ML20065R650
| ML20065R650 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/05/1990 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065R654 | List: |
| References | |
| NUDOCS 9012190122 | |
| Download: ML20065R650 (60) | |
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UNI TED STATES
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4 CAROLINA POWER & LIGHT COMPANY. et.al,
' DOCKET NO. 50-325 BRUNSWICK STEAM. ELECTRIC PLANT. UNIT 1 AMENDMENT T0 FACILITY OPERATING LICENSE Amendment No.149 License No. DPR-71 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
-The application for amendment filed by Carolina Power & Light Company (the licensee), dated February 29, 1988, superseded September 20, 1989, as supplemented December 5,1989, February 15, August 9, and 0ctober 24, 1990, complies with the standards and requirements of i
the Atomic Energy Act of-1954, as amended (the Act), and the Commission's rules and 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 rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering tne health and safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
-The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. DPR-71 is hereby amended to read as fcilows:
9012190122 901205 PDR ADOCK 0500032S P
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. (2) Technical Specifications i
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.149, are-hereby incorporated in the license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment'is effective as of the date of its issuence and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 6
4 Original Signed By:
Elinor G. Adensam, Director q
Project DirEitorate 11 1 Division of Reactor Projects. 1/11 Office of Nuclear Reactor Regulation i
Attachment:
Changes-to the Technical i
Specifications Da_te of-Issuance: -December 5,1990 i
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2 (2)
Technical Specifications
- The Technical Specifications contained in Appendices A and B, as revised through Amendment No.149, are hereby incorporated in the.
license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This. license amendment is effective as of the>date of its issuance and shall be implemented'within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Changes.to the Technical Specifications Date.of Issuance: December 5, 1990 1
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ATTACHMENT TO LICENSE AMENDNENT NO. 149 FACILITY OPERATING LICENSE NO. OPR-71 DOCKET NO. 50-325 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages X1 XI 1-5 1-5 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-17a 3/4 3-17a 3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 3/4'3-21 3/4 3-21 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26.
3/4 3-26 3/4 3-27 3/4 3-27 3/4-3-28 3/4 3-28 l
3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-32 3/4 3-32 3/4 6-14 3/4 6-14
-3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 3/4 6-17 3/4 6-17 l
3/4 6-22 3/4 6 ;
3/4 6-24 3/4 6-24 B3/4 6-5 B3/4 6-5 B3/4 6-6' B3/4 6-6 l
l
o INDEX BASES SECTION PACE 3/4.4 REACTOR COOLANT SYSTEM (Continued) 3/4.4.4 CHEMISTRY........................................... B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY................................... B 3/4 4-2 3/4.4.6 PRES SURE / TEMPERATURE LIMITS......................... B 3 /4 4-3 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.................... B 3/4 4-7
'3/4.4.8 STRUCTURAL INTEGRITY................................ B 3/4 4-7 3/4.5 EMERCENCY CORE COOLING SYSTEM 3/4.5.1 HICH PRESSURE COOLANT INJECTION SYSTEM.............. B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSUR12ATION SYSTEM (ADS)............. B 3/4 5-1 3/4.5.3 LOW PRES SURE COOLI NG S YSTEMS........................ B 3 /4 5-2 3/4.5.4 SUPPRESSION P00L.................................... B 3/4 5-4 1
3/4.6 CONTAINMENT SiSTEMS 3/4.6.1 PRIMARY CONTAINMENT................................. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS................ B 3/4 6-3 3/4.6.3 PR IM ARY CONTAINMENT ISOLATION VALVES................ B 3 /4 6-4 3/4.6.4 V A C U UM R E L I E F....................................... B 3 / 4 6 - 5 3/4.6.5 SECONDARY CONTAINMENT............................... B 3/4 6-5 l
3/4.6.6 CONTAINMENT ATMOSPHERE CONTR0L...................... B 3/4 6-6 l
l 3/4.7 PLANT SYSTENS, 3/4.7.1 SERVIC6 WATER SYSTEMS............................... B 3/4 7-1 l
3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM............ B 3/4 7-la l
l l
l l
1 I
BRUNSWICK - UNIT 1 XI Amendment No. 733,149
e DEFINITIONS OPERABLE - OPERABILITY ( Cont inued )
Implicit in this definit ion shall be t'
gion that all necessary attendant inst rument at ion, cont rol s, nc J emergency electric power sources, cooling or seal water, lubric:
. ar other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of perf orming their related support function (s).
OPE:lAT10NAL CONDITION An OPERATIONAL CONDITION hall be any one inclusive combination of mode switch position and average reactor coolant t emperature a s indicated in Table 1.2.
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the tundamental nuclear characteristic s of the reactor core and rs '
~j instrumentation and are 1) described in Sect ion 14 of the Updated FSA.
. authorized under the provisions of 10 CPR 50.59, or 3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEARACE PRESSURE BOUNDARY LE AK ACE shall be leakage through a non-isolatable fault in a reactor coolant system component body, pipe wall, or vessel wall.
PRlHARY CONTAINMENT INTECRITY PRIMARY CONTAINMENT INTECRITY shall exist whea:
a.
All penetrations required to be closed during accident conditions are eit her:
- 1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1.
I b.
All equipment hatches are closed and sealed.
c.
Each containment air lock i6 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.
BRUNSWICK - UNIT 1 1-5 Amendment No. 62. I24, 121, 149
TABLE 3.3.2-1 E
E ISOLATION ACTUATION INSTRUMENTATION E
E VALVE CROUPS MINIMUM NUMBER-APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)(c) CONDITION ACTION E
[I 1.
PRIMARY CONTAINMENT ISOLATION a.
1.
Low, Leiel'1 2, 6 2
1, 2, 3 20 8
2 1,2,3 27 2.
Lou, Level 3 1
2 1, 2, 3 20 b.
Drywell Pressure - High 2, 6 2
1, 2, 3 20 l
c.
1.
Radiation - High 1
2 1, 2, 3 21 5
kiI u
2.
Pressure - Low I
2 1
22 l
e i
I5) 2/line 1
22 l
3.
. Flow - High I
d.
Main Steam Line Tunnel I5) 2(d) 1, 2, 3 21 l
Temperature - High I
e.
Condenser Vacuum - Low II5}
2 1, 2 * }
21 l
I f.
Turbine Building Area Temperature - High II5) 4(d) 1 2, 3 21 l
g.
Main Stack Radiation - High (h) 1 1,2,3 28 U-El h.
Reactor Building Exhaust g
Radiation - High 6
I 1, 2, 3 20 l
_N ?
d
"'_=;n b
e
o TABLE 3.3.2-1 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION 2
E' VALVE GROUPS MINIMUM NUMBER APPLICABLE OPERATED BY' OPERABLE CilANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)(c) CONDITION ACTION Ey 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Bu iding Exhaust Radiation - High (1) 1 1, 2, 3, 5, and* 2 3 6
1 1, 2, 3
20 b.
Drywell Pressure - liigh (1) 2 1, 2, 23 2, 6 2
1, 2, 3 20 c.
Low, Level 2 (1 )
2 1, 2, 3 23 w
3 2
1, 2, 3 24 2
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
a Flow - liigh 3
1 1, 2, 3 24 b.
Area Temperature - High 3
2 1, 2, 3 24 Area Ventilation a Temperature - liigh 3
2 1,2,3 24 l
c.
II}
d.
SLCS Initiation 3
NA 1, 2, 3 24 e.
- 3>
Low, Level 2 3
2 1, 2, 3 24
~R
[.
f.
6 Flow - Iligh - Time Delay Relay NA 1
1, 2, 3 24 l
e.
T
TABLE 3.3.2-1 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION E
5 VALVE CROUPS MINIMUM NUMBER APPLICABLE.
OPERATED BY OPERABLE CllANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)(c) CONDITION ACTION E
N 4.
CORE STANDBY COOLING SYSTEMS ISOLATION a.
High Pressure Coolant Injection System Isolation 1.
HPCI Steam Line Flow - High 4
1
- 1. 2, 3 25 2.
IIPCI Steam Line Flow - Ifigh l
Time Delay Relay NA 1
1, 2, 3 25 3.
HPCI Steam Supply Pressure - Low 4
2 1,2,3 25 7(k) 1 1, 2, 3 25 l
w 5
w 4.
HPCI Steam Line Tunnel f
Temperature - High 4
2 1, 2, 3 25 5.
Bus Power Monitor NA(8}
1/ bus 1.,
2, 3' 26 6.
HPCI Turbine Exhaust Diaphragm Pressure - High 4
2 1,2,3 25 7.
HPCI Steam Line Ambient Temperature - liigh 4
1 1,2,3 25 8.
