ML20065R246
| ML20065R246 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/26/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065R249 | List: |
| References | |
| NUDOCS 9405100281 | |
| Download: ML20065R246 (42) | |
Text
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4 UNITED STATES 5
5
- NUCLEAR REGULATORY COMMISSION 4,,..... j/
WASHINGTON, D.C. 20555-0001 PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 A G DMENT TO FACILITY OPERATING LICENSE Amendment No. 69 License No. NPF-39 1.
The Nuclear Regulatory Comission.(the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated April 19, 1993, as supplemented by. letter dated April 18, 1994, complies with the standards and requirements of the.
Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, 1e provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be l
conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will' not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
9405100281 940426 PDR ADOCK 05000352 P
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l~- i 2.
Accordingly, the license is amended by changes to the Technical i
Specifications as indicated in the attachment to this license amendment, and paragraph 2,C.(2) of Facility 0perating License..No. NPF-39 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 69
, are hereby incorporated into this license.
j Philadelphia Electric Company shall operate the facility in accordance j
with the Technical Specifications and the Environmental Protection Plan.
~
3.
This license amendment is effective as of its date of issuance.
I FOR THE NUCLEAR REGULATORY COMMISSION Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II i
Office of Nuclear Reactor Regulation I
Attachment:
}
Changes to the l
Technical Specifications Date of Issuance: April 26, 1994 4
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ATTACHMENT TO LICENSE AMENDMENT N0. 69 FACILITY OPERATING !.ICENSE NO. NPF-39 DOCKET NO. 50-352 Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.*
Remove Insert xvii xvii
- xviii xviii XIX XIX l
XX XX*
l 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10*
3/4 3-15 3/4 3-15*
3/4 3-16 3/4 3-16 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*
B 3/4 3-1 B 3/4 3-1*
B 3/4 3-2 B 3/4 3-2 B 3/4 3-3 B 3/4 3-3 B 3/4 3-4 B 3/4 3-4*
INDEX -
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADI0 ACTIVE EFFLUENTS (Continued)
Explosive Gas Mixture...................................
3/4 H-15 Main Condenser..........................................
3/4 U-16 The information on page 3/4 11-17 has been intentionally omitted.
Refer to note on th page........................................is 3/4 n-17 i
3/4.11.3 (Deleted) The information on pages 3/4 11-18 through t
3/4 n-20 has been intentionally omitted.
Refer to note on page 3/4 n-18.
t 3/4.u.4 (Deleted)...............................................
3/4 11-18 3/4.12 (Deleted) The information on pages 3/4 U-1 through 3/4 u-14 has been intentionally omitted.
Refer to note on page 3/4 U-1................
3/4 U-1 l
l LIMERICK - LMIT 1 xvii Amendment No. 48
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hshill y m ^j' f /99$
INDEX BASES SECTION PME 3/4.0 APPLICABILITY...................................................
B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN............................................
B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.......................................
B 3/4 1-1 3/4.1.3 CONTROL R0DS...............................................
B 3/4 1-2 3/4.1.4 CONTROL R00 PROGRAM CONTR0LS...............................
B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM..............................
B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.......................................................
B 3/4 2-1 3/4.2.2 (DELETED)..................................................
B 3/4 2-2 LEFT INTENTIONALLY BLANK..............................................
B 3/4 2-3 l
3/4.2.3 MINIMUM CRITICAL POWER RATI0...............................
B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE................................
B 3/4 2-5 3/4.3 INSTRUMENTATION
]
3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION..................
B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION........................
B 3/4 3-2 1
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION............................................
B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION..........
B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION............................................
B 3/4 3-4 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION..........................
B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.....................
B 3/4 3-5 LIMERICK - UNIT 1 xviii Amendment No. 7, 33, 66, h4
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INDEX l
BASES SECTION PEE JHSTRUMENTATION (Continued)
Seismic Monitoring Instrumentation.......................
B 3/4 3-5 3
(Deleted)................................................
B 3/4 3-5 l
Remote Shutdown System Instrumentation and Controls......
B 3/4-3-5 Accident Monitoring Instrumentation......................
B 3/4 3-5 Source Range Monitors....................................
B 3/4 3-5 Traversing In-Core Probe System..........................
B 3/4 3-6 l
Chlorine and Toxic Gas Detection Systems.................
B 3/4 3-6 Fire Detection Instrumentation...........................
B 3/4 3-6 Loose-Part Detection System..............................
B 3/4 3-7 (Deleted)................................................
B 3/4 3-7 Offgas Moni toring Instrumentation........................
B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM........................
B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................................
B/3/4 3-7 Bases figure B 3/4.3-1 Reactor Vessel Water Level.............................
B 3/4 3-B 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM.......................................
B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.......................................
B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................
B 3/4 4-3 Opera ti on al Le akag e......................................
B 3/4 4-3 3/4.4.4 CHEMISTRY..................................................
B 3/4 4-3 LIMERICK - UNIT 1 xix Amendment No. 48, 53,69 j
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INDEX BASES s
SECTION PAGE j
REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY.......................................
B 3/4 4-4 l
l 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.............................
B 3/4 4-4 i
Bases Table 8 3/4.4.6-1 Reactor Vessel Toughness.................
