ML20117P657

a.

Neutron Flux - High 2

3 1

3 4 2

2 Eb S'

3 3

b.

Inoperative 2

3 1

3, 4 2

2 5

3 3

2.

Average Power Range Monitor:")

a.

Neutron Flux - High, Setdown 2

2 1

3 2

2 5*)

2 3

b.

Flow Biased Simulated Thermal Power-Upscale 1

2 4

j c.

Fixed Neutron Flux-High 1

2 4

d.

Inoperative 1, 2 2

1 t

3 2

2 l

5 2

3 l

3.

Reactor Vessel Steam Dome Pressure - High 1, 2")

2 1

f 4.

Reactor Vessel Water Level - Low, Level 3 1, 2 2

1 5.

Main Steam Line Isolation Valve -

Closure 1")

4 4

6.

DELETED LA SALLE - UNIT 1 3/4 3-2 Amendment No. 115

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1

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

f 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 CORE 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

- Deleted ACTION 6

- Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to s 140 psig, equivalent to THERMAL i

POWER 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 .

ACTION 7

- Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8

- Lock the reactor mode switch in the Shutdown position within I hour.

ACTION 9

- Suspend all operations involving CORE ALTERATIONS,* and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within I hour.

Except movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.

LA SALLE - UNIT 1 3/4 3-4 Amendment No. 115

TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds) 1.

Neutron Flux

-High*

NA a.

b.

Inoperative NA 2.

Average Power Range Monitor

a.

Neutron Flux - High, Setdown NA b.

Flow Biased Simulated Thermal Power-Upscale 5 0.09**

c.

Fixed Neutron Flux - High 5 0.09 d.

Inoperative NA 3.

Reactor Vessel Steam Dome Pressure - High

< 0. 55

4.

Reactor Vessel Water Level - Low, level 3 5 1.0 5*'

5.

Main Steam Line Isolation Valve - Closure 5 0.06 6.

Deleted 7.

Primary Containment Pressure - High NA 8.

Scram Discharge Volume Water Level - High NA 9.

Turbine Stop Valve - Closure 5 0.06

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low

$ 0.08#

11. Reactor Mode Switch Shutdown Position NA

12. Manual Scram NA 13.

Control Rod Drive a.

Charging Water Header Pressure - Low NA b.

Delay Timer NA

Neutron detectors are exempt from response time testing.

Response time shall be measured ~ from the detector 1

output or from the input of the first electronic component in the channel.

- Not including simulated thermal power time constant.

Measured from start of turbine control valve fast closure.

1. Sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.

LA SALLE - UNIT 1 3/4 3-6 Amendment No. 115

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEN INSTRUNENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION'*)

SURVEILLANCE REQUIRED 1.

Intermedlate Range Monitors a.

Neutron Flux - High S/U(*), S S/U(*), W R

2*

S W

R 3*,

4, 5 b.

Inoperative NA W

NA 2*, 3*,

4, 5 2.

Average Power Range Monitor:")

a.

Neutron Flux - High, Setdown S/U(b),S S/U(*), W SA 2*

S W

SA 3*,

5 b.

Flow Biased Simulated Thermal Power-Upscale S, D("

S/U(*), Q W(d"", SA, R 1

N c.

Fixed Neutron Flux -

High S

S/U(*), Q W(d), SA 1

d.

Inoperative NA Q

NA 1, 2, 3, 5 3.

Reactor Vessel Steam Dome Pressure - High NA Q

Q 1, 2 4.

Low, level 3 NA Q

R 1, 2 5.

Main Steam Line Isolation Valve - Closure NA Q

R 1

l 6.

Deleted 7.

Primary Containment Pressure -

High NA Q

Q 1, 2

\\

LA SALLE - UNIT 1 3/4 3-7 Amendment No. 115

TABLE 3.3,2-1 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CONDITION ACTION A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level Low, Level 3 7

2 1,2,3 20 Low Low, Level 2 2, 3 2

1,2,3 20 Low Low Low, Level 1 1, 10 2

1,2,3 20 b.

Drywell Pressure - High 2, 7, 10 2

1,2,3 20 c.

Main Steam Line l

1 DELETED 2

Pressure - Low 1

2 1

23 3

Flow - High 1

2/line")

1, 2, 3 21 d.

