ML20198E010
| ML20198E010 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 12/14/1998 |
| From: | Capra R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198E017 | List: |
| References | |
| NUDOCS 9812230313 | |
| Download: ML20198E010 (22) | |
Text
._. - _..
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g UNITED STATES s
EU E
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20655 4001 PECO ENERGY COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.132 License No. NPF-39 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by PECO Energy Company (the licensee) dated August 8,1996, as supplemented June 30,1997 and August 26,1998, 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activitics will be 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.
9812230313 981214 PDR ADOCK 05000352 P
- 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-39 is hereby amended to read as follows:
Technical Soecifications i
1 The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.132, are hereby t
incorporated into this license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Prctection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION SN G.
Robert A. Capra, Director Project Directorate 1-2 Division of Reactor Projects -!/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 14, 1998 l
i
l ATTACHMENT TO LICENSE AMENDMENT NO.112 FACILITY OPERATING LICENSE NO. NPF-39
{
DOCKET NO. 50-352 Replace the folkmng 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.
Remove Insert 3/434 3/4 3-6 1
3/4 3-23 3/4 3-23 1
3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3'4 3-39 3/4 3-39 B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 l
l l
I l
l 4
i l
\\
l
TABLE 3.3.1-2 r-
- E REACTOR PROTECTION SYSTEM RESPONSE TIMES 9
RESPONSE TIME M
FUNCTIONAL UNIT (Seconds) n l.
E a.
Neutron Flux - High N.A.
b.
Inoperative N.A.
~
2.
Average Power Range Monitor *:
a.
Neutron Flux - Upscale, Setdown N.A.
b.
Neutron Flux - Upscale 1)
Flow Biased s0.09 2)
High Flow Clamped s0.09 c.
Inoperative N.A.
d.
Downscale N.A.
3.
Reactor Vessel Steam Dome Pressure - High s0.55 4.
Reactor Vessel Water Level - Low, Level 3 sl.05#
l
{
5.
Main Steam Line Isolation Valve - Closure s0.06 i'
6.
DELETED DELETED e
7.
Drywell Pressure - High N.A.
8.
Scram Discharge Volume Water Level - High a.
Level Transmitter N.A.
b.
Float Switch N.A.
9.
Turbine Stop Valve - Closure s0.06 10.
Turbine Control Valve Fast Closure, Trip 011 Pressure - Low s0.08**
11.
Reactor Mode Switch Shutdown Position N.A.
z
?
12.
Manual Scram N.A.
L
- Neutron detectors are exempt from response time testing. Response time shall be measured M
from the detector output or from the input of the first electronic component in the channel.
- Measured from start of turbine control valve fast closure.
- Sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.
fv TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP _ FUNCTION RESPONSE TIME (Seconds)#
1.
MAIN STEAM LINE ISOLATION a.
Low, Low - Level 2 N.A.
i 2)
Low, Low, Low - Level 1 s1.0###*
b.
DELETED DELETED c.
Main Steam Line Pressure - Low
$1.0###*
l d.
Main Steam Line Flow - High
$0.5###*
l e.
Condenser Vacuum - Low N.A.
f.
Outboard MSIV Room Temperature - High N.A.
l g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
h.
Manual Initiation N.A.
2.
RHR SYSTEM SHUTOOWN COOLING MODE ISOLATION a.
Reactor Vessel Water Level Low - Level 3 N.A.
l b.
Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.
c.
Manual Initiation N.A.
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
RWCS a Flow - High N.A.##
l b.
RWCS Area Temperature - High N.A.
c.
RWCS Area Ventilation A Temperature - High N.A.
d.
SLCS Initiation N.A.
e.
Low, Low - Level 2 N.A.
l f.
Manual Initiation N.A.
LIMERICK - UNIT 1 3/4 3-23 Amendment No. 29,89,132
o TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
4.
HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION a.
HPCI Steam Line Pressure - High N.A.
l A
b.
HPCI Steam Supply Pressure - Low N.A.
l c.
HPCI Turbine Exhaust Diaphragm Pressure - High N.A.
d.
HPCI Equipment Room Temperature - High N.A.
e.
HPCI Equipment Room a Temperature - High N.A.
f.
HPCI Pipe Routing Area Temperature - High N.A.
g.
Manual Initiation N.A.
5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.
RCIC Steam Line a Pressure - High N.A.
l b.
RCIC Steam Supply Pressure - Low N.A.
l c.
RCIC Turbine Exhaust Diaphragm Pressure - High N.A.
d.
RCIC Equipment Room Temperature - High N.A.
e.
RCIC Equipment Room A Temperature - High N.A.
f.
RCIC Pipe Routin Area Temperature - Hi h N.A.
g.