HPCI Steam Line Area a Temperature - High 4
1 1, 2, 3 25 l
0$
~ 5-9.
HPCI Equipment Area 3!
Temperature.High 4
1 1, 2, 3 25 l
un g z
10.
Dryvell Pressure - High 7(k) 1 1,2,3 25 l
s SS
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o TABLE 3.3.2-1-(Continued)
E
~
Q ISOLATION ACTUATION INSTRUMENTATION VALVE CROUPS MINIMUM NUMBER APPLICABLE Q
OPERATED Br OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)(c) CONDITION ACTION E
4 CORE STANDBY COOLING SYSTEMS ISOLATION (Continued)
U b.
Reactor Core Isolation Cooling System Isolation 1.
RCIC Steam Line Flow - High 5
1
- 1. 2. 3 25 2.
RCIC Steam Line Flow - liigh l
Time Delay Relay NA 1
1, 2, 3 25 3.
RCIC Steam Supply Pressure - Low 5
2 1, 2, 3 25 9(k) 1 1,2,3 25 l
4.
RCIC Steam Line Tunnel t[
Temperature - High 5
2
- 1. 2, 3 25 IR 1/ bus 1, 2, 3 26 5.
Bus Power Monitor HA 6.
RCIC Turbine Exhaust. Diaphragm Pressure - liigh 5
2
- 1. 2. 3 25 7.
-RCIC Steam Line Ambient Temperature - High 5
1 1, 2, 3 25 8.
RCIC Steam Line Area a Temperature - Ifigh 5
1 1, 2, 3 25 l
9.
RCIC Equipment Room Ambient Temperature - High 5
1 1, 2, 3 25 l
- g 10.
RCIC Equipment Room yu S a Temperature - liigh 5
1 1,2,3 25 l
o.
212 11.
RCIC Steam Line Tunnel ya R Temperature - High Time Delay Relay NA 1
1, 2, 3 25
_, z tm O 12.
Drywell Pressure - High 9(k)
I 1
2, 3 25
")
w
TABLE 3.3.2-l'(Continued)
E
~
E ISOLATION ACTUATION INSTRUMENTATION PE E
VALVE GROUPS HINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)(c) CONDITION ACTION E
i 5.
SHUTDOWN COOLING SYSTEM ISOLATION a.
2 'i 2
- 1. 2, 3 20 Low, Level I b
2
- 1. 2. 3 27 I) b.
Reactor Steam Dome Pressure - liigh 8
1
- 1. 2, 3 27 l
Z e
DEI ea.
L i:2 us o n
o s
1-l TABLE 3.3.2-1 (Continued)
_lSQLAT,10N ACTUAT!cr. INSTRUMENTATION I
N07ES When handling irradiated fuel in the secondary containment (a)
See Specification 3.6.3.1 Table 3.6.3-1 for valves in each valve group.
(b)
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> f or required surveillance without placing t he t rip syst em in the t ripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.
(c)
With only one channel per t rip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip runc t ion t o occur.
In t hese cases, the inoptrable channel
.shall be rest ored t o OPERABLE st 8tus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION I
requi red by Table 3.3.2-1 f or t hat Trip function shall be taken.
(d)
A channel is OPERABLE it 2 of 4 instruments in that channel are OpCRA BLE.
(e)
Wit h reac tor st eam pressure 3 $00 psig.
1 (f).
Closes only RWCU outlet isolation valve.
(g)
Alarm only.
(h)
Isolates containmtnt purge and vent valves.
(i) sees not isolate Ell-F015A.B.
(j)
Does not isolate B32-0019 or B32-F020.
(k)
Valve isolation depends upon low steam supply pressure coincident wit h high drywell pressure.
(1)
Secondary containment isolaticn dampers as listed in Table 3.6.5.2-1.
BRUNSWICK - UNIT 1 3/4 3-17a Amendment No. 149
~
TAPLE 3.3.2-2 E'
E ISOLATION ACTUATTON INSTRUMENTATION SETPOTNTS P
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ALLOWABLE e
-2 RIP FUNCTION TRIP SETPOINT VALUE E
1.
PRIMARY CONTAINMENT ISOLATION
~
-a a.
1.
Low, Level 1 3 + 162.5 inches (a) 162.5 inches (a }
+
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2.
Low, Level 3 3 + 2.5 inches *}
3 + 2.5 inches ")
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b.
Dryvell Pressure - High
< 2 psig
$ 2 psig c.
Radiation - High
$3m full power 1 3.5 m full p >we r v
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Pressure - Low 3 825 psin 3 825 psig 3.
Flow - High 1 140% of rated flow
$ 140% of rated flow d.
Main Steam Line Tunnel Temperature - High i 200*F
$ 200*F e.
Condenser Vacuum - Low 37 inches Hg vacuum
> 7 inches Hg vacuum f.
Turbine Building Area Temperature - High
$ 200*F
$ 200*F l
g.
Main Stack Radiation - High (b)
(b) h.
Reactor Building Exhaust Radiation - High
$ 11 mr/hr 1 11 mr/hr l
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i TABLE 3.3.2-2 (Continued) l ISOLATION ACTUATION INSTRUMENTATION SETPOINTS f
E ALLOWABLE
~o TRIP SETPOIPIT VALUE TRIP FUNCTION t
4.
CORE STANDBY COOLING SYSTEMS ISOLATION jE e
High Pressure Coolant Injection System Isolation a.
1.
HPCI Steam Line Flow - flich
$ 3001 of rated flow
$ 3001 of rated flow l
f 2.
HPCI Steam Line Flow - High Time Delay Relay 3$t $ 7 seconds 3$t $ 12 secondr.
100 psig
_> 100 psig 3.
IIPCI Steam Supply Pressure - Low 4.
HPCI Steam Line Tunnel Temperature - Ifigh
$ 200*F
$ 200*F v
7:
5.
Bus Power Monitor NA NA 6.
IIPCJ Turbine Exhaust Diaphragm c)
Pressure - High
$ 10 psig
$ 10 psig 7.
HPCI Steam Line Ambient Tempera t ure - High
$ 200*F
$ 200*F l
8.
IIPCI Steam Line Area a Temperature - High
$ 50*F
$ 50*F l
9.
HPCI Equi pment Area Tesperature - High
$ 175*F
$ 1/5*F l
10.
Drywell Pressure - fligh 1 2 psig
$ 2 psic l
- = 3 M) O m
On z
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m M
%2 55 2
4 TABLC 3.3.2-2 (Cont inued)
=
- =
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS m
i 35 n
ALLOWAR LE TRIP FUNCTION TRIP SETPOINT VALUE 4
CORE STANDBY COOLING SYSTEMS ISOLATION (Cont inued)
~i b.
Reactor Core Isolation Cooling System Isolation I.
RCIC Steam Line Flew - High
$ 3001 of rated flow
$ 3001 of rated flow 2.
RCIC Steam Line Flow - liigh l
Time Delay Relay 3$t $ 7 seconds 3$t $ 12 sctonds 3.
RCIC Steam Supply Pressure - Low
,3 50 peig 3
'a nsig 4
RCIC Steam Line Tunnel 7emperaturc - High
$ 175*F
$ 175'F l
w R
5.
Bus Power Monitor NA NA w
4 6.
RCIC Turbine Exhaust Diaphragm
~
Pressure - High
$ 10 psig
$ 10 psig 7.
RCIC Steam Line Ambient Temperature - High
$ 200*F
$ 200*F l
8.
RCIC Steam Line Area e Temperature - High
$ 50*F
$ 50*F l
9.
RCIC Equipment Room Ambient Temperature - High
$ 175'F
$ ly$*F l
l 10.
RCIC Equipment Room A Temperature - liigh v> a
~< $0*F
< 50*F l
ca
~
E II.
RCIC Steam Line Tunnel Temperature - High Time Delay Relay
-< 30 minutes
- 30 minutes z
12.
Drywell Pressure - High
$ 2 psig
$ 2 psig en M
_sa
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1 TABLE 3.3.2-3 ISOLA 110N SYSTEM INSTRUMENTATION RLSPONSE TIME TRIP FUNCTION RESI'ONSE TIME (Seconds)(a)(e)l 1.
PRIMARY CON 1 AINMENT !$0LATION a.
Reacto Vessel Wa te r Level -
1.
Lo e, Level 1 313 (d) 2.
Le w. Level 3
( l. gI )
713 b.
Drywell Pressure - High
$13 c.
Main St eam Line 1.
Radiat ion - Higb(l')
<1.0 ldA 313(f) l 2.
Pre 6sure - Low 313 (0. Id) 3.
