ML19274D658
| ML19274D658 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/12/1979 |
| From: | Utley E CAROLINA POWER & LIGHT CO. |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| GD-79-375, NUDOCS 7902210231 | |
| Download: ML19274D658 (100) | |
Text
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Carolina Power & Light Company February 12, 1979 FILE: NG 3514 (B)
SERIAL: GD-79-375 Office of Nuclear Reactor Regulation ATTENTION:
Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 United States Nuclear Regulatory Commission Washingten, D. C.
20555 RE: BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 LICENSE NOS. DPR-71 AND DPR-62 MISCELLANEOUS TECHNICAL SPECIFICATION CHANGES
Dear Mr. Ippolito:
In accordance with the Code of Federal Regulations, Title 10, Parts 2.101 and 50.90, Carolina Power & Light Company hereby requests a revit on to the Technical Specifications for the Brunswick Steam Electric Plant (BSEP) Units 1 and 2.
The bases for the revision requests for Units 1 and 2 are tabulated on a page-by-page basis in Attachments IA and IIA, respectively.
Revised pages to implement the requested changes for Units 1 and 2 are included in Attachments IB and IIB.
The changed por-tions are marked by a vertical bar in the right hand margin.
The changes requested are administrative in nature and have no safety or environmental significance. Therefore, in accordaace with 10CFR170.22, the requests constitute one Class I and one Class II Amendment. A license fee of $1,600.00 is enclosed.
Yours very truly, f.E Utley
[
E Senior Vice President Power Supply SP/jc Attachments Sworn to and subscribed before me this 12th day of February, 1979.
N Notary Public j
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ATTACHMENT IIA UNIT #2 CHANGES A.
The following is a list by page and item number of the changes due to typographical errors:
Page Number Item 2-5 10 2-6 Note (7) 3/4 1-8 4.1.3.5.b.1 3/4 1-8 4.1.3.5.b.2 3/4 3-3 7
3/4 3-3 8
3/4 3-3 11 3/4 3-3 12 3/4 3-6 7
3/4 3-6 8
3/4 3-6 11 3/4 3-6 12 3/4 3-7 6
.s/4 3-7 7
3/4 3-8 8
3/4 3-8 11 3/4 3-8 12 3/4 3-11 1.b 3/L 3-11 1.c.1 3/4 3-12 2.b 3/4 3-13 4.a.3 3/4 3-15 5.b 3/4 3-17 1.b 3/4 3-17 1.c.1 3/4 3-18 2.b 3/4 3-19 4.a.6
. 3/4 3-22 1.b 3/4 3-22 1.c.1 3/4 3-22 1.d 3/4 3-22 2.b 3/4 3-25 1.b 3 /4 3-25 1.c.1 3/4 3-25 1.e 3/4 3-26 2.b 3/4 3-26 3.b 3/4 3-33 4.e 3/4 3-37 7.b 3/4 3-43 1.d 3/4 3-46 1
3/4 3-46 2
3/4 3-48 2
3/4 3-49 2
Attachment IIA (continued) 0 1
Page Number Item 3/4 3-51 6
3/4 3-52 2
3/4 3-52 6
3/4 3-56 2
3/4 3-57 2
3/4 3-58 2
3/4 3-59 4.3.5.7.2 3/4 4-18 4.4.6.2 3/4 5-2 4.5.1.b 3/4 5-2 4.5.1.c.2 3/4 5-3 4.5.2.b 3/46-7 4.6.1.5 B.
The folloving is a list by page and item number of the chenges due to incorrect instrument numbers in the orig h 1 issue of the Standard Technical Specifications:
Page Number Item 3/4 3-37 2.d 3/4 3-48 1
3/4 3-49 1
3/4 4-6 4.4.3.2.a 3/4 4-6 4.4.3.2.b C.
The following explanations are given for individual changes since the reasons for these r hmges cannot b? grouped together generically:
1.
Page No. 3/4 3-7, Item No. 2.a.
See Attachment IA, Item C.l.
2.
Page No. 3/4 3-7, Item No. 2.c.
See Attachment IA, Item C.2.
3.
Page No. 3/4 3-15, item No. 5.a.
Groups changed to indicate those applicable.
4.
Page No. 3/4 3-15, Item No. 5.b.
Page No. 3/4 3-29, Item No. 5.b.
See Item No. C.3 above and also Attachment IA, Item No. C.3.
5.
Page No. 3/4 3-57, Item No. 2 See Attachment IA, Item No. C.S.
Attachment IIA (continued)
D.
The following is a list by pages and iten nunber for changes to the " Definitions" of Appendix A:
?aee Nu=her l-6 Add definition of STAGGEPID TEST BASIS as Iten 1.31.
This addition necessitates renu = bering existing itens as follows:
Ther=al Power to 1.32.
Total Peaking Factor to 1.33.
Unidentified Leakage to 1.34.
1-7 1.
Change Frequency of SA to 184 days.
2.
Add A (Annual) Frequency of 366 days.
3.
Change Frequency of R to 550 days.
E.
Other Changes Add to Appendix A, new paragraph as follows:
6.13 Labeling - In lieu of 10CFR20.203(f), entrances to each building in which radioactive =aterials are used, stored, or handled, shall have signs bearing the legend, EVERY CONTAINER OR VESSEL IN TEIS APIA MAY CONTAIN RADIOACTIVE MATERIAL.
ATTACIDENT IIB Revised pages to implement the requested changes for Unit No. 2
DEFINITIONS SHUTDOW MARGIN 1.3 SHUTDOWN MARGIN sbAll be the amount of reactivity by which the reactor would be suberitical assuming that all control rods capable of insertion are fully inserted except for the analytically deternined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68 F, and Xenon free.
1.31 Staggered Test Basis A Staggered Test Basis shall consist of:
A test schedule for n systems, subsystems, trains or designated a.
components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train or designated components at the beginning of each subinterval.
THERMAL POWER 1.32 THERMAL power shall be the total reactor core heat transfer rate to the reactor coolant.
TOTAL PEAKING FACTOR 1.33 The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LEGR 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 1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
i BRUh3 WICK - UNIT 2 1-6
TABLE 1.1 FREOUENCY NOTATION NOTATION FREOUENCY S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days.
A At least once per 366 days.
R At least once per 550 days.
S/U Prior to each reactor startup.
N.A.
Not applicable.
BRUNSWICK - UNIT 2 1-7
Ei TABLE 2.2.1-1 E
gg REACTOR PROTECTION SYSTEM INST,3yi.ENTATION SETPOINTS ALLOWABLE gj FUNCTIONAL UNIT AND INSTRUMENT NUMBER TRJPSETPOINT VALUES Turbine Stop Valve - Closure (6)
,10% closed
< 10% closed 9.
(EllC-SVOS-l X,2X,3X,4X )
- 10. TurbineControlValveFast{gjosure, Control Oil Pressure - Low
~> 500 psig
-> 500 psig l
(ENC-PSL-1756,1757,1758,1759)
.m i
l
TABLE 2.2.1-1 (Continued)
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 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.
MwwSG e @ h O
e BRUNSWICK - UNIT 2 2-6
REACTIVITY CONTROL SYSTEMS CONTROL R0D SCRAM 4CCUMULATCP,5 1.IMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.
APPLICABILITY:
CONDITIONS 1, 2 and 5*.
ACTION:
a.
In CONDITION 1 or 2 with one control rod scram accumulator inoperable, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s:
1 The inoperable accumulator,is restored to OPERABLE status, or 2.
The control rod associated with the inoperable accumu-lator is declared inoperable, and the requirements of Specification 3.1.3.1 are satisfied.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In CONDITION 5* with a withdrawn control rod scram accumulator inoperable, fully insert the affected control rod and elec-trically disarm the directional control valves within one hour.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1. 3. 5 The control rod scram accumulators shall be determined OPERABLE:
a.
At least once per 7 days by verifying that the pressure and leak detectors are not in the alarmed condition, and b.
At least once per 18 months by performance of a:
1.
CHANNEL FUNCTIONAL TEST of the leak detectors (Cl2-LS-129-xxxx), and 2.
CHANNEL CALIBRATION of the pressure detectors (c12-PS-130-xxxx).
- At least the accumulator associated with each withdrawn control rod.
Not V
applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
BRUNSWICK - UNIT 2 3/41-8
TABLE 3.3.1-1 (Continued) o, E
F.
REACTOR PROTECTION SYSTEH IllSTRUMENTATION 5
APPLICABLE MINIMUN tlUMBER OPERATIONAL OPERABLE CllANNELS FUNCTI0tlAL UNIT AllD INSTRUMENT NUMBER CONDITIONS PER TRIP SYSTEM (a)(b[ ACTION c
7.
Drywell Pressure - liigh 1,2(e) 2 6
(C72-PS-N002 A,B,C,D) 8.
Scram Discharge Volume Water Level -
liigh (Cl2-LSil-N013 A,B,C,D) 1,2,5(7) 2 5
9.
Turbine Stop Valve - Closure 1(9}
4 8
(EllC-SV05-1 X.2X,3X,4 X )
10.
Turbine Control Valve Fast Closure, t'
Control Oil Pressure low 1(g) 2 8
(EllC-PSL-1756,1757,1758,1759) 9' 11.
Reactor Mode Switch in shutdown Position (C72A-SI) 1,2,3,4,5 1
9 l
- 12. Manual Scram (C72A-S3A,B) 1,2,3,4,5 1
10
TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES E
E FUNCTIONAL UNIT AND INSTRUMENT NUMBER RESPONSE TIME y
. Seconds)
(
n*
1 Intermediate Range Monitors (C51-IRM-K601 A,B,C,D,E,F,G,H):
a.
Neutron Flux - High*
NA E
b.
Inoperative NA 2.
Average Power Range Monitor * (C51-APRM-CH.A,B,C,0,E,F):
a.
fleutron Flux - High,15%
< 0.09 b.
Flow Biased Neutron Flux - High fiA c.
Neutron Flux - High, 120%
< 0.09 d.
Inopera tive ITA e.
Downscale NA f.
LPRM NA 3.
Reactor Vessel Steam Dome Pressure - High (B21-PS-N0?3 A,B,C,D)
< 0.55 4.
Reactor Vessel Water Level - Level #1 (B21-LIS-N017 A,B,C,D)
< l.05 2
5.
Main Steam Line Isolation Valve-Closure (B21-F022 A,B,C,D and 821-F028 A,B,C,0) < 0.06 Y'
6.
Main Steam Line Radiation - Hioh (D12-RM-K603 A,B,C,D)
NA o,
l 7.
Drywell Pressure - High (C72-PS-N002 A,B,C,D)
NA 8.
Scram Discharge Volume Water Level - High (C12-LSil-N013 A,B,C,D)
NA l
9.
Turbine Stop Valve - Closure (EHC-SV0S-lX,2X,3X,4X)
< 0.06
- 10. Turbine Control Valve Fast Closure, Control Oil Pressure - Low (EllC-PSL-1756,1757,1758,1759)
< 0.08
- 11. Reactor Mode Switch in Shutdown Position '(C72A-SI)
NA
- 12. Manual Scram (C72A-S3 A,B)
NA l
- Neutron detectors are exempt from response time testing.
Response time shall be measured from detector output or input of first electronic component in channel.
n n
n
~
TABLE 4.3.1-1 REACTOR PROTECTION SYSTEH INSTRUMENTAT10ft SURVEILLAllCE REQUIREMENTS E
CllANNEL OPERATIONAL E
FUllCTI0llAL UNIT CllANilEl FUNCTI0llAL CllAPNEL CONDITIONS IN WillCll y
AllD INSTRUMENT NUMBER CllECK TEST CALIBRATI0ft(a)
SURVEILLANCE REQUIRED F;
^
1.
Intermediate Range Honitors:
(C51-IRM-k601 A,0,C,0,E,F,G II) h, c) a.
fleutron Flux - liigh D
S/U R
2 D
W R
3,4,5 b.
Inoperative flA W
NA 2,3,4,5 2.
Average Power Range Monitor:
(C51 - A PRM-Cll. A,B.C,0,E,F)
S/U(b),g(d)
Q 2
a.
Neutron Flux - liigh 15%
5 (e)(f)
/U(b) b.
Flow Blased Neutron Flux-Illgh S y
q c.
Fixed Neutron Flux - Illgh, S/U(b),W W(e),Q l
I m2 120%
S d.
Inoperative NA W
NA 1, 2, 5 y',
e.
Downscale NA W
NA 1
f.
LPRM D
NA (g) 1, 2, 5 3.
Reactor Vessel Steam Dome Pressure.- liigh NA H
Q 1, 2 (021-PS-N023 A,0,0,D) 4 Reactor Vessel Water level - Low Level #1 (B21-LIS-H017 A,0,C,0)
D H
Q 1, 2 5.
