ML20212H060
| ML20212H060 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/21/1999 |
| From: | Clifford J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212H065 | List: |
| References | |
| NUDOCS 9909300198 | |
| Download: ML20212H060 (11) | |
Text
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UNITED STATES s
j NUCLEAR REGULATORY COMMISSION l
t WASHINGTON. D.C. 2056!WX101 k,...../
PUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.123 License No. NPF-57
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1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by the Public Service Electric & Gas Company (PSE&G) dated May 24,1999, as supplemented June 21,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, i
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 set forth in 10 CFR Chapter I; 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 