Text

Amends 81 & 65 to Licenses NPF-11 & NPF-18,respectively
	ML20086H621
Person / Time
Site:	LaSalle
Issue date:	11/27/1991
From:	Barrett R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20086H624	List: ML20086H621; ML20086H628;
References
	NUDOCS 9112090251
	Download: ML20086H621 (45)
	v • d • e

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT-TO FACILITY OPERATING LICENSE Amendment No. 81 License No. NPF-11 1;

.The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment filed by the Comonwealth Edison-Company (the licensee), dated October. 10, 1990, as an. ended October 16

' Energy Act of 1954, as amended (the Act)quirements-of the Atomic-1991, complies with-the standards and re

, and the Comission's regulations set forth in.10 CFR Chapter I; B.

.The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission;

.C.

There is reasonable assurance: (1)thattheec\\.ivitiesauthorizedby this amendment can be conducted without endangering.the health and

' safety of the public,; and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter 1; D.-

The issuance of this amendment will not be-inimical to the common defense and security or to the health and safety of the public; and E.

The-issuance of this amendment is.in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions.as indittted ir the enclosure to this license amendment and paragraph.

2.C.(2) of the acility: Operating License No. NPF-11 is hereby amended tb read as follows:

9112090251 911127 PDR ADOCK 05000373 P

PDR

1-u-

(2)i-Technical Specifications and Environmental Protection Plan 4

The-Technical. Specifications contained _in. Appendix A, as revised

-through Amendment No.- 81, and the Environmental Protection Plan contained in. Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical. Specifications and the-Environmental Protection Plan.

3.-

This amendment is effective upon date of issuance to be--implemented prior to startup following the L1R05 refueling outage.

FOR THE NUCLEAR'REGUtA Y COMMISSION 1

4

/j/d' 0

Rich r

. Barrett, Director Project Directorate 111-2 Division of. Reactor Projects --lil/IV/V' Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical

Specifications

Date of Issuance:

November 27, 1991-1 1

l, t

L

[{

't ATTACHMENT-TO LICENSE AMENDMENT NO. 81 FACILITY OPERATING LICENSE NO.-NPF-11 DOCKET NO.- 50-373

_ Replace the following pages of the Appendix "A" Technical Specifications with

- the enclosed pages.. The revised pages are-_ identified by amendnent: number and contain a-vertical:line-indicating the area of change.

REMOVE:

INSERT 4

3/4L3-26 3/4 3-26

-3/4 3-27'

'3/4 3-27 3/4 3 '3/4 3-30 "3/4-3-30a 3/4 3-30a 3/4'3-32 3/4 3-32 3/4 3 3/4 3-33 3/4 3-34 3/4 3-34 3/4 5-5 3/4-5-5 3/4 5-6 3/4 5-6

-r

-3/4 5-7 3/4.5-7 3/4~5 3/4 5.3/4_5-9

.3/4'5-9 3/4 6-33 3/'.-33

.3/4 6 '34

4 6-34

/

.-3/4. 6-34a 3/4 8-30 3/4 8-30:

B 3/4.5-1:

'B 3/4 5-1=

B 3/4 5 ~

B 3/4 5-2 B 3/4 6-3a B 3/4 6-3a f

1 TABLE 3.3.3'

-(Continued) 5-EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENIATION vs?5 MINIMUM OPERABLE APPLICABLE CHANNELS PER TRIP OPERATIONAL i

!E TRIP FUNCTION FUNCTION (a)

CONDITIONS ACTICN Z

C.

DIVISION 3 TRIP SYSTEM

~

1.

HPCS SYSTEM a.

Reactor Vessel Water Level - Low, Low, Level 2 4

1, 2, 3, 4*, 5*

35 b.

Drywell Pressure - High 4(cj 1,2,3 35 c.

Reactor Vessel Water !evel-High, Level 8 2

1, 2, 3, 4*, 5*

32 d.

Deleted e.

Deleted f.

Pump Discharge Pressure-High (Bypass) 1 1,2,3,4*,5*

31 m

a.

HPCS System Flow Rate-Low (Permissive) 1 1, 2, 3, 4*, 5*

31 1

6.

Manual Initiation 1/ division 1, 2, 3, 4 *, 5*

34 5

D.

LOSS OF POWER _

MINIMUM APPLICABLE TOTAL NO.

INSTRUMENTS OPERABLE OPERATIONAL I

OF INSTRUMLNTS TO TRIP INSTRUMENTS ")CONDITIONS ACTION 1.

4.16 kv Emergency Bus Undervoltage 2/ bus 2/bu:

2/ bus 1, 2, 3, 4**, 5**

37 (Loss of Voltage) 2.

4.16 kv Emergency Bus Undervoltage 2/ bus 2/ bus 2/ bus 1, 2, 3, 4**: 5**

37 (Degraded Voltage)

(a) A channel instrument 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 during periods of required surveillance without placin g

one other OPERABLE channel /g the trip system / channel / instrument in the tripped condition provided at least a

instrument in the same trip system is raonitoring that parameter.

g (b) Also actuates the associated division diesel generator.

5 (c) Provides signal to close HPCS pump discharge va!ve only on 2-out-of-2 logic.

j

- Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.

2o Required when ESF equipment is required to be OPERABLE.

Not requird to be OPERABLE when reactor steam dome pressure is < 122 psig.

?

e.

a TABLE 3.3.3-1 (Continued)

EMERGENCYCOREC00LJpuSYSTEMACTUATIONINSTRUMENTATION ACTION

-ACTION 30 -

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

a.

With one' channel inoperable, place the inoperable channel in the tripped condition within one hour

or declare the associated system inoperab~1e.

b.

With'more than one channel inoperable, declare the associated system inoperable.

ACTION 31 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function, place the inoperable channel in the tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated system inoperable.

ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum 0PERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.

ACTION 33 -

With the number of OPERABLE chtnnels less than the Minimum OPERABLE Channels per Trip Furction requirement, place the inoperable' channel in the tripped condition within one hour.

ACTION 34 -

With the number of OPERABLE channels less than required by the Minimurr OPERABLE Channels per Tria Function requirement, restore

_the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the associated ADS trip-system or ECCS inoperable.

ACTION 35 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function. requirement a.

For one-trip system, ) lace that trip system in the tripped condition within one lour

or declare the HPCS system

' inoperable, b.

For both trip systems, declare the HPCS system i~ ierable.

ACTION 36 -

Deleted ACTION 37 --

With the number of OPERABLE instruments less than the Minimum Operable Instruments, place the inoperable instrument (s) in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

or declare the associated l

emergency diesel generator inoperable and take the ACTION l

required by Specification 3.8.1.1 or 3.8.1.2 as appropriate.

l l

The provisions of Specification 3.0.4 are not applicable.

