ML20101P841
| ML20101P841 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/04/1996 |
| From: | Skay D NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20101P847 | List: |
| References | |
| NUDOCS 9604100005 | |
| Download: ML20101P841 (18) | |
Text
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UNITED STATES y=
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20086 4001
\\...../
COMONWEALTH EDISON COMPANY DOCKET NO. 50-373 j
LASALLE COUNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 111 License No. NPF-11 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by the Commonwealth Edison Company (the licensee), dated January 18, 1996, as supplemented by letters dated March 1, March 22, March 26, and April 3, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Comission; 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 Comission's regulations set forth in 10 CFR Chapter I; j
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; i
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 Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:
4 9604100005 960404 PDR ADOCK 05000373 P
. (2)
Technical Snecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.111, and the Environmental Protection Plan 1
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 and shall be implemented prior to startup from refueling outage LIR07.
FOR THE CLEAR REGULATORY COMISSION x
I 7). hu'^
ga., 7 Donna M. Skay, Proje d 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:
April 4, 1996
vai aj Q
ATTACHMENT TO LICENSE AMENDMENT NOS. 111 FACILITY OPERATING LICENSE NO. NPF-11 l
DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Technical Specifications with l
the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change.
REMOVE INSERT l
3/4 3-11 3/4 3-11 3/4 3-14a 3/4 3-14a 3/4 3-15 3/4 3-15 l
3/4 3-18 3/4 3-18 3/4 3-20 3/4 3-20 8 3/4 3-2 B 3/4 3-2 1
l 0
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 CONTAIMENT SOLATION a.
Low, level 3 7
2 1, 2, 3 20 f
Low Low, Level 2 2, 3 2
1,2,3 20 f,
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.
Radiation - High 1
2 1,2,3 21 3
2 1,2,3 22 2
Pressure - Low 1
2 1
23 3
Flow - High 1
2/line")
1, 2, 3 21 d.
DELETED l
e.
Main Steam Line Tunnel 1",3(33 2""l)
ATemperature - High 1
2
,3,33 f.
Condenser Vacuum - Low 1
2 1, 2*, 3*
21 2.
SECONDARY CONTAINMENT ISOLATION Reactor Building Vent Exhaust a.
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(*"*)
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. 111 L
- h TABLE 3.3.2-1 (Continued)
NOTES (Continued)
(g) Requires RCIC steam supply pressure-low coincident with drywell pressure-high.
(h) Manual initiation isolates IE51-F008 only and only with a coincident reactor vessel water level-low, level 2, signal.
(i) Both channels of each trip system may be placed in an inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required reactor building ventilation system corrective maintenance, filter changes, damper cycling and surveillance tests, other than Surveillance Requirement 4.6.5.1.c, without placing the trip system in the tripped condition.
(j)
Both channels of each trip system may be placed in an inoperable status for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to loss of reactor building ventilation or for performance of Surveillance Requirement 4.6.5.1.c without placing the trip system in the tripped condition.
i 1
LA SALLE - UNIT 1 3/4 3-14a Amendment No.
111
TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE A.
AUTOMATIC INITIATION 1.
PRIMARY CONTAllqEUT ISOLATION
- p.
a.
Low, Level 3 2 12.5 inches
- 2 11.0 inches
- 2)
Low Low, Level 2 2 -50 inches
- 1 -57 inches
- 3)
Low Low Low, Level 1 2 -129 inches
- 1 -136 inches
- b.
Drywell Pressure - High s 1.69 psig s 1.89 psig c.
Radiation - High s 3.0 x full power background s 3.6 x full background 2)
Pressure - Low 2 854 psig 1 834 psig 3)
Flow - High 5 111 psid s 116 psid d.
DELETED e.
Main Steam Line Tunnel l
A Temperature - High 5 65'F s 70*F f.
Condenser Vacuum - Low
> 7 inches Hg vacuum
> 5.5 inches Hg vacuum 2.
SECONDARY CONTAIPMENT ISOLATION a.