B 3/4 4-7 Bases Figure B 3/4.4.6-1 Fast Neutron Fluence i
3 (E>l Mov) At 1/4 T As A Function of Service Life......................
8 3/4 4-8 2
4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................
8 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY....................................
B 3/4 4-6 l
3/4.4.9 RESIDUAL HEAT REMOVAL..................................,
g 3/4 4 6 a
Jl j
3/4.5 EMERGENCY CORE COOLING SYSTEMS 2
3/4.5.1 and 3/4.5.2 ECCS - OPERATI M and SHUTDOWN............
8 3/4 5-1 1
i 3/4.5.3 SUPPRESSION CHAM ER................................
8 3/4 5-2 I
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAIMENT i
l Primary Containment Integrity......................
8 3/4 6-1 1
Primary Containment Leakage........................
B 3/4 6-1 s
Primary Containment Air Lock.......................
B 3/4 6-1 4
MSIV Leakage Control Systes........................
B 3/4 6-1 Primary Containment Structural Integrity...........
B 3/4 6-2 Drywell and Suppression Chamber Internal 3
Pressure.........................................
B 3/4 6-2 i
Drywell Average Air Temperature....................
8 3/4 6-2 Drywel l and Supp re s s i on Chambe r Purge Sys ten.......
8 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...........................
8 3/4 6-3 LIMERICK - UNIT 1 xx Amendment No. 33 OCT 3 01989 m
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INSTRUMENTATION 4
3/4.3.2.
ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2.-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.
APPLICABILITY: As shown in Table 3.3.2-1.
ACTION:
a)
With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
b)
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirements for one trip system:
1.
If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If this cannot be accomplished, the ACTION required by Table -
3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.
or 2.
If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within:
a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for trip functions common
- to RPS Instrumentation.
b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for trip functions not common
- to RPS Instrumentation.
The provisions of Specification 3.0.4 are not applicable.
Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.
LIMERICK - UNIT 1 3/4 3-9 Amendment No. 53, 69
INSTRUMENTATION LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
(Continued)
With the number of OPERABLE channels less than required by the Minimum c.
~
OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and take the ACTION required by Table 3.3.2-1.
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SURVEILLANCE REQUIREMENTS 1
4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.
4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
M 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown !
in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 -J' months. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
i
)
i The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.
i LIMERICK - UNIT 1 3/4 3-10 Amendment No. 53 Decur)DM I%/f9/
1 TABLE 3.3.2-1 g
ISOLA N (Continued)
INSTRtN NTATION E
n NIdINlM APPLICABLE ISOLATIgI'IcI OPERABLECHANNELgI OPERATIONAL TRIP FUNCTION SIGNAL PER TRIP SYSTEN CONDITION ACTION EQ 7.
SEColSARY CONTAIISWIT ISOLATION ss a.
Reacter Vessel Water Level Law. Low - Level 2 8
2 1,2,3 25 b.
Drywell Pressure - High H
2 1,2,3 25 c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiatten - High
.R 2
25 l
- 2. Refueling Area Unit 2 Ventilatten wi Exhaust Duc' Radiatten - Nigh R
2 25 l
s.,
U' d.
Reacter Enclosure Ventilatten Exhaust Ouct Radiatten - Nigh 5
2 1,2,3 25 e.
Outside Atmosphere Te Reacter i
Enclosure a Pressure - Law U
1 1,2,3 25 f.
Outside Atawphere To Refeeling Area 4 Pressure - Law T
1 25 g.
Reector Enclosure Manuel Initiatten MA 1
1,2,3 24
(-
h.
Refueling Area Manuel Initiatten IIA 1
25 4
y u
o gy
=
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS 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 ACTION 20 next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 21
- Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 /> 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 />.
i ACT10N 22
- Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves ACTION 23 l
are closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and declare the affected system inoperable.
In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 24
- P.estore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or close the affected system isolation valves within the next hour and declare l
the affected system inoperable or be.in at least HOT SHUTDOWN within the next I
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 />.
ACTION 25
- Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ACTION 26 Close the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
TABLE NOTATIONS Required when (1) handling irradiated fuel in the refueling area secondary o
containment, or (2) during CORE ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel-head removed and fuel in the vessel.
May be bypassed under administrative control, with all turbine stop valves closed.
- o l
During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
(a)
See Specification 3.6.3, Table 3.6.3-1 for primary containment isolation valves which are actuated by these isolation signals.
l (b)
A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.
In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillance without placing the channel or trip system in the tripped condition.
l LIMERICK - UNIT 1 3/4 3-16 Amendment No. 23. 40, 53, 69
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TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH e
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE y
1.
MAIN STEAM LINE ISOLATION h
a.
?!
1)
Low, Low, Level 2 S
Q R
1, 2, 3
[
2)
Low, Low, Low - Levei i S
Q R
1,2,3 b.
Main Steam Line Radiation ## - High S
Q R
1, 2, 3 l
L.
c.
Main Steam Line Pressure - Low S
Q R
1 l
d.
Q R
1, 2, 3 l
t' Flow - High Y
e.
Condenser Vacuum - Low 5
Q R
1, 2**, 3**
l 0
f.
Outboard MSIV Room 5
Q R
1,2,3 Temperature - High g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - Jiigh S
Q R
1, 2, 3 l
N.A.
