ML20072P481
| ML20072P481 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/29/1994 |
| From: | Thadani M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072P483 | List: |
| References | |
| NUDOCS 9409070313 | |
| Download: ML20072P481 (22) | |
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3 UNITED STATES li j
NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 2055W1 y
l PHILADELPHIA ELECTRIC COMPANY g
DOCKET N0. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. NPF-39 o
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated January 10, 1994, as supplemented by letter dated July 20, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as arrended (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 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.
9409070313 940829 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:
i Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
75, 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 the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION C
Mohan C. Thadani, Acting 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:
August 29, 1994
ATTACHMENT TO LICENSE AMENDMENT N0. 7s FACILITY OPERATING LICENSE N0. 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 iX iX x
x*
XiX XiX XX XX*
3/4 3-67 3/4 3-67*
3/4 3-68 3/4 3-68 3/4 3-69 3/4 3-70 3/4 3-71 3/4 3-72 B 3/4 3-5 B 3/4 3-5 B 3/4 3-6 B 3/4 3-6*
l 1
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS l
l SECTION PAGE INSTRUMENTATION (Continued)
Table 4.~3.7.1-1 Radiation Monitoring Instrumentation Surveillance l
Requirements.....................
3/4 3-66 I
The information from pages 3/4 3-68 through 3/4 3-72 has been intentionally omitted. Refer to note on page 3/4 3-68..........
3/4 3-68 The information from pages 3/4 3-73 through 3/4 3-75 has been intentionally omitted.
Refer to note on page 3/4 3-73.........
3/4 3-73 Remote Shutdown System Instrumentation and Controls....
3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls.....
3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System Instrumentation Surveillance Requirements.....................
3/4 3-83 Accident Monitoring Instrumentation....................
3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-tation..........
3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrumenta-tion Surveillance Requirements...
3/4 3-87 Source Range Monitors.................................
3/4 3-88 Traversing In-Core Probe System........................
3/4 3-89 Chlorine Detection System.....................
3/4 3-90 Toxic Gas Detection System...........................
3/4 3-91 Fire Detection Instrumentation.........................
3/4 3-92 I
I l
i LIMERICK - UNIT 1 ix Amendment No. 48, 75
i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE i
INSTRUMENTATION (Continued)'
J Table 3.3.7.9-1 Fire Detection Instrumentation....
3/4 3-93 Loose-Part Detection System.............................
3/4 3-97 The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted.
Refer to note on page 3/4 3-98...........
3/4 3-98 l
1 Offgas Monitoring Instrumentation.......................
3/4 3-103 Table 3.3.7.12-1 Offgas
~
Monitoring I'istrumentation.......
3/4 3-104
{
Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements........
3/4 3-107 1
t 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM......................
3/4 3-110 1
3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION..........................................
3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation....
3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpoints....................
3/4 3-114 1
Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-i tion Surveillance Require-ments.............................
3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation loops.....................................
3/4 4-1 LIMERICK - UNIT 1 x
Amendment No.48
\\
Wh2ht.
<9. 89$
INDEX BASES SECTION PAGE fNSTRUMENTATION (Continued)
(Deleted)..............................................
B 3/4 3-5 l (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 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 Of fgas 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 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM...................................
B 3/4 4-1 i
3/4.4.2 SAFETY / RELIEF VALVES...................................
B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE i
Leakage Detection Systems..............................
B 3/4 4-3 j
Operational Leakage....................................
B 3/4 4-3 i
3/4.4.4 CHEMISTRY..............................................
B 3/4 4-3 LIMERICK - UNIT 1 xix Amendment No. #S, 33, 69, 75
INDEX 8ASES i
1 SECTIdN pg REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY.......................................
B 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURE LINITS.............................
8 3/4 4-4 Bases Table 8 3/4.4.6-1 Reactor Vessel i
Toughness.................
8 3/4 4-7 Bases Figure 8 3/4.4.6-1 Fast Neutron Fluence (E>l Mov) At 1/4 T As A Function of Service Life......................
8 3/4 4-8 3/4 4.7 MA !P '"" '. !NE 150'. '"'" "* LvE5........
B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY....................................
B 3/4 4-6 3/4.4.9 RESIDUAL HEAT REM VAL...................................
B 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN............
8 3/4 5-1 3/4.5.3 SUP P RE S S I ON CHAM ER................................ B 3/4 5-2 l
l 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAI N NT Primary Containment Integrity......................
8 3/4 6-1 Primary Containment Leakage........................
I 3/4 6-1 Primary Containment Air Lock.......................
B 3/4 6-1 M51V Leakage Control Systes........................
S 3/4 6-1 Primary Containment Structural Integrity...........
t 3/4 6-2 Drywell and Suppression Chamber Internal Pressure.........................................
8 3/4 6-2 Orywel l Average Ai r Temperature....................
8 3/4 6-2 Orywell and Suppression Chamber Purge Systas.......
