall be removed and examined to determine reactor pressure vessel fluence as a function of time and THERMAL t

POWER as reeuired by 10 CFRja0, Appendix H in acc;prdance with the schedule in g

! M able 4.4.5.1.3-1. The results of these fluence determinations shall be used Part toupdatethecurvesofFigurey3.4.6.1-1 0 40.0.1-24.4.0.1. l 4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80*F: '

~

a. In OPERATIONAL CONDITION 4 when the reactor coolant temperature is:

l

1. $ 100*F, at least once per lit hours.

, 2. S 85'F at least once per 30 minutes.

/) b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

1 .

l '

i I

l .

l l

i t

^g 1

l .

LA SALLE - UNIT 1 3/4 4-17

(.

./

E g Table 4.4.6.1.3-1 r-E Reactor Vessel Material Survelliance Program Withdrawal Schedule E

q Specimen holder Vessel location Lead factor

~

Withdrawal time

(%-~'ce Y::=}- @ffECTWC t=UU-117C4936G010 300* 0.6 ,.le- 6 117C4936G011 120* 0. 6 117C4936G012 30*

E IS 0.6

Spare, Neutron Dosfeeter 30* -

1st Refuel Outage M

Jun 3

9

, 3

\

<a

4

, . o .

' i

~

g REACTOR C00LANT SYSTDI -

! N

. , 3/4.4.9 RESTDUAL HEAT REMOVAL -

HOT SHUTDOWN

.a LDtITING CONDITION FOR OPERA'. ION e

3.4.9.1 Two# shutdown coo 11ng mode loops of the residual heat removal (RHR) system shall be OPERA 8LE and at least one shutdown cooling mode loop shall be '

inoperation*W with each loop consisting of at least: I

a. One OPERABLE RHR pump, and

b. One OPERA 8LE RHR heat exchanger.

APPLICA8ILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint. -

~

ACTION:

a. With less than the above required RHR shutdown cooling mode loops OPERABLE, inmediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Withinpadhourandatleastonce I

~

per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operabtitty of at least one alternate method capable of decay heat removal for each inoperable RHR

, shutdown cooling mode loop. Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**

b. With no RHR shutdown cooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible. Within hour establish reactor coolant circulation by an l alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE REQUIREMENTS l -

4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal Li

' system or alternate method shall be determined to be in operation and circulating

reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I i t e

j. #0ne RHR shutdown cooling mode loop say be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for

surveillance testing provided the other loop is OPERA 8LE and in operation.

l "The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

, per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

l NThe RHR shutdown cooling mode loop say be removed from operation during j hydrostatic testing.

! **Whenever two or more RHR subsystems are inocerable, if unable to attain COLD

}* SHijiCC'aN as required by this AC I:N, t.afntain reacter ceolant tamperature as

,, low as practical by use of alta- .a a .. sat removal met.' ces.

i 8

/

j i LA SALLE - UNIT 1 3/4 4-23 y emese e _m- o gymme w .go e mumm e to m * * * . * *

. ee g **

w - w.-- w g i.-.m-, 9.,.,,mm,. - . , _ , -- , . - -,mm,y..-,-,-,u, - , , - , , -.,_w, m,ww. ,..,,--,,w, - -_+- .--. . .. _, - - - - - ,

, ... , ;._ . .. ( , ~ ,.. ; .. ] . ,,,, , , , , , , , , , ., , , , . . , , , , . . . .,} ,, ,,

l

i, REACTOR COOLANT SYST94 s

NLh * .

LDlITING CONDITION FOR OPERATION 3.4.9.2 Two# shutdown cooling mode loopa of the residual heat removal (RHR) l systes shall be OPERA 8LZ" and at least one shutdown cooling mode loop shall be in operation with each loop consisting of at least: I

a. One OPERA 8LE RHR pump, and l

l b. One OPERA 8LE RHR heat exchanger.

APPLICA81LITY: OPERATIONAL CONDITION 4.

ACTION: ~

a. With less than the above required RHR shutdown cooling mode loops OPERABLE, within hour and at least once per*24 hours thereafter, demonstrate l the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop. -

b. 1

, With no RHR shutdown cooling mode loop in operation, within. ped hour 1 i establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once ped hour.

SURVEILLANCE REQUIREMENTS, 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and

.j circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

t -

l l , 0ne RHR shutdown cooling mode loop may be inoperabie for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERA 8LE and in operation.

j

, "The normal or emergency power source may be inoperable.

- The shutdown cooling pump may be removed free operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

MThe shutdown coclin; : ode loop may be removed from ocerstion during

, hydrostatic tasti.g.

e' LA SALLE - UNIT 1 3/4 4-24

,,% o. m se e m+ 6 N 8 * "

w- w >p. o. gw-- y-n gy-4 --~~<yy ,.

l EMERGENCY CORE CDOLING SYSTEMS

, LDUTING CON 0TTION POR OPERATION (Continued) l ACTION: (Continued)

d. For ECC5 divisions 1 and 2, provided that ECCS division 3 is OPERA 8LE:

, l 1. With LPCI subsystas "A" and af ther LPCI subsystem "8" or "C"

' inoperable, restore at least the inoperable LPCI subsystas "A" or fnoperable LPCI subsystem "8" or "C" to OPERA 8LE status within j M km.

2. With the LPCS system inoperable and either LPCI subsystems "8" or "C" inoperable, restore at least the inoperable LPCS system or inoperable LPCI subsystas "8" or "C" to OPERA 8LE status within i

l 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. ..Oth.grwise be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in 15 ldh within tarfo-11owing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *. .

l

e. For ECCS divisions 1 and 2, provided that ECCS division 3 is OPERA 8LE and divistans 1 and 2 are otherwise OPERA 8LE:

', 1.

With one of the above required AOS valves inoperable, restore

the inoperable ADS valve to OPERA 8LE status within 14 days or be in at least HDT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to~ < 122 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

v 2. With two or som of the above required AOS valves inoperable.

l '

be in at least HDT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to 5122 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . <

I -

f. With an ECCS discharge Ifne "kaep filled" pressure alam instrumenta-tion channel inoperable, perfom Surveillance Requirement 4.5.1.a.1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1

g. With an ECCS header delta P instrumentation channel inoperable, restore the inoperable channel to CPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

! or determine ECCS header delta P toca11y at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; j otherwise, declare the associated ECCS inoperable.

h. With survef11ance Requirement.4.5.1.d.2 not performed at the required interval due to low reactor steam pressure, the provisfens of Specifi-

! I catien 4.0.4 are not appifcable provided the survefliance is perfomed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor staas pressure is adequata to perfom l l the test. .

l !

t

i. In the event an ECC5 systas is actuated and injects water into the Reactor Coolant Systam, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.6.C within

, 90 days describing the circumstances of the actuation and the total

accueuiated actuation cycles to date. The current value of the j -

,, usage factor for each affected safety infection nozzle shall be

providea in tais Special Re: or: . T:ever its vs.h;e ex:ae:s 0. 70. g,3,g V

.s "knanever t o or ., ore idiR subsystems are inaparacia, if w,acle La attain COLD $

SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of af ternate heat removal methods. ] C-

! LA SALLE - UNIT 1 3/4 5-3 s- ** *

-.-r,

, -- ,,""Mlllle

--,7 ..,-n-,---, ,, .,--,r-,--,- --,m-n.-,-

"IAJSet.T i:o e. PA6-e 3/4 s-3 j ., With one or more ECCS corner room watartigtn doors inoperable,'resto'e r

all the inoperable ECCS corner room watertight doors to OPERABLE ~~'i-

. status within 14 days, othenvise, be in at least HOT SHUTDOWN'within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the followihg 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i l

l l

l t

I 1

l

n. k .:. u_-. . . -, ; , . .: z.,

.. .. ~ . , . . . . . . . . . ^.... .

, a 5

EpKRGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS l 4.5.1 ECC5 divisions 1, 2 and 3 shall be demonstrated OPERA 8LE by:

a. At least once per 31 days for .the LPC5, LPCI and HPCS systems:

?

1. Verifying by venting at the high point vents that the systes piping from the pump discharge valve to the system isolation l valve is filled with water.

2. Performance of a CHANNEL FUNCTIONAL TEST of the:

a) Discharge line " keep filled" pressure alarm instrumentation, and b) Header delta P instrumentation.

3. Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. -

i b.

, Verifying that, when tested pursuant to Specification 4.0.5, each:

1.

LPC5 pump develops a flow of at least 6350 gpa against a test line pressure greater than or equal to 290 psig.

2.

LPCI pump develops a flow of at least 7200 gps against a test

! line pressure greater thu or equal to 130 psig.

3. HPCS pump develops a flow of at least 6250 gpe against a test line pressure greater than or equal to 370 psig.

c. For the LPCS, LPCI and HPCS systems, at least once per 18 months:

j { 1. Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolar.t into the reactor vessel say be. excluded from this test.

I

4. Verifying that each ECCS corner room watertight door is closed, except during encry to and exit from the room.

m l

i LA SALLE - UNIT 1 3/4 5-4 l .

, ,.,,,,y ,- _ _ - . _ , , . . . r.e, - - - - - - - -

. . Ju - -

i EMERGENCY CORE' COOLING SYSTEMS .

3. .

SURVEILLANCE REQUIREMENTS (Continued)

~

2. Perfoming a CHANNEL CALIBRATION of the:

a) Discharge line " keep filled" pressure alam instrumentation and verifying the:

1) - High pressure setpoint and the low pressure setpoint I

of the:

I (a) LPCS systes to be 1 500 psig and 1 55 psig, respectively.

(b) LPCI subsystems to be 1 400 psig and > 55 psig, respectively. -

2), Low pressure setpoint of the HPCS system to be 1 63 psig.

b) Header delta P instrumentation and verifying the setpoint .

of.the:

1) LPCS system and LPCI subsystems to be 1 1 psid.

'. a. 0

2) HPCS system to be 5 s.f,4 psid greater than the l nomal indicated .iP.

3. Verifying that the suction for the HPCS system is automatically transferred from the condansate storage tank to the suppression chamber on a condensate storage tank low water level s,ignal and on a suppression chamber high water level signal.

y susui"

d. For the ADS by: peda#

1. At least once per 31 days, perfoming a CHANNEL FUNCTIONAL TEST

, of the accumulator backup compressed gas system low pressure alam system.

t

2. At least once per 18 sonths: ,

a) Performing a system functional test which includes simulated 8

automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.

,' b) Manually opening each ADS valve and observing the expected 1 change in the indicated valve position.

' Performing a 0!ANNEL CALIBRATICN of ne ac umulator ::ackuc

_ c)

compressec gas system low pressure alarm system and verifying an alam setpoint of 500 + 40, - O psig on decreasing pressure.

. LA SALLE - UNIT 1 3/4 5-5 *

/

e w g O - = *

=w, ,,www-w- ,- ,----,--w,,- y---,w-q -, , , - , ,m- - - --g,+w----wg a t.g e--------eg-y y y- m--n-

l TNSetT FoK fggy 3/4 5.g 4.

Visually inspecting the ECCS corner room watertight door seals and roomorpenetration damage, seals obstructions. , and verifying no ab .arsal degradation, i

i l

, ;--_---,-------.n, , - - - r-,-., -,.-v. --.. ---, y. ..-

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

i.

' EMERGENCY CORE COOLING SYSTEMS ,

+ %' .

3/4.5.3 SUPPRESSION CHAMBER' LimimG CONDITION FOR OPERATTON

~

, 3.5.3 The suppression chamber shall be 4PERABLE:

a. In OPERATIONAL CONDITION 1, 2 or 3 with a contained water volume of at least 128,800 ft3, equivalent to a level of 26 821:$ip, l Si

b. In OPERATIONAL CONDITION 4 or 5* with a contained water volume of at 3

least70,000ft,equivalenttoalevelof14f0Vexceptthatthe {

suppression chamber level may be less than the limit or may be

, drained in OPERATIONAL CONDITION 4 or 5* provided that:

l l 1. No operations are performed that have a potential for draining the reactor vessel,

2. The reactor mode switch is locked in the Shutdown or Refuel position,

3. The condensate storage tank contains at least 135,000 r- ute .

gallons of water, equivalent to a level of 14.5 feet, and S 4. The HPCS system is OPERA 8LE per Specification 3.5.2 with an 0PERA8LE flow path capable of taking suction from the condensate storage tank and transferring the watt through the spray l sparger to the reactor vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4,and 5*. I

, ACTION:

i j a. In OPERATIONAL CONDITION 1, 2, or 3 with the suppression chamber water level less than the above Itait, restore the water level to i within the Ifnit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within i the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I

! b. In OPERATIONAL CONDITION 4 or 5* with the suppression chamber water

+

level less than the above limit or drained and the above required

! conditions not satisfied, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock the reactor mode switch in the Shutdown position. Establish SECONDARY

! CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . *

! j #5ee specification 3.6.2.1 for pressure suppression requirements.

I ! "The suoeression chamber is not required to be OPERABLE orovided that the

, reactor vessel head is removed, the cavity is floede er :eing ficocea

. . fr:i the suceression pool, the s::ent fuel pool gatas ir? reacved wnen t.9e ,

~~

, cavity is flooded, and the water level is maintained witnin the limits of

. Specifications 3.9.8 and 3.9.9.

l ;

t

! 1 LA SALLE - UNIT 1 3/4 5-8

~~

p 3.-. , . - . - - - -

y--y gme.-g----c+----r-- -

gw- , -

m ,-,.-w.ew---w-w-e--..w- . - ,-_m_ -- e -__ -- __. - . _ - - - _ _ - - - * --

s

5. ,

BEtGENCY CDRE COOLING SYST9tS s'

l LDGTING CONDITION FOR OPERATION (Continued) i~ .

M (Continued) i l c. With one suppression chuber water level instrumentation channel l - inoperable, restore the inoperable channel to OPERABLE status within 7 days or verify the suppression chamber water level to be greater l l than or equal to 2 ,,,or 14'0L as applicable, at least once per l l

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local i ication

!! - d. With boi.h sgpression chamber water level instrumentation channels inoperable, restore at least one inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the' next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and verffy the suppression chamber water level to be greater than or equal to 25 a or 14'0L as applicable, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i by Tocal i ication SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying: -

3 a.' The water Tevel to be greater than or equal to, as applicable:

1. 25,# 2 4at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

{

. 2. 14jo{at 1, east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l i

b. Two suppression chamber water level instrumentation channels OPERABLE by performance of a:

4

1. CHANNEL CHECX at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , I 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and i i i

j 3. CHANNEL CALIBRATION at least once per 18 months, I

with the low water level alars setpoint at greater than or equal to

26"4",;, == 1 1

' 4.5.3.2 With the suppression chamber level less than the above limit or drained in OPERATIONAL CONDITION 4 or 58, at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

l

a. Verify the required conditions of Specification 3.5.3.b to be l 3 satisfied, or

~

b. Verify footncte conditions

to be satisfied. -

3 - - - -iti:: :::::i ,:. c' :! ::eci-t : :: : :d e ' : "- . : : :: : ; :-

j 11 :ui-d ch- ; t- thf: ::t--f-t :h:11 5: :e.:itt:d t: t.t tri;;ir.

-w4n4 - n--an ^~ - ~* *__* - " .-+.4.....

4~ '.""Y

-7 LA SALLE - UNIT 1 3/4 5-9 m e g emesem ans . e.e gangne

zussef Fo

PAGB J/y 5- 9 "The suppression chamber is not required to be OPERABLE provided' that the rea'ctor vessel head is removed, the cavity is flooded or being f1'ooded from the ,suppres -

sion pool' the spent fuel pool gatas are removed when the cavity is flooded, and the water level is saintained within the limits of Speciff' cations 3.9'.8 and 3.9.9. -

l l

i l

l l

i-l .

. CONTAINMENT SYSTBes O

s/ PRIMARY CONTAIW9ENT LEAKAGE - - - - - - - - - - - - -

1 LD4ITING CONDITION FOR OPERATION

3. 6.1. 2 Primary containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less,than or equal to L,,

0.635 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P,,

39.6 psig.

h. A combined leakage rate of less than or equal to 0.60 L, for all penetrations and all valves Ifsted in Table 3.6.3-1, except for main steam isolation valves and Salves which are hydrostatically leak .

tested per Table 3.6.3-1, sibject.to Type 8 and C tests innen pressurized to P , 39.6 psig.

c. *Less than or ualtoNscfperhourfor .

main staas lives 4hroegkh isolation valve when tasted at 25.0 psig. 1

d. A combined leakage rate of less than or equal to 1 gpa times the total number of ECCS and RClc containment isolation valves in hydro-statically tested ifnes which penetrate the primary containment, when tested at 1.10 P ,, 43.6 psig.

APPLICA8ILITY: When PRIMARY CONTAINMENT INTEGRITY is required per i

Specification 3.6.1.1.

ACTION:

i .

.'. With: -

l[ a. The measured overall integrated primary containment leakage rate

'i exceeding 0.75 L,, or

'! b. The measured combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam isolation valves and valves which are hydrostatically leak tested per j Table 3.6.3-1, subject to Type 8 and C tests exceeding 0.60 L,, or i /dd The measured leakage rate exceeding 15 scf per hour for rj =fege. pcdro sfeaw all

. c.

llues @ fsolation valvej or

d. The measured combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 1 gpm times the total number of such valves,

" ~

{ ,

" Exemption to Appendix "J" of 10 CFR 50.

l LA SALLE - UNIT 1 3/4 6-2

CENTAI19 eft SYSTEMS - -

..- LIMITING CONDITION FOR OPERATION (Continued}

ACTION (Continued) .

a

restore: e

a. The overall integrated leakage rate (s) to less than or equal ts 0.75 i La, and

~

b. The combined leakage' rate for all penetrations and all valves listed

! in Table 3.6.3-1, except for main staas isolation valves and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type 8 and C tasts to less than or equal to 0.60 L , and

. Me 5 all c.

1;ws 4ho[The x =h .t. leakage isolationrate valve?to and less than or equal to 25-sef per h

d. The combined leakage rata for all ECCS and'RCIC containment isolation I ~ valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpa times the total number of -

l

such valves, ,

l prior to increasing reactor coolant system temperaturn above 200*F.

SURVEILLANCE REQUIREMENTS

,l '

4. 6.1. 2 The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in confomance with the criteria s

specified in Appendix J of 10 CFR,50 using the methods and provisions of ANSI (

y N45.4[-[1972: Nr I

a. Three Type A Overall Integrated Containment Leakage Rate tests shall l be conducted at 40 210 month intervals during shutdown at P,,

f 39.6 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year

plant inservice inspection.

i b. If any periodic Type A test fails to meeto.75 L,, the test schedule I for subsequent Type A tests shall be reviewed and approved by the

! Commission. If two consecutive Type A tests fail to meeto.75 L,, a l

,; Type A test shall be performed at least every 18 months until two l ; consecutive Type A tests meet 0.75 L,, at which time the above test t

! schedule may be resumed.

I

c. The accuracy of tach Type A test shall be verified by a supplemental

! test which: '

l

1. Confirms the accuracy of the test by verifying that the l difference between the supplemental data and the Type A test g data is within 0.25 L,.

! 2. Has duration sufficient to establish accurately the change in l 1eakage rate between the Type A test and the supplemental test.

' 3. Requires the cuanthy cf gas injected :. : -he contain :ent er-bled frem the contai.- int during the st ;;arental tast : te

._, equivalent to at feast 25.gszeent of the total measured leakage l l

at P,, 39.6 psig. %

! LA SALLE - UNIT 1 3/4 6-3

, . . - - . , . . . - , - . - ,,,,.,.n-,_.n.

s CONTAINMENT SYSTEMS '

3'../

[ PRIMARY CONTAlfWENT AIR LOCKS i -- -

LIMITING CONDITION FOR OPERATION i -

l 3.6.1.3 Each primary containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normat

transit entry and exit through the containment, then at least one .

e i

dr lock door shall be closed, and t b.

An overall air lock leakage rata of less than or equal to 0.05 L, at P,, 39.6 psig.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2",and 3. l ACTION:

~

! a. With one primary containment air lock door inoperable:

j 1. Maintain at least the OPERABLE air lock door closed and either .

. restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERA 8LE air lock door closed.

../ 2. Operation may chen continue until perfomance of the next required

, overall air !ock leakage test provided that the OPERABLE air lock j door is verified to be locked closed at least once per 31 days.

. 3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. The provisions of Specification 3.0.4 are not applicable.

i

b. With the primary containment air lock inoperable, except as a result of i

an inoperable air lock door, maintain at least one air lock door closed;

' restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I l

l 1

.] asee special Test Exception 3.10.1.

l %

e LA SALLE - UNIT 1 3/4 6-5

, . , . . - , . - , - - . .., - - - - - - - . .w--n. - - . -- - - - - - -- --- --- --- - - - - - - - - - - - - - -- - - - -

, i CONTAINMENT SYSTEM 5 l m '

PRIMARY CONTAI!S9ENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION l i l l

3.8.1.5 The structure 1 integrity of the primary containment shall be maintained !

r

at a level consistent with the acceptance criteria in Specification 4.6.L5.

APPLICA8ILITY: ! l OPERATIONAL CONDITIONS 1, 2;and 3.

ACTION:

With the structural

'. tagrity of the primary containment not confere- i

. ing to the above requ nts in that tested tandon lift-off force

'. s individual tendons of es below the predicted lower limit but is greater than the design fait, perform an engineering evaluation of -

the primary containment demonstrate its structural integrity within 15 days; if the measured ft-off force of a tandon(s) is less than the design limit, perform engineering evaluation of the primary containment to demonstrate s structural integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ;

' otherwise, be in at least HQ 5HUTDCWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and .

in COLD SHUTDOWN within the f 11owing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the structural integrity o the primary containment otherwise not conforming to the above requ nts, perform an engineering evaluation of the primary contai nt to demonstrate Its structural integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; othervi , be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in CD SHUTDOWN within the following

. . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i a. With more than one tendon with an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit or with l

i one tendon below 90% of the predicted lower limit, restore the tendon (s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with

' Specification 6.6C. or be in at itast HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I b. With any other abnormal degradation of the structural integrity at l a level below the acceptance criteria of Specification 4.6.1.5, i restore the containment vessel to the required level of integrity i within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perform an engineering evaluation of the contain-l ment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.6C. or be in at least HOT STANDBY i i within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following

' 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

. P. .

{

LA SALLE - UNIT 1 3/4 6-8 e

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

i

~ , , - - - ~ <,-,,,,.w -

.-.:.a.......

l

~

CONTAINMENT SYSTEMS 5

, SURVEILLANCE REOUIREMENTS

4. 6.1. 5 Primary Containment Tendons. The primary c neent st tural integ-

- rity shall shall be demonstrated at the end of M .onta; thred years after I the initial structural integrity test (ISIT) and every years thereafter i in accordance with Table 4.6.1.5-1. The structural integrity shall be demon .

strated by: -

a. Determining that a representative s' ample of at least 13 tendons, 8 horf-

, zontal and 5 vertical, selected in accordance with Table 4.6.1.5-1 have a e val de ' lift-off force 3: M :- t % ri-"- r f 9 '- values listed in Table 4.6.1.5-2 l

' ,, gu. at the first year inspection. For subsequent inspections, for tendons g and periodicities per Table 4.6.1.5-1, tt r' r 'i et y= "'t :" t j f:r;;. ch !'

d----

nd by th: :-nnt ".1 N : %4; fr " trfx: rd (MWUMJ' H ?:;; t t ' 'r; t:nde..; ;nd the minia'um lift-off forces shall be where t is the time interval in y, ears from initial tensioni{ng of the tendonde to the current testing date.and the values X1, X2, Y1 and Y2 are in accord-Vance with the values listed in Table 4.6.1.5-2 for the surveillance tendon.

M f, is As.1 This test shall include essentially a complete detensioning of tendons "N I selected in accordance with Table 4.6.1.5-1 in which the tendon is deten-stoned to determine if any wires or strands are broken or damaged. Tendons W yeses M found acceptable during this test shall be retensioned to their observed R h l feasteleg lift-off force 2 3%. During retensioning of these tendons, the change g 4 gy c in load and elongation shall be seasured simultaneously at a minimum of a

gg I three, approximately equally spaced, levels of force between the seating If elongation corresponding to a specific load differs

, $ " + T , orce and zero.more than 5% from that recorded during installation of tendons, an W is s p i 4 investigation should be made to ensure that such difference is not related to wire failures or slip of wires in anchorages. If the lift-off force Qar8 of any one tendon in the total sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon, shall be checked for their lift-off force. If both 8

these adjacent tendons are found acceptable, the surveillance program may proceed considering the- single deficiency as unique and acceptable. The twndon(s) shall be restored to the required level of integrity. More than

! one tendon zi x; ;- below the predicted bounds out of the original sample I t

II population or the lift-off force cf a selected tendon lying below 90% of the prescribed lower limit is evidence of abnormal degradation of the con-

,j tainment structure. I i

j i b. Performing tendon detensioning and material tests and inspections of a l previously stressed tendon wire or strand from one tendon of each group, hoop and V, and determining that over the entire length of the removed f wire or strand that:

1. The tendon wires or strands are free of corrosion, cracks and damage.

t l 2. A minimum tensile strength value of 240 ksi, the guaranteeri ultimate j strengtn of tne tendon material, for at least three wire or strand samoles, ona f em each end and one at mid-leng*5, cut from exh

! removed wire or strand. Failure of any one of the wire or strand samples to meet the minimum tensile strength test is evidence of abnormal degradation of the primary containment structure.

LA SALLE - UNIT 1 3/4 6-9 me e

=wwwv'wyw- wevp,,,. .

4---- v n -p 9 yy a

i

, TABLE 4.6.1.5-1 s TENDON SURVEILLANCE B

_ TENDON NupeERS i

Years After lj g aj,,,5,tr g al 1 3 5 10 15

$h[M5.,, H V H V H V H V H V Visual Inspection 48AC 15C 48AC 15C 48AC 15C 48AC 15C 48AC 15C of End Anchorages 56C8 15A 2C8 6C 38A 28A 48A 308 50CB Adjacent Concrete 19A 12C8 20A 14AC 17A 128A 23A 41C8 22A 538A 138 Surface and Pre- 708 47C 248A 32C 21C8 58 50AC 57AC l

stress Monitor- 20C8 29A 37C8 42C 238A 31C ing Tests IC8 47C8 38C8 12AC ' 57C8 49AC 568A 508 688 l 21AC Detensicning and 20C8 47C 2C8 42C 238A 31C 48A 22A 50C8 19A Material Tests l

TENDON Nul6ERS

! Years After l Initial Structural 20 25 30 35 40 l

Integrity Test Type of Inspection H V H V H V H V H V TYd" Inspection 48AC 15C 48AC 15C 48AC 15C 48AC 15C 48AC 15C l of End Anchorages 39CB 258 18A 38 48C8 78 49C8 25A 36C8 13A Adjacent Concrete 498A 11A 47AC 12A 51AC ISA 518A 188 488A 278 Surface and Pre- 71D 578A 588A 590 690 stress Mohitor-ing Tests Detensioning and 481A 12A 47AC 38 48C8 18A 518A 188 36CB 13A' /

Material Tests i 1

~.

l LA SALLE - UNIT 1 3/4 6-11 Amendment 1I l

i

-r -

g- - ---i.- , , - , , -eg-,_w,. .-,r_.--,..-.m , . , _,,,,,,--w,.,, ., - - - . - __ .,,_-,,__+-..--,,-.-.,---,,.m---,.,,,m,e w.m , - , -,w-- .

. . , ; . g;.

. g . ,. :w . . .. ._,._z~.... . . . .

1

. i CONTAD00ENf SYSTDt5

. N .

ORYWELL AND SUPPRESSION CHAMER PURGE SYSTEM LDu a mis CONDITION POR OPERATION

3. 6.1. 8 The drywell and suppression chamber purge system may be in operation with the drywell and/or suppression chamber purge supply and exhaust butterfly isolation valves open for inerting, de-inerting,and pressure control, provided I

- that 6mLeach na.buttarfly(ees

%4% Trem4*nart sys+en %sti be ess4,.k.kQ 4e less 4hw or avalve is blocked sr+o 90 hoors APPLICA8ILITY: OPERATIONAL ColCITIONS 1, 2 and 3. per '36 5" days. '

f ACTION:

With a drywell and/or suppression chamber purge supply and/or exhaust buttarfly .

isolation valve open for other than inerting, de-inerting,or pressure control, I or not blocked to isss than or equal to 50* open, close the buttarfly valve (s) within$omehourorbeinatleastHOTSHUTDOWNwithinthenext12hoursandin l COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

m SURVEILLANCE REQUIR98ENTS

. 4.6.1.8.1 When being opened, ths drywell and suppression chamber purge supply l and exhaust butterfly isolation valves shall be verified to be blocked so as to open to less than or equal to 50* open, unless o verified within the previous 31 days. ,

, 4.6.1.8.2 Each drywell and suppression chamber purge supply and exhaust butterfly isolation valve shall be desanstrated OPERA 8LE at least once per 92 days by verifying that the measured leakage rate is less than or equal to 0.05 L,.

4.6.1.8.3 The cumulative time that the drywell and suppression chamber purge system has been in operation purging through the Standby Gas Treatment System j shall be verified to be less than or equal to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days prior to

use in this mode of operation.

l l5 1

', LA SALLE - UNIT 1 3/4 6-15 l:

l - .. . . _ _ _

,a-, m ---- --.,y_, .p._. 3, , y- , ,,--,._e --- , -..-,w,.. ,y -,,-,- -.,w - ---e.- y - - - - -

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

l

. , CQNTAIMENT SYSTEMS

\

3/4.6.2 DEPRESSURIZATION SYSTEMS

' SUPPRESSION CHANOER LINITING CONDITION FOR OPERATION I -

-_. 3.5.2 L. The suppression chammer shall be OPERABLE with:

~a r The pool water:

l 1. Volume between 131,9003 ft and 128,800 3ft , equivalent to a t

level between 29510 Land 267,,2%, and a l

2. Maximia average temperature of 100*F" during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature l may be pem,jged to increase to:

a) 105'FF, during testing which adds heat to the suppression l

, chamber.

b) 110*FFwith THERMAL POWER 1ess than or equal to 1% of i 1 RATED ERMAL POWER.

. c) 120* with the main steam Ifne isolation valves closed I i following T scran. ~ ~

b. Drywell-to-suppression chamber bypass leakage less than or equal to 105 of the acceptable ANE design value of 0.03 f't2, APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

j ACTION:

i a. With the suppression chamber water level outside the above limits.

... restore the water level to within the Ifmits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. In OPERATIONAL CONDITION 1 or 2 with the suppression chamber average water temperature greater than or equal to 100*F, restore the average temperature to less than or equal to 100*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in l l at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , except, as pe mitted above:

1. With the suppression chamber average water temperature greater than 105'F during testing which adds heat to the suppression chamber, stop all testing which adds heat to the suppression chamber and restore the average tamperature to less than or

, equal to 100*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to COLD SHUTDOWN within the following

! !' 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I 2. With the suppression chamber average water temperature greater j than 110*F, place the reactor mode switch fn the Shutdown

position and operate at least one residual heat removal loop in the suppression pool cooling mode.