LA SALLE - UNIT 1 3/4 3-27 Amendment No. O, 81

. TABLE 3.3.3-2 (Continued) i r*

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS l

sg ALLOWABLE 8

TRIP FUNCTION TRIP SETPOINT VALUE C

3 C.

DIVISION 3 TRIP SYSTEM 1.

HPCS SYSTEM a.

Reactor Vessel Water level - Low Low, level 2

>- 50 inches *

>- 57 inches

b.

Drywell Pressure - High 7 1.69 psig

< 1.89 psig c.

Reactor Vessel Water Level - High,~ Level 8

{55.5. inches *

{56 inches

d.

Deleted e.

Deleted f.

Pump Discharge Pressure - High

> 120 psig

> 110 psig g.

HPCS System Flow Rate - tow i 1000 gpm i 900 gpm h.

Manual Intiation NA NA

,s*

D.

LOSS OF POWER m.

E 1.

4.16 kV Emergency # Bus Undervoltage (Loss of Voltage) a.

4.16 kV Buses

1) Divisions I and 2 2625 131 volts with 2625 262 volts with

$ 10 seconds time delay i 11. seconds time delay 2496 125 volts with 2496 250 volts with

> 4 seconds time delay

> 3 seconds time delay p

<oa

2) Division 3 2870 143 volts with 2870 287 volts with j

1 10 seconds time delay 1 11 seconds time delay c+

E

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

3

These are inverse time delay voltage relays or instantaneous voltage relays with a time delay. The voltages shown are the maximum that will not result in a trip.

Lower voltage conditions will result in decreased trip times.

I

)

TABLE 3.5.3-2 (Continuad) 9 E!1ERGENCY CORE COOLING SY5 TEM ACTUATION INSTRUMENTATION SETPOINTS E

ALLOWASLE TRIP FUNCTION TRIP SETPOINT VALUE c-3 2.

4.16 kV Emergency Eus Undervoltage (Degraded Voltage) a.

4.16 kV Buses i

1) Divisions 1, 2 and 3 3814 76 voits with 3814 76 velts with 10 1 seconds time 10 1 seconds time delay with LOCA signal delay with LOCA signal or or 5

0 5 minutes time 5

0.5 minutes time delay without L0tA delay withaut LOCA signa; signal 1

Y8 w

a 3-

-I W

i' TABLE 4.3.3.1-1

~

2-m EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANC.E REQUIREMENTS l

G CHANNEL OPERATIONAL

[-

CHANNFL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH g

TRIP FUNCT!UN CHECK TEST CALIBRATION SURVEILLANCE REQUIRED i

A. DIVISION I TRIP SYSTEM

1. RHR-A (LPCI MODE) AND LPCS SYSTEM l

i a,

Reactor Vessel Water Level -

Low Low Low, level 1 NA H

R 1, 2, 3, 4*, 5*

i b.

Drywell Pressure - High NA M

Q 1, 2, 3..

c.

LPCS Pump Discharge Flow-Low NA M

Q 1, 2, 3, E *, 5*

d.

LPCS and LPCI A Injection Valve Injection Line Pressure Low l

Interir-k NA M

R 1, 2, 3, 4*, 5*

w e.

LPCS an/ cPCI.A Injection Valve D

Rear' tr Pressure Low Interlock NA M

R 1, 2, 3, 4*

5*

f.

LPCI rump A Start Time Delay Relay NA M

Q 1, 2, 3. '*, 5*

w J,

g.

LPC' Pumo A Flow-Low NA M

Q 1, 2, e,,r*, 5*

l t

Mr.ual Initiation NA R

NA 1, 2, 3, 4*, 5*

l I

2. AUTOViTIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

i I

Reactor Vessel Water Level -

I a.

Low Low Low, level 1 NA M

R 1,2,3 l

b.

Drywell Pressure-Hign NA M

Q 1, 2, 3 c.

Initiation Timer NA M

Q 1,2,3 l

d.

Reactor Vessel Water Level -

s 37 Low, Level 3 NA M

R 1,2,3 l

e.

LPCS Pump Discharge Pressure-High NA M

Q 1, 2, 3 i

f.

LPCI Pump A Discharge Pressure-High NA M

Q 1,2,3 l

F g.

Manual Initiation NA R

NA 1, 2, 3 l

h.

Drywell Pressure Bypass Tim, NA' M

Q 1,2,3 l

3 i.

Manual Inhibit NA R

NA 1,2,3

t I

TABLE 4.3.3.1-1 (Continut 2,

i g;

EMERGENCY CORE COOLING SYSTEM ACTUATION TNSTRUMENTATION SURVEILLANCE RE4?IREMERTS r-E

^

CHANNEL OPERATIONAL c:

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WK t

l

!$ TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE FE0i j'_

B. DIVISION 2 TRIP SYSTEM

1. RHR B AND C (LPCI MODE) a.

Reactor Vessel fater Level -

Low Low Low, Level 1 NA M

R 1, 2, 3, 4*, 5*

i.

b.

Dryvell Pressure.- High NA M

Q 1, 2, 3 c.

LPCI B and C Injection Valve Injection Line Pressure Low Interlock NA M

R 1, 2, 3, 4*, 5*

i os d.

LPCI Pump B Start Time Delay Relay NA M

Q 1, 2, 3, 4*, 5*

3s e.

LPCI Pump Discharge Flcv-Low NA M

Q 1, 2, 3, 4*, 5*

e, f.

Manual Initiation NA R

NA 1, 2, 3, 4*, 5*

l o',

g.

LPCI B and C Injection Valve Reactor Pressure Low Interlock NA M

R 1, 2, 3, 4*, 5*

2. AUTCKATIC DEPPCSSURIZATION SYSTEM TRIP SYSTEM "B"#

a.

Reactor Vessel Water Level -

Low Low Low, level 1 NA M

R 1,2,3 b.

Drywell Pressure-High NA M

Q 1, 2, 3 c.

Initiation Timer NA M

Q 1,2,3 3I d.

Reactor Vessel Water Level -

I E

Low, Level 3 NA M

R 1, 2, 3 Et e.

LPCS Pump B and C Discharge i

Pressure-High NA M

Q 1, 2, 3 f.

Manual Initiation NA R

NA 1, 2, 3 l

i EF h.

Drywell Pressure Bypass Timer NA M

Q 1,2,3 i.

Manual Inhibit NA R

NA 1,2,3 j

??

t

TABLE 4.3.3.1-1 (Continued) 5 g;

ENERGENCY CORE CDOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS t

CHANNEL OPERATIONAL c:

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 15' TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED C. DIVISION 3 TRIP SYSTEM

1. HPCS SYSTEM f

a.