DELETED e.

Main Steam Line Tunnel 0 33 1,3[3,3 2""3)

ATemperature - High 1

2 f.

Condenser Vacuum - Low 1

2 1, 2*, 3*

21 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Vent Exhaust Plenum Radiation - High 4""

2 1, 2, 3 and **

24 b.

Drywell Pressure - High 4""*)

2 1, 2, 3 24 c.

Reactor Vessel Water Level - Low Low, level 2 4""*3 2

1, 2, 3, and "

24 d.

Fuel Pool Vent Exhaust Radiation - High 4""*)

2 1, 2, 3, and **

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

. Amendment No. 115

TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION a.

Low, Level 3 2 12.5 inches

2 11.0 inches

2)

Low Low, Level 2 2 -50 inches

2 -57 inches

3)

Low Low Low, level 1 2 -129 inches

2 -136 inches

b.

Drywell Pressure - High

$ 1.69 psig 1 1.89 psig c.

DELETED 2)

Pressure - Low 2 854 psig 2 834 psig 3)

Flow - High I 111 psid s 116 psid d.

DELETED e.

Main Steam Line Tunnel tL Temperature - High s 65*F s 70*F f.

Condenser Vacuum - Low

> 7 inches Hg vacuum

> 5.5 inches Hg vacuum 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Vent Exhaust Plenum Radiation - High

$ 10 mr/hr s 15 mr/hr b.

Drywell Pressure - High s 1.69 psig 5 1.89 psig c.

Reactor Vessel Water Level - Low Low, Level 2 2 -50 inches

2 -57 inches

d.

Fuel Pool Vent Exhaust Radiation - High s 10 mr/hr s 15 mr/hr 3.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

tL Flow - High 5 70 gpm 5 87.5 gpm b.

Heat Exchanger Area Temperature

- High 5 181*F s 187*F c.

Heat Exchanger Area Ventilation AT - High s 85*F 5 91*F d.

SLCS Initiation NA NA e.

Low Low, level 2 2 -50 inches

2 -57 inches

LA SALLE - UNIT 1 3/4 3-15 Amendment No. 115

f TABLE 3.3.2-3

o ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME j

f TRIP FUNCTION RESPONSE TIME (Seconds)#

}

A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAIMENT ISOLATION a.

Low, Level 3 N/A 2)

Low Low, Level 2 N/A l

3)

Low Low Low, Level 1 s 1.0*, ##

]

b.

Drywell Pressure - High N/A c.

DELETED

\\

2)

Pressure - Low

$ 2.0*, ##

3)

Flow - High s 0.5*, ##

j d.

DELETED i

e.

Condenser Vacuum - Low N/A f.

Main Steam Line Tunnel ATemperr.ture - High N/A

{

2.

SECONDARY CONTAINMENT ISOLATION N/A I

a.

Reactor Building Vent Exhaust Plenum Radiation - High i

b.

Drywell Pressure - High c.

Reactor Vessel llater Level - Low, Level 2 i

d.

Fuel Pool. Vent Exhaust Radiation - High i.

l 3.

REACTOR WATER CLEANUP SYSTEM ISOLATION N/A a.

AFlow - High b.

Heat Exchanger Area Temperature - High I

c.

Heat Exchanger Area Ventilation AT-High d.

SLCS Initiation e.

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

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION N/A 1

a.

RCIC Steam Line Flow - High b.

RCIC Steam Supply Pressure - Low c.

RCIC Turbine Exhaust Diaphragm Pressure - High d.

RCIC Equipment Room Temperature - High 4

e.

RCIC Steam Line Tunnel Temperature - High f.

_RCIC Steam Line Tunnel ATemperature - High g.

Drywell Pressure - High 4

h.

RCIC Equipment Room ATemperature - High 5.

RHR SYSTEM STEAM CONDENSING MODE ISOLATION N/A a.

RHR Equipment Area ATemperature - High b.

RHR Area Cooler Temperature - High c.

RHR Heat Exchanger Steam Supply Flow - High 4

LA SALLE - UNIT 1 3/4 3-18 Amendment No. 115

5 TABLE 3.3.2-3 (C:ntinu;d)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME 4

TRIP FUNCTION RESPONSE TIME (Seconds)*

]

6.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION N/A a.