I 3/4 6-2 3/4.6.2 DEPRES$URIZATION SYSTEM 5.........................
8 3/4 I LIERICK - UNIT 1 u
Ameneent No. 33 OCT 3 01989 x.
~
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TABLE 4.3,7.1 1 (Continued)
RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREM TABLE NOTATIONS "When irradiated fuel is being handled in the secondary containment.
(a) With fuel in th'e spent fuel storage pool.
(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with N85.
These standards shall permit calibrating the system over its intended range of energy and measurement range.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
O I
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LIMERICK - UNIT 1 3/4 3-67
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Section 3.3,7,7 (Deleted)
S THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
TECHNICAL SPECIFICATIONS PAGES 3/4 3-69 THROUGH 3/4 3-72 0F THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
LIMERICK - UNIT 1 3/4 3-68 Amendment No. II, 40, 75
INSTRUMENTATION BASES 3/4.3.7 HONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in t!e 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.
The specified surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor has been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21,1992).
3/4.3.7.2 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.
3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
3/4.3.7.4 REMOTE SHUTOOWN SYSTEM INSTRUMENTATION AND CONTROLS Tin JPERABILITY of the remote shutdown system instrumentation and controls ensures 0 ;t sufficient capability is available to permit shutdown and maintenance of H0T SHUT %1 of the unit from locations outside of the control room.
This capability is requin M in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, Appendix A.
3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABI;ITY of the accident monitoring instrumentation ensures that sufficient inforr.ation is available on selected plant parameters to monitor and assess important variables following an accident.
This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"
December 1975 and NVREG-0737, " Clarification of TMI Action Plan Requirements,"
November 1980.
3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown.
At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.
LIMERICK - UNIT I B 3/4 3-5 Amendment No. M, 33, 70, 75
l INSTRUMENTATION B2SES 3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY.of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.
The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e.,
detectors) prior to performing an LPRM calibration function.
Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE.
The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).
3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective tctions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection.
Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection.
The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95, " Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.
Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D.
Binz, IV, from C.E. Rossi dated July 21,1992).
. s c.
3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge extin-quishing agent in a timely manner.
Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warr.ing and notification.
Consequently, the minimum number of OPERABLE fire detectors must be greater.
LIMERICK - UNIT 1 B 3/4 3-6 Amendment No. 48. M. 70
-.nna 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 Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
36, 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 the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
Mohan C. Thadani, Acting 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:
August 29, 1994 i
l
a asav 54 a
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UNITED STATES i{'
NUCLEAR REGULATORY COMMISSION s.s,#
WASHINGTON. D.C. 20555-0301
+....
PHILADELPHIA ELECTRIC COMPANY i
DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 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 January 10, 1994, as supplemented by letter dated July 20, 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; 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 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 and all applicable requirements have been satisfied.
l I
ATTACHMENT TO LICENSE AMENDMENT NO. 36 FACILITY OPERATING LICENSE N0. NPF-85 DOCKET NO. 50-353 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 ix ix x
x*
xix xix xx xx*
3/4 3-67 3/4 3-67*
3/4 3-68 3/4 3-68 3/4 3-69 3/4 3-70 3/4 3-71 3/4 3-72 B 3/4 3-5 8 3/4 3-5 B 3/4 3-6 8 3/4 3-6*
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE INSTRUMENTATION (Continued)
Table 4.3.7.1-l Radiation Monitoring Instrumentation Surveillance Requirements.................
3/4 3-66 The information from pages 3/4 3-68 through 3/4 3-72 has been intentionally omitted.
Refer to note on page 3/4 3-68..........
3/4 3-68 The information from pages 3/4 3-73 through 3/4 3-75 has been intentionally omitted.
Refer to no te on page 3/4 3-73..........
3/4 3-73 Remote Shutdown System Instrumentation and Controls....
3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls....
3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System Instrumentation Surveillance Requirements...............
3/4 3-83 Accident Monitoring Instrumentation............
3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-tation..........................
3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrument.a-tion Surveillance Requirements..
3/4 3-87 Source Range Monitors...................................
3/4 3-88 i
Traversing In-Core Probe System........................
3/4 3-89 j
Chlorine Detection System..............................
3/4 3-90 Toxic Gas Detection System.............
3/4 3-91 Fire Detection Instrumentation............
3/4 3-92 LIMERICK - UNIT 2 ix Amendment No II, 36
l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREENTS 8
SECTION PAGE INSTRUMENTATION (Continued)
Table 3.3.7.'9-1 FireDetectionInstrumentation....
3/4 3-93 Loose-Part Detection Systas.............................
3/4 3-97 The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted.
Refer to note on page 3/4 3-98.....................
3/4 3-98 Offgas Monitoring Instrumentation.......................
3/4 3-103 Table 3.3.7.12-1 Offgas Monitoring Instrumentation.......
3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements........
3/4 3-107 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM......................