3. With the suppression chamber average water temperature greater

than 120*F, depressurtze the reactor pressure vessel to less l than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

+

5ee Specification 3.5.3 for ECCS requirements.

! i: M Jrf;d . 'O'- C ;;; ;' ; ,;; :;^.f; ; ;' 2; C7 ;..n;i ;; ;; ' . ; ;;; . .

- Weh e " t: :-- e: t:f:n T ? n re a :n u e : n ;' ;,--: -' n ;L

. "" " ~7dN. ! CM""" *" *2f ;^2i'iC:ti- . M'. ! Z ;. ;,.;;C i"'.1 ;

'2 : :. . ^ % c,;.

{

fsee Special Test Exception 3.10.8.

i

]' LA SALLE - UNIT 1 3/4 6-16

--g,yy-- --w m,w-- w-r 9 9,--e-s-;-----=-+-*PM--tr- -

-m'w g=gewe=e---dM-. P1N-.-'-^"'"-e"-"-"'*"'""*'*T*~

e - _

t 9

. CONTAINMENT SYSTEMS -

l -- s LIMITING CONDITION FOR OPERATION (Continued)

ECTION: (Continued) [

, c_ lith one suppression chamber water level instrumentation channel -

inoperable and/or with one suppression pool water temperature instruentation division @'inop'erable, restore the inoperable l

instrumentation to CPERA8LE status within 7 days or verify suppres-4 sion chamber water level and/or temperature to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local indication.

d. With both suppression chamber water level instrumentation channels inoperable and/or with both suppression pool water temperature instrumentationdivisions& inoperable,restoreatleastone l

' inoperable water level channel and one water targerature division to OPERA 8LE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDCWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . -

e. With the drywell-to-suppression chamber bypass leakage in excess of

! the limit, restore the bypass leakage to within the limit prior to l . increasing reactor coolant temperature above 200*F.

t

, SURVEILLANCE REQUIREMENTS -

.. 4.5.2.1 The suppression chamber shall be demonstrated OPERA 8LE:

. - a. By verifying the suppression chamber water volume to be within the

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

l b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in OPERATIONAL CONDITION 1 or 2 by verifying i i the suppression chamber average water tamperature to be less than or equal to 100*F, except: -

1

1. At least once per 5 minutes during tasting which adds heat to i the suppression chamber, by verifying the suppression chamber

average water tamperature less than or equal to 105'F.

i 2. At least onca per 60 minutes when suppression chamber average water temperature is greater than 100*F, by verifying suppression chamber average water temperature less than or equal to 110*F 8

and THERMAL POWER 1ess than or equal to 1% of RATED THERMAL

POWER.

I

3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 100*F by verifying suppression chamber average water tamparature less than or equal to 120*F.

= =; = =. ..=1 =;m c u ==i= .= m==i

e' ce---',:t'

_ g. . i u . ~ , . . .

e'.. ..!____.__.._m..

5 !' t - :: ste.a...

....__s e ::2.__..

._ -- ru--d

m. t me d'"* *-- te2.-.

.... ._ . , , ............o er ?

~

e I

'pa*'%e

_ , . ^ . ,_ , f_ h*

, within=90 de"s--e* *-- -t- 8 e-/

1 I

lj- LA SALLE - UNIT 1 3/4 6-17

( -

l *

. . . . ~

s I

i

. CoffTAINMee SYSTBes SURVEILLANCE REDUIR999175 (Continued) l .

I c. By verifying at least two suppressionf.hamber water level instru-I --

mantation channels and at least .faue4 Wen suppression pool water

-- f j

tamperature OPERABLE instrumentation by performance of a: cflannels, y in each of two divisions, i

l 1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

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

3. CHANNEL CALIBRATION at least once per 18 months, with the water level and tamperature alam setpoint for:

l 1. High water level i 26g#

l

2. Law water level > 29,"47

3. High temperature i 100' T

d. By conducting drywell-to-suppression chamber bypass leak tests and verifying that the A/4 calculated from the seasured leakage is

~

... within the specified limit when drywell-to-suppres'sien chamoer bypass leak tests are conducted:

1. At least once per 18 months at an initial differential pressure of 1.5 psi, and

2. At the first refueling outage and then on the schedule required for Type A Overall Integrated Containment Leakage Rate tests by Speci-fication 4.6.1.2.aj at an initial differential pressure of 5 psi, l l

, except that, if the first two 5 psi teak tests performed up to that I

i !

tima result in:

I 1. A calculated A/8 within the specified limit, and l-

2. The A/ 4 calculated from the leak tests at 1.5 psi is i 20% of the specified limit, then the leak tests at 5 psi may be discontinued.

' n-l

! -snim.- wnd--F4aal-::T int-t: h O te.;!aed A.-fa; :tectop-tu t ;;. %. .

i j "r; n ;f'7d ern;;; t: 'J.i; '.,,eint er.el l :,. . ; i M i L= C;--; .. l wn j . - utua *^ d:; : ef test :---1 ti:n. ,

.i 2

.;t .

LA SALLE - UNIT 1 3/4 6-18 e .

.em m. ass . -en .. .e,ee. ese, e * * * "*"" **""**** * * ** * **

ee_ ee - *w * * * ** ******4**- I

yoe.- -mar,*- n au- , ,,.m--g,r,.e-- , . - -

.,,y,-,, ,,.,.,,me, -

m.y,.p ,,..,,,e,y ,y_,,p,,, ,,p __ , - , ,.,, , - - , . , n. ---_-- - - , . - , , _ ,

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

-N - -

If any 1.5 pri or 5 psi leak test results in:

!7 - -

1. A calculated A//E greater than the specified limit, or

.j 6 ,

2. AcalculatedA//Efroma1.5psileaktest>20%ofthe specified limit, then the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 1.5 psi leak tests result in a calculated A//k greater than the specified limit, then:

~

1. A 1.5 psi leak test shall ba performed at least once per 9 months until two consecutive 1.5 psi leak tests result in the calculated A//E within the specified limits, and

2. A 5 psi leak test, performed with the second consecutive successful 1.5 psi leak test, results in a calculated A//E within the specified limit, after which the above schedule for only 1.5 psi leak tests may be resumed.

If two consecutive 5 psi leak. tests result in a calculated A//E greater thanthespecifiedlimit,thena5psileaktestshallbeperformedat -

l e least once pert 18' months until two consecutive 5 psi leak tests result in a calculated A/Jk within the specified limit, after which the above

' schedule for only 1.5 psi leak tests may be resumed.

s e

i l

LA SALLE - UNIT 1 3/4 6-19

... . -..- .N .; . - - - . - _ . . .. . a- -:

' ' (

CONTAINNENT SYSTEMS

(

- SUPPRESSION POOL SPRAY

.I LINITTNG CONDITION FOR CPERATION S

3. 6. 2. 2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERA 8.LE with two independent loops, each loop consisting of:

1

a. One OPERA 8LE RHR pump, and i

h. An CPERABLE flow path capable of recirculating water from the suppression chamber.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION: .

a. With one suppression pool spray loop inoperable, restore the inoperabie locp to OPERABLE status within 7 days or be in at least HOT SHUTDCWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t

b. With both suppression pool spray loops inoperable, restore at least

one loop to OPERA 8LE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN" within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

4 SURVEILLANCE REQUIREMENTS

4. 5. 2. 2 The suppression pool spray mode of the RHR systes shall be demonstrated OPERA 8LE:

a. At least once per 31 days by verifying that each valve manual, power-operated,or automati$ in the flow path that is not locked, sealed j or otherwise secured in position, is in its correct position.

I b. By verifying that each of the required RHR pumps develops a flow of I

at least 450 gpa on recirculation flow through the suppression pool spray sparger when tasted pursuant to Specification 4.0.5.

i, .

'idhenever notn AHR subsystems are inoperable, if unable to attain COLD SHUTDCWN

as required by this ACTION, maintain reactor coolant temperature as .ow as j practical by use of alternate heat ceroval metheds.

} '

e li LA SALLE - UNIT 1 3/4 6-20

. .2-.-.

1

l _. . w .

. . . . .-s .,.

. 8

. i.

. , CONTADMENT SYSTDt3 ,

SUPPRESSION POOL COOLING i 4

4' LIMITING CONDITION FOR OPERATION l

3.6.2.3 The suppression pool cooling mode of the residual heat renoval (RHR) .

, systes shall be OPERA 8LE with two independent loops, each loop consisting of:

l a. One OPERA 8LE RHR pump; and

b. An OPERA 8LE flow path capable of recirculating water from the suppression chamber through an RHRSW heat exchanger.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION: -

a. With one suppression pool cooling loop inoperable; restore the inoperable loop to 0PERA8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

b. With both suppression pool cooling loops inoperable, be in at 1 east HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN" within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS i

4.6.2.3 The suppression pool cooling mode of the RHR systes shall be demonstrated OPERA 8LE:

i a. At least once per 31 days by verifying that each valvegiaanual, powe& operated,or automatid in the flow path that is not locked, sealedj or otherwise secured in position, is in its correct position.

b. By verifying that each of the required RHR pumps develops a flow of at least 7200 gpa on recirculation flow through the RHR heat exchanger and the suppression pool when tested pursuant to Specif1 cation 4.O.5.

lI -

i 'Nhenever both RHR subsystems are inoperable, if unable to attain COLD SHUTD0hN

! as required by this ACTION, maintain reactor c:olant temperature as 1cw as

. practical by use of altarnate heat removal methods.

?, '

l l

LA SALLE - UNIT 1 3/4 6-21

.m 9 TA8tE 3.6.3-1 m

5 .

PRIMARY CONTAIMENT ISOLATION VALVES G

' MAXIMUM

c. VALVE FUNCTION AND NUMBER ISOLATION TIME '

3 VALVE GROUP (*)

' * (Seconds)

a. Automatic Isolation Valves w

1. Main Steam Isolation V4.1ves* 1 5* I 1821-F022A, 8, C, D(b) 1821-F028A, B, C, DID) -

2. Main Steam Line Drain Valves # 1 1821-F016 1821-F019 1 15 l 1821-F067A, B, C, D(b) -

< 15 R 3. Reactor Coolant S stem Sample 1 23 I Line ValvesICI 3 4

1833-F019 $5 1833-F020

4. Drywell Equipment Drain Valves 2 .

1RE024 _< 20 1RE025 1RE026 l 1RE029

5. Drywell Floor Drain Valves 2 1RF012 _< 20 1RF013

6. Reactor Water Clean'p Suction Valves 5 1G33-F001(d) -< 30

' 1G33-F004

7. RCIC Steam Line Valves 8 IE51-F008I ')

1E51~F063 < 20

! 1E51-F064 7 15 IE51-F076 _ 15 i $ 15

~~

' a l

TABLE 3.6.3-1 (Continued) 2 .

v. PRIMARY CONTAllMENT ISOLATION VALVES E

m MAXIMUM ISOLATION TIME VALVE FUNCTION AND NUMBER VALVE GROUP (a) (Seconds)

[ Automatic Isolation Valves (Continued)

8. Containment Vent and Purge Valves # 4 IVQO26 < 4MlO #

1VQO27 5130*IO*

IVQO29 < 439-/o8*

IVQO30 7 .HO-lod" IVQO31 7 43FIO""

IVQO32 7 40 f w IVQO34 7 400'10 #

1 IVQO35 I5 m IVQO36 7 480~/#

r'. IVQ040 7 430-10*

IVQ042 7,90 /*

IVQ043 7 40'/O**

IVQ047 7.3& 6' IVQ048 7 .3&-f IVQ050 IJ3-f IVQ051 5 33-#

IVQ068 <5

9. RCIC Turbine Exhaust Vacuum Breaker 9 -IIA #.S .

Line Valves 1E51-F080 1E51-F086

10. LPCS, HPCS, RCIC, RHR Injection i Testable Check Bypass Valves IU) 2 -fiA- AJ. A . -

IE21-F333 I

IE22-F354 1E12-F327A, B, C -

i 1E51-F354

! 1E51-F355

. . . . . . - . . . - . . ~ . --e~...... . . - -- . .

g

' le

.- .- i s : (,

! .l

. . I.

I TABLE 3.6.3-1 (Continued) !

E PRIMARY CONTAllelENT ISOLATION VALVE 5' .

= ,

N 90ANIIGM C ISOLATION TIIE VALVE GROUP I8)

VALVE FUNCTION Ale fluleER (Seconds) .

E '

q Automatic Isolation Valves (Continued)

11. Containment Monitoring Valves 2

<5 ICllul 7A,8 W ~

ICillslisA,BF r 1Ca t:ll'sA,88 '

]Cidj20A,8 * -

ICtl021D((h) 1CIWs72 h *

, 3 11 8 '

} l'h *'4 A i

g_4 .

, 1CI:h?M 4

m ICI:02 font 10.lo? /

I C. . ..u . .

101.' . 'J -

10 iWsti .

ICl a si ICilJt? .

1C hil3 ,

ICf tiis4 .

i

12. Drc.s.:ll Pneumatic Valves 2 IlithulA and 8 t

11:1011 l $ 40 '

l i tlis /-l * < 30 IIrin/'#

I 30 1 lilin s t* 7 30 7

1111is111

13. R ..intdown Cooling Mode Valves 6

! IEI. 1008 < 41 IE 1/ -I n09 Ili/ I023

< 41 11 I/-I '153 A and 8 I 90 2 29 Il I / -l #199A I 3

. 30 .

, i

.Me e e

i

. r' s., ) -

(' '

/ -

TA8tE 3.6.3-1 (Continued)

E ~

g P8tMA8V CONTAll80ENT ISOLATION VALVES I~

m flANIfE81 I *I ISOLATI0ft TIfE c VALVE fullCTIDII Ale NINSE8 VALVE GnouP (Seconds) 5 g .Au_f.o.m.itfc Isolation Valves (Contfnued)

14. Tigi Guide Tube Valve 7 -il4 AI.A.

Ball Valve l

?

1C>1 .1004 .

15. Reat. Lor Building Closed Cooling Water

System Valves 2 1 Wild 29 1 30 ,

, 1Wl'040 1:' IWal/9

** ]Witlhu l

}

16. Prom.asy Contalneent Chilled U.l c. Inlet Valves

2 Ivas1.s A and 8

l i

IVI'idel A and 8 1 90 -

j 17. Prim.iry Containment Chilled 1 40 3 trater Outlet Valves

2 IVIus3 A and 8 l 1

IVPili A and 8 i 40 l 18. Re. ii n.. liydraulic F16w Control ~ 1 90 f Lis.e V.alvesIUI 2 l 15 i 18Il i138 A and 8 ,

16 0 1139 A and 8 1

1811-1140 A and 8 l 18s1-l141 A and 8 1

18's1-1342 A and 8 j 1813-1's43 A and B .

i IBi: l 144 A and 8 '

4 i 10s 1645 A and 8

19. Fe. ii.e..t er Testable Check Valves 2 N4 M.O - k IB.'l-l .s32 A arul 8 1

)

l

~...z- . . .... - . . . . . . - . . . . . - . - - . . . . . . .

k t

I

.. s.v., ,

,[,j, *

l TABLE 3.5.3-1 (Continued)

- 'l ,

E

i s

g PRIMARY CONTA!W4ENT ISOLATION VALVES r-r; nnxImun .

. 150LAT10N TIIE

[ VALVE FUNCTION AND NtseER VALVE GROUP IU (Seconds) i z m .

-8 b. Manual Isolation Valves @.

1. 1FC086 m.#.4 4

, 2. IFCll3 .N # #.A.

3. 1FC114 Jet-g. A. -'

4. IFCll5 .

NArW.A.

5. 1HCO21(1) tegd. A.

6. 1HCO33(1) pegg.G. n

7. 15Au42(1) sekg.A.

u 8. F.A01G(1) .

444 af.h.

} .

T' .

~ ,

(S .

e 9

D j

4 I

, 93 l .

i j . I i l

. E

Y-

%../

g TABLE 3.6.3-1 (Continued) ' l PRIMARY CONTAlletENT ISOLATION VALVES g VALVE FINICT10N AND NISSER e !

., e d. Ollier Isolation Valves '

f.

-4 l

1. MSIV Leakage Control System I 1 1E32-F001A, E, J N(b) !

2. Rezictor Feedwater and 8WCU Systee Return #

4 1821-F010A, 8 e 1821-F065A, B .

3G33-F040 es '

4:s 3. $.lilo41HeatRemoval/LowPressureCoolantInfecticaSystes (J, 1E12 lH42A, 8, C 1Elz-lul6A, B N 1E12-1017A, 8 . .

IEl2-lug 4A, O U

IE12-lH27A, g 8 jg } .

/

g j

1E12-Iu24 )8 y) 1E12-1021 i 1E12-1to2 Il3 .

i 1E12-1064A, B I 1El/ lullA, 8g l i

1El/-la.88A, 8 C '

1E12-lo25 3,C 8

i 1E12-IH30 l 7 1E12-tu05III >

IE12-1013A,8g)g g 1El2-F074A, 8g y) ,

1Elz-1055A, Bg )))

1EI2-1036A, 8 y 1E12-1311A, BISI ~

P l 1E12-1041A,8(kh(k)

IElz-lu50A, 8 lj

. I j

1

- i.r g SIWh o g S

s ,

( .

l TABLE 3.6.3-1 (Continued) :

E

I g PRIllARY CONTAll0EllT ISOLATI006 VALVES l.

F m ! .

. VALVE Ft#4CII000 AND flulGER k-4 Other Isolation Valves (Continued) .

0 Low Pressure Core Spray System 4.

IE21-F005 IE21-f00l III r IE21-f012 ISI

~~

IE21 foll IE21-I018 III 1E21-1011

'~ 'c 1[21-IDO6 II -

3:'

5. liigh.. Pressure Core Spray System

.I. IE22-F004

'd IE22-F015 ISI ' '

IE22-l023 I$I IE22-t012(I '

. IE22-Inl4

i It~22-1005

6. Re.ir.ts.r Core Isolation Cooline Systee i IE ?.1 -i n13 .

/TE ',1 -l u86 . .

LIE? 1-l 080 - l 1E51-1069 , T lt's1-1028 '

If 51-1068 (,

IELI-1040 g IE51-f031 III IEbl-1619 '

itsI-I065 IkI l ';

IE51-1066 II (,

. I N

. . e . .. e . . tu . .. ... . . _ .> ,... . . ..

3 ....

.. g.

. . l

. . 1

.~

N - CONTAllegfT SYSTEMS 3/4.6.4 VACUtM RELIEF

.4 LIMITING CONDITION FOR OPERATION -

3.8.4 All suppression chamber - drywell. vacuum breakers shall be OPERA 8LE and closed.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2;and 3.

i ACTION:

a. With one suppression chamber - drywell vacuum breaker inoperable and/or open, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close the manual isolation valves on ,

both sides of the inoperable and/or open vacuum breaker. Restore

, the inoperable and/or open vacuum breaker to CPERA8LE and closed status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least NOT SHUTDOWN within the -

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one position indicator of any 0PERA8LE suppression chamber -

drywell vacuum breaker inoperable, restore the inoperable position indicator to OPERA 8LE status within 14 days or visually verify the .

., vacu m breaker to be closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise,

. declare the vacuum breaker inoperable.

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each suppression chamber - dell vacuum breaker shall be:

$ a. Verified closed at least once per 7 days.

i b. Demonstrated OPERA 8LE:

1. At least once per 31 days and within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any discharge l , of steam to the suppression chamber from the safety-relief l

valves, by cycling each vacuum breaker through at least one

completa cycle of full travel.

.i i' 2. At least once per 31 days by verifying both position indicators OPERA 8LE by performance of a CHANNEL FUNCTIONAL TEST.

j 3. At least once per 18 months by; l Verifying ti.e force required to open the vacuum breaker, from

.j a) the closed position, to be less than or equal to 0.5 psid, and 1

1 b) Verifying both cosition indicators OPE.iA3LE by performance of a e

- CHANNEL CALIBRATIO,N.

s LA SALLE - UNIT 1 3/4 6-35

.D

-~-,--~e----n--,--,-----we.,-,,._-,,--,-g---wr,-- ---e- - w - -,,.,.- + --~ ,

w w a - - - - - - - - -~w-- - , , -wwn-.-w, - ,

........._..a- ,.

f v

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

4. 6. 4. 2 The manual isolation valves on both sides of an inoperable and/or

open suppression chamber-drywell vacuum'breakar shall be verified to be closed at least once per 7 days.

l 4.6.4.3 Vacuus breaker header flanges which have been broken shall be leak tested after re-making by 5dth 0..; et e p. ..... . ef i 5 ;:f g.-- ,

~l~yp B M OA 39.6 ps! pc F ciCcdN 4.6.1.2,d ,

O l

e I

i .

e l 1 l l i .

I I.

i.

J LA SALLE - UNIT 1 3/4 6-36

~

e. . n. .. . .. .s.:. y .;. ;-

' ~r.u.u:...

. . m .. ..s. ...

B .

~,.' CONTA!!9eENT SYSTEMS j 3/4.5.5 SECONOANY CONTAINNENT ,

{ SEC0fCARY CONTAI! STENT INTEGRITY -

i LIMITING CONDITION FOR OPERATTON ,

l 3.6.5.1 SECONDARY CONTAINNENT INTEGRITY. shall be asintained.

APPLICA8ILITY: OPERATIONAL CWWITIONS 1, 2, 3 and *. l

/

l ACTION:

Without SECONDARY CONTAIl# TENT INTEGRITY:

a. In CPERATIONAL CONDITION 1, 2 or 3, restore SECONDARY CONTAINNENT l INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least NOT SHUTDOWN within the

, next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b. In Operational Condttion *, suspend handling of frndf ated fuel in

i the secondary containment, CORE ALTERATIONS and operations with a -

i potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

e

, SURVEILLANCE REQUIREMENTS i

( '

4.6.5.1 SECONDARY CONTAIl#EENT INTEGRITY shall be demonstrated by:

3

a. Verifying at lasst once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months i. hat the pressure within the

socordary containment is less than or equal to 0.25 incher of [

l vacuum water gauge.# .

l b. Verifying at least once per 31 days that:

i 1. At least one door in each access to the secondary containment -

is closed.

j .

2. All secondary containment penetrations not capable of being l

closed by OPERABLE secondary containment automatic isolation I

i dampers and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers secured in position.

I c. At least once per la months:

, 1. Verifying that one standby gas treatment subsystem will draw

! down the secondary containment to greater than or equal to j 0.25 incret of vacuum water gauge in less than or equal to 'l

{ 300 seconds, and

2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inches of vacuum l 1

water gauge in the secondary containment at a flow rata not j exceeding 4000 CFM z 10%.

I l r "When irractatec %al is being har.::le: 'n . e secondsay e nt.21-ce-t anc d'.r': -

CORE ALTERATION. s ' operations witn a potenstal for oraining tne reactor vessei.

[] .

f5ECONDARY CONTAIrm: O INTEGRITY is maintained when secondary containment vacuum is less than requirh for up to sner hour solely due to Reactor Building ventilation l j system failure. 1 I

i l , LA SALLE - UNIT 1 3/4 6-37

. . . . . .. . . . . . . .. -- * . - - - . - - - . . . - - - ~ . --

~

.'....~ . .

-: .r u . .

i .

i m .

.'- CONTAllW8ENT SYSTDes SECDICARY CONTAINIENT AUTONATIC ISOLATION CAMPER 3 i

, ; LDf! TIME CON 0! TION FOR OPERAT' ION 3.6.5.2 The secondary containment ventilation systas automatic isolation despers shown in Table 3.6.5.2-1 shall be OPERA 8LE with isolation times equal to er less than shown in Table 3.6.5.2-1.' i APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *.

ACTION:

With one or more of the secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.2-1 inoperable:

a. Maintain at least one isolation damper OPERA 8LE in each affected -

. penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

-- - T

( Restore the inoperable damper to OPERA 8LE status, or - '

2. Isolate each affected penetration by use of at least one deactivated automatic damper secured in the Tsolation position. -

or

':

3. Isolata each affected penetration by use of at least one closed .

.s annual valve or blind flange.

J b. Otherwise, in OPERATIONAL CON 0! TION 1, 2,or 3 be in at least HOT l SHUTDOWN within the.next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . '

c. Othenvise, in Operational Condition *, suspend handifng of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations

'. with a potential for draining the reactor vessel. The provisions of .

Speciffcat1on 3.0.3 are not applicable.

SURVEILLANCE REQUIRENENTS 1

4.6.5.2 Each secondary containment ventilation system automatic isolation j damper shown in Table 3.6.5.2-1 shall be demonstrated OPERA 8LE:

l: a. Prior to returning the damper to service after maintenance, repair or replacement work is psrformed on the damper,or its associated

} actuator, control or power circuit by cycifng the damper through at

! 1 east one completa cycle of full travel and verifying the specified i isolation time.

l i

l 4

b. During COLD SelUTDOWN or REFUELING at least once per 18 months by l

' l verifying that on a containment isolation test signal each isolation

damper actuates to its isolation position.

I c. Sy verifying the isolation time to be within !e limit when tested I ! pursuant to Spectfication 4.0.5.

1 (

"When irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

LA SALLE - UNIT 1 3/4 6-38

= ..

, 1 j

l 1 *

,.- CONTAD01ENT SYSTDes STANOBY GAS TREATNENT SYSTEM I'

i-LDEITING CONDITION FOR OPERATION l .'.

I

l. 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERA 8LE.#

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, j3 and 8 I

, ACTION:

1

a. With one standby gas treatment subsystem inoperable, restore the

]. inoperable subsystem to OPERA 8LE status within 7 days, or:

i i 1. In OPERA 8LE CONDITION 1, 32 cr 3, be in at least HOT SHUTDOWN l 1

' within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the -

following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. In Operational Condition *, suspend handling of irradiated

, fuel in the secondary containment, CORE ALTERATIONS and opera-

' tions with a potential for draining the reactor vessel. The .

., provisions of Specification 3.0.3 are not applicable.

b. With both standby gas treatment subsystems inoperable in Operational

' Condition *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for l: draining the reactor vessel. The provisions of Specification 3.0.3

! are not applicable.

l1 I

l I SURVEILLANCE REQUIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERA 8LE:

I

a. At least orce per 31 days by initiating, from the control room, flow l through the HEPA filters and charcoal adsorters and verifying that the l subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERA 8LE.

l "When irraciated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

I o, The e.or?.a1 or emergency power sou :a .?.ay be inoperable in Operational

,j .

Condition *. '

LA SALLE - UNIT 1 3/4 6-40

s g'

CONTAIWIENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or j chemical release in any ventilation zone [

communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a. C.5.c and C.5.d of Regulatcry Guide 1.52 Revision 2, March 1978, and the system flow rate is 4000 cfm 2 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2 March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

3. Verifying a subsystem flow rate of 4000 cfm + 105 during system operation when tested in accordance with ANST N510-1975.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2. March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than or equsi to 8 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm 2 105.

2. Verifying that the filter train starts and isolation dampers open on each of the following test signals:

a. Reacter Building exhaust plenum radiation - high,

b. Drywell pressure - high,

c. Reactor vessel water level - low low, level 2, and

d. Fuel pool vent exhaust radiation - high.

3. Verifying that the heaters dissipate 20 1 2.0 kw when tested in accordance with ANSI M510-1975.

i LA SALLE - UNIT 1 3/4 6-41

<- , - e . -n - - ,- ,..,,,.__.,.,w,..--,_ , , . - - . - - - . . , , , . . -,,,,,,,,,,,,-,,--e__-,-,, - - -... - . - - , , ,,- -- , . . _ . . - , - - - - - -

.. ,. l l

i PLANT SYSTEMS l

.- SURVEILLANCE REQUIREMENTS

c. At least once per 18 months by:

1. Performing a system functional test which includes simulated l

_ _, automatic actuation and verifying that each automatic valve in

'E=- - - - -

the flow path actuates to its correct position, but may exclude

- "-' ~

actual injection of coolant into the reactor vessel.

2. Verifying that the system is capable of providing a flow of greater than or equal to 600 gpa to the reactor vessel when steam is supplied to the turbine at a pressure of 150.2 15 psig using the test flow path.# l

3. Perfoming a CHANNEL CALIBRATION of the discharge line " keep filled" pressure alarm instrumentation and verifying the low pressure setpoint to be ?,62 psig.

d. By demonstrating MCC-121y and the 250-volt batter $nd charge # l OPERABLE:

1. At least once per 7 days by verifying that:

a) MCC-121y is energized, and has correct breaker alignment, indicated power availability from the charger and battery, and voltage on the panel with an overall voltage of greater than or equal to 250 volts.

b) The electrolyte level of each pilot cell is above the plates, c) The pilot cell specific gravity, corrected to 77'F, is greater than or equal to 1.200, and f

d) The overall battery voltage is greater than or equal to 250 volts.

2. At least once per 92 days by verifying that:

a) The voltage of each connected battery is greater than or equal to 250 volts under float charge and has not decreased more than 12 volts from the value observed during the original test, b) The specific gravity, corrected to 77'F, of each connected cell is greater than or equal to 1.195 and has not decreased more than 0.05 from the value observed during the previous test, and c) The electrolyte level of each connected cell is above the plates.

3. At least once per 18 months by veri *ying that:

a) The battery shows no visual indication of physical damage or abnomal deterioration, and b) Battery teminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

"The provisions of Specification 4.0.4 are not applicably provided the surveillance is perfomed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is

_ adequate to perform the tests. ,

! o hN [.'. b.bEM5955 5N b555i[_N5. 2dfb55 3IdNU" 53 '

s e Z 476.64_E-KI..~a.Z.Z Z Zi:n. G_.'i.'_.'Z ._'T7!_!.. C _'L.'..~.

_ _ _ _ . _ _ _ _ . . _ -- _ --- - . _ _ _ . . _ _ . A".E_

. . . . .I.. r_

! LA SALLE - UNIT 1 3/4 7-8 I Amendment 5 1

m_ . m ._ ,

i T PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.5.1.1 The fire suppression water systes shall be demonstrated OPERABLE: ,

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 l flow path through at least one complete cycle of full travel.

c. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1. Verifying that each automatic valve in the flow path actuates to its correct position, ,

2. Verifying that each fire suppression pump develops at least l

gpaatasystesheadofgfeet.

3. Cycling each valve in the flow path that is not tastable during

, plant operation through at least one complete cycle of full -

x travel, and

~

4. Verifying that each fire suppression pump starts sequentially to maintain the fire suppression water systas pressure greater thanorequalto)IISD8' psig. l

d. At least once per 3 years by perfoming a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook,

14th Edition, published by the National Fire Protection Association.

! 4.7.5.1.2 Each diesel driven fire suppression pump shall be demonstrated OPERABLE:

I.

j a. At least once per 31 days by:

,j l

1. Verifying the fuel day tank contains at least 130 gallons of fuel.

l 2. Starting: ,

} a) The fuel transfer pump and transferring fuel from the storage tank to the day tank.

I b) The diesel driven pump from ambient conditions and operating for at least 30 minutes on recirculation flow.

0 l

,5 LA SALLE - UNIT 1 3/4 7-12

I i -

. ., ..,.a.-

. . .. a ... u .. a .: p . . e c... ..,..._ . .