R N.A.
- 1. 2, 3 g
h.
Manual Initiation
.t gg 2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION e
i a.
Reactor Vessel Water Levelff l
S Q
R 1, 2, 3 Low - Level 3 h$
b.
Reactor Vessel (RHR Cut-In S
Q R
1, 2, 3 h
Permissive) Pressure - High c.
Manual Initiation N.A.
R N.A.
1,2,3
.. ~
.. _. _ -. - _ -. -. ~ -
ISOLRTION ACTUATIGP! INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATf0NAL C
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH l
55 TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE
%3.
REACTOR WATER CLEANUP SYSTEM ISOLATION W
a.
RWCS a Flow - High S
Q R
1, 2, 3 l
I E
b.
RWCS Area Temperature - High S
Q R-1,2,3 l
M t
c.
RWCS Area Ventilation A Temperature - High S
Q R
1, 2, 3 l
d.
SLCS Initiation N.A.
R N.A.
1, 2, 3 e.
Q R
1, 2, 3 l
Low, Low, - Level 2 f.
Manual Initiation N.A.
R N.A.
1, 2, 3 t
" 4.
HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION i
U a Pressure - High S
Q R
1,2,3 l
Y a.
HPCI Steam Line b.
HPCI Steam Supply S
Q R
1,2,3 l
i Pressure, Low 4
c.
HPCI Turbine Exhaust Diaphragm Pressure - High S
Q R
1, 2, 3 l
[
d.
HPCI Equipment Room S
Q R
1, 2, 3 l
Temperature - High 3
m P,
e.
HPCI Equipment Room S
Q R
1, 2, 3 l
A Temperature - High g
f.
HPCI Pipe Routing Area S
Q R
1,2,3 l
w Temperature - High g.
Manual Initiation N.A.
R N.A.
1,2,3 i
i N.A.
Q R
1, 2, 3 l
l i
h.
HPCI Steam Line i
a Pressure Timer
-i i
TABLE 4.3.2.1-1 (Continued)
C.
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL.
R CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE 4A 5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION 4-a.
RCIC Steam Line l
Pressure - High S
Q R
1,2,3 1
A b.
RCIC Steam Supply Pressure - Low S
Q R
1,2,3 c.
RCIC Turbine Exhaust Diaphragm Pressure - High S
Q R
1,2,3 l
l d.
RCIC Equipment Room Q
Temperature - High 5
Q R
1,2,3 l
Ll[
e.
RCIC Equipment Room Temperature - High S
Q R
1,2,3 l
je A
f.
RCIC Pipe Routing Area Temperature - High S
-Q R
1,2,3 g.
Manual Initiation N.A.
R N.A.
1, 2, 3
.. i
!O h.
RCIC Steam Line
!k A Pressure Timer
.N.A.
Q
'R 1, 2, 3 lN
[
! E.F
,e i
i t-
TABLE 4.3.2.1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH e
j TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE 5j 6.
PRIMARY CONTAINMENT ISOLATION i
c:
a.
1)
Low, Low - Level 2 S
Q R
1, 2, 3 l
2)
Low, Low, Low - Level 1 S
Q R
1,2,3 b.
Drywell Pressureff - High 5
Q R
1, 2, 3 l
c.
North Stack Effluent Radiation - High S
Q R
1,2,3 d.
Deleted
- t' e.
Reactor Enclosure Ventilation Exhaust Duct - Radiation - High S
Q R
1,2,3 Y
E f.
Outside Atmosphere to Reactor Enclosure a Pressure - Low N.A.
H Q
1,2,3 g.
Deleted h.
Drywell Pressure - High/
(o Reactor Pressure - Low S
Q R
1,2,3 4
I k
i.
Primary Containment Instrument Gas to Drywell A Pressure - Low N.A.
M Q
1, 2, 3
, z P
N.A.
R N.A.
1, 2, 3
,a j.
Manual Initiation 61 e
TABLE 4.3.2.1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE
[ 7.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level Low, Low - Level 2 S
Q R
1, 2, 3 l
9 i
b.
Drywell Pressure ## - High S
Q R
1,2,3 l
E
. M c.1.
Refueling Area Unit i Ventilation Exhaust Duct Radiation - High 5
Q R
l t
2.
Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S
Q R
l d.
Reactor Enclosure Ventilation Exhaust Duct Radiation - High S
Q R
1, 2, 3 l
e.
Outside Atmosphere To Reactor E.
Enclosure A Pressure - Low N.A.
M Q
1, 2, 3 c
f.
Outside Atmosphere To Refueling Area A Pressure - Low N.A.
M Q
~
g.
Reactor Enclosure Manual Initiation N.A.
R N.A.
1, 2, 3 h.
Refueling Area k
Manual Initiation N.A.