Flow - liigh
[13gII l
d.
Main Steam Line Tunnel Temprature - High
$13 e.
Condenser Vacuum - Low (13 f.
Turbine Building Area Temperature - High NA g.
Main Stack Radiation - HighIDI
< l.O(d) h.
Reactor Building Exhaust Radiation - Higb(b)
NA l
2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Exhaust Radiation - Higb(b) cg3 i
b.
Drywell Pressure - High
$13 c.
Reactor Vessel Water Level - Low, Level 1
$13 l
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High
<45(C) l b.
Area Temperature - High
$13 c.
Area Ventilation a Temperature - High
$13 l
d.
SLCS Initiation NA e.
Reactor Vessel Water Level - Low, Level 2 313 l
f.
6 Plow - High - Time Delay Relay NA l
BRUNSWICK - UNIT 1 3/4 3-23 Amendment No. gg, 722, 172, 149
o s
t TABLE 3.3.2-3 (Cont ir utd) i ISOLAl!ON S)SifM INSTRUMENTATION RESPONSE T!g i
TRIP FUNCTION RESPONSE TIME (Seconds)(a)(c) l 4.
00RC STANDRY COOLINC SYSTEMS ISOLATION i
a.
High Pressure Coolant Injection System Isolation 1.
HPCI Steam Line Flow - High
$13(C) l 2.
HPCI Steam Line flow - High Time Delay Relay NA l
3.
HPCI Steam Supply Pressure - Low
$13 4
HPCI Steam Line Tunnel Temperature - High
$13 5.
Bus Power Monitor NA 6.
HPC1 Turbine Exhaust Diaphragm Pressure - High NA 7.
HPCI Steam Line Ambient Temperature - High NA 8.
HPCI Steam Line Area o Temperat ure - High NA l
9.
HPCI Equipment Area Temperature - liigh NA l
10.
Drywell Pressure - High NA l
b.
Reactor Core isolation Cooling System isolation
- 1.. RCIC Steam Line Flow - High
$13(C) l 2.
RCIC' St eam 1,ine Plow - High Time Delay Relay NA 3.
RCIC St eam Supply Pressure - Low-NA 4.
RCIC Steam Line Tunnel Temperature - liigh NA l
Bus Power Monitor k'A 6.
RCIC Turbine Exhaust Diaphram Pressure - High NA 7.
RCIC Steam Line Ambient Temperature - High NA S.
RCIC Steam Line Area o Temperat ure - High NA l
9.
RCIC Equipment Room Ambient Temperature - High NA l
10.
RC4 C Equipment Room a Temperature - High NA l
11.
RCIC Steam Line Tunnel Temperature - High NA Time Delay Relay 12.
Drywell Pressure - High NA BRUNSWICK - UNIT 1 3/4 3-24 Amendment No. gg, 73g,.
149
.=
9 4
1 ABLE 3.3.2-3 (Cont inued)
_1_ SOLATION SYS1FM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RE:SPONSE TlHf; (Seconds)(a)(c) l SHUTIOWN CCCLINC SYSTEM ISOLAilON a.
kcattor Vessel Water Level
- Low, Level 1 NA ti.
Reactor Steam Done Pressure - High NA BRUNSWICK - UNIT 1 3/4 3-25 Amendment No. $0, 115, 122, 130, 149
e i
I 1
TAbtE 3. 3.2-3 (Continued )
ISOLAllON SYSTEM INSTRUHfNTATION RESPONSE TIME l
NOTES
)
(a)
The isolation system instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
Isolation system instrumentation response time specified includes any delay for diesel generator starting assumed in the accident analysis.
(b)
Radiation monitors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electroiic component in the channel.
(c) includes t ime delay added by t he t ime delay relay.
(d)
Isolat ion actuat ion inst rumentat ion response t ime f or MSlVs only. No diesel gen rat or delays assumed.
(e)
Isolation system instrumentation response time specified f or the Trip function actuating each valve group / damper shall be added to the isolation time f or valves in each valve group shown in Table 3.6.3-1 and secondary containment isolat ion dampers shown in Table 3.6.5.2-1 to obt ain IS01.ATION SYSTEM RESPONSE TIME f or each valve / damper.
(1)
Isolation system instrumentation response time f or associated valves except MSIVs.
BRUNSWICK - UNIT 1 3/4 3-26 Amendmen t No. "", 122, 130, 149
TABLE 4.3.2-1 h
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS vi3 CHANNEL OPERATIONAL Q
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH
~
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRET)
E 1.
PRIMARY CONTAINMENT ISOLATION Q
a.
1.
Low, Level 1 Transmitter:
NA(a)
NA R(b)
- 1. 2, 3 Trip Logic:
D H
M 1, 2, 3 2.
Low, Level 3 NA *I NA R
1, 2, 3 I
IdI Transmitter:
Trip Logic:
D M
M 1, 2, 3 b.
Drywell Pressure - High y
Transmitter:
NA NA R{g 1, 2, 3 Trip Logic:
D M
M I, 2, 3 c.
1.
Radiation - High D
W R(d) 3 7,,
w 4
2.
Pressure - Low Transmitter:
NA(a)
NA R(b) y
+
Trip Logic:
D M
M 1
3.
Flow - High Transmitter:
NA(a)
NA R(b) j Trip Logic:
D M
M 1
j d.
Main Steam Line Tunnel Temperature - High NA M
R 1, 2, 3 1
e.
Condenser Vacuum - Low 4
y gg Transmitter:
NA NA R
1, 2 Oy Trip Logic:
D M
M I, 2 *I I
me E
f.
Turbine Building Area
~
Temperature - High NA M
R 1, 2, 3 l
- E g.
Main Stack Radiation.High NA Q
R
- 1. 2, 3 h.
Reactor Building Exhaust
[
Radiation - High D
. M R
I, 2, 3 l
P 4
TABLE 4.3.2-1 (Cont inued )
es h
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E3 CHANNEL Q
OPERATIONAL CII/sNNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED E
2.
SECONDARY CONTAINMENT ISOLATION N
a.
Reactor Building Exhaust Radiation - High D
M R
I.2.3.5, andI I. I l
b.
Drywell Pressure - Iligh Transmitter:
NA(a) g4 p(b)
I, 2, 3 Trip Logic D
M M
1, 2, 3 c.
Reac t or Vessel Water Level -
Low, Level 2 Transmitter:
NA(*
NA R(b)
I, 2, 3 Trip Logic:
D M
M 1, 2, 3 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
a Flow - High D
M R
1, 2, 3 y
~
b.
Area Temperature - High NA M
R 1, 2, 3 Area Ventilation a Temperature - High NA M
R 1, 2, 3 l
c.
d.
SLCS Initiation NA R
NA 1, 2, 3 c.
Low, Level 2 Transmitter:
NA(a) g4 p(b)
- 1. 2, 3 Trip Logic:
D M
M 1, 2, 3 f.
A Flov - High - Time Delay Relay NA M
R 1, 2, 3 l
$$ $il R
E z
P
~
E
i 0
t s
TABLE 4.3.2-1 (Continued)
E 1
E ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMDITS E
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED E
G 4
CORE STANDBY COOLING SYSTEMS ISOLATION
)
a.
High Pressure Coolant Inject ion Syst em Isolation 1.
HPCI Steam Line Flow - High Id}
Transmitter:
NA NA R(IN 1, 2, 3 Trip Logic:
D M
M I,2,3 I
2.
HPCI Steam Line Flow - High l
Tine Delay Relay NA R
R 1, 2, 3 w
3.
HPCI Steam Supply Pressure - Low NA M
R I, 2, 3 N
w 4
HPCI Steam Line Tunnel O
Temperature - High NA M
Q 1, 2, 3 e
5.
Bus Power Monitor NA R
NA 1, 2, 3 6.
HPCI Turbine Exhaust Diaphragm Pressure - High NA M
Q 1, 2, 3 4
7.
HPCI Steam Line Ambient Temperature - High NA M
R 1, 2, 3 l
8.
HPCI Steam Line Area a Temperature - Ifigh NA M
R 1, 2, 3 l
u)
E 9.
HPCI Equipment Area S
Temperature - High NA M
Q 1,2,3 l
3 z
10.
Drywell Pressure - liigh Transmitter:
NA(a) pg p(b)
I,2,3 g
Trip Logic:
D
~ M M
1, 2, 3 0
4
.m m
4 TABLE 4.3.2-1 (Continued)
E ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E.
CHANNTL OPERATIONAL Q
CHANNEL FUNCTIOf4AL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED E
4.
CORE STANDBY COOLING SYSTEMS ISOLATION (Continued)
~+
b.
Reactor Core Isolation Cooling System Isolation m.