(- ,

l 1 l l,3 PLANT sYsTees l ,

OELUGE AN0/0R SPRINKLER SYSTDES ,

I l -

LIMITING CONDITION FOR OPERATION I 3.7.5.2 The deluge and sprinkler systans of Unit 1 and Unit 2 shown in I

Table 3.7.5.2-1 shall be OPERA 8LE.* *

, APPLICABILITY: Whenever equipment protected by the deluge / sprinkler systems l are required to be OPERA 8LE.

ACTION:

a. 1 tith one or more of the deluge and/or sprinkler systems shown in Table 3.7.5.2-1 inoperable, within hour estabitsh a continuous l l i fire watch with backup fire suppression equipment for those areas in I ! which redundant systems or components could be damaged; for other -

l ! areas, estabitsh an hourly fire watch patrol. Restore the system to OPERA 8LE status within 14 days or, in lieu of any-other report (j; required by Specificattori 6.6.5, prepare and submit a Special Report to the Coemission pursuant to Specification 6.6.C within the next 30 days outlining the action taken, the cause of the inoperabil.ity and .

the plans and schedule for restoring the system to CPERA8LE status.

3 .--

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

1I ' -

l SURVEILLANCE REQUIREMElfT5 4.7.5.2 Each of the above required deluge and sprinkler systems shown in Table 3.7.5.2-1 shall be demonstrated 0PERA8LE:

l I

a. At least once per' 31 days by verifying that each valveksanual, power-operated y or automatigin the flow path is in its correct position. i

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: .

l 1. By performing a system functional test which includes simulated l 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 l

b) Cycling each va've in :.:e fi w cath that is not testa:'e

., during plant caerstien :.veuin at im; or.e c:=lete :,cle j ,, of full travel.

8 -

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

i i k

, LA SALLE - UNIT 1 3/4 7-14

\

m -

PLANT SYSTEMS s

CD, SYSTEMS LIMITING CONDITION FOR OPERATION f

3.7.5.3 The following low pressure C0 systems of Unit 1 and Unit 2 shall be OPERA 8LE.*

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 roce.

APPLICA81LITY: Whenever equipment protected by the low pressure CO2 systems is required to be OPERA 8LE.

ACTION:

a. With one or more of the above required low pressure CO: systems I

inoperable, within hour establish a continuous fire watch with j 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, in lieu of any other report required by Specification 5.6.8, 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 I plans and schedule for restoring the system to OPERA 8LE status, The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

b.

{ SURVEILLANCE REQUIREMENTS

, 4.7.5.3 Each of the above required low pressure CO2 systems shall be

! demonstrated OPERA 8LE:

a. At least once per 7 days by verifying CO2 storage tank level to be

greater than 505 full and, pressure to be greater than 290 psig, and

b. Atleastonceper31daysbyverifyingthateachvalveg(manual, power-operated, or automatie f )in the flow path is in the correct position.

I' c. At least once per 18 months by verifying:

l j 1. The system valves and associated motor operated ventilation i dampers actuate, manually and automatically, upon receipt of

'I a simulated actuation signal, and ,

- 2. Flow from each nozzle during a " Puff Test." ,

i "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 Amenhent No.13 L: _ _ _ _ , . . ,

j l

FIRE 2 5E STAT 70NS

.4 Linuusis 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 OPERABLE.

APPLICA8ILITY: Whenever equipment in the areas protected by the fire hose l stations is required to be OPERA 8LE.

ACTION:

l 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 i to the unprotected area (s)/ zone (s) from an OPERA 8LE hose station within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable f,f re 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 GPERA8tE status within 14 days or, in lieu of any other report required my Specification 6.6.8, 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 systes to OPERA 8LE status.

, b. The provisions of specifications 3.0.3 and 3.0.4 are not appitcable.

I SURVEILLANCE REQUIREMENTS .

l -

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:

I I . 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.

1

b. At least once per 18 months by:

i .

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

,l.

2. Removing the hose for inspection and re b king, and l

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

c. At least once per 3 years by partially opening each hose station valve t

to verify valve OPERAB*L*TY and no f t:w 31ccks;2. ,

[ '

l

d. Within 5 years and beteeen 5 and 8 years aftar purcnase cata ano 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 s operating pressure, whichever is greater.

l LA SALLE - UNIT 1 3/4 7-18

.-.--.---,----,---r------- -.-

l

,l k

s

/3 PuMT SYSTEMS 4 -

3/4.7.6 FIRE RATED A55 BELIES

~

LIMITING CON 0! TION FOR OPERATION

sc/v/kg.

. 3.7.6 All fire rated assemblies 4 ea11s, f floor / ceilings, cable tray enclosures i and other fire barrierWseparating safety related fire areas or 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.

i APPLICA8ILITY: At all times.

. M ;_ .

i a. With one or mere of the above required fire rated assembifes and/or sealing devices inoperable, within one hour either estaolish a con-tinuous fire watch on at least one side of the affected assembly (s) and/or device (s) or verify the OPERA 8ILITY of fire detectors on at l : 1 east one side of the inoperable assembly (s) and/or sealing device (s) .

l s and establish an hourly fire watch patrol. Restore the inoperable fire rated assembly (s) and/or sealing desfee(s) to OPERA 8LE status

, D' within 7 days or, in lieu of any other report required by Specifica-tion 6.6.8, prepare and submit a Special Report to the Commission pursuant to Specification 6.6.C within the next 20 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 appitcable.

SURVEILLANCE REQUIREMENTS l

l ! 4.7.6.1 Each of the above required fire rated assembifes and sealing devices I

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

a. The exposed surfaces of each fire rated assembifes.

l *

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

c. At least 10 percent of each type, of sealed penetration. If apparent l changes in neearance or aar:r ' ' :!;-1:2:1en3 are fc.rd, a vir.a inspec-f an ci an additiona; 0 . :;rt ;f u: :c:a cf seglec h penetrat. ion s.. ail oe mace. Tais tr.4 pac. ion process snail consint.a

'd-until a 10 percent sample with no apparent changes in appearance or abnormal degradation is f.ound.

b .

I LA SALLE - UNIT 1 3/4 7-22

. g 4

3 -

l I

, PUWT aiaicF=

l

_2/4.7.7 AREA TEi+ERATURE MONITOPING <

,_ .l- ..

f L a iiNG CONDITION FOR OPERATION ,

i d Unit 2 shown in Table 3.7.7-1 3.7.7 The temperature of each area of Unitis Irequired anin Table to be 3.7.7-1.

shall be saintained Whenever the equipment in an affected area within the limits indicated APPLICASILIT_Y:

6PERABLE.

g: e lief t(s) shown in Table iff-3.7.7-1:

l-With one or more areas exceeding the temperaturhour For more than 30 days providing a

a. cation 6.6.8, prepare and submit a Special Repori e time the pursuant to Specification 6.6.C within the nextanalysis to record of the amount by whichd equipment. and the cumulat v l

D in the affected area exceeded the continued OPERA 8ILITY ofReport therequired itsaffecte limit above, and an J

b.

By more than 30*F, in addition to the Specialithin its temp inoperable.

l within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area to wlimit o i

SURVEILLANCE REQUIREMENT 5

s shown in Table 3.7.7-1 i

4.7.7 The temperature in each of the above required arealeast on j

shall be determined to be within ita limit at h

l j -

l

?.

9

'e , .

as*

. _ _ . .._ . .. . _ _ _ . . ._ . .. . 3/4.7-24 ,_ . . .

m..,

2L . E. o

_=_ . _

l S, ..

TA8LE 3.7.7-1 4

AREA TEMPERATURE MONITORING _

TEMPERATURE t.IMIT (*F)

ERVICED EQUIPMtNi AREA NOT OPERAiING OPERATING A. Unit 1 Area Temperature Monitorino

1. Control Room 104 < 104

2. Auxiliary Electric Equipment Room h104 < 104

3. Otesel Generator Roos 122 < 122

4. Switchgear Room M-104 < 104

5. N'T.S. LPC5, RHR & RCIC Rooms 150 < 150

6. Prisary Containment

' 50 *

a. Drywell 76-150 < 150

b. Beneath Reactor Pressure Vessel 185 < 185

8. Unit 2 Area Temeerature Monitoring Recuired For Unit 1

1. Auxiliary Electric Equipment Room 104 < 104

2. Ofesel Generator 2A Room -

-122 < 122 l . 3. Division 1 and 2 Switchgear Rooms -104 < 104 l

l l .

i .

i I

, 4 l

l k

t f

LA SALLE - UNIT 1 3/4 7-25 l .-

- 5

. t 3/4.7.9 SMUSSERS *- *

}nser t A N e

fAf

~

l LIMITING CONDITION FOR OPERATION y4:e me MeckM l

. 3.7.9 All'snubbers "-tM ' . Ne ? '."-1 ;.-4 :.7.; shall be OPERABLE. I I

i

\

APPLICA8ILITY: OPERATIONAL C0ff0ITIONS 1, 2 and 3 .ene f OPERATIONAL CONDITION" 4 l and 5 for snubbers located on stems required OPERABLE in those OPERATIONAL

'; CONDITIONS.

ACTION:

With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber (s) to 0PERA8LE tatus and perform an engineering evaluation per Specification 4.7.94$on the su(pported component or declare the supporte j system inoperable and follow the ap' priat'e ACTION statament for that system. ,

SURVEILLANCE REQUIRE 4ENTS 4.7.9 Each snubber shall be demonstratb OPERA 8LE by performance of the

~

following augmented inservice inspection; program and the requirements of Specification 4.0.5.

a. VisualInspe$ tion

! The first inservice visual inspection of snubbers shall be performed after 4 months but within 10 months of commencing POWER OPERATION and shall include all snubbers listed in Tanles 3.7.9-1 and 3.7.9-2.

If less than two snubbers are foun4 inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months 2 25% f the date of the first inspection. Otherwise, subsequent vi ual inspections shall be i performed in accordance with the foll ing schedule:

llo. Inoperable Snubbers ubsequent Visual per Inspection Period nspection Period *#

0 18 months 2 25%

~

! 1

' months 2 25%

2 months 2 25%

! 3, 4 12 days 2 25%

l 5,6,7 6 days 2 25%

8 or more. 31 ays 2 25%

i The snutbers ay be catageri:ed f .to twc g cu; : Those sc:essib'a and then irc::assible cLr ; . u: cr c:erttie.. E.!:. caccc ny :e inspected incepenoently in accoraance witn tne we scneoule.

"The inspection interval shall not be lengthened more than ne step at a time. *

, #The provisions of Specification 4.0.2 are not applicable.

LA SALLE - UNIT 1 3/4 7-27

__ n .

e ..: .

e g

m bseer MiecWac P%ir ,

. ' PLANT SYSTEMS SURVEILLANCE REQUIR 5 (Continued)

b. Visual Inspection keekance Criteria Visual inspections s 11 verify (1) that there are no visible indications or supporting structuof that attacaments to the foundation aredamage secure, anor impaire OPERABILITY,-(2)d (3) in those snubber sovement can b manually induced without disconnecting the snubber,

) .

that the snubber has f edom of sovement and is not frozen up. Snubbers l which appear inoperable s a result of these visual inspections may be detamined OPERABLE for e purpose of establishing the next visual inspection interval pro ding that (1) the cause of the rejection is clearly established,and r! died for that particular snubber and for other snubbers that say be gener ally susceptible, and is functionally tested in se as found condition an(d determined OPERA per Surveillance Requiremen 4.7.9.d and 4.7.9.e as a when a fluid part of a hydr lic snubber is found,to be.pplicable. .uncavatedHowever, the snubber shall be declared i erable and cannot be determined OPERABLE by ~7 functional testing for the p sose of tstablishing the next visual inspection interval. All sn ers connected to an inoperable common hydraulic fluid reservoir shal be counted as inoperable snubbers.

. c. Functional Tests -

~

_.' During the first refueling shutd and at least once per 18 months

~

thereafter during shutdown, a esentative sample of snubbers shall be functionally tested.

1 For hydraulic snubbers, a represen .tive sasole of at least 10% of the total of hydraulic snubbers listed n Table 3.7.9-1 shall be functionally l -

tested either in place or in a bene test. For each hydraulic snubber that does not meet the functional te t acceptance criteria of Surveil- .

lance Raquirement 4.7.9.d or 4.7.9.e, an additional 10% of the hydraulic snubbers shall be functionally tested.

! For mechanical snubbers, a representat ve sample of that number of mee anical i

2 snubbers listed in Table 3.7.9-2 which allowstheexpression35(1+ ),

3 where c* is the allowable number of nec nical snubbers not meeting th

acceptance criteria selected by the oper tor shall be functionally

tested either in-place or in a bench test foreachnumberofmechanical

snubbers above c which does not meet the nctional test acceptance t

2

, criteria of Specifications 4.7.9.d. or 4.7 9.e an additional sample selectedaccordingtotheexpression35(1 j) (e { J (a 2

{

- c) shah where a is the total number of sechanical be functionally snubbers tested,le found inoperab during the functic 1 testing of the i representative saeple.

! Functionaltestingshallcontinueaccordingtotheexpression b [ 35 (1 + j) (g)2] where b is th: number of techanical snubbers found ino

! ' _. aecherable in the previous anical snuccars ts-sa are fot.;.a i;ainale until r.occ a sas:ple 1 r..il aitional i::perible all ::ecnanical snubbers in Table 3.7.9-1 and 3.7.9-2 have been f ctionally tested. .

. "C=2 1

LA SALLE - UNIT 1 3/4 7-28 -

. ( ,

. . 1 PLANT SYSTEMS *- ##

SUINEILLANCE REWIR S (Continued) .

Functional Tests (conti ) ,

e The representati sample selected for functional testing shall include the vario configurations, operating environments and the

.' rance of size and apacity of hydraulic and mechanical snubbers.

At Teast 25". of snubbers in the representative samples shall include snubbers f the foll'owing three categories:

L The first ber away from each reactor vessel nozzle.

2. Each snubber within 5 feet of heavy equipment (valve, pump, turbine actor,etc.).

3. Each snubber w thin 10 feet of the discharge from a safety relief valve. ,

In eddition to the regular ample, snubbers which failed the previous functional test shall be re sted during the next test period. If-a spare snubber has been insta led in place of a failed snubber, then -

e_ both the failed snubber if i is repaired and installed in another position, and the spare,snubb shall be retested. Test results of these snubbers may not be incl ed for the re-sampling.

If any snubber selected for f ional testing either fails to lockup or fails to move, i.e., f zen in place, the cause will be evaluated and if caused by manufa turer or design deficiency all snubbers,ofthesamedesignsubje to the same defect shall be functionally tested. This testing quirement shall be independent

of the requirements stated above fo snubbers not setting the

, functional test acceptance criteria.

I For any snubber (s) found inoperable, engineering evaluation shall be performed on the components which a supported by the snubber (s).

The purpose of this engineering evaluat n shall be to determine if the components supported by the snubber ( ) were adversely affected by the inoperability of snubber (s) in or r to ensure that the supported component remains capable of me ing the designed service.

d. Hydraulic Snubbers Functional Test Accentan e Criteria The hydraulic snubber functional test shall rify that:

! - 1. Activation (restraining action) is achiev d within the specified range of velocity cr ac:eh iti:n in ::tb ansi:n rd c:.-- us :-

2. Snutbar bleec, or ra;easa . ta,

hare n; i.2:, is .i :.9.. :..2 specified range in compression or tension. or snubbers specifi-cally required to not displace under continuo s load, the l

ability of the snubber to withstand load with t displacement shall be verified.

LA SALLE - UNIT 1 3/4 7-29

e. w e , , , .

, - . , , , , , - - - - , . - - , , - - - , ----+.--n ,--,--w-----v-w ,. ~---------v

l

. (

l

^

1

-taJseer AVBCMED PME 1 PLANT SYSTEMS z .

SURVEILLANCE REDUIRO9ENTS (Continued)

! e. Mechanical Snubbers Functiona Test Acceptance Criteria The mechanical snubber function 1 test shall verify that:

L The force that initiates f sovement of the snubber rod in

', either tension or compressio is less than the specified maximum breakaway friction drag force Breakaway friction drag force shall not have increased more than 5 since the last surveillance test.

2. Activation (restrainingaction) s achieved within the specified range of velocity or acceleratio in both tension and compression.

3. Snubber release rate, where requi d is within the specified range in compression or tension. For sn bars specifically required not to displace under continuout load, e ability of the snubber to

, withstand load without displacement hall be_ verified.

f. Snubber Service Life Monitorina A record of the service life of each snubber the date at which the -

m designated service life commences and the ins llation and maintenance

' records on which the designated service life i based shall be saintained as required by Specification 6.5.8. .

' Concurrent with the first inservice visual inspe tion and at least

. once per 18 sonths thereafter, the installation a d maintenance I records for each snubber listed in Tables 3.7.9-1 nd 3.7.9-2

!

shall be reviewed to verify that the indicated serv'ce life has not l

been exceeded or will not be exceeded prior to the t scheduled snubber service life review. If the indicated servi life will be l} exceeded prior to the next scheduled snubber service ife review l .

thesnubberservicelifeshallbereevaluatedorthesubbershall be replaced or reconditioned so as to extend its servic life beyond i the date of the next scheduled service life review. Th1 reevaluation, I

replacement or reconditioning shall be indicated in the cercs.

i 4 .

l.

lI LA SALLE - UNIT 1 3/4 7-30 y , - - - - - - , , v,- - - - , , . - , - - -,,7 . , , -- _. # ,w,,_.-_ .-m ..._,,,_,m_ -._- . ,..,_,.v_.,,, , . - - . _ - . - - _ _ _ _ _ - , - _ _ _

i

) .

-- ) '.) '

.i I

l TABLE 3.7.9-1 - -

I E SAFETY RELATED hydraulic $NUBBER$*

C E HIGi RADIATION

, . SNUB 8ER SYSTEM SNUBBER INSTALLED ACCES$18tE OR ZONE ESPECIALLY SIFFICULT g NO. ON, LOCATION AND ELEVATION INACCESSIBLE DURING $110100W0**

10 RENDVE

-4 (A or I) (Yes or No) (Yes or No) .

" None r .

5 - ~

Y -

U .

, I

\s -

W hers may be h to safat re systems without prior License Amendeent to Table 3.7.9-1 ~

, provided that a revision to T 1 3.7.9-1 is included with the next License Amendment request. .

- Moditiration to thi unn due to cha'nges in high radiation areas may be made without prior Licenw Ames provided that a revision to Table 3.7.9-1 is included with the next License .

Amendsen st. ;

I n

a I

I

. j e

-- j l

I l

PLANT SYSTEMS

.m."

3/4.7.9 SNUSSERS

%.V LIMITING CONDITION FOR OPERATION 3.7.9 All hydraulic and mechanical snubbers shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERA 8LE in those OPERATIONAL CONDITIONS.

ACTION:

With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.9g. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that j system.

SURVEILLANCE REQUIREMENTS 4

l 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a. Inspection Types As used in this specification, type of snubber shall mean snubbers of the same design and manu'facturer, irrespective of capacity.

f*3 f '- b. Visual Inspections Snubbers are categorized as inal:cessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be

, inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed

~

after 4 months but within 10 months of commencing POWER OPERATION and shall include all hydraulic and mechanical snubbers. If all snubbers of each type on any system are found OPERABLE dudng the.4e4e1 inser-vice visual inspection, the second inservice visual inspection of that systes shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given system shall be performed in accordance with the following schedule:

No. Inoperable Snubbers of Each Type Subsequent Visual On Any System per Inspection Period Inspection Period" #

0 18 montns 2 25%

1 12 months 2 25%

2 6 months 1 25%

3,4 124 days t 25%

5,6,7 62 days t 25%

8 or more 31 days 2 25%

'The inspection interval for each type of snubber on a given system shall not

, be lengthened more than one step at a time unless a generic problem has been

/. identified and corrected; in that event the inspection interval may be lengthened one step the first time and two steps thereafter if no inoperable snubbers of that type are found on that system. l -

The provisions of Specification 4.0.2 are not applicable.

LA SALLE - UNITA [ 3/47% 91-

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

1 l

l t

j PLANT SYSTEMS I

t e.

SURVEILLANCE REQUIREMENTS (Continued)

L \ _, . '

c. Visual Inspection Acceptance Criteria t

Visual inspections shall verify that: there are no visible indica-tions of damage or impaired OPERA 82LITY and (2) attachments to the foundation or supporting structure are secure, and (3) fasteners '

' for attachment of the snubber to the component and to the snubber anchorage are secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that: *

(1) the cause of the rejection is clearly established and remedied for that particular snubber and for othe snubbers irrespective of type on that system that say be generically susceptible; and (2) the j l affected snubber is functionally tested in the as-found condition and determined OPERA 8LE per Specification 4.7.9f. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers. For those snubbers common to more than one system, the OPERA 8ILITY of such snubbers shall be considered in

assessing the surveillance schedule for each of the related systems. ;

l d. Transient Event Inspection An inspection shall be perfomed of! all hydraulic and mechanical snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a

review of operational data and a visual inspection of the systems within 6 months following such an event. In addition to satisfy-ing the visual inspection sacceptance criteria, freedoe-of-motion of sechanical snubbers shall be verified using at least one of the l following: (1) manually induced snubber movement; or (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical l

snubber through its full range of travel.

e. Functional Tests ,

During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected prior to the test period or the sample plan used in the prior test period shall be i implemented: ,

1) At least 10% of the total of each type of snubber shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.9f., an additional 10%

of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or A representative sample of each type of snubber shall be fune-

{ 2) tionally tested, in accordance with Figure 4.7-1. "C" is the

. I LASALLE-UNITgg f 1 3/4 7-)as )

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

m __-

}

~

I. i

. PLANT SYSTEMS

( ; l SURVEILLANCE REQUIREMENTS (Continued) '

e. Functional Tests (Continued) '

total number of snubbers of a type found not meeting the accep- .

tance requirements of Specification 4.7.9f. The cumulative

' number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (pre-vious day's total plus cyrrent day's increments) shall be plotted on Figure 4.7-1. If at any time the point plotted

. falls in the " Reject" region, all snubbers of that type may be functionally tested. If at any time the point plotted falls in ,

the " Accept" region, testing of snubbers of that type may be !

terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be .

t tested until the point falls in the " Accept" region or the

" Reject" region, or all the snubbers of that type have been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing i

to resume. anew at a later time provided all snubbers tested with the failed equipment during the day of equipment failure are ratested; or

3) An initial representative sample of 55 snubbers shall be func-l -

tionally tasted. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the factor, 1 + C/2, where "C" is the number of snuobers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be plotted I

using an " Accept" line which follows the equation N = 55(1 +

C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the

" Accept" line, testing of that type of snubber may be terminated.

If the point plotted falls above the " Accept" line, testing must continue until the point falls in the " Accept" region or all the snubbers of that type have been tested.

The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative of the various con-figurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same location as snubbers which failed the previous functional test shall be ratested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due tn failure of only one type of snubber, the func-l tional test results shall be reviewed at that time to determine if l additional samples should be limited to the type of snubber which

', has failed the functional testi.ng. .

~

l LASALLE-UNIT /l 3/4 7-

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

. - . , - - - - - - - - - -- -- - ~ ~ - - - - - - - - - - - - - - - - - - - - -^ -

\ : .. - -

a. .--. .. . :.. .--... .. .. - -

(

l ( ,. PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i f. Functional Test Acceptance Criteria l

l The snubber functional test shall verify that:

1) Activation (restraining action) is achieved within the specified '

range in both tension and compression; l

2) Snubber bleed, or release. rate where required, is present in both tension and compression, within the specified range;

3) Where required, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel; and

4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load

( without displacement.

Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods,

g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the l '. - functional test acceptance criteria to determine the cause of the

' failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.

For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the compment remains capabic of meeting the designed service.

If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.9e. for snubbers not meeting the functional test acceptance criteria.

h. Functional Testino of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the i

fuactional test results shall be tested to meet the functional test .

LASALLE-UNITyI 3/47-Jf

- , - , , . , - - - - - - - -. - -n,,, - - - - - ,_, , , , , , , , , . , . -

e an- - - , . , - - - - - , - - . - - - - - - - - - ,

4

~

n I.

( ,_ PLANT SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) i h.

Functional Testino of Repaired and Replaced Snubbers (Continued) criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent ser-vice, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.

i.

. Snubber Service Life Procram The service life of hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be deter-mined and established based on engineering information and snall be ureextended history. orCritical shortened based on monitored test results and fail-parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be docu-mented Specification and the documentation shall be retained in accordance with 6.58.

k o

l LA SALLE - UNIT gg 2/4 7,X 31

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

- - - . . - - - - . . . .- .,._x.. __ _ _ _

e i \

l'O- .

9 8

7 REJECT 6 '

' p.+1

' g.h i j (' .

G

l

'. 4 f CONTINUE 3 2 TESTING , [

2 / #eM<

1 ,/ ACCEPT

)

i 0 10 20 30 40 50 60 70 80 90 100 N

FIGURE 4.7-1

/ , SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST t l

LASALLE-UNITgj 3/4 7 y 39.

t

~

1

, s Table 3.7.9-2 Safety Related Mechanical Snubbers

\

Snubber System snubber installed on Snubber System snubber installed No. ocation and elev. No. on location and elev.

{

DG15-100$$ DG Reactor 704 FWO2-H735 W DG18-00nS Containment 779 ' I DG Diesel Generator 746 FWO2-11745 N Containment 779 l DG18-00145 DG Diesel Generator 735 NO2-H755 FW Containment 792 DG34-10025 DG Diesel Generator 734 .N02-n765 OG34-10035 FW Auxiliary 714 DG fesel Generator 734 FW u-10035 FW Auxiliary 753 DG34-10045 DG esel Generator 745 FWu-10045 FW Auxiliary 753 DG34-1005S DG Di el Generator 745 FW11-10055 FW Auxiliary 754 DG34-18005 DG Die 1 Generator 740 FWu-10065 FW Auxiliary 754 FWO2-10375 FW Conta nment 792 HG01-10035 HG Reactor 799 i FWO2-10435 FW Contai nt 792 NG01-10055 HG Reactor 797

) FWO2-10445 N Contai nt 787 NG01-1006S HG Reactor 797 FWO2-10455 FW Containes t 779 HG01-10095 NG Reactor 797 FWO2-1046S FW Containmen 779 NG01-10105 HG Reactor 797 N02-10475 FW Containment 779 NG01-10465 HG Containment 803 FWO2-10505 FW Containment 792 NG02-2016S HG Reactor 798 FWO2-10555 FW Containment 755 NG02-20225 Reactor HG 798 N02-10605 FW Auxiliary 744 HG02-20275 HG Reactor 798 l FWO2-10715 FW Containment 778 NG02-20315 Reactor NG 798 l FWO2-u335 FW Auxiliary 743 NG05-10165 HG Reactor 756 N02-u435 FW Containment NG05-10255 NG Reactor 772

Wo2-11455 FW Containment 7 HG05-10295 NG Reactor 782 W 2-n465 FW Containment 7 NG05-10415 HG Reactor 734 rWO2-u485 FW Containment 784 NG06-10465 HG Reactor 757 FWC2-11495 FW Containment 778 6-10915 NG Reactor 726 FWO2-11505 FW Containment 778 -10985 HG Reactor 737 FWO2-n525 FW Containment '

785 21055 NG Reactor 772 FWO2-11535 FW Containment 785 21095 HG Reactor 769 FWO2- u545 FW Containment 780 HP01 .0075 HP Reactor 675 FWO2-11555 FW Containment 780 HP01-: 175 HP Reactor 675 FWO2-11565 FW Containment 783 HP02- HP Reactor 685 FWO2-11575 FW Containment 777 HP02-10 HP Reactor 764 NO2-11595 FW Containment 758 HP02-1019 HP Reactor 735 FWO2-H695 FW Containment 780 HP02-1020$ HP Reactor 690 N02-u705 FW Containment 754 HP02-10215 HP Reactor 685 FWO2-u715 FW Containment 791 HP02-10255 HP Containment 774 FWO2-u725 FW Containment 792 HP02-10265 HP Containment 789 i

i HP02-10275 HP Containment 774 LC01-10245 LC Auxiliary 690 NP02-15005 HP Containment 789 LC01-10265 Auxiliary 690 HP02-1502S HP Reactor 753 LC01-10315 L Auxiliary 690 HP02-15035 HP Reactor 753 LC01-10475 LC Auxiliary 682 HP02-15075 HP Containment 774 LC01-10505 LC Auxiliary 681 HP02-15085 HP Containment 774 LC01-10515 LC Auxiliary 681 HP02-1510$ HP Containment 774 LC01-10525 LC Auxiliary 681 HP02-15135 HP Reactor 690 LC01-10535 LC 111ary 682 l HP02-15195 HP Reactor 768 LC01-10545 LC A iliary 682 l HP02-15205 HP Reactor 769 LC01-10555 LC Au 'liary 681 l *02-15235 HP Turbine 708 LC01-10565 LC Auxi iary 682 LA Salle - UNIT 1 3/4 7-32 l

l l .

Table 3.7.9-2 (Continued) i, Snubber System snubber installed on Snubber System snubber installed No. location and elev. No. on location and elev.