R N.A.
u k
- Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE
?,
ALTERATIONS, or (3) during operations with a' potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
g:
- When not administratively bypassed and/or when any turbine stop valve is open.
g h
- During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
N. ##These trip functions (1b, 2a, 6b, and 7b) are common to the RPS actuation trip function.
l
!ws?auwtN7ait0N 3/4 33 EwERGENCY CORE COCLINC SYSTEw acTUA?!CN IN578UMEN?A7!0N I
i c:u!*!NG CON 0! TION FOR ODE 3ATION 4
i JJ3 No e=e gem:y :sre cooling system (ECCS) actsation tastrumentation e.ma-eis sne-n in isole 3.3.3 1 sna11 ce OPERA 8LE ita treir trio setooints j
set ::asistaat itn the values sno a in tse Trip Setooint colwan of facte 3.3 3 2 i
anc witn EMERGENCY CCRE C00 LING SYSTEM RESPONSE TIME as snown in Teole 3.3.3-3.
i i
A pet. !!A8 ! L ITY:
As snown in Table 3.3.3 1, 4
AC*!0N:
a.
witn an ECCS actuation instrumentation enannel trip setooint less conservative snan tne value snown in the A11ewsole values column of f
Teole 3.3.3-2. ceclare the channel ineseraele until the enannel is restorea to CPERA8LE status with its trip setooint acjusted consistent j
witn tne Trip 5etpoint value.
j i
i n.
with one or more ECCS actuation instrumentation enannels inoperaole, tame the ACTICN recuireo ey Table 3.3.3 1.
c.
Wita eitner A05 trip system suosystee inoperaale, restore the 4
5l inoperaole trip system to CPERA4LE status witnin:
/
4 1.
7 days, provided that the NPCI and RCIC systems are OPERA 8LE.
t 2.
72 neurs.
Otnerwise. De in at least M0T SMUT 00WN within the nest 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam come pressure te less than er eeuel te l
100 psig within the following 24 neurs.
I j
SURVE!LLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuatten instrumentatten channel shall be deoenstrated OPERA 8LE by the perfensance of the CHANNEL CHECX, CHANNEL FUNCTIONAL ftST and j
CHANNEL CAL 18AAT!0N emerations for the OPERATIONAL CON 0! TION $ and at the j
frequencies shous in Tatte 4.3.3.1 1.
4 3.3.2 LOGIC 5957tM FUNCTIONAL TESTS and simulated auteestic operation of
.all channels shall be perfereed at least once per 18 monthe.
4.3.3.3 The ICCS RESPONSE TINE of each ICCS trip function shown in Table 3.3.3 3 j
sna11 te desenstrated to to within the 11eit at least once per 18 months. Each test shall include at least one channel per trip systes susn that all channels are tested at least once every N times 18 months where N is the total numeer of redundant channels in a specific (CCS trip systas.-
i 4
LIMERICK - UNIT 1 3/A 3*38 e
O m.-,
..m.,,-.-..._,,_y-g y
4*
g,
9* gM e es
3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
Preserve the integrity of-the fuel cladding.
a.
b.
Preserve the integrity of the reactor coolant system.
Minimize the energy which must be absorbed following a c.
loss-of-coolant accident, and d.
Prevent inadvertent criticality.-
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable.
for brief intervals to conduct required surveillance.
The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined j
in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intr.nt j
of IEEE-279 for nuclear power plant protection systems. Specified s
surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specificatica i
Improvement Analyses for BWR Reactor Protection System,"-as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) (letter to T. A.
i Pickens from A. Thadani dated July 15, 1987.. The bases for the trip settings of RPS are discussed in the bases for Specification 2.2.1.
i i
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
1 l
l t
LIMERICK - UNIT 1 B 3/4 3-1 Amendment No. 53 p' %@unbtL / 'd / f4/
1NSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to l
mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and l
ECCS Instrumentation," as approved by the NRC and documented in the NRC Safety l
Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6, 1989) and NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," as approved by the NRC and documented in the NRC SER (letter to S. D. Floyd from C. E. Rossi dated June 18,1990).
Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high of low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
Except for the MSIVs, the safety analysis does not address individual sensor l
response times or the response times of the logic systems to which the sensors l
are connected.
For D. C. operated valves, a 3 second delay is assumed before the I
valve starts to move.
For A.C. operated valves, it is assumed that the A.C.
l power supply is lost and is restored by startup of the emergency diesel I
generators.
In this event, a time of 13 seconds is assumed before the valve starts to move.
In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay.
It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
1 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION j
1 The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the l
l ability of the operator to control.
This specification provides the OPERABILITY l
requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
l LIMERICK - UNIT 1 B 3/4 3-2 Amendment No. 33, 33, 69 l
INSTRUMENTATION BASES f
3/6.3.3 EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION (Continued)
Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER i
(letter to D. N. Grace from A. C. Thadani dated December 9,1988 (Part 1) and letter to D. N. Grace from C. E. Rossi dated De:: ember 9,1988 (Part 2)).
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip sysam provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.
The response of the i
plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.
The end-of-cycle recirculation pump trip (E0C-RPT) system is a supplement to the reactor trip.
During turbine trip and generator load rejection events, the E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level 2.
Each E0C-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events.
The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system.
For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps.
Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled.
The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.
The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,
175 ms.
Included in this time are:
the response time of the sensor, the time
-allotted for breaker arc suppression, and the response time of the system logic.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
LIMERICK - UNIT 1 B 3/4 3-3 Amendment No. 53, 69
l j
INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to dssure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel.
This instrumentation does not provide actuation of any of l
the emergency core cooling equipment.
Specified surveillance intervals and maintenance outage times c. ave been specified in accordance with recoinnendations made by GE in their letter to the BWR Owner's Group dated August 7, 1989
SUBJECT:
" Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is a r & so that a trip in any one of the inputs will result in a control rod block.