4 1.
RCIC Steam Line Flow - High NA "}
NA-R(b) 1, 2, 3 I
Transmitter:
Trip Logic:
D M
M 1, 2, 3 2.
RCIC Steam Line Flow - High l
Time Delay Relay NA P
R 1, 2, 3 3.
RCIC Steam Supply Pressure - Lov NA M
Q 1, 2, 3 4.
RCIC Steam Line Tunnel t',
Temperature - High NA M
R 1, 2, 3 l
v 5.
Bus Power Monitcr NA R
NA 1, 2, 3 0
6.
FCIC Turbine Exhaust Diaphragm
+
Pressure - High NA M
R 1, 2, 3 7.
RCIC Steam Line Ambient Temperature - High NA M
R I, 2, 3 l
8.
RCIC Steam Line Area
& Temperature - High NA M
R 1, 2, 3 l
{
i 9.
RCIC Equipment Room Ambient j
Temperature - High NA M
Q 1, 2, 3 l
l i' y 10.
ECIC Equipment Room A Temperature - High NA M
Q 1, 2, 3 l
o.
11.
RCIC Steam Line Tunnel Temperature - High E
Time Delay Relay NA M
R 1, 2, 3 5
12.
Dryvell Pressure - High i
Transmitter:
NA(a)
}g Trip Logic:
D NA R(b) g, y, y M
M 1, 2, 3 D
ya
e TAELE 4.3.2-1 (Cont inued )
ISOLATION ACTUATION INSTRUMf MT ATION SURVEILLANCE REQUIREMENTS l
NOTES l
(a) The transmitter channel check is satisfied by the trip unit channel chcck.
A separate t ransmitter check is not required.
(b) Transmitters are exempted f rom the monthly channel calibration.
(c) If not performed within the previous 31 days.
(d) Testing shall verify that the mechanical vacuum pump rips and the mechanical vacuum pump line valve closes.
(e) When reactor steam pressure > 500 psig.
1 (f ) When handling irradiat ed f uel in t he secondary containment.
l BRUNSWICK - UNIT 1 3/4 3-32 Amendment No. 87, 95, 730, 149
o TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES Table 3.6.3-1 has been deleted.
Refer to Plant Procedure RCl-02.6.
Fages 3/4 6-15 through 3/4 6-17 han been deleted.
(Next page is 3/4 6-18) l BRUNSWICK - (TNIT 1 3/4 6-14 Amendment No.
149
=- ~~.. - -.- -
CONTAINMENT SYSTEMS SECONDARY CONI AINHENT AUTOMATIC ISOLATION DAMPl:RS j
LIMITING CONDITION FOR OPERATION 3.6.$.2 The secondary containn.ent automatic isolation dampers shown in Table 3.6.$.2-1 shall be OPERABLE.
APPLICABILITY OPERATIONAL CONDITIONS 1, 2, 3, 5, and *.
l ACTIONI With one or more of the secondary containment isolation dampers specified in Table J.b.$.2-1 inoperable, operation may cont inue and the p rovisions of Specitication 3.0.4 are not applicable, provided that at, east one isolation damper is neintained OPERABLE in each af fected penetration that is open, andl a.
The inoperabir damper is restored to OPERABLE 6tatus within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or
.b.
The affected penetration is isolated by use of a closed damper within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or c.
SECONDARY CONTAINMENT INTECRITY is demonstrated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and the damper is restored to OPERABLE st at us within 7 days.
Otherwise, in OPERATIONAL CONDITIONS 1, 2, or J, be.n at least HOT l
SHUTDOWN wi t hi n t he ne xt 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in CF.D SHUTDOWl: within the i
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Ot herwise, in OPERATIONAL CONDITION $ or *, suspend irradiated fuel l
handling.in the secondary containment, C0kE ALTERATIONS, or activities that could reduce the SHUTDOWN MARCIN. The provisions of Specification 3.0.3 are not applicable.
- When irradiated fuel is being handled in the secondary containment.
BRUNSWICK - UNIT 1 3/4 6-22 Amendment No. 149
5 MBLE 3.6.5.2-1 P
SECONDARY CONTAINMENT AUTOHATIC_ ISOLATION DAMPERS 4
3
. Table 3.b.5.2-1 has beth delet ed.
Ref er to Plant Procedure RCl-02.6.
. k.
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BRUNSWICK - UNIT 1 3/4 6-24 Amendment No.149
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l CORTAIRMENT SYSTEMS j
BASES 3/4.6.3 PRTHARY CONTAINMENT !$0LATION VALVES (Cont inued )
A list of automatic closing primary containment isolation valves and their associated closure times shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50.
The addition and deletion of primary i
contairment isolation valves shall be made in accordance with Section $0.$9 of 10 CFR Part 50.
3/4.6.4 VACUUM REllEF Vacuum relief breakers are provided to equalize the pressure between the drywell and suppression pool and the suppression pool and reactor building.
This system wiil maintain the st ructural integrity of the containment under conditions of large differential pressures.
The vacuum breakers between the drywell and the suppression pool must not be tnoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. There are an adequate number of valves to provide some redundancy so that operation nay continue with no more than 2 vacuum breakers inoperable and secured in the closed position.
Each set of vacuum relici valves between the suppression chamber and reactor building provides 1001 relief, which may by required in the unlikely event that negative pressures develop in the primary containment.
The Nitrogen Backup System provides backup motive power for these suppression pool-reactor building vacuum breakers on a loss of instrument air.
The normal non-interruptible instrument air system for these vacuum breakers is designed as a Seismic Class I system supplied by air compressors powered from the emergency buses.
The Nitrogen System serves as a backup to that air system and thus the loss of the Nitrogen System, or portions thereof, does not make the vacuum breakers inoperable. The design allows f or the out of service times in Actions b and c.
The Nitrogen Backup System is added to the Suppression Pool-Reactor Building Vacuum Breaker specification to satisf y NRC concerns relative to 10 CFR $0.44(c)(3) as addressed in the Brunswick Safety Evaluation Report dated October 30, 1986 concerning Generic Letter 84-09.
Pressurization to 1130 psig assures suf ficient system capacity to provide 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation with design valve actuation and system leakage.
3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioact ive material which may result from an accident. The reactor building provides secondary containment during normal operation when the drywell is sealed and in service.
When the reactor is shut down, or during ref ueling, the drywell may be open and the ranctor building then becomes the primary containment.
BRUNSWICK - UNIT 1 B 3/4 6-5 Ameno.nent No. 76, fif, 149
O I
CONTAINMENT SYSTEMS i
BASES (Continued) 3/4.6.$
SECONDARY CONTAINMENT (Cont inued )
{
Establishing and maintaining a vacuum in the building with the standby gas treatment syst em, once per 15 nont h6 Along with the surveillance of the 4
valves, is adequate to ensure that there are no violations of the integrity of the secondary centainment.
A list of secondary containment automatic isolation dampers shall be available at the plant in accordance with Section 50.71(c) et 10 CFR Part 50.
The addition and deletion of secondary containment automatic isolation dampers shall be made in accordance with Section 50.59 of 10 CfR Part 50.
3/t. 6.6 CONTAINMENT ATMOSNIERE CONTROL l'he OPEPABILITY ot the containn,ent iodine filter t rains ensures that
)
sufficient iodine removal capability will be available in the evcnt of a LOCA. The reduc t ion in cont ainment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage.
The operation of this system and resultant iodine removal capacity are consistent with the assumpt ions used in the LOCA analyses.
The OPERABILITY of the equipment and systen;s required for the detection and cont rol of hydrogen pas ensures that t hi s equi pment will De available to maintain the hydrogen concentrat ion within containment below its flannable limit du ring pos t-LOCA condi t ions.
The containment inerting system is capable of cont rolling t he expected hydrogen generat ion associat ed wit h 1) zirconium-water reactions, 2) radiolytic decomposition of water, and 3) corrosion of me t al s wit hin cont ainment.
The hydrogen control system is consistent with t he recommendations of Regulatory Guide 1.7, " Cont rol of Combust ible Cas Concentrations in Containment following a LOCA."
BRUNSWICK - UNIT 1 B 3/4 6-6 Amendment No.149
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i CAROLINA POWER & LIGHT COMPANY, et al.
DOCKET NO. 60 324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICEkSE Amendment No.179 License No. DPR 62 3.
The Nuclear Regulatory Comission (the Comission) has found that:
F A.
The-application for amendment filed by Carolina Power & Light Company
)
-(the' licensee),datedFebruary 29.-1988,' superseded September 20,
'3989, as supplenented Deceinber 5.1989, February 15, August.9, and October 24, 1990, complies with the standards and requirements of i
L-the Atomic' Energy Act of 1954, as amended (the Act), and the.