NP02-15245 Containment 690 LC01-10575 LC Auxiliary 681 MP02-18005 Reactor 686 LC01-10585 LC Auxiliary 681 HP08-10095 Reactor 704 LC01-10595 LC Auxiliary 706 HP08-10245 HP Reactor 699 LC01-10655 LC Auxiliary 722

- NP08-10265_ .____HP Reactor 700 LC01-1066S LC Auxiliary 706 NP08-10285 HP Reactor . 699 "LC01-10685 LC Auxiliary 722 MP08-10295 HP tainment 703 LC01-10695 LC Auxiliary 722 HP09-10135 HP actor 688 LC01-10705 LC Aux 11 bry 721 or

- HP09-10365 --HP Re 693 LC01-10895 LC Auxiliary 732 HP14-10035 HP Rea r. 681 LC01-10965 LC Auxiliary 733 HP22-10035 HP Reac r' 768 LC01-11065 LC Auxiliary 731 HP75-10035 HP Diesel enerator 735 LC01-11205 LC Auxiliary 732 HP75-10055 HP Diesel nerator 746 LC01-11215 LC Auxiliary 732 LC01-10025 LC Auxilia 707 LC01-11235 LC Auxiliary 732 LC01-10035 LC Auxiliary 707 LC01-11265 LC Auxiliary 732 LC01-10045 LC Auxiliary 706 LC01-18075 LC Reactor 680 LC01-10055 LC Auxiliary 706 LC01-18085 LC Reactor 680 LC01-10075 LC Auxiliary 707 LC01-18095 LC Reactor 680 LC01-10085 LC Auxiliary 707 LC01-18105 LC Reactor 680 l LC01-1013S LC Auxiliary 706 LC03-18175 LC Reactor 680 LC01-10145 LC Auxiliary 06 LC03-18205 LC Reactor 680 LC01-10155 LC Auxiliary LC03-18235 LC Reactor 680 LC01-10195 LC Auxiliary 7 LC03-1826S LC Reactor 680

.C01-10215 LC Auxiliary 681 LC09-10015 LC Reacter 681 l

LC01-10225 LC Aux 11,iary 706 LC09-10135 LC Auxiliary 679 LC09-10185 LC Auxiliary 681 00-10155 MS Containment 780 LC09-10315 LC Auxiliary 702 -10165 MS Containment 780 LC09-10365 LC Auxiliary 702 MS -10175 MS Containment 775 LC09-10495 LC Auxiliary 681 M500 0195 MS Containment 763

LC09-10505 LC Auxiliary 681 M500- 0205 MS Containment 743 LC09-10525 LC Auxiliary 681 M500-10215 MS Containment 743 LC09-10535 LC Auxiliary 737 M500-10225 MS Containment 743 LP01-10145 LP Reactor 681 M500-1023S MS Containment 769 LP02-10135 LP Containment 778 M500-10245 MS Containment 786 LP02-10155 LP Containment 774 M500-10255 MS Containment 787 LP02-10165 LP Containment 774 MS00-10265 MS Containment 780 LP32-10175 LP Containment 774 MS00-10295 MS Containment 743 LP02-10185 LP Containment 774 M500-1030s S Containment 772 LP02-10195 LP Containment 774 MS00-10315 MS Containment 785 LP02-10205 LP Reactor 777 M500-10325 Mt Containment 790 l LP02-10255 LP Reactor 703 M500-10335 MSk Containment 790 l LP02-10545 LP Reactor 777 M500-10345 MS \ Containment 780 LP02-10555 LP Reactor 777 M500-10385 MS

\ Containment 743 LP02-10575 LP Reactor 777 M500-10395 M5 fontainment 743 i LP02-10595 LP Reactor 777 M500-10405 MS C ntainment 772 LP02-10625 LP Reactor 689 M500-10415 MS ainment 786 LP02-10675 LP Containment 788 M500-1042S MS Con inment 790 l LP19-10115 LP Reactor 763 MS00-10435 MS Cont nment 790

.P20-10255 LP Reactor 733 M500-10445 MS Contal ment 780 LA SALLE - UNIT 1 3/4 7-33 Amendment 1 l

s

{

Table 3.7.9-2 (Continued) .' '

\ !

l nubber System snubber installed on Snubber System snubber installed

2. l location and elev. No. on location and elev. l U20-10285 W20-10295 b Reactor 728 MS00-10455 MS Containment 785 l L Reactor 727 MS00-10495 MS Containment 743 1 020-10305 & Reactor 726 M500-10505 MS Containment 737 W20-10355 & Reactor 767 MS00-1051S MS Containment 769 M500-10065 MS Containment 775 M500-10525 MS Containment 775 MS00-10075 MS Containment 780 M500-10535 MS Containment 787 M500-10095 MS ntainment 748 M500-10545 MS Containment 786 l MS00-10105 MS tainment 774 M500-10555 MS Containment 780 l MS00-10115 MS Con inment 780 M501-1273S MS Auxiliary 691 M500-10125 MS Con neent 780 M504-n355 MS Containment 760 M500-10135 MS Conta nt 780 MSO4-n375 MS Containment 771 M500-10145 MS Contal nt 774 M504-n385 MS Containment 781 M504-11395 MS Contai t 781 M504- n925 MS Containment 760 M504-n415 MS Containee 758 M504-n935 MS Containment 751 M504-11425 MS Containment 758 MSO4-11945 MS Containment 766 M504-n435 MS Containment 775 M504-n955 MS Containment 779 MSO4-11455 MS Containment 779 MSO4- n965 MS Containment 779 MSO4- n475 MS Containment 779 M504-11975 MS Containment 778 M504- n505 MS Containment 748 MSO4-12015 MS Containment 753 M504-H515 MS Containment 770 MSO4-12025 MS Containment 755 M504- u52S MS Containment 0 M504-1203S MS Containment 760 M504- u535 MS Containment MSO4-12045 MS Containment 781 MSO4-11545 MS Containment 77 M504-12055 MS Containment 783

'504-n55S MS Containment 781 MSO4-12075 MS Containment 783

.504-n575 MS Containment 760 04-12105 MS Containment 783 MSO4-11585 MS Containment 773 04-12nS MS Containment 760 MSO4-11595 MS Containment 773 M. -12125 MS Containment M504-11615 Containment 7574 MS 781 MS -12135 MS Containment 758 MSO4-11635 MS Containment 748 M504 12155 MS Containment 748 .

M504- u655 MS Containment 762 MS04-1465 MS Containment 756 MS04- n665 MS Containment 763 MS0a-12175 MS Containment 748 M504-n695 MS Containment 749 MSO4-1218S MS Containment 764 M504- n705 MS Containment 751 M504-12199 MS Containment 756 M504- n715 MS Containment 766 MSO4-12205\ MS Containment 757 M504- n725 MS Containment 766 MSO4-12215 \ MS Containment 764 MSO4- n735 MS Containment 774 MSO4-12225 MS Containment 772 M504-1174S MS Containment 772 M504-12235 \MS Containment 751 M504-11755 MS Containment 774 MSO4-12255 MS Containment 750 MSO4- n775 MS Containment 758 M504-12295 MS Containment 779 M504-n785 MS Containment 761 M504-12305 MS- Containment 779 MSO4- n795 MS Containment 761 M504-1232S MS *. Containment 746 M504-n805 MS Containment 778 M504-12335 MS \ Containment 746 M504-11835 MS Containment 749 MSO4-12345 MS 'gContainment 770 MSO4-1284S MS Containment 772 M504-12355 MS Containment 770 M504-n855 MS Containment 773 MSO4-12375 MS Chntainment 780 M504-11865 MS Containment 772 MSO4-12385 MS Containment 779 M504- n895 MS Containment 781 M504-12395 MS Containment 747 M504-n905 MS Containment 781 M504-12405 MS Contalnment 747 MSO4-12415 MS Containment 747 MSO4-12955 MS Contaibent 744 o i LA Salle - UNIT 1 3/4 7-34 Amendmenti I 4

, - - . . , - - - = - - -

v r w-e- *y *-- w w e r e-r ww=--*t---v-ee---- e-em* - - - - - --we- - - - + - - - - - - - - - - - - - - - - - - - - - - -

l 1

Table 3.7.9-2 (Continued) 4 ubber System snubber installed on Snubber System snubber installed location and elev. No. on location and elev.

. MSO4 455 MS Containment 780 M504-12985 MS Containment 761 MSO4- 465 MS Containment 779 M504-12995 MS Containment 770 M504-12 5 MS Containment 752 M504-13005 MS Containment 772 M504-124 MS Containment 752 M504-13015 MS Containment 772 M504-1250S MS Containment 771 MSO4-13025 MS Containment 781

MSO4-12515 MS Containment 771 M504-1304S MS Containment 781 M504-12525 MS Containment 771 MSO4-13055 MS Containment 774 M504-12535 MS Containment 778 M504-13075 MS Containment 771 MSO4-12565 Containment 766 MSO4-13085 MS Containment 773 ;

M504-12585 Containment 761 M504-13205 MS Containment 746 MSO4-12595 MS Containment 780 M504-13215 MS Containment 768 MSO4-12615 MS Containment 774 M504-13225 MS Containment 766 MSO4-1262S MS Containment 748 M504-13235 MS Containment 766 MSO4-12655 MS ntainment 745 M504-13245 MS Containment 749 MS04-12665 MS tainment 764 MSO4-13255 MS Containment 766 MSO4-12675 MS Co inment 767 M504-13285 MS Containment 751 M504-12695 MS Con inment 780 M504-13295 MS Containment 772 MSO4-12705 MS Conta nt 781 MSO4-1330S MS Containment 772 M504-12715 MS Contai nt 747 MSO4-13315 MS Containment 774 r MSO4-12725 MS Contai t 747 MS04-1332S MS Containment 745 1

MSO4-12745 MS Containee 769 MSO4-13335 MS Containment 748 MS04-12755 MS Containment 769 MSO4-13345 MS Containment 774 MSO4-12765 MS Containment 772 MSO4-13375 MS' Containment 749 M504-12775 MS Containment 773 MSO4-13415 MS Containment 748 MSO4-12785 MS Containment 773 M504-1344S MS Containment 782 M504-12795 MS Containment 780 M504-13455 MS Containment 743 M504-12815 MS Containment 3 M504-13465 MS Containment 743 MSO4-1282S MS Containment M504-13475 MS Containment 757

~

MS04-12835 MS Containment 75 MSO4-13495 MS Containment '742 Containment MSO4-12885 MS 770 MSO4-13505 MS Containment 759 MSO4-1289S MS Containment 773 04-13515 MS Containment 758 M504-12905 MS Containment 772 04-1353S MS Containment 749 M504-12915 MS Containment 780 MS -13545 MS Containment 749 M504-12935 MS Containment 781 M50 13555 MS Containment 759 M504-12945 MS Containment 744 MSO4- 565 MS Containment 766 MSO4-13585 MS Containment 775 MSO4- S MS Containment 760 MSO4-13595 MS Containment 780 MSO4-154 MS Containment 760 MSO4-13625 MS Containment 751 MSO4-1550 MS Containment 760 MSO4-13635 MS Containment 759 MSO4-15515 MS Containment 781 M504-13645 MS Containment 759 M504-1553S MS Containment 748 M504-13665 MS Containment 757 MSO4-16835 MS Containment 769 MSO4-13675 MS Containment 775 M504-17055 MS Containment 751 M504-13685 MS Containment 743 M505-10025 Containment 810 MSO4-13695 MS Containment .779 M505-10045 M Containment 804 M504-13715 MS Containment 783 M505-10055 MS Containment 804 MSO4-13725 MS Containment 749 M505-10085 MS Containment 304 M504-13735 MS Containment 766 M505-10105 MS Containment 804 9504-13765 MS Containment 745 M506-10065 MS ontainment 737

~

A504-13785 MS Containment 761 M506-10165

MS ntainment 776 '

MSO4-13795 MS Containment 760 MS06-10175 MS Co tainment 779 LA SALLE - UNIT 1 3/4 7-35 Amen ent 1

- - _ _ - - --__________._,-____.__-,,.-,----_,.-_--,,,,,-,,.-.r. - . , , , , . . . , . .--,,_y, _ ~ , , _ _ . . , . , , .

. . . s Table 3.7.9-2 (Continued)

. i.

s \

Snubber System snubber installed on Snubber System snubber installed l l No. location and elev. No. on location and elev. l MSO4-13845 MS Containment 749 MS10-10025 MS Centainment 761 M504-13905 MS Containment 783 M510-10075 MS Containment 761 M504-13915 MS Containment 783 M510-10135 MS Containment 759 MSO4-13925 Containment 783 M514-10305 MS Containment 738

~

MS04-13935 Containment 783 MS14-10315 MS Containment 738 M504-13945 MS Containment 751 .MS14-10345 MS Containment 738 M504-13955 MS Containment 751 M514-10375 MS Containment 739 MSO4-13965 MS Containment 747 M514-10385 MS Containment 738 MSO4-13975 MS ontainment 744 MS14-10395 MS Containment 739 MSO4-13985 MS ntainment 758 M514-10445 MS Containment 739 MSO4-13995 MS Co inment 760 M514-10475 MS Containment 740 M504-15025 MS Con inment 760 M514-10485 MS Containment 741 M504-1503S MS Conta nt 763 MS14-1050S MS Containment 739 M504-15045 MS Contai nt 771 MS14-10515 MS Containment 740 MSO4-1505S MS Contai nt 770 MS14-10525 MS Containment 741 M504-15065 MS Containee 758 MS14-10545 MS Containment 741 '

i MS04-15085 MS Containeen 748 MS14-10555 MS Containment 741 M504-15095 MS Containment 756 M514-10565 MS Containment 741 MSO4-15105 MS Containment 749 M514-10585 MS Containment 741 MSO4-15115 MS Containment 760 MS14-10595 MS Containment 741 MSO4-15135 MS Containment 761 M514-10635 MS Containment 739 MS14-10665 MS Containment 741 M15-10025 N8 Containment 830 215-10055 NB Containment 828

. AS25-10235 MS Auxiliary NB15-10085 N8 Containment 827 MS25-10615 MS Auxiliary 68 N816-10025 N8 Containment 810 MS88-10055 MS Containment 787 NB16-10055 N8 Containment 810

'MS88-10065 MS Containment 787 16-10065 NS Containment 809 MS88-10115 MS Reactor 740 3-10035 N8 Containment 808 -

MS88-1012S MS Reactor 740 N8 -10025 NB Containment 808 MS88-10135 MS Reactor 740 PC01 0145 PC Reactor 741 MS88-1015S MS Reactor 740 PC01- 155 PC Reactor 741 MS88-10205 MS Reac+.or 740 PC01-1 65 PC Reactor 741 MSC6-10035 MS Containment 761 PC01-101 PC Reactor 741 MSC6-10045 MS Reactor 761 PC01-1019 PC Reactor 747 MSC6-10055 MS Containment 760 PC01-10205 PC Reactor 747 MSC6-10065 MS Reactor 761 PC01-18005 PC Reactor 751 MSC6-10095 MS Containment 761 RG10-00145 RG Auxiliary 795 MSC6-10135 MS Containment 760 RG21-00125 RG Auxiliary 797 MSC6-10155 MS Containment 760 RH01-10055 H Reactor 675 MSC6-10165 MS Containment 761 RH01-10065 R Reactor 675 MSC6-10185 MS Containment 760 RH01-10085 RH Reactor 681 MSC6-10215 MS Containment 760 RH01-10135 RH Reactor 681 MSC6-1024S MS Containment 761 RH01-10175 RH Reactor 681 MSD1-10015 MS Containment 787 RH01-10185 RH Reactor 581 MSF9-10025 MS Containment '785 RH01-10255 RH actor 678 MSF9-10045 MS Containment 784 RH02-1010S RH R ctor 705 MSF9-1006S MS Containment 787 RH02-10125 RH Rea tor 636 LA SALLE - UNIT 1 3/4 7-36 Amendment 1

.i e.

.,.,-.n _..__.__y . - - , - . -,,.,-..-,,_,_m.,-.- - - - _ . , ~ _ . - - . . . _ . - . . _ . - . _ _ . - . . _ _ _ - _ , - - - - - - .

Table 3.7.9-2 (Continued) g Snubber System snubber installed on No.

Snubber System snubber installed location and elev. No. on location and elev.

HSF9-10075 Containment 788 RH02-10175 M811-10035 M RH Reactor 732 Containment 808 RH02-10185 RH Reactor 686 G13-1001S M8 Containment 832 RH02-10195 M813-10025 RH Reactor 686 !

2 Containment 832 RH02-10245 RH Reactor 735 213-10045 W Containment 829 RH02-10255 NB13-1006S RH Reactor 695 2 tainment 828 ~ RH02-1026S RH Reactor 703

, N813-10255 M inment 813 RH02-1027S RH Reactor 695 l i-N813-10275 M8 Co inment 811 RH02-10435 RH Reactor 735 NB13-1028S M8 Con ineent 814 RH02-10475 RH Reactor 725 M813-1031S M Conta nt 811 RH02-10485 RH Reactor 722 RH02-10515 RH Reacto 732 RH03-15175 RH Containment 741 RH02-10525 RH Reactor 712 RH03-15245 Reactor RH 733 RH02-10565 RH Reactor 710 RH03-15255 RH Reactor 736 RH02-10575 RH Reactor 696 RH03-15265 Reactor RH 736 RH02-10585 RH Reactor 686 RH03-1527S RH Reactor 700 RH02-10605 RH Reactor 710 RH03-15285 RH Reactor 730 RH02-10625 RH Reactor' 715 RH03-1530S RH Reactor 736 RH02-10635 RH Reactor 727 RH03-15325 RH Reactor 736 RH02-10645 RH Reactor 696 RH03-15335 RH Reactor 715 RH02-10655 RH Reactor 5 RH03-15345 RH Reactor 736 RH02-10675 RH Reactor RH03-15375 RH Reactor 734 RH03-10345 RH Reactor 7 RH03-1540S RH Reactor 718 RH03-10355 RH Reactor 733 RH03-15415 RH Reactor 700 RH03-10365 RH Reactor 719 RH03-15445 RH Containment 738 RH03-10375 RH Reactor 717 SH04-10205 RH Reactor 682 RH03-10385 RH Reactor 718 Rh04-10215 RH Reactor 681 RH03-10445 RH Reactor 734 RH04-1022S RH Reactor 700 RH03-10465 RH Reactor 703 RH04-10235 RH Reactor 703 RH03-10475 RH Reactor 736 RH04-10245 RH Reactor 703 RH03-10495 RH Reactor 700 RH04-10255 RH Reactor 703 RH03-10515 RH Reactor 729 RH04-10275 RH Reactor 703 RH03-15005 RH Containment 738 RH04-10285 RH Reactor 703 RH03-15025 RH Containment 738 RH04-10295 RH Reactor 703 RH03-15035 RH Containment 738 RH04-10315 RH Reactor 703 l RH03-15045 RH Containment 738 RH04-10325 s RH Reactor 703 RH03-15055 RH Containment 740 RH04-10335 '

. RH Reactor 703 RH03-15065 RH Containment 742 RH04-10355 RH Reactor 700 RH03-15075 RH Containment 728 RH04-10365 RH Reactor 692 RH03-15085 RH Containment 738 RH04-10385 RH Reactor 688 RH03-15095 RH Containment 738 RH04-10395 RH Reactor 688 RH03-15115 2H Containment 738 RH04-10405 RH , Reactor 688 RH03-15125 RH Containment 738 2H04-10445 RH Reactor 682 RH03-15135 RH Containment 738 RH04-10455 RH tReactor 682 RH03-15145 RM Containment 738 RH04-10465 RH ' Reactor 688 RH03-15155 RH Containment 741 RH04-1051S RH Reactor 682 RH03-15165 RM Containment 741 RH04-1416S RH Containment 738 RH04-14185 RH Containment 738 RH12-10715 RH Reactor 688 RH04-14205 RH Containment 738 RH12-10725 RH Reacter 697

, RH04-15005 RH Containment 738 RH12-10775 RH Reactor 682 '

LA SALLE - UNIT 1 3/4 7-37 Amendment 1

i Table 3.7.9-2 (Continued) ,'

Nubber stas snubber installed on Snubber System snubber installed

. . 1 ation and elev. No. on location and elev.

RH04-15015 RM Containment 738 RH13-10095 RH Reactor 720 RH04-15025 RH Containment 738 RH13-10HS RH Reactor 719 RH04-15035 RH Containment 738 RH13-10415 RH Reactor 702 RH04-15065 RH inment 754 RH13-10855 RH Reactor 687 RH04-15085 RH tainment 744 RH13-10865 RH Reactor 682 RH04-15135 RH Rea r 703 RH13-10885 RH Reactor 698 RH04-15155 RH Reac r 688 RH13-10895 RH Reactor 698 RH04-15165 RH Reac 682 RH13- n005 RH Reactor 686 RH04-15185 RH Reacto 734 RH13- n015 RH Reactor 695 RH04-15195 RH Reactor 734 RH13- n105 RH Reactor 690 RH04-15205 RH Reactor 703 RH13- u nS RH Reactor 702 RH04-15215 RH Reactor 703 RH13-in25 RH Reactor 703 RH04-15225 RH Reactor 688 RH13-1H3S RH Reactor 683 RH05-10105 RH Reactor 679 RH13-in45 RH Reactor 683 RH05-10125 RH Reactor 679 RH13-In95 RH Reactor 689

'RH05-10485 RH Reactor 683 RH13-u205 RH Reactor 705 RH05- n10S RH Reactor 683 RH13-11225 RH Reactor 705 RH05- n135 RH Reactor 683 RH13-u23S RH Reactor 705 RH05- H145 RH Reactor RH13-n265 RH Reactor 701 RH05-11205 RH Reactor 703 RH13- D275 RH Reactor 701 RH05-11225 RH Reactor 6 RH13-11315 RH Reactor 701 RH12-10075 RH Reactor 687 RH13- n325 RH Reactor 701 RH12-10095 RH Reactor 697 RH13- n345 RH Reactor 706 912-1010S RH Reactor 697 RH13- n355 RH Reactor 706

.12-1011S RH Reactor 697 RH13-n375 RH Reactor 705 RH12-10125 RH Reactor 697 RH13- n385 RH Reactor 708 RH12-10135 RH Reactor 695 RH13hn395 RH Reactor 707 RH12-10145 RH Reactor 695 RH13-n405 RH Reactor 708 RH12-10185 RH Reactor , 682 RH13-1141S RH Reactor 712 RH12-10605 RH Reactor 690 RH13-n425 RH Reactor 713 RH12-10675 RH Reactor 692 RH13-n545 RH Reactor 719 RH12-10695 RH Reactor 695 RH13- n555' RH Reactor 705 RH12-1070S RH Reactor 695 RH13- n565 RH Reactor 705 RH13- n615 RH Reactor 7n RH23-10395 RH Reactor 706 l RH13-n65S RH Reactor 720 RH23-1040s RH Reactor 704

, RH13- n665 RH Reactor 719 RH23-10415 RH Reactor 705

\

RH14-1012S RH Reactor 724 RH23-10425 RH Reactor 705 RH14-10135 RH Reactor 724 MH23-10435 RW Reactor 720 RH14-10145 RH Reactor 724 RH23-10445 Reactor 721 RH\ Reactor RH14-1016S RH Reactor 723 RH23-10455 RH 720 RH14-10185 RH Reactor 707 RH23-10465 RH Reactor 720 RH14-10465 RH Reactor 717 RH23-10475 RH Reactor 707 RH14-10475 RH Reactor 724 RH23-10485 RH sReactor 707 RH14-1050S RH Reactor 708 RH23-10525 RH heactor 717 RH14-1051S RH Reactor 723 RH23-10535 RH Rhetor 720 RH14-10525 RH Reactor 718 RH26-10075 RH Reactor 703 RH14-1C555 RH Reactor 734 RH28-10015 RH Reactor 720 RH15-10135 RH Reactor 703 RH29-18045 RH React'or 696 ,

\

LA Salle - UNIT 1 3/4 7-38 g

,-n.--,.,,-,v,m,-,-, ,,ew,,v,r,,wo-- ,,,-,-e--,-.---.wy we-, - , - . . , a.,

l Table 3.7.9-2 (Continued) [,

I '

' Snubber System snubber installed on No. Snubber System snubber installed location and elev. No. on location and elev.

RH18-10075 Reactor 703 RH18-10095 RH33-10535 RH Reactor 746 Reactor 709 RH33-1054S RH18-10115 RH Reactor RH Reacter 747 715 RH33-1055S RH Reactor RH18-10155 RH Reactor 746 705 RH34-10325 RH Reactor RH19-10095 RH Reactor 740 718 RH34-10335 RH Reactor RH19-10105 RH Reactor 740 715 RH39-10255 RH Reactor RH19-10135 RH eactor 732 718 *RH39-10265 RH Reactor 732 RH19-10145 RH ctor 719 RH39-10275 RH Reactor 732 RH19-10155 RH Re tor 719 RH40-10285 RH Reactor 733 RH19-10165 RH Rea or 718 RH40-10295 RH Reactor 734 RH19-10335 RH Reac n 723 RH40-10305 RH Containment 794 RH19-10345 RH Reactok 722 RH40-10325 RH Containment 792 RH19-10365 RH Reactor 717 RH40-10335 RH Containment 782 RH19-10375 RH Reactor 721 RH40-10345 RH Containment 778 RH19-10415 RH Reactor 717 RH40-10365 RH Reactor 757 RH19-1042S RH Reactor 716 RH40-10375 RH Reactor 729 i RH19-10435 RH Reactor 717 RH40-10395 RH Reactor 734

! RH19-10445 RH Reactor 717 RH40-10405 RH Reactor 734 RH19-10455 RH Reactor 715 RH40-10415 RH Reactor 734 RH23-10335 RH Reactor 712 RH40-10425 RH23-10365 RH Reactor 733 RH Reactor 721 RH40-10435 RH Reactor RH40-10445 729 RH Reactor 30 RH50-10025 RH Reactor RH40-10475 675 RH Reactor 734 RH50-10045 RH Reactor 675 RH40-10485 RH Reactor 750 RH50-1017S

.d440-10495 RH Reactor 679 RH Reactor- 731\ RH53-10165 RH Reactor RH40-14015 695 RH Reactori 729 \ RH53-10185 RH Reactor 718 RH40-15005 RH Containment 799 W 53-10205 RH40-15015 RH Reactor 722 RH Containment 799 AH53-10215 RH Reactor 725 RH40-15045 RH Containment 774 RN53-10225 RH Reactor 725 RH40-15055 RH Containment 774 RH5 -10245 RH40-1506S RH Reactor 731 RH Containment 784 RH53 0255 RH Reactor 763 RH40-15225 RH Containment 797 RH53- 265 RH Reactor 7E0 RH40-15235 RH Containment 797 RH53-1 7S RH Reactor 775 RH40-15245 RH Containment 793 RH53-10 S RH Reactor 780 l

RH40-15265 RH Containment 797 RH53-1029 i

RH Reactor 800 RH40-15395 RH Reactor 734 RH53-10305 RH Reactor 780 RH40-15405 RH Reactor 734 RH53-10315 RH Reactor 777 RH40-15415 RH Reactor 734 RH53-1032S RH Reactor 776 RH40-15435 RH Reactor 733 RH53-10695 RH Reactor 725 RH40-15445 RH Reactor 734 RH53-15505 RH Containment 796 RH40-15485 RH Reactor 734 RH53-15515 Containment 796 RH40-15495 RH Reactor 735 RH53-15535 Containment 798 RH40-15505 RH Reactor 757 RH53-15545 RH Containment 796 RH40-15515 RH Auxiliary 732 RH53-15575 RH Containment 794 RH40-15525 RH Reactor 747 RH53-15615 RH Containment 783

! RH40-15535 RH Reactor 750 RH53-15625 RH Containment 779 RH40-15545 RH Reactor 757 RH53-15655 RH Reactor 776 RH40-15575 RH Reactor 764 RH53-15685 RH ntainment 793.

RH40-15595 RH Containment 771 RH53-15735 RH R ctor 770J 1.A Salle - UNIT 1 3/4 7-39 M

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

. . l Tablo 3.7.9-2 (Continued) .

' Snubber stem snubber installed on Snubber No. I ation and elev. 'estas snubber installed !

No. on location and elev. l RH40-15605 RH Containment 794 RH53-15745 RH Reactor 771 !

RH40-15615 RH Containment 799 RH53-15755 RH40-15725 RH Reactor 771 RH Reactor 734 RHS6-1003S RH Reactor 688 RH40-15735 RH eactor 728 RH56-10075 RH41-10915 RH RH Reactor 688 r 708 RH59-10305 RH Reactor 762 RH42-10325 RH Re&ctor 705 RH59-10315 RH42-10335 RH Reactor 762 RH Readtor 706 RH59-1048S RH Reactor 758 RH42-10375 RH Reacthe 735 RH59-10495 RH Reactor 738 RH59-10525 RH 758 RI01-1072S RI ReactoK Reactor 702 RH59-10565 RH Reactor 754 RI01-10735 RI Reactor 701 RH82-10305 RH Reactor 708 RI01-10745 RI Reactor 701 RH82-10375 RH Reactor 687 RIO1-10765 RI Reactor 692 RH82-10385 RH Reactor 686 RI01-10775 RI Reactor 693 RH82-10405 RH Reactor 686 RI01-10805 RI Reactor 685 RH82-10415 RH Reactor 687 RI01-10815 RI Reactor 685 RH82-10465 RH Reactor 695 RI01-10835 RI Reactor 676 RH82-10685 RH Auxiliary 698 RI01-10845 RI Reactor 678 RH82-1074S RH Auxiliary 698 R101-10855 RI Reactor 678 RH83-10115 RH Reactor 9 RIO1-10885 RI Reactor 679 RH83-10145 RH ' Reactor 689 RIO1-10895 RI Reactor 683 RH83-10155 RH Reactor 68K R101-10905 RI Reactor 683 RH83-10165 RH Reactor 684 RI01-10915 RI Reactor 683

~

l RH83-10175 RH Reactor 684 RI01-10925 RI Reactor 677 i RH83-10185 RH Reactor 684 101-10935 RI Reactor 678 RH83-10375 RH Reactor 689 R101-11015 RI Containment 718 RHA6-10035 RH Reactor 715 RIO1-1102S RI Containment 718 RHB4-10025 RH Containment 738 RI Containment RI0k11035 750 RH84-10055 RH Containment 738 RIO1-11065 RI Containment 776 RHB4-10075 RH Containment 742 RI01-11085 RI Containment 779

RH84-10085 RH Containment 742 RIO2-10675 RI Reactor 680 RHB4-10115 RH Containment 745 RIO2-10095 RI Reactor 687 R101-1006S RI Containment 753 RI Containment 742 RIO9-1005$g RI01-10075 RI Containment 771 RIO9-10075 RI Containment 742 RI01-10085 RI Containment 769 RIO9-10085 RI Containment 747 RI01-10095 RI Containment 769 RIO9-10095 RI Containment 747 RI01-10105 RI Containment 775 RIO9-10115 RI Containment 743 RIO1-10115 RI Containment 775 RIO9-10165 RI Containment 743 RI01-10125 RI Containment 774 RIO9-10215 I Containment 743 RIO1-10635 RI Reactor 743 RIO9-10245 R1 Containment 747 RIO1-10645 RI Reactor 743 RIO9-10255 Containment RIO1-10655 RI Reactor RIk 746 RI01-10675 RI Reactor 744 740 RIO9-10265 RIO9-10275 RI \

RI Containment 747 743 RI01-10695 RI Reactor 736 RI16-10165 RI \ Containment Reactor 688 s

RI01-10705 RI Reactor 736 RI16-10225 RI Reactor 683 RI16-10235 RI Reactor 683 RI41-10675 RI Reactor 703 RI16-10255 RI Reactor ^687 RI41-10685 RI 703 ReKctor RI24-10155 RI containment 793 RI41-10895 RI Reaqtor 703 RI24-10165 RI Containment 793 RI41-1092S RI React r 711 l i LA SALLE - UNIT 1 3/4 7-40 Amendment l

_ _ _ _ __ -y- - - - -

' ' ' ' ' ' ' ' * ^ ^ ^ ' " ' ' ' ' '

Table 3.7.9-2 (Continued) .

\ '

' Snubber System snubber installed on Snubber System snubber installed No. location and elev. No. on location and elev.