Reactor Vessel Water Level - Low, Level 3 b.

Reactor Vessel (RHR Cut-In Permissive) Pressure - High c.

RHR Pump Section Flow - High d.

RHR Area Cooler Temperature - High e.

RHR Equipment Area AT - High B.

MANUAL INITIATION N/A i

1.

Inboard Valves 2.

Outboard Valves 3.

Inboard Valves l

4.

Outboard Valves 5.

Inboard Valves 6.

Outboard Valves 7.

Outboard Valve J

TABLE NOTATIONS Isolation system instrumentation response time for MSIVs only.

No diesel generator delays assumed.

Isolation system instrumentation response time specified for the Trip Function actuating the MSIVs shall be added to MSIV isolation time to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

Sensor is eliminated from response time testing for the MSIV actuation logic circuits. Response time testing and conformance to the admi'sistrative limits for the remaining channel including trip unit and re'.ay logic are required.

l 1

N/A Not Applicable.

4 LA SALLE - UNIT 1 3/4 3-19 Amendment No. 115

TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION a.

Low, Level 3 5

Q R

1, 2, 3 2)

Low Low, Level 2 NA Q

R 1, 2, 3 3)

Low Low Low, Level 1 S

Q R

1, 2, 3 b.

Drywell Pressure - High NA Q

Q 1, 2, 3 c.

Main Steam Line i

1)

DELETED l

2)

Pressure - Low NA Q

Q l

3)

Flow - High NA Q

R 1, 2, 3 d.

DELETED e.

Condenser Vacuum - Low NA Q

Q 1, 2*, 3*

f.

Main Steam Line Tunnel A Temperature - High NA Q

R 1, 2, 3 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Vent Exhaust Plenum Radiation - High S

Q R

1, 2, 3 and **

b.

Drywell Pressure - High NA Q

Q 1,2,3 c.

Reactor Vessel Water Level - Low Low, level 2 NA Q

R 1, 2, 3, and

d.

Fuel Pool Vent Exhaust Radiation - High S

Q R

1, 2, 3 and **

3.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High S

Q R

1, 2, 3 b.

Heat Exchanger Area Temperature - High NA Q

Q 1, 2, 3 c.

Heat Exchanger Area Ventilation AT - High NA Q

Q 1, 2, 3 d.

SLCS Initiation NA R

NA 1, 2, 3 e.

Reactor Vessel Water Level - Low Low, level 2 NA Q

R 1, 2, 3 LA SALLE - UNIT 1 3/4 3-20 Amendment No. 115

e t @ CEGO fi UNITED STATES 1

E NUCLEAR REGULATORY COMMISSION f

WASHINGTON, D.C. 20645 0001 s...../

COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 100 License No. NPF-18 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee), dated April 9, 1996, complies with the standcris and requirements of the Atomic Energy Act of 1954, as a:nenden (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;

)

B.

The facilsty will operate in conformity with the appl h ' ion, the provisions of the Act, and the regulations of the CE. :;s ion; C.

There is reasonable assurance: (i) that the activities authorized by 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 Commission's regulations set forth in 10 CFR Chapter I; a

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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:

l

\\

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 100, 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.

l 3.

This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION

. p-y-L

.-- D. j{ec, v" g4 Donna M. Skay, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications Date of Issuance: September 20, 1996 j

ATTAC'JMENT TO LICENSE AMENDMENT NO. inn f821LITY OPERATING LICENSE NO. NPF-18 l

DOCKET NO. 50-374 I

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

Pages indicated by an l

asterisk are provided for convenience.

REMOVE INSERT 2-4 2-4 B 2-11 B 2-11

B 2-12

B 2-12 3/4 3-2 3/4 3-2 3/4 3-4 3/4 3-4 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-11 3/4 3-11 3/4 3-15 3/4 3-15 3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 i

l

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWA8LE FUNCTIONAL UNIT TRIP SETPOINT VALUES 1.

Interwediate Range Monitor, Neutron Flux-High s 120 divisions of s 122 divisions full scale of full scale 2.

Average Power Range Monitor:

a.

Neutron Flux-High, Setdown s 15% of RATED THERMAL s 20% of RATED POWER THERMAL POWER b.