Main Steam Line Isolation Valve -
g)
Closure (021-F022 A,B,C,0 and NA H
R 1
021-F028 A,B,C,D)
III 6.
Main Steam Line Radiation - liigh S
H R
1, 2 l
(012-RM-K601 A,B,C,0) 7.
Drywell Pressure - liigh NA H
Q 1, 2 l
(C72-PS-N002 A,B,C,D)
TABLE 4.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL OPERATIONAL FUNCTIONAL UNIT CHANNEL FUNCTIONAL CilANNEL CONDITIONS IN WilICH
- o AND INSTRUMENT NUMBER CllECK TEST CALIBRATION SURVEILLANCE REQUIRED 5
8.
Scram Discharge Volume Water i
5 Level - liigh NA Q
R 1, 2, 5 l
Q (c12-LSil-N013 A,B,C,D) 9.
Turbine Stop Valve - Closure NA M
R(h) j E
(EllC-SV05-l X,2X,3X,4X )
- 10. Turbine Control Valve Fast Closure, Control Oil Pressure -
low (EllC-PSL-1756,1757,1758,1759)
NA M
R 1
- 11. Reactor Mode Switch in Shutdown HA R
NA 1,2,3,4,5 Position (C72A-S1)
- 12. Manual Scram NA Q
NA 1,2,3,4,5 (C72A-S3A,B) l a.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
The IRM channels shall be compared to the APRM channels and the SRM instruments for overlap during w
c.
2 each startup, if not performed within the previous 7 days.
Y d.
When changing from CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter entering CONDITION 2.
e.
This calibration shall consist of the adjustment of the APRM readout to conform to the power values calculated by a heat balance during CONDITION 1 when TilERMAL POWER > 25% of RATED lilERMAL POWER, f.
This calibration shall consist of the adfustment of the APRM flow biased setpoint to conform to a calibrated flow signal.
g.
The LPRH's shall be calibrated at least once per effective full power month (EFPM) using the TIP system.
h.
This calibration shall consist of a physical inspection and actuation of these position switches.
1.
Instrument alignment using a standard current source.
J.
Calibration using a standard radiation source.
I n
n n
TABLE 3.3.2-1 g
ISOLATION ACTUATION INSTRUMENTATION h
VALVE GROUPS MINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL E
TRIP FUNCTION AND INSTRUMENT NUMBER SIGNAL (a)
PER TRIP SYSTEM (b)(c)
CONDITION ACTION e
g 1.
PRIMARY CONTAINMENT ISOLATION
[
a.
Reactor Vessel Water Level - Low 1.
Level #1 (B21-LIS-N017 A,B,C,D) 2,6,7,8 2
1,2,3 20 2.
Level #2 (821-LIS-N024 A,B and 1, 3 2
1,2,3 20 B21-LIS-N025A,B) b.
Drywell Pressure - liigh 2,6,7 2
1,2,3 20 (C72-PS-N002 A,B,C,D)
R c.
Radiation - High (d) 1 2
1,2,3 21 y
(D12-RM-K603A,B,C,D)
[
2.
Pressure - Low 1
2 1
22 (B21-PS-N015 A,B,C,D) 3.
Flow - liigh 1
2/line 1
22 (B21-dPIS-N006 A,B,C D; B21-dPIS-N007 A,B,C,0; B21-dPIS-N008 A,B,C,0; and B21-dPIS-N009 A,B,C,0) 4.
Flow - High 1
2 2, 3 21 (821-dPIS-N006A; B21-dPIS-N007B; B21-dPIS-N008C and B21-dPIS-N0090) d.
Main Steam Line Tunnel Temperature - High 1
2(e) 1,2,3 21 (B21-TS-N010 A,B,C,0; B21-TS-N011 A,B,C,D; B21-TS-N012 f.,B,C,0 and B21-TS-N013 A,B,C,D) e.
Condenser Vacuum - Low 1
2 1,2(f) 21 (B21-PS-N056A,B,C,D) f.
Turbine Building Area Temperature-High 1
4(e) 1, 2, 3 21 (B21-TS-3225 A,B,C,D; B21-TS-3226 A,B,C,0; B21-TS-3227 A,B,C,D; B21-TS-3228 A,B,C,0; B21-TS-3229 A,B,C,0; B21-TS-3230 A,B,C,D; B21-TS-3231 A,B,C,0 and B21-TS-3232 A,B,C,D)
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION m
E5 VALVE GROUPS MINIMUM NUMBER APPLICABLE 5
OPERATED BY OPERABLE CIIANNELS OPERATI0flAL p
TRIP FUNCTION AND INSTRUMENT NUMBER SIGNAL (a)
PER TRIP SYSTEM (b)(c)
CONDITION ACTION h
2.
SECONDARY CONTAINMENT ISOLATION s
H a.
Reactor Building Exhaust N
Radiation - liigh 6
1 1, 2, 3, 5 and
- 23 (012-RM-N010 A,B,)
b.
Drywell Pressure - liigh 2, 6, 7 2
1,2,3 23 (C72-PS-N002 A,B,C,D) c.
Reactor Vessel Water Level - Low, Level #2 1, 3 2
1,2,3 23 w1 (B21-LIS-H024 A,B and B21-LIS-N025 A,B) 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - liigh 3
1 1,2,3 24 (G312dFS-N603-1A 1B) b.
Area Temperature - liigh 3
2 1,2,3 24 (G31-TS-N600A,B,C,0,E,F)
Area Ventilation A Temp. - liigh 3
2 1,2,3 24 c.
(G31-TS-N602A,B,C,D,E,F) d.
SLCS Initiation (C41A-S1) 3 (g)
NA 1,2,3 24 e.
Low, Level # 2 1, 3 2
1 2, 3 24 (B21-LIS-N024A,B and B21-LIS-N025A,B)
O n
n
TABLE 3.3.2-1 (Continued) y ISOLATION ACTUATION INSTRUMENTATI0fi 5:c VALVE GROUPS MINIMUM fiUMBER APPLICABLE E
OPERATED BY OPERABLE CilANNELS OPERATIONAL 7
TRIP FUNC110N AND It!STRUMEflT NUMBER SIGilAL(a)
PER TRIP SYSTEM (b)(c)-
CONDITI0fl ACTION E
4.
CORE STANDBY COOLING SYSTEMS ISOLATION q
m a.
liigh Pressure Coolant Injection Isolation 1.
HPCI Steam Line Flow - liigh 4
2 1,2,3 25 (E41-dPIS-N004 and E41-dPIS-N005) 2.
HPCI Steam Supply Pressure -
Low (E41-PSL-fl001A,B,C,D) 4 2
1,2,3 25 w
3.
HPCI Steam Line Tunnel 1
Temperature - liigh 4
2 1,2,3 25 (E41-TS-3314; E41-TS-3315; E41-TS-3316; E41-TS-3317; E41-TS-3318; w
h E41-TS-3354; E41-TS-3488 and E41-TS-3489) 4.
Bus Power Moriitor NA (h) 1/ bus 1, 2, 3 26 (E41-K55 and E41-K56) 5.
HPCI Turbine Exhaust Diaphragm Pressure - High 4
2 1,2,3 25
( E41 -PSH-H012A,B,C,D) 6.
IIPCI Steam Line Ambient Temperature - High 4
2 1,2,3 25 (E51-TS-N603C,D) 7.
HPCI Steam Line Area A Temp. -
liigh 4
2 1,2,3 25 (E51-dTS-N604C,D) 8.
Emergency Area Cooler Temperature - liigh 4
2 1,2,3 25 (E41-TS-N602A,B)
TABLE 3.3.2-1 (Continued)
ISOLATI0ft ACTUATION INSTRUMENTATION E
VALVE GROUPS MINIMUM NUMBER APPLICABLE Q
OPERATED BY OPERABLE CilANNELS OPERATIONAL TRIP fullCTION AND INSTRUMENT NUMBER SIGNAL (a)
PER TRIP SYSTEf1(b)(c)
CONDITION ACTION 5.
SIlUT00Wii COOLING SYSTEM ISOLATION a.
Reactor Vessel Water - Low, level #1 2,
6, 7, 8 2
3, 4, 5 20 l (B21-LIS-N017A,B,C,0) b.
Reactor Steam Dome Pressure-liigh 7, 8 1
3, 4, 5 27l (B32-PS-N018A,B) 5 Y
G
TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS Si ALLOWABLE TRIP FUNCTI0ff AND INSTRUMENT NUMBER TRIP SETPOINT VALUE 1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Vessel Water level - Low 1.
Level #1 (B21-LIS-H017 A,B,C,D)
> +12.5 inches
> + 12.5 inches 2.
Level #2 (B21-LIS-N024 A,B and I -38 inches s - 38 inches 821-LIS-N025A,B)
~
b.
Drywell Pressure - High 1 2 psig i 2 e ig s
(C72-PS-N002 A,B,C,D) c.
(D12-RM K603 A,B,C,D)
~< 3 x full power background
< 3.5 x full power 1.
Radiation - liigh Eackground l
Y' (B21-PS-N015A,B,C,D)
-> 825 psig
> 825 psig 2.
Pressure - Low
~
d 3.
Flow - High
< 140% of rated flow
< 140% of rated flow (B21-dPIS-N006 A,B,C,0; B21-dPIS-N007 A,B,C,0; B21-dPIS-N008 A.B.C.D; and 821-dPIS-N009 A.B.C,0)
-< 40% of rated flow
< 40% of rated flow 4.
Flow - High
~
(B21-dPIS-N006A; B21-dPIS-N007B; B21 dPIS-N008C and 821-dPIS-N009D) d.
Main Steam Line Tunnel Temperature - High
< 200*F
< 200*F (B21-TS-N010 A B.C,0; B21-TS-N011 A.B.C.E; B21-TS N012 A,B,C,D; and B21-TS-fiO13 A,B,C,0) e.
Condenser Vacuum - Low
> 7 inches Hg vacuum
> 7 inches lig vacuum (821-PS-NOS6A,B,C,0)
< 200'F f.
Turbine Building Area Temp - High
< 200*F (B21-TS-3225 A,B,C,D; B21-TS-3226 A B.C,6; B21-TS-3227 A,B.C,0; B21-TS-3228 A,B,C,0; B21-TS-3229 A,B,C,0; B21-TS-3230 A,B,C D; B21-TS-3231 A,B,C,D and B21-TS-3232 A,B,C,D)
TABLE 3.3.2-2 (Continued) h ISOLATION ACTUATION INSTRUMENTATI0fl SETPOINTS 0;
5 ALLOWABLE 9
TRIP FUllCTION AND INSTRUMENT flUMBER TRIP SETPOINT VALUE g
- 2. SECONDARY C0tiTAINMENT ISOLATI0fl a.
Reactor Building Exhaust Radiation - High 1 11 mr/hr i ll mr/hr (D12-RM-N010 A,B) b.
Drywell Pressure - High 1 2 psig 1 2 psig l
(C72-PS-N002 A,B,C,D) c.
Reactor Vessel Water Level - Low, level #2
> -38 inches
{
(021-LIS-N024 A,8 and B21-LIS-N025 A,B) -
-> -38 inches 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High
-< 53 gal / min
-< 53 gal / min (G31-dFS-N603-1A,lB) b.
Area Temperature - High
-< 150*F
< 150*F (G31-TS-N600A,B,C,D,E,F)
Area Ventilation Temperature ATemp-High < 50*F
< 50*F c.
(G31-TS-N602A,B,C,D,E,F) d.
NA NA e.
Reactor Vessel Water - Low, level #2
> -38 inches
> -38 inches (821-LIS-N024A,B and 821-LIS-N025A,B) -
TABLE 3.3.2-2 (Continued) m g
ISOLAT10ft ACTUATI0il IllSTRUMENTATION SETPOINTS cN ALLOWABLE 7
1 RIP FUllCTI0li AND INSTRUMENT f! UMBER TRIP SETPOINT VALUE E
4.
CORE STAllDDY COOLING SYSTEMS ISOLAT10ft n>
a.
liigh Pressure Coolant injection isolation 1.
IIPCI Steam Line Flow - liigh
< 300% of rated flow
~< 300% of rated flow (E41-dPIS-fl004 and E41-dPIS-N005)~
1 00 psig 1
1 00 psig 1
2.
IIPCI Steam Supply Pressure - Low (E41-PSL-N001A.D.C.D) 3.
IIPCI Steam Line Tunnel Temperature -
g (E41-TS-3314; E41-TS-3315; E41-TST3316; E41-TS-3317; E41-TS-3310;
~< 200*F Illgh
< 200*F E41-TS-3354; E41-TS-3488 and E41-TS-3489) n w
4.
Bus Power Monitor HA flA 1
(E41-K55 and E41-K56) 5.
IIPCI Turbine Exhaust Diaphragm Pressure - liigh
< 10 psig
< 10 psig (E41-PSil-N012A,B.C,0) 6.