RI24-10175 k Containment 779 RI42-10105 RI Reactor 742 RI24-10185 RL Containment 780 RR00-10015 RR Containment 743 RI24-10195 RI Containment 759 RR60-1002S RR Containment 739 RI24-10205 RI Containment 760 RR00-10035 RR Containment 743 RI24-10215 RI containment 754 RR00-10045 RR Containment 759 RI24-10225 RI Containment 754 RR00-10055 RR Containment 743 RI24-10695 RI actor 741 AR00-10065 RR Containment 739 RI24-10705 RI actor 736 RR00-10075 RR Containment 739 RI24-10715 RI ctor 726 RR00-10085 RR Containment 739 RI24-10725 RI Re tor 726 RR00-10095 RR Containment 743 RI24-10805 RI Reac' tor 742 RR00-10105 RR Containment 748 RI24-11205 RI Conta nt 821 RR00-10115 RR Containment 748 RI24-11215 RI Contai nt 821 RR00-10125 RR Containment 759 RI24-u225 RI Containagnt 821 RR00-10:25 RR Containment 758 RI24-11245 RI containeett 826 RR00-10145 RR Containment 759 RI24-11305 RI Containment 810 RR00-10155 RR Containment 743 RI24-11315 RI Containment \ 809 RR00-10165 RR Containment 753 RI24-11325 RI containment 810 RR00-10175 RR Containment 753 RI24-15115 RI Containment 810 RR00-10185 RR Containment 743 RI24-15125 RI Containment 810 RR00-10195 RR Containment 765 R124-15135 RI Containment 785 RR00-10205 RR Containment 765 RI24-15145 RI containment 764 RR00-10215 RR Containment 761 RI24-15155 RI Containment 749 RR00-10225 RR Containment 761 1141-1056S RI Reactor 40 RR00-10235 RR Containment 758 RI41-10575 RI Reactor 712 RR00-10245 RR Containment 757 RI41-10585 RI Reactor 713s RR00-10255 RR Containment 743 RI41-10595 RI Reactor 708 RR00-10265 RR Containment 762 >

RI41-1060S RI Reactor 703 RR00-10275 RR Containment 762 RI41 10615 RI Reactor 705 R00-10285 RR Containment 765 RI41-10645 RI Reactor 705 RR00-10295 RR Containment 765 RI41-10655 RI Reactor 703 RR00-10305 RR Containment 743 RI41-10665 RI Reactor 703 RR00-10315 RR Containment 746

. _RR00-10325 RR Containment 746 RR28'10165 RR Containment 750 RR00-10415 RR Containment 748 RR28-10175 RR Containment 750 RR00-1042S RR Containment 750 RT01-10855 RT Containment 748 r RR00-10435 RR Containment 757 RT01-1087S RT Containment 739 '

l RR00-10445 RR Containment 759 RT01-10885 RT Containment 738 RR00-10455 RR Containment 743 RT01-10895' RT Containment 738 RR00-10465 RR Containment 753 RT01-10915 RT Containment 738 RR00-10475 RR Containment 754 RT01-10935 RT Containment 738 RR00-10485 RR Containment 743 RT01-10945 RT Containment 738 RR00-10495 RR Containment 768 RT01-10965 RT Containment 738 RR00-10505 RR Containment 765 RT01-10975 RT Containment 738 i RR00-10515 RR Containment 764 RT01-11005 T Containment 738 l RR00-10525 RR Containment 764 RT01-11295 Containment 738 RR00-1053R RR Containment 743 RT01-11305 RT Containment 739 RR00-10545 RR Containment 762 RT17-10025 RT Containment 748 RR00-10555 RR Containment 762 RT17-10085 RT Containment 746 RR00-1056S RR Containment 765 SC02-10155 Reactor SC 814 !,

l s -

l

! LA Salle - UNIT 1 3/4 7-41 [

knendment 1 I l

l l

.. e l Table 3.7.9-2 (Continued) (

l Snubber System snubber installed on Snubber System snubber installed'

__. No. location and elev. No. on location and elev.

RR00-10575 Containment 765 SC02-10275 RR00-10585 SC Reactor 781 R Containment 739 SC02-10365_ SC Containment 774 RR00-10595 RR Containment 743 SC02-10385 RR00-10605 SC Containment 770 RR Containment 746 SC02-10475 SC Containment 757 RR00-10615 RR Containment 746 SC02-10555 RR00-10625 SC Containment 750 RR Containment- 748 VG01-00015 VG Reactor 831 RR01-10325 RR ntainment 743 VG01-00055 RR07-14325 VG Reactor 831 RR ntainment 738 VG01-00065 VG Reactor 831 RR17-10015 RR Co tainment 738 VG01-00085 RR17-10025 VG Reactor 831 RR Con inment 737 VG01-00105 VG Reactor 831 RR17-10035 RR Con inment 737 VG02-1004S RR17-10045 VG Reactor 794 RR Conta nt 737 VG02-10055 VG Reactor 794 RR17-10055 RR Contai nt 737 VG02-18005 RR17-10065 VG Reactor 802 RR Contai nt 737 VG04-10035 VG Reactor 821 RR17-10075 RR Containee t 737 VG04-1005S RR17-10085 VG Reactor 809 RR Containeen 737 VG04-10115 VG Reactor 794 RR28-10075 RR Containment 756 V604-10145 VG Reactor 794 RR28-10125 RR Containment 750 VG04-10155 VG Reactor 798 RR28-10155 RR Containment 741 VG04-10195 VG Reactor 798 VQO2-10355 VQ Reactor 805 LP-25-H045 LP VQ02-10405 VQ Reactor 810 N.P. 30A RR-59 VQ05-10015 VQ Reactor 733 RR68-N-4 RR VQ05-10095 VQ Reactor 38 RR68-N-6 RR ,

N.P. 83 RR-59 N.P. 85 RR-59 RR69-H-4 RR FHP1204-H02 HP

\N.P.164 N.P. 170 RR-59 RR-59 FHP1204-H03 HP RR-59 FHP1204-H02

% P. 175 HP N.R. 180 RR-59 FHP1204-H03 HP N.PA 240 RR-59 N.P. 288 MS-51A HG-08-H035 NG N. P. 38 MS-51A HG-081H085 HG N.P. 41 MS-51A HG-08-H075 HG N.P. 48A MS-51A N.P. 114s HG-61 N.P. 2968 MS-51A N.P. 115 \ HG-61 N.P. 336B MS-51A N.P. 116 \ HG-61 N.P. 376B MS-51A HG21-PO4 \ HG M.P. 855 MS-51A RH N.P. C43 FRH1213-H15 \

MS-51A FRH1213-H14 s RH N8-125-H075 RH FRH1213-H12 \ RH N.P. 465 VG-03 FRH1213-H11 \RH N.P. 475-X VG-03 FRH1213-H09 RH N.P. 475-Z VG-03 FRH1213-H07 N.P. 308 RN-75A FRH1213-H08 RK RH N.P. 30A RH-76 FRH1213-H06 RH N.P. 308 RH-76 FRH1213-H03 RH N.P. 135B RH-76 FRH1213-H02 RH 4.P. 112 HG-61 FRH1213-H01 RH LA SALLE - UNIT 1 3/4 7-42 ndment 1

Tablo 3.7.9-2 (Continued) (

., z .. , .

Snubber Sys snubber installed on snubber .4fsierii snubber insta11.d No. locati n and elev. -

No,. 7 , j. 4 bH 10 cation and elev.

HP-M-H02S HP- U-H045 HP HP Phdijhij d l PR$209903 i i

N.P. 125 RR-59 FRM18Q$* i$ i M.P. 135 RR-59 RRH120 o 05 s NG-06-H125 HG NP$$ i-26 FAH1IO - hl i N.P. 52 RM-68 RRH 3 40 1 N.P. 63 RM-68 FRH 0}-7 H N.P. 95 RH-68 FRN $f-Q 08 y RH52-H07 RH a FR11210* 401 RH52-H08 RH FR21209-H05 L RH52-H09 RH FRN1232H10 RH52-H06 RH FNH1fji-H04 :J l RH52-H05 RH PR 1232-H06 7 RH52-H02 RH l-7:,

@a a 23 RH52-H03 RH N. : Aid -C-71 N.P. 133 RH-68 FAHif06-Nil .H.

FRH1211-H01 RH FRH1206 H18 FRH12 H-H04 RH Id. P. . ; [H 0-FRH1211-H03 RH N.Pr ~: I . kt-FRH42-1037 RH N'. P. . l,8*! $4a l

H-4RH-87 H-2RN-88 RH \ N.Pi f'3 : LO-RH \ N.P. 23-1 LC 85 N.P. 120 RH-69 \N.P. 23?X C-58 N.P. 135 RH-69 FRI1207dhdi..- ,

FRH1214-N-125 CS 1RIO78-2 1-H08 h

! FRH1230-H07 RH 1RIO78-2 1-W06 R:

l FRH1230-H06 RH 1RIQ78-2-1-H05 RI l FRH1230-H05 RH 1RIO78-2-1-H16 RI . -

l FRH1230-H03 RH N.P. 53 RH-fi

! FRH1230-H02 RH N. P. 90 . RH-21 i FRH1230-H01 RM N.P.-100 RH-21 l FRH1206-H03 RH N.P. 125\ RH-21 FRH1233-H05 RI MS-52-H065 MS FRH1233-H04 RI MS-53-H06$s MS FRH42-1031 RH FRH42-1032 RH MS-53-H055\

MS-50-H025 .

Pl$

MS FRH42-1033 RH MS-50-H045 \ MS N.P. 135 RH-16 MS-51-H055 s MS FRH1231-H09 RI RT-33-H085 FRH1231-H06 RI RT-33-H095

\M$

tS FRH1231-H04 RI RT-33-H105 P5 RH25-H01 RH N. P. $6 RA C$

RH25-H02 RH N.P. 52 R 3 RH25-H04 RH M-1362-if-169 $d RH25-H03 RH M-130faff=110 RH RH25-H05 RH 19-1302-22-112 Ri4 RH25-H06 RH M-1302-22-113 RH HG-04-H045 HG M-1302-23-96 MS N.P. 117 HG-61 N.P. 60A RH-C3 LA SALLE - UNIT 1 3/4 7-43 Amendment 1

_____.__.___..._._._-_..-,..,,..,,-,__.,-.._,-.,.,.,,...~..,_v.. , . . . . . , , . . . - . _ _ , - . . . . . . _ _ . _ _

~

8 Table 3.7.9-2 (Continued) .'

I.

c t Snubber 40.

Sy tem snubber installed on I Situbber j tion and elev. No. , U)thit endbdr installed N.P. 118

, ,. t, iscation and elev.

HG-6

=

N.P. 119 HG-61 PPM 0 =24 144 7 N.P. 15 NS-83 5180a]*14*149 E P. 558-Y # 1402 24 181 RH-G1 N.P. 558-Z W1302*84*188 RN-G1 N.P. 40 HP-Al 51302*f4e181 M 1302-14-48 NB W1304*24*1H I M=1302*24=41 s v

N.P. 2558 RN-G1 6

l. N.P. 50 l N.P. 65 NB-C1 NB-C1

, W1302*255l

'W130$A8* 3 f

! M-1302-15-3 Mt1302*28* 2 RF RH M-1302-15-56 . ll-1302=28*110 Af 1

RH l

M-1302-15-72

\ M-1302- A W1302-16-35 RH N8 EP.08249134 16 R -C8 l W1302-16-26 IS M-110i=ffIis nf W 1302-16-54 NB W1402=27-12% N

\ W1302*21*128 R M-1302-16-55 NB M-1302-16-23 RH

\ $ 1302 21*128 A

\ M-1302*27*129 Ri

>1302-16-51 RH \ M-1302*37*$2 M-1302-16-52 RH R M-1302dg7a13g g

. N.P. 1708 HP \ M-1302-28974 l

W 1302-20-205 l

M-1302-20-208 MS MS \g M-1302-28-78 dlJ M-1302-28-84 26 M-1302-20-209 MS

~

61302-20-212 g M-1302-30-52 R)

MS s M-1302-30-54 Rt

.r-1302-20-145 MS .

\M-1302-30k41 R6 M-1302-20-131 N8 U

N.P. 375A-Y It-1302-35-41 HP-Al M 1302-35-43 U N.P. 375A-Z HP-Al ' 51302-36-45 R1 M-1302-36-138 RI M-1302-36-131 Al t

M-1302-36-141 RI M-1302-23-95 Mi 5 1302-36-142 RI M-1302-23-140 Mt

> 1302-36-148 RI M-1302-23-143 M!

M-1302-36-152 RI M-1302-23-124 M1 M-1302-36-154 RI

M-1302-36-169 RI M-1302-36-94 RI i

M-1302-36-167 RI M-1302-9-1 N8 M-1302-36-133 RI M-1302-36-146 RI l

M-1302-36-147 RI M-1302-36-150 RI M-1302-36-151 RI M-1302-36-125 RI M-1302-36-126 RI M-1302-36-170 RI W 1302-21-40 RR N. P. 45 RR-A7 LA Salle - UNIT 1 3/4 7-44 Amendment l pane *eem

w T.----gg-r y9 -.-.ar *---g -my-e -- -ge-.w-. e-r-- ------w--------.c- ve- 6 mew---..-o--.w. .------wr

~

e Tablo 3.7.9-2 (Continued) .

Snubber Sys snubber installed on Snubber sydted$hubberihstalled

%. loca on and elev. . No , , . DM 1egation and elev.

N.P. 118 NG-51 M-1302-34*i i

N.P. 119 NG-61 ih

>1302*f4*149 RR N. P. 15 NO-83 M 1302-244137 N.P. 558-Y RN-G1 M1302d24'13d N.P. 558-Z RN-G1 M-1302*24?151 N. P. 40 HP-Al M= UO2 24 152 f M-U02-14-48 NO W 1302 25-41 96 N.P. 2558 RWG1 M-u02-25-8 RPl N.P. 50 NO-C1 M1302626-10$ RH N.P. 65 NO-C1 M-1302-264132 RN M-1302-15-3 RH M-1302 26-110 Nf M-1302-15-56 RH M-1302-26-134 RE W1302-15-72 RH N.P. C8-10.. . RH C8 W 1302-16-35 NB M-1302-27-113 RH i

W 1302-16-26 NO M-1302-27-121 RH M-1302-16-54 NO M-1302-27-122 RH M-1302-16-55 2 5 1302-27-128 RH W U02-16-23 RH M-1302-27-129 RH M-1302-16-51 RH M-1302-27-83 RH l W 1302-16-52 RH M-1302-27-139 RH N.P. 1708 HP M-1302-28-74 LP M-1302-20-205 MS M-1302-28-76 LP M-1302-20-208 MS W 1302-28-84 RH M-1302-20-209 MS M-1302-30-52 RH

~ S 1302-20-212 MS j A-1302-20-145 MS \M-1302-30-54 M-1302-30-41 RH RH M-1302-20-131 NB .

h1302-35-41 LC N.P. 375A-Y HP-Al W 1302-35-43

, LC N.P. 375A-Z HP-Al M-1302-36-45 RI M-1302-36-138 RI M-1302-36-131 RI M-1302-36-141 RI M-1302-23-95 MS W1302-36-142 RI W 1302-23-140 MS M-U02-36-148 RI M-1302-23-143 MS M-1302-36-152 RI M-1302-23-124 MS M-1302-36-154 RI M-1302-36-169 RI \

M-1302-36-94 RI M-1302-36-167 RI M-1302-9-1 M M-1302-36-133 RI M-1302-36-146 RI M-1302-36-147 RI g M-1302-36-150 RI \

M-1302-36-151 RI M-1302-36-125 RI M-1302-36-126 RI W 1302-36-170 RI M-1302-21-40 RR N.P. 45 RR-A7 LA Salle - UNIT 1 3/4 7-44 Amendmen 1l

t. .

Table 3.7.9-2 (Continued) !

s Snubber Systes snubber installed on Snubber System snubber installed No. locatio and elev. No. on location and elev.

M-1302-22-83 RR M-1302-22-33 RR 5 1302-22-127 RR M-1302-24-110 RR M-1302-24-111 RR M-1302-24-106 RR

i. 5 1302-24-107 RR M-1302-24-103 RR M-1302-21-183 MS M-1302-21-181 MS N.P. 858 MS-C1 N.P. 65-Z MS-C2 N. P. 6'i-Y MS-C2 M-1302-21-162 MS

,, M-1302-21-163 MS M-1302-21-74 MS M-1302-21-189 MS l

HP75-20095 HP Ofesel Generator 49 2HG-24-H02 2HG-62 HP75-20145 HP Ofesel Generator 5 2HG-16-H01 2HG-62 HP75-20175 HP Diesel Generator 747 RH-27H-H035 2HG-61 l HP75-28005 HP Diesel Generator 743\ RH-27H-H065 2HG-61 l RH83-20065 RH Reactor 684 \ 2HG-4H-H105 2HG-61 l 'tH83-20075 RH Reactor 689 .2HG-15H-H03S 2HG-61 l AH83-20095 RH Reactor 684 '2HG-15-H055 2HG-61 RH83-20125 RH Reactor 689 2HG-15H-H015 2HG-61 RH83-20135 RH Reactor 684 2DG-66-H03 2CS-75 RH83-20165 RH Reactor 689 2HG23-H06 2HG-70 l SC02-29675 SC Reactor 782 2HG22-H09 2HG-70 VG01-00155 VG Reactor 831 2VQ-10,-H02 2VG-03 VG01-00165 VG Reactor .

831 2VQ-10-H04 2VG-03 VG01-20195 VG Reactor 830 2VQ-11-H01 2VG-03 VG01-20215 VG Reactor 803 2VQ-11-HQ3 2VG-03 VG02-00125 VG Reactor 831 2VQ-11-H06 2VG-03 VG02-20195 VG Reactor 803 2VQ-11-H04\ 2VG-03 VG02-20215 VG Reactor 831 \

VG02-20225 VG Reactor 794 \

g

! VG02-2023S VG Reactor 826 i l VG02-20245 VG Reactor 831 \

l VG03-20015 VG Reactor 794 \

\

VQ04-20195 VQ Reactor 738 g VQ04-20205 VQ Reactor 733 s ,

VQ05-20235 VQ Reactor 762 \

VQ05-20245 VQ Reactor 762 VQ05-20255 . VQ Reactor 787 \

VQ05-20265 VQ Reactor 732 VQ05-20315 VQ Reactor 733 VQ04-28005 VQ Reactor 733 VQ06-20025 VQ Auxiliary 810 LA SALLE - UNIT 1 3/4 7-45 Ame ent 1

PLANT SYSTEMS k 3/4.7.10 MAIN TUR8INE 8YPASS SYSTEM LIMITING CONDITION FOR OPERATION 3.7.10 The main turbine bypass system shall be OPERA 8LE.

APPLICABILITY: OPERATIONAL CONDITION 1 w ereyel -to ar$ ef it.ATso Twen4Ac. ke, how&ea Therm gt 4. PousaK is3 ACTION: With the main turbine bypass system inoperable, within.$wo hours l restore the system to OPERA 8LE status or reduce THERMAL POWER to less than 25%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.10 The main turbine bypass system shall be demonstrated OPERA 8LE at least once per:

a. 7 days by cycling each turbine bypass valve through at least one complete cycle of full travel,

b. 18 months by:

( 1. Performing a system functional test which includes simulated l automatic actuation and verifying that each automatic valve actuates to its correct position.

t ""-*^ =ia; e ""^r'~L CALI""^ TIN e' _

+h- -ia te-biae Ep;rrr I

.u,+- +, +<-- t..+. - -. ss--

42. Demonstrating TUR8INE SYPASS SYSTEM RESPONSE TIME to be less than or equal to 200 milliseconds t; ; .;h: ;;;. i ti e,.

- - ' ^ ' -

mt::h..;. ^

l

! 6 2-et m = -- -- --! et- + 1 +- - e te--f-e e 8 : St--te- +- t :- :- -

5 "' ? ::^--'-t '- *--;; -' ' MftetM velre - Si tf^- 95!' ' 5""-4+t-d te

" "- -- 8 :! '^ ef tM* - M ty: -* tt:t :: ;?:t' LA SALLE - UNIT 1 df I 3/4 7-)33 Amenement 1 i @ >g-

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

l

+

i N- ..

,/ -

L 3/4.8 ELECTRICAL POWER SYSTEMS

!. .~

3/4.8.T A.C.~ s0uRCEs 9

A.C. SOURCES - OPERATING ,

t -

, LIMITING CONDITION FOR OPERATION

3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and

b. Separate and independent diesel generators 0,1A, 2A and 18 with:

1. For diesel generator 0, IA and 2A

a) A separata day fuel tank containing a minimum of

~

250 gallons of fuel.

~

b) A separate fuel storage systas containing a minimum of 31,000 gallons of fuel.

2. For diesel generator 18, a separate fuel storage tank / day tank .

containing a minimum of 29,750 gallons of fuel.

_/ 3. A separate fuel transfer pump. -

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

j a. iiHth either one offsite circuit or diesel generator 0 or IA of the j above required A.C. electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remaining A.C. ources by performing Surveil-

! i lance Requirements 4.8.1.1.1.a within hour, and 4.8.1.1.2.a.4., I i for one diesel gen.erator at a time, within four hours, and at least i once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two offsite circuits j and diesel generators 0 and IA to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or i

' be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one offsite circuit and diesel generator 0 or IA of the above required A.C. electrical power sources inoperable, demonstrate the q

OPERASILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within hour, and 4.8.1.1.2.a.4, for one 1

j -

diesel generator at a time, within three hours, and at least once per .

l 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable A.C.

sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be irt at least NOT j SHUTCCWN witnin tne next 12 hJurs and in COLD SH'TW1 within tne j ,, followinc .'; :urs. Restors at least two offsita -ir::uits are diesel gener.U.crs G anc 1A ta OPERABLE status wnain 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frcm

~ the time of initial loss or be in at least HOT SHUTDOWN within the

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDCWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l I

?

! [

LA SALLE - UNIT 1 3/4 8-1

ELECTRICAL POWER SYSTEMS i.

LIMITING CONDITION FOR OPERATION (Continued) g ACTION (Continued) i l

c. With both of the above required offsite circuits inoperable. l demonstrate the OPERABILITY of the remaining A.C. sources by per- ;

forming Surveillance Requirement 4.8.1.1.7fa.4, for one diesel l l generator at a time, within four hours, and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months '

thereafter, unless the diesel generators are already operating; restore at least one of the ino~perable offsite circuits to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the i

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to OPERABLE status wfthin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With diesel generators 0 and 1A of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remain-ing A.C I withinM.sourcesbyperformingSurveillanceRequirements4.8.1.1.1[a 5e hour and 4.8.1.1.2.a.4, for one diesel generator at a

' l time, within two hours and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; a

restore at least one of the inoperable diesel generators 0 and 1A to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within P,

the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Restore both diesel generators 0 and 1A to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of ini'ti:1 loss or be in at least HOT SHUTDOWN V within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

e. With diesel generator 1B of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.34a within bee hour, and 4.8.1.1.2.a.4, for one diesel generator at a l time, within three hours, and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore the inoperable diesel generator 18 to OPERABLE status within

, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

f. With diesel generator 2A of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.34a and l

, 4.8.1.1.2.a.4, for diesel generator 1A, within one hour, and at I

least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore the inoperable diesel ,

generator 2A to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare standby gas treatment system subsystem B, Unit 2 drywell and ' suppression chamber hydrogen recombiner system, and control room and auxiliary electric equipment room emergency filtration system train 8 inoperable 3.6.6.1.,

, and take eo.+;poed and 3.7.2j the ACTION required perforwasca of by Specifications Surestileuce. 3.6.5.3,ts 4.g. l.l. lo_.

R,p,trww l *"a 4.8. t .t. JRa.4 4e d,esel Om sys4 ems ars deched Senem+oe tA is noE reydr*l ?"A*O*- .

spa:th+: ens is +streu. pisMe.g uncnw f 4k& res(ocke- w LA SALLE - UNIT 1 3/4 8-2

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. At least once per 18 months during shutdown by: J

1. Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesel generator capability to reject a load of greater than or equal to 1190 kw for diesel generator 0, greater than or equal to 638 kw for diesel generators 1A and 2A, and greater than or equal to 2381 kw for diesel generator 1B while maintaining engine speed less than or equal to 75% of the difference between nominal speed and the overspeed trip setpoint or 15% above nominal, whichever is less.

3. Verifying the diesel generator capability to reject a load of 2600 kw without tripping. The generator voltage shall not exceed 5000 volts during and following the load rejection.

4. Simulating a loss of offsite power by itself, and:

a) For Divisions 1 and 2 and for Unit 2 Division 2:

1) Verifying de-energi2ation of the emergency busses and load shedding from the emergency busses.

2) Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 13 seconds, energizes the auto-connected loads and operates for greater than or equal to 5 minutes 6.while its generator is so loaded. After v energization, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 2 150 volts and 60 1 1.2 Hz during this test.

b) For Division 3:

1) Verifying de-energization of the emergency bus.

2) Verifying the diesel generator starts on the auto-start signal, energizes the emergency bus with its loads with-l in 13 seconds anc' operates for greater than or equal to l 5 minutes while its generator is so loaded. After energization, the steady state voltage and frequency t

of the emergency bus shall be maintained at 4160 t j 150 volts and 60 1 1.2 Hz during this test.

I

5. Verifying that on an ECCS actuation test signal, without loss of offsite power, diesel generators 0, IA and IB start on the auto-start signal and operate on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 + 416, -150 volts and 60 + 3.0, -1.2 Hz within 13 seconds after the auto-start signal; the steady state generator voltage and I

frequency shall be maintained within these limits during this test.

t 6. Verifying that on a simulated loss of the diesel generator, I

with offsite power not available, the loads are shed from the k emergency busses and that subsequent loading of the diesel generator is in accordance with design requirements.

y .

LA SALLE - UNIT 1 3/4 8-4 I

l l

l ELECTRICAL POWER SYSTEMS l SURVEILLANCE REQUIREMENTS (Continued) i J. Simulating a loss of offsite power in conjunction with an ECCS I actuation test signal, and:

a) For Divisions 1 and 2:

1) Verifying de-energization of the emergency busses and load shedding from the emergency busses.

2) Verifying the diesel' generator starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 13 seconds, energizes the auto-connected emergency loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses I shall be maintained at 4160 1 416 volts and 60 1 1.2 Hz during this test.

b) For Division 3:

1) Verifying de-energization of the emergency bus.

l 2) Verifying the diesel generator starts on the auto-start signal, energizes the emergency bus with its loads within 13 seconds and operates for greater.than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady l state voltage and frequency of the emergency bus l shall be maintained at 4160 1 416 volts and 60 1 1.2 Hz l during this test.

14 Verifying that all diesel generator 0, lA and 18 automatic trips l except the following are automatically bypassed on an ECCS actuation signal:

a) For Divisions 1 and 2 - engine overspeed, generator l

differential current, and emergency manual stop.

b) For Division 3 engine overspeed, generator differential or overcurrent, and emergency manual stop.

8*

l 4 Verifying the diesel generator operates for at least 24 hnurs. l l During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded to greater than or equal to 2860 kw and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to 2600 kw. The generator voltage and frequency shall

! be 4160 + 420, -150 volts and 60 + 3.0, -1.2sHz within 13 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test, perform Surveillance Requirement 4.8.1.1.2.d.4.a).2) and b).2).*

LA SALLE - UNIT 1 3/4 8-5

t

(

_ .. . . W. .

i c , . . -

-- } , .

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

I .s

, T ,W. Verifying that the auto-connected loads to each diesel generator l do not exceed the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months reting of 2860 Ipr.y /

JCL. R . Verifying the diesel generator's capability to: l

a) Synchronize with the offsite power source while the i generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Se restored to its standby status.

fl. ,3C Verifying that with diesel generator 0, IA and IB operating in a l test mode and connected to its bus:

i a) For Divisions 1 and 2, that a simulated ECCS actuation signal overrides the test mode by returning the diesel -

generator to standby operation.

b) For Division 3, that a simulated trip of the diesel generator overcurrent relay trips the SAT feed breaker to bus 143 and that the diesel generator continues to supply normal bus loads. -

L 1

- g [13. Verifying that with all diesel generator air start receivers

~

f pressurized to less than or equal to the comoressors auto-start }

j setpoint and the compressors secured, diesel generators 0, IA l

and 2A start at least 5 times and diesel generator 18 starts at least 3 times from ambient conditions and accelerates to l 900 rpa + 5%, -2%, in less than or equal to 13 seconds.

j 11,34: Vertrying snat the automatic load sequence timer is OPERABLE I~

with the interval between each load block within 210% of its i design interval for diesel generators 0 and 1A.

IJ.AS'. Verifying that the following diesel generator lockout features }

prevent diesel generator operation only when required:

a) Generator underfrequency.

1 b) Low lube oil pressure.

lf c) High jacket cooling temperature Genet ator reverse power.

!l d) -

l e) Generator overcurrent.

t I -

f) Generator loss of field.

.[

l g) Engine cranking lockout.

l '

If surveillance Requirement 4.S.I.l.2.d.4.a)2) and/o- C ara not satisfacurily

) m co m leted. it is not necessary to receat the pre:e:ir.- F nour est. bs eae.

(. the diesel generator may be operated at 2600 ip/ for pae nour or until operating temperature has stabilized. y s

i 1

i

, LA SALLE - UNIT 1 ' 3/4 8-6 .

w -

m- s.,.-_%-v-m t- %e-W- C+--tw --r- e- -g o w- -y ,- w_ .ny,-wea- +-r ,w-g-ei- s-ew -,w m--ww%mv-- n*--w--,we-s-,--r-mwvw-----t-----------m--'w-- v--- *

.._ _ _ _... . .~ . ._

s i

N ~

.j ' ELECTRICAL POWER SYSTEMS .

3URVEILLANCE 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, IA and 18 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 oli storage tank, removing the act.umulated ,

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 systes designed to Section III, subsection NO, of the

ASME Code in accordance with ASME Code Section 11, Article IM)-5000. .

~

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.6.8. Reports of diesel generator failures shall include the information recommended in Regula-tory Postion 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,

.. 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 I DIESEL GENERATOR TEST SCHEDULE l . .

I Number of Failures in i

Last 100 Valid Tests

Test Frecuency I

<1 At least once per 31 days ll 2 At least once per 14 days

! 3 At least once per 7 days

>4 At least once per 3 days l'

"Griteria for aetermining number of failures and number of valid l

tests shall be in accordance with Regulatory Position C.2.e of

} Requiatory 3#d$ 1.103. Revision 1, August 1977, wrece the last ll.

100 tasts are aster:nined on a per nuclear unit basis.

m#

I l

LA SALLE - UNIT 1 3/4 8-7

, --, -, ,,--n-y ,- , . . - - , . , - - . . , - . , . , , , , . , _ - . , , , . -

l l

ELECTRICAL POWER SYSTEMS

{

l

, A.C. SOURCES - SHUTDOWN

\

LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class IE distribution system, and

b. Diesel generator 0 end/or 1A, and diesel generator 1B when the HPCS l system is required to be OPERABLE, and diesel generator 2A when the of'fsite power scurce for standby gas treatment system subsystem B or control room and auxiliary electric equipment room emergency filtra-tion system train B is inoperable and either or both systems are required to be OPERABLE, with each diesel generator having:

1. For diesel generator 0, IA and 2A:

a) A separate day fuel tank containing a minimum of 250 gallons of fuel.

b) A separate fuel storage system containing a minimum of 31,000 gallons of fuel.

8'

2. For diesel generator 18, a separate fuel storage tank / day tank -

containing a minimum of 29,750 gallons of fuel.

3. A fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *.

ACTION:

a. With all offsite circuits inoperable and/or with diesel generators 0

.end/or IA inoperable, suspend CORE ALTERATIONS, handling of irradiated \

fuel in the secondary containment and operations with a potential for draining the reactor vessel.

i b. With diesel generator 18 inoperable, restore the inoperable diesel i

generator 18 to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPCS

! system inoperable and take the ACTION required by Specification 3.5.2 and 3.5.3.

When handling irradiated fuel in the secondary containment.

