ML20032E821
| ML20032E821 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/18/1981 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20032E819 | List: |
| References | |
| NUDOCS 8111230327 | |
| Download: ML20032E821 (20) | |
Text
,
DEFINITIONS PRIMARY CONTAINMENT INTEGRITY (Continued) 1.24.3 Each containment air lock is OPERABLE pursuant to Specification 3 6.*.3.
1 24.4 The containment leakage rates are within the limits of Specification 3.6.1.2.
1 24.5 The sealing mechanism associated with each penetration (e.g.,
welds, bellows or 0-rings is) is OPERABLE.
Rt. FED THERMAL POWER 1.23 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2436 MWT.
REACTOR PROTECTION SYSTEM RESPONSE TIME 1.26 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.
REFERENCE LEVEL ZERO 1.27 The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel zero point. This REFERENCE LEVEL ZED is approximately mid-point on the top fuel guide and is the single reference for all specifications of vessel water level.
REPORTABLE OCCURRENCE 1.28 A REPORTABLE OCCURRENCE shall be any of those conditions specified in l
Specification 6.8.1.8 and 6.3.1.9.
ROD DENSITY 1.29 ROD DENSITY shall be the number of control rod notches inserted as a l
fraction of the total number of notches. All rods fully inserted is equivalent to 100% ROD DENSITY.
SECONDARY CONTAINMENT INTEGRITY l
1.30 SECONDARY CONTAINMENT INTEGRITY shall exist when:
l 1.30.1 All automatic reactor builidng ventilation system isolation l
valves or dampers are OPERABLE or secured in the isolated position, 1.30.2 The standby gas treatment system is OPERABLE pursuant to l
Specification 3.6.6.1.
1.30.3 At least one door in each access to the reactor building is l
closed.
BRUNSWICK - UNIT 1 1-5 Amendment No.
8111230327 811118 PDR ADOCK 05000324 P
_ DEFINITIONS 1.30.4 The sealing mechanism associated with each penetration (e.g.,
l welds, bellows or 0-rings) is OPERABLE.
SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor l
would be suberitical assuming that all control rods capable of insertion are fully inserted except for the analytically determined high worth rod which is assumed to be fully withdrawn, and the reactor is in the rtdown condition, cold, 68'F, and Xenon free.
STAGGERED TEST BASIS "1 32 A STAGGERED TEST BASIS shall consist of:
l a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the l
TOTAL PEAKING FACTOR 1.34 The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LHGR for any l
apecific location on a fuel rod divided by the average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.
UNIDENTIFIED LEAKAGE 1.35 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED l
LEAKAGE.
i BRUNSWICK - UNIT 1 1-6 Amendment No.
I
g TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS O
ALLOWABLE
' FUNCTIONAL UNIT AND INSTRUMENT NUhBER TRIP SETPOINT VALUES 1.
Intermediate Range M>nitor, Neutron Flux - HigLII)
<120 divisions of full scale
<120 divisions (C51-IRM-K601 A,B,C,D,E,F,G,H) of full scale y
2.
Average Power Range Monitor (C51-APRM-CH.A,B,C,D,E,F)
Neutron Flux - High 15%(2)
<15% of RATED THERMAL POWER
<15% of RATED a.
THERMAL POWER b.
Flow Biased Neutron Flux - High(3)(4)
<(0.66 W + 54%)
<(0.66 W + 54%)
Fixed Neutron Flux - High(4)
<120% of RATED DIERMAL POWER
<120% of RATED c.
y THERMAL POWER c.
3.
Reactor Vessel Steam Dome Pressure - High (B21-PTM-N023A-1,B-1,C-1,D-1)
<1045 psig
<1045 psig 4.
Reactor Vessel Water Level - Low, Leve1#1(7)
>+162.5 inches
>+162.5 inches (B21-LTM-N017A-1,B-1,C-1,D-1) 5.
Main Steam Line Isolation Valve - Closure (5)
<10% closed
<10% closed (B21-F022 A,B,C,D; B21-F028 A,B,C,D) 6.
Main Steam Line Radiation - High
<3 x full power background
<3.5 x full power
( D12-RM-K603 A,B,C,D) background N.
Q.
E F
TABLE 2.2.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION
.e (1)
The Intermediate Range Monitor scram functions are automatically bypassed when the reactor mode switch is placed in the Run position and the Average Power Range Monitors are on scale.
(2)
This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is placed in the Run position.
(3)
The Average Power Range Monitor scram function is varied, Figure 2.2.1-1, as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.