IIPCI Steam Line Ambient Temp -
liigh (E51-TS-N603C, D)
< 200*F
< 200*F l
7.
IIPCI Steam line Area A Temp -
liigh (51-dTS-fl6040,D)
< 50*F
< 50*F (E51-dTS-N604C,0) 8.
Emergency Area Cooler Temp - liigh ~< 175'F
--< 175'F (E41-TS-N602A,0)
i ISOLATION SYSTEM RESPONSE TIME TRIP FUNCTION AND INSTRUMENT NUMBER RESPONSE TIME (Seconds) 1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level - Low 1.
Level #1 (B21-LIS-N017 A,B,C.D)
<13 2.
Level #2 (B21-LIS-N024 A,B and
- 71. 0**
B21-LIS-N025 A,B) b.
Drywell Pressure - High 113 (C72-PS-N002 A,B,C,D) c.
Radiation - Hich*
<l.0**
( D12-RM-K603" A, B, C, D)
[
~
2.
Pressure - Low
<13 (B21-PS-N015 A,B,C,D) 3.
F1ow - Hich
<0.5**
(B21-dPiS-N006 A,B,C,D; B21-dPIS-N007 A,B,C,D;
~
B21-dPIS-N00B A,B,C,D and B21-dPIS-N009 A,B,C D) 4 Flow - High
<0.5" (B21-dPIS-N006A; B21-dPIS-N007B; B21-dPIS-B00B'C and B21-dPIS-N009D) d.
Main Steam Line Tunnel Temperature - High
<13 (B21-TS-N010 A,B.C,D; B21-TS-N0ll A,B,C,D; B21-TS-N012 A,B,C,D; and B21-TS-N013 A,B,C,D)
I e.
Condenser Vacuum - Low
<13 (B21-PS-N056 A,B,C,D)
~
f.
Turbine Building Area Temperature - High NA (B21-TS-3225 A,B,C,D; B21-TS-3226 A,B,C,D; B21-TS-3227 A,B,C,D; B21-TS-322B A,B,C,D; B?l-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) 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Exhaust Radiation - High*
113 (D12-RM-N010 A,B) b.
Drywell Pressure - High
<13 (C72-PS-N002 A,B,C,D) c.
Reactor Vessel Water Level - Low, Level # 2
-< l. 0" (B21-LIS-N024 A,B and B21-LIS-N025 A,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.
" Isolation actuation instrumentatien response time only.
BRUNSWICK-UNIT 2 3/4 3-22
TABLE 4.3.2-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREHEilTS en E
CllANNEL OPERAT10tlAL 4
CilAllNEL FUNCTIONAL CilANNEL CONDITIONS IN WillCil y
TRIP FUNCTI0il AND IllSTRUMENT NUMBER CllECK TEST CAllBRATION SURVEILLANCE REQUIRED n
i 1.
PRIMARY C0llTAINMENT ISOLATION C5 a.
Reactor Vessel Water Level - Low 1.
Level #1 D
H Q
1, 2, 3 (B21-LIS-H017 A B.C.D) 2.
Level #2 0
M Q
1, 2, 3 (B21-LIS-N024 A,0 and 021-LIS-N025 A,B) b.
Drywell Pressure - liigh NA H
Q 1, 2, 3 l
(C72-PS-N002 A,B,C,D) c.
1.
Radiation - liigh D
W R
1, 2, 3 (012-RM-K603 A,B,C,0) l i'
2.
Pressure - Low NA H
Q 1
I?>
(B21-PS-N015 A,B,C,0) 3.
Tlqw - liigh NA H
Q 1
(B21-dPIS-N006 A,B,C,D; B21-dPIS-H007 A B.C,0; B21-dPIS-H008 A,B.C.D; and 021-dPIS-N009 A,B,C,0) 4.
Flow - liigh NA H
Q 2, 3 (B21-dPIS-N006A; B21-dPIS-N0070; B21-dPIS-N000C and 021-dPIS-N0090) d.
Main Steam Line Tunnel Temperature - liigh NA H
R 1, 2, 3 (B21-15-N010 A,B,C,D; il21-TS-N011 A,B,C,D; B21-TS-N012 A,B,C,0 and 821-IS-N013 A,B,C,0) e.
Condenser Vacuum - Low NA H
R 1, 2 g
(871-PS-H056 A,0,C,0) r.
Turbine Building Area Temp-Illgh NA H
R 1, 2, 3 (B21-i5-3225 A,B,0,0; B21-TS-3226 A,B,C,0; B21-TS-3227 A,B,C,0; B21-TS-3220 A,B,C,0; B21-TS-3229 A,B,C,0; B21-TS-3230 A,B,C,0; B21-TS-3231 A,B,C,D and 021-TS-3232 A,0,C,0)
ITFen reactor steam pressure > 500 psig.
TABLE 4.3.2-1 (Continued)
ISOLATI0il ACTUATION INSTRUMENTATI0ff SURVEILLANCE REQUIREMENTS
,n h
CilANNEL OPERATIONAL 5
CilANNEL FUNCTIONAL CllANNEL CONDITIONS IN WillCll p
1 RIP FUNCTION AND IllSTRUMENT NUMBER CilECK TEST CALIBRATION _
SURVEILLANCE REQUIRED 2.
SECONDARY CONTAINMENT ISOLATI0ft 2
- 1 Reactor Building Exhaust a.
Radiation - liigh D
M R
1, 2, 3, 5 and
- m (012-RM-i1010A,B) b.
Drywell Pressure - liigh NA M
Q 1, 2, 3 (C72-PS-H002 A,B,C,D)
I c.
Reactor Vessel Water Level - Low, level #2 D
M Q
1, 2, 3 (021-LIS-N024 A,B and 021-LIS-N025 A,B) k'o 3.
RE?. TOR WATER CLEANUP SYSTE!! ISOLATION y,
y a.
A flow - liigh 0
M R
1, 2, 3 (G31-dFS-N603-1A,10) b.
Area Temperature - liigh NA M
R 1, 2, 3 (G31-TS-N604A, B, C, D, E, F) l c.
Area Ventilation A Temp -
Illgh (G31-TS-N602A,B,C,D,E,f) ilA M
R 1, 2, 3 d.
NA R
NA 1, 2, 3 e.
low, level #2 D
M Q
1, 2, 3 (B21-LIS-N024A,0 and 021-LIS-N025A,B) liihen handling irradiated fuel in the secondary containment.
TABLE 4.3.2-1 (Continued)
E ILL_AT10ft ACTUATION IllSTRUMENTATION SURVEILLAllCE REQUIREMENTS M
CilAllNEL OPERATIONAL CilANNEL FUNCTIONAL CilANNEL CONDITIONS IN WillCll TRIP FUNCTI0li AND INSTRUMElli flUMBER CllECK TEST CALIBRATION SURVEILLAllCE REQUIRED E
Z 5.
SilUIDOWil COOLING SYSTEM ISOLATION
~
a.
Reactor Vessel Water - Low, tevel #1 D
M Q
3,4,5 (021-LIS-N017A,0 C D) b.
Reactor Steam Dome Pressure -
Illgh (032-PS-fl010A,0)
NA S/U*, M R
3, 4, 5 l
5 W not performed within the prevfous 31 days.
Y G
('
TABLE 3.3.3-1(Continue 41 E
g EMERGENCY CORE COOLING SYSTEM ACTUATI0ft INSTRUMENTATION k
!!INIMUM NUMBER APPLICABLE 7
OPERABLE CHANNELS OPERATIONAL E
TRIP FUNCTION AND INSTRUMENT NUMBER PER TRIP SYSTEM CONDITIONS _
Z 3.
IIPCI SYSTEM m
a.
Reactor Vessel Water Level - Low, Level #2 2
1,2,3 (821-LIS-N031A,B,C,0) b.
Drywell Pressure - High (Ell-PS-N0llA,B,C,D) 2 1,2,3 c.
Condensate Storage Tank level-Low **(E41-LS-H002, E41-LS-N003) NA*
1,2,3 d.
Suppression Chamber Water Level-High** (E41-LSH-N015A,B)
NA*
1, 2, 3 e.
Bus Power Monitor # (E41-K55 and E41-K56) 1/ bus 1, 2, 3 4.
ADS a.
Drywell Pressure - High, coincident with (Ell-PS-N010A,B,C.D) 2 1,2,3 b.
Reactor Vessel Water Level - Low, level #3 2
1, 2, 3 wL (821-LIS-N031A,B,C,0) c.
ADS Timer (B21-TDPU-KSA,B) l 1, 2, 3 d.
Core Spray Pump Discharge Pressure - High (Permissive) 2 1,2,3 (E21-PS-N008A,B and E21-PS-N009A,B) e.
RHR (LPCI MODE) Pump Discharge Pressure - High (Permissive) 2/ pump 1, 2, 3 (Ell-PS-N016A,B,C,0 and Ell-PS-N020A,B.C D) f.
Bus Fower Monitor # (B21-KlA,B) 1/ bus 1, 2, 3
- Alarm only.
When inoperable, verify power availability to the bus at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- When either channel of the automatic transfer logic is inoperable, align HPCI pump suction to the tuppression pool.
- Provides signal to HPCI pump suction valves only.
I
e TABLE 4.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5;
l5 CilAflNEL OPERATIONAL n
CHAllNEL FUilCTIONAL CHANNEL CONDITIONS IN WHICH
[
TRIP FUNCTION AND IllSTRUMENT flUMBER CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 5
H 1.
CORE SPRAY SYSTEM ro a.
Reactor Vessel Water Level - Low, level #3 (B21-LIS-NO31A,B,C,0)
D M
Q 1,2,3,4,5 b.
Reactor Steam Dome Pressure -
Low (821-PS-H021A,B,C,0)
NA M
Q 1,2,3,4,5 c.
Drywell Pressure - High NA M
Q 1, 2, 3 (Ell-PS-N0llA,B,C.D) d.
Time Delay Relay ilA R
R 1,2,3,4,5 e.
Bus Power Monitor (E21-KlA,B)
NA R
NA 1,2,3,4,5 2.
a.
Drywell Pressure - High NA M
Q 1, 2, 3 y
(E11-PS-H011A,B,C,D) b.
Reactor Vessel Water Level - Low, Level #3 (021-LIS-N031A,B C,0)
D M
Q 1, 2, 3, 4*, 5*
I c.
Reactor Vessel Shroud Level NA M
Q 1, 2, 3, 4*, 5*
B21-LITS-NO36 and B21-LITS-il037) d.
Recctor Steam Dome Pressure - Low
( n21-PS-N021A, P,, C, D) 1.
RHR Pump Start and LPCI Injection Valve Actuation NA M
Q 1, 2, 3, 4*, 5*
2.
Recirculation loop Pump Discharge Valve Actuation NA M
Q 1, 2, 3, 4*, 5*
e.
RHR Pump Start-Time Delay Relay NA R
R 1, 2, 3, 4*, 5*
f.
Bus Power Monitor (Ell-K106A,B)
NA R
HA 1, 2, 3, 4 *, 5 *
- Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.
TABLE 4.3.4-1 00lliROL ROD WITilDRAWAL BLOCK lllS1RUMEllTATION SURVEILLAllCE REQUIREHENTS CllANNEL OPERAT10flAL CilANilEl FUNCTIONAL CllAfillEl CONDIT10ris IN WillCll y
MIP FUNCTIOff AtlD INSTRUMENT flUMBER CllECK TEST CAllBRATIOil"I SURVEILLANCE REQUIRED I
F, 1.
APRM (C51-APRM-Cll. A.B.C,0,E,f) h a.
Upscale (Flow Blased) flA S/U H
R(b) y C
c,Q NA 1, 2, S 7
b.
Inopere t ive flA S/U i
E c.
Downscale NA S/U,c,H HA 1
'll d.
Upscale (fixed)
NA S/U' W
Q
- 2. 5 l
2.
ROD BLOCK M0til10R (CSI-RBM-Cll.A,B)
C a.
Upscale llA S/U
,H R
1*
C b.
Inopera tive NA S/U
,Q NA 1*
c.
Downscale NA S/U
,H R
1*
3.
SOURCE RAllGE H0illTORS (C51-SRM-K600A,B,C D) a.
Detector not full in flA S/U
)W NA 2, 5 l'N NA 2, 5 b.
Upscale ng 5I
l' c.
Inoperative r;g 2, 5
')'
d.
Downscale pig r
3{
4.
lHTERNEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,0,E,F,G,II)
S/U(CI,W(d) e a.
Detector not full in NA g
NO W
NA 5
b.
Upscale ng gjg(c)
(d)
NO W
NA 5
c.
Inoperative NA S/U(CI,W(d) g NA W
NA 5
sfu(C)y(d) d.
Downscale NA g3 NA W
NA 5
'T CliAliftfl CKillilihTI0lIS are electronic.
a b.