Manual Initiation N.A.
l l
l LIMERICK - UNIT 1 3/4 3-24 Amendment R). 3},132
TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
6.
PRIMARY CONTAINMENT ISOLATION a.
Low, Low - Level 2 N.A.
2)
Low, Low, Low - Level 1 N.A.
b.
Drywell Pressure - High N.A.
l c.
North Stack Effluent Radiation - High N.A.
d.
Deleted e.
Reactor Enclosure Ventilation Exhaust Duct - Radiation - High N.A.
f.
Deleted g.
Deleted h.
Drywell Pressure - High/
Reactor Pressure - Low N.A.
i.
Primary Containment Instrument Gas to Drywell a Pressure - Low N.A.
j.
Manual Initiation N.A.
7.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level Low,. Low - Level 2 N.A.
b.
Drywell Pressure - High N.A.
c.l.
Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High N.A.
2.
Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High N.A.
d.
Reactor Enclosure Ventilation Exhaust Dutt Radiation - High N.A.
e.
Deleted i
LIMERICK - UNIT 1 3/4 3-25 Amendment No. T5,412,132
TABLE 3.3.2-3 (Continued) l ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME l
TRIP FUNCTION RESPONSE TIME (Seconds)#
f.
Deleted g.
Reactor Enclosure Manual Initiation N.A.
i h.
Refueling Area Manual Initiation N.A.
TABLE NOTATIONS (a)
DELETED l
(b)
DELETED Isolation system instrumentation response time for MSIV only.
No diesel generator delays assumed for MSIVs.
DELETED l
Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
1 With 45 second time delay.
Sensor is eliminated from response time testing for the MSIV actuation logic circuits.
Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.
l l
LIMERICK - UNIT 1 3/4 3-26 Amendment No. 6,89,442,132 l
a l
TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES i
1 E
RESPONSE TIME (Seconds) 1.
CORE SPRAY SYSTEM s 27#
l 2.
LOW PRESSURE COOLANT INJECTION MODE 0F RHR SYSTEM l
s 40#
l 3.
AUTOMATIC DEPRESSURIZATION SYSTEM N.A.
4.
HIGH PRESSURE COOLANT INJECTION SYSTEM s 60#
l 5.
LOSS OF POWER N.A.
l l
- ECCS actuation instrumentation is eliminated from response time testing.
l LIMERICK - UNIT 1 3/4 3-39 Anendment No. 492,132
3/4.3 INSTRUMENTATION 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 adsorbed 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 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 Resort (SER) (letter to T.A.
Pickens from A. Thadani dated July 15, 1987. The )ases for the trip settings of RPS are discussed in the bases for Specification 2.2.1.
Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15,1991).
The measuement 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.
Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NED0-32291-A.
Response time testing for the remaining channel components is required as noted.
l LIMERICK - UNIT 1 B 3/4 3-1 Amendment No. 53,89,132
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 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 SpecificationImprovementAnalysisforBWkInstrumentationCommontoRPSand ECCS Instrumentation" as approved by the NRC and documented in the NRC Safet Evaluation Report (SER) (letter to D.N. Grace from C.E. Rossi dated January 1989 and NEDC-31677P-A, " Technical S Isola)tionActuationInstrumentation"pecificationImprovementAnalysisforBk the NRC SER (letter to S.D. Floyd fr,om C.E. proved by the NRC and documented in as ap Rossi dated June 18,1990).
Automatic closure of the MSIVs upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring, System was removed as the result of an analysis performed by General Electric in NED0-31400A.
The NRC approved the results of this analysis as documented in the SER letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated (May15, 1991).
Some of the trip setting,s 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 or low end of the setting have a direct bearing on safety,inadvertentare established at a level away from the normal operating range to prevent actuation of the systems involved.
Except for the MSIVs the safety analysis does not address individual sensor response times or the respon,se times of the logic systems to which the sensors are connected.
For D.C. operated valves, a 3 second delay is assumed before the valve starts to move.
For A.C. operated valves, it is assumed that the A.C.
power suppl generators.y is lost and is restored by startup of the emergency dieselIn this event, a tim 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) loading delay.
is concurrent with the 10-second diesel startup and the 3 second load center analysis considers an al'lowable inventory loss in each case which in turn The safety 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.
Response time testing for sensors are not required based on the analysis in HED0-32291-A.
Response time testin required as noted in Table 3.3.2-3.g of the remaining channel components is 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 orovided to initiate actions to mitigate the conse ability of the operator to control. quences of accidents that are beyond the l
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.