Specified surveillance intervals and maintenance outage times have been determined in accordance with MEDC-30851P, Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation "
as approved by the NRC and documented in the SER (letter to D. N. Grace from C.
E. Rossi dated September 22,1988).
j Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures thatt (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A General Design Criteria 19, 41, 60, 61, 63, and 64.
LIMERICK - UNIT 1 B 3/4 3-4 Amendment No. 48,53 W~ Dl'Cb17?t% /7, /99/
fa accoq
.fm E 'I Df )* !
S UNITED STATES 3$
NUCLEAR REGULATORY COMMISSION k..v
/
WASHINGTON, D.C. 2055A001 PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 32 License No. NPF-85 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the
~
licensee) dated April 19, 1993, as supplemented by letter dated April 18, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and
)
safety of the public, and (ii) that such activities will be i
conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:
i Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 32
, are hereby incorporated into this license.
Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 26, 1994 l
I*-
ATTACHMENT TO LICENSE AMENDMENT NO. 32 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 i
Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and j
contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.*
Remove Insert xvii xvii *-
xviii xviii xix xix xx xx*
3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10*
3 l
3/4 3-15 3/4 3-15*
3/4 3-16 3/4 3-16 3/4 3-27 3/4 3-27
{
3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 1
3/4 3-32 3/4 3-32*
l 1
B 3/4 3-1 B 3/4 3-1*
i B 3/4 3-2 B 3/4 3-2 1
2 B 3/4 3-3 B 3/4 3-3
]
B 3/4 3-4 8 3/4 3-4*
i 4
i I
i
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREME SECTION PAGE RADIOACTIVE EFFLUENTS (Continued) k Explosive Gas Mixture...................................
3/4 11-15 4
Main Condenser..........................................
3/4 u-16 The information on page 3/4 11-17 has been i
intentionally ositted.
Refer to page............................. note on this I
3/4 11-17 3/4.11.3 (Deleted) The information on pages 3/4 11-18 through 3/4 11-20 has been intentionally omitted.
i Refer to note on page 3/4 11-18.
j 3/4.'11.4 (Deleted)...............................................
3/4 11-18 3/4.12 j
(Deleted) The information on pages 3/4 12-1 through 3/4 12-14 has been intentionally ositted.
.i Refer to note on page 3/4 12-1................
3/4 12-1 o
a i
l l
l i
l J
i.
E I
i I
4 LIMERICK - UNIT 2 xvii Amendment No. 4 N! Ell
& /ff/
INDEX BASES f_%E SECTION 3/4.0 APPLICABILITY......................................................
B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN...............................................
E 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES..........................................
B 3/4 1-1 J
3/4.1.3 CONTROL R0DS..................................................
B 3/4 1-2 3/4.1.4 CONTROL R0D PROGRAM CONTR0LS..................................
B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.................................
B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..........................................................
B 3/4 2-1 3/4.2.2 APRM SETP0lNTS................................................
B 3/4 2-2 LEFT INTENTIONALLY BLANK..................................................
B 3/4 2-3 l
3/4.2.3 MINIMUM CRITICAL POWER RATI0..................................
B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE...................................
B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.....................
B 3/4 3-1 3/4.3.2 IS07% TION ACTUATION INSTRUMENTATION...........................
B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...............................................
B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.............
B 3/4 3-3 l
3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...............................................
B 3/4 3-4 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION.............................
B 3/4 3-4 l
l 3/4.3.7 MONITORING INSTRUMENTATION l
Radiation Monitoring Instrumentation..........................
B 3/4 3-5 l
l LIMERICK - UNIT 2 xvili Amendment No. 4,32
BASES SECTION PAGE INSTRUMENTATION (Continued)
Seismic Monitoring Instrumentation............................
B 3/4 3-5 (Deleted).....................................................
B 3/4 3-5 Remote Shutdown System Instrumentation and Controls...........
B 3/4 3-5 Accident Monitoring Instrumentation...........................
B 3/4 3-5 Source Range Monitors.........................................
B 3/4 3-5 Traversing In-Core Probe System...............................
~ B 3/4 3-6 l-Chlorine and Toxic Gas Detection Systems......................
B 3/4 3-6 Fire Detection Instrumentation................................
B 3/4 3-6 i
l Loose-Part Detection System..................................
B 3/4 3-7 (Deleted).....................................................
B 3/4'3-7 Offgas Monitoring Instrumentation.............................
B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM...........................
-B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION...............................................
B 3/4 3-7 Bases Figure B 3/4.3-1 Reactor Vessel Water
+
~
Level...........................
B.3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..........................................
B 3/4 4-1 1
3/4.4.2 SAFETY / RELIEF VALVES..........................................
B 3/4 4-2
)
3/4.4.3 REACTOR COOLANT SYSTEM LcAKAGE Leakage Detection Systems....................................
B 3/4 4-3 Op e ra t i on al Le a k ag e..........................................
B 3/4 4-3 3/4.4.4 CHEMISTRY....................................................
B 3/4 4-3a LIMERICK - UNIT 2 xix Amendment No.