1his calibration shali com?:' of the adjustment of the APRM flow biased setpoint in conform to a calibra ted n.ow sf gnal.
c.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days, d.
When changing from CONDITION I to 00flDITION 2, pt_rform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering 00ilDIT10N 2.
When TilfRMAI. POWER is greater than the preset power level of the RWM and RSCS.
TABLE 4.3.5.1-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REOUI1LN'TS CHAhTEL INSTRUMENTS, SENSOR CHANNEL FUNCTIONAL CHANNEL LOCATIONS AND INSTRUMENT NUMBER CHECK TEST CALIBRATION 1.
Triaxial Peak Shock Recorders a.
Reactor Building -17' level NA NA R
(DN-IRB-823-1) b.
Reactor Building +20' level NA NA R
(DN-IRH-823-2) c.
Reactor Building +17' level NA NA R
(DU-IRE-823-3) 2.
TLiarial Time-History Accelographs a.
Reactor Building +65' level on Drywell M*
SA R
(DN-II-823-2) b.
Reactor Building -17' level M*
SA R
(ENV-IT-823-1 and ENV-IT-823-3)
- Except seismic trigger BRUNSk'ICK-UNIT 2 3/4 3-46
TABLE 3.3.5.2-1 REMOTE SilUTDOWN HONITORIllG lilSTRUMENTATION s
n
[
HINIHUM z
READOUT CilAHilELS "4
FUllCil0HAL UNIT AllD INSTRUMENT NUMBER LOCATION OPERABLE s,
1.
Reactor Vessel Pressure RSP*
I (021-PI-3332 and C32-PT-110050 g
2.
Reactor Vessel Water Level RSP*
1 (021-LI-3331, 021-LI-R604AX, 021-LT-H027 and B21-LT-N026A) l 3.
Suppression Chamber Water level RSP*
1 (CAC-LI-3342 and CAC-LT-2601) g 4.
Suppression Cliamber Water Temperature RSP*
1 (CAC-TR-778-7) o u
E 5.
Drywell Pressure RSP*
1 (CAC-PI-3341 and CAC-PT-2599) 6.
Drywell Temperature RSP*
1 (CAC-TR-778-1,3,4) 7.
Drywell Oxygen Concentration local Panel 1
(CAC-AT-1259-2)
TRenute Shutdown Panel, Reactor Building 20' Elevation
!ABLE 4.3.5.2-1 REMOTE SilVIDOWN MONI10RiflG lilSTRUMENTATION SURVEILLANCE REQUIREMEllIS c
Bi5 9
CilAllNEL CllANNEL E-IUtlCTI0llAL UtilT AND INSTRUMENT NUMBER CIIECK CALIBRATION O
I 1.
Reactor Vessel Pressure M
Q (021-PI-3332 and c32-PT-H005A) 2.
Q (021-LI-3331. B21-LI-R604AX, B21-LT-N027 and 021-LT-IIO26A) l 3.
Suppression Chamber Water Level M
R (CAC-LI-3342 and CAC-LT-2601)
R 4.
Suppression Chamber Water Temperature M
R
{
(CAC-TR-778-7) e 5
Drywell Pressure (CAC-PI-3.Wi and CAC-PT-2599)
M Q
6.
Drywell Tempera tur~e (CAC-TR-778-1,3,4)
M R
7.
Drywell Oxygen Concentration (CAC-AT-1259-2)
M Q
Ta bl e 3. 3. 5. 3-1 POST-ACCIDENT MONITORING INSTRUMENTATION MINIMUM NO.
OF OPERABLE INSTRUMENT INSTRUMENT AND INSTRUMENT NUMBER CHANNELS 1.
(B21-LITS N026A,B; B21-LR-615; B21eLI-R604A,,B and B21-LI'rS-NO37) 2.
Reactor vessel pressure 2
(B21-PI-R004A,B; C32-LPR-R608 and C32-PT-N005A,B) 3.
Containment pressure 2
(CAC-PI-2599; CAC-PT-2599; CAC-PR-1257-I anc CAC-PT-1257-1) 4 Containment pressure 2
(CAC-TR-1258-1 thru 13,22,23,24 and C91 P602) 5.
Suppression chamber atmosphere temperature 2
(CAC-TR-1258-17 thru 20 and C91-P602) 6.
Suppression enamber water level 2
[
(CAC-LI-2601-3; CAC-LR-2602; CAC-LT-2601; CAC-LT-2602 and CAC-LY-2601-1) 7.
Suppression chamber water temperature 2
(CAC-TR-1258-14, 21 and C91-P602) 8.
Containment radiation 2
(CAC-AR-1260; CAC-AQH-1260-1,2,3; CAC-AR-1261; CAC-AQH-1261-1,2,3; CAC-AR-1262 and CAC-AQH-1262-1,2,3) 9.
Containment oxygen 2
(CAC-AT-1259-2; CAC-AR-1259; CAC-AT-1263-2 and CAC-AR-1263) 10.
Containment hydrogen 2
(CAC-AT-1959-1; CAC-AR-1259; CAC-AT-1263-1 and CAC-AR-1263)
BRUNSWICK-UNIT 2 3/4 3-51
TABLE 4.3.5.3-1 POST-ACCIDENT HONITORING lilSTRUMENTATION SURVEILLAllCE REQUIREMENTS E55 CilANNEL CllANflEL Q
INSTRUMENT AND INSTRUMENT fiUMBER CllECK CALIBRATION e5 1.
R
]
(B21-LITS-fl026A,0; 021-LR-R615; B21-LI-R604A,B and B21-LITS-NO37) 2.
Reactor vessel pressure M
R l
(021-PI-R004A,0; C32-LPR-R608 and C32-PT-N005A,0) 3.
Containment pressure H
R (CAC-PI-2599; CAC-PT-2599; CAC-PR-1257-1 and CAC-PT-1257-1) 4.
Containment temperature H
R wh (CAC-TR-1258-1 thru 13,22,23,24 and 091-P602)
[,
5 Suppression chamber atmosphere temperature M
R (CAC-TR-1258-17 thru 20 and C91-P602) 6.
Supression chamber water level H
R (CAC-LI-2601-3; CAC-LR-2602; CAC-LT-2601; CAC-LT-2602 and CAC-LY-2601-1) 7.
Suppression chamber water temperature H
R (CAC-TR-1258-14, 21 and C91-P602) 8.
Containment radiation H
R (CAC-AR-1260; CAC-AQll-1260-1,2,3; CAC-AR-1261; CAC-AQll-1261-1,2,3; CAC-AR-1262 and CAC-AQil-1262-1,2,3) 9.
Containment oxygen concentration H
R (CAC-AT-1259-2; CAC-AR-1259; CAC-AT-1263-2 and CAC-AR-1263) 10.
Containment hydrogen concentration M
R (CAC-AT-1259-1; CAC-AR-1259; CAC-AT-1263-1 and CAC-AR-1263)
TABLE 3.3.5.6-1 Si CHLORIDE INTRUSION MONITORS 5
Q MINIMUM NUMBER FUNCTIONAL UNIT AND INSTRUMENT NUMBER OPERABLE CHANNELS (,)
b 1.
Chloride leak detectors in the condenser 4
]
hotwell outlet headers (C0-CR24) 2.
Chloride leak detector in the condensate 1
pump discharge (C0-CIS-3075-1) (TS-CR-863) 3.
Chloride leak detector in the inlet to the 1
condensate filter demineralizer (CFD-CIT-1) 4.
Chloride leak detector in the inlet to the I
bed demineralizer (CDD-CIT-1) w E
Chloride intrusion can be detected if any of the functional units have their required a.
minimum number of channels CPERABLE.
TABLE 3.3.5.6-2 5
CilLORIDE IllTRUSI0tl M0filTORS SETP0lllTS E
R7 filflCTI0tlAL UillT Afl0 IllSTRuf1EilT lluf1BER ALARM SETP0lflT ALLOWABLE LIMIT E
- )
1.
Chloride leak detectors in the condenser i 1.0 pmhos/cm 1 2.0 pmhos/cm hotwell outlet headers (C0-CR24) no 2.
Chloride leak detector in the condensate pump discharge (00-CIS-3075-1 Wide Rane,e) 1 (2.0 pmhos/cm for wide
~< (10.0 nmhos/cm for wide range nonitor) range monitor)
(TS-CR-863 Narrow Range) 1 0.3 pmhos/cm 1 0.5 pmhos/cm 3.
Chloride leak detector in the inlet to the 1 0.3 umhos/cm 1 0.5 pmhos/cm filter demineralizer (CFD-CIT-1)
M i'
the deep bed demineralizer (CDD-CIT-1)
~~< 0.3 pmhos/cm
-< 0.5 pmhos/cm 4.
Chloride leak detector in the inlet to "I
o, TABLE 4.3.5.6-1 E
Bi CHLORIDE INTRUSION MONITORS SURVEILLANCE REQUIREMENTS 5
9 CHANNEL 5
CHAL'!EL FUNCTIONAL CHANNEL
[
FUNCTIONAL UNIT AND INSTRUMENT NUMBER J;dCK_
TEST CALIBRATION 1.
Chloride leak detector in the condenser D
M R
hotwell outlet headers (C0-CR24) 2.
Chloride leak detector in the D
M SA condensate pump discharge (C0-C15-3075-1)
[
(TS-CR-863) 3.
Chloride leak detector in the inlet to D
M SA the condensate filter demineralizer j
(CFD-CIT-1)
Y 4.
Chloride leak detector in the inlet to D
M SA the deep bed demineralizer (CDD-CIT-1)
INSTRUMENTATION FIRE 0: 1:CTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.7 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3.5.7-1 shall be OPERABLE.
APPLICABILITY:
Whenever equipment in that fire detection zone is required to the OPERABLE.
ACTION:
With one or more of the fire detection instrument (s) shown in Table 3.3.5.7-1 inoperable:
a.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, increase the inspection frequency for the zone (s) with the inoperable instrument (s) tc at least once per hour, and b.
Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.5.7.1 Each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST.
4.3.5.7.2 The non-supervised circuits between the local panels l
associated with the detector alarms of each of the above required fire detection instruments and the control room shall be demonstrated OPERABLE at least once per 31 days in accordance with approved procedures.
BRUNSWICK - UNIT 2 3/4 3-59
' REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING COND! TION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
b.
5 gpm UNIDENTIFIED LEAKAGE averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
c.
25 gpm total leakage averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
APPLICABILITY:
CONDITIONS 1, 2 and 3.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within' the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With any reactor coolant system leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within the limits within B hours or be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REOUIREMENTS 4.4.3.2 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
a.
Monitoring the drywell and equipment drain sump flow rates at least once oer 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> nnd (G16-FQ-K603; G16-FQ-K601; G16-FY-M50 2; y
G16-FY-K601; G16-FT-N013 and G16-FT-N003) b.
Monitoring the primary containment atmospheric particulate and gaseous radioactivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(CAC-AQH-1260-1,2,3; CAC-AQH-1262-1,2,3 andCAC-AQH-1261-1,2,3) l BRUNSWICK - UNIT 2 3/4 4-6
C REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1020 psig.
APPLICABILITY:
CONDITION 1* and 2*.
ACTION:
With the reactor steam dome pressure exceeding 1020 psig, rectuce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMETUS 4.4. 6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Not applicable during anticipated transients, reactor isolation or reactor trip.
C BRUNSWICK - UNIT 2 3/4 4-18
EMERGEflCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMErlTS (Continued) 2.
Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct
- Dosition, b.
At least once per 92 days, by verifying that the system develops a flow of at least 4250 gpm for 3 system head corresponding to a reactor pressure of a 1000 psig when steam is being supplied to the turbine at 1000, +20, -80, psig c.
At least once per 18 months by:
1.
Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verfying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel is excluded from this test.
2.
Verifying that the system develops a flow of at least 4250 gpm for a system head corresponding to a reactor pressure of E 165 psig when steam is being supplied to the l
turbine at 165, + 15, psig.
3.
Verifying that the suction for the HPCI system is auto-matically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal or suppression pool high water level signal.
BRUNSWICK - UNIT 2 3/4 5-2
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM LIMITING CONDITION FOR OPERATION 3.5.2 The Automatic Depressurization System (ADS) shall be OPERABLE with at least seven OPERABLE ADS valves.
APPLICABILITY:
CONDITIONS 1, 2 and 3 with reactor vessel steam dome pressure > 113 psig.
ACTION:
a.
With one ADS valve inoperable, POWER OPERATION may continue provided the HPCI, CSS and LPCI systems are OPERABLE; restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With two or more ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, c.
With the Surveillance Requirement of Specification 4.5.2.b not performed at the required interval due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applicable provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor vessel steam pressure is adequate to perfonn the tests.
SURVEILLANCE REQUIREMENTS 4.5.2 The ADS shall be demonstrated OPERABLE at least once per 18 months by:
a.
Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
b.