Flow Biased Simulated Thermal Power - Upscale 1)

Two Recirculation Loop Operation a) Flow Biased s 0.58W + 59% with a 5 0.58W + 62% with a maximum of a maximum of b) High Flow Clamped s 113.5% of RATED s 115.5% of RATED THERMAL POWER THERMAL POWER 2)

Single Recirculation Loop Operation a) Flow Biased s 0.58W + 54.3% with 5 0.58W + 57.3%

a maximum of with a maximum of b) High Flow Clamped s 113.5% of RATED s 115.5% of RATED THERMAL POWER THERMAL POWER c.

Fixed Neutron Flux-High s 118% of RATED s 120% of RATED i

THERMAL POWER THERMAL POWER 3.

Reactor Vessel Steam Dome Pressure - High s 1043 psig s 1063 psig 4.

Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above 2 11 inches above instrument zero*

instrument zero*

i 5.

Main Steam Line Isolation Valve - Closure s 8% closed s 12% closed 6.

DELETED 7.

Primary Containment Pressure - High s 1.69 psig s 1.89 psig 8.

Scram Discharge Volume Water Level - High s 767' Sk" s 767' 5%"

9.

Turbine Stop Valve - Closure s 5% closed s 7% closed

See Bases Figure B 3/4 3-1.

IA RAl l r IINTT 7

?_A AmonAmont Nn ihn

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 4.

Reactor Vessel Water Level-Low i

The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure l

limits.

I 5.

Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIV's are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature and low steam line pressure. The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.

6.

DELETED 7.

Primary Containment Pressure-Hiah High pressure in the drywell could indicate a break in the primary pres-sure boundary systems. The reactor is tripped in order to minimize the possi-bility of fuel damage and reduce the amount of energy being added to the coolant. The trip setting was selected as low as possible without causing spurious trips.

LA SALLE - UNIT 2 B 2-11 Amendment No. 100

e LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECT 0N SYSTEM INSTRUMENTATION SETPOINTS (Continued) 8.

Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered.

The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great i

enough to accommodate the water from the movement of the rods at pressures i

below 65 psig when they are tripped.

9.

Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.

With a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained.

i l

10.

Turbine Control Valve Fast Closure, Trip Oil Pressure-Low

)

The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failures of the turbine bypass va?ves.

The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting sole-noid valves and in less than 30 milliseconds after the start of control valve fast closure.

This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to I

the Reactor Protection System.

This trip setting, a nominally 50% greater closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve.

Relevant transient analyses are discussed in Section 15.1.0 of the Final Safety Analysis Report.

11.

Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional' manual reactor trip capability.

12.

Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

LA SALLE - UNIT 2 B 2-12

TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERABLE OPERATIONAL CHANNELS PER FUNCTIONAL UNIT CONDITIONS TRIP SYSTEM fa)

ACTION 1.

a.

Neutron Flux - High 2

3 1

3 4 2

2 I

S ")

3 3

b.

Inoperative 2

3 1

3, 4 2

2 5

3 3

2.

Average Power Range Monitor:(*)

a.

Neutron Flux - High, Setdown 2

2 1

3 2

2 5(6) 2 3

- b.

Flow Biased Simulated Thermal Power-Upscale 1

2 4

c.

Fixed Neutron Flux-High 1

2 4

d.

Inoperative 1, 2 2

1 3

2 2

5 2

3 3.

Reactor Vessel Steam Dome Pressure - High 1, 2'd) 2 1

4.

Reactor Vessel Water Level - Low, Level 3 1, 2 2

1 5.

Main Steam Line Isolation Valve -

Closure 1(')

4 4

6.

DELETED LA SALLE - UNIT 2 3/4 3-2 Amendment No. 1og

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 1 Verify all insertable control rods to be inserted in the core ACTION 2 and lock the reactor mode switch in the Shutdown position within I hour.

Suspend all operations involving CORE ALTERATIONS

and ACTION 3 insert all insertable control rods within one hour.