Reactor Vessel Water Level -

Low Low, level 2 NA M

R 1, 2, 3, 4*, 5*

b.

Drywell Pressure-High NA M

Q 1, 2, 3 l

c.

Reactor Vessel Water Level-High Level 8 NA M

R 1, 2. 3, 4*, 5*

d.

Deleted e.

Deleted 20 f.

Pump Discharae Pressure-High NA M

Q 1, 2, 3, 4*, 5*

g.

HPCS System flow Rate-Low NA M

Q 1, 2, 3, 4*, 5*

1

'f h.

Manual Initiation NA R

NA 1, 2, 3, f~, 5*

w 2n D. LOSS OF POWER l

1.

4.16 kV Emergency Bus Under-voltage (Loss of Voltage)

NA NA R

1, 2, 3, 4**, 5**

2.

4.16 kV Emergency Bus Under-NA NA R

1, 2, 3, 4* *, 5*

voitage (Degraded Voltage) i t

i

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 122 psig.

Er

When the system is required to be OPERABLE after being manually realigned, as applicable, per i

g Specification 3.5.2.

Et

- Required when ESF equipment is required to be OPERABLE.

E.

i

.F t

b r

w EMERGENCY CORE COOLING SYSTEMS SURVEILLANCEREQUIREHENTS(Continued)

(a) LPCS system to be 5 500 psig and 1 55 psig, respectively.

(b) LPCI subsystems to be i 400 psig and 1 55 psig, respectively.

2)

Low pressure setpoint of the HPCS system to be >

63 psig.

b)

Header delta P instrumentation and verifying the setpoint of the:

1)

LPCS system and LPCI subsystems to be i I psid.

2)

HPCS system to be 5 1 2.0 psid greater than the normal indicated AP.

3.

Deleted.

4.

Visually inspecting the ECCS corner room watertight door seals and room penetration seals and verifying no abnormal degradation, damage, or obstructions, d.

For the ADS by:

1.

At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the accumulator backup compressed gas system low pressure alarm system.

2.

At least once per 18 months:

a)

Performing a system functional test which includes simulated i

automatic actuation of the system throuahout its emergency operatingsequence,butexcludingactuaTvalveactuation, b)

Manually opening each ADS valve and observing the expected change in the indicated valve position, c)

Performing a CHANNEL CAllBRATION of the accumulator backup compressed gas system low pressure alarm system and verifying an alarm setpoint of 500 + 40, - O psig on decreasing pressure.

L i

LA SALLE - UNIT 1 3/4 5-5 Amendment No.-18, 81

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:

The low pressure core spray (LPCS) system with a flow path capable a.

of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel, b.

Lowpressurecoolantinjection(LPCI) subsystem"A"oftheRHRsystem with a flow path capable of takir.g suction from the suppression chamber upon being manually realigned and transferring the water to the reactor vessel.

Low pressure coolant injection (LPCI) subsystem "B" of the RHR system c.

ith a flow path capable of taking suction from the suppression chamber w

upon being manually realigned and transferring the water to the reactor

vessel, Lowpressurecoolant-injection d.

withaflowpathcapableoftak(ingsuctionfromthesuppressioncham-LPCI) su of the RHR system ber upon being manually realigned and transferring the water to the reactor vessel, The high pressure core spray (HPCS) system with a flow path capable e.

of taking suction from the sup3ression pool and transferring the water through the spray sparger to tie reactor vessel.

APPLICABILITY:

OPERATIONAL CONDITION 4 or 5*.

ACTION:

With one of the above required subsystems / systems inoperable, restore a.

at least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all operations that have a potential for draining the reactor

vessel, b.

With both of tne above requirt ! subsystems / systems inoperable, suspend CORE ALTERATIONS and all o)erations that have a potential for diuining the reactor vessel, lestore at least one subsystem /

system to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Ihe ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and 1

water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

LA SALLE - UNIT 1 3/4 5-6 AMENDMENT NO. 81 l

u

- - - - ~

~

n,.,

,--.--w,-nmn

9 %

EMERGENCY CORE COOLING SYSTEMS S_tRVE!LLANCE REQUIREMEN15 u

4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requiretrent 4.5.1, except that the header delta P instrumentation is not required to be OPERABLE.

i l

l l

LA SALLE - UNIT 1 3/4 5-7 AVr. nMENT HO. 81 i

e-

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

EMERGENC,Y_, CORE COOLING SYSTEMS 3/4.5.3 SUPPRI:SSION CHAMBER #

LIMITING CONDITION FOR 0,PERATION 3.5.3 The suppression chamber shall be OPERABLE:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with a contained water volume of H

3 at least 128,800 ft, equivalent to a level of -4 1/2 inches.**

b.

In OPERATIONAL CONDITION 4 or 5* with a contained water volume of at least 70,000 f t8, equivalent to a level of -12 feet 7 inches.**

APPLICABILITY:

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

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with the suppression chamber water level less than the above limit, restore the water level to within the limit 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 4 or 5* with the suppression chamber water level less than the above limit, suspend CORE ALTERATIONS and all l

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 .

4

See Specification 3.6.2.1 for pressure suppression requirements.

The_ suppression chamber is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

- Level is referenced to a plant elevation of 699 feet 11 inches (See Figure B 3/4.6.2-1).

' LA SALLE - UNIT 3 3/4 5-8 Amendment No.

59 81

EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued) c.

With one suppression chamber water level instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 7 days or verify the suppression chamber water level to be greater than or equal to -41/2 inches ** or -12 feet 7 inches **, as applicable, 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, 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 :sext 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 verify the suppression chamber water level to be greater than or equal to -41/2 inches ** or -12 feet 7 inches **, as applicable, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local indication.

i SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying:

a.-

The water level to be greater than or equal to, as applicable:

1.

--4 1/2 inches ** at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

-12 feet 7 inches ** at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Two suppression chamber water level instrumentation channels OPERABLE by performance'of a:

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 low water level aiarm setpoint at greater than or equal to

-3 inches.**

4.5.3.2 With the su TIONAL CONDITION 5*,ppression chamber level les. than the above limit in OPERA at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify footnote conditions

to be satisfied.

The suparession chamber-is not required to be OPERABLE provided that the reactor vessel lead is removed, the cavity is flooded or being flooded from the suppres-sion pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

- Level is referenced to a plant elevation of 699 feet 11 inches (See Figure B 3/4.6.2-1).

LA SALLE - UNIT 1 3/4 5-9 Amendment No. M, 81

i

.1

)

TABLE 3.6.3-1 (Continued)

[

PRIMARY CONTAINMENT ISOLATION VALVES E

VALVE FUNCTION AND NUMBER

.g Other Isolation Valves (Continued)

[

4.