(4)
The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
(5)
The Main Steam Line Isolation valve-Closure scram function is auto-matically bypassed when the reactor mode switch is in other than the Run position.
(6)
These scram functions are bypassed when THERMAL POWER is less than 30%
of RATED THERMAL POWER.
(7)
Vessel water levels refer to REFERENCE LEVEL ZERO.
BRUNSWICK - UNIT 1 2-6 Amendment No.
TABLE 3.3.2-2 m
- n h
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 5
1 Q
ALLOWABLE TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT VALUE h
1.
PRIMARY CONTAINMENT ISOLATION H
f a.
Reactor Vessel Water IAvel - Iow*
1.
Level #1 (B21-LTM-N017A-1, B-1, C-1, D-1)
> +162 5 inches
> +162.5 inches 2.
Level #2 (B21-LTM-N024A-1, B-1 and
[+112 inches
[+112 inches
]
B21-LTM-N025A-1, B-1) b.
Drywell Pressure - High
< 2 psig
< 2 psig (C71-PS-N002 A,B,C,D) c.
Radiation - High R
(D12-RM-K603 A, B, C, D)
< 3 x full power background
< 3.5 x full power background l
Y 2.
Pressure - Low
> 825 psig
> 825 psig l
[
(B21-PS-N015A,B,C,D) 3.
Flow - High
< 140% of rated flow
< 140% of rated flow (B21-dPIS-N006 A,B,C,D; B21-dPIS-N007 A,B,C,DT B21-dPIS N008 A,B,C,D; and B21 dPIS-N009 A,B,C,D) t d.
Main Steam Line 'Ibanel Temperature - High
< 200 F
< 200 F (B21-TS-N010 A, B, C, D; B21-TS-N011 A, B, C, D; B21-TS-N012 A, B, C, D; and B21-TS-N013 A, B, C, D) d e.
Condenser Vacuum - Low
> 7 inches Ng vacuum
-> 7 inchee Hg (B21-PS-N056 A,B,C,D) vacuum i
f.
Turbine Building Area Temp. - High
< 200 F
< 200 F E
(B21-TS-3225 A,B,C,D; B21-TS-3226 A,B.C.D; B21-TS-3227 A,B,C,D; B21-TS-3228 A,B,C,D; B21-TS-3229 A,B C.D; B21-TS-3230 A,B,C,D; B21-TS-3231 A,B,C,D; B21-TS-3232 A,B,C,D) a F
- Vessel water levels refer to REFERENCE LEVEL ZERO.
l
.. _ =
m TABLE 3.3.2-2 (Continued) m ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 5
Q ALLOWABLE TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT VALUE l
2.
SECONDARY CONTAINMENT ISOLATION s
a.
Reactor Building Exhaust Radiation - High
< 11 mr/hr
< 11 mr/hr
~
(D12-R!!-N010 A, B) b.
Drywell Pressure - High (C71-PS-NOO2 A,B,C,D)
~-< 2 psig
""< 2 psig i
l c.
Reactor Vessel Water Level - Low, level #2*
.i (B21-LTM-NO24A-1,B-1 and B21-LTM-N075A-1,B-1)
--> +112 inches
> +112 inches
~-
Ei 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION w
Y a.
A Flow - High
< 53 gal / min 5
(G31-dFS-N603-1A, IB)
--< 53 gal / min
~-
b.
Area Temperature - High
< 150" F j
(G31-TS-N600A, B, C, D, E, F)
-- 150 F I
c.
Area Ventilation Temperature A Temp - High
< 50" F l
(G31-TS-N602A, B, C, D, E, F)
-- 50 F
)
{
d.
SLCS Initiation (C41 A-SI)
NA NA 1
I l
m e.
Reactor Vessel Water - Low, Level #2*
j l.
(B21-LTM-N024A-1, B-1 and B21-LTM-N025A-1, B-1) --> +112 inches
> +112 inches
~~
i s
t n
k
- Vessel water levels refer to REFERENCE LEVEL ZERO.
I t
i J
0 m
t;
-TA'LE 3.3.2-2 (Continued)
- o E
w ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M
Q ALLOWABLE
-e TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT VALUE 5.
SHUTDOWN COOLING SYSTEM ISOLATION
>-3 a.
Reactor Vessel Water - Iow, Level #1*
~> +162.5 inches
> +162.5 inches
~
(B21-LTM-N017A-1, B-1, C-1, D-1) b.
Reactor Steam Dome Pressure - High
-< 140 psig
~< 140 psig.