Manually opening each ADS valve when the reactor steam 6cce I
pressure is > 100 psig and observing that either; 1.
The control valve or bypass valve position responds accordingly, or 2.
There is a corresponding change in the measured steam flow.
BRUNSWICK - UNIT 2 3/4 5-3
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment internal pressure shall be maintained between -0.5 and 1.75 psig.
APPLICABILITY:
CONDITIONS 1, 2 and 3.
ACTION:
With the containment internal pressure outside of the specified limit, restore the internal pressure to within the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4. 6.1.5 The primary containment internal pressure shall be determined l
to within the limits at least once per 12 leurs.
BRUNSWICK - UNIT 2 3/4 6-7
J ADMINISTRATIVE CONTROLS 6.12 HICH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by papagraph 20.203(c)(2) of 10 CFR 20, each High Radiation Area in which the intensity of radiation is 1000 crem/hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a.
A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made know-ledgeable of them.
c.
An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation "ork Permit.
6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.
In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or the Plant Health Physicist.
6.13 Labeling - In lieu of 10CFR20.203(f), entcances to each building in which radioactive materials are used, stored, or handled, shall have signs bearing the legend, EVERY CONTAINER OR VESSEL IN THIS AREA MAY CONTAIN RADIOACTIVE MATERIAL.
- Health Physics personnel shall be exempt from the RWP issuance require-ment during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection pro-cedures for entry into high radiation areas.
BRUNSWICK - UNIT 2 6-21
ATTACHMENT 'd
~
UNIT #1 CHANCES A.
The following is a list by pages and ite= numbers of the changes due to typographical errors:
Pace Number Item 3/4 1-8 4.1.3.5.b.1 3/4 1-8 4.1.3.5.b.2 3/4 3-3 7
3/4 3-3 8
3/4 3-3 11 3/4 3-3 12 3/4 3-6 7
3/4 3-6 8
3/4 3-6 11 3/4 3-6 12 3/4 3-7 6
3/4 3-7 7
3/4 3-8 8
3/4 3-8 11 3/4 3-8 12 3/4 3-11 1.b 3/4 3-11 1.c.1 3/4 3-12 2.b 3/4 3-13 4.a.3 3/4 3-17 1.b 3/4 3-17 1.c.1 3/4 3-18 2.b 3/4 3-19 4.a.6 3/4 3-22 1.b 3/4 3-22 1.c.1 3/4 3-22 1.d 3/4 3-22 2.b 3/4 3-25 1.b 3/4 3-25 1.c.1 3/4 3-25 1.e 3/4 3-26 2.b 3/4 3-26 3.b 3/4 3-33 4.e 3/4 3-37 2.b 3/4 3-43 1.d 3/4 3-46 1
3/4 3-48 2
3/4 3-49 2
Attachment IA (continued)
Pace Number Item 3/4 3-51 6
3/4 3-52 2
3/4 3-52 6
3/4 3-56 2
3/4 3-57 2
3/4 3-58 2
3/4 3-59 4.3.5.7.2 3/4 4-18 4.4.6.2 3/4 5-3 4.5.2.b 3/4 6-7 4.6.1.5 B.
The following is a list by page and item number of the changes due to incorrect instrument numbers in the original issue of the Standard Technical Specifications.
Pace Number Item 3/4 3-48 1
3/4 3-49 1
3/4 4-6 4.4.3.2.a 3/4 4-6 4.4.3.2.b
ATTACFE NT IA (cont'd)
UNIT #1 CHANGES C.
The following explanations are given for individual changes since tne reasons for these changes cannot be grouped together generically.
1.
Page No. 3/4 3-7, Ite= No. 2.a Changing the channel calibration frequency for this item will conform it to the same frequency as other trips on the same instruments.
2.
Page No. 3/4 3-7, Item No. 2.c The 120% fixed high flux trip cannot be tested in Conditions 2 and 5 because its associated alarms and indications would be masked by the 15% flux trip at these power levels. The 1207. trip is also not required tu be operable in other than Condition 1. (Refer to Item 2.c, Table 3.3.1-1, page # 3/4 3-2.)
3.
Page No. 3/4 3-15, Ite= No. 5.b. and Page No. 3/4 3-29, Item No. 5.b.
In operational conditions 1 and 2 reactor steam dome pressure is always greater than 140 psig and administrative 1y and operationally the valves in Group 8 are closed.
Therefore, operational conditions 3,4 and 5 are the only times group 8 valves can be opened and functionally checked.
4.
Page No. 3/4 3-17, Item No. 1.C.4 Page No. 3/4 3-22, Item No. 1.c.4 Page No. 3/4 3-25, Item 1.c.4 There is no 40% high steamflow trip on Uhit No. 1.
The reason for not having this trip is given in the FSAR on Page No. 7.3-26, item 20).
5 Page No. 3/4 3-57, Item No. 2 This change establishes an ALL0b'ABLE LIMIT for the vide range monitor.
ATTACHMENT IA (continued)
D.
The following is a list by pages and item number for changes to the " Definitions" of Appendix A:
Pane Nteber 1-6 Add definition of STAGGERED TEST BASIS as Item 1.31.
This addition necessitates renumbering existing items as follows:
Thernal Power to 1.32.
Total Paaking Factor to 1.33.
Unidentified Leakage to 1.34.
1-7 1.
Change Frequency of SA to 184 days.
2.
Add A (A =al) Frequency of 366 days.
3.
Change Frequency of R to 550 days.
E.
Other Changes Add to Appendix A, new paragraph as follows:
6.13 Labeling - In lieu of 10CFR20.203(f), entrances to each building in which radioactive materials are used, stored, or handled, shall have signs bearing the legend, EVERY CONTAINER OR VESSEL IN THIS AREA MAY CONTAIN RADIOACTIVE MA'f m AL.
ATTACMENT IB Revised pages to it:plement the requested changes fcr Unit No. 1 O
DEFINITIONS SHUTDOW MARGIN 1.3 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be suberitical assuming that all control rods capable of insertion are fully inserted except for the annlytically 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.
1.31 Stancered Test Basis A Staggered Test Basis shall consist of:
a.
A test schedule for n systems, subsystems, trains or designated co=ponents obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one syste=, subsysten, trair or designated components at the beginning of each subinterval.
THDL$1AL POWER 1.32 Tmmut power shall be the total reactor core heat transfer rate to the reactor coolant.
TOTAL PEAKING FACTOR 1.33 The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LEGR 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 1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
BRUNSWICK - UNIT 1 1-6
TABLE 1.I FREOUENCY NOTATION NOTATION FREOUENCY S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days.
A At least once per 366 days.
R At least once per 550 days.
S/U Prior to each reactor startup.
N.A.
Not applicable.
BRUNSWICK - UNIT 1 1-7
REACTIVITY CONTROL SYSTEMS CONTROL R0D SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.
APPLICABILITY:
CONDITIONS 1, 2 and 5*.
ACTION:
a.
In CONDITION 1 or 2 with one control rod scram accumulator inoperable, the provisions of Specification 3.0.4 are not appli-cable and operation may continue, provided that within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s:
1.
The inoperable accumulator is restored to OPERABLE status, or 2.
The control rod associated with the inoperable accu-mulator is declared inoperable, and the requirements of Specification 3.1.3.1 are satisfied.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
V b.
In CONDITION 5* with a withdrawn control rod scram accumulator inoperable, fully insert the affected control rod and elec-trically disarm the directional control valves within one hour.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.3.5 The control rod scram accumulators shall be determined OPERABLE:
a.
At least once per 7 days by verifying that the pressure and leak detectors are not in the alarmed condition, and b.
At least once per 18 months by performance of a:
1.
CHANNEL FUNCTIONAL TEST of the leak detectors (C11-LS-129-xxxx), and 2.
CHANNEL CALIBRATION of the pressure detectors (c11-PS-130-xxxx).
- At least the accumulator associated with each withdrawn control rod.
Not applicable to control rods removed per Specification 3.9.10.1 or
[
3.9.10.2.
V BRUNSWICK UNIT i 3/4 1-8
TABLE 3.3.1-1 (Continued) h REACTOR PROTECTION SYSTEM lilSTRUMENTATI0ff tb5 APPLICABLE MilllHUM NUMBER p
OPERATIONAL OPERADLE CllANNELS FUllCTIO!!AL UillT AllD ItiSTRUMElli ilUMBER C0!!DITI0fiS PER TRIP SYSTEM (a)(IO ACTI0il 7.
Drvwell Pressure - Illgh 1, 2
- 2 6
(C71-PS-N002 A,B,C,D) 8.
Scram Discharge Volume Water level -
liigh (011 -LSit-N013 A,B,C,D) 1,2,5(7) 2 5
9.
Turbine Stop Valve - Closure 1 9 4
8 (LilC-SV05-lX,2X,3X 4X) 10.
Turbine Control Valve fast Closure, Control Oil Pressure - Low 1(g) 2 8
g (EllC-PSL-1756,1757,1758,1759)
Y' 11.
Iteactor Mode Switch in Shutdown Position (C71A,-SI) 1,2,3,4,5 1
9 l
12.
Manual Scram (C71 A -S3A,B) 1,2,3,4,5 1
10 l
TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES 5E C5 FUNCTI0flAL UllIT AND INSTRUMENT ilUMBER RESPONSE TIME y
(Seconds) 1.
Intermediate Range Monitors (C51-IRM-K601 A,B,C,D,E,F,G,H):
5 a.
Neutron Flux - High*
NA
]
b.
Inoperative NA 2.
Average Power Range Monitor * (CSI-APRM-Cil.A,B,C,D,E,F):
a.
Neutron Flux - liigh,15%
1 0.09 b.
Flow Biased Neutron Flux - liigh NA c.
Neutron Flux - liigh, 120%
$ 0.09 d.
Inoperative NA e.
Downscale NA f.
LPRM NA 3.
Reactor Vessel Steam Dome Pressure - liigh (821-PS-N023 A,B,C,0)
< 0.55 4.
Reactor Vessel Water Level - Level #1 (B21-LIS-N017 A B.C.D)
< 1.05 w
]
5.
Main Steim Line Isolatinn Valve-Closure (B21-F022 A,B,C,0 and B21-F028 A,B,C,D) < 0.06 a
6.
Main Steam Line Radiation - High (D12-RM-K603 A,B,C,0)
NA 7.
Drywell Pressure - High (071 -PS-N002 A,B,C,D)
NA l
8.
Scram Discharge Volume Water Level - liigh (C11-LSil-N013 A,B,C,D)
NA l
9.
Turbine Stop Valve - Closure (EllC-SV05-lX,2X,3X,4X)
< 0.06 10.
Turbine Control Valve Fast Closure, Control Oil Pressure - Low (EHC-PSL-1756,1757,1758,1759)
< 0.08 11.
Reactor Mode Switch in Shutdown Position (C71 A -SI)
NA l
12.
Manual Scram (C71 A.-S3 A,B)
NA l
- Neutron detectors are exempt from response time testing.
Response time shall be measured from detector output or input of first electronic component in channel.
I
TABLE 4.3.1-1 REACTOR PROTECT 10ft SYSTEM IllSTRUMENTATION SURVEILLAtlCE REQUlt;MEllTS U2 E
CllAffilEL OPERATIONAL Ci fUticT10ilAL UtilT CilAllilEL FUtlCT10tlAL CilAtlNEL CollDIT10lls til WillCil
-)-
AllD lllS1RUMElli ilUMBER CllECK TEST CALIBRATI0ff "
SURVEILLANCE REQUIRED n
i 1.
Intennediate llange Monitors:
g (C51-IRM-K601 A,B,C,D,E,f,G II)
IC U
a.
lieutron Flux - liigh D
S/U R
2 D
W R
3,4,5 b.
Inoperative flA W
tlA 2,3,4,5 2.
Average Power Range Monitor:
(C51-APRM-Cll. A,0,C,D,E,F)
S/U '), W(d)
It 9
2 l
a.
Heutron Flux - Illgh 15%
S h.
Flow Blased fleutron Flux-liigh 5
/U U, W (0,Q mg c.
Fixed fleutron Flux - liigh, S/U(b),W W(e),Q I
l 120%
S o,
d.
Inoperative flA W
flA 1, 2, 5 e.
Downscale flA W
NA 1
f.
LPRM D
HA (g) 1, 2, 5 3.
Reactor Vessel Steam Dome
' ressure - liigh flA H
Q 1, 2
'B21-PS-fl023 A,B,C,D) 4.
Reactor Vessel Water Level - Low level #1 (B21-LIS-fl017 A,B,C,0)
D H
Q 1, 2 5.
Main Steam Line Isolation Valve -
Closure (021-F022 A,B,C,0 and ilA
'M R(h) 1 021-002B A,B,C,D)
II)
RISI 6.
Main Steam Line Radiation - liigh S
H 1, 2 l
(012-RM-K603 A,B,C,0) 7.