3/4 3-110 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION..........................................
3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation....
3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrimentation Setpoints...........................
3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation Surveillance Requirements.........
3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.....................................
3/4 4-1 LIMERICK - UNIT 2 x
Amendment No.11
INDEX BASES SECTION PAGE INSTRUMENTATION (Continued)
(Deleted).....................................................
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 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 Leve1...........................
B 3/4 3-B 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..........................................
B 3/4 4-1 j
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 Operational Leakage...........................................
B 3/4 4-3 3/4.4.4 CHEMISTRY.....................................................
B 3/4 4-3a LIMERICK - UNIT 2 xix Amendment No. YI, I2, 77, 32, 36
T
.IM BASE 5 5,fgT,lg!
3 REACTOR C00VWrf $YSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY.......................................
B 3/4 4-4 3/4.4.6 PRE 55URE/TDFERATURE LIMIT 5.............................
B 3/4 4-4 Bases Table B 3/4.4.6-1 Reacter Vessel Toughness.................
B 3/4 4-7 Bases Figure B 3/4.4.5-1 Fast Neutron Fluence (E)1 MeV) At 1/4 T As A Function of Service Life......................
B 3/4 4-5 3/4.4.7 MAIN STEAM LINE ISCLATION VAi.VE5........................
B 3/4 4-6 3/4.4.B STRUCTURAL INTEGRITY....................................
B 3/4 4-6 3/4.4.9 RE5IOUAL MAT RDOVAL...................................
B 3/4 4-6 s
3/4.5 EERGENCY CORE C00 LIM 5Yuem 3/4. 5.1 arW 3/4. 5. 2 ECCS - OPERATIM and SNLTTD0lSI............
B 3/4 5-1 3/4.5.3 SUPPRESSION QWWER................................
B 3/4 5-2 3/4.6 CONTAll00ENT SYSTEMS 3/4.6.1 PRIMARY CONTAIWWIT Primary Containment Integrity......................
B 3/4 6-1 Primary Containment Laakage........................
B 3/4 6-1 Primary Containment Air Lock.............'..........
B 3/4 6-1 MSIV Leakage Centrol Systaa........................
B 3/4 6-1 Primary Containment Structural Integrity...........
B 3/4 6-2 Drywell and Suppression Chamber laternal Pressure.........,...............................
B 3/4 6-2 Drywell Average Ai r Temperature....................
B 3/4 6-2 Drywell and Suppression Chamber Purge Systaa.......
B 3/4 6-2 3/4.6.2 DEPRE550RIZATION SYSTEM 5...........................
B 3/4 6-3
' " 8'CE - UNIT 2 xx
~
8 TABLE 4.3.7.1-1 (Continued)
RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS
- When irradiated fuel is being handled in the secondary containment.
(a) With fuel in the spent fuel storage pool.
(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.
These standards shall permit calibrating the system over its intended range of energy and measurement range.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
)
l LIMERICK - UNIT 2 3/4 3-67
Section 3.3.7.2 (Deleted) l THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
TECHNICAL SPECIFICATIONS PAGES 3/4 3-69 THROUGH 3/4 3-72 0F THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
i 4
i i
i i
LIMERICK - UNIT 2 3/4 3-68 Amendment No. 36
INSTRUMENTATION BASES 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONIT0'ING INSTRUMENTATION R
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 variable following an accident.
This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.
The specified surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor has been determined in accordance with GENE-770-06-1,
" Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specification," as approved by the NRC and documented in the SER (letter to R. D. Binz, IV, from C. E. Rossi dated July 21, 1992).
3/4.3.7.2 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.
3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room.
This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, Appendix A.
The Unit 1 RHR transfer switches are included only due to their potential impact on the RHRSW system, which is common to both units.
3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.
This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors )rovide the operator with information of the status of the neutron level in tie core at very low power levels during startup and shutdown. At these power levels, reactivity additions shall not be made without this flux level information available to the operator.
When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.
LIMERICK - UNIT 2 B 3/4 3-5 Amendment No. II, 17, 33, 36
9 INSTRUMENTATION BASES
)
3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM j
The OPERASILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use l
of this equipment accurately represent the spatial neutron flux distribution of t
the reactor core.
+
The TIP system OPERA 8ILITY is demonstrated by normalizing all probes (i.e.,
detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors tu monitor selected areas of the reactor core, thus all detectors any not be required to be OPERABLE.
The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).
3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilating system will automatically be placed in the chlorine isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of. operation to provide the required protection. The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95 ' Pro-taction of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.
Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and dot.umented in the SER (letter to R.D.
Binz, IV, from C.E. Rossi dated July 21,1992).
3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERASILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge extin-quishing agent in a timely manner.
Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.
Consequently, the minimum number of OPERA 8LE fire detectors must be greater.
LIMERICK - UNIT 2 8 3/4 3-6 Amendment No. JJ, 23 33 l
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