(B32-PS-N018A, B)
R>
Y M
?1 f
- Vessel water levels refer to REFERENCE LEVEL ZERO.
l P,
.if
=
g TABLE 3.3.3-2 Eb y
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS
,4 b
ALLOWV3LE
~
TRIP FUNCTION AND INSTRITdENT NUMBER TRIP SETPOINT' VAlly.
E y
1.
CORE SPRAY SYSTEM Reactor Vessel Water Level - Inw, level #3*
~> +2.5 inches
> +2.5 inches a.
(B21-LTS-NO31A-4, B-4, C-4, D-4)
~
b.
Reactor Steam Dome Pressure - Iow (B21-PTS-N021 A-2, B-2, C-2, D-2) 410 + 15 psig 410 + 15 psig c.
Drywell Pressure - High
< 2 psig
< 2 psig (El1-PS-N011A,B,C,D) d.
Time Delay Relay 14 _< t _<_ 16 secs 14 _< t_< 16 secs i
N e.
Bus Power Monitor (E21-KI A, B)
NA NA Yy 2.
Drywell >ressure - High
< 2 psig
< 2 psig (Ell-PS-N0llA,B,C,D) b.
Esactor vessel Water Ievel - Iow, Ievel #3*
~> +2.5 inches
~> +2.5 inches (B21-LTS-N031A-4, B-4, C-4, D-4) l c.
Reactor Vessel Shroud level *
> -53 inches
> -53 inches (B21-LTM-NO36-1 and B21-LTM-NO37-1) g it m
S
- Vessel water levels refer to REFERENCE LEVEL ZERO.
~
o L
1
J m
TABLE 3.3.3-2 (continutd) 4 W
h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS s
Q ALLOWABLE TRIP FUNCTION AND INSTRUMEffr NUMBER TRIP SETPOINT VALUE 5
i A
3.
HPCI SYSTEM 2
H
'a.
Reactor Vessel Water level - Iow, Ievel #2*
-> +112 inches
> +112 inches (B21-LTS NO31A-2, B-2, C-2, D-2)
}
b.
Drywell Pressure - IEgh
-< 2 psig
-< 2 psig j
(Ell-PS-N0llA,B,C,D)
T c.
Condensate Storage Tank Ievel - Iow l
(E41-LS-N002, E41-LS-N003)
-> 23'4"
> 23'4" f
d.
Suppression Chamber Water level - High**
l (E41-LSH-N015A, B)
-< -2 feet
< -2 feet w
e.
Bus Power Monitor (E41-K55 and E41-K56)
N/A N/A E
w 4.
LDS d,
a.
Drywell pressure - High
< 2 psig
-< 2 psig
( E11-l'S-N010A, B, C,D) b.
Reactor Vessel Water level - Iow, Level #3*
-> +2.5 inches
> +2.5 ira.no l
(B21-LTS-NO31A-3, B-3, C-3, D-3) c.
Reactor Vessel Water level - Iow, Ievel #1*
-> +162.5 inches
-> +162 5 inches l
(B21-LTM-N042A-1, B-1) d.
ADS Timer (B21-TDPU-KSA, B)
< 120 seconds
< 120 seconds E.
e.
Core Spray Pump Discharge Pressure - High
> 100 psig
> 100 psig S
(E21-PS-N008A, B and E21-PS-NOO9A, B)
E
- 2:
f.
RHR (LPCI Mode) Pump Discharge Pressure - High
> 100 psig
-> 100.psig (E11-PS-N016A, B, C, D and E11-PS-N020A, B, C, D)-
- Vessel water levels refer to REFERENCE LEVEL ZERO.
- Suppression chamber water level zero in. the torus centerline.-minus 1 inch.
t
1 en TABLE 3.3.6.1-2
- o E
ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION-SETPOINTS g
s b
TRIP ALLOWABLE e
TRIP FilNCTION AND INSTRUMENT NUMBER SETPOINT VALUE 1.
Reactor Vessel, Water Level -
d Low low, Level #2*
(B21-LTM-N024A-2,B-2 and
-> +112 inches
-> +112 inches B21-LTM-N025A-2,B-2) 2.
Reactor Vessel Pressure - High
-< 1120 psig
-< 1120 psig (B21-PS-N045A,B,C,D) i 5
U a
- ~
4 y
8 I
l 1
I i
i t
(
k
- Vessel water levels refer to REFERENCE LEVEL ZERO.
R 1
if i
i
DEFINITIONS PRIMARY CONTAINMENT INTEGRITY (Continued) b.