Low Pressure Core Spray System IE21-F005 I5)

IE21-F001 1E21-F012((3) i)

1E21-F011I 1E21-F018(lI 1E21-F031(3) k)

1E21-F006 5.

High Pressure Core Spray System R

1E22-F004, 1E22-F015'(5) 3)

T 1E22-F023 ti

~1E22-F012(d) fi) 1E22-F014 IE22-F005(k) 6.

Reactor Core Isolation Cooling System 1E51-F013 1E51-F069 IE51-F028 IE51-F068 E'

1E51-F040 1E51-F031(l)

I 1E51-F019 ))

IE51-F065(k)

IE51-F066(*k) 1E51-F059( )

F IE51-F022(*):

q 1E51-F3625 1r51-F3631 i

B i

t i

TABLE 3.6.3-1-(Continued).

y.>

g PRIMARY CONTAINMENT ISOLATION VALVES

.p.

'm VALVE FUNCTION AND NUMBER

- t Other Isolation Valves (Continued) r E:

Z

.7.

Post LOCA Hydrogen Control w

1HG001A,.B l

IHG002A, B 1HG0054,-B f

1HG006A, B L

j-8.

Standby Liquid Control System IC41-F004A, B-IC41-F007 -

w L

1 9.

Reactor Recirculation Seal Injection i

m 4*

1833-F013A,B(I)

II)

IB33-F017A, B i

10.

Drywell' Pneumatic System

'i j.

IIN018 t

l

[

l i

3 l

r

?

l l

1

-h

?

i S

f i

1 I

a i

c

TAB,E 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES But 2 3 seconds.

(a) See Specification 3.3.2, Table 3.3.P.-1, for isolation signal (s) that operates each valve group.

(b) Not included in total sum of Type B and C tests.

(c) May be opened on an intermittent basis under administrative control.

(d) Not closed by SLCS actuation.

(e) Not closed by Trip Functions Sa, b or c, Specification 3.3.2, Table 3.3.2-1.

(f) Not closed by Trip Functions 4a, c, d, e or f of Specification 3.3.2, Table 3.3.2-1.

(g) Not subject to Type C leakage test.

(h) Opens on aa isolation signal.

Valves will be open during Type A test.

No Type C test required.

(i) Also closed by drywell pressure-high signal.

(j) Hydraulic leak test at 43.6 psig.

(k) Not subject to Type C leakage test - leakage rate tested per Specifica-tion 4.4. 3.2.2.

(1) These penetrations are provided with removable spools outboard of the outboard isolation valve.

During operation, these linet will be blind flanged'using a-double 0-ring and a type B leak test.

In addition, the packing of these isolation valves will be soap-bubble tested to ensure insignificant or no leakage at the containment test pressure each refueling outage.

(m) If valves IE51-F362 and 1E51-F363 are locked closed and acceptably leak rate. tested, then valves 1E51-F059 and 1E51-F022 are not considered to be primary containment isolatian valves and are not required to be-leak rate tested.

(n) Either the 1E51-F362 or the 1E51-F363 valve may be open when the RCIC system is in the standby mode of operation, and both valves may be open during opetation of the RCIC system in the full flow test mode, providing that:

1) valve IE51-F022 is acceptably' leak rate tested, and 2) valve IE51-F059 is deactivated, locked closed and acceptably leak rate tested, and 3) the spectacle flange, installed immediately downstream of the IE51-F059 valve, is closed and acceptably leak rate tested.

.LASALLE - UNIT 1 3/4 6-34a Amendment No. 81

TABLE 3.8.1.3-1 (Continued)

MOTOR OPERATED VALVES THERMAL DVERLOAD PROTECTION BYPASS DEVICE SYSTEM (S)

VALVE NUMBER (Continuous)(Accident Conditions)

AFFECTED 1.

1E32 - F001A Accident Conditions MSIV-LCS 1E32 - F002A Accident Conditions 1E32 - F003A Accident Conditions 6

1E32 - F001E Accident Conditions 1E32 - F002E Accident Conditions 1E32 - F003E Accident Conditions 1E32 - F001J Accident Conditions 1E32 - F002J Accident Conditions IE32 - F003J Accident Conditions IE32 - F001N Accident Conditions IE32 - F002N Accident Conditions

IE32 - F003N Accident Conditions 1E32 - F006 Accident Conditions 1E32 - F007 Accident Conditions 1E32 - F008 Accident Conditions 1E32 - F009 Accident Conditions m.

1E22 - F004 Accident Conditions HPCS system 1E22 - F012 Accident Conditions 1E22 - F015 Continuous 1E22 - F023 Accident Conditions LASALLE - UNIT 1 3/4 8-30 AMENDMENT NO. 81 l

3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN ECCS Division I consists of the low pressurecoolantinjectionsubsystem"A"pressurecorespraysystem,lowof the RHR system depressurization system (ADS) as actuated by ADS trip system "A".

ECCS Division 2consistsoflowpressurecoolantinjection Asystems B"and "C" of the RHR system and the automatic depressurization syaem as actuated by ADS trip system B".

The low pressure core spray oreisadequatelycooledfollowin(galoss-of-coolantaccidentLPCS) system is provided and provides adeqtate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for transients or smaller breaks following depressurization by the A05.

The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that t' ' LPCS sys-tem will be OPERABLE when required.

Although all active components

.'e testable and full flow can be demonstrated by recirculation through a test loop during reactor o]eration, a complete functional test requires reactor shutdown.

The pump discierge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

Thelowpressurecoolantinjection(LPCI)modeoftheRHRsystemispro-vided to assure that the core is adequately coolcd following a loss-of-coolant accident.

Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for transients or small breaks following depressurization by the A05.

The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.

The pump discharge piping is maintained 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 system.

The high pressure core spray (HPCS) system is provided to assure that tha reactor core is adequately cooled to limit fuel clad temperature in the event.of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.

The HPCS system permits the reactor to be shut down while maintaining sufficient reactor 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.

The caaacity of the HPCS system is selected to provide the required core coolin$.

Tie HPCS pump is designed to deliver greater than or ecual to 516/15 0/6200 gpm at differential pressures of 1160/1130/200 psic, Water is-takenfromthesuppressionpoolandinjectedintothereactor.

LA SALLE - UNIT 1 B-3/4 5-1 Amendment No. 27, 31

l-EMERGENCY CORE COOLING SYSTEHS BASES ECCS-OPERATINGandSHUTDOWN(Continued) l With the HPCS system inoperable, ade @ n e core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system,ically provide makeu) a system for which no credit is taken in the hazards analysis, will automat at reactor operating pressures on a reactor low water level condition.

The iPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required.

Although all active components are test-able and full flow can be demonstrated by recirculation through a test loop during rtactor operation, a complete functional test with reactor vessel iniec-tion requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCS system to function properly, if required, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in-time to limit fuel cladding temperature to less than 2200'F.

ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 122 psig even though low pressure core cooling systems provide adequate core cooling up to 350 psig.

ADS automatically controls seven selected safety-relief valves.

Six valves are required to be OPERABLE since~the LOCA analysis assumes 6 ADS valves in addition to a single failure.

It is therefore appropriate to permit one of the recuired valves to be out-of-;ervice for up to 14 days without materially recucing system reliability.

3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is also required to be OPERAB;.E as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA.

This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow.to the core (See Figure B 3/4.6.2-1).

The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.2.1.

Repair work might require making the-suppression chamber inoperable.

This specification will permit those repairs to be made and at the same time give. assurance that-the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable in OPERATIONAL CONDITION 4 or 5.

In OPERATIONAL CONDITION 4 and 5 the suppression chember minimum required water volume is reduced because the reactor coolant is maintained at or below 200 F.

Since pressure suppression is not required below 212 F, the minimum water volume is based on NPSH, recirculation volume, vortex prevention plus a 2'-4" safety margin for conservatism.

LA SALLE - UNIT 1 B 3/4 5-2 Amendment No. H, 81 1

I

m Control Suppression Plant Room / Local Chamber Level Elevation Indication 26' 10" 700' 2"

+3" High Level LCO (Volume 3

131,900 ft )

26' 9" 700' 1"

+2" High Level Alarm 26' 7" 699' 11" 0" Instrument Zero 26' 4" 699' 8"

-3" Low Level Alarm 26' 2 1/2" 699' 6 1/2"

-4 1/2" Low Leve'i LC0 Opera-tional Condition 1, 2, 3 or 3 (Volume 128,800 ft )

14' 687' 4"

-12' 7" Low Level LCO Opera-tional Condition 4 or 3

5 (Voiume 70,000 ft )

F SUPPRESSION POOL LEVEL SETPOINTS BASES FIGURE B 3/4.6.2-1 LA SALLE - UNIT 1 B 3/4 6-3a Amendment No. 59, 81

n.

o s mo.

o UNITED STATES

+

- [' 3 p,

^,

NUCLEAR REGULATORY COMMISSION ti

-t W AsHINGTON, D. C. 20b55 i

s%,.'... + /

COMM0tiWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No NPF-18 1.

The-Nuclear Regulatory Comission (the Comissioni has found that:

A.

The application for amendment filed by the Comonwealth Edison Company (thelicensee),datedOctober 10, 1990, as amended October 16, 1991, complies with the standards and requircaents of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the. Act, and the regulations of the Comission; C.

-There'is reasonable assurance: (i)that'theactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of-the public, and (ii)-that-such activities will be conducted in compliance wit >: the Conrnission's regulations set forth in 10 CFR Chapter I;.

D.-

The issuance of this amendment will not be inimical to the.comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amendeo by changes to the Technical Specifica-tions as indicated in the enclosure to this license amendu.ent and paragraph-2.0.(2) of the Facility Operating License No.-.NPF-18 is hereby anended to read as follows:-

U E,

- v 4

, ~~,-.-

r.,

e..,r-,--,

.-,r-e-,,-r,nny.r--

%. - - -I

2 (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 65, and the Environmental Protection Plan contained in Aspendix B, are hereby incorporated in the license.

The licensee sia11 operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective upon date of issuance to be implemented prior to startup following the L2R04 refueling outage.

FOR T11E HUCLEAR REGULATOR COMMIS$10N

/

h\\

RichardJ.Berre(n O

t, Director Project Directorate III-2 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 27, 1991 v v

,m e

w~

--,,,-------n.---

ATTACH!iENTTOLICEllSEAMENDMEtiTNO.J5_

FACILITY OPERATING LICENSE NO. Nt'r-18 DOCKET NO._50-374 Replace the following pages of the Appendix "A* Technical Ipecifications with the enclosed pages.

11e revised pages are identified by arnendment nurnber and contain a vertical lir.e indicating the area of change.

Pages ind)cated by an asterisk are provided for convenience.

REMOVE INSERT 3/4 3 26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-30 3/4 3-30 3/4 3-30a 3/4 3-30a 3/4 3-34 3/4 3-34 3/4 5-5 3/4 5-5 3/4 5-6 3/4 5-6 3/4 5-7 3/4 5-7 3/4 5-8 3/4 5-8 3/A 5-9 3/4 5-9 3/4 6-36 3/4 6-36 3/4 6-37 3/4 6-37 3/4 6-37a

3/4.6-38

3/4 6-38 3/4 8-30 3/4 8-30 B 3/4 5-1 B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 B 3/4 6-3a B 3/4 6-3a

^

E

- TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION G

MINIMUM OPERABLE APPLICABLE e

CHANNELS PER TRIP OPERATIONAL g

FUNCTION (a)

CONDITIONS ACTION Z

TRIP FUNCTION m

C.

DIVISION 3 TRIP SYSTEM 1.

HPCS SYSTEM a.

Reactor Vessel Water Level - Low, Low, Level 2 4

1, 2, 3, 4*, 5*

35 b.

Drywall Pressure - High 4(C).

1, 2, 3 35 c.

Reactor Vessel Water level-High, Level 8 2

1, 2, 3, 4*, 5*

32 d.

Deleted e.

Deleted w

f.

Pump Discharge Precsure-High (Bypass)-

1 1, 2, 3, 4*, 5*

31 1

g.

HPCS System Flow P 'e-Low (Permissive) 1 1, 2, 3, 4*, 5*

31 w

h.

Manual Initiation 1/ division 1, 2, 3, 4*, 5*

34 l

m D.

LOSS OF POWER MINIHLH TOTAL NO.

INSTRU-OPERABLE APPLICABLE OF INSTRU-MENTS TO INSTRU-OPERATIONAL MENTS TRIP MENTS(a)

CONDITIONS ACTION 1.

4.16 kV Emergency Bus'Undervoltage 2/ bus 2/ bus 2/ bus 1, 2, 2, 4**, 5**

37 (Loss of Voltage) 2.

4.16 kV' Emergency Bus Undervoltage 2/ bus 2/ bus 2/ bus 1, 2, 3, 4**, 5**

37 (Degraded Voltage)

TABLE NOTATION (a) A channel / instrument may be'placen in.in inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of required surveillance without placing the trip system /channelfinstrument in the tripped condition provided at least F

one other OPERABLE channel / instrument in the same trip system is conitoring that parameter.