Reactor Building Vent Exhaust Plenum Radiation - High 5 10 nr/hr s 15 mr/hr 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/hr s 15 mr/hr 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High 5 70 gpa 5 87.5 gpa b.
Heat Exchanger Area Temperature
- High s 181*F
$ 187*F c.
Heat Exchanger Area Ventilation AT - High 5 85*F s 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. 111
k TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
A.
AUTOMATIC INITIATION 1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level-1)
Low, Level 3 N/A 2)
Low Low, Level 2 N/A 3)
Low Low Low, level 1 s 1.0*
b.
Drywell Pressure - High N/A c.
Radiation - High'**)
s 1.0*
2)
Pressure - Low s 2.0*
3)
Flow - High s 0.5*
d.
DELETED c.
Condenser Vacuum - Low N/A 1
f.
Main Steam Line Tunnel ATemperature - High N/A 2.
SECONDARY CONTAINMENT ISOLATION N/A a.
Reactor 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 a.
RHR Equipment Area ATemperature - High b.
RHR Area Cooler Temperature - High c.
RHR Heat Exchanger Steam Supply Flow High LA SALLE - UNIT 1 3/4 3-18 Amendment No. 111
TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIRENENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST
[ALIBRATIQN SURVEILLANCE REQUIRED A.
AUTOMATIC INITIATION 1.
PRIMARY CONTAllMENT ISOLATION a.
Low, level 3 S
Q R
1, 2, 3 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.
Radiation - High S
Q R
1, 2, 3 2)
Pressure - Low NA Q
Q l
3)
Flow - High NA Q
R 1, 2, 3 i
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 SYSTEN ISOLATION a.
A Flow - High S
Q R
1, 2, 3 f
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. 111 m
sne
INSTR 181ENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This' specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERA 8ILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Both channels of each trip system for the main steam tunnel ventilation system differential temperature may be placed in an l
inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required reactor building ventilation system maintenance and testing and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to loss of reactor building ventilation or for the required secondary containment Leak Rate test without placing the trip system in the tripped condition. This will allow for maintaining the reliability of the ventilation system and secondary containment.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, Supplement 2 " Technical Specification Improvement Analyses for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation", March 1989, and with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", July 1990. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains primary containment isolation capability. Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct +
bearing on safety, are established at a level away from the normal operating}*
range to prevent inadvertent actuation of the systems involved.
Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected.
For A.C. operated valves, it is assumed that the A.C.
power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 13 seconds is assumed before the valve starts to move. The safety analysis considers an allowable inventory loss which in turn determines the valve speed in conjunction with the 13 second delay.
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERASILITY requirements, trip setpoints and response times that will ensure effectiveness 7ef the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send'the actuation signal to more than one system at the same time.
Specified surveillance intervals and surveillance and maintenance outage times save been determined in accordance with NEDC-30936P-A, " Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)", Parts 1 and 2, December 1988, and RE-025 Revision 1, " Technical Specification Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for LaSalle County Station, Units 1 and 2", April 1991. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains ECCS initiation capability.
LA SALLE - UNIT 1 B 3/4 3-2 Amendment No. 111
ts49 51 UNITED STATES g
g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001
/
t COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 96 License No. NPF-18 l
1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment filed by the Comonwealth Edison Company (the licensee), dated January 18, 1996, as supplemented by letters dated March 1, March 22, March 26, and April 3, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), 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 Comission; 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 Comission'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 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 lC}
. i (2)
Tarknical Snecifications and Environmental Protection Plan i
The Technical Specifications contained in Appendix A, as revised through Amendment No. 96, 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 and shall be implemented prior to startup from refueling outage L2R07.
FOR THE NUCLEAR REGULATORY COMISSION W
d6 l
.. Un Donna N. Skay, Project ager Project Directorate 1 2
Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 4, 1996
g l
o ATTACHMENT TO LICENSE AMENDMENT NO. 96 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 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.