All equipment hatches are closed and sealed.
c.
Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.
d.
The containment leakage rates are within the limits of Specification 3.6.1.2.
e.
The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.
RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2436 MWT.
REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.
REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the vessel zero point. This REFERENCE LEVEL ZERO is approximately mid-point on the top fuel guide and is the single reference for all specifications of vessel water level.
REPORTABLE OCCURRENCE A REPORTABLE OCCURRENCE shall be any of those conditions specified in Spccifications 6.9.1.8 and 6.9.1.9.
t ROD DENSITY ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of notches.
All rads fully inserted are equivalent to 100% ROD DENSITY.
SECONDARY CONTAINMEN'T INTEGRITY SECONDARY CONTAINMENT INTEGRITY shall exist when:
a.
All automatic Reactor Building ventilation system isolation valves or dampers are OPERABLE or secured in the isolated position, b.
The standby gas treatment system is OPERABLE pursuant to Specification 3.6.6.1.
c.
At least one door in each access to the Reactor Building is closed.
BRUNSWICK - UNIT 2 1-5 Amendment No.
DEFINITIONS d.
The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.
SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount or reactivity by which the reactor would be suberitical assuming that all control rods capable of insertion are fully inserted except for the analytically determined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68'F, and Xenon free.
STAGGERED TEST' BASIS A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TOTAL PEAKING FACTOR The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local URGR for any specific location on a fuel rod divided by the average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.
UNIDENTIFIED LEAKAGE UNIDF IED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
l BRUNSWICK - UNIT 2 1-6 Amendment No.
~
en TABLE 2.2.1-1
- o h
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS s
Q ALLOWABLE l
_ FUNCTIONAL UNIT AND INSTRUMENT NUMBER TRIP SETPOINT VALUES h
1.
Intermediate Range Monitor, Neutron Flux - High(I)
< 120 divisions of full scale
< 120 divisions i
(C51-IRM-K601 A,B,C,D,E,F,G,H) of full scale 2.
Average Power Range Monitor l
(CSI-APRM-CH.A,B,C,D,E,F) a.
Neutron Flux - High, 15%(2)
< 15% of RATED THERMAL POWER
< 15% of RATED THERMAL POWER b.
Flow Bf ased Neutron Flux - High(3)(4)
< (0.66 W + 54%)
< (0.66 W + 54%)
Fixed Neutron Flux - IUgh(4)
< 120% of RATED THERMAL POWER
< 120% of RATED c.
THERMAL POWER 3.
Reactor Vessel Steam Dome Pressure - High
-< 1045 psig
-< 1045 psig w
E (B21-PS-N023 A,B,C,D) 4.
Reactor Vessel Water Lcvel - Iow, Ievel #1(7) j (B21-LIS-N017 A,B,C,D)
-> +162.5 inches
> +162.5 inches 1
I 5.
Main Steam Line Isolation valve - Closure (5) i 10% closed I 10% closed l
(B21-F022 A,B,C D; B21-F028 A,B,C,D)
I 6.
Main Steam Line Radiation - High
-< 3 x full power background
-< 3.5 x full power
( D12-RM-K603 A,B,C,D) background g
Drywell Pressure - High
< 2 psig
< 2 psig 7.
(C72-PS-N002 A,B,C,D) m i
5.
j 8.
Scram Discharge Volume Water level - High
-< 109 gallons
-< 109 gallons S
(Cl2-LSH-N013 A,B,C,D)
?
4
TABLE 2 2.1-1 (Continu:d)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION (1) The Intermediate Range. Monitor scram functions are automatically bypassed when the reactor mode switch is placed in the Run position and the Average Power Range Monitors are on scale.
(2) This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is placed in the Run position.
(3) The Average Power Range Monitor scram function is varied, Figure 2.2.1-1, as a function of recirculation loop flow (W).
The trip setting of this function must be maintained in accordance with Specification 3.2.2.
(4) The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds.
The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
(5) The Main Steam Line Isolation Valve-Closure scram function is automatically bypassed when the reactor mode switch is in other than the Run position.
(6) These scram functions are bypassed when THERMAL POWER is less than 30% of RATED THERMAL POWER.
(7) Vessel water levels refer to REFERENCE LEVEL ZERO.
I BRUNSWICK - UNIT 2 2-6 Amendment No.
to TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS c
s 9
ALLOWABLE 5
TRIP FUNCTION AND INSTRUMENT NUMBER TRIF SETPOINT VALUE i
1.