Drywell Pressure - liigh flA H
Q 1, 2 l
(071.-PS-N002 A, B,C,D)
TABLE 4.3.l_
(Continued)_
REACTOR PROTECTION SYSTEM IfiSTRUMENTATI0ff SURVEILLAllCE REQUIREMEflTS CilANNEL OPERATIONAL g
FUNCTIONAL UNIT CilANNEL FUNCTIONAL CllANNEL CONDITIONS Ill WilICH g
AND IriSTRUMEllT NUMBER CHECK TEST
_ CALIBRATION SURVEILLANCE REQUIRED 8.
Scram Discharge Volume Water p
Level - liigh NA Q
R 1,2,5 g
(Cll-LSil-fl013 A,B,C,D)
I E
9.
Turbine Stop Valve - Closure NA M
R(h}
l 5
(EHC-SV05-lX,2X,3X,4X) 10.
Turbine Control Valve Fast Closure, Control Oil Pressure -
Low (EHC-PSL-1756,1757,1758,1759)
NA H
R 1
11.
Reactor Mode Switch in Shutdown NA R
NA 1,2,3,4,5 Position (C71 A-st) l
- 12. Manual Scram NA Q
NA 1,2,3,4,5 (C71A-S3A,B) l a.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
c.
The IRM channels shall be compared to the APRM channels and the SRM instruments for overlap during each startup, if not performed within the prevfous 7 days.
g En d.
When changing from CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering CONDITION 2.
This calibration shall consist of the adjustnent of the APRM readout to conform to the power values e.
calculated by a heat balance during CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER.
f.
This calibration shall consist of the adjustment of the APRM flow biased setpoint to conform to a calibrated flow signal.
g.
The LPRM's shall be calibrated at least once per effective full power month (EFPM) using the TIP system.
h.
This calibration shall consist of a physical inspection and actuation of these position switches.
1.
Instrument alignment uslag a standard current source.
J.
Calibration using a standard radiation source.
n n
n
- w
_ TABLE 3.3.2-1 g
I_ SOLATION ACTUATION INSTRUMEllTATION E
w VALVE GROUPS MilllMUM NUMBER APPLICABLE E
OPERATED BY OPERABLE CilANNELS OPERATI0flAL R
TRIP FUNCTI0ft AND INSTRUMENT NUMBER
_ SIGNAL (a)
PER TRIP SYSTEM (b)(c)
CONDITION ACTION g
1.
_P_RIMARY CGNTAlflMENT I_ SOLATION a.
Reactor Vessel Water level - Low 1.
Level #1 (821-LIS-N017 A,B,C,D) 2,3,6,a 2
1,2,3 20 2.
Level #2 (B21-LIS-N024 A,B and 1
2 1,2,3 20 B21-LIS-N025 A,B) b.
Drywell Pressure - liigh 2,6,8 2
1,2,3 20 (C71 -PS-N002 A,B,C,D) l w
c.
1.
Radiation - liigh (d) 1 2
1,2,3 21 (012-RM-K603 A,B,C,0) l u
L 2.
Pressure - Low 1
2 1
22 (B21-PS-H015 A,B,C,D) 3.
Flow - liigh 1
2/line 1
22 (B21-dPIS-N006 A.B.C.D; B21-dPIS-N007 A.B.C.D; B21-dPIS-N008 A,B,C,D; and B21-dPIS-N009 A,B,C,D) 4.
Flow - High 1
2 2, 3 21 (821-dPIS-H006A; B21-dPIS-N007B; B21-dPIS-N008C and B21-dPIS-N009D) d.
Main Steam Line Tunnel Temperature - liigh 1
2(e) 1, 2, 3 21 (B21-TS-N010 A,B,C,D; B21-TS-N011 A,B,C,0; B21-TS-N012 A,B,C,D and B21-TS-N013 A,B,C D) e.
Condenser Vacuum - Low 1
2 1,2(f) 21 (B21-PS-N056 A,B,C,0) f.
Turbine Building Area Temperature-High 1
4(e) 1, 2, 3 21 (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,0; B21-TS-3230 A,B,C,D; B21-TS-3231 A,B,C,0 and 821-TS'-3232 A,B,C,D)
~
ISOLA 1 aC T N iST l1El ATION 5
9 VALVE GROUPS MINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OP$RATf0NAL TRIP FUNCTION AND INSTRUMEllT NUMBER SIGNAL (a)
PER TRIP SYSTEM (b)(c)
CONDIrION ACTION c-5
]
2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Building Exhaust Radiation - liigh 6
1 1, 2, 3, S and
- 23 (D12-RM-fl010 A,B,)
b.
Drywell Pressure - High 2,6,8 2
1,2,3 23 l
(C71 -PS-N002 A,B,C,D)
I c.
R e or Vessel Water rel - Low, Level #1 2,3,6,a w
2 1,2,3 23 L
'(Bh LIS-N024 A,8 and B21-LIS-N025 A,B) 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High 3
1 1,2,3 24 (G31-dFS-N603-1A,1B) b.
Area Temperature - High 3
2 1,2,3 24 (G31-TS-N600A,0,C,D,E,F) c.
Area Ventilation a Temp. - High 3
2 1,2,3 24 (G31-TS-H602A,B,C,0,E,F) d.
SLCS Initiation (C41A-SI) 3 (g)
NA 1,2,3 24 e.
Low, Level # 1 2,3,6,8 2
1,2,3 24 (B21-LIS-fl024A,B and 821-LIS-N025A,B)
JABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION E;;
VALVE GROUPS MINIMUM NUMBER APPLICABLE 7:
OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION AND INSTRUMENT NUMBER SIGNAL (a)
PER TRIP SYSTEM (b)(c)
CONDITION ACTION E
- 1 4.
CORE STANDBY COOLING SYSTEMS ISOLATION a.
High Pressure Coolant Injection Isolation 1.
HPCI Steu Line Flow - High 4
2 1,2,3 25 (E41-dPIS-N004 and E41-dPIS-N005) 2.
HPCI Steam Supply Pressure -
Low (E41-PSL-N001A,B,C,0) 4 2
1,2,3 25 us 3;
3.
HPCI Steam Line Tunnel Temperature - High 4
2 1,2,3 25 u,
2.
(E41-TS-3314; E41-TS-3315; E41-TS-3316; E41-TS-3317; E41-TS-3318; l
E41-TS-3354; E41-TS-3488 and E41-TS-3489) 4.
Bus Power Monitor NA (h) 1/ bus 1, 2, 3 26 (E41-K55 and E41-K56) 5.
HPCI Turbine Exhaust Diaphragm Pressure - High 4
2 1,2,3 25 (E41-P5H-N012A,B,C,0) 6.
HPCI Steam Line Ambient Temperature - High 4
2 1,2,3 25 (E51-TS-N603C,0) 7.
HPCI Steam Line Area a Temp. -
Hig'i 4
2 1,2,3 25 (Ebi-dTS-N604C,D) 8.
Emergency Area Cooler Temperature - High 4
gp 2
1,2,3 25 (E41-TS-N602A,B)
TABLE 3.3.2-1 (Continued) m E
ISOLAT10ft ACTUAT10il INSTRUMENTATION R
VALVE GROUPS MINIMUM NUMBER APPLICABLE i
OPERATED BY OPERABLE CllANNELS OPERATIONAL g
TRIP _ FUllCII0tl Afl0 INSTRUMEllT ilUMBER
_ SIGNAL (a)
PER TRIP SYSTEM (b)(c)
CONDITION ACT10fl
[
5.
SilVIDOWil COOLING SYSTEM ISOLATION a.
Reactor Vessel Water - Low, level #1 2,3,6,3 2
3, 4, 5 25 (021-LIS-N017A,0,0,D) h.
Reactor Steam Dome Pressure-Illgh B
1 3,4,5 27 l
(B32-PS-N018A,0) t' ta
TABLE 3.3.2-2 ISOLAT10ft ACTUATION INSTRUMENTATI0il SETPOINTS v,
'5 ALLOWACLE Q
1 RIP FUllCTION AllD INSTRUMENT NUMBER TRIP SETPOINT VALUE 1.
PRIMARY CONTAINMENT ISOLAT10ft U
a.
Reactor Vessel Water level - low 1.
Level #1(B21-LIS-N017A,0,C,0)
> +12.5 inches
>t 12.5 inches 2.
Level #2 (021-LIS-N024 A,0 and 7 -38 inches s - 38 inches 021-LIS-H025 A,B) b.
Drywell Pressu e - Illgh
< 2 psig
< 2 psig (C71 -PS-NOCT A,B,C,D) c.
1.
Radiation - Idgh
~< 3 x full power background
< 3.5 x full power
[
(D12-RM-K603 A,B,C D)
Iiackground i'
2.
Pressure - Low
> 825 psig
> 825 psig C
(021-PS-N015 A,0,C.D) 3.
Flow - liigh
< 140% of rated flow
< 140% of rated flow (B21-dPIS-N006 A,B,C,0; Bdl-dPIS-N007 A,B,C,0; B21-dPIS-H008 A,B,C,0; aiid B21-dPIS-H009 A,B,C,D)
I d.
Main Steam line Tunnel Temperature - fligh
< 200*F
< 200"F (B21-IS-N010 A,B,C,D; B21-TS-N011 A,B,C,6; B21-TS N012 A,0,C,D; and B21-TS-II013 A,B,C.D) e.
Condenser Vacuum - Low
> 7 inches lig vacuum
> 7 inches lig vacuum (B21-PS-N056 A,B,C,0) t.
Turbine Building Area Temp - Illgh
< 200"F
< 200*F (B21-TS-3225 A,B,C,0; B21-TS-3226 A,3,C,6; B21-TS-3227 A,B,C,D; B21-TS-3228 A,B,C,0; 1121-TS-3229 A,B,C,0; B21-TS-3230 A,B,C,D; "21-TS-3231 A,B,C,D and B21-TS-3232 A,B,C,0)
TABLE 3.3.2-2 (Continued)
E g
ISOLATI0tl ACTUATI0li INSTRUMENTATI0ff SETPOINTS S
Q TRIP FUllCTION Afl0 IllSTRUMENT NUMBER ALLOWADLE TRIP SETP0lfli VALUE
- 2. SEC0tIDARY CONTAINMENT ISOLATI0ft c-Reactor Building Exhaust a.
Radiation - liigh
< 11 mr/hr
< 11 mr/hr (012-RM-N010 A,B) h.
Drywell Pressure - liigh
< 2 psig
< 2 psig l
(C71-PS-N002 A,B,C,D) c.
Reactor Vessel Water l evel - Low, level #1 (B21-LIS-N024A,B and B21-LIS-N025A,B)
-> 412.5 inches
> t12.5 inches 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow,- liigh
< 53 gal / min
< 53 gal / min (G31-dFS-N603-1A,10) b.
Area Temperature - Illgh
< 150*F
< 150*F (G31 -TS-il600A,B,C,D,E,F )
Area Ventilation Temperature ATemp-liigh < 50*F c.
< 50*F (G31-TS-N602A,B,C,0,E,F) d.
NA flA Reactor Vessel Water - Iow, Level #1
> 412.5 inches
> 112.5 inches e.
(B21-LIS-N024A,B and B21-LIS-N025A,D)
TABLE 3.3.2-2 (Continued)
_ISOLAT10ll AC10AT10N INSTRUMENTATIOl1 SETPOINTS ALLOWABLE g
TRIP FullCTI0ff AND INSTRUMENT llUMBER TRIP SETPOINT val.UE h
4.
CORE STAfIDBY COOLING SYSTEMS ISOLATI0fl 5
]
a.
liigh Pressure Coolant Injection Isolation 1.
IIPCI Steam Line Flow - liigh
< 300% of rated flow
< 300% of rated flow (E41-dPIS-fl004 and E41-dPIS-N005)~
2.
IIPCI Steam Supply Pressure - Low
> 100 psig
> 100 psig (E41-PSL-H001A,B,C,0) 3.
IIPCI Steam Line Tunnel Temperature -
liigh 5 200*F 5 200*F (E41-TS-3314; E41-TS-3315; E41-TS-3316; E41-TS-3317; E41-TS-3318; R
E41-TS-3354; E41-TS-3488 and E41-TS-3489) 4.
Bus Power Monitor NA NA y
(E41-K55 and E41-K56) 5.
IIPCI Turbine Exhaust Diaphragm Pressure - Illgh
< 10 psig
< 10 psig
( E41 -PSil-fl012A,B,C,0) 6.
IIPCI Steam Line Ambient Temp -
liigh (E51-TS-N603C D)
< 200*F
< 200 F l
7.
IIPCI Steam Line Area A Temp -
liigh (SI-dTS-H604C,D) 5 50*F 5 50 F (E51-dTS-N604C,D) 8.