(b) Also actuates the associated division diesel generator.

D (c) Provides signal to close HPCS pump discharge valve only on 2-out-of-2 logic.

g Applicable when the' system is required to be 0PERABLE per Specification 3.5.2 or 3.5.3.

g; Required when ESF equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is < 122 psig.

_..=:.....a

. ~.. ~...... - - -

~

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 -

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

a.

With one channel inoperable, place the inoperable channel in the tripped condition within one hour

or declare the associated system inoperable, b.

With more than one channel inoperable, declare the associated system inoperable.

ACTION 31 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function, place the inoperable channel in the tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated system inoperable.

ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.

ACTION 33 -

With the number of OPERABLE channels less than the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

ACTION 34 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels.per Trip Function requirement, restore the inoper*. ale channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or

= declare the associated ADS trip system or ECCS inoperable.

ACTION 35 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement a.

For one trip system, place that trip system in the tripped condition within one hour

or declare the HPCS system inoperable.

bi For both trip systems, declari. the HPCS system inoperable.

ACTION 36 -

Deleted ACTION 37 -

With the number of OPERABLE instruments less than the Minimum OPERABLE INSTRUMENTS, place the inoperable instrunent(s) in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

or declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2 as appropriate.

The provisions of Specification 3.0.4 are not applicable.

LA SALLE - UNIT 2 3/4 3-27 Amendment No. 27, 65 1

I

^

TABLE 3.3.3-2 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 2-ALLOWABLE y,

TRIP FUNCTION TRIP SETPOINT VALUE DIVISION 3 TRIP SYSTEM 1.

HPCS SYSTEM c5 a.

Reactor Vessel Water level - Low Low, Level 2

>- 50 inches *

>- 57 irches*

N b.

Drywell Pressure - High 7 1.69 psig i 1.89 psig t

c.

Reactor Vessel Water Level - High, Level 8 7 55.5 inches

7 56 inches

d.

Deleted

~

e.

Deleted f.

Pump Discharge Pressure - High

> 120 psig

> 110 psig g.

HPCS System Flow Rate - Low

> 1000 gpm

> 900 gpm h.

Manual Intiation N.A.

R.A.

'D.

LOSS OF POWER i

R 1.

4.16 kV Emergency Bus Undervoltage (Loss of Voltage)#

3 a.

4.16 kV Buses l

1) Divisions 1 and 2 2625 131 volts with 2625 262 volts with

[

$ 10 second time delay 1 11 second time delay 2496 125 volts with 2496 1 250 volts with

> 4 second time delay

> 3 second time de.ay

2) Divisinn 3 2870 143 volts with 2870 1 287 volts with 1 10 second time delay 1 11 second time delay

?

i i

(

TABLE NOTATIONS f

a

~;

See Bases Figure B 3/4 3-1.

m

[

These are inverse time delay vo tage relays or instantaneous voltage relays with a time delay.

The voltages l

,o shown are the maximum that will not result in a trip.

Lower voltage conditions will result in decreased trip i

times.

N. A. Not Applicable -

S i

I

TABLE 3.3.3-2 (Continued)

E EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS wR ALLOWABLE E

TRIP FUNCTIOM TRIP SUPOINT VALUE i

g D.

LOSS OF POWER (Continued)

Z 2.

4.16 kV Emergency Bus Undervoltaga (Degraded Voltage) a.

4.16 kV Buses l

I

1) Divisions 1, 2 and 3 3814 76 volts with 3814 76 volts with 10 i 1 seconds time 10 1 seconds time delay l

delay with LOCA signal with LOCA signal l

or Or

(

5 0.5 minutes time 5 1 0.5 minutes time delay i

j delay without LOCA without LOCA signal signal

[

k l

M I

{

8 i

a a

i E

t s

4 r

g 5

TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL c

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH j

TRIP FUNCTION CtfECK TEST CALIBRATION SURVEILLANCE REQUIRED C.

DIVISION 3 TRIP SYSTEM 1.

HPCS SYSTEM a.

Reactor Vessel Water Level -

Low 1,ow, Level 2 NA M

R 1, 2, 3, 4*, 5*

b.

Drywell Pressure-High NA -

M Q

1, 2, 3 c.

Reactor Vessel Water Level-High Level 8 NA M.

R 1, 2, 3, 4*, 5*

d.

Deleted w

e.

Deleted 5

f.

Pump Discharge Pressure-High NA M

Q 1, 2, 3, 4*, 5*

w g.

HPCS System Flow Rate-Low NA M

Q 1, 2, 3, 4*, 5*

y h.

Manual Initiation-NA R

NA 1, 2, 3, 4*, 5*

D.

LOSS OF POWER l.

4.16 kV Emergency Bus Under-NA' NA R

1, 2, 3, 4**, 5**

voltage (Loss of Voltage) 2.

4.16 kV Emergency Bus Under-NA NA R

1, 2, 3, 4**, 5**

voltage (Degraded Voltage)

{

TABLE NOTATIONS

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 122 psig.

5

When the system is required to be OPERABLE after being manually realigned, as applicable, per g

Specification 3.5.2.

{

- Required when ESF. equipment is required to be OPERABLE.

T S

u

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

Performing a CHANNEL CAllBRATION of the:

a)

Discharge line " keep filled" pressure alarm instrumentation and verifying the:

1)

High pressure setpoint and the low pressure setpoint of 4he:

(a) LPCS system to be 5 500 psig and 1 55 psig, respectively.

(b) LPCI subsystems to be $ 400 psig and 1 55 psig, respectively.

i 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 i 1 psid.

2)

HPCS system to be 5 1 2.0 psid greater than the normal indicated AP.

3.

Deleted 1

4.

Visually inspecting the ECCS corner room watertight door seals and room penetration seals and verifying no abnormal degradation, damage, or obstructions.

d.

For the ADS by:

1.

At least once per 31 days, performing 4 CHANNEL FUNCTIONAL TEST of the accumulator backup compressed gas system low pressure alarm system.

2.

At least once per 18 months:

a)

Performing a system functional test which includes simulated 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 change in the indicated valve position.

c)

Performing a CHANNEL CALIBRATION of the accumulator backup compressed gas system low pressure alarm system tnd verifying an alarm setpoint of 500 + 40, - O prig on decreasing pressure, l

LA SALLE - UNIT 2 3/4 5 S AMENDMENT NO. 65 l

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS - SHUTDOWN tlMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:

a.

The low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

-b.

Low pressure coolant injection-(LPCI) subsystem "A" of the RHR system with a flow path capable of taking suction from the suppression chamber upon being manually realigned and transferring the water to the reactor vessel, c.

