ML20117A613
| ML20117A613 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/30/1985 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20117A615 | List: |
| References | |
| NUDOCS 8505080285 | |
| Download: ML20117A613 (22) | |
Text
O 1
[
UNITED STATES es I D7). t NUCLEAR REGULATORY COMMISSION 2-j 2%
WASHINGTON, D. C. 20555 qtv/e
- /
....+
COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-373 LA SALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment 22 License No. NPF-11 1.
The Nuclear Regulatory Comission (the Comission or the NRC) having found that:
A.
The application for amendment filed by the Comonwealth Edison Company, dated February 21, 1985, 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; 8.
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 attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 22 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 i
Technical Specifications and the Environmental Protection Plan.
$$$hy3 ADOC PDR
\\
l
3.
This amendment is effective as of November 30, 1985.
FOR THE NUCLEAR REGULATORY COMMISSION (M($$$V A. S'chwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
Changes to the Technical Specifications Date of Issuance: April 30,1985
1 3.
'This amendment is effective as of November 30, 1985.
FOR THE NUCLEAR REGULATORY COMMISSION A'. Schwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
Changes to the Technical
- Specifications Date of Issuance: April 30,1985 I
1 j
n 4
[
j
^~
e5 x
9)
- f,'
9 LBj(
/LA LB#2/DL/PM OEL
'LB#2/DL/BC'
/
[
E on ABournia:1b ASchwpncer" T
vak '
O'/l/85 04/if/85-04/gp/85 04/lg/85 04/ /85 l
ENCLOSURE TO LICENSE AMENDMENT N0. 22 FACILITY OPERATING LICENSE NO. NPF-ll DOCKET NO. 50-373 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 3-7 3/4 3-7 3/4 3-20 3/4 3-20 3/4 3-22 3/4 3-22 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-49 3/4 3-49 J
,e.-
,,, ~ -
TABLE 4.3.1.1-1 9
g REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH i
FUNCTIONAL UNIT CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED 1.
Intermediate Range Monitors a.
Neutron Flux - High S/U(b) S S/U(c),W R
2 S
W R
3,4,5 b.
Inoperative NA W
NA 2,3,4,5 2.
Average Power Range Monitor:f f) a.
Neutron Flux - High, Setdown S/U(b) S S/U(c),W SA 1, 2 S
W SA 3, 5 b.
Flow Biased Simulated Thermal w
f9)
S/U(c),y y(d)(e),SA,R(h)
)
A Power-Upscale S, O w
c.
Fixed Neutron Flux -
4 High 5
S/U(c),y y(d), SA 1
d.
Inoperative NA W
NA 1, 2, 3, 5 3.
Reactor Vessel Steam Dome Pressure - High NA M
Q 1, 2 4.
Low, Level 3 NA M
R 1, 2 l
S.
Main Steam Line Isolation Valve - Closure NA M
R 1
6.
Main Steam Line Radiation -
2 High S
M R
1, 2 E
R 7.
Primary Containment Pressure -
5 High NA M
Q 1, 2
TABLE _4.3.2.1-1 9
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL F
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 7
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED g A.
AUTOMATIC INITIATION M
1.
PRIMARY CONTAINMENT ISOLATION a.
Low, Level 3 NA M
R 1, 2, 3 2)
Low Low, Level 2 NA M
R 1, 2, 3 b.
Drywell Pressure - High NA M
Q 1,2,3 c.
Radiation - High S
M R
1, 2, 3 2)
Pressure - Low NA M
Q l
3)
Flow - High NA M
R 1,2,3 l
d.
Main Steam Line Tunnel Temperature - High NA M
R 1, 2, 3 w1 e.
Condenser Vacuum - Low NA M
Q 1, 2*, 3*
w f.
Main Steam Line Tunnel h
A Temperature - High NA M
R 1,2,3 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Vent Exhaust Plenum Radiation - High S
M R
1, 2, 3 and **
b.