LIMERICK - UNIT 1 8 3/4 3-2 Amendment No. 33,53,69,89,132
l 1.C KlO
[
UNITED STATES s
g NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20555-0001
[
PECO ENERGY COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AfdENDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 License No. NPF-85 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by PECO Energy Company (the licensee) dated August 8,1996, as supplemented June 30,1997 and August 26,1998, 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurcnce (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 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 an,d all applicable requirements have been satisfied.
W
.- l 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:
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 93
, are hereby incorporated in the license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Pl:a.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION U d- [ W Robert A. Capra, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: December 14, 1998 l
r
._ _ _ _._._. ___ ~__
j ATTACHMENT TO LICENSE AMENDMENT NO 93 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-351 l
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.
Remove -
Ln,gg I
3/4 3-6 3/4 3-6 3/4 3-23 3/4323 3/4324 3/4 3-24 3/4 3-25 3/4325 3/4 3-26 3/4 3-26 3/4 3-3g 3/4 3-39 B W4 >1 B W4 Si B 3/4 3-2 B 3/4 3-2 l
l
\\
l I
~
TABLE 3.3.1-2
.E-REACTOR PROTECTION SYSTEM RESPONSE TIMES y
RESPONSE TIME
~
p;
' FUNCTIONAL UNIT (Seconds) x-l.
g a.
Neutron Flux - High N. A.~
q b.
Inoperative N. A.-
2.
Average Power Range Monitor *:
a.
Neutron Flux - Upscale, Setd.ivn N.A.
b.
Neutron Flux - Upscale i
1)
Flow Biased-s0.09 t
2)
High Flow Clamped s0.09 c.
Ir,9 perative N.A.
d.
Downscale N.A.
3.
Reactor vessel Steam Dome Pressure - High so.55 4.
Reactor Vessel Water Level - Low, Level 3 sl.05#
- l !
{
5.
Main Steam Line Isolation Valve - Closure s0.06 y
6.
DELETED DELETED e
7.
Drywell Pressure - High N.A.
8.
Scram Discharge Volume Water Level - High a.
Level Transmitter N.A.
8 b.
Float Switch N.A.
l jii 9.
Turbine Stop Valve - Closure so.06 g;
10.
Turbine Control Valve Fast Closure, i
g Trip 011 Pressure - Low so.08**
w I
11.
Reactor Mode Switch Shutdown Position N.A.
12.
Manual Scram N ?..
- Neutron detectors are exempt from response time testing.
Response time shall be measured from the detector output or from the input of the first electronic component in the channel.
- Measured from start of turbine control valve fast closure.
- Sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit i
tnd relay logic are required.
=
TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
-1.
MAIN STEAM LINE ISOLATION a.
Low, Low - Level 2 N.A.
2)
Low, Low, Low - Level 1
$1.0###*
b.
DELETED DELETED c.
Main Steam Line Pressure - Low
$1.0###*
l d.
Main Steam Line Flow - High 10.5###*
l e.
Condenser Vacuum - Low N.A.
f.
Outboard MSIV Room Temperature - High N.A.
g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
h.
Manual Initiation N.A.
2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Reactor Vessel Water Level Low - Level 3 N.A.
l b.
Reactor Vessel (RHR Cut-In
' Permissive) Pressure - High N.A.
_c.
Manual Initiation N.A.
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
RWCS A Flow - High N.A.##
b.
RWCS Area Temperature - High N.A.
c.
RWCS Area Ventilation a Temperature - High N.A.
d.
.SLCS Initiation N.A.
.e.
Low, Low - Level 2 N.A.
l f.
Manual Initiation N.A.
4
- LIMERICK - UNIT 2 3/4 3-23 Amendment No. 52,93 4
-w, m
TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
4.
HIGH PRESSURE CCOLANT INJECTION SYSTEM IS0lATION a.
HPCI Steam Line A Pressure - High N.A.
l b.
HPCI Steam Supply Pressure - Low N.A.
l c.
HPCI Turbine Exhaust Diaphragm Pressure - High N.A.
d.
HPCI Equipment Room Temperature - High N.A.
e.
HPCI Equipment Room A Temperature - High N.A.
f.
HPCI Pipe Routing Area Temperature - High N.A.
g.
Manual Initiation N.A.
5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.
RCIC Steam Line Pressure - High N.A.
l 4
b.
RCIC Steam Supply Pressure - Low N.A.
[
c.
RCIC Turbine Exhaust Diaphragm Pressure - High N.A.
d.
RCIC Equipment Room Temperature - High N.A.
e.
RCIC Equipment Room A Temperature - High N.A.
f.
RCIC Pipe Routing Area Temperature - High N.A.
g.
Manual Initiation N.A.
LIMERICK - UNIT 2 3/4 3-24 Amendment No. 9?
~
TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPCNSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
6.
PRIMARY CONTAINMENT ISOLAT10N a.
Low, Low - Level 2 N.A.