Low pressure coolant injection (LPCI) subsystem "B" of the RHR system with a flow path capable of taking suction from the suppression chamber upon being manually realigned and transferring the water to the reactor

vessel, d.

Low pressure coolant injection (LPCI) subsystem "C" of the RHR syste.

with a flow path capable of taking suction from the suppression chamber upon being manually yealigned and transferring the water to the reactor vessel, e.

The high pressure core spray (HPCS) system with a flow path ca)able of taking suction from the suppression pool and transferring tie water through the spray sparger to the reactor vessel.

APPLICABILITY:

OPERATIONAL CONDITION 4 or 5*.

ACTION:

a.

With one of the'above required subsystems / systems inoperable, restore at least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all cperations that have a potential for draining the reactor vessel.

b.

With both of the above required subsystems / systems inoperable, suspend CORE ALTERATIONS and a'l operations that have a potential for draining the reactor vessel.

Restore at least one subsystem /

. system to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish-SECONDARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The ECCS is not required to be OPERABLE provided that the reactor vessel head is remuved, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

LA SALLE - UNIT 2 3/4 5-6 AMENDMENT NO. (25

_A

i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 1

4,5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1, except that the header delta P instrumentation is not required to be OPERABLE.

l r

6 i

p 4

- LA_SALLE - UNIT 21 3/4 5-7 AMENDMENT NO. 65 i

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSION CHAMBER #

LIMITING CONDITION FOR OPERATION 3.5.3 The suppression chamber shall be OPERABLE:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with a contained water volume of at least 128,800 f ta, equivalent to a level of -41/2 inches.**

-b.

In OPERATIONAL CONDITION 4 or 5* with a contained water volume of at least 70,000 ft3, equivalent to a level of -12 feet 7 inches.**

l APPLICABILITY:

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

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with the suppression chamber water > level less than the above limit restore the water level to withinthelimitwithinIhourorbeInatleastHOTSHUTDOWNwithin 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 4 or 5* with the suppression chamber water level less than the above limit, suspend CORE ALTERATIONS and all l

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 .

See Specification 3.6.2.1 for pressure suppression requirements.

The suppression chamber is not reavired to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of i

Specifications 3.9.8 and 3.9.9.

- Level is referenced to a plant elevation of 699 feet 11 inches (see Figure B 3/4.6.2-1).

LA.SALLE' - UNIT 2 3/4 5-8 Amendment No. M, 65

EMEaM.

CORE COOLING SYSTEF 11Mll A CC M' FOR O H RATION (Continued) y

}

.wed)

"If dit ura suppression chamber water level instrumentation channel i

.noperalle, restore the i. perable channel to OPERABLE status within 7 days or verify the suppression chamber water level to be greater

>, on equal to -41/2 inches ** or -12 feet 7 inche:**, as applicable, s

" 'i' /- J I o<st once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by incal indication.

'th both supp.

chamber water level instrumentation channels V"i nuperable, rer

<t least one inoperable channel to OPERABLE E

.tatus within 4 s or be in at least HOT SHUTDOWN with'n the next 12 hcur. and in C0LD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and verify the suppression chamber water level to be greater than or j

!ual to -4 1/2 iaches** o -12 feet 7 inches **, as applicable, at

. east once per 12 ho. irs t;y local indication.

&VEILLANCEREQUIREMENTS 4.5.3.1 The suppression chamber shall be de'. ermined OPERABLE by verify 16.g:

a.

The water level to be greater than or equal to, as applicable:

1.

-4 1/2 inches ** at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

-12 feet 7 inche3** at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

Two suppression chamber water level instrumentation channels OPERABLE hV by performance of a:

3.

CHANNEL CliECK 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 por 31 says, and 3.

CPANNEL CALIBRATION at least on-er 18 months, with the low water level alarm setpoint at greater than or equal to

-3 i rii.he s. *

4.5.3.2 With the suppression chambe~ level less than the above limit in OPERA-3 TIONAL CONDITION 5*, at least onc

r 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify footnote conditions

to s

be sat.,fied.

f I

The suppression chamber is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppres-r sion ouol, the spent fuel pool gates are removed when the cavity is flooded, and i

the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

- Level is referenced +o a plant elevation of 699 feet 11 inches (See Figui e B 3/4.6.2-1).

LA SALLE - UNIT 2 3/4 5-9 Amendment No. BS, 65 I

+

TABLE 3.6.3-1 (Continued) l g

' PRIMARY CONTAINMENT ISOLATION VALVES r-E VALVE FUNCTION AND NUMBER Other Isolation/alves (Continued)

[

4.

Low Pressure Core Spray' System 2E21-F005, 2E21-F001'3) 2E21-F012(3) d)

2E21-F011($)

2E21-F010(

2E21-F031(3) 2E21-F006(k) 3 5.

High Pressure Core Spray System M

2E22-F004(I) 2 D F015(5)

T 2E22-F023 2E22-F012((I)

M 2Ez2-F014(3) k)

2E22-F005 t

6.

Reactor Core Isolation Coolir.; System 2E51-F013 2E51-F069 2E51-F028 2E51-F068 F

2E51-F040(O E

2E51-F031(31 2E51-F019 2E51-F065(k) 2E51-F066(k)

'+

'2E51-F059(*)

2E51-F022(*)

~

' 2E51-F362('")

2E51-F363(;)

TABLE 3.6.3-1 (Continued).

g PRIMARY CONTAINMENT ISOLATION VALVES'..

[

VALVE FUNCTION AND NUMBER Other' Isolation Valves (Continued)

E 7.

Post LOCA Hydrogen Control-

.HG001A, B N

2HG002A, B 2HG005A, B 2HG006A, B 8.

Standby Liquid Contrc: System 2C41-F004A, B 2C41-F007 9.

Reactor Recirculation Seal Injection cn 2833-F013A,B(i) f

/>

d) 2833-F017A, B Drywell Pneumatic Valves 2IN018 b

a

.??

e i

6 >

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES E

TABLE' NOTATIONS e

E

But 2 3 seconds.

Z (a) See Specification 3.3.2, Table 3.3.2-1, for isolation signal (s) that operates each valve group.

(b)- Not included in total sum of: Type B and C tests.

m (c) May be opened on an intermittent basis under administrative control.

(d) Not closed by SLCS actuation.

(e) Not closed by Trip Functions Sa, b, or c, Specification 3.3.2, Table 3.3.2-1.

~(f) Nat closed by Trip Functions 4a, c, d, e,'or f of Specification 3.3.2, Table 3.3.2-1.

(g) Not subject to Type C leakage test.

(h) Opens on an isolation signal.

Valves will be open during Type A test.

No Type C test required.

(i) Also closed by drywell pressure-high signal.