Emergency Area Cooler Temp - liigh
< 175*F
< 175'F (E41-TS-N602A,0)
ISOLATION SYSTEM RESPONSE TIME TRIP FUNCTION AND INSTRUMENT NUMBER RESPONSE TIME (Seconds) 1.
PRIMARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level - Low 1.
Level #1 (B21-LIS-N017 A,B,C,D)
<13 2.
Levei #2 (B21-LIS-N024 A,B and 71.0**
B21-LIS-N025 A,B) b.
Drywell Pressure - High 5,13 (C71 -PS-N002 A,B,C,D) c.
Radiation - High*
1 3
0**
g (D12-RM-K603 A,B,C,D) 2.
Pressure - Low
<13 (B21-PS-N015 A,B,C,D) 3.
Flow - High
<0.5**
(B21-dPIS-N006 A,B,C,D; B21-dPIS-N007 A,B,C,D; B21-dPIS-N008 A,B,C,D and B21-dPIS-N009 A,B,C,D) d.
Main Steam Line Tunnel Temperature - High (B21-TS-N010 A,B,C,D; B21-TS-N0ll A,B,C,D;
-<l3 B21-TS-N012 A,B,C,D; and B21-TS-N013 A,B,C,D)
I e.
Condenser Vacuum - Low
<l3 (B21-PS-N056 A,B,C,D) f.
Turbine Building Area Temperature - High NA (B21-TS-3225 A,B,C,D; B21-TS-3226 A,B,C,D; B21-TS-3227 A,B,C,D; B21-TS-322B 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 and B21-TS-3232 A,B,C,D) 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactu' Building Exhaust Radiation - High*
<l3 (D12-RM-N010 A,B) b.
Drywell Pressure - Hig)h
-<13 l
(C71-PS-N002 A,B,C,D t
c.
Reactor Vessel Water Level - Low, Level # 1
~ ~ <13 (B21 -LIS-NO 24 A.B and B21-LIS-N025A,B)
W Radiation monitors are exempt from response time testing.
Resconse time shall be measured from detec or output or the inout of the first electronic component in the channel.
- Isolation actuation instrumentation resconse time only.
BRUNSWICX-UNIT 1 3/4 3-22
TABLE 4.3.2-1 ISOLATION ACTUATI0li INSTRUMO.TATION SURVEILLANCE REQUIREMENTS CllAfillEL OPERATI0llAL E
CilANNEL FullCTIONAL CilANNEL CONDITIONS IN WillCil TRIP fullCT10tl AND INSTRUMENT NUMBER CllECK TEST CAllDRATION SURVEILLAtlCE REQUIRED
{
1.
PRIMARY C0ffTAINMENT ISOLATION a.
Reactor Vessel Water Level - Low E
1.
Level #1 D
H Q
1, 2, 3 ti (021-LIS-N017 A,B,C,D) 2.
I.evel #2 D
H Q
1, 2, 3 (B21-LIS-N024 A,B and 021-LIS-N025 A,0) b.
Drywell Pressure - liigh NA H
Q 1, 2, 3 (071 -PS-N002 A,B,C,D) c.
Radiation - liigh D
W R
1, 2, 3 l
(012-RM-K603 A,B,C,D) w2 2.
Pressure - Low NA H
Q 1
(B21-PS-H015 A,0,C,0) m4 3.
Flow - liigh NA M
Q 1
(B21-dPIS-N006 A,0,C,0; B21-dPIS-N007 A,B,C,D; B21-dPIS-N000 A,B,C,0; and 021-dPIS-N009 A,B,C,D) d.
Main Steam Line Tunnel Temperature - liigh NA M
R 1, 2, 3 (B21-IS-N010 A,B,C,D; B21-TS-fl011 A,B,C,D; B21-TS-N012 A,0,C,0 and 021-TS-N013 A,B,C,D) e.
Condenser Vacuum - Low NA H
R 1, 2 (B21-PS-N056 A,B,C,0) f.
Turbine Building Area Temp-fligh NA ft R
1, 2, 3 (B21-IS-3225 A,B,C,D; B21-IS-3226 A,B,C,0; B21-TS-3227 A,B,C,0; B21-TS-3220 A,B,C,D; B21-IS-3229 A,B,C,0; B21-IS-3230 A,B,C,D; B21-TS-3231 A,F,C,0 and 021-TS-3232 A,B,C,0)
TWFen reactor steam pressure > 500 pstg.
TABLE 4.3.2-1 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE IlEQUIREMENTS 10 p;
CilANNEL OPERATIONAL CilANilEL FUNCTIONAL nlANNEL CONDl!!0NS IN WillCil 1 RIP fullCT10ft AND INSTRUMENT NUMBER CllECK TEST
. CALIBRATION SURVEILLANCE REQUIRED
'1 2.
SEC0tIDARY C0flTAINHENT ISOLATION a.
Reactor Building Exhaust fla<liation - liigh D
M R
1, 2, 3, 5 and *
(Pl2-ItM-N010 A,B) b.
Drywell Pressure - liigh NA H
Q 1, 2, 3 (C71-PS-N002 A,B,C,D) l c.
Reactor Vessel Water Level - Low, Level #1 D
M Q
1, 2, 3
's (B21-LIS-N024 A,B and B21-LIS-l1025 A,B) 3.
REAC10R WATER CLEANUP SYSTEM ISOLATION a.
a Flow - liigh D
M R
1, 2, 3 (G31-DIS-N603-1A,10) b.
Area Temperature - liigh NA H
R 1, 2, 3 (G31-TS-N604 A,B,C,0,E,F) l c.
Area Ventilation A Temp -
liigh (G31-TS-N602A,B,C,D,E,F)
NA H
R 1, 2, 3 d.
NA R
NA 1, 2, 3 e.
Low, Level #1 D
H Q
1, 2, 3 (B21-LIS-fl024A,B and 821-LIS-fl025A,B)
AWiiliaidiWiij Trradiated fuel in the secondary containment.
TABLE 4.3.2-1 (Continued) m E
2
_ISOLAT10tl ACTUATION INSTRUMENTATI0tt SURVEILLANCE REQUIREMENTS v,
5p CilANNEL OPERATIONAL CllANNEL FUNCTIONAL CllANilEL CONDITI0riS IN WillCil TRIP FUNCTIOil AND IllSTRUMENT NUMBER CllECK TEST CAllDRATION SURVEILLANCE REQUIRED c5
]
S.
_511UIDOWil C00LillG SYSTEM ISOLATI0tt a.
Reactor Vessel Water - Low, level #1 D
M Q
3, 4, 5 (B21-LIS-fl017A,0,C.D) b.
Reactor Steam Dome Pressure -
liigh (032-PS-N018A,0)
NA S/U*, M R
3,4,5 l
M Ilf not performed within tt:e previous 31 days.
lb
TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION IflSTRUMENTATION E
MINIMUM NUMBER APPLICABLE 7
OPERABLE CilANNELS OPERATIONAL E
TRIP FUNCTION AND INSTRUMENT NUMBER PER TRIP SYSTEM CONDITIONS
-4 3.
IIPCI SYSTEM a.
Reactor Vessel Water Level - Low, Level #2 2
1,2,3 (B21-LIS-H031A,B,C,D) b.
Drywell Pressure - liigh (Ell-PS-H0llA,B,C,0) 2 1, 2, 3 Condensate Storage Tank Level-Low **(E41-LS-N002, E41-LS-H003) NA*
1, 2, 3 c.
d.
Suppression Chamber Water Level-liigh** (E41-LSil-H015A,B)
NA*
1, 2, 3 e.
Bus Power Monitor # (E41-K5S and E41-K56) 1/ bus 1, 2, 3 4.
ADS a.
Drywell Pressure - liigh, coincident with (Ell-PS-N010A,B,C,D) 2 1, 2, 3 w
b.
Reactor Vessel Water Level - Low, level #3 2
1,2,3 d,
(021-LIS-H031A,B,C,D) c.
ADS Timer (021-TDPU-K5A,B) 1 1, 2, 3 d.
Core Spray Pump Discharge Pressure - liigh (Permissive) 2 1, 2, 3 (E21-PS-N008A,0 and E21-PS-N009A,B)
RilR (LPc1 MODE) Pump Discharge Pressure - liigh (Permissive) 2/ pump 1, 2, 3 l
e.
(Lil-PS-N016A,B,C,0 and Ell-PS-N020A,B,C,D) f.
Bus Power Monitor # (B21-KIA,B) 1/ bus 1, 2, 3
- Alann only.
When inoperable, verify power availability to the bus at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- When either channel of the automatic transfer logic is inoperable, align IIPCI pump suction to the suppression pool.
- Provides signal to llPCI pump suction valves only.
)
TABLE 4.3.3-1 g
EMERGENCY CORE C0OLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMEllTS h
CHANNEL OPERATIONAL M
CllANflEL FUNCTIONAL CHANNEL C0flDITIONS Ifl WilICil TRIP FUNCTION AND INSTRUMENT NUMBER CHECK TEST
,CALIBRATI0ft SURVEILLANCE REQUIRED q
1.
CORE SPRAY SYSTEM a.
Reactor Vessel Water Level - Low, level #3 (B21-LIS-NO31A,B,C,P',
D M
Q 1,2,3,4,5 b.
Reactor Steam Dome Pressure -
Low (B21-PS-N021A,B,C.Q NA M
Q 1,2,3,4,5 c.
Drywell Pressure - High NA M
Q 1, 2, 3 (Ell-PS-N0llA,B,C,D) d.
Time Delay Relay NA R
R 1,2,3,4,5 e.
Bus Power Monitor (E21-KlA,B)
NA R
NA 1,2,3,4,5 2.
y
-LPCI MODE OF RHR SYSTEM w
w a.
Drywell Pressure - High NA M
Q 1, 2, 3 2,
(Ell-PS-N0llA,B,C,D) b.
Reactor Vessel Water Level - Low, Level #3 (B21-LIS-N031A,B,C,D)
D H
Q 1, 2, 3, 4*, 5*
l c.
Reactor Vessel Shroud Level NA H
Q 1, 2, 3, 4*, 5*
B21-LITS-N036 and B21-LITS-N037) d.
Reactor Steam Dome Pressure -
Low (B21-PS-N021A,B C.D)
NA H
Q 1, 2, 3, '4 *, 5
- e.
RilR Pump Start-Time Delay Relay NA R
R 1, 2, 3, 4*, 5*
f.
Bus Power Monitor (Ell-K106A,8)
NA R
NA 1, 2, 3, 4 *, 5*
- Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.
1 i
~
TABLE 4,.
.4-1 CONTROL RDD WITilDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIR
~
CllANNEL OPERATIONAL CilANNEL FUNCTIONAL CilANNEL 00hblT10NS IN WillCll y
TRIP FullCTION AllD INSTRUMENT NUMBER CllECK TEST CAllBRATION
- SURVEILLANCE REQUIRED v.
1.
APR_H,(C51-APRM-Cll.A B.C.D.E.F)
Upscale (Flow Blased)
NA S/U
,H R
1 n
a.
C 7
h.
Inoperative NA S/U
,Q NA 1, 2, 5 C
E c.
Downscale NA S/U HA 1
S/U(C} H Q
d.
Upscale (Fixed)
NA
,W Q
2, 5 2.
ROD DLOCK HONITOR (CSI-RBH-Cll.A,B) a.
'Jpsca le flA S/U c,H R
1*
C}
h.
Inoperative ilA S/U Q
NA 1*
c.
Downscale NA S/U c,H R
1*
3.
SOURCE ItANGE H0lilTORS (C51-SRM-K600A,B.C.D) a.
Detector not full in NA S/U C} W NA b.
Upscale NA S/U c,W 2, 5 NA 2, 5 c.
Inoperative NA S/U W
NA d.
Downscale NA S/Ug,W 2, 5 u
tlA 2, 5 4.
INTERHEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,0,E,F,G.II)
,w a.
Detector not full in NA S/U(C}W(d) f(A 2
flA W
NA 5
b.
Upscale NA S/U
,W NA 2
C NA W
NA 5
c.
Inoperative flA S/U(c) 9(d) f1A 2
NA W
NA 5
d.
Downscale NA S/U IC,W(d) flA 2
te W
NA 5
CllAtillEL CAllDRATibtf5 are electronic, a.
b.
This calibration shall consist of the adjustment of the APRH flow biased setphtnt to conform to a calibrated flow signal.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
c.
d.
When changing from CONDITION 1 to C0flDIT10N 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a f ter entering CONDITION 2.
When TilEllHAL POWER is greater than the preset power level of the RWH and RSCS
TABLE 4.3.5.1-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL INSTRUMENTS, SENSOR CHANNEL FUNCTIONAL CHANNEL LOCATIONS AND INSTRUMENT NUMBER CHECK TEST CALIBRATION 1.
Triaxial Time-History Accelographs a.