Comission's rules and regulations set forth in 10 CFR Chapter I; 1
B.
The facility will operate in conformity with the application, the l
provisions of the_Act, and the rules and regulations of the' s
Comission; C.
There'1s' reasonable' assurance: (1)'thattheactivitiesauthorized by this amendnent can be conducted without endangering the health p
and safety of the public, anc (ii) that such activities will be conducted in-compliance with the Comission's regulations;7 D.
The issuance of this amendment will no't 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 $1 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 Facility Operating License No. DPR-62 is hereby amended to read as follows:
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l (2).3chnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.179, are hereby incorporated in the license. Carolina Fower t Light Company shall operate the freility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMIS$10N Original Signed By:.
Elinor G. Adensam. Director Project Directorate 11 1 Division of Reactor Projects. 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications, Date of Issuance: December 5, 1990 A
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e ATTACHMENT TO LICENSE AMENDMEhT NO. 179 FACILITY OFERATING LICENSE NO. DPR-62 DOCKET NO. 50 324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by matginal lines.
Remove Pages insert Pages XI XI 1-5 1-5 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-17a 3/4 3-17a 3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 3/4 3-21 3/4 3-21 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-32 3/4 3-32 3/4 6-14 3/4 6-14 3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 3/4 6-17 3/4 6-17 3/4 6-22 3/4 6-22 3/4 6-24 3/4 6-24 8 3/4 6-5 6 3/4 6-5 8 3/4 6-6 B 3/4 6-6
i e
i INDEX i
BASr$
SECTION PACE 3/4.4 REACTOR C001. ANT SYSTEM (Continued) 3/4.4.4 C li E M I S T R Y............................................ B 3 / 4 4 - 2 3/4.4.$
SPECIFIC ACTjVITY.................................... b 3/4 4-2 3/4.4.6 PRESSURE / TEMPERATURE LI MI TS.......................... B 3 /4 4-3 3/4.4.7 M A) N STEAM LI NE I SOLATI ON VALVES..................... B 3 /4 4-7 3/4.4.8 S TR UCTU RA L I N TE CR I T Y................................. B 3/4 4-7 3/4.5 EMERCENCY CORE COOLING SYSTEM 3/4.5.1 HICH PRESSURE C001. ANT I NJ ECT!M :iYSTEM............... B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS).............. B 3/4 5-1 3/4.5.3 LOv! PRES S UR E C00L I NC S Y STEM S......................... B 3 /4 5-2 3/4.5.4 S U P PR ES S 10h P00 L..................................... B 3 / 4 5 -4 3/4.6 CONTAINMENT SYSTl:MS 3/4.6.1 PRIMARY CONIAINMENT.............................
.... B 3/4 6-1 3/4.6.2 DEPRESSURIZ ATION AND COOLING SYSTEMS................. B 3 /4 6-3 3/4.6.3 PRI MARY CONTAINMENT I SOI.ATION V ALVES................. B 3 /4 6-4 3/4.6.4 V AC UUM R E L I E F........................................ B 3 / 4 6 - 5
=3/4.6.5 S ECON DARY CONTA I N ME NT................................ B 3 / 4 b -5 3/4.6.b CONTAINMENT ATMOS PHER E CONTR0L....................... B 3 /4 6-b l
3/4.7 PLANT SYSTEMS 3/4.7.1 S ERVI CE W ATER S YSTEMS................................ B 3 /4 7 -1 3/4.7.2 CONTROL ROOM EMERGENCY f! LTRATION S YSTEM............. B 3 /4 7-la BRUNSWICK - UNIT 2 XI Ame ndme nt No.-f63, 179
e e
DE r.t NI TI ONS OFFSITE LOSE CALCULATION M ANUAL (ODCM)
The OFFSITE 105C CALCULATIONAL NANUAL (ODCM) is a manual which contains the current methodology and parameters to be used to calculate of f site doses resulting f rom the release of radioactive gaseous and liquid ef fluent s! the methodology to calculate gaseous and liquid effluent monitoring inst rumentation ala rm/t rip set point s! and, the requirements of the environmental radiological monitoring program.
OPEkABLE - OPERABILITY A system, subsystem, t rain, component, or device shall be OPERABLE or have OPERABILITY when it is capable of perf orming its specified f unction (s).
Implicit in this definition shall be the assumption that all necessary attendant inst rument at ion, control s, nornal and emergency elec t ric power sources, cooling or seal water, lubrication or other auxiliary equipment that are required f or the system, subsystem, t rain, component, or device to perf orm its f unction (s) are also capable of perf orming their related support function (s).
OPERATIONAL CONDITION An OPERATIONAL CONDl'Il0N shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.
PHYSICS TESTS PHYSICS TESTS shall be those tests perf ormed to measure the fundamental nuclear charact;sistics of the reactor core and related instrumentation and are 1) described in Section 14 of the Updated FSAR. 2) authorized under the provisions of 10 CFR 50,59, or 3) otherwise approved by the Commission.
PRES $URE BOUNDARY LEAKACE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable f ault in a reactor coolant system component body, pipe wall, or vessel wall.
PRIMARY CONTAINMENT INTECRITY PRIMARY CONTAINMENT INTECRITY shall exist whent All penetrations required to be closed during accident conditions are a.
eithert 1.
Capable of being closed by an OPERABLE containment automatic isolat.on valve system, or 2.
Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1.
l t
BRUNSWICK - UNIT 2 1-5 Amendment No. $$, f49, 179
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s i
o e
u E
s e e
s Li r
a d -
R l
M sL L
e t
u i n N
e r
ma s
k uo I
V,
P ai s
u c
Bi A
v v
ed e
a e
s a
t T
r o o
l t a r
l 1
t e s
ee t
ra N
oL L
l SR P
F F
S p n
np S
oi O
t e
m e
i m
t d N
C c
v n
ne d
b e n
ca O
a y
i iT n
rT i
aR I
Y e.
r a.
a o
u a
e T
R R1 2
D M1 2
3 4
M C
T M
R C
A N
M U
I F
R P
a b
c d
e f
g h
P I
R T
1 n T M ' E;: u y,o EEe:
F af-2:
M.
D ;;n [,
L l
~
E TABLE 3.3.2-1 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION N
VALVE CROUPS MINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL j
E TRIP FUNCTION SICWAL(a)
PEP TRTP SYSTEM (b)(c) CONCITION ACTION
~I o
2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Exhaust (1) 1 1, 2, 3 23 l
Radiation - High and
- 6 1
- 1. 2. 3 20 l
b.
Drywell Pressure - High (1) 2 1,2.3 23
- 2. 6 2
1, 2. 3 20 c.
(1) 2
- 1. 2 3 23 w
Low, Level 2 3
2
- 1. 2. 3 24 N
w 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION 1
A Flow - High 3
1 1, 2, 3 24 a.
b.
Area Temperature - High 3
2
- 1. 2. 3 24 c.
Area Ventilation A Temperature - High 3
2 1, 2, 3 24 l
II d.
SLCS Initiation 3
NA 1, 2. 3 24 i
e.
l Low, Level 2 3
2
- 1. 2. 3 24
=>
'" 5 f.
A Flow - High - Time Delay Relay NA I
I, 2, 3 24 l
-+ E RI!
1 Zz
,n o, e
_a Gl*
EE TABLE 3.3.2-1 (Continued) i E!
E ISOLATION ACTUATION INSTRUMEN' ATION f
T E
7:
VALVE GROUPS MINIMUM NUMBER APPLTCABLE i
OPERATED BY OPERABLE CHANNELS OPERATIONAL l
E TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)(c) CONDITION ACTTON
+
s u
4.
CORE STANDBY COOLING SYSTEMS ISOLATION 4
High Pressure Coolant Inject ion Syst em Isolation a.
1.
HPCI Steam Line Flov - High 4
I I, 2, 3 25 2.
HPCI Steam Line Flow - High l
Time Delay Relay NA 1
1, 2. 3 25 3.
HPCI Steam Supply Pressure - Lov 4
2 1, 2, 3 25 Ik}
4 w
7
}
I, 2. 3 25 l
i
- 2 t
u 4
HPCI Steam Line Tunnel l
2.
Temperature - High 4
2 1,2,3 25 I
s~
IR i
5.
Bus Power Monitor NA 1/ bus 1, 2, 3 26 i
6.
HPCI Turbine Exhaust l
Diaphragm Pressure - High 4
2 1, 2, 3 25 t
7.
HPCI Steam Line Ambient Temperature - High 4
I, 2, 3 25
[
8.