Drywell Pressure - High NA M
Q 1,2,3 c.
Reactor Vessel Water y
Level - Low Low, Level 2 NA M
R 1, 2, 3, and l
d.
Fuel Pool Vent Exhaust Radiation - High S
M R
1, 2, 3 and **
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High S
M R
1,2,3 E
b.
Heat Exchanger Area Temperature - High NA M
Q 1, 2, 3 c.
Heat Exchanger Area Ventilation AT - High NA M
Q 1, 2, 3 d.
SLCS Initiation NA R
NA 1, 2, 3 I
e.
Reactor Vessel Water Level - Low Low, Level 2 NA M
R 1,2,3 l
m"
t TABLE 4.3.2.1-1 (Continued) 9 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS u,
N F-CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION S_URVEILLANCE REQUIRED c:i'i
-4 6.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION s
a.
Low, Level 3 NA M
R 1, 2, 3 l
b.
Reactor Vessel (RHR Cut-in Permissive)
Pressure - High NA M
Q 1, 2, 3 c.
RHR Pump Suction Flow - High NA M
Q 1, 2, 3 d.
RHR Area Temperature - High NA M
Q 1,2,3 e.
RHR Equipment Area AT - High NA M
Q 1,2,3 R> B.
MANUAL INITIATION i>
1.
Inboard Valves NA R
NA 1,2,3 pg 2.
Outboard Valves NA R
NA 1,2,3 3.
Inboard Valves NA R
NA 1, 2, 3 and **,#
4.
Outboard Valves NA R
NA 1, 2, 3 and **,#
5.
Inboard Valves NA R
NA 1,2,3 6.
Outboard Valves NA R
NA 1,2,3 7.
Outboard Valve NR R
NA 1,2,3 "When reactor steam pressure > 1043 psig and/or any turbine stop valve is open.
- When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
f
- During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
Br
TABLE 4.3.3.1-1 9
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS m?
E CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED c-5 A.
DIVISION I TRIP SYSTEM s
1.
RHR-A (LPCI MODE) AND LPCS SYSTEM a.
Low Low Low, Level 1 NA M
R 1, 2, 3, 4*, 5*
l b.
Drywell Pressure - High NA M
Q 1, 2, 3 c.
LPCS Pump Discharge Flow-Low NA M
Q 1, 2, 3, 4*, 5*
d.
LPCS and LPCI A Injection Valve Injection Line Pressure Low Interlock NA M
R 1, 2, 3, 4 *, 5*
5:'
e.
LCPS and LPCI A Injection Valve Reactor Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
Y f.
LPCI Pump A Start Time Delay M
Relay NA M
Q 1, 2, 3, 4*, 5*
g.
LPCI Pump A Flow-Low NA M
Q 1, 2, 3, 4*, 5*
i h.
Manual Initiation NA R
NA 1, 2, 3, 4*, 5*
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
a.
Low Low Low, Level 1 NA M
R 1, 2, 3 l
b.
Drywell Pressure-High NA M
Q 1, 2, 3 i
c.
ADS Timer NA M
Q 1,2,3 d.
Low, Level 3 NA M
R 1,2,3 l
e.
LPCS Pump Discharge l
y Pressure-High NA M
Q 1,2,3 f.
LPCI Pump A Discharge y
Pressure-High NA M
Q 1, 2, 3 g
g.
Manual Initiation NA R
NA 1,2,3 e
=
TABLE 4.3.3.1-1 (Continued) 9 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS us?
G; CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED c:
i5;j B.
DIVISION 2 TRIP SYSTEM 1.
Low Low Low, Level 1 NA M
R 1, 2, 3, 4*, 5*
l b.
Drywell 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*
d.
LPCI Pump B Start Time Delay u,
3:
Relay NA M
Q 1, 2, 3, 4*, 5*
e.
LPCI Pump Discharge Flow-Low NA M
Q 1,2,3,4*,5*
u, J,
f.