2)
Low, Low, Low - Level 1 N.A.
b.
Drywell Pressure - High N.A.
[
c.
North-Stack Effluent i
Radiation - High N.A.
d.
Deleted e.
Reactor Enclosure Ventilation Exhaust Duct --Radiation - High N.A.
f.
Deleted g.
Deleted h.
Drywell Pressure - High/
Reactor Pressure - Low N.A.
i.
Primary Containment Instrument Gas to Drywell A Pressure - Low N.A.
j.
Manual Initiation N.A.
7.
SECONDARY CONTAINMENT ISOLATION a.
Low, Low - Level 2 N.A.
b.
Drywell Pressure - High N.A.
c.1.
Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High N.A.
c 2.
Refueling Area Unit 2 Ventilation l
Exhaust Duct Radiation - High N.A.
d.
Reactor Enclosure Ventilation Exhaust l
Duct Radiation - High N.A.
e.
Deleted 1
i 9
LIMERICK - UNIT 2 3/4 3-25 Amendment No. 74,93
1
~
TABLE 3.3.2 1 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Se. ronds)#
f.
Deleted g.
Reactor Enclosure Manual Initiation N.A.
h.
Refueling Area Manual Initiation N.A.
TABLE NOTATIONS (a)
DELETED l
(b)
DELETED Isolation system instrumentation res generator delays assumed fur MSIVs. ponse time for MSIV only.
No diesel i
DELETED l
Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
With 45 second time delay.
Sensor is eliminated from response time testing for the MSIV actuation logic circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.
I l
N LIMERICK -' UNIT 2 3/4 3-26 Amendment No. 52,74,93
c TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ELES RESPONSE TIME (Seconds)
I.
CORE SPRAY SYSTEM s 27#
l 2.
LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM s 40#
l
.3.
AUTOMATIC DEPRESSURIZATION SYSTEM N.A.
4.
HIGH PRESSURE COOLANT INJECTION SYSTEM s 60#
l 5.
LOSS OF POWER N.A.
i
- ECCS actuation instrumentation is eliminated from response time testing.
l i
LIMERICY - UNIT 2-3/4 3-39 Amendment No. 66,93 I
r
- 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATIQH 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 adsorbed following a c.
a 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 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 systwas. 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 approve; 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.
. Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, LBWR Owner's Group from A.C. Thadani, NRC, dated May 15,1991).
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 npplicable.
. Response ~ time may be demonstrated by any series of sequentiai, overlapping or l total channel ' test measurement, provided such tests demonstrau 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.
Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NED0-32291-A.
Response time testing for the remaining channel components is required as noted.
9
. LIMERICK'- UNIT-2 B 3/4 3 1 Amendment No. 9,52,93
o INST.RUMENTATION 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 setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
t determined in accordance with NEDC-30851Pcified surveillance intervals e times have been Supplement 2 Specification Improvement Analysis for BWk Instrumentation Common to RPS and ECCS Instrumentation" as approved by the NRC and documented in the NRC Safet Evaluation Report (SER) (letter to D.N. Grace from C.E. Rossi dated January 1989)tionActuationIns{rumentation"pecificationI rovementAnalysisforBk and NEDC-31677P-A " Technical S Isola as ap the NRC and documented in the NRC SER (letter to S.D. Floyd fr,om C.E. proved bRossi d ted June 18, 1990).
Automatic closure of the MSIVs upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated (May 15,1991).
Some of the trip settings may have tolerances explicitly stated where both i
the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety,inadvertentare established at a level away from the normal operating range to prevent actuation of the systems involved.
Except for the MSIVs the safety analysis does not address individual sensor
. response times or the respon,se times of the logic systems +o which the sensors are connected.
For D.C. operated valves, a 3 second delay is assumed before the valve starts to move.
For A.C. operated valves, it is assumed that the A.C.
power suppl generators.y is lost and is restored by startup of the emergency dieselIn this event, a time of i
starts to move.
In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sentor response) ding delay.
is concurrent with the 10-second diesel startup and the 3 second load center loa 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 anc the 13-seccnd time for emergency power
-establishment will establish the response time for the isolation functions.
Response time testing for sensors are not required based on the analysis in NED0-32291-A. Response time testin required as noted in Table 3.3.2-3.g of the remaining channel components is 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.
j 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION i
The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the conse ability of the operator to control. quences of accidents that are beyond theThis specification provid
)
j requirements, trip setpoints and response times that will ensure effectiveness j
of the systems to provide the design protection. Although the instruments are i
listed by system, in some cases the same instrument may be used to send the 4
actuation signal to more than one system at the same time.
LIMERICK - UNIT 2 B 3/4 3-2 Amendment No. 47,32,52,93 4
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