HPCI Steam Line Area
- g a Temperature - High 4
1 1, 2, 3 25 l
t ea e
~ E 9.
HPCI Equipment Area
$5 Temperature - High 4
1 1, 2, 3 25 sa o
- e Ik}
I 1, 2, 3 25 l
f
-^ z 10.
Drywell Pressure - High 7
u>.
%d i
SQ I
l s
f 42 l
I e
a
=
i or w
TABLE 3.3.2-1 (Continued)
Ey ISOLATION ACTUATION INSTRUMENTATION
~n 7:
VALVE CROUPS MINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNEuS OPERATIONAL TRIP FUNCTION SICNAL(a)
PER TRIP SYST'4M(b)(c ) CONDITION ACTION 5
4.
CORE STANDBY COOLING SYSTEMS ISOLATION (Continued) u b.
Reactor Core Isolation Cooling System Isolation 1.
RCIC Steam Line Flev - High 5
1 1, 2, 3 25 2.
RCIC Steam Line Flov - High l
Time Delay Relay NA I
1, 2, 3 25 3.
RCIC Steam Supply Pressure - Low
?[()
2
- 1. 2 3 25 I
1, 2, 3 25 4.
RCIC Steam Line Tunnel
}*
Temperature - High 5
2 1, 2, 3 25 IE}
5.
Bus Power Monitor NA 1/ bus 1, 2, 3 26 6.
RCIC Turbine Exhaust Diaphragm Pressure - High 5
2 1, 2, 3 25 7.
RCIC St 1.ne Ambient Temper.-
.e - High 5
1 1, 2, 3
25 l
8.
RCIC Steam Line Area a Temperature - High 5
1 1, 2, 3 25 l
9.
RCIC Equipment Room Ambient i
Temperature - High 5
I I, 2, 3 25 l
~20{
10.
RCIC Equipment Room a Temperature - High 5
1 1, 2, 3 25 l
3 wa
$55 11.
RCIC Steam Line Tunnel NA 1
I, 2, 3 25 E
Temperature - High
~
$$z Time Delay Relay G2 O
~ '
12.
Drywell Pressure - High 9(k)
I 1, 2, 3 25 m en M3=
=
1 i
E-TABLE 3.3.2-1 (Continued)
E-E ISOLATION ACTUATION INSTRUMENTATION 5*
VALVE CkOUPS MINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL E3 TRIP FUNCTION SICNAL(a)
PER TRIP SYSTEM (b)(c) CONDITIGd ACTION s
u 5.
SHUTDOWN COOLING SYSTEM ISOLATION 2
~
a.
2, 6 2
1,2,3 20 Low, Level 1 8
2 1, 2. 3 27 II) b.
Resctor Steam Dome Pressure - High 8
I I, 2, 3 27 l
E asy
.em.
i 7
"3
.~
~
... ~.
vi TABLE 3.3.2-1 (Cont inued )
I SOLATION AC'UATION I NSTRUMENTATION r
NOTES When handling irradiatrd fuel in the secondary containment.
(a)
See Speci fication 3.6.3.1, Table 3.6.3-1 f or valves in each valve group.
(b).
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required-surveillance without placing the t rip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.
(c)
With only one-channel per t rip system, an inoperable channel need not be placed in the. tripped condition where this would cause the Trip Function to occur.
In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
L(d)
A channel is'0PERABLE it' 2 of 4 instruments in that channel are OPE RABLE.
(e)
Wit h reactor steam pressure > 500 psig.
(f)
Closes only RWCU outlet isolation valve.
L(g)
Alarm only.
(h)
Isolates containment purge and vent-valves.
(i)
Does not isolate Ell-F015A,B.
(,) )
Does not isolate B32-F019 or B32-F020.
(k)
Valve isolation depends upon low steam supply pressure coincident with high drywell pressure.
(1)
Secondary containment isolation dampers as listed in Table 1.6.5.2-1.
4 BRUNSWICK - UNIT 2 3/4 3-17a Amendment No. 179
E TABLE 3.3.2-2 E
E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS Fi*
ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE E
.5 1
PRIMARY CONTAINMENT ISOLATION u
a.
' Reactor Vessel Water Level -
1.
. Low, Level 1 162.5 inches (a)
_> + 162.5' inches (*)
>+
2.
Lou, Level 3 3 + 2.5 inches (a:
+ 2.5 inches ("}
b.
Drywell Pressure - High
$ 2 psig 1 2 psig c.
Radiation - liigh
$3x full wer
$ 3.5 x fu power background {l background os 5
w 2.
Pressure - Low 3 825 psig 3 825 psig 3.
Flow - High
$ 140% of rated flow
$ 140% of rated flow 4
Flow - High 1 40% of rated flow 1 40% of rated flow d.
Main Steam Line Tunnel Temperature - High
$ 200*F
$ 200*F e.
Condenser Vacuum - Low 37 inches flg vacuum 3 7 inches Hg vacuum f.
Turbine Building Area Temperature - High
$ 200*F
$ 200*F l
t g.
Main Stack Radiation - liigh (b)
(b) s= a ea e
~ E h.
Reactor Building Exhaust Radiation - liigh
- 11 mr/hr
< 11 mr/hr 8
- m. 3 l
" Z
+
.m y
^-
, ~. ',
N TABLE 3.3.2-2 (Continued) ll E
E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS
~n
. ALLOWABLE F-TRIP SETPOINT
_ VALUE TRIP FUNCTION Ey 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Exhaust Radiation - High
$ II mr/hr 5 11 mr/hr u
b.
Drywell Pressure - High
$ 2 psig
$ 2 psig I
I c.
Reactor Vescel Water Level - Low, Level 2 3 + 112 inches ")
3 + 112 inches "
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High
$ 53 gal / min 5 53 gal / min b.
Area Temperature - liigh
$ 150*r
$ 150'F Area Ventilation A Temperature - High 5 50*F
$ 50*F l
c.
d.
SLCS Initiation NA NA e.
Reactor Vessel Water Level - Lou, Level 2
> + 112 inches (a )
> + II2 inche.(a) f.
A Flou - High - Time Delay Relay
$ 45 seconds
$ 45 seconds g
wa >
. o Q.
l sq a
.s$
n Z
$A ?
N t%
q)*
r
~'
ag E
TABLE 3.3.2-2 (Continued)
.E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 5*
ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE E
Q 4
CORE STANDBY COOLING SYSTEMS ISOLATION u
a.
High Pressure Coolant Injectic, System Isolation 1.
HPCI Steam Line Flow - High 1 3001 of rated flow
$ 300% of rated flow s
2.
HPCI Steam Line Flow - High l
Time Delay Relay 3$t $ 7 seconds 3$t $ 12 seconds y
3.
HPCI Steam Supply Pressure - Low 3 100 psig 3 100 psig I
4.
HPCI Steam Line Tunnel Temperature - Iligh
$ 200*F
$ 200*F w
N 5.
Bus Power Monitor NA NA w
E 6.
HPCI Turbine Exhaust Diaphragm Pressure - High
$ 10 psig
$ 10 psin 7.
HPCI Steam Line Ambient Temperature - High
$ 200*F
$ 200*F l
8.
HPCI Steam Line Area a Temperature - High 5 50*F
$ 50*F l
9.
'llPCI Equipment Area Temperature - liigh
$ 175*F
$ 175*F' l
10.
Drywell Pressure - High
$ 2 psig
$ 2 psig l
E$ g>
en o
_a E
z b
M
=
E TABLE 3.3.2-2 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M*
ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE-E M
4.
CORE STANDBY COOLING SYSTEMS ISOLATION (Continued) o b.
Reactor Core Isolation Cooling System Isolation
- 1.. RCIC Steam Line Flow - High j 300 of rated flow
$ 3001 of rated flow 2.
RCIC Steam Line Flow - Iligh l
Time Delay Relay-3$t i 7 seconds 3$t i 12 seconds 3.
RCIC Steam Supply Pressure - Low 3 50 psin 2 50 psin 4.
RCIC Steam Line Tunnel Temperature - Iligh
$ 175*F
$ 175*F l
w v
5.
Bus Power Monitor NA NA w
E
~
6.
RCIC Turbine Exhaust Diaphragm Pressure - High 5 10 psig
$ 10 psig 7.
RCIC Steam Line Ambient Temperature - High 5 200*F
$ 200*F l
8.
RCIC Steam Line Area a Temperature - High 1 50 J
$ 50*F l
9.
RCIC Equipment Room Ambient Temperature - liigh
$ 175*F
$ 175'F l
10.
RCIC Equipment Room
- g a,
A Temperature - High
< $0*F
< 50*F l
n-
-gg @
11.