- 11. 12. 17. 32
1 l
IEEX BASES f
SECTION g
]
REACTOR COOLANT SYSTEM (Continued)
]
3/4.4.5 SPECIFIC ACTIVITY.......................................
s 3/4 4-4 i
3/4.4.6 PRES $URE/ TEMPERATURE LIMIT 5.............................
8 3/4 4-4 Bases Table 8 3/4.4.6-1 Reactor ' Vessel Toughness.................
8 3/4 4-7 i
Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) At 1/4 T As A Function of Service 1
Life......................
B 3/4 4-8 i
J
.j 3/4.4.7 MAINSTEAMLINEISOLATIONVAi.VES........................
B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY....................................
8 3/4 4-6 1
3/4.4.9 RESIDUAL HEAT RB GVAL...................................
8 3/4 4-6
]
3/4.5 EMERGENCY CORE C06 LING SYSTEMS I
3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTD0 W............
B 3/4 5-1 a
3/4.5.3 SUPPRESSION CHMSER................................
8 3/4 5-2 3/4.6 CONTAI MENT SYSTEMS 3/4.6.1 PRIMARY CONTAI N.NT Primary Containment Integrity......................
8 3/4 6-1 l
Primary Containment Leakage........................
8 3/4 6-1 Primary Containment Ai r Lock.............'..........
B 3/4 6-1 Y
i i
MSIV Leakage Control Systen........................
B 3/4 6-1 Primary Contai nment Structural Integri ty...........
S 3/4 6-2 Drywell and Suppression Chamber Internal Pressure.........,...............................
B 3/4 6-2 i
l
}
Drywell Average Ai r Temperature....................
B 3/4 6-2 1
Drywell and Suppression Chamber Purge System.......
8 3/4 6-2
{
3/4.6.2 DEPRESSURIZATION SY5TEMS...........................
B 3/4 6-3 1
~
' mentcg - UNIT 2 xx I
INSTRUMENTATION 3/4.3.2.
ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2.-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table i
3.3.2-3, 1
(
l l
APPLICABILITY: As shown in Table 3.3.2-1.
l ACTION:
a)
With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value, b)
With the number of OPERABLE channels less than required by the Minimum OPERABLE l
Channels per Trip System requirements for one trip system:
1.
If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If this cannot be accomplished, the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.
or 1
2.
If placing the inoperable channel (s) in the tripped condition would not cause !
an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within:
a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for trip functions common
- to RPS Instrumentation, b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for trip functions not common
- to RPS Instrumentation.
The provisions of Specification 3.0.4 are not applicable.
l i
Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.
LIMERICK - UNIT 2 3/4 3-9 Amendment No.17,32
J NSTRUMENTAT TON LIMITING CONDITION FOR OPERATION (Continued)
~
ACTION:
(Continued) 1 l
With the number of OPERABLE channels less than required by the Minimum c.
OPERABLE Channels per Trip System requirement for both trip systems,
_ place at least one trip system ** in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.-
and take the ACTION required by Table 3.3.2-1.
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be denenstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.
4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of al1~
channels shall be performed at least once per 18 months.
i 4.3.2.3 The ISOLATION' SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 i
months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.
LIMERICK - UNIT 2 3/4 3-10 Amendment No. 17 DGAM1 W I?, I f 9 7
TABLE 3.3.2-1 (Continued)
,_g ISOLATTUiFXCTUATION INSTRIM NTATION E
p; NINIfRM APPLICABLE ISOLATIgI'ICI OPERABLECHANNELgI OPERATIONAL x
TRIP FUNCTION SIGNAL PER TRIP SYSTEN C00mITION ACTION c
7.
SEC0fSARY CONTAll0ENT ISOLATION a.
Reactor Vessel Water Level Low, Low - Level 2 8
2 1,2,3 25 b.
Drywell Pressure - High H
2 1,2,3 25 c.1. Refueling Area Unit 1 Ventilation Exhaust Doct Radiation - High R
2 25
- 2. Refueling Area Unit 2 Ventilation g
Exhaust Duct Radiation - High R
2 25 ia d.
Reactor Enclosure Ventilation Exhaust g
Duct Radiation - High 5
2 1,2,3 25 e.
Outside Atmosphere To Reactor Encidsure a Pressure.- Low U
1 1,2,3 25 f.
Outside Atmosphere To Refueling Area A Pressure - Low T
1 25 g.
Reactor Enclosure Manual Initiation MA 1
1,2,3 24 h.
Refueling Area Manual Initiation NA 1
25 t
J
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 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 SHUIDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 21 Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least H0T 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 />.
ACTION 22 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 23 In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within I hour and declare the affected system inoperable.
In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOW4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 24 Restore the manual initiation ' unction to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least H0T 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 />.
ACTION 25 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within I hour.
ACT10N 26 Close the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
TABLE NOTATIONS Required when (1) handling irradiated fuel in the refueling area secondary o
containment, or (2) during CORE ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
May be bypassed under administrative control, with all turbine stop valves closed.
During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
(a)
See Specification 3.6.3, Table 3.6.3-1 for primary containment isolation valves which are actuated by these isolation signals.
(b)
A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.
In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for tnat valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for required surveillance without placing the channel or trip system in the tripped condition.
LIMERICK - UNIT 2 3/4 3-16 Amendment No. 17,32
~
I
TABLE 4.3.2.1-1 g
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS
.. g; CHANNEL OPERATIONAL Q
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE l
1.