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

ACTION 4 ACTION 5 DELETED ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to s;140 psig, equivalent to THERMAL POWER less than 30% of RATED THERNAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

1 ACTION 7 Verify all insertable control rods to be inserted within I hour.

Lock the reactor mode switch in the Shutdown position within ACTION 8 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

]

l ACTION 9 Suspend all operations involving CORE ALTERATIONS,* and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within I hour,

Except movement of IRM, SRM, or special movable detectors, or. replacement i

of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.

LA SALLE - UNIT 2 3/4 3-4 Amendment No. 100

TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds) 1.

Intermediate Range Monitors a.

a.

Neutron Flux - High*

NA b.

Inoperative NA 2.

Average Power Range Monitor

i a.

Neutron Flux - High, Setdown NA b.

Flow Biased Simulated Thermal Power-Upscale s 0.09,,

c.

Fixed Neutron Flux - High 5 0.09 i

d.

Inoperative NA 3.

Reactor Vessel Steam Dome Pressure - High

< 0.55" 4.

Reactor Vessel Water Level - Low, Level 3 2 1.05*'

5.

Main Steam Line Isolation Valve - Closure s 0.06 l

6.

DELETED I

7.

Primary Containment Pressure - High NA 8.

Scram Discharge Volume Water Level - High NA 9.

Turbine Stop Valve - Closure s 0.06

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low s 0.08' 11.

Reactor Mode Switch Shutdown Position NA

12. Manual Scram NA t

13.

Control Rod Drive a.

Charging Water Header Pressure - Low NA b.

Delay Timer NA

Neutron detectors are exempt from response time testing.

Response time shall be measured from the detector output or from the input of the first electronic component in the channel.

- Not including simulated thermal power time constant.

Measured from start of turbine control valve fast closure.

1. Sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.

LA SALLE - UNIT 2 3/4 3-6 Amendment No. 10.0

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

~

CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)

SURVEILLAKE REQUIRED 1.

Neutron Flux - High S/U(b),S S/U(*), W R

2*

S W

R 3*, 4, 5 b.

Inoperative NA W

NA 2*, 3*,

4, 5 2.

Average Power Range Monitor:")

a.

Neutron Flux - High, Setdown S/U(b),5 S/U(*), W SA 2*

S W

SA 3*,

5 b.

Flow Biased Simulated Thermal Power-Upscale S, D("

S/U(*), Q W(d"*), SA, R*)

1 c.

Fixed Neutron Flux -

High S

S/U(*), Q W(d), SA 1

d.

Inoperative NA Q

NA 1,2,3,5 3.

Reactor Vessel Steam Dome Pressure - High NA Q

Q 1, 2 4.

Low, Level 3 S

Q R

1, 2 5.

Main Steam Line Isolation Valye - Closure NA Q

R 1

6.

DELETED 7.

Primary Containment Pressure -

High NA Q

Q 1, 2 LA SALLE - UNIT 2 3/4 3-7 Amendment No. 100

=

o TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CONDITION ACTION A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level Low, i.evel 3 7

2 1, 2, 3 20 Low Low, level 2 2, 3 2

1,2,3 20 Low Low Low, Level 1 1, 10 2

1,2,3 20 b.

Drywell Pressure - High 2, 7, 10 2

1,2,3 20 c.

Main Steam Line 1

DELETED l

2 Pressure - Low 1

2 1

23 3

Flow - High 1

2/line(*

1, 2, 3 21 d.

DELETED e.

Main Steam Line Tunnel ATemperature - High 1

2 1"',$3 33d>2""3),

f.

Condenser Vacuum - Low 1

2 1, 2*, 3*

21 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Vent Exhaust Plenum Radiation - High 4(*"

2 1, 2, 3 and **

24 b.

Drywell Pressure - High 4(*"

2 1, 2, 3 24 c.

Reactor Vess'el Water Level - Low Low, Level 2 4(*"')

2 1, 2, 3, and "

24 d.

Fuel Pool Vent Exhaust Radiation - High 4 ( * "')

2 1, 2, 3, and **

24 LA SALLE - UNIT 2 3/4 3-11 Amendment No.100

~.

a TABLE 3.3 2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS i

ALLOWA8LE TRIP FUNCTION TRIP SETPOINT VALUE A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION a.

Low, Level 3 2 12.5 inches

2 11.0 inches

2)

Low Low, Level 2 2 -50 inches

2 -57 inches

3)

Low Low Low, level 1 2 -129 inches

2 -136 inches

b.