RCIC Steam Line Tunnel Temperature - High
$ 30 minutes 5 30 minutes A
Time Delay Relay OZ u) 12.
Drywell Pressure - High 3 2 psig
$ 2 psig E
~
- ~
~
TABLE 3.3.2-3 ISOLATION SYSTEH INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)(a)(e) l j
1..
PRIMARY CONTAINMENT ISOLATION a.
' Reactor Vessel Water Level -
1.
Low. Level 1 (13
< l. (d) 2.
Low, Level 3 213gI) b.
Drywell Pressure - High
$13 c.
Main Stearr. Line I
1.
Radiation - Higb(b)
~(i, (d)
_13 l
I}
2.
Pressure - Low
,l3 3 '. Flow - High (d)
<o, I) 313 l
313g(d) l 4.
Flow - High
<0.
f) d.
Main Steam Line Tunnel Temperature - High
$13 e.
Condenser Vacuum - Low
,13 f.
Turbine Building Area Temperature - High NA g.
Main St.ck Radiation - High(b)
<l.0(d)
ID) h.
Reactor. Building Exhaust Radiation - High NA l
2.
SECONDARY CONTAINME "t ISOLATION Reactor Building Exhaust Radiation - High(b)
<i3 a.-
b.
Drywell Pressure - High
$13' c..
Reactor Vessel Water Level - Low, Level 2
<l3 l'
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
.A Flow - High
$45(C) l b.
Area Temperature High-
$13 c.
Area Ventilation a Temperature - High
<l3 l
d.
SLCS Initiation NA e.
Reactor Vessel Water Level - Low, Level 2 313 l
l
.f.
6 Flow - High - Time Delay Relay NA t
BRUNSWICK - UNIT 2 3/4 3-23 Amendment No. 78, 97, 746, 162, 179
~
-1 TABLE 3.3.2-3 (Continued)
ISOIATION SYSTEM INSTRUMENTATION RESPONSE TIME RESPONSE TIME (Seconds)(a)(e)l TRIP FUNCTION 4.-
CORE STANDBY COOLING SYSTEMS 1 SOLATION a.
High Pressure Coolant injection System Isolation 1.
IIPCI Steam Line Flow - High
$13(C) l 2.-
liPCI Steam 1.ine Flow - lii gh T u..: Delay Relay NA l
3.
HPCI Steam Supply Pressure - Low
$M 4.
IIPCI St eam Line Tunnel Temperat ure - liigh
$13 5.
Bus Power Monitor NA 6.
IIPC f Turbine Exhaust Diaphragm Pressure - High NA 7.
IIPCI Steam Line Ambient Temperature - High NA 6.
IIPCI Steam Line Area a Temperature - High NA l
9.
IIPCI Equi pment Area Temperature - High NA l
10.
Drywell Pressure - High tJA l
b.
Reactor Core Isolation Cooling System Isolation 1.
RCIC Steam Line Flow - High
$13(c) l 2.
RCIC Steam Line Flow - liigh Time Delay Relay NA l
3._
RCIC Steam Supply Pressure - Low NA 4
RCIC Steam Line Tunne1 Temperature - High NA l
5 '. Bus Power Monitor NA
~
6.
RCIC Turbine Exhaust Diaphram Pressure - High NA 7.
RCIC Steam Line Ambient Tempe ra ture ' - liigh NA 8.
RCIC Steam Line' Area a Temperature - High NA l
9.
RCIC Equipment Room Ambient Temperature - High NA l
10.
RCIC Equi pment ' Room A Temperature - High NA l
11.
RCIC Steam Line Tunnel Temperature - High NA Time Delay Relay 12.
Drywell Pressure - liigh NA BRUNSWICK - UNIT 2 3/4 3-24 Amendment'No. 97, 160, 179
e TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME RESPONSE TIME' (Second s )(a )(e) l
.TRI P FLINCTION 5.
SHUTDOWN COOT.INC SYSTEM ISOLATION a.
React or Ves sel Wate r Level - Low, Level 1 NA b.
Reactor Steam Dome Pressure - High NA BRUNSWICK - UNIT 2 3/4 3-25 Amendment No. 46, 78, 97, 141, 142, 146, 160, 179
i
+
TABLE 3.3.2-3 (Continued )
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME NOTES (a)
The isolation system instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
Isolation system instrumentation response time specified includes any delay for diesel generator starting a.sumed in the accident analysis.
(b)
Radiation monitors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in the channel.
(c)
Includes time delay added by the time delay relay.
(d)
Isolation actuation inst rument at ion re sponse time for MSIVs only. No diesel generator delays assumed.
(e)
Isolation system instrumentation response time specified f or t he Trip function actuating each valve group / damper shall be added to the isolation time f or valves in each valve group shown in Table 3.6.3-1 and secondary containment isolation dampers shown in Table 3.6.5.2-1 to obtain IS0lATION SYSTEM RESPONSE TIME for each valve / damper.
(f)
Isolation system instrumentation response time for associated valves except MSIVs.
BRIJNSWICK - UNIT 2 3/4 3-26 Amendment No. ($, 78, 97, 160, 179
oo e
m
. TABLE 4.3.2-1 E
'y ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS R
x CHANNEL OPERATIONAL CHANNEL' FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED G
1.
. PRIMARY CONTAINHENT ISOLATION w
a.
1.
Low, Level'l
' Transmitter:
NA(a)
NA R(b) y, y, 3 Trip Logic:
D M
M
- 1. 2, 3 2.
Low, Level 3 Transmitter:
NA(a)
NA R(b)
- 7. 7 3 Trip Logic D
S M
1, 2, 3 b.
Drywell }lressure - liigh
)
b)
Transmitter:-
NA NA R
1, 2, 3 Trip Logic:
D M
M 1, 2, 3 c.
Main Steam Line R(d) 1.
Radiation - High D
W 1
2, 3 2.
Pressure - Lew Transmitter:
NA(a)
NA R(b) y Trip Logic:
D M
M 1
3.
Flow - High
)
y Transmitter:
NA NA R
1 Trip Logic:
D M-M 1
4 Flow - High D
M M
2, 3 d.
Main Steam Line Tunnel Temperature - liigh NA M
R 1, 2, 3 pa [j e.
Condenser Vacuum - Low
)
R(b) 1, 2(I n-Transmitter:
h' A NA e)
Trip Logic:
D M
M 1, 2 * )
u> g f.
Turbine Building Area z
Temperature - High NA M
R 1, 2, 3 l
o h$
g.
Main Stack Radiation - High NA Q
R 1, 2, 3 20 h.
~
Reactor Building Exhaust Radiation - High-D M
R 1, 2, 3 E?
E TABLE 4.3.2-1 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH E
TRIP FUNCTION CHECK TEST CALIBRATION
' SURVEILLANCE REQUIRED-U u
2.
. SECONDARY CONTAINMENT ISOLATION 4
Reactor Building Exhaust a.
II) l Radiation - High D
M R
1,2,3,5, and b.
Drywell Pressure - High I)
Transmitter:
NA NA R(b) 1, 2, 3 Trip Logic:
D M
M 1, 2, 3 c.
Low, Level 2' w
NA ")
NA R(b) 1 2 3 I
S Transmitter:
Trip Logic:
D M
M 1, 2, 3 w
0 3.
REACTOR WATER' CLEANUP SYSTEM ISOLATION a.
A Flow - High D
M R
1, 2, 3 b.
Area Temperature - High NA M
R 1, 2, 3 l
Area Ventilation a Temperature - High NA M
R 1, 2, 3 c.
d.
SLCS Initiation NA R
NA 1, 2, 3 e.
- g oc e Low, Level 2
[E Transmitter:
NA NA R(b) 1, 2, 3 fa)
- j$
Trip Logic:
D M
M 1, 2, 3 e
2:
f.
'a Flow - High - Time Delay Relay NA M
R 1, 2, 3 l
.Bs
~.
E
' TABLE 4.3.2-1 (Continued)
E!
E ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE ' REQUIREMENTS
~n*
CHANNEL OPERATIONAL CIIANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICII E!
TRIP FUNCTION CHECK TEST Ci.LIBRATION SURVEILLANCE REQUIRED 4
[
w 4.-
CORE STANDBY COOLING SYSTEMS ISOLATION a.
liigh I' essure Coolant Injection System Isolation 1.
IIPCI Steam Line Flow - Iligh Transmitter:
NA(a)
NA R(b) 1, 2, 3 Trip Logic:
D M
M 1, 2. 3 2.
11PCI ' Steam Line Flou - Iligh l
Time Delay Relay NA R
R 1, 2, 3 3.