Manual Initiation NA R
NA 1, 2, 3, 4*, 5*
g.
LPCI B and C Injection Valve Reactor Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
a.
Low Low Low, Level 1 NA M
R 1,2,3 l
b.
Drywell Pressure-High NA M
Q 1,2,3 c.
ADS Timer NA M
Q 1, 2, 3 d.
Low, Level 3 NA M
R 1, 2, 3 l
jr e.
LPCI Pump B and C Discharge Pressure-High NA M
Q 1, 2, 3 j[
f.
Manual Initiation NA R
NA 1,2,3 5
.E
1 TABLE 4.3.3.1-1 (Continued)
I 5
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS u,
E l
G CHANNEL OPERATIONAL 1
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED g
]
C.
DIVISION 3 TRIP SYSTEM 1.
HPCS SYSTEM a.
Low Low, Level 2 NA M
R 1, 2, 3, 4*, 5*
l 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*
l d.
Condensate Storage Tank Level -
Low NA M
Q 1, 2, 3, 4*, 5*
R e.
Suppression Pool Water 4
Level - High NA M
Q 1, 2, 3, 4*, 5*
Y f.
Pump Discharge Pressure-High NA M
Q 1, 2, 3, 4*, 5*
1 g.
HPCS System Flow Rate-Low NA M
Q 1, 2, 3, 4*, 5*
l h.
Manual Initiation NA R
NA 1, 2, 3, 4*, 5*
D.
LOSS OF POWER 1.
4.16 kv Emergency Bus Under-NA NA R
1, 2, 3, 4**, 5**
voltage (Loss of Voltage)
Not required to be OPERABLE when reactor steam done pressure is less than or equal to 122 psig.
When the system is required to be OPERABLE after being manually realigned, as applicable, per
.i Specification 3.5.2.
4 Required when ESF equipment is required to be OPERA 8LE.
l a
F 4
TABLE 4.3.5.1-1 5
REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS g
r-E CHANNEL CHANNEL FUNCTIONAL CHANNEL 3
FUNCTIONAL UNITS CHECK TEST CALIBRATION j
g a.
]
Low Low, Level 2 NA M
R l
1 b.
Reactor Vessel Water NA M
R I
Level - High, Level 8 c.
Manual Initiation NA R
NA R
+
Y 8
E an i
~
f A
UNITED STATES y
g NUCLEAR REGULATORY COMMISSION 3.
- j WASHINGTON, D. C. 20555
%, v...../
COMMONWEALTH EDIS0N COMPANY DOCKET N0. 50-374 LA SALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment 10 License No. NPF-18 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for amendment filed by the Commonwealth Edisor, Company, dated February 21, 1985, 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 Commission; C.
There is reasonable assurance:
(1) 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; 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 attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 10 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 i
i
- 3.
This amendment is effective as of November 30, 1985.
FOR THE NUCLEAR REGULATORY COMISSION
//NI' A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
Changes to the Technical Specifications Date of Issuance: April 30,1985 d
4 g-
--w,-
-.- a.
+m---,,-
I.
.3.
This amendment is effective as of November 30, 1985.
FOR THE NUCLEAR REGULATORY COMMISSION A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
Changes to the Technical Specifications Date of Issuance: April 30,1985 LBf2 L/LA LB#2 L/PM i
LB 2 L/BC EHd n ABournia:1b ASch neer ak 04/{k/85 04/j8'/85 M
85>-04/ /85-
/85 04 k
ENCLOSURE TO LICENSE AMENDMENT NO.10 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 3-7 3/4 3-7 3/4 3-20 3/4 3-20 3/4 3-22 3/4 3-22 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-49 3/4 3-49 l
5 TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS r-7 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (3) SURVEILLANCE REQUIRED c
w N
1.
Intermediate Range Monitors a.
Neutron Flux - High S/UID) S S/U(c),W R
2 S
W R
3,4,5 b.
Inoperative NA W
NA 2,3,4,5 2.
Average Power Range Monitor:(I) a.