Drywell Pressure - High s 1.69 psig s 1.89 psig c.

DELETED l

2)

Pressure - Low 2 854 psig 2 834 psig 3)

Flow - High 5 111 psid s 116 psid d.

DELETED i

e.

Main Steam Line Tunnel A Temperature - High s 65*F s 70*F f.

Condenser Vacuum - Low

> 7 inches Hg vacuum

> 5.5 inches Hg vacuum 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Vent Exhaust Plenum Radiation - High s 10 mr/h s 15 mr/h b.

Drywell Pressure - High s 1.69 psig s 1.89 psig c.

Reactor Vessel Water Level - Low Low, Level 2 2 -50 inches

2 -57 inches

d.

Fuel Pool Vent Exhaust Radiation - High s 10 mr/h s 15 mr/h

[

3.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

aFlow - High s 70 gpm s 87.5 gpm b.

Heat. Exchanger Area Temperature

- High s 181*F s 187*F c.

Heat Exchanger Area Ventilation AI - High s 85*

s 91*F d.

SLCS Initiation N.A.

N.A.

e.

Low Low, level 2 2 -50 inches

2 -57 inches

LA SALLE - UNIT 2 3/4 3-15 Amendment No.100

TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAIMENT ISOLATION a.

Low, Level 3 N/A l

2)

Low Low, Level 2 N/A 3)

Low Low Low, Level 1 s 1. 0** "

b.

Drywell Pressure - High N/A c.

Main Steam Line l

l 1)

DELETED I

2)

Pressure - Low s 2. 0* * "

3)

Flow - High s 0. 5** "

d.

DELETED e.

Condenser Vacuum - Low N/A l

f.

Main Steam Line Tunnel ATemperature - High N/A 2.

SECONDARY CONTAINMENT ISOLATION N/A a.

Rcactor Building Vent Exhaust Plenum Radiation - High b.

Drywell Pressure - High c.

Reactor Vessel Water Level - Low, Level 2 d.

Fuel Pool Vent Exhaust Radiation - High 3.

REACTOR WATER CLEANUP SYSTEM ISOLATION N/A a.

AFlow - High b.

Heat Exchanger Area Temperature - High c.

Heat Exchanger Area Ventilation AT-High d.

SLCS Initiation e.

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

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION N/A a.

RCIC Steam Line Flow - High b.

RCIC Steam Supply Pressure - Low c.

RCIC Turbine Exhaust Diaphragm Pressure - High d.

RCIC Equipment Room Temperature - High e.

RCIC Steam Line Tunnel Temperature - High f.

RCIC Steam Line Tunnel ATemperature - High g.

Drywell Pressure - High h.

RCIC Equipment Room ATemperature - High 5.

RHR SYSTEM STEAM CONDENSING MODE ISOLATION N/A l

a.

RHR Equipment Area ATemperature - High b.

RHR Area Cooler Temperature - High c.

RHR Heat Exchanger Steam Supply Flow High LA SALLE - UNIT 2 3/4 3-18 Amendment No.100

4 TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

4 6.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION N/A a.

Reactor Vessel Water Level - Low, Level 3 b.

Reactor Vessel (RHR Cut-In Pcruissive) Pressure - High c.

RHR Pump Suction Flow - High d.

RHR Area Cooler Temperature High e.

RHR Equipment Area AT High B.

MANUAL INITIATION N/A 1.

Inboard Valves 2.

Outboard Valves 3.

Inboard Valves 4.

Outboard Valves 5.

Inboard Valves 6.

Outboard Valves i

7.

Outboard Valve TABLE NOTATIONS Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed.

Isolation system instrumentation response time specified for the Trip Function actuating the MSIVs shall be added to MSIV isolation time to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

Sensor is eliminated from response time testing for the MSIV actuation logic circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.

N/A Not Applicable.

LA SALLE - UNIT 2 3/4 3-19 Amendment No.100

TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION a.

Low, Level 3 S

Q R

1, 2, 3 2)

Low Low, level 2 NA Q

R 1, 2, 3 3)

Low Low Low, level 1 S

Q R

1, 2, 3 b.

Drywell Pressure - High NA Q

Q 1, 2, 3 c.