REMOVE INSERT 3/4 3-11 3/4 3-11 3/4 3-14a 3/4 3-14a 3/4 3-15 3/4 3-15 3/4 3-18 3/4 3-18 3/4 3-20 3/4 3-20 B 3/4 3-2 B 3/4 3-2
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.
Low, Level 3 7
2 1,2,3 20 J
Low Low, level 2 2, 3 2
1,2,3 20 d
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.
Radiation - High 1
2 1, 2, 3 21 3
2 1,2,3 22 2'i Pressure - Low 1
2 1
23 3ll Flow - High 1
2/line")
1, 2, 3 21 d.
DELETED e.
Main Steam Line Tunnel 1"),(l3 d >2""l)
)
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(*"*)
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. 96 is
w TABLE 3.3.2-1 (Continued)
NOTES (Continued)
(g) Require's RCIC steam supply pressure-low coincident with drywell pressure-high.
(h) Manual initiation isolates 2E51-F008 only and only with a coincident reactor vessel water level-low, level 2, signal.
1 (1) Both channels of each trip system may be placed in an inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required reactor building ventilation system corrective l
maintenance, filter changes, damper cycling and surveillance tests, other l
than Surveillance Requirement 4.6.5.1.c, vsthout placing the trip system in j
the tripped condition.
(j)
Both channels of each trip s., stem may be placed in an inoperable status for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to loss of reactor building ventilation or for' l
performance of Surveillance Requirement 4.6.5.1.c without placing the trip l
system !n the tripped condition.
l 4
s k
I l
l l
l 28u ih l
LA SALLE - UNIT 2 3/4 3-14a Amendment No. 96
TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE A.
AUTOMATIC INITIATION 1.
PRIMARY CONTAINMENT ISOLATION ReactorVessklWaterLevel a.
1)
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.
Radiation - High s 3.0 x full power background s 3.6 x full background 2)
Pressure - Low 2 854 psig 2 834 psig 3)
Flow - High s 111 psid s 116 psid d.
DELETED e.
Main Steam Line Tunnel A Temperature - High s 65'F s 70*F l
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 AT - 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. 96 m-~
TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION -
RESPONSE TIME (Seconds)#
A.
AUTOMATIC INITIATION 1.
PRIMARY CONTAINMENT ISOLATION a.
Low, Level 3 N/A 2)
Low Low, Level 2 N/A 3)
Low Low Low, Level 1 s 1.0*
b.
Drywell Pressure - High N/A c.
Radiation - High'**)
s 1.0*
2)
Pressure - Low s 2.0*
3)
Flow - High s 0.5*
d.
DELETED l
e.
Condenser Vacuum - Low N/A f.
Main Steam Line Tunnel ATemperature - High N/A 2.
SECONDARY CONTAINMENT ISOLATION N/A a.
Reactor 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 i
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 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. 96
I l
TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIRENENTS CHANNEL OPERATIONAL l
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED A.
AUTOMATIC INITIATION 1.
PRINARY CONTAllRENT ISQLATION 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.
Radiation - High S
Q R
1, 2, 3 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 CONTAIMENT 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 SYSTEN 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 2 3/4 3-20 Amendment No. 96 g
N i
INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Both channels of each trip system for the main steam tunnel ventilation system differential temperature may be placed in an l
inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required reactor building ventilation system maintenance and testing and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to loss of reactor building ventilation or for the required secondary containment Leak Rate test without placing the trip system in the tripped condition. This will allow for maintaining the reliability of the ventilation system and secondary containment.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, Supplement 2, " Technical Specification Improvement Analyses for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation", March 1989, and with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", July 1990. When a channel is placed la an inoperable status solely for performance of required surveillances, entryt into LCO and required ACTIONS may be delayed, provided the associated function maintains primary containment isolation capability.
Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety.
The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected.
For A.C. operated valves, it is assumed that the A.C.
power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 13 seconds is assumed before the valve starts to move.
The safety analysis considers an allowable inventory loss which in turn determines the valve speed in conjunction with the 13 second delay.
3/4.3.3 EF"NCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
LA SALLE - UNIT 2 B 3/4 3-2 Amendment No. 96