Neutron Flux - High' Setdown S/U(b) S S/U(c),W SA 1, 2 S
W SA 3, 5 A
Flow Biased Simulated Thermal (9)S/U(c),y y(d)(e), SA, R(h) j w
b.
Power-Upscale S, D w
c.
Fixed Neutron Flux -
O High S
S/U(c),y y(d), SA 1
d.
Inoperative NA W
NA 1,2,3,5 3.
Reactor Vessel Steam Dome Pressure - High NA M
Q 1, 2 4.
Low, Level 3 NA M
R 1, 2 l
5.
Main Steam Line Isolation Valve - Closure NA M
R 1
N m
6.
Main Steam Line Radiation -
h High S
M R
1, 2 m
E 7.
Primary Containment Pressure -
g:
High NA M
Q 1, 2 E;
g TABLE 4.3.2.1-1 g
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS F
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH
[
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 3
A.
AUTOMATIC INITIATION m
1.
PRIMARY CONTAINMENT ISOLATION a.
Low, Level 3 NA M
R 1, 2, 3 2)
Low Low, Level 2 NA M
R 1, 2, 3 b.
Drywell Pressure - High NA M
Q 1, 2, 3 c.
Radiation - High S
M R
1, 2, 3 2)
Pressure - Low NA M
Q 1
3)
Flow - High NA M
R 1, 2, 3 l
d.
Main Steam Line Tunnel R
Temperature - High NA M
R 1,2,3 e.
Condenser Vacuum - Low NA M
Q 1, 2*, 3*
T f.
Main Steam Line Tunnel El A Temperature - High NA M
R 1,2,3 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Vent Exhaust Plenum Radiation - High S
M R
1, 2, 3 and **
b.
Drywell Pressure - High NA M
Q 1,2,3 c.
Reactor Vessel Water g
Level - Low Low, Level 2 NA M
R 1, 2, 3, and l
d.
Fuel Pool Vent Exhaust Radiation - High S
M R
1, 2, 3 and **
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION g
a.
A Flow - High S
M R
1,2,3 b.
Heat Exchanger Area Temperature - High NA M
Q 1,2,3 c.
Heat Exchanger Area F
Ventilation AT - High NA M
Q 1, 2, 3 d.
SLCS Initiation NA R
NA 1, 2, 3 o
e.
Reactor Vessel Water Level - Low Low, Level 2 NA M
R 1, 2, 3 l
g TABLE 4.3.2.1-1 (Continued) y ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS-l E
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED "e
6.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION m
a.
Low, Level 3 NA M
R.
1, 2, 3 l
i b.
Reactor Vessel I
(RHR Cut-in Permissive)
Pressure - High NA M
Q 1,2,3 c.
RHR Pump Suction Flow - High NA M
Q 1, 2, 3 d.
RHR Area Temperature - High NA M
Q 1,2,3 i
e.
RHR Equipment Area AT - High NA M
Q 1,2,3 i
R i
B.
MANUAL INITIATION Y
i y
1.
Inboard Valves NA R
NA 1,2,3 2.
Outboard Valves NA R
NA 1,2,3 3.
Inboard Valves NA R
NA 1, 2, 3 and **,#
l 4.
Outboard Valves NA R
NA 1, 2, 3 and **,#
5.
Inboard Valves NA R
NA 1,2,3 i
6.
Outboard Valves NA R
NA 1,2,3 i
7.
Outboard Valve NA R
NA 1,2,3 i
j "When reactor steam pressure > 1043 psig and/or any turbine stop valve is open.
- When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
l g
- During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
3
- s B
g n
I i
E 1
TABLE 4.3.3.1-1 g
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS-F CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH g
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED G
A.
DIVISION I TRIP SYSTEM m
1.
RHR-A (LPCI MODE) AND LPCS SYSTEM a.
Low Low Low, Level 1 NA M
R 1, 2, 3, 4*, 5*
l b.
