ML20073K796
| ML20073K796 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/01/1991 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20073K794 | List: |
| References | |
| NUDOCS 9105130013 | |
| Download: ML20073K796 (10) | |
Text
.
ENCI.OSURE 2 IIRUNSWICK STEA51 El.ECTRIC l'IANT, UNITS 1 ANI) 2 NRC DOCKET NOS. 50 325 & 50 324 OPERATING 1.lCENSE NOS. Di'R.71 & Di'it.62 SUPPLE 51ENT TO Hl: QUEST l'OR TE51PORARY WAIVElt OF COhll'LIANCE REACTOR WATER Cl.EANUI' SYSTEh! DilTERENTIAI, Fl.OW ISO!ATION INSTRUSTENT TECilNICAi; SI'l:Cil'ICATION l' AGES UNIT 1 9105130013 9105o1 PDR ADOCK 05000324 P
pga
8 TABLE 3.3.2-2 (Continued)
E I
E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 0
un Et i.
E ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE I
E 2.
SECONDARY CONTAINMENT ISOLATION 3
a.
Reactor. Building Exhaust Radiation - High
$ 11 mr/hr i 11 arr/hr
]
b.
Drywell Pressore - High 1 2 psig
$ 2 psig 3 + 112 inches (a) 112 inches *)
I i
c.
Reactor Vessel' Water Level - Lov, Level 2
+
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION
&)
100 a.
A Flow - High 1 M gal / min j
b.
Area Temperature - High
$ 150*F
$ 150*F g
e f
y c.
Area Ventilation a Temperature - High 1 50*r 1 $0*F l
1 f
e d.
SLCS Initiation NA NA 1
112 inches *)
3 + 112 inches (a)
I e.
Reactor Vessel Water Level - Low. Level 2
+
f.
S Flow - High - Time Delay Relay 145 secor.Js 1 45 seconds l
i e
k a
l 3
E z
.o
>W
I TABLE 3.3.2-2 (Continued)
E E
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E
R ALLOWABLE
~
VALUE TRIP SETPOINT TRIP FUNCTION E
Q 5.
SifUTDOWN COOLING SYSTEM ISOLATION 162.5 inches (a) 162.5 inches (a)
Reactor Vessel Water Level - Low Level 1 a.
i b.
Reactor Steam Dome Pressure - Ifigh
$ 140 psig
$ 140 psig U
r-Y u
M (a) Vessel. water level s ref er to REFERENCE LEVEL ZERO.
the methodology contained in the OFFSITE DOSE CALCULATION MANUAL (ODCM).
(b) Establish alarm / trip setpoint s per reaches 20% of (c ) The if ydrogen Water Chemi st ry (IfWC) syst em shall not be placed in service until reactor power background radiation 2
RATED THERMAL POWER.
After reaching 20! of RATED THERMAL POWER. the normal full power 3
E level and associated trip setpoints may be increased t o compensate for increased radiation levels as a result A
of full power operation with hydrogen injection.
Prior to decreasing power below 20 of RATED THERMAL POWER 3
and af ter the !!WC system bas been shut off, the baci< ground level and associated setpoint shall be ret urned to the normal full power values.
Control rod notion sha!I be suspended, when the reactor powcr is below 201 of RATED THERMAL POWER. until the necessary ad justment is made (except for scram or other emergency action).
z bSdsble'sk dr.'e sed.*d per -fke-Me.N e d e I.$y M e a e. rl L J sk h a y I k g,y, p
- }L M M *T t
i i
.,)
iNSTRlNENTATION
- a'.
BASES
- ml q~;.
',Q
- 2 3/4.3.2 ISOLATION ACTlRTION INSTRlHENTATION
.,l This specification ensures the ef fectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the trip settings for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Some of the trip settings have tolerances explicitly stated where both the high and low values are critical and may have a substantial ef fect on safety.
The setpoints of other instrumentation, where only the high or low end of the i
setting has a direct bearing on the safety, are established at a level away f rom the normal operating range to prevent inadvertent actuation of the systems involved.
DMech A - see hlIswIn) pak 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTl% TION INSTRlHENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to cont'rol.
This specification provides the trip point settings that will ensure ef fectiveness of the systems to provide the design protection.