Reactor Building +65' level on Drywell M*
SA R
(ENV-IT-823-2) b.
Reactor Building -17' level (ENV-IT-823-1 and M*
SA R
ENV-IT-823-3)
- Except seismic trigger BRUNSWICK-UNIT 1 3/4 3-46
TABLE 3.3.5.2-1 E
REMOTE SilVTDOWN HONITORING INSTRUMENTATION E
'n 7
HINIHUM E
READ 0UT CilANNELS tj FUNCTIONAL UNIT AND INSTRUMENT NUMBER LOCATION OPERABLE 1.
Reactor Vessel Pressure RSP*
1 (B21-PI-3332 and c32 PT-N005A) l 2.
Reactor Vessel Water Level RSP*
1 (021-LI-3331, B21-LI-R604AX, B21-LT-N027 and B21-LT-H026A) l 3.
Suppression Chamber Water Level RSP*
1 (CAC-LI-3342 and CAC-LT-2601) 4.
Suppression Chamber Water Temperature RSP*
1 u,
];
(CAC-TR-778-7)
[,
S.
Drywell Pressure RSP*
I (CAC-PI-3341 and CAC-PT-2599) o' 6.
Drywell Temperature RSP*
1 (CAC-TR-778-1,3,4) 7.
Drywell Oxygen Concentration Local Panel 1
(CAC-AT-1259-2) 3-Remote Shutdown Panel, Reactor Building 20' Elevation
TABLE 4.3.5.2-1 g
IREMOTE Sil0TDOWN HONITORING INSTRUMENTATIDH SURVEILLA4CE REQUIREHENTS s
n 7
CllANNEL CllANNEL E
FUNCTIONAL UNIT AND INSTRUMENT NUMBER CllECK CAllDRATION 1.
Reactor Vessel Pressure H
Q l
(021-PI-3332 and C32 -PT-N005A) 2.
Q g
(021-LI-3331, 021-LI-R604AX, 021-LT-N027 and B21-LT-N026A) i 3.
Suppression Chamber Water Level H
R (CAC-Li-3342 and CAC-LT-2601) 4.
Suppression Chamber Water Temperature H
R (CAC-TR-778-7) i, 7
5.
Drywell Pressure (CAC-PI-3341 and CAC-PT-2599)
H Q
6.
Drywell Temperature (CAC-TR-778-1,3,4)
H R
7.
Orywell Oxygen Concentration (CAC-AT-1259-2)
H Q
Table 3.3.5.3-1 POST-ACCIDENT MONITORING INSTRUMENTATION MINIMUM NO.
OF OPERABLE INSTRUMENT
, INST.7 : MENT AND INSTRUMENT NUMBER CHANNELS 1.
Reacter vessel wa:er level 2
(B21-LITS-N026A,B; B21-LR-615; B21-LI-R604A,B and B21-LITS-NO37) 2.
Reactor vessel pressure 2
(B21-PI-R004A,B; C32-LPR-R608 and C32-PT-N005A,B) 3.
Containment pressure 2
(CAC-PI-2599; CAC-PT-2599; CAC-PR-1257-1 and CAC-PT-1257-1) 4.
Containment pressure 2
(CAC-TR-1258-1 thru 13,22,23,24 and C91-P602) 5.
Suppression chamber atmosphere temperature 2
(CAC-TR-1258-17 thru 20 and C91-P602) 6.
Suppression chamber water level 2
(CAC-LI-2601-3; CAC-LR-2602; CAC-LT-2601; g
CAC-LT-2602 and CAC-LY-2601-1) 7.
Sucpression chamber water temperature 2
(CAC-TR-1258-14, 21 and C91-P602) 8.
Containment radiation 2
(CAC-AR-1260; CAC-AQH-1260-1,2,3; CAC-AR-1261; CAC-AQH-1261-1,2,3; CAC-AR-1262 and CAC-AQH-1262-1,2,3) 9.
Containment oxygen 2
(CAC-AT-1259-2; CAC-AR-1259; CAC-AT-1253-2 and CAC-AR-1263) 10.
Containment hydrogen 2
(CAC-AT-1959-1; CAC-AR-1259; CAC-AT-1263-1 and CAC-AR-1263)
BRUNSWICK-UNIT 1 3/4 3-51
TABLE 4.3.5.3-1 g
POST-ACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5e p;
CilANNEL CilANNEL p
INSTRUMENT AND INSTRUMENT NUMBER CllECK CALIBRATION Eq 1.
R (B21-LITS-N026A,0; B21-LR-R615; B21-LI-R604A,B and 821-LITS-N037) 2.
Reactor vessel pressure H
R (B21-PI-R004A,B; C32-l.PR-R608 and C32-PT-N005A,B) g 3.
Containment pressure H
R (CAC-PI-2599; CAC-PT-2599; CAC-PR-1257-1 and CAC-PT-1257-1)
Id 4.
Containment temperature M
R (CAC-TR-1258-1 thru 13,22,23,24 and C91-P602)
Y";
5.
Suppression chamber atmosphere temperature H
R (CAC-TR-1250-17 thru 20 and C91-P602) 6.
Supression chamber water level H
R (CAC-L1-2601-3; CAC-LR-2602; CAC-LT-2601; CAC-LT-2602 and CAC-LY-2601-1) 7.
Suppresstan chamber water temperature H
R (CAC-TR-1258-14, 21 and C91-P602) 8.
Containment radiation H
R (CAC-AR-1260; CAC-AQll-1260-1,2,3; CAC-AR-1261 ; CAC-AQil-1261-1,2,3; CAC-AR-1262 and CAC-AQil-1262-1,2,3) 9.
Containment oxygen concentration H
R (CAC-AT-1259-2; CAC-AR-1259; CAC-AT-1263-2 and CAC-AR-1263) 10.
Containment hydrogen concentration H
R (CAC-AT-1259-1; CAC-AR-1259; CAC-AT-1263-1 and CAC-AR-1263)
i TABLE 3.3.5.6-1 CHLORIDE INThUSION HONITORS Fir E
HINIMUM NUMBER 7
FUNCTIONAL UNIT AND INSTRUMENT NUMBER OPERABLE CHANNELS (a) 1.
Chloride leak detectors in the condenser 4
notwell outlet headers (C0-CR24)
H 2.
Chloride leak detector in the condensate 1
pump discharge (C0-CIS-3075-1)
(TS-CR-8bJ) l 3.
Chloride leak detector in the inlet to the 1
condensate filter demineralizer (CFD-CIT-1) 4.
Chloride leak detector in the inlet to the 1
bed demineralizer (CDD-CIT-1) m h
Chloride intrusion can be detected if any of the functional units have their required a.
minimum number of channels OPERABLE.
O O
O
y TABLE 3.3.5.6-2 CliLORIDE IllTRUS10tl M0tilTORS SETPOINTS
?
E
-FUtiCTI0ilAL UtilT AtlD INSTRUMEllT llUMBER ALARM SETP0litT ALLOWABLE LIMIT r;
1.
Chloride leak detectors in the condenser
< l.0 pmhos/cm hotwell outlet headers (CO-CR24)
~
~< 2.0 pmhos/cm 2.
Chloride leak detector in the condensate pump discharge (00-CIS-3075-1 wide Range) 1 (2.0 pmhos/cm for wide range monitor)
-~< (10 pmhos/cm for wide range monitor)
(TS-CR-863 Narrow Range) 1 0.3 pmhos/cm
< 0.5 pmhos/cm 3.
Chloride leak detector in the inlet to the 1 0.3 pmhos/cm 1 0.5 pmhos/cm
{
filter demineralizer (CFil-CIT-1) i' 4.
Chloride leak detector in the inlet to
< 0.3 pmhos/cm
~< 0.5 pmhos/cm t'!
the deep bed demineralizer (CDD-CII-1)
~~
TABLE 4.3.5.6-1 a,
E5 CHLORIDE INTRUSION MONITORS SURVEILLANCE REQUIREMENTS S
Ra CilANNEL E
CilANNEL FUNCTIONAL CHANT (EL
]
FUNCTIONAL UNIT AND INSTRUMENT NUMBER CllECK TEST
_ CALIBRATION 1.
Chloride leak detector in the condenser D
M R
hotwell outlet headers (C0-CR24) 2.
Chloride leak detector in the D
M SA condensate pump discharge (co-cIS-3075-1) g (TS-CR-863) 3.
Chloride leak detector in the inlet to D
M SA the condensate filter demineralizer R
(CFD-CIT-1) a y
4.
Chloride leak detector in the inlet to D
M SA g
the deep bed demineralizer (CDD-CIT-1)
O O
U
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.7 As a minimum, the fire detection instrttmentatten for each fire detection zone shown in Taele 3.3.5.7-1 shall be OPERABLE.
APPLICABILITY:
Whenever equipment in that fire detection zone is requireo to the OPERABLE.
ACTION:
With one or more of the fire detection instrument (s) shown in Table 3.3.5.7-1 inoperable:
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, increase the inspection frequency for the a.
zone (s) with the inoperable instrument (s) to at least once per hour, and b.
Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status.
The provisions of foecifications 3.0.3 and 3.0.4 are not c.
applicable.
SURVEILLANCE REOUIREMENTS 4.3.5.7.1 Each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST.
- 4. 3.5. 7. 2 The non-supervised circuits between the local panels I
associated with the detector alarms of each of the above required fire detection instruments and the control room shall be demonstrated OPERABLE at least once per 31 days in accordance with aporoved procedures.
BRUNSWICK - UNIT 1 3/4 3-59
REACTOR COOLANT SYSTD4 OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
b.
5 gpm IINIDENTIFIED LEAKAGE averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
25 gpm total leakage averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
c.
APPLICABILITY:
CONDITIONS 1, 2 and 3.
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN a.
within 12 nours and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With any reactor coolant system leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within the limits within 8 hcurs or be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SU,RVEILLANCE REOUIREMENTS 4.4.3.2 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
a.
Monitoring the drywell and equipment drain sumo flow rates at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and (G16-FQ-K603; G16-FQ-K601; G16-FY-K602; G16-FY-K601; G16-FT-N013 and G16-FT-N003) b.
Monitoring the primary containment atmospheric particulate and gaseous radioactivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(CAC-AQH-1260-1,2,3; CAC-AQH-1262-1,2,3 and CAC-AQH-1261-1,2,3)
BRUNSWICK - UNIT 1 3/4 4-6
4 REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1020 psig.
APPLICABILITY: CONDITION 1* and 2*,
ACTION:
With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4. 6.2 The reactor steam dome pressure shall be verified to be less l
than 1020 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Not applicable during anticipated transients, reactor isolation or reactnr trip.
BRUNSWICK - UNIT 1 3/4 4-18
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM LIMITING CONDITION FOR OPERATION 3.5.2 The Automatic Depressurization System (ADS) shall be OPERABLE with at least seven OPERABLE ADS valves.
APPLICABILITY:
CONDITIONS 1, 2 and 3 with reactor vessel steam dome pressure >l13 psig.
ACTION:
a.
With one ADS valve inoperable, POWER OPERATION may continue provided the HPCI, CSS and LPCI systems are OPERABLE; restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With two or more ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
With the Surveillance Requirement of Specification 4.5.2.b not
~'
performed at the required interval due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applicable provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor vessel steam pressure is adequate to perform the tests.
SURVEILLANCE REOUIREMENTS 4.5.2 The ADS shall be demonstrated OPERABLE at least once per 18 months by:
a.
Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
b.
Manually opening each ADS valve when the reactor steam dome l
pressure is > 100 psig and observing that either; 1.
The control valve or bypass valve position responds accordingly, or 2.
There is a corresponding change in the measured steam fl ow.
BRUNSWICK - UNIT 1 3/4 5-3
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment internal pressure shall be maintained between -0.5 and 1.75 psig.
APPLICABILITY:
CONDITIONS 1, 2 and 3.
ACTION:
With the containment internal pressure outside of the specified limit, restore the internal pressure to within the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REOUIREMENTS 4.6.1.5 The primary containment internal pressure shall be determined l
to within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
BRUNSWICK - UNIT 1 3/46-7
.N k
i ADbCNISTRATIVE CONTROLS 6.12 HIGE RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by papagraph 20.203(c)(2) of 10 CFR 20, each High Radiation Area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiaticm area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a.
A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made know-ledgeable of them.
c.
An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.
6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.
In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the arbinicerative control of the Shif t Foreman on duty and/or the Plant Health Physicist.
6.13 Labeling - In lieu of 10CFR20.203(f), entrances to each building in which radioactive materials are used, stored, or handled, shall have signs bearing the legend, EVERY CONTAINER OR VESSEL IN THIS AREA MAY CONTAIN RADIOACTIVE MATERIAL.
- Health Physics personnel shall be exe=pt from the RWP issuance require-ment during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection pro-cedures for entry into high radiation areas.
BRUNSWICK - UNIT 1 6-21