LA SALLE - UNIT 1 3/4 8-8

ELECTRICAL POWER SYSTEMS

{ LIMITING CONDITION FOR OPERATION (Continued)

'O .

ACTION: (Continued)

c. With diesel generator 2A inoperable, declare standby gas treatment system subsystem 8 and control room and auxiliary electric equipment room emergency filtration system train B inoperable and take the ACTION required by Specifications 3.6.5.3 and 3.7.2.

d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS t.1

4. 8 At At least the above required A.C. electrical power scurces shall be }

demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1; 4.8.1.1.2 and 4.8.1.1.3, except for the requirement of 4.8.1.1.2.a.5.

1 I

t 1

1 l

l LA SALLE - UNIT 1 3/4 8-9 l

r. u . . - . --. .

e ,

. . ~

l l 5

. . i l gS -

ELECTRICAL POWER arasam -

t . .

l. ,

3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS l ; A. C. DISTRIBUTION - OPERATING e

LIMITING CONDITION FOR OPERATION ,

3.8.2.1 The following A.C. distribution system alectrical divisions shall be OPERABLE and energized:

a. Division 1, consisting of; i

1. 4160-volt bus 141Y.

2. 480-volt buses 135X and 135Y.

l i 3. 480-volt MCC's 135X-1, 135X-2, 135X-3, 135Y-1 and 135Y-2. I

4. 120-volt A.C. distribution panels in 480 volt MCCs 13SX-1, -

, 135X-2, 135X-3 and 135Y-1. .

~

! b. Division 2, consisting of; -

i -

i 1. 4160. volt bus 142Y.

! 2. 480-volt buses 136X and 136Y.

~

i c 3. 480-vcit MCC.'s 136X-1, 136X-2, 136X-3, 136Y-1 and 136Y-2. I

, 4. 120 volt A.C. distributtort panels in 480 volt MCCs 136X-1, 136X-2,136X-3 and 136Y-2.

c. Division 3, consisting of; j 1. 4160-volt bus 143.

l 2. 480-volt MCC 143-1.

3. 120-volt A.C. distribution panels in 480 volt MCC 143-1.

d. Unit 2, Division 1, consisting of;

1. 4160-volt bus 241Y.

2. .. h . 233X. S rcalter- B414 OP5eA81.E or' closek.

s =1t =: +-assx4-and-aa6x4,

a S. re! t ^. C. CL;. LtL., ,,;.i.eh P 'SO '?el t TC': 225% 2 end 2SEM4:

e. Unit 2 Division 2, consisting of; .

1. 4160-volt bus 242Y.

2. 480-volt buses 236X and 236Y.

3. 480-volt MCC's 236X-1, 236X-2, 236X-3, 236Y-1, and 236Y-2. I

4. 120 volt A.C. distribution panels in 480 volt MCC #

s 236X-1, i

225X-2, 236X-3, and 235Y-2.

.- APPLIC.'3!'.!TY: GPEMT!;.i.AL CCN01 TION 3 1, nd 3.

'LA SALLE - UNIT 1 3/4 8-10 i

.&+ - - - , ,- -,,,.,p__ -

.w.,,my y_y, , ,,,.,-,.,,.w-,_._,_-.,..-,_,,__-.,.,,,-.w. -

ELECTRICAL POWER SYSTEMS

( -A.C. DISTRIBUTION - SHUTDOWN

. . j LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, Division 1 W or Division 2, and Division 3 when the 1 HPCS system is required to be OPERABLE, and Unit 2 Division 2 when the standby gas treatment system and/or the control room.and auxiliary electric equipment room emergency filtration system are required to be OPERABLE, of the A.C.

distribution system shall be' OPERABLE and energized with:

a. Division 1, consisting of;

1. 4160 volt bus 141Y.

2. 480 volt buses 135X and 135Y.

3. 480 volt MCC's 135X-1, 135X-2, 135X-3, 135Y-1 and 135Y-2. I

4. 120 volt A.C. distribution panels in 480 volt MCCs 135X-1, 135X-2, 135X-3 and 135Y-1.

b. Division 2, consisting of;

1. 4160 volt bus 142Y.

2. 480 volt buses 136X and 136Y.

3. 480 volt MCC9 136X-1, 136X-2, 136X-3, 136Y-1 and 136Y-2. I

4. 120 volt A.C. distribution panels in 480 volt MCCs 136 X-1, 136X-2, 136X-3 and 136Y-2.

ls c. Division 3, consisting of; V #

1. 4160 volt bus.143

2. 480 volt MCC 143-1.

3. 120 volt A.C. distribution panels in 480 volt MCC 143-1.

d. Unit 2 Division 2, consisting of;

1. 4160 volt bus 242Y.

2. 480 volt buses 236X and 236Y.

3. 480 volt MCCM 236X-1, 236X-2, 236X-3, and 236Y-1. I

4. 120 volt A.C. distribution panels in 480 volt MCC's 236X-1, I 236X-2, and 236X-3.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5,and *. !

"When handling irradiated fuel in the secondary containment.

LA SALLE - UNIT 1 3/4 8-12 l l

i l

i ELECTRICAL' POWER SYSTENS

.. D.C. DISTRIBUTION - CPERATING l k* LIMITING CONDITION FOR OPERATION g 3.8.2.3 The following D.C. distribution system electrical divisions shall be !

OPERABLE and energized:

l

_. _ ._. .._ a ... Division 1, consisting of;

- ~ ~ ~ ~ "

1. 125 volt battery 1A.

2. 125 volt full capacity charger.

3. 125 volt distribution panel 111Y.

Division 2, consisting of; b.

1. 125 volt battery 18.

2. 125 volt full capacity charger.

3. 125 volt distribution panel 112Y.

c. Division 3, consisting of;

1. 125 volt battery 1C.

,2. 125 volt full capacity charger.

125 volt distribution panel 113.

3.

~

r (ii. Unit 2 Division 1, consisting o D O - '

i 1. 125 volt battery 2A.

. 2. 125 volt full capacity charger.

> 3. 125 volt distribution panel 211Y.

J' . Unit 2 Division 2,consisi.ingof;

1. 125 volt battery 28.

I

2. 125 volt full capacity charger.

3. 125 volt distribution panel 212Y.

[ APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

l l ACTION:

a. With either Division 1 di. trit _ti;n p;;;l 111Y or Division 2 1 t @ +- N >t'- re c' 11?" inoperable or not energized, restore the

! inoperable division di: trit _ti;; ;;;;l to OPERABLE and energized status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in4 COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With Division 3 m t ' Mt'r ?:n;1 113 inoperable or not energized, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

l With one division battery and/or battery charger inoperable, either:

{c.

I

1. Operation may continue provided the unit tie breakers for the affected division are OPERABLE and aligned to supply power to the affected distribution panel from the associated Unit 2 6

LA SALLE - UNIT 1 3/4 8-14 4

- , , , ,--.--~--,,-,,.-~,.,..-.,.~.a -,.y,- ,.. -

, - , .,_- ,,- e--. ,,_ , - - - - - - . - - . -

. l EECTRICAL POWER SYSTEMS s , LIMITING CONDITION FOR OPERATION (Continued)

I Z ~ " ACTION: (Continued) y 125 voit DC distribution panel *, restore the inoperable battery and/or battery charger to OPERABLE status within .72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or l

2. Restore the inoperable battery and/or battery charger to OPERABLE l status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

O.

- With eith.. L'rit 2 O h hi: I d h tri h ti;;. p...el 211Y :: Unit 2 Division 2 dia. .unim. p.. e 2127 inoperable or not energized, restore the inoperable division dhtriktie.. px'. to OPERABLE and energized status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7

e. With both Unit 2 Division 1 distribution panel 211Y and Unit 2 Division 2 distribution panel 212Y inoperable or not energized, restore at least one of the inoperable distribution panels to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEfLLANCEREQUIREMENTS

4.8.2.3.1 Each of the above required D.C. distribution system electrical .

l divisions shall be determined OPERABLE and energized at least once per 7 days l by verifying correct breaker alignment, indicated power availability from the charger and battery, and voltage on the panel with an overall voltage of greater than or equal to 125 volts.

4.8.2.3.2 Each 125-volt battery and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

l 1. The parameters in Table 4.8.2.3.2-1 meet the Category A limits, i and

2. Total battery terminal voltage is greater than or equal to 128 volts on float charge, 4

l l

" Unit 2 equipment in service to supply Unit 1 shall be demonstrated OPERABLE l

7 per Unit 1 Technical Specifications. ACTIONS "a" and "b" shall be revised and ACTION "c" and this footnote shall be deleted upon issuance of an 9

erating License for Unit 2. ,

LA SALLE - UNIT 1 3/4 8-15

.c . . .

1

._ ^(,1 -

ELECTRICAL POWER SYSTBe5 '

,, SUWEILLANCE REQUIREMENTS (Continued) 4 1 .

b. At least once per 92 days and within 7 days after a battery discharge i with battery voltage below 110 volts, or battery overcharge with

, battery tarsinal voltage above 150 volts, by verifying that:

, 1. The parameters in Table 4.8.2.3.2-1 seet the category 8 limits,

. 2. There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10*ohap,and I i 3. The average electrolyte temperature of at least 10 connected i

, cells is above 60*F.

c. At least once per 18 months by verifying that:

i 1. The cells, call plates and battery racks shE no visual indication of physical damage or abnomal deterioration,

'. L The call-to-cell and terminal connections are clean, tight, free of corrosfor. .and costad with anti-corrosion material,

~

I

~' '

3. The resistance of each cell and terminal connection is less than or equal to 150 x 10.e che/,and l i 4. The battery charger will supply at least 200 amperes for i division 1, 75 asperes for division 2 and 50 amperes for

! division 3 at a minious of 130 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

I

d. At least once per 18 months, during shutdown, by verifying that either:

1. The battery capacity is adequata to supply and maintain in j OPERA 8LE status all of the actual emergency loads for the design cycle when the battery is subjected to a battery service j test, or i

2. The battery capacity is adequate to supp19 a dummy load, which is verified to be greater than the actual emergency load, of the following profile while maintaining the battery terminal voltage greater than or equal to 105 vcits.

a) Division 1, greater than or equal to:

1) 483.4 asceres for tre first 50 secones, l ! ., 2) 251.2 amperes for the nex: 3 .inu as,

' 3) 227.7 as+eres for tne next 15 minur.as, l

151.7 asperes for the next 30 minutes, and l

-'

4) l ; 5) 83.7 amperes for the last 180 minutes.

i i

j 4

. LA SALLE - UNIT 1 3/4 8-16 .

l . . . - ... . . . - .: f .- .. .

-w w- --w -- yr y ,- - - ~ - - -,y 3,y w-_ . , - .,--.------,,,,,,---,,c,.--m--,-,., ,, , .-,,,%,y-r, ,y--w-%- - y- --.-- -

ELECTRICAL POWER SYSTEMS

_ SURVEILLANCE REQUIREMENTS (Continued) b) Division 2, greater than or equal to:

1) 488.5 amperes for the first 50 seconds,

2) 237.6 amperes for the next 14 minutes,

3) 177.6 amperes for the next 15 minutes, and

4) 141.6 amperes for the next 30 minutes, and

5) 54.4 amperes for the last 180 minutes.

c) Division 3, greater than or equal to:

1) 58.4 amperes for the first 60 seconds,

2) 11.1 amperes for the next 239 minutes.

d) Unit 2 Division 1, greater than or equal to:

1) 483.4 amperes for the first 60 seconds,

2) 251.2 amperes for the next 14 minutes, !

3) 227.7 amperes for the next 15 minutes,

4) 151.7 amperes for the next 30 minutes, and

5) 83.7 amperes for the last 180 minutes.

d Ad Unit 2 Division 2, greater than or equal to:

1) 488.5 am' peres for the first 60 seconds,

2) 237.6 amperes for the next 14 minutes,

3) 177.6 amperes for the next 15 minutes, .

141.6 amperes for the next 30 minutes, and i 4)

5) 54.4 amperes for the last 180 minutes.

e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturers rating when subjected to a performance discharge test. Once per 60 month interval, this performance discharge test may be performed in lieu of the battery service test.

f. Annual performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life eFpected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated l

capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

l LA SALLE - UNIT 1 3/4 8-17

f ELECTRICAL POWER SYSTEMS I

D.C. DISTRIBUTION - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, Division 1.endfor Division 2, and Division 3 when the l HPCS system is required to be OPERABLE, and Unit 2 Division 2 when the standby gas treatment system and/or the control room and auxiliary electric equipment room emergency filtration system are required to be OPERABLE, of the D.C.

distribution system shall be OPERABLE and energized with:

a. Division 1, consisting of;

1. 125 volt battery 1A.

2. 125 volt full capacity charger.

3. 125 volt distribution panel 111Y.

b. Division 2, consisting of;

1. 125 volt battery 1B.

2. 125 volt full capacity charger.

. 3. 125 volt distribution panel 112Y.

c. Division 3, consisting of;

1. 125 volt battery IC.

2. 125 volt full capacity charger.

L 3. 125 volt distribution panel 113.

d. Unit 2 Division 2, consisting of;

1. 125 volt battery 28.

2. 125 volt full capacity charger.

3. 125 volt distribution panel 212Y.

t APPLICABILITY: OPERATIONAL CONDITIONS 4, 5;and *.

ACTION:

l a. With both Division 1 distribution panel 111Y and Division 2 distribution panel 112Y of the above required D.C. distribution system inoperable or not energized, suspend CORE ALTERATIONS, handling l of irradiated fuel cask in the secondary containment and operations with a potential for draining the reactor vessel.

! b. With Division 3 distribution panel 113 of the above required D.C.

distribution system inoperable or not energized, declare the HPCS system inoperable and take the ACTION required by Specifications 3.5.2 and 3.5.3.

1 "When handling irradiated fuel in the secondary containment.

l LA SALLE - UNIT 1 3/4 8-19 l

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

o ..

. ,e .

. =

.. . . .'.i_ = . . .

l

. s l

. l

. i '

~

ELECTRICAL POWER SYSTEMS ~

3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. CIRCUTT5 INSIDE PRIMARY CDNTAINMENT _

LIMITING CONDITION FOR OPERATION 3.8.3.1 At least.the following A.C. circuits inside primary containment shall be de-energized *:

a. Installed welding grid systems IA and 18, and

b. All drywell lighting circuits.

c. All drywall koi 43 M trAusa cirw M s .

APPLICA8ILITY: OPERATIONAL ColeITIONS b-2;and-3.--- J ACTION:

itith any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel (s) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

l .

C y ,

SURVEILLANCE REQUIREMENTS

4.8.3.1 Each of tne above required A.C. circuits shall be determined to be I

de-energized at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ** by verifying that the associated circuit breakers are in the tripped condition.

l 1

. "Except during entry into the d nwell.

le .

l

- Except at least once ?ar 31 days if Icekif. rested or etherwise f ecurse

', the tripped coccition. -

(.s

(:

4 l

LA SALLE - UNIT 1 3/4 S-21

I l

. g .

$ b -

. TABLE 3.8.3.2-1 1

PRIMARY '.0NTAINMENT PENETRATION CONDUCTOR

, OVERCURRENT PROTECTIVE DEVICES ~

~

TRIP RESPONSE SYSTEM /

OEVICE NUMBER SETPOINT TIME AND LOCATIO,h_ (Amperes) (Milliseconds / Cycles)(*) COMPONENT POWERED

a. 6.9 KV Circuit Breakers

1. Swgr. 151 (Compt. 4) 840 CC) 83.3/5 RR Pump 1A

2. Swgr. 152 (Compt. 4) 840 CC) 83.3/5 RR Pump 18

3. Swgr. 151-1 (8kr. 2A) 720(b) g3,375 gg pu,,_yg _.

Iow speed .

4. Swgr.152-1 (Bkr. 28) 720(b) 83.3/5 RR Pump 18, low speed

b. 480 VAC Cfreuit Breakers *

1. Swgr.136Y (Compt. 160 CC) 50/3 VP/Prf. Cont.

403C) Vent Supply Fan IB

'- 2. Swgr. 135Y (Compt.

, 160(C) 50/3 VP/Pri. Cont.

203A) Vent Supply Fan IA -

l l ; c. 480 VAC (Molded Case) Circuit Breakers .e

! i Type X-M Cat # NZ Mi-160/ZM6 w P f

/

l

! i a) MCC 134Y-3 (Compt. A10) 36 N.A. ItC/Hofst Motor i b) MCC 134X-2 25 N.A. HC/EQpt. Handling l (Compt. A4) Platform

! c) MCC 134X-2 60 N.A. HC/ Crane &

! Compt. 03) Trolley Power 1 I J. Type X-M Cat # NZ Mi-160/ZM6C .

l i i'j a) MCC 136Y-2 174 N.A. RR/MOV 1833-F067B

.(Compt. C4) l b) MCC 136Y-2 72 N.A. RMCV 1833-F0238

j (Compt. A3) c) MCC 134X-1 10 N.A. NS/MOV1 1321-FC01

?

(Compt. 83) -

i.

[d .

d) MCC 134X-1 (Compt. 84) 10, N.A. NS/MOV 1821-F002

.i

LA SALLE - UNIT 1 3/4 8-24

m

3 s . .

fl.ECTRICAL POWER SYSTEMS 9570R OPERATED VALVES THERMAL OVERLOAD PROTECTION LIMITINGCONDITIONFOROPERAi!0N

~~ : . . ' . ~~:K .

3.8.3.3 The thermal overload protection of each valve shown in Table 3.8.4.2-1 shall be bypassed continuously or under accident conditions, as applicable, by an CPERABLE bypass device integral with the actor startar.

APPLICA8ILITY: Whenever the motor operated valve is required to be OPERABLE.

ACTION:

e tHth the thermal overload protection for one or more of the above required valves not bypassed continuously or under accident conditions, as applicable, by a1 OPERA 8LE integral bypass device, take administrative action to continuously bypass the thermal everload within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the affected valve (s) ~

inoperable and apply the appropriate ACTION statement (s) for the affected systan(s).

a

k. Thr. provislaus f c.i.ia4b 3.0 4 4,e. Mo+

A f fL4 h /8-.

SURVEILLANCE REQUIREMENTJ -

('

4.8.3.3.1 The thermal overload protection for the above required valves shall be verified to be bypassed continuously or under accident conditions, as applicable,

' by an OPERABLE integral bypass device by the performance of a CHANNEL FUNCTIONAL TEST of the bypass circuitry for those thermal overloads which are normally in force during plant operation and bypassed under accident conditions and by verifying that the thermal overload protection is bypassed for those thermal overloads which are continuously bypassed and temporarily placed in force only ,

, when the valve motors are undergoing periodic or maintenance testing:

a. At least once per 18 months, and

b. Following maintenance on the actor startar.

i ! 4.8.3.3.2 The thermal overload protection for the above required valves which

{ are continuously bypassed shall be verified to be bypassed following testing j during which the thermal overload protection was temporarily placed in force.

l i -

T f

4 l .

LA SALLE - UNIT 1 . 3/4 8-26 ee

.y- , , - - - , - - ,y-y,-

-. ,.-,,- ,,w-- - . - - _ - - - - . - - - , - , . - , . - . - . - , - - - , - . , - . - - , , - - . , . - - - - - - - - - - - - - , - - - - - -

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

w. . a b'

.M .. .

t ,1 .

I TABLE 3.8.3.3-1 MITOR OPERATED VALVES THERMAL OVERLOAD PROTECTION .

i 8YPASS DEVICE SYSTEM (S)

VALVE NUMBER (Continuous)(Accident Conditions) AFFECTED

a. IVGG01 Accident Conditions 58GTS IVG003 - Accident Conditions

, 2VG001 Accident Conditions .

2VG003 Accident Conditions

b. IVP113A Accident Conditions Primary containment IVP1138 Accident Conditions chilled water coolers IVP114A Accident conditions IVP1148 Accident Conditions -

IVP053A Accidant Conditions IVP0538 Acci d nt Conditions -

IVP063A Accident Conditions IVP0638 Accident Conditions

c. IVQ040 Accident Conditions Primary containment IVQO36 Accident Conditions vent and purge system y' .IVQ026 Accident conditions IVQ029 Accident Conditions IVQO38 Accident Conditions i

_ "70^t ^cc M -t C x t.'^.' ... l IVQ031 Accident Conditions '

IVQ032 Accident Conditions

, IVQ034 Accident Conditions IVQ035 Accident conditions IVQ027 Accident Conditions .

t l IVQ042 Accident Conditions l j IVQ043 Accident Conditions i , IVQ047 Accident Conditions i

' IVQ048 Accident Conditions IVQ050 Accident Conditions

VQ051

. Accident Conditions IVQ068 Accident Conditions IVQO30 Accident Conditions -

IVQ037 Accident Conditions

d. 1WR179 Accident Conditions R8CCW system

{ IWR180 Accident Conditions 1 IWR040 Accident Conditions l 1WR029 Accident Conditions

' e. 1821 - F057A Accident Canditions Main steam syst n 1821 - FC67B Acefrdent Conditiens 1B21 - FC57C Accicant Concitions

'd 1821 - F067D Accident Conditions j 1821 - F019 Accident Conditions 1821 - F016 Accident Conditions I

LA SALLE - UNIT 1 3/4 8-27 l -

- _ _ - . _ __ J._..___ ._ ._ _.._._._ __ __._...._.. . . _ _ _ . _ _ . _ , _ _ . _ __. . _ _ _ _ _ _ _ _ _

. . ..~.: ..

~.

S ELECTRICAL POWER SYSTDes

()

REACTOR PROTECTION SYSTEM ELECTRICAL F7 ER NDNITORING l LIMITING CONDITION POR OPERATION

3 l 3.8.3.4 Two RPS efectric power monitoring assembifes for each inservice RPS I Mi set or alternate power supply shall be OPERA 8LE.

APPLICA8ILITY: At all tiens.

ACTION:

a. With one RP5 electric power monitoring assembly for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power monitoring assembly to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS MG set or alternate power supply fros

. . service. ,

l b. With both RPS electric power monitoring assemb11es for an inservice RPS MG set or alternate power supply inoperable, hstore at least one electric power monitoring assembly to OPERA 8LE status within 30 minutes or remove the associated RPS MG set or alternate power suppiy from service. -

SURVEILLANCE REQUIREMENTS s/

i

4.8.3.4 The above specified RPS electric power monitoring assemblies shall be

! determined OP'RA8LE: -

' B

a. ." 'M :.-- -- C _.r. 4y performance of a CHANNEL FUNCTIONAL i TEST lowb is ia Col.D

! -ke.)A M decf4ime Anes,uplest es.

per g! forma in 4ks ysvieasS#t/TDow#

(, n eas. </ec A pan

, b. At feast once per 13 months by demonstrating the OPERABILITY of overtvoltage, undertvoltage, and under! frequency protective I instrumentation by performance of a CHANNEL CALIBRATION including sieuiated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following i

i setpoints.

j 1. Ovedvoltagei132VAC, i ! -

2. Unde (voltage 1108VAC,

3. Under{ frequency 157Hz.

4

,J LA SALLE - UNIT I 3/4 8-31

~ l REFUELING OPERATIONS

i SURVEILLANCE REQUIREMENTS (Continued) i i

)

b. Performance of a CHANNEL FUNCTIONAL TEST: ;

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of CORE ALTERATIONS, and

2. At least once per 7 days.

c. Verifying that the channel count rate is at least 0.7 cpsFt l

1. Prior to control rod withdrawal,

2. Prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS, and

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

d. Verifying that the RPS circuitry " shorting links" have been removed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during:

1. The time any control rod is withdrawn," or

2. Shutdown margin demonstrations.

i l

l "Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

$ frottded signal-to-nstse t-ind is 2 2. Other+se, 3 cp. l LA SAL.E - UNIT 1 3/4 9-4 Amendment No. 2

, . ,4 . 5

. . i

. . t Tg MPUELING OPERATIONS .

, 3/4.9.H RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION NIGH WATER LEVEL -

LIMITING CONDITION FOR OPERATION .

I 3.9.11.1 At least one shutdo*E cooling mode loop of the residual heat removal

(RHR) system sha11 he OPERABLE and in operation

with at least:

! a. One OPERA 8LE' RHR pimp, and

. b. One OPERA 8LE RHR heat exchanger.

APPLICA8ILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor i vessel and the water level is greater than or equal to 22 feet above the top -

i of the reactor pressure vessel. flange.

t

ACTION:

l

a. With no RHR shutdown cooling mode locp OPERA 8LE, within.omr hour and at L

l .

i

' least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at

~

least one alternate method capable of decay heat removal. Otherwise,

suspend all operations involving an increase in the reactor decay heat j load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

I b.

l With no RHR shutdown cooling mode loop in operation, within hour [

establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour. "

l

! SURVEILLAN'E C REQUIREMENTS t

i 4.9.11.1 At Teast one shutdown cooling mode loop of the residual heat removal j systas or alternate method shall be verified to be in operation and circulating i reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i

'l 1 .

I I

"The snute:wn :: 'ing can be removed fr: :.senicr. f:r r: to 2 .icurs :sr 8-nuar period.

,j The normal or emergency power source say be inoperable.

I

! LA SALLE - UNIT 1 3/4 9-16 l

, e_ : -- - -- - -

1 . . -. - ,

...s.. , . .. * .

I -

)

REFUELING OPERATTORS -

I

LOW WATER LEVEL - .

. 1

, LIMITING CONDITION FOR OPERATION -

. l l

3. 9." u. 2 Two shutdown egoling mode loops of the residual heat removal (RHR)

I system shall be OPERA 8LE" and at least one loop shall be in operation," with

each loop consisting of at least:

a. One OPERA 8LE RHR pump, and

b. One OPERA 8LE RHR heat exchanger.

h APPLICA8ILITY: OPERATIONAL ColeITION 5, when irradiated fuel is in the reactor l,

vessel and the water level is less than 22 feet above the top of the reactor ,

pressure vessel flange.----

l! ACTION: * -

a. With less than the above, required shutdown cooling mode loops of the RHR system OPERA 8LE, within tas hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, l demonstrate the operability of at least one alternate method capable of

-- - decay heat removal for each inoperable RHR shutdown cooling mode loop.

b. With no RHR shutdcwn cooling mode loop in operation, within .p.L ne hour g i establish reactor coolant circulation by an alternate method and monitor 1 reactor coolant temperature at least once per hour.

Ii g

I l l

,, SURVEILLANCE REQUIREMENTS ii i,

! 4.9. u.2 At least one shutdown cooling mode loop of the residual heat removal

l systas or alternate method shall be verified to be in operation and circulating

{ reactor coolant at least once per 12 hcurs. -

1 .

I l

l 4 .

I "The sautccwn cooling c:.: D may be ruev=c f.m :::2-acicr. "cr 2:: 1 2 hcurs -

^ , per 3-hcur period.

l

/ #The normal or emergency power source may be ineperable for each loop.

l l

~

LA SALLE - UNIT 1 3/4 9-17 e em og e o e ' .e w emee- eesme o *

y -- ---- * - - - -w n---w.-----,,--e--rv -- -

t -_..

s- - _. e i

S ..

TA8LE 4.U.1-1 RADICACTTVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

~

g~ ~. ~. ~'

. ' - ~ '

Minim,um Type of Lower Limit Liquid Release Sampling Analysis Activity of Detection Type Frequency Frequency Analysis (LLD)

(pci/ml),

A. Satch Waste P P Principal Gamma 5x10-7 Release Each Batch Each Batch Esitters#

d Tanks I-131 -

1x10'6 .

P M Dissolved and 1x10

-5 One Batch /M Entrained Gaies (Gamma esitters)

-5 P M H-3 1x10 .

Each 8atch Composite h Gross Alpha 1x10'7

( [

Sr-89, Sr-90 -8 P Q b 5x10 ,,

Each Batch Composite - -

Fe-55 -8 3

1x10

8. Continuous W Principal Gassa -7 5x10 Composite c e

l Releases

Continuous Esitters f I-131 1x10

' -5 M M Dissolved and 1x10

, Grab Sample Entrained Gases j (Ganna Esitters)

I M H-3 1x10'5 l !! Continuous C

Composite" Gross Alpha- 1x10'I b

m '

o Q .Sr-89, ir-90 l 5x10' !

" C -6

. Continuous Composite C Fe-55 1x10 i

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

- -- - - , - , -w .w, , ,m--,,rn--err,--,,,-,,,-,--,-,-r- ,.e--------,-w---,------n----,---,e.--- - - , - - - - - - - - - - - - -- --

~ ~ ~ ~ ~ ~

.~ . - . . . . .. ,

. s

( ,

). RADI0 ACTIVE EFFLUENTS 3/4.11.2 GA$EOUS EFFLUENTS .i i 4 005E RATE -

LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 5.1.1-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 area /yr to the total body and less than or equal to 2000 stes/yr to the skin, and

. b. For all radiciodines and for all radioactive saterials in particulate form and radionuclides (other than noble gases) with half lives greater than 8 days: Less than or equal to 1500 aree(yr to any organ vsa the mkaI4tson pathway. -

APPLICA8ILITY: At all times. -

l

_ ACTION:

tiith the dose rate (s) exceeding the above limits, immediately decrease the

] release rata to within the above Ifnit(s).

! SURVEILLANCE REQUIREMENTS i

4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM.

. 4.11.2.1.2 The dose rate due to radioactive materials, ether than noble i gases, in gaseous effluents shall be determined to be within the above Ifmits in accordance with the methods and procedures of the 00CM by obtaining represen-tative samples and performing analyses in accordance with sampling and analysis program specified in Table 4.11.2-1.

p m

l J '

'I LA SALLE - UNIT 1 3/4 11-9

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

.w j ' ~

N ..

./ - -

TABLE 4.u.2-1 (Continued)

TABLE NOTATION .

-~ -- -~ -

b. Analyses shall also be performed following shu'Jown, startup, or a THERMAL POWER change exceeding 15.p y estff of the RATED THERMAL POWER f i within a pne' hour period. .

7* I

' .i.

c. Whenever there is flow through the $8GTS.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing /or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startuo or THERMAL POWER change exceeding 15 percent cf RATED THERMAL POWER in hour and l analyses completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be .

increased by a factor of 10. ,

~

e. Tritits grab samples shall be taken at least once per 7 days from the plant vent to detamine tritius releases in the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool. ~

f. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation sede in accordance with Specifications 3. U.2.1, 3. u.2.2 j and 3. n.2.3.

i g. The principal gamma esitters for which the LLD specification applies -

! wdvda.+' r t.i.J, n 2: following radionuclides: Kr-87, Kr-88, Xe-133, l l Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59,

! Co-58, co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for

! particulate emissions. This list does not mean that only these i nuclides are to be detected and reported. Other peaks which are measurable and identifiable, at the 95% confidence level, together

with the above nuclides, shall also be identified and reported.

l ll; ,

This requirement does not apply if (1) analysis shows lt that the DOSE EQUIVALENT I-131 concentration in the primary coolant l'l l ,

has not increased more than a factor of 3; and (2) the nocle gas monitor shows that effluent activity has not increased more than a I

i factor of 3.

lj -

,. t

~

i i

LA SALLE - UNIT 1 3/4 u-12

,,.,e_.w.,---n. , _ , , . . _ , _ , , , , , . , , , . . , , . , _ _ _ _ , , . _ , _ , , , . . _ _ _ , _ _ _ . _ . , _ _ , , _ . _ _ _ . , , _ _ _ - , . _ . _ , - , . . , , . _ _ _ , _ , - _ _ ,.