Although the instrtsnents are listed by system, in some cases the same instrument is used to mend the start signal to several systems at the same time.
The out-of-service times for the instrtsnents are consistent with q,,
the requirements of the specifications in Section 3/4.5.
1 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRlNENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Rod Program Controls.
and Section 3/4.2 Power Distribution Limits.
The trip 1031c is arranged so l ]
that a trip in any one of the inputs will result in a rod block, bj 3/4.3.5 MONITORING INSTRINENTATION
]
3/4.3.5.1 SEISMIC MONITORING INSTRlNENTATION M
The OPERABILITY of the seismic monitoring instrunentation ensures that
- )
sufficient capability is available to promptly deterwe the magnitude of a JJ seismic event and evaluate the response of those f eatures important to
[i safety.
This capability is required to permit comparison of the rae a s u red
[
response to that used in the design basis f or the f acility.
2, il) i
.5 j
,.j BRtlNSWICK - UNIT 1 B 3/4 3-2 l
/
RETYPED TECH. SPECS.
)
Updated Thru. Amend. 53 l
INSF.RT A:
Reactor Water Cleanup Syssm DiffcIrntial Flow liigh The reactor water cleanup system (RWCU) high differential flow signal is provided to limit the release of radioactivity to maintain offsite and control room doses within the limits of 10 CFR 100 and 10 CFR 20. A time delay relay is provided to prevent spurious trips during most RWCU operational transients.
The high differential flow signals are initiated from transmitters that are connected to the inlet and outlets of the RWCU system. The outputs of the transmitters are arranged, compared and, if too large, initiate a trip signal. One channel of the high differential flow function is available and is required to be OPERABl.E in OPERATIONAL CONDITIONS 1,2, and 3.
The RWCU Differential Flow liigh Allowable Value has been determined using the ISA S67.041988 methodology. The Trip Setpoint is established far enough below the Allowable Value to accommodate the cumulative effects of instrument drift uncertainty, calibration uncertainty, all other random uncertainties and biases introduced by process effects, and a maximum of 10 percent analytical margin for the normal operating conditions. Seismic and accident effects are not included. A more restrictive tolerance is established as an acceptance criteria for periodic testing of the loop. That value excludes immeasurable loop uncertainty factors such as flow orifice effects, transmitter static pressure effects, and process temperature biases. The test acceptance criteria is controlled in engineering documents and plant procedures.
This function actuates the Group 3 valves.
I l
l
I ENCL.OSUlti: 3 liitUNSWICK STEAM ELECTitlC l'I. ANT, UNITS 1 AND 2 NitC DOCKET NOS. 50 325 & 50 324 Ol'EllATING 1.lCENSE NOS.1)1'It 71 & Dl'It 62 SUI'I't.EMENT TO itEQUEST l'Olt TEMl'OllAltY WAlVElt Ol? COMl'I. LANCE ItEACTOlt WATElt Cl.EANUl' SYSTEM DilTElli:NTIAl,l' LOW ISOLATION INSTI(UhlENT TECilNICAl. Sl'l:Cll'ICATION l' AGES UNIT 2 I
Y A
I Of TABLE 3.3.2-2 (Continued)
- n i
E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E
E ALLOWABLE x
VALUE TRIP SETPOINT TRIP FUNCTION E
3 2.
SECONDARY CONTAINMENT ISOLATION Il mr/hr
$ 11 mr/hr u
a.
Reactor Building Exhaust Radiaticn - High b.
Dryvell Pressure - liigh
$ 2 psig i 2 psig c.
Reactor Vessel Water Level - Low, Level 2 3 + 112 inches (*
3 + 112 inches
- 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION
-(N '--
_< M esIimin 100 a.
A Flow - High
< 150*F 1 150*F b.
Area Temperature - High
< 50*F l
l w
< 50*F Area Ventilation a Temperature - High 0
c.
e NA NA d.
SLCS Initiation 3 + 112 inches (a) 3 + 112 inches *)
I Reac* or Vessel **later Level - Low, Level 2 e.
f.
6 F'. ow - H i gh - Ti me De l a y Re l a y
$ 45 seconds 1 45 seconds l
1 1
a i
R E
z h
%D
l TABLE 3.3.2-2 (Continued) 1 g
E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 6
.C.