(j) Hydraulic leak test at 43.6 psig.

(k) Not subject to Type C leakage test - leakage rate tested per Specification 4.4.3.2.2.

(1) These penetrations.are provided with removable spools outboard of the outboard isolation valve.

w D

During operation, ther-lines will be blind flanged using a double 0-ring and a type B leak en test.

In addition, the packing of these isolation valves will be soap-bubble tested to ensure O

insignificant or no leakage at the containment test pressure each refueling outage.

Ef (m) If valves 2E51-F362 and 2E51-F363 are locked closed and acceptably leak rate tested, then valves 2E51-F059 and 2E51-F022 are not considered to be

{

primary containment isolation valves and are not required to be leak rate tested.

(n) Either the 2E51-F362 or the 2E51-F363 valve may be open when the RCIC system is in the standby mode of operation, and both valves may be open r

during operation of the RCIC system in the full flow test mode, providing that:

k 1) valve 2E51-F022 is acceptably leak rate tested, and E

2) valve 2E51-F059 is deactivated, locked closed and acceptably leak E

rate tested, and 3) the spectacle flange, installed immediately downstream of the g

2E51-F059 valve, is closed and acceptably leak rate tested.

.~

CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF LIMITING CONDITION FOR OPERATION 3.6.4 All suppression chamber - drywell vacuum breakers shall be OPERABLE 'and closed.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With one suppression chamber - drywell vacuum breaker inoperable and/or open, within-4 hours 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 OPERABLE 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 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 position indicator of any OPERABLE suppression chamber -

drywell vacuum breaker inoperable, restore the inoperable position indicator to OPERABLE status within 14 days or visually verify the vacuum 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 - drywell vacuum breaker shall be:

a.

Verified closed at least once per 7 days, b.

-Demonstrated OPERABLE:

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 of steam to the suppression chamber from the safety-relief valves, by cycling each vacuum breaker through at least one complete cycle of full trcvel.

2.

At least once per 31 days by verify: ] both position indicators OPERABLE by performance of a CHANNEL FUNCTIONAL TEST.

3.

At least once per 18 months by;

-a)

Verifying the force required to open the vacuum breaker, from the closed position, to be less than or equal to 0.5 psid, and b)

Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.

)

LA SALLE - UNIT 2 3/4 6-38

'~

..:s TABLE 3.d.3.3-1 (Continued)

MOTOR OPERATED YALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICE SYSTEM (S)

VALVE NUMBER (Continuous)(f.ccident Conditions)

AFFECTED n-2E32 - F003N Accident Conditions 2E32 - F006 Accident Conditions 2E32 - F007 Accident Conditions 2E32 - F008 Accident Conditions 2E32 - F009

-Accident Conditions m.

2E22 - F004 Accident Cond.tions HPCS system 2E22 - F012 Accident Cenditions 2E22 - F015 Continuous 2E22 - F023 Accident Conditions LASALLE i -UNIT 2 3/4 8-30 Amendment No.

25, 65

...~

,=

.?

3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTOOWN ECCS Division 1 consists of the low pressure' core spray system, low pres-sure coolant injection subsystem "A" of the RHR system, and the automatic de-presturization system (ADS) as actuated by ADS trip system "A".

ECCS Division 2 consistsoflowpressurecoolantinjectionsubsystems"B"and"C"oftheRHRsys-tem and the automatic depressurization system as actuated by ADS trip system B".

The low oressure core spray (LPCS) system is provided to assure that the core is adequately 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 transients or smaller breaks following depressurization by the ADS.

The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that the LPCS sys-tem will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a-complete functional test requires reac w shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressura coolant injection (LPCI) mode of the RHR system is pro-vided to assure that the ccre is adequately cooled following a loss-of-coolant accident.

Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double ended reactor recirculation line break, and for transients or small breaks following depressurization by the A05.

The surveillance requirew ats provide adequate assurance that the LPCI system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete-functional test requires r m tor shutdown.

The-pump discharge piping is maintained 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 system.

The high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.

The HPCS system i

permits'the reactor to be shut down while maintaining sufficient reactor vessal 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.

The capacity of the HPCS syste;n is selectri to trovide the required core cooling.

The HPCS pump is designed to deliver greater than or equal to 516/1550/6200 gpm at differential pressures of 1160/1130/200 psid.

Water is taken from the suppression pool and injected into the reactor.

LA SALLE - UNIT 2_

B 3/4 5-1 Amendment No. 27, 65

0-EMERGENCY CORE COOLING SYSTEMS BASES ECCS-0PERATING and SHUTDOWN (Continued) l With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the hazards analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition.

The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactoc vesst:,

injection requires recctor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCS system to function properly, if required, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cuoling systems can enter tne core in time to limit fuel cladding temperature to-less than 2200 F.

ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 122 psig even though low pressure core cooling systems provide adequate core cooling up to 350 psig.

ADS automatically controls seven selected safety-relief valves.

Six valves are required to be OPERABLE since the LOCA analysis assumes 6 ADS valves in addition to a single failure.

It is therefore appropriate to permit one of the required valves to be out-of-service for up to 14 days without materially reducing system reliability.

3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is alsc required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of-a LOCA.

This limit on suppression chamber minimum wattc.* volume ensures that sufficient water is available to permit recirculation cooling flow to the core-(See Figure B 3/4.6.2-1).

The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.2.1.

Repair nrk might require making the suppression chamber inoperable.

This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply

when the suppression chamber must be made inoperable in OPERATIONAL CONDITION 4 or 5.

In-0PERATIONAL CONDITION 4 and b the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained 6t or below 200 F.

Since pressure suppression is not required below 212 F, the minimum water volume is based on NPSH, recirculation volume, vortex prevention plus a 2'-4" safety margin for conservatism.

LA SALLE - UNIT 2 B 3/4 5-2 Amendment No. O s 65

Control Suppression Plant Room / Local Chamber Level Elevation Indication 26' 10" 700' 2"

+3" High Level LC0 (Volume 3

131,900 ft )

26' 9" 700' 1"

+2" High Level Alarm 26' 7" 699' 11" 0" Instrument Zero 26' 4" 699' 8"

-3" Low Level Alarm 26' 2 '/2" 699' 6 1/2"

-4 1/2" Low Level LCO Opera-tional Condition 1, 2, 3 or 3 (Volume 128,800 ft )

14' c'

687' 4"

-12' 7" Low Level LCO Opera-tional Condition 4 or 5 3

(Volume 70,000 ft )

SUPPRESSION POOL LEVEL SETPOINTS BASES FIGURE B 3/4.6.2-1 LA SALLE UNIT 2 8 3/4 6-3a AMENDMENT N0. 39, 65

ML20086H621

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