, MAIN STEAM LINE ISOLATION a.
Low, Low, Level 2 5
Q R
1,2,3 2)
Low, Low, Low - Level 1 5
Q R
1,2,3 b.
Main Steam Line Radiation ## - High S
Q R
1,2,3 l
l c.
Main Steam Line Pressure - Low S
Q R
1 l
d.
Main Steam Line Flow - High S
Q R
1,2,3 l
Y e.
Condenser Vacuum - Low S
Q R
1, 2**, 3**
l f.
Outboard MSIV Room Temperature - High S,
Q R
1, 2, 3 l
g.
Turbine Enclosure - Main Steam-
^
1,2,3 l
Line Tunnel Temperature - High S
Q R
h.
Manual Initiation N.A.
R N.A.
- 1. 2, 3 ag 2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION e
a.
l l
Low - Level'3 S
Q R
1, 2, 3 i
=
b.
-Reactor Vessel (RHR Cut-In S
Q R
1,2,3 U
Permissive) Pressure - High c.
Manual Initiation.
N.A.
R N.A.
1, 2, 3
TABLE 4.3.2.1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH r
y TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE 9 3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
RWCS a Flow - High S
Q R
1,2,3 l
l y
[
b.
RWCS Area Temperature - High S
Q R
1,2,3 l
l 5
A c.
RWC1 Area Ventilation a Temperature - High S
Q R
1, 2, 3 l
d.
SLCS Initiation N.A.
R N.A.
1, 2, 3 e.
Reactor Vessel Water Level Low, Low, - Level 2 5
Q R
1, 2, 3 l
f.
Manual Initiation N.A.
R N.A.
1, 2, 3 4.
HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION a.
HPCI Steam Line a Pressure - High S
Q R
1,2,3 l
w5 b.
HPCI Steam Supply w
h Pressure, Low S
Q R
1, 2, 3 l
c.
HPCI Turbine Exhaust Diaphragm Pressure - High 5
Q R
1, 2, 3 l
d.
HPCI Equipment Room Temperature - High S
Q R
1, 2, 3 l
h e.
HPCI Equipment Room g
a Temperature - High S
Q R
1,2,3 l
r
(&
f.
HPCI Pipe Routing Area Temperature - High S
Q R
1, 2, 3 l
0 L
g.
Manual Initiation N.A.
R N.A.
1, 2, 3 m
h.
HPCI Steam Line a Pressure Timer N.A.
Q R
1, 2, 3 l
l
TABLE 4.3.2.1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS C
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH A TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE i
E 5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION 4
w a.
RCIC Steam Line a Pressure - High S
Q R
1,2,3 l
b.
RCIC Steam Supply Pressure - Low S
Q R
1, 2, 3 c.
RCIC Turbine Exhaust Diaphragm Pressure - High S
Q R
1,2,3 l
I w
S d.
RCIC Equipment Room Y
Temperature - High S
Q R
1,2,3 l
~
M i
e.
RCIC Equipment Room a Temperature - High S
Q R
1,2,3 l
j 1
f.
RCIC Pipe Routing Area Temperature - High S
Q R
1, 2, 3 l
1 g.
Manual Initiation N.A.
R N.A.
1,2,3 E
h.
RCIC Steam Line k
a Pressure Timer N.A.
Q R
1, 2, 3 l
E3g i
o h
1 4
1 i
B*
e
._m._
m.
.___.__._m.m..
m.
TABLE 4.3.2.1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS C
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CHECK TEST CALIBRATION SURVEILLANCE REQUIRE E TRIP FUNCTION I
g 6.
PRIMARY CONTAINMENT ISOLATION S
Q R
1, 2, 3 l
w 1)
Low, Low - Level 2 2)
Low, Low, Low - Level 1 S
Q R
1,2,3 5
Q R
1,2,3 l
b.
Drywell Pressure ## - High l
c.
North Stack Effluent S
Q R
1,2,3 Radiation - High d.
Deleted
,g Reactor Enclosure Ventilation 6
Exhaust Duct - Radiation - High S
Q R
1,2,3 e.
a o
f.
Outside Atmosphere to Reactor N.A.
M Q
1,2,3 Enclosure A Pressure - Low i
g.
Deleted 4
h.
Drywell Pressure - High/
S
-Q R
1,2,3 lyg Reactor Pressure - Low
- a3 i.
Primary Containment Instrument N.A.
M Q
1,2,3 Gas to Drywell a Pressure - Low i "
5 N.A.
R N.A.
1,2,3 j
{
j.
Manual Initiation i
d i
TABLE 4.3.2.1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL
[
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH gj CHECK TEST CALIBRATION SURVEILLANCE REQUIRE
$ TRIP FUNCTION R
7.
SECONDARY CONTAINMENT ISOLATION i
a.
Q R
1, 2, 3 g
Low, Low - Level 2 b.
Drywell Pressure #f - High 5
Q R
1,2,3 l
4 m
c.1.
Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High 5
Q R
- p l
2.
Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S
Q R
l d.
Reactor Enclosure Ventilation 2
Exhaust Duct Radiation - High S
Q R
1,2,3 l
w Y
w e.
Outside Atmosphere To Reactor Enclosure A Pressure - Low N.A.