ALLOWABLE l
C)n VALUE TRIP SETPOINT TRIP FUNCTION E*j 5.
' SHUTDOWN COOLING SYSTEM ISOLATION
> 162.5 inches (a)
> 162.5 inches (a)
Reactor Vessel Water Level - Low Level 1 4.
Reactor Steam Dorie Pressure - High
$ 140 psig
$ 140 psig b.
l l
l l l l
U V
Y (a) Vessel water levels refer to REFEREP4CE LEVEL ZEHO.
the OFFSITE DOSE CALCULATION MANUAL (ODCM).
(b) Establish alarm / trip setpoints per t he methodology contained in reaches 201 of system shall not be placed in service until reactor power 2
(c ) The Hydrogen Wat er Chemi st ry (HWC)After reaching 201 of RATED THERMAL POWER, the normalin background radiation full power E
RATED THERMAL POWER.
increased to com;ensate for 1evel and as.ociated trip setpoints may be below 20% of RATED THERMAL POWER s
E of f ull power operation with hydrogen injection.
Prior to decreasing power shall be returned to off, the background level and associated setpoint and after the HWC system has been shut the normal full power values. Control rod motion shall be suspended, when the reactor power is below 20% of z
other emergency action).
RATED THERMAL POWER, until the necessary ad justment is made (except for scram or base 3fy,3,q, M *!4Y per fka m4kel,(
y MC 5c Nh)
(b b5 NSk b5f a
P
INSTRUKENTATION l
BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the affectiveness of the instrumentation used to mitigste the consequences of accidents by prescribing the trip settings f or isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Soms of the trip settings have tolerances explicitly stated where both the high and low values are critical and may have a substantial ef fect on safety.
The setpoints of other instrumentation where only the high or low end of the setting has a direct bearing on the saf ety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
(*1614tr4 A - see fododeh pa%
3/4.3.3 EMERGENCY CORE glING SYSTEM ACTUNIION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to control.
This specification provides the trip point settings that will ensure ef fectiveness of the systems to provide the design protection.
Although the instruments are listed by system, in some cases the same instrument is used to send the start signal to several systems at the same time.
The out-of-service times for the instruments are consistent with t he requirements of the specifications in Section 3/4.5.
3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in !ection 3/4.1.4, Rod Program Controls and Section 3/4.2 Power Dis tribution.dmit s.
The trip logic is arranged so that a trip in any one of the inputs will result in a rod block.
3/4.3.5 MONITORING INSTRUMENTATION 3/4.3.5.1 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumention ensures that sufficient capaht11ty is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.
This capability is required to permit comparison of the measured response to that used in the design basis for the facility.
BRUNSWICK - UNIT 2 B 3/4 3-2 RETYPED TECH. SPECS.
Updated Thru. Amend. 78
9 lleactor Water Clelump SntenLDifferential Flow - 1ligh The reactor water cleanup system (ItWCU) high differential flow signal is provided to limit the release of radioactivity to maintain offsite and control room doses within the limits of 10 CFil 100 and 10 CFil 20. A time delay ielay is provided to prevent spurious trips during most itWCU operational transients.
The high differential flow signals are initiated from transmitters that are connected to the inlet and outlets of the RWCU system. The outputs of the transmitters are arranged, compared and, if too large, initiate a trip signal. One channel of the high differential flow function is available and is required to be OPERAllt.E in OPEllATIONAL CONDITIONS 1,2, and 3.
The ItWCU Differential Flow liigh Allowable Value has been determined using the ISA-S67.041988 methodology. The Trip Setpoint is established far enough below the Allowable Value to accommodate the cumulative effects of instrument drift uncertainty, calibration uncertainty, all other random uncertainties and biases introduced by process effects, and a maximum of 10 percent analytical margin for the normal operating conditions. Seismic and accident effects are not included. A more restrictive tolerance is established as an acceptance criteria for periodic testing of the loop. That value excludes immeasurable loop uncertainty factors such as flow orifice effects, transmitter static pressure effects, and process temperature biases. The test acceptance criteria is controlled in engineering documents and plant procedures.
This function actuates the Group 3 valves.
_- - _ ______ - _ -____ - _ __