!!PCI Steam Supply Pressure - Low NA M
R 1,2,3 w
0 4.
IIPCI Steam Line Tunnel Temperature - liigh NA M
Q 1, 2, 3 5.
Bus Power Monitor NA R
N/
1, 2, 3 6.
HPCI Turbine Exhaust Diaphragm Pressure - liigh NA M
Q 1, 2, 3 7.
IIPCI Steam Line Ambient Temperature - liigh NA M
R 1, 2, 3 l
-gjg 8.
hPCI Steam Line Area A Temperature - High NA M
R 1,
2.,
3 l
a m a.
~4 3m 9.
IIPCI Equipment Area
- g E Temperature - liigh NA M
Q 1, 2, 3 l
e z
' ?
10.
Dryvell Pressure - High hgg Transmitter:
NA(a)
NA R(b) 1, 2, 3 Trip Logic:
D M
M 1, 2, 3 m
TABLE 4.3.2-1 (Continued)
Ey ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
y, CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WiiICli TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED U
4.
CORE STANDBY COOLING SYSTEMS ISOLATION (Continued) o b.
Reactor Core Isolation Cooling System Isolation 1.
RCIC Steam Line Flow -' lli eh Transmitter:
NA(a)
NA R(b) 1, 2, 3 Trip Logic-D M
M 1, 2, 3 2.
RCIC Steam Iine Flow - High l
Time Del,y Relay NA R
R 1, 2, 3 3.
FCIC c eam Supply Pressure - Low NA M
Q 1, 2, 3 Z
4.
aCIC Steam Line Tunnel Temperature - High NA M
R 1, 2, 3 l
h 5.
Bus Power Monitor NA R
NA 1, 2, 3 6.
RCIC Turbine Exhaust Diaphragm Pressure - High' NA M
R 1, 2, 3
7.
RCIC Steam Line Ambient Temperature - !!igh NA M
R 1, 2, 3 l
8.
RCIC Steam Line Area a Temperature - High NA M
R 1, 2, 3 l
9.
RCIC Equipment Room Ambient Temperature - High NA M
Q 1, 2, 3 l
10.
RCIC Equipment Room A Temperature - High NA M
Q I, 2, 3 l
f, 11.
RCIC Stes.n Line Tunnel Tempera-ture - High Time Delay Relay NA M
R 1, 2, 3 12.
Drywell Presseure - High Q
Transmitter.
NA(a)
NA R(b) 1 2, 3 Trip Logic:
D M
M 1, 2, 3 5
e
=
TABLE 4.3.2-1(ContinueQ ISOLATION ACTUATION INS 1RUMENTATION SURVEILLANCE Kt.QUIREMENTS NOTES l
(a') The t ransi t t er channel check is satisfied by the t rip unit <hannel check.
A separate transmitter check is not required.
(b) Transmitters are exempted f rom the monthly channel calibrt t ion.
(c) If not perf ormed within the previous 31 days.
(d) Testing shall verif y t hat the mechanical vacuum pump trips and the mechanical vacuum pump line valve closes.
(e) When react or steam pressure > $00 psig.
(f ) When handling irradiated f uel in the secondary containment.
l BRUNSWICK - UNIT 2 3/4 3-32 Amendme nt No. 60, 72, 78, 97, 120, 124, 160, 179
4 - e TABl.E 3.6.3-1 PR] MARY CONTAINMENT ISOLATION VALVES Table 3.6.3-1 has been deleted.
Refer to Plant Procedure RCI-02.6.
I Pages 3/4 6-15 through 3/4 6-17 have been deleted.
l 1
179l (Next page is 3/4 6-18)
BRUNSWICK - UNIT 2 3/4 c-14 Amendment No.
O e
o CONTAIRMENT SYSTEMS SECONDARY CONTAINMEN1 AUTOMA1-
- -^LATION DAMPERS LlHITING CONDITION FOR OPERATION 3.6.5.2 The secondary containment aut omat ic,i solat ion dampe rs shown in Tabl e 3.6.5. 2-1 shal l be OPERAbl.E.
APPLICABILITY:
OPERA 110NAL CONDITIONS 1, 2, 3, 5, and *.
l ACT10Nt With one or more of the secondary containment isolation dampers specified in Table 3.6.5.2-1 inoperable, operat ion may cont inue and the provisions of Specification 3.0.4 are not applicable, provided that at least one isolation damper is maintained OPERABLE in each at f ected penet ration that is open, andI The inoperable damper is restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, a.
or b.
The af f ected penetration is isolated by use of a closed damper within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or SECONDARY CONTAINMENT INTEGRITY is demonstrated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and c.
the damper is restored to OPERABLE status within 7 days.
Otherwise, in OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT l
SilUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTLOWN within the following 24. hours.
Otherwise, in OPERATIONAL CONDITION 5 or *, suspend irradiated f uel l
handling in the secondary containment, CORE ALTERATIONS, or ac t ivi ties that could reduce the SHUTDOWN MARCIN. The provisions of Spec i f icat ion 3.0.3 are not applicable.
- When irradiated fuel is being handled in the secondary containment.
BRUNSWICK - UNIT 2 3/4 6-22 Amendment No.179
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T ABl.E 3. 6. '. 2-1 SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS Tabl e 3.6.5.2-1 ha s been del et ed.
Refer to Plant Procedure RCl-02.6.
BRUNSWICK - UNIT 2 3/4 6-24 Amendment No. 179
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4 CONTAIRMENT SYSTEMS BASES 3/4.6.3 PRIMARY CONTAINMENT I SOLATION VALVES (Cont inued )
A list of automatic closing primary containment isolation valves and their associated closure times shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50.
The addition and deletion of primary containment isolation valves shall be made in accordance with Section 50.59 of 10 CFR Part 50.
3/4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the drywell and suppression pool and the suppression pool and reactor building.
This system will maintain the structural integrity of the containment under conditions of large differential pressures.
The vacuum breakers betueen the drywell and the suppression pool must not be inoperable in the open position since this would allow bypassing of the suppression pool in. case of an accident.
There are an adequate number of valves to provide some redundancy so that operation may cont inue with no more than 2 vacuum breakers inoperable and secured in the closed position.
Each set of vacuum relief valves between the suppression chamber and reactor building provides 100% relief, which may be required in the unlikely event that negative pressures develop in the primary containment.
The Nitrogen Backup System provides backup motive power for these suppression pool reactor building vacuum breakers on a loss of instrument air.
The normal non-interruptibla instrument air system for these vacuum breakers is designed as a Seismic Class I system supplied by air compressors powered from the emergency buses.
The Nitrogen System serves as a backup to the air system and thus the loss of the Nitrogen System, or portions thereof, does not make the.
vacuum breakers inoperable. This design allows for the out of service times in Actions b and c.
The Nitrogen Backup System is added to the Suppression t
Pool-Reactor Building Vacuum Breaker specification to satisf y NRC concerns relative to 10 CFR 50.44(c)(3) as addressed in the Brunswick Saf et y Evaluation Report dated October 30, 1986 concerning Generic Letter 84-09.
Pressurization to 1130'psig assures sufficient system capacity to provide 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation with design valve actuation and system leakage.
3/4.6.5 SECONDARY CONTAINHENT Secondary containment is designed to minimize any ground level release of I
radioactive material which may result from an accident. The reactor building l
provides secondary containment during normal operation when the drywell is sealed and in service. When the reactor is shut down or during refueling the drywell may be open and the reactor building then becomes the primary containment.
1 l
BRUNSWICK - UNIT 2 B 3/4 6-5 Amendment No. 47, 738, l
l 179
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CONTAINMENT SYSTEMS BASES (Continued) 3/4.6.5 SECONDARY CONTAINMENT (Continued) l Establishing and maintaining a vacuum in the building with the standby gas treatment system, once per 18 months, along with the surveillance of the valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
)
A list of secondary containment automatic isolation dampers shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50.
The addition and deletion of secondary containment automatic isolation dampers shall be made in accordance with Section 50.59 of 10 CFR Part 50.
3/4.b.6 CONTAINMENT ATMOSPHERE CONTROL The OPERABILITY of the containment iodine filter trains ensures that sufficient iodine removal capability will be available in the event of a LOCA. The reduction of containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.
The Ol'ERABILITY of the equipment and systems required f or the detect ion and cont rol of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditiens.
The containment inerting system is capable
- of controlling the expected hydrogen generation associated with 1) circonium-water reactions, 2) radiclytic decomposition of water, and 3) corrosion of metals within containment. The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, " Cont rol of Combust ible Gas Concentrations in Containment following a LOCA."
BRUNSWICK - UNIT 2 8 3/4 6-6 Amendment No.179 l
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