Drywell Pressure - High NA M
Q 1, 2, 3 c.
LPCS Pump Discharge Flow-Low NA M
Q 1, 2, 3, 4*, 5*
d.
LPCS and LPCI A Injection Valve Injection Line Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
R e.
LPCS and LCPI A Injection Valve Reactor Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
y f.
LPCI Pump A Start Time Delay g
Relay NA M
Q 1, 2, 3, 4*, 5*
g.
LPCI Pump A Flow-Low NA M
Q 1, 2, 3, 4*, 5*
h.
Manual Initiation NA R
NA 1, 2, 3, 4*, 5*
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
a.
Low Low Low, Level 1 NA M
R 1, 2, 3 l
b.
Drywell Pressure-High NA M
Q 1, 2, 3 c.
ADS Timer NA M
Q 1,2,3 d.
Low, Level 3 NA M
R 1,2,3 3g e.
LPCS Pump Discharge g
g Pressure-High NA M
Q 1,2,3 g
f.
LPCI Pump A Discharge g
Pressure-High NA M
Q 1,2,3 g.
Manual Initiation NA R
NA 1, 2, 3 2
O
= -
i
~
g TABLE 4.3.3.1-1 (Continued).
W EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS l--
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED i
4 m
B.
DIVISION 2 TRIP SYSTEM j
1.
1 Low Low Low, Level 1 NA M
R 1, 2, 3, 4*, 5*
l 3
b.
Drywell Pressure - High NA M
Q 1, 2, 3 c.
LPCI 8 and C Injection Valve Injection Line Pressure Low 1
Interlock NA M
R 1, 2, 3, 4*, 5*
i d.
LPCI Pump B Start Time Delay w
j Relay NA M
Q 1, 2, 3, 4*, 5*
w e.
LPCI Pump Discharge Flow-Low NA M
Q 1, 2, 3, 4*, 5*
1 i
J, f.
Manual Initiation NA R
NA 1,2,3,4*,5*
g.
LPCI B and C Injection Valve l
Reactor Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
l l
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
a.
i Low Low Low, Level 1 NA M
R 1,2,3 l
b.
Drywell Pressure-High NA M
Q 1, 2, 3 j
c.
ADS Timer NA M
Q 1,2,3 d.
k Low, Level 3 NA M
R 1, 2, 3 l
l e.
LPCI Pump B and C Discharge
=
I Pressure-High NA M
Q 1,2,3 g
f.
Manual Initiation NA R
NA 1, 2, 3
- o 1
g TABLE 4.3.3.1-1 (Continued)
W EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS l-CHANNEL OPERATIONAL-CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E
TRIP FUNCTION CHECK TEST' CALIBRATION SURVEILLANCE REQUIRED M
C.
DIVISION 3 TRIP SYSTEM m
1.
HPCS SYSTEM a.
Low Low, Level 2 NA M
R 1, 2, 3, 4*, 5*
l 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*
l d.
Condensate Storage Tank. Level -
Low NA M
Q 1, 2, 3, 4*, 5*
R e.
Suppression Pool Water Level - High NA M
Q 1, 2, 3, 4*, 5*
Y 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*
h.
Manual Initiation NA R
NA 1, 2, 3, 4*, 5*
0.
LOSS OF POWER 1.
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)
(Division 3)
Ng TABLE NOTATIONS
&g
- Not required to be OPERA 8LE when reactor steam dome pressure is less than or equal to 122 psig.
- When the system is required to be OPERABLE after being manually realigned, as applicable, per P
Specification 3.5.2.
g
5-TABLE 4.3.5.1-1 N
REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS F
CHANNEL CHANNEL FUNCTIONAL CHANNEL ji FUNCTIONAL UNITS CHECK TEST CALIBRATION M
n, a.
Low Low, Level 2 NA M
R l
b.
Reactor Vessel Water NA M
R l
Level - High, Level 8 c.
Manual Initiation NA R
NA R
+
Ye F
R R
a E
l t