PRIMARY CONTAINMENT ISOLATION 1
8 a.
Reactor Vessel Water level - Iow*
l.
Level #1 (B21-LIS-N017 A,B,C,D)
> +162.5 inches
> +162.5 inches 2.
Level #2 (B21-LIS-NO24 A,B, and I +112 inches i
B21-LIS-N025 A,B)
~
I +112 inches
]
b.
Drywell Pressure - High
< 2 psig
< 2 psig (C72-PS-N002 A,B,C,D) c.
Radiation - High
-~< 3 x full power background
< 3.5 x full power
( D12-RM-K603 A,B,C,D) background w
3 2.
Pressure - Low
,_ 825 psig
> 825 psig (B21-PS-N015 A,B,C,D) w 3.
Flow - 111gh
< 140% of rated flow
< 140% of rated flow 1
O (B21-dPIS-N006 A,B,C,D; B21-dPIS-NOO7 A,B,C.D; B21-dPIS-NOO8 A,B,C,D; and B21-dPIS-N009 A,B,C.D) 4.
Flow - High
< 40% of rated flow
< 40% of rated flow (B21-dPIS=N006A; B21-dPIS=N0078; B21-dPIS-N008C and B21-dPIS-N009D)
'~
1 d.
Main Steam Line Tunnel 4
Temperature - High
< 200*F
< 200*F (B21-TS-N010 A,B,C,D; B21-TS-N011 A,B,C,D; B21-TS N012 A,B,C,D; and B21-TS-N013 A,B,C,D)
]
e.
Condenser Vacuum - Iow
> 7 inches Hg vacuum
> 7 inches Hg vacuum (B21-PS-N056 A,B,C,D)
~~
E.
f.
Turbine Building Area Temp - High
< 200*F
< 200*F (B21-TS-3225 A,B,C,D; B21-TS-3226 A,B,C,D; B21-TS-3227 A,B,C,D; B21-TS-3228 A,B,C,D; ft B21-TS-3229 A,B,C,D; B21-TS-3230 A,B,C,D; B21-TS-3231 A,B,C,D and B21-TS-3232 A,B,C,D) e
- Vessel water levels refer to REFERENCE LEVEL ZERO.
es TABLE 3.3.2-2 (Continutd) x v3 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M
O ALLOWABUE i
TRIP FUNCTION AND INSTRITMENT NUMBER TRIP SETPOINT VALUE 5A 2.
SECONDARY CONTAINMENT ISOLATION H
a.
Reactor Building Exhaust Radiation - lugh
~< 11 mr/hr
~< 11 ar/hr (D12-RM-N010 A,B) b.
Drywell Pressure - High (C72-PS-N002 A,B C.D)
< 2 psig
< 2 psig c.
Reactor Vessel Water Level - Low, Level #2*
> +112 inches
> +112 inches (B21-LIS-N024 A,B and B21-LIS-N025 A,B) 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION y
a.
A Flow - High
< 53 gal / min
< 53 gal / min g
(G31-dFS-N603-1A,lB) b.
Area Temperature - High
~~< 150*F
--< 150*F (G31-TS-N600A,B,C,D,E,F) c.
Area Ventilation Temperature A Temp-High
~~< 50*F
< 50*F (G31-TS-N602A,B,C,D,E,F)
-~
d.
SLCS Initiation (C41 A-SI)
NA NA e.
Reactor Vessel Water - Low, Level #2*
},+112 inches
}_ +112 inches g
g (B21-LIS-N024A,B and B21-LIS-N025A,B)
O' 8n
- Vessel water levels refer to REFERENCE LEVEL ZERO.
l 4j 1
en TABLE 3.3.2-2 (Continu d)
E l
ISOLATION ACTU'iTION INSTRUMENTATION SETPOINTS i
N ALLOWABLE.
8 TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT VALUE E-j 5.
SHUTDOWN COOLING SYSTEM ISOLATION H
a.
Reactor Vessel Water
. Iow, tevel #1* '
-l (B21-LTM-N017A-1, B-1, C-1, D-1)
-> +162.5 inches
-> +162.5 inches I
b.
Reactor Steam Dome Pressure - High
-< 140 psig
-< 140 psig (B32-PS-N018A, B) l 4
1 T
3 i
I d
i w
N y
s I
i i
1 4
i 1
No C.
a a
{
g
- Vessel water levels refer to REFERENCE LEVEL ZERO.
i I
l I
i J
4 i
t to TABLE 3.3.3-2 b
j-fA EMERGENCY CORE COOLING SYSTEA ACTUATION INSTRUMENTATION SETPOINTS i
b 9
ALLOWABLE TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT VALUE I
$g 1.