.. . ... . t. . . .. . .. . . . . . . . . -. .

t..

J' ' RADI0 ACTIVE EFFLUENTS DOSE - N0BLE GASES .[ -

~

LIMITING CONDITION FOR OPERATION l , 3.11.2.2 The air dose due to noble gases released in l each reactor unit, from the site (see Figure 5.1.1-1) gaseous shall beeffluents, Itaited tofrom the i following: -

s. During any calendar quarter: Less than or equal to 5 arad for gamma l radiation and less than or equal to 10 arad 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 arad for beta radiation. ~

1 .

APPLICA8ILITY: At all times. ~~

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in Ifeu of any other l

report required by Specification 6.6. A or 6.6.8, prepare and submit l to the Commission within 30 days, pursuant to Specification 6.6.C, a i

Special Report which identifies the cause(s) for exceeding the l limit (s) and defines the corrective actions to be taken to reduce ;

i the releases M rdhrth; x:;h ;nn da ~~rr :ffh;;.G d.rks

'2: -hir Of 2: c T rt ce M M e-

r-': = r t' -ta" tM ~+ = " de='-

rMr;xc.t 0.cx ceM-6 e"-"ht'g":tMdre h d th'- 10 i.4 .".. ;-- : rd h' h . x d 20 r--d 4- h* ;d ht h ..

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

11

! 3 SURVEILLANCE REQUIREMENTS 1

i l I 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

l with the 00CM at least pnce per 31 days.

! gyo 'ntE PadPMED COR9ecnE ACn0MI TT GE TAKE l

TD 4150 Re TIMT ,50G5eGvfaT* (ELFAScT wm 60-J lA) GM 9La 9^)LG wsTH TH5 A60/E LsM (73.

I t

LA SALLE - UNIT 1 -

3/4 11-13

.___.v.. _ _ _ _ _ _ . _ _ _ _ _ _ _

~

- .. a ..-

~ . .

'N' . .

- RADI0 ACTIVE EFFLUENTS IN15E g HER -THAN RADIGICOINES. RADI0 ACTIVE MATERIALS IN PARTICULATE FORM. AND RADIONUCLIDES NOBLE GA5E5 -

~

, LIMITING CCNDITION FOR CPERATION 3.11.2.3 The dose to an individual from radiofodines and radioactive materials in particulata fom, and radienuclides, other than noble gases, with half-lives greater than 8 day.1 in gaseous affluents released, from each reactor unit, from the site (see Figure 5.1.1-1) shall be Ifeited to the following:

a. During any calendar quarter: Lessthanorequalto7.5are(toany organ, and )

b. s -

During any calendar year: Less than or equal to 15 ares to any g organ. ..

l APPLICA8ILITY: At all times.

ACTION: "

.. a. With the calculated dose from the release of radiofodines, radioactive materials in particulata form, or radionuclides (other than noble gases) with half Ifves 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 or 6.6.5, prepare and submit to the Commission within 30 days, pursuant to Specification 6.6.C. a m Special Report which identifies the cause(s) for exceeding the Ifmit M have. bees M and defines the corrective actions * " "'- '- --d"-= the rela = = i I

g " g A misant 4 r:di:i: din. ;ad r;;ie. J .. -.6.,;.;; in ;;r.ini;t. Tec;, =d-Aasd A F' g - d'==Iids, 'l:r C= n;;te. ;nn, 40 he!f !!;n si.O-- th C fagun.;:i =t: duri;; Ce . - _;num T G . c.cc n t-l g , d J m(*liw s 4' ra w d-- xcr ud d;ri g 0; e ;;eg .;. . . ., Cr;; nind:r ;;r.:r; l

*t 7 -'I;ti} de.- vi uv== 6w n ="w i=i.;dv.I-from Q w 4e Assure. * , ,

..:n re ...... i s wi w i n u = == w -ni ein.n.

w .n

& shs*}&g j I

435,3 w;lt be. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

lgg ;Ame, iw.3L + 0'"

l 4;y4s SURVEILLANCE REQUIREMENTS l .

I 1

i i 4.11.2.3 Dese Calculations Cumulative dose contributic7s for the current j

calendar cua-ter anc cur ent cale wit'. the C':C:t at least once perJ4f."nr 5. year snall ne cnarm'ned in accor ance J

LA SALLE - UNIT 1 -

3/4 11-14

- . - - - - - , , - - , . . w.-p,-.,--.4-. , - ,,--_,w g,-~,,---,--,,..,_-,,e.,w- - , - . , , - - , , . _

.u . . . .., 1 .1.': :.:.....:.L:ek::%. . ~~ . '= -

. i.

'N -

i

. _ , RADI0 ACTIVE EFFLUENTS VENTING OR PURGING ,

LIMITING CONDITION FOR OPERATION l l

3.11.2.8 VENTING or PURGING of the containment drywell shall be through the ;

. Primary Containment Vent and Purge Systes or the Standby Gas Treatment System. '

APPLICA8II.ITY: Whenever the drywell is vented or purged.

ACTION:

a. With the requirements of the above specification not satisfied, suspend all VENTING and PURGING of the drywell. ,

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

SURVEILLANCE REQUIREMENTS -

5

4. .8 The containment drywe shall be determined to be aligned for VENTING or PU % through the Primary C tainment Vent and Purge System or the Standby Gas Trea t System within 4 hou prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months dur VENTING or PURGING f the drywell. -

4.11.2.8.1 The containment drywell shall be determined to be aligned for VENTING or PURGING through the Primary Containment Vent and Purge System or the Standby Gas Treatment System within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months daring VENTING or PURGING of the drywell.

t I .4.11.2.8.2 Prior to use of the Purge System through the Standby Gas Treatment System in OPERATIONAL CONDITION 1, 2 or 3 assure that:

l a. Both Standby Gas Treatment System trains are OPERABLE, and l .t b. Only one of the Standby Gas Treatment System trains is used for i PURGING.

d

.LA SALLE - UNIT 1 3/4 11-19 s - .,n., ,v- _,, , . , - , .

e--._. .-m. . , _ . _ . _ _ . . _ , _ _ , , , , . _ _ _ . , . . , , _ , _ , . , _ , , , , . . _

~'

~ . . . --......a..z.. .. .: ._ . . . . . . . . . . .._:.._...

(. l s .

3/4.12 RA0!0 LOGICAL ENVIRONMENTAL M)MITORING J

, 3/4.12.1 NONITORING PROGRAM . .

Leiund CONDITION FOR OPERATION -

3.12.1 The radiological environmental s' onitoring program shall be conducted as specified in Table 3.12.1-1.

. APPLICA8ILITY: At all times.

ACTION:

a. With the radiological environmental sonitoring program not being 4,g, g e . 1 conducted as specified in Table 3.12.1-1, in Ifeu of any othery '

report required by Specification &&rl, prepare and subett to the l#

b ' 8 j

Commission, in the Annual Radiological Operating Report, a descrip-tion of the reasons for not conduct,ing the program as required and .

the plans for preventing a recurrence.

b. Wii.hthelevelofradioactivityinanenvironsenialsamplingmedium j exceeding the reporting levels in Table 3.12.1-2 when averaged over

any calendar quarter, in Ifeu of any other report required by apec1rication'.5 Set; prepare and subeit to the Comeission wtthin 30 ~

s' b'.b

(, 6. 6 D days from the end of the affected calendar quarter a Report pursuant

i. to Specification 6.9.1.13. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) , concentration (2) ,***3- g,g .

l

[ 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 subettted 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 1 and 3.11.2.3. This report is not required if the measured level of i radioactivity was not the result of plant effluents; however, in l such an event, the condition shall be reported and described in the j , Annual Radiological Environmental Operating Report.

} c. With afik or fresh leafy vegetable samples unavailable from one or j more of the sample locations required by Table 3.12.1-1, in fiou of (,. 6. A e i ! any other report required by Specification.,5,9'ti, prepare and submit g,a j 1 to the Commission within 30 days, pursuant to Specification 6 vert -

l Special Report which identifies the cause of the unavailability oh, a g*g j samples and identifies locations for obtaining replacement samples.

. The locations from which samples were unavailable may then be deleted fecm those eeuf red by 1'19e 3.12.1-1, previta: t.5e !ccaticns frem wnica tne replacement s u 'es were cbtainea are ac:ec to tne environmental monitoring ;rc.; ram as replace ect icesticas.

w..

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

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

, . _ , . . _ , , _ , , . . . - - - - - + -

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

,! . i -

( . .'a .

)

TABLE 3.12.1-1 , . .

E RADIOLOGICAL ENVIRONENTAL MDNITORING panc.nau i: ! ,

,! j j i Y

P m

Number of samples .

e Exposure Pathway and Sampitag and i Type and Frequency g and/or Sample Sample Locations Y Collection Frequency of Analysis ; -

" 1. AIRBORNE 6 locn4;ons l'-

i R.iellulodinc and -{'rth:- 13) Cantinuous operatten of Particulates Radialodine canister. .l l t !

sampler with sample col- Analyze at least once : .

1ectlen as required by per 7 days for I-131.

dust loading but at least once per 7 days. Particulate sampler. ,,

Analyze for gross beta 11 radioactivity > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months !.

fellowing filter change. 8 Perform gamma isoteptc analysis on each sample l. ,

g when gross beta activity is > 10 times the yearly :,

l _

mean of control samples.

4 y

Perfore gamma isotopic i

i sa analysis on composite l * . !

! (by location) sample l at least once per j

. 3 0 1oc44;o.2 r 92 days.

l ,.

j 2. DIRfCT RADIATION (*.:::th:: '- O'.) At least once per 31 days. Gamme dose. At least l

> 2 desleeters or > or' once per 31 days. .

i Tnstrument for con _ 1 or :

i tinuously esasuring At least once per 92 days. Gamma dose. At least i and recording dose (Reed-out frequencies are once per 92 days.

j rate at each determined by type of desfa- '

location. eters selected.) i l

I -

-ib lwC& liens 4,6 y$.ea .6 17.6 s $yus u .,4 i.I,16 l,i iN'~~ '

l < i g

l-l

.t

. ENDee g e

l \ n

C ,., I

~ s .

, 1 r TABLE 3.12.1-1 (Continued) ,

E g RADIOLOGICAL ENVIR00 MENTAL MONITORING PROGRAM r-e- ,

en ,

. Minfaus ! I j

g Exposure Pathuay -

Number of Samples ' '

and Sampling and Type and Frequency N and/or Sample Sample locations

Collection Frequency of Analysis i'

?

3. WAIER80RNE ' .

j .

, . +

i

a. Surface 2 locations . Composite sample collected Gamma isotopic analysis l over a period of < 31 days. of each composite sample. ' *

Tritium analysis ~of com- I

. . ,osite .aspie ai. ieast once per 92 days. !

R

! b. Ground 5 locations At least once per 92 days. Gamma is.otopic and .

, $.1 trittua analyses of i

each sample.

i

c.

! Sedleent from 1 location At least once per 184 days. Gamma isotopic analysts -

! Shoreline of each sample.

8 1

I

S c!: !:::th;_ :n 9- m *2: f tp= h L gen. .

t l i

i

I e

i

+

. I

! l I i : .

l lr

. i

. . .... . . .~. . ... .

k m

\g l 3/4.0 APPLICA8ILITY sASEs __

The speciffcations of this section provide the general requirements appifcable to each of the Lfatting Conditions for Operation and Survef11ance

Requirements within Section 3/?.

l p; sect A -**

3.0.1 This specification states the applicability of each specification in terms of defined OPHATIONAL CONDITION or other specified applicaoflity condition and is provided to delfneate speciffcally when saca speciffcation is

, applicante.

l

, 3.0.2 This speciffcation defines those conditions necessary to constitute compliance with the tems =f an f adividual Lf atting Condf tf on for Operation and associated ACTION requirement. .

'l 3.0.3 This specification delineates the asasures to be taken for circiastances not directly provfded for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 3.7.2 requires two control room and auxiliary electric equipment room emergency ffitration trains to be OPERA 8LE and provides explicit ACTION requirements if one train is inoperable. Under the requirements oftSpecifica-w tion 3.0.3, if both of the required trains are fnopersale, within one hour l esasumsr sust be inttf ated to place the unit in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least NOT SMUTD0tdM within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0km within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . As a further example, Speciffcation 3.6.8.1 requires two primary containment hydrogen reconciner systans to be OPDA8LE and provides explicit ACTIch requirement 2 if one recomeiner systan is fnoperable. Under the requirensetsgof Spectff cat 1on 3.O.3, I if both of the required systnes are inoperable, within .end' hour sessures must l be initiated to place the unit in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and

, A $ in at least NOT SHUTDonM within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l 3.0.4 This specification provides that entry into an OPERATIONAL CONDITION l

sust te sede with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as speciff ed in the Limiting 4 Conditions for Operation being set without regard for allowable deviations and out of service provisions contained in the ACTION statements.

i i The fntent of this provis1on is to ensure that unit operation is not

! fnitf ated with either required equipsont or systass inoperable or other limits ,

! being exceeded. '

t Exceptions to this provision have been provided for a Ifmited numeer of l; specifications when startup with fnoperable ecufpment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specif f cations.

l N,

. s f

LA SALLE - UNIT 1 8 3/4 0-1 I

IvseeTS rce PAse B 3/4 o-l

-Diset.T 8 In the event of a disagreement between the requirements stated in these Technical Specifications and those stated in an applicable Regulation or Act, the requirements stated in the applicable Regulation or Act, shall take precedence and shall be met.

D ise e.7 8

. It is acceptable to initiate and complete a reduction in OPERATIONAL CONDITIONS in a shorter time interval than required in the ACTION statement and to add the unused portion -

of this allowable out-of-service time to that provided for operation in sub-sequent lower OPERATIONAL CONDITION (S). Stated allowable out-of-service times i

are applicatie regardless of the OPERATIONAL CONDITION (S) in which the inopera-bility is discovered but the times provided for achieving a CONDITION l reduction are not applicable if the inop vability is discoverad in a CONDI-l TION Tower than the applicable CONDITION.

(

EACTIVITY CONTROL SYSTEMS I 4

i 3/4.1 RDCTIVITY CONTROL SYSTEMS J 7 BASES .

3/4.1.1 $HUTDOW MRGIN i

A sufficient SHUTDOW MAGIN ensures that(1) the reactor can be made I suberitical from all operatir.g conditions,(2) the reactivity transients 8 associated with postulated accident conditions are controllable within acceptable 1 faits, and(1) the reactor will be maintained sufficiently I subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel decletion and poison burnup, the demonstration of SHUTDOW MARGIN will be performed in the cold, xenon-free condittori and shall show the core to be subcritical by at least R + 0.385 delta K or R + 0.285 delta K, as appropriate.

The value of R in units of % delta K is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R eust be positive or zero and sust be determined for each fuel loading cycle.

s Two different values are supplied fr the Limiting Condition for Operation

._.- to provide for the different methods of demonstration of the SHUTDOW MARGIN.

The highest worth rod say be determined analytically or by test. The SHUTDOW MAGIN is demonstrated by an insequence control rod withdrawal at the beginning-of-life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicatas that the recuired margin could '

l be reduced as a function of exposure. Observation of subcriticality in this condition assures suberiticality with the most reactive control rod fully

( withdrawn.

lI This reactivity characteristic has been a basic assumption in the

{' analysis of plant performance and can be best demonstrated at the time of fuel i, loading, but the margin sust also be determined anytime a control rod is l} incapable of insertion.

\

i 3/a.1.2 RD CTIVITY ANOMALIES 1 Since the SHUTDOW MARGIN requirement for the reactor is small, a careful l i check on actual conditions to the predicted conditions is necessary, and the i i changes in reactivity can be inferred from these comparisons of rod patterns.

I Since the comparisons are easily done, frequent checks are not an imposition i j on nonsal operations. A 1% change is larger than is expected for normal

operation so a change of this magnitude should be thoroughly evaluated. A i : change as large as 15 would not exceed the design conditions of the reactor i, and is on the safe side of the postulated transients.

l --

l I

i LA SALLE - UNIT 1 8 3/A 1-1

=+ , . - _ . . - - . - . . , , , , . , , . - . , _ . _ . , ,,,._.,y, , , . _ . , , _

,_,.,y__, ,,,_-.,,,, , - _ , , - , . _ . _ - - , _ , , . , . _ - - ,-

I (

i s REACTIVITY CONTROL SYST995

[~ .

i i l

BASES t

. 3/4.1.3 CONTROL RODS

- , The specificationsof this section en'ure s that (1) the sinimus SHUTDOWN I (2) the control rod insertion times are consistant with l

MARGIN those usedisinmaintained.ident the acc analysis, and (3)44sft the potential effects of the red dron accident,a The ACTION statements permit variations from the basic ,

1 resents but at the same time impose more restrictive criteria for continued g ,

g.,M ,.g operation. A limitation on inoperable rods is set such that the resultant effect on total rnd worth and scras shape will be kept to a sinisus. The requirements for the various scras time seasurements ensure that any indication j of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a ceneric probles, therefore with a control rod issovable because of excessive friction or mechanical intarfarence, operation of the reactor is limited to a time period which is i reasonable to datamine the cause of the inoperability and at the same time prevent operation with a large number of inoperable centrol rods.

I i Control rods that are inoperable for other reasons are persitted to be

, taken out of service provided that those in the nonfully-insertad position are

,j consistant with the SHUTDOWN MARGIN requirements.

The maber of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable 4 rods could be indicative of a generic probles and the reactor sust be shutdown j for investigation and resolution of the probles. _

The control rod systas is dest to brin the reactor suberitical at a M l rats fast enough to prevent the MCP ros bocas ng less than h06*during tne clAdd Wt

' '! limiting power transient analyzed in Section 15.0 of the FSAR. This analysis ae shows that the negative reactivity ratas resulting from the scras with the r

p"* .'

-l average response of all the drives as given in the specifications, provide the ~

ll required protection and MCPR remains greater than-he6.nte ocm me ai l ; scras times longer then those specified should be viewed as an indication of a

' systemic probles with the rod drives and therefore the surveillance interval i is reduced in order to prevent operation of the reactor for long perious of i time with a potentially serious probles.

t i The scras discharge volume is required to be OPERABLE so that it will be i available when needed to accept discharge water from the control rods during a reactor scras and will isolata the reactor coolant system from the environment

. when required.

l Control rods with inoperable accumulators are declared inoperable and

' i Specification 3.1.3.1 then applies. This prevents a pattern of inoperabia

accumulators that would result in less reactivity insertion on a scras than

., has been analyzed even though control rods with inoperable ace:mulators say

, still be inserted with normal drive water pressure. Operability of the accumu- '

' . later ensures that there is a means available to insert the control rods even l under the most unfavorable depressuri:ation of the reactors.

LA SALLE - UNIT 1 8 3/4 1-2

p.

_ REACTTVITY CONTROL SYSTDS SASES

---'.-- .7 .. . .

3/4.1.6 ECONOMIC G DERATION CONTROL SYSTB Operation with the econcate generation control (EGC) system, autematic flow control, is limited to the range of 65% to 100% of cated core flow. In this flow range and with THERMAL POWER > 20% of RATED THERMAL PCWER, the reactor could safely tolerate a rate of change of load of 8 MWe/pc (referer:ce FSAR Section P1h4).

62.4 Limits within the EGC and the flow control systas prevent rates of change greater than approximately 4 MWe/sec. When EGO is ir. oceration, this fact l will be indicated on the main control room console.

s t

A 1

.s LA SALLE - UNIT 1 8 3/4 1-5

l 1

( ,

1

,m 3/4.2 P0k B OISTRIBUTION LIMITS BASES l

The specifications of this section assure that the peak cladding temperature .

followin the 2200gF the Ifnitpostulated specified in design 10 CFR basis loss of-coolant accicent will not axceed 50.46.

3/4.2.1 AVERAGE PUWAR t.INEAR HEAT GENERATICN RATE This specification assures that the peak cladding temperature 'following '

the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rata of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assemoly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. <

This LHGR times 1.02 is used in the heatup code along with the exposure 4 dependent staady state gas conduc*.ance and rod-to-rod local peaking factor.

' The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this UiGR of the highest powered red divided by its local peaking >

factor. The Ifatting value for APLHG3 is shown in Figure 3.2.1-1,4.c two

loog oper*Nou.

i The calculational procedure used to establish the APLHGR shown on Figure 3.2.1-1 is based on a loss-of-coolant accident analysis. The analysis was perfomed using General Electric (GE) calculational models which are i consistent with the requirements of Appendix K to 10 CFR 50. A complete i

j' discussion of each code employed in the analysis is presented in Reference 1.

Differences in this analysis compared to previous analyses perfomed with

! i Reference 1 are: (1) the analysis assumes a fuel assembly planar power consistent with 1025 of the MAPLHGR shown in Figure 3.2.1-1, (2) fission product decay is computed assuming an energy release rate of 200 MEV/ fission; 4

I asak s(tagnation periodpr(4) the effects of core spray entrainment and counter-3) pool boiling current flow limitation as described in Reference 2, are included in the reflooding calculations.

1 A list of the significant plant input parameters to the loss-of-coolant l accident analysis is presented in Bases Tacle B 3.2.1-1.

F Cd #

1 i se. usana shall ha Anulli(lkJ.

c o Wd O.E w b s-1

,( , . . w ,s. a ce , a u sewwwarea~A*+iwI*q*d%;"b*

l 1

l LA SALLE - UNIT 1 8 3/a 2-1

. i

f 7 POWER DISTRIBUT10M SYSTEMS t

8ASES 3/4.2.2 APRM SE90INTS

~

The fuel cladding integrity safety Limits of Specification 2.1 were based

,; on a power distribution which would yield the design LHGR at RATED THERMAL I. POWER. The flow biased simulated themal power upscale seras setting and - ,meAr 1

' control rod block functions of the APRM instruments r aust ce acjusted to ensure ha.$

that the MCPR does not become less thanAM or that > 1% plastic strain does gegr.

g 4,.) not occur in tne degraded situation. The scras settin,gs and rod bicek settings gp, d r are adjusted in accordance with the fomula in this specification when the com-bination of THERMAL POWER and MFLPD indicates a higher peaked power distribution 5%.Y 1:n to ensure that an UiGA transient would not be increased in the degraded condition.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel l cladctng integrity safety Limit MCPR-f '.%, and an analysis of abnc~ sal oporttional transients. For any abnormal operating transient analysis evalua- I tion with the initial condition of the reactor being at the steady-state operating limit, it is recuired that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instru-

.A sent trip setting given in Specification 2.2.

l To assure that the fuel cladding integrity Safety Limit is not exceeded l; during any anticipated abnormal operational-transient, the most limiting

', translents have been analyzed to datamine which result in the largest reduc-tion in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and

- coolant temperature decrease. The limiting transient yields the largest delta I MCPR. When added to the Safety Limit MCPR d '.O, the required minimum operating j

limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-1.

- The evaluation of a given transient begins with the systas initial parameters

showniin FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient j computar program. The code used to evaluata pressurization events is described in NED0-24154(3) ar.d the program used in non6ressurization events is described I in NED0-10802(2) ..

The outputs of this program along with the initial MCPR fom the input for further analyses of tne thermally limiting bundle with the single channel transient themal hydraulic TASC code described in NEDE-25149(#

The principal result of this evaluation is the reduction in MCPR c&. sed by the l transient.

I The need to adjust the MCPR operating limit as a function of scram time i

I arises from the statistical approach used in the implementation of tne 00YN

-i computer code for analyzing rapid pressurization events. Generic statistical

- analyses were performed for plant groupings of similar design which consicered the statistical variation in several parameters, i.e. , initial power level, CR0 scram insertion time, and andel uncertainty. These analyses, which are j'

u.;

$ LA 5ALLE - UNIT 1 8 3/4 2-3 ,

t

-..,,.,...,.,.,_.m. . . . . _ , . .._..,,,,m.,,,_,.,.,__,_,.__._.___,,,,,,,_,_.m,,,,.____,,.,.%,,._,__m__,,,_,,y._-_-. . , - .

- . - = . . . . _ _

TdifET foe PAGg B y/y g 3 Y NA ko miranIns~ I. g ap,,4;w gag ,; y reeiu.I.% [.,7 , g o

O e

4 l

l l

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

' * ~~

i INSTRtmgNTAT_LON 8ASES l

FIRE DETECTION INSTRUMENTATION (Continued) l l

In the event that a portion of the fire detection instrumentation is '

inoperable, increasing the frequency of fire watch patrols in the affected areas in required to provida detection capability until the inoperacle instrumentation is restored to OPERA 8ILITY. !

3/3.3.7.10 RADICACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION The radioactise liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid affluents during actual or potential releases of liquid effluents. The l alarm / trip setpoints for these instruments shall be calculated in accordance '

with the procedures in the 00CM to ensure f. hat the alans / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERASILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.7.11 RAOICACTIVE GASECUS EFFLUENT 4ONITORING INSTRUMENTATION l 0 t* r aia etiv e == friu at annitor and control, as appitcable, the releases of radioactive materials in aftarias ia tr= atatiaa is Provia a ==

gaseous effluents during actual or potential releases of gaseous affluents. The alam/ trip setpoints for these insrtments shall be calculated in accordance

. with the procedures in the 00CM to ensure that the alarm / trip will occur prior

. to exceeding the limits of 10 CFR Part 20. This instrumentation also includes

! provisions for monitoring (and controlling) the concentrations of potentially j explosive gas sixtures in the waste gas holdup system. The OPERA 8ILITY and

use of this instrumentation is consistent with the requirements of General

. Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

i j 3/4.3.7.12 t.0OSE-PART DETECTION SYSTEM i

The OPERA 8ILITY of the loose-part detection system ensures that sufficient

! capability is available to detect loose metallic parts in the primary systes and avoid or sitigate damage to primary system components. The allowable out-

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

3/4.3.8 FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main tuitine trip system actuation instrumentation is provided to initiate the feedwater systas/ main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 setpoint associated with a feedwater controller failurerfo pvem!: ovu ftllia . i e

mek.yessa uk;a.k m resul+ ;a kt possae tiyu;d d;scl.aqs 41sf roo ,

-ya lanaF talat disciary.hoez ;

LA SALLE - UNIT 1 8 3/4 3-6 4

wy- --- - , -

= e- t-w-- r-w yer c ww *e -w----,+r-- -,. , - - , -w,w-,-e- ==~t = = =--r---+--'*w-----T N-a, ---w -w - w w w-ee>.se-swme+-er-s--ww *+--e-me ver eeres -e-- - ---

~ .. -,

i.

3/4.4 REACTOR COOLANT SYSTEM 8ASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable -4e- I 7-""*M =ti' = ;::!=t = ;f th; ;;rfe. ;;; ;f t.5; :::: _.i a ea; le--

~=H aa

Mr ;;rf;M, :":!=tM ;.- ;;; min ;; ;;; ;;;;;*dh.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design-basis-accident by increasing the blowdown area and reducing the capacility of reflooding the core; thus, the requirement for snutcown of the facility with a jet ptmo inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed scheduled for significant degradation.

Recirculation loop flow sisaatch limits are in compliance with the ECOS LOCA analysis design criterion. The limits will ansure an adequata core flow coastdown from either recirculation loop following a LOCA. {

In order to prevent undue stress on the vessel nozzles and bottom head smr region, the recirculation loop temperatures shall be within 50*F of each otner (ellow(.ar pMa prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal f

' 0 shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the. vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the a temperature differenca was greater than 145'F.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety-relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordanca with the ASME Code. A total of 18 OPERA 8LE safety /

relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

l Demonstration of the safety-relief valve lift settings will occur only

' during shutdown and will be pe A rmed in accordance with the provisions of Specification 4.0.5.

k %ea sududed nad Lea Saud to be acceghale. dudag k

.(; rs + ko.\ da. o , gra&& 4 ouil- is o enled.la saardene.

y a p - answo., e m spw-km..a LA SALLE - UNIT 1 8 3/4 4-1

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

2Nseer von PA&o 89/44-I Where the recir-culation loop flow mismatch limits can not be maintained during the recir; culation loop operation, continued operation is permitted in the single ..

recirculation loop operation mode. .. . ,

e w.,-- ,y w - - - . , , < ,,w.-,,--, ,r----,-,,.,m,. ..r,, - - - , - - .--..,.._,.,,,-----n, . - - - - - - - - . - - - , - . - - - - , . - - . . , - - - , - - - - - , - - - - , -

l t

)

!. 3/4.5 EMERGENCY CORE COOLING SYSTEM

!. BASES i 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDCWN U r r . r - ECCS Division 1 consists of the low pressure core spray systas, low i pressure coolant injection subsystes "A" of the RHR systes, and the automatic depressurization systas (A05) as actuated by A05 trip systas "A". ECCS Division 2 consists of low pressure coolant injection subsystems "B"and "C" of the RHR system and the automatic depressurization systas as actuated by ADS trip systas "B".

The low pressure core spray (LPCS) systas is provided to assure that the

, core is adequataly cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the A05.

The LPCS is a primary source of emerge.ncy core cooling after the reactor veuel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequata assurance that the LPCS systan will be OPERABLE when required. Although all active components are

.n testable and full flow can be demonstrated by recirculation through a test

,' loop during reactor operation, a completa functional test requires reactor -

shutdown. The pump discharge piping is saintained full to prevent water

)1asserdamagetopipingandtostartcoolingattheearliestmoment.

The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Three subsystans, each with one pump, provide adequata core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the A05.

}- The surveillanca requirements provide adequata assurance that the LPCI

system will be OPERA 8LE when required. Although all active components are I tastanle and full flow can be demonstrated by recirculation through a test l

loop during reactor operation, a completa functional test requires reactor i shutdown. The p op discharge piping is saintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

ECCS Division 3 consists of the high pressure core spray systes. The j high pressure core spray (HPCS) systas is provided to assure that the reactor j core is adequately cooled to limit fuel clad temperature in the event of a sas11 break in the reactor coolant systas and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCS system pemits the reactor to be shut down while saintaining sufficient reactor t

vessel water level inventory until the vessel is depressurized. The HPCS system operates over a range of 1160 psid, differential pressure between reactor vessel and HPCS suction source, to O psid.

i The capacity of the HPCS systas is selected to provide the required core i

.] cooling. HPCS pump is designed to deliver greater than or equal to

's 516/1550/ . gpa at differential pressures of 1160/1130/200 psid. Initially,

, water from the condensata storage tank is used instead of injecting . eater from

&200

{ LA SALLE - UNIT 1 8 3/4 5-1 1

o ;

) . I 4 ,

l 0 caMrat>iar svstras 1 BASES 3/4.6.1.5 PRIMAltY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment wilt be amintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 45 psig in the event cf a LOCA. The

.j sessurement of containment tendon lift-off force, the ta'sile n tests of the g tendon wires or strands, the visual examination of tendons, anchorages and

, exposed interior and exterior surfaces of the containment, the chemical and visual examination of the sheathing filler grease, and the Type A leakage test are sufficient to demonstrate this capability.