M Q
1,2,3 f.
Outside Atmosphere To Refueling N.A.
M Q
Area a Pressure - Low g.
Reactor Enclosure N.A.
R N.A.
1, 2, 3 Manual Initiation h.
Refueling Area x
N.A.
R N.A.
8 Manual Initiation A
- Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE 8
ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel o
g head removed and fuel in the vessel.
- When not administratively bypassed and/or when any turbine stop valve is open.
O
'u
- During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
- These trip functions (1b, 2a, 6b, and 7b) are common to the RPS actuation trip function.
INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATION LINITING CONDITION FOR OPERATION l
3.3.3 The emergency core cooling system (ECCS) actuation instrumentation j
channels shown in Table 3.3.3-1 sha.11 be OPERA 8LE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with ENERGENCY CORE COOLING SYSTEM RESPONSE TINE as shown in Table'3.3.3-3.
APPLICABILITY:
As shown in Table 3.3.3-1.
ACTION:
a.
With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERA 8LE status with its trip setpoint adjusted consistent j
with the Trip Setpoint value.
b.
With one or more ECCS actuation instrumentation channels inoperable.
take the ACTION required by Table 3.3.3-1.
c.-
With either ADS trip system subsystem inoperable, restor's the
, inoperable trip system to OPERA 8LE status within:
1.
7 days, provided that the HPCI and RCIC systems'are OPERA 8LE.
2.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
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 /> and reduce reactor steam done pressure to less than or equal to 100 psig 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 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations.for the OPERATIONAL CONDITIONS and at the i
i frequencies shown in Table 4.3.3.1-1.
4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ECCS trip system.
LIMERICK - UNIT 2 3/4 3-32
3/4.3 [NSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
a.
Preserve the integrity of the fuel cladding.
b.
Preserve the integrity of the reactor coolant system.
Minimize the energy which must be absorbed following a c.
loss-of-coolant, accident, and d.
Prevent inadvertent criticality.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
i The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," as approved by the NRC and documented'in the NRC Safety Evaluation Report (SER) (letter to T. A.
Pickens from A. Thadani dated July 15, 1987.
The bases for the trip settings of RPS are discussed in the bases for Specification 2.2.1.
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests desonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
LIMERICK - UNIT 2 B 3/4 3-1 Amendment No.17 Q' blCDnt% /q /Yfl
INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip When setpoints and response times for isolation of the reactor systems.
necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Specified surveillance intervals and maintenance outage times have been a
l determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and ECCS Instrumentation," as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6, 1989) and NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," as approved by the NRC and documented in the NRC SER (letter to S. D. Floyd from C. E. Rossi dated June 18,1990).
3 Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high of low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors i
are connected.
For D. C. operated valves, a 3 second delay is assumed before the l
valve starts to move.
For A.C. operated valves, it is assumed that the A.C.
power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 13 seconds i^s assumed before the valve In addition to the pipe break, the failure of the D.C. operated starts to move.
valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay.
The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay.
It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.
l Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.
This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
I LIMERICK - UNIT 2 B 3/4 3-2 Amendment No.17. 32
d INSTRUMENTATION
' BASES 3/4.3.3 EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION (Continued)
Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D.
N. Grace from A. C. Thadani dated December 9, 1988 (Part 1) and letter to D. N.
Grace from C. E. Rossi dated December 9, 1988 (Part 2)).
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.
system is a supplement to The end-of-cycle recirculation pump trip (E0C-RPT)d rejection events, the the reactor trip.
During turbine trip and generator loa E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level 2.
Each E0C-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
i i
A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system.
Similarly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system.
For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves.
The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps.
4 Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.
The E0C-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,
175 ms.
Included in this time are:
the response time of the sensor, the time allotted for breaker arc suppression, and the response time of the system logic.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
LIMERICK - UNIT 2 B 3/4 3-3 Amendment No.17,32
s i
{
BASES a
3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is i
provided to initiate actions to assure adequate core cooling in the event of j
reactor isolation from its primary heat sink and the loss of feedwater flow to i
the reactor vessel.
This instrumentation does not provide actuation of any of f
the emergency core cooling equipment.
Specified surveillance intervals and maintenance outage times have been a
specified in accordance with recomendations made by GE in their letter to the BWR Owner's Group dated August 7,~1989
SUBJECT:
" Clarification of Tech in cal j
Specification changes aiven in ECCS Actuation Instrumentation Analysis."
Operation with a 'cri;, set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the i
difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety i
analyses.
4 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION 4
The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 i
Instrumentation. The trip logic is arranged so that a trip in any one of the e
inputs will result in a control rod block.
Specified surveillance intervals and maintenance outage times have been determined in accor. dance with NEDC-30851P, Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation "
as approved by the NRC and documented in the SER (letter to D. N. Grace from C.
E. Rossi dated September 22,1988).
t Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety i
i analyses.
3/4.3.7 MONITORING INSTRUMENTATION i
3/4.3.7.1 RADIATION MONITORIhG INSTRUMENTATION i
The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A General Design Criteria 19,~41, 60, 61, 63, and 64.
d i
LIMERICK - UNIT 2 B 3/4 3-4 Amendment No. II,17 hte.unkM I7: Mh m-
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