CORE SPRAY SYSTEM
__ 2.5 inches
> 2 5 inches a.
Reactor Vessel Water level - law level #3*
j (B21-LIS-N031A,B,C,D) b.
- Reactor Steam bome Pressure - Low 410 + 15 psig 410 + 15 psig (B21-PS-N021A,B,C,D)
~
c.
Drywell ?ressure - High
< 2 psig
. < 2 psig (E11-PS-N011A,B,C,D) d.
Time Delay-Pelay 14 < t < 16 secs 14 < t < 16 aecs 4
i e.
Bus Power Monitor (E21-K1 A,B)
NA-NA 1
2.
g a.
Drywell Pressure - High
< 2 psig
< 2 psig S-(E11-PS-N011A,B,C D)
)
y b.
Reactor Vessel Water level'- Inw, Level #3*
__> +2.5 inches
> +2.5 inches l
(B21'LIS-N031A,B,C D) y c.
Reactor Vessel Shroud level *
> -53 inches
>'-53 inches
_l, (B21-LITS-NO36 and B21-LITS-N037) t d.
Reactor Steam Dome Pressure - Iow
]
(B21-PS-N021A,B,C,D) 1.
RHR Pump Start and LCPI Valve Actuation
.410 + 15 psig 410 + 15 psig i
2.
Recirculation Pump Discharge Valve 310[15'psig 310[15psig Actuation e.
RHR Pump Start - Time - Delay Relay 9<t< 11 seconds 9 < t < 11 seconds f.
Bus Power Monitor (E11-K106A,B)
NA NA 3[ -
a e
- Vessel water levels -refer to, REFERENCE ' LEVEL ZERO.
i
)
0
P g
TABLE 3.3.3-2 (Continued)
O
~
g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMCNTAfION SETPOINTS
~
O ALLOWABLE TRIP FUNCTION AND INSTRUMENr NUMBER TRIP SETPOINT VALUE Ey 3.
HPCI SYSTEM a.
Reactor Vessel Water level - Iow, Ievel #2*
-> +112 inches
- +112 inches (B21-LIS-N031A,B,C,D) b'.
Drywell Pressure-High
< 2 psig
< 2 psig (E11-PS-N011A,b,C,Di c.
Condensate Storage Tank level - Iow
-> 23'4"
-> 23'4" (E4:-LS-N002, E41-LS-N003) d.
Suppression Chr.mber Water Ievel - High**
-< -2 feet
-< -2 feet (E41-LSH-N015.1 B) e.
Bus Power Monitor (E44-K55 and E41-K56)
NA NA R
4.
ADS Tg a.
Drywell Pressure-High
< 2 psig
< 2 psig (E11-PS-N010A,B,C,D) b.
Reactor Vessel Water Level '- Iow, level #3*
> +2.5 inches
> +2.5 inches l
(B21-LIS-N031A,B,C,D) l c.
Reactor Vessel Water level - Iow level #1*
-> + 162.5 inches
-> + 162.5 inches (B21-LIS-N042 A,B) d.
ADS Timer (B21-TDPU-K5A,B)
< 120 seconds
< 120 seconds e.
Core Spray Pump Diacharge Pressure - High
[100psig
[100psig (E21-PS-N008A,B and E21-PS-N009A,B) f.
RhR (LPCI MODE) Pump Discharge Pressure - High
-> 100 psig
-> 100 psig (E11-PS-N016 A,B,C,D and E11-PS-N020A,B,C.D) y g.
Bus Power Monitor (B21-K1 A,B)
NA NA 0
a-
- Vessel water levels refer to REFERENCE LEVEL ZERO.
p
- Suppression chamber water level -zero is the torus centerline minus 1 inch.
es TABLE 3.3.6.1-2 W.
ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS n
x TRIP ALLOWABLE TRIP FUNCTION AND INSTRUMENT NUMBER SETPOINT VALUE H
--> +112 inches
~-> +112 inches 1.
Reactor Vessel, Water Level - Low low, Level #2*
(B21-LIS-N024 A, B; B21-LIS-N025 A, B) 2.
Reactor Vessel Pressure-High
<1120 psig
<1120 psig (B21-PS-N045 A, B, C, D)
U t~
Y l
N i
o E
- Vessel water levels refer to REFERENCE LEVEL ZERO.
E
.