~

l The surveillance requirements for demonstrating the prismry containment's

. structural integrity and the method of predicting the pre-stress loses are in j p campliance with the recommendations of Regulatory Guide 1.35.1, " Inservice

-p9 C - * " :: of Ungrouted Tendons in Prestressed Concrete Containment Structures,"

.I ril 1979 with the following clarification: the tested lift-off fome of I individual tendon tension shall be greater than or equal to the initial j pre-stress minus the loses, as predicted in the as-built design, which occur j between the initial pre-operational structural integrity test and the time of O

~

subsequent surveillance.

4---tu ser$

1/4.6.1.6 DRWELL AND SUFPRESSION CHAMBER INTERNAL PRES 5URE - Mfogs

" ")

The limitations on drywell and suppression enenbar internal pressure I ensure that the containment peak pressure of 39.6 psig sses not excesd the .

design pressure of 45 psig during LOCA conditione or that the external pres-sure differential does not exceed the design mesteus external pressure differen-tial of 5 psid. The 11af t of 2.0 peig for initial positive primary containment pressure will limit the total pressure to 39.6 psig which is less than the design pressure and is consistant with the accident analysis.

l 3/4.6.1.7 DRWELL AVERAGE AIR TEMPGATURE t The limitation on drywell average air tammerature ensures that the containment peak air tesserature does not exceed the design tamperature of 340*F during LOCA conditions and is consistent with the accident analysis.

l 3/4.6.1.8 ORWELL ANO SUP*RESSION CHAM 8ER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required

for inerting, de-inerting and pressure control. Until thesa valves have been comonstrated capable of closing during a LOCA or steam line break accicent, they shall be blocked so as not to open more than 50*.

Amst t T1 %% 6veds l. 35.1, beht>rin o fre sbegsic J Torcas.o $erb .% A d hfrefosA 5fakasoof '

s+ Cu 15 CwksnArf Chf wse ,

LA 5ALLE - UNIT 1 S 3/4 6-2 l

l t

m., m- -_._,__._._-,-..__..m.... _ . . _ _ _ . , _ _ . . _ _ _ _ _ _ . _ . . . . _ _ . , _ . _ _ . __- - _ _ . . _ _ _ _ _ _ - , _ _ . - _ . . _ . . _ _ .

T U StLr PMs 8 s/4 6 -G -

The required Special Reports from any engineering evaluation or contain-ment abnormalities shell include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerances on cracking, the results of the engineering evalua-tion, and the corrective action taken.

l 1

i i

e o

I

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

~- .- - - - _

~ ~

l l i

l I

i

)

.s.+ , l

- PLANT SYSTDf5 BASES 3/4.7.6 FIRE' P.ATED ASSEMBLIES i.

i. The OPERA 8ILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited. These design features minimize the l possibility of a single fire involving more then one fire area prior to det.?ction and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.

} 3/4.7.7 AREA TEMPERATURE MONITORING i

i The area temperature limitations ensure that safety-related equipment will I not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can

,:ause loss of its OPERA 8ILITY. The temperature limits include allowance for an j instrument error of 2 7*F.

_. 3 3/a.7.8 STRUCT1)RAL INTEGRITY OF CLASS 1 STRUCTURES. .

J _

In order to assure that settlement does not exceed predicted and allowable settlement values, a program has been established to conduct a survey at the i site. The allowable total differential settlement values are based on original i settlement predictions. In establishing these tabulated values, an assumption i is made that pipe and conduit connection have been designed to safely withstand the stresses which would develop due to total and differential settlement.

j I

I i

i

, I l

l ~

! , er+

fello.s ny f8

.e i s3

~~

l I LA SALLE - UNIT 1 8 3/4 7-3

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

TA/50ET~ Foc. PAFE $ 7/V 7-3 W4 1.9. Sull66EGS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effe.-t on any safety related system. ;

size. Snubbers are classified and grouped by design and manufacturer but not by For example, mechanical snuboers utilizing the same design features of )

the 2-kip,10-kip, and 100* kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purpose of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.

(

l l

l

N N

' . ./ PLANT SYSTEMS SES s

3/a. 9 SNUBBERS A snubbers are required OPERA 8LE to ensure that the structural integrity of the actoriiroolant spystem and all other safety related systems is maintained l during an following a seismic or other event initiating dynamic loads. Snubbers excluded f this inspection program are those installed on nonsafety related systems and 5en only if their failure or failure of the system on whfen they are installed, would have no adverse effect on any safety related system.

The visual nspection frequency is based upon maintaining a constant level p of snubber prot 'on to systems. Therefore, the required inspection interval varies inversely w< h the observed snubber failures and is detemined by tne number of inoperaal snubbers found during an inspection. Inspections performed before that interval s elapsed say be used as a new reference point to determine the next inspection. ver, the results 'of such early inspections performed before the original reg red time interval has elapsed, nominal time less 25%, say not be used to lengthen required inspection intarval. Any inspection whose results require a shortar pection interval will override the previous schedule.

6

. When the cause of the re ion of a snubber is clearly established and

- remedied for that snubber and r any 'othof snubbers that may be generically susceptible, and verified by ins ice functional testing, tnat snuboer may be

- exempted from being counted as i rable. Generically susceptible snubbers j are those snubbers which are of a s ific make or model and have the same

design features directly related *a ection of the snubber by visual inspec-

. tion, or are sfeilarly located or expo to the same environmental conditions

' such as tamperature, radiation, and vib tion.

l l When a snubber is found inoperable, engineering evaluation is performed, j in addition to the determination of the s er mode of failure, in order to i determine if any safety-related component or stes has been adversly affected

by the inoperability of the snubber. The engi ring evaluation shall detarsine

whether or not the snubber mode of failure has arted a significant effect j or degradation on the supported component or syst .

t i To provide assurance of snubber functional rol flity, a representative j sample of the installed snubbers shall be functional tested during plant shut-downs at 18 month intervals. Selection of a represen tive sample according to the expression 35 (1 + h)) provides a confidence level f approximately 95% that i

905 to 100% of the snubbers in the plant will be OPERA 8LE ithin acceptance limits.

Observed failures of these sample snubbers will require fun tional testing of additional units.

Hydraulic snubbers and mechanical snubbers may each be tre tad as a j different entity for the above surveillance programs.

V; %

f l

LA SALLE - UNIT 1 8 3/a 7-4 h

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

- +

%, ,m Pt. ANT SYSTEMS l

BASES i $NUSSERS (Continued)

The service life o a snubber is evaluated via manufacturer input and information through consi ration of the snuceer service conditions and associated installation an maintenance records (newly installed snuceer, seal replaced, spring replaced, high radiation area, in high temperature area, etc.). The requirement to ao tar the snubner service life is included to ensure that the snubbers peri cally undergo a performance evaluation in view of their age and operating condi fons. Tnese records will provide statistical bases for future consideration of nubber service life. The requirements for the saintenance of records and the ubber service life review are not intanced to affect plant operation.

~

4 3/4.7.10 MAIN TUR$INE BYPASS SYSTEM

.- The main turbine bypass system is req red OPERA 8LE as assumed in the feedwater controller failure analysis.

C i

l l

l

,h i

i i !

!I .

\

3 i

-- . LA SALLE - UNIT 1 , 8 3/4 7-5 gg, ,,m . y ea e e .e .e ee. 4"*6 #*"

W

, - . , , - - - . ,.-,-,--,,-w,c.- . -,.--w,, n,.. - - - , e ,-.n,,,,,, - , - , w.,,,,.wm-, , , . .w, ,,-yw_,wi,,__.,_.,.,.. - .

PLANT SYSTEMS BASES SNUBBERS (Continued)

A list of individual snubbers with detailed information of snubbers loca-tion and size and of system affected shall be available at the plant in accord-ance with Section 50.71(c) of 10 CFR Part 50. . The accessibility of each snubber shall be determined and approved by the Onsite Review and Investigative Function.

. The determination shall b2 based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g. ,

temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guide 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

The visual inspection' frequency is based upon maintaining a constant level of snubber protection to each safety-related system. Therefore, the required inspection interval varies inversely with the observed snubber failures on a given system and is determined by the number of inoperable snubbers found during an inspection of each system. In order to establish the inspection frequency for each type of snubber on a safety-related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber on that system could cause the system to be v"- unprotected and to result in failure during an assumed initiating event.

Inspections performed before that iqterval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection l interval will override the previous schedule.

The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared i inoperable and shall not be determined OPERABLE via functional testing.

To provide assurance of snubber functional reliability, one of three functional testing methods is used with statoo acceptance criteria:

1. - Functionally test 10% of a type of snubber with an additional 10%

tested for each functional testing failure, or

2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or l 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.

(

s LASALLE-UNIT [i 8 3/4 7-4 l

m

,-,-...,._-,o,-- -

- .., ,-y-,,--,.,,,,w -v.w----,.--,ce-,m,,,r-mw,,.v,-,..w.,

m , .

l PLANT SYSTEMS

~.

A. BASES

't SNUBBERS (Continued)

Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. Duncan.

l Permanent or other exemptions free the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption l is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the com-I pletion of their fabricati'on or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.

t The service life of a snubber is established via manufacturer input and infomation through consideration of the snubber service conditions and I associated installation an~d maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.- These records will provide statistical 4- bases for future consideration of snubber service life.

3/4.7.10 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system is required OPERABLE as assumed in the feedwater controller failure analysis.

l .

i l

l l

~

p .

( .

w.

j LA SALLE - UNIT /l B 3/4 7-5

,-,-..--.r-e, er-,-,- , gm, - , .,,.-n.w-r -

-..,,w,,, -

. a k

m 3/4.'u RADI0 ACTIVE EFFLUENTS 8ASES 3/4.u.1 LIQUID EFFLUENTS -

3/4.u.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in ifquid wasta effluents from the sita will be less than the concentration levels specified in 10 CFR Part 20, Appendix S, Table II, Col un 2. This 11mitation provides additional assurance that the levels of radioactive materials in bodies of water outsica the site will result in exposure within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration 1 faits for dissolved or entrained noble gases '

were determined by converting their MPC's ivi air to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICAP) Pubitcation 2.

3/4.u.1.2 005E

~

This specification fs provided to taplement the requirements of Sections II.A. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation faplessnts to guides set forth in Section II. A of Appendix I. The ACTION stateme9ts provide the required operating flexibility and at the same time fapienent the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid -

effluents will be kept "as low as is reasonably achievable." Also, for fresh l

' water sitas with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calcula-tions in the 00CM implement the requirements in Section III.A of Appendix I that confomance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriata pathways is unlikely to be suestantially

, underestimated. The equations specified in the 00CM for calculating the doses due to the actual release ratas of radioactive saterials in liquid affluents are consistant with the methodology provided in Regulatory Guide 1.109,

,' " Calculation of Annual Ooses to Man free Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, October 1977 and Regulatory Guide 1.u3, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Isplementing Appendix I," April 1977. rat hceMM$4tOS N This specification applies to the release ofaliquid effluents from each I reactor at the site. For units with shared radwasta treatment systans, the l.

Ifquid effluents from the shared systas are proportioned among ene units v sharing that system.

L LA SALLE - UNIT 1 8 3/4 u-1 r w,- . . - - - _ _ _ _ _ . . _ _ - , - _ - . , _ , . - _ ~ ._,,.,,.,,.,,,,.__,.,,.._.c.. ..,._.._._..-__,.m.,,,,-_---,.-,-w- .#,..,., ,,-w. ,. _...e

1 I.

.s %,

- l

. RADICACTTVE EFFLUENTS 1 5

l BASES 3

I

. 005E RATE (Continued) infant via the cow-silk-infant pathway to less than or equal to 1500 arem/

. year for the nearest cow to the plant. #gg i diogr.44, id g AW5 This specification applies to the release of.geoeous affluents 3 from all (

reactors at the site. For units within shared raawaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that systas.

3/4.11.2.2 005E - N08LE GASES This specification is provided to impiosent the requirements of Sections II.8, III. A and IV. A of Appendix I,10 CFR Part 50. The Lief ting Conditions for Op4 ration are the guides set forth in Section II.8 of Appendix I. The ACTION

, ] st.ttaments provide the required operating flexibility and at the same time

implement the guides set forth in Section IV.A of Appendix I to assure that h,. the relaases of radioactive estarial in gaseous affluents will be kept "as low as is reasonably achievable." The Surveillance Requirements fuglement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data l 4 such that the actual exposure of an individual through appropriata pathways is l j unlikely to be sucstantially utidorestimated. The dose calculations estaclished

.. in the 00CM for calculating the doses due to the actual release rates of -

i' ' radioactive noble gases in gaseous affluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from .

l .l' Routine Releases of Reactor Effluents for the Punose of Evaluating Compitance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory 8

Guide 1.111. " Methods for Estinating Atmospheric Transport and Dispersion of

'! Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,"

Revision 1. July 1977. The 00CM equations provided for determining the air j doses at the site boundary are based upon the historical average atmospheric conditions.

j i

l ! 3/a.11.2.3 005E - RADICIODINES, RADICACTTVE MATERIALS IN PARTICULATE FORM

! AND RADICNUCUDE5 OTHER THAN NC8LE GASES I

i The specification is provided to implement the requirements of Sections II.C.

j III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for

Operation are the guides set forth in Section II.C of Appendix I. The ACTION i- ! statements provide the required operating flexibility and at the same time i implement the guides set forth in Section IV.A of Appendix I to assure that

q the releases of radioactive materials in gaseous effluents will be koot "as

.s low as is reasonacly achievable." The 00CM calculational methods specified in the Surveillance Requirements isolement the requirements in Section III. A of ,

LA SALLE - UNIT 1 8 3/4 11-3

g. e um o m , . e>e g

i s

s . ...

3/4. U RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4. U.1 MONITORING PROGRAM i

The radiological monitoring program required by this specification provices seasurements of radiation and of radioactive materials in those exposure path-ways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring progras by verifying that the asasurable concentrations of radioactive materials and levels of raciation are not higher than expected on the basis of the affluent seasure-sents and sodaling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first ures ~3 l years of commercial operation, as defined i,n the ODCM.

The detection capabilities required by Table 4.12-1 are stata-of-the-art for routine environmental asasurements in industrial laboratories. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as "a posteriori" (after the fact) limit for a particular esasurement. Analyses shall be

_T perfomed in such a manner that the stated LLDs will be achieved under routine V conditions. Occasionally background fluctuations, unavoidahl,y small sample sizes, the presence of interfering nuclides, or other uncontrollacle circum-stances osy render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

. 3/4.12.2 t.ANO USE CENSUS .

! This specification is provided to ensure that changes in the use of i unrestricted areas are identified and that modifications to the monitoring {

! program are made if required by the results of this census. The best survey aerial 4r consulting with local agricu -

U f]

i infomation fromsnaii the door-to-doof,,This census satisfies the requirteents of i e tural authorities ne usea. '

g, I Section IV.S.3 of Appendix I to 10 CFR Part 50.

l l

i

r~

V LA SALLE - UNIT 1 B 3/4 12-1

( ..- .

i

5. 0 OESIGN FEATURES 1 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1. 2 The low population zone shall be as shown in Figure 5.1.2-1.

SITE SOUNDARY FOR GASEOUS EFFLUENTS .

5.1. 3 The site boundary for gaseous effluents shall be as shown in Figure 5.1.1-1.

SITE BOUNDARY FOR LIQUIO EFFLUENTS

5. L 4 The site boundary for liquid affluents shall be as shown in Figure 5.1.1-1.

5.2 CONTAINMENT s CONFIGURATION

5. 2.1 The primary containment is a steel ifned post-tensioned concrete

~

structure consisting of a drywell and suopression chamber. The drywell is a steel-lined post-stressed concreta vessel in the shape of a truncated cone closed by a steel done. The drywell is above a cy1!ndrical steel-lined post stressed concrete suporession chamber and is attached to the suppression chamber through a series of downconer vents. The drywell has a minious free air volume of tet;4 E cubic feet. The suppression chamber has an air region of 164,800 to l 2%gd 168,100 cubic feet and a water region of 128,800 to 131,900 cubic feet.

, DESIGN TEMPERAT1JRE ANO PRESSURE

5.2.2 The primary containment is designed and shall be maintained for:

, a. Maximum internal pressure 45 psig.

l b. Maximum internal tamperature: drywell 340*F.

! suppression chammer 275"F.

4

, j c. Maximum external pressure 5 psig.

d. Maximus floor differential pressure: 25 psid, downward.

, 5 psid, upward.

SECONOARY CONTAINMENT .

_ 5.2.3 The secondary containment consists of the Reactor Building, the equimment access structure and a portion of the main steam tunnel and has a minimum free

'} volume of 2,875,000 cubic feet.

L.A SALLE - UNIT 1 ,

5-1 w v w, --e c - --

-e.v-iu--,ey y f, , , - --e ~ie a, tec=-- -e=e- ww -- -- rww.w.,r-t e-v--*e--nv --e,e e--emewmun r- - - -*- =- -w-we

!h ADMINISTRATIVE CONTROLS l, Any devi~ation from the above guidelines shall be authorized by the (Station i

~

Superintendent or his deputy, or higher levels of management, in accordance with i es tablished procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual

, overtime shall be reviewed monthly by the Station Superintendent or his designee

to assure that excessive hours have not been assigned. Routine deviation from the

! above guidelines.is not authorized.

O. Qualifications of the station management and operating staff shall meet minimum acceptable levels as described in ANSI N18.1, " Selection and Training of Nuclear Power Plant Personnel," dated March 8, 1971. The

. Rad /Chen Supervisor shall meet the requirements of radiation protection manager of Regulatory Guide 1.8, September, 1975. The ANSI N18.1-1971 l, qualification requirements for Rad /Chen Technician may also be met by t

either of'the following alternatives:

1. Individuals who have coepleted the' Rad /Chen Technician training program and have accrued 3ne' year of working experience in the {

specialty, or A.

l _. 2. Individuals who have completed the Rad /Chen Technician training program, but have not yet accrued % year of working experience l

' in the specialty, who are supervised by on-shift health physics

!Q supervision who meet the requirements of ANSI N18.1-1971 Section 4.3.2, " Supervisor Not Requiring AEC Licenses," or (j Section 4.4.4, " Radiation Protection."

'I E. Retraining and replacement training of Station personnel shall be iri accordance with ANSI N18.1, " Selection and Training of Nuclear Power

Plant Personnel", dated March 8, 1971 and Appendix "A" of 10 CFR Part 55, i and shall include familiarization with relevant industry operational experience identified by the 0NSG.

Retraining shall be conducted at intervals not exceeding 2 years.

F.

1 l G. The Review and Investigative Function and the Audit Function of activities t affecting quality during facility operations shall be constituted and l have the responsibilities and authorities outlined below: '

i 1. The Supervisor of the Offsite Review and Investigative Function shall be appointed by the Director, Nuclear Safety. The Audit i Function shall be the responsibility of the Manager of Quality Assurance and shall be independent of operations.

a. Offsite Review and Investigative Function The Supervisor of the Offsite Review and Investigative Function i 'shall: (1) provide directions for the review and investigative i function and appoint a senior participant to provide appropriate direction, (2) select each participant for this function, (3) select a complement of more than one participant who collectively -

O possess background and qualifications in the subject matter under review to provide coeprehensive interdisciplinary review coverage LA SALLE - UNIT 1 6-3 Amendment No.14

1 I

m r pauown tus s'. t l .

~

~ t

! li

(-

i i "$155 ifl il l: .

.[ ..li.ill l'> t-------

a -

!=s --

les:--

gIj3 l

.. - -- I l [

, i k! - it!

l l ha ih< -

i i I. -

lH .-

t is .

F

sg i .r. . .

I l H.gl -

i1 - ' -

, ,. .o.!. : 1 ,

3 l

i i

l .' =-

li g

<. g s.

8

, s I;

. ij i.

In;3 . -

i: i _ .

t i e g I -

i i

- ',- s[ig

. js - i ..

'gi l

e cls '3 ,

t z e, '-

g s

l, \ i, i

w I

~

a es ,

m II

~

l Eg3 4

'ga l'.

., e l! ,

. ., ogl --

g

, . . , . \ !! ,!u,

~~

T T' he

' -i u a e I t- :: it

.lrg -- li -

3 r.t ga: . ..,

j -

Ii1l. '.: ,l 1

ii -

H 1 is -J

+

i

_ -, a

~

[% { g r g ;

.- 1 -

u=s I' . :s r

- -- - 1 I b- - lg 3

l c

n,

.! in l ti m ..s :: 31 -

, I i

1

[3ll a si. 8 zg l' 4 .' -

l >> 8; llc 5 ! . . . _ .s '

, ggi -

lg 3

=

g i.

I -

{

lli E]1

's .

~

{. i 1

l.A SALLE - WIT 1 f-11 w v------v----r-w +~ +n- , -- ,m e- w ,e- ,.www w-- n-------.*n--------=-vrm---9+-- =- ~owwwwwt-e--w '--'s#wr T --Yw ' - ' - *- ~ ' - * ' ~ ' ' w

~ -

~~._.... l . ~~ . . .: -.

s "

l .

h

e. #1 3j .

t.

D

. . .- g lI ~~

lI. .

f, Ij i

e 3. :

~

I l.

p'

,j< -

Jg 1 3 11 L

r ,-

i II r-

.t ., Lu a ki l;i

                                                                                                                                           }I-5 1

Il

f. 1.

[.

                                                                            -                           i
                                                                                                                                                               .s .;
                                                                                                  ,                                                               }
                                                                                                                    ;            I I;                          -

11 ,J

                                    .                         ..                 1 2            -

3. 1 Il ji l[ ljl ll1 lj I ~

I l i . .ll ;Il p .

                                                                                                                                 ,11   -

Is ens,.mm q y .,

                                                                         .I jf}

i (g.- is ri 3 1 r - LA SALLE - UNIT 21 6-11

                  ,                       . . ._ ,,,- -- , . - -- .----,                                                                   --v     ,   - . - ,       .-w - .- -    -- -,- - , -- . - --,. . - - - , - .- , , . - - -

m. . _m .

l

              . s,                                                ,

gg ,w er Pe# l

                                                                                                    ~1N5G'C_'

Ffoure 6.1-3 MINIMLM SHIFT CREV COMPOSMON i. 3 P05 m 0N g NUMBER OF I;CIVIDUALS REQUIRED TO FILL POSITION kCONDITION1,2and3 CON 0! TION 4 and 5 SE 1 1 SF 1 None M 2 1

    ;                                                 M                           2                              1 i

SCRE 1 None ; l i i i f

             ]__                                  or, whenever a SCRE (SR3/STA)            not included in the shift crew
   .                                              cosposition, the sintem shift               composition shall be as follows:                             .

i t. , I POSITICM MLsWER OF IMIVIDUALS REQU%ED TO FILL POSITION j CONDITION 1, 2 and 3 lkCONDITION4and5 l- SE 1 1 1

   ,                                                  SF                          1                            None M                           2                              1 M                           2                              1 STA                         1                               ne l

l'-O . LA SALLE - UNIT 1 6-13

h l _ l I L

T

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

Fiaure 6.1-3

         ,,       ..                                              MINIMUM SHIFT CREW COMPOSITI0h WITH UNIT Jg IN CONDITION 1, 2, OR 3 PO5ITION               NUMBER OF INDIVIDUAL 5 REQUIRED TO FILL POSITION CONDITIONS 1, 2 -and 3                                      CONDITIONS 4 and 5 8                                                  8 SE                                1                                                     1 8
     '_.                                           SF                                1                                                    None M                                5                                                      1 M                                U                                                      1 SCRE                                  a

, l None or, whenever a SCRE (SRO/STA) is not included in the shift crew composition, the minimum shift crew composition shall be as follows: WITHUNIT4INCONDITION1,2,OR3 r POSITION NUMBER OF INDIVIDUAL 5 REQUIRED TO FILL POSITION CONDITIONS 1, 2 and 3 CONDITIONS 4 and 5 8 8 SE 1 1

   -                                                                                    a SF                               l                                                     None b

R0 2 y M h 1 STA 8 1 None WITN UNIT 4 IN CONDITION 4 OR 5 OR DEFUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2 and 3 CONDITIONS 4 and 5 8 8 SE 1 1 SF 1 None RO 2 1 A0 b 2 2 STA 1 None

LA SALLE - UNIT fl 6-13 gg7 ; .g

                -      .--,e,    y -- .#, - - , ,             e..  .e ,w,,.e-    - - - .-,,,we-.,,e*---- ,w         e -,.=ve                  -,,--e-w.--,,    w w       ,-___w-----____s   --

                                                         *~      ' " ' '               '*
               ~. .

Fiaure 6.1-3 (Continued) MININUM SHIFT CREW COMPOSITION NOTES SE - Shift Supervisor (Shift Engineer) with a Senior Reactor Operators License on Unit 1. SF - Shift Foreman with a Senior Reactor Operators License on Unit 1. R0 - Individual with a Reactor Operators License on Unit 1. 5 AO - Auxiliary Operator. SCRE - Station control Roca Engineer with a Senior Reactor Operators License. 4 i Except for the Shift Supervisor, the Shift Crew Composition may be one less

      ;                             than the minious requirements of Figure 6.1-3 for a period of time not to
                    ,               exceed 2 hours in order to accommodata unexpectad absence of on cuty snift crew members provided i W iata action is taken to restore the shift crew composition to within the minimum requirements of Figure 6.1-3. This
      ,                             provision coes not permit any shift crew position to be unmanned upon shift change due to an oncoming shift cre man being late or absent.

While the unit is in OPERATIONAL CONDITION 1, 2, or 3, an individual with a valid SRO license shall be designated to assume the Control Room direction

]'

function. While the unit is. in OPERATIONAL CONDITION 4 or 5, an individual with a valid SRO or R0 Itcense shall be designated to assume the Control Roos direction function. . O I I' l e"

                        / f ,,Inesvidual y f111 the sa.e positinn on unit a b/
                                - One of the two required individuals may fill the same position on Unit 1 I                                                                                                   -

l LA SALLE - UNIT 1 6-14

s ,.

     .                       A0ftINISTRATIVE CONTROLS

Pi. ANT OPERATING RECORD 5 (Continued)

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

plant: .

             ~           ~

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

     ;                               L      Changes sede to the plant as it is described in the SAR;
     .                               3. Records of new and spent fuel inventon and assembly histories; i

4 4. Updated, corrected, and as-built drawings of the plant; j 5. Records of plant radiation and contamination surveys; i 6. Records of offsite environmental 1sonitoring surveys; I 7. includin [- Records of radiation exposure for all plant personnel,ith contractors and visitors to the plant, in accordance w Ry0; 10 C(g

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

9. Records of transient or operational cycling for those components that

 -   -                                         have been designed to operate safety for a limited number of transient I                                         or operational cycles (identified fn Table 5.7.1-1);
     .f                               10. Records of individual staff members indicating qualifications, J                                          experience, training, and retraining; l                                11. Inservice inspections of the reactor coolant system;

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

13. Records of reactor tests and experiments;

14. Records of Quality Assurance activities required by the QA Manual;

. 15. Records of reviews performed for changes made to procedures on ecuip- ! ment or reviews of tests and exper h nts pursuant to 10 CFR 50.59; and l 15.. Records of the service lives of all hydraulic and sechanical snubcers

                                               'h'd r T b. 0.7.11-1 -,4 0.L L 2 including the date at which the i

l

    .! reguired by                             service life commences and associated installation and maintenance
      ! 5 peut: d'oo                           mcords.

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

      !         .s
       ;       G LA SALLE - UNIT 1                                    6-20 em            - ,e e   -

r

                       . . . .     .                                                                                                           ~
                   .r .

N .-

           ,                    ADMINISTRATIVE CONTROLS Thirty-Day Written Recorts (Continued)

e. An unplanned offsite release of 1) more than 1 curie of radio-active asterial in liquid effluents, 2) nore than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of

         "                                              radiofodine in gaseous affluents. The report of an unplanned offsite release of radioactive material shall include the
        ,                                               following infomation:

1. A description of the event and equipment involved.

        ,                                               2.           Cause(s) for the unplanned release.

i

3. Actions taken to prevent recurrence.

         ,                                              4.           Consequences of the unplanned release.
        .'                                     f.       Measured levels of radioactivity in an environmental sampling j        j                                               mediumidetermined to exceed the reporting level values of i

Table 3.12-2 when averaged over any calendar quartar sampling period.

        ;                       C.       Unique Reporting ~ Requirements l                                                                        .

l , 1. Special Reports shall be submitted to the Of rector of the Office of t Inspection and Enforcement (Region III) within the time period l .

       ,                                      speciffed for each report.

g i 6. 7 PROCESS CONTROL PROGRAM (PCP)*" l

      !                         6. 7.1 The PCP shall be approved by the Commission prior to implementation.

i l i 6.7.2 Licensee initiated changes to the PCP:

a. ShallbesubmittedtotheCommissioninthesemibnualRadioactive l

Effluent Release Report for the period in wnich the change (s) was l ! : mode. This submittal shall contain: l ' j 1. Sufficiently detailed information to totally support the rationale I for the change without benefit of additional or supplemental i infomation; i 2. A detarsination that the change did not reduce the overall conformance of the solidified waste product to existing criteria

j for solid wastes; and

   *i                                         3.       Documentation of the fact that the change has been reviewed and l

1 n found acceptaale by the Onsite Review and Investigative Function.

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

(P :- u . v rs i.a st e. d. :r 16-28ad L 3 s.l e. % ~' z LA 5 LLE - UNIT 1 l

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

r--

                     ~~
                                                        ~                                                                            ,

l

                       ^
    .        .        'J                                                                                                     '

l

             ,                     ADMINISTRATIVE CONTROLS '

6. 8 0FFSITE DCSE CALCULATION MANUAL (00CM W 6.8.1 The 00CM shall be approved by tho' Commission prior to implementation. .

6.8.2 Licensee initiated ' changes to the 00CM:

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

           '                                   1. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

i 2. A datamination that the change will not reduce the accuracy or

      ,j                                             reifability of. dose calculations or setpoint dataminations; and 3                     3. Documentation of the fact that the change has been reviewed and found acceptable by the ansita Review and Investigative Function.

b. Shall becnos effective upon review and acceptance by the Onsite Review and Investigative Functico.

6.9 %10R CHANGES TO RADI0 ACTIVE W.5TE TREATEMENT SYSTEMS 3 6.9.1 Licensee initiated safor changes to the radioactive wasta systems

        ;                         (liquid, gaseous and solid):

i : l a. Shall be reported to the Commission in the Monthly Operating Report i for the period in which the'avaluation was reviewed by the Onsite [ i Review and Investigative Functien. The discussion of each change

         !                                    shall ebatain                  ,

1. A summar/ of the evaluation that led to the datamination that

        !                                           the change could be made in accordance with 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason l ! for the change without k nefit or additional or supolamental l ; infomation; I :

i i 3. A detailed desc-iption of the equipment, components and processes i involved and the interfaces with other plant systems; O t ODcH) C.mou to h&Ile 0~rI su L.1 %[g Q, 7?_ s LA SALLE - UNIT 1 6-29 l

  .                         . - -                           ,_ _ _                  _.      _      _ _ _ _ .}}

ML20087N316

Text

7.0 seconds

= =; = =. ..=1 =;m c u ==i= .= m==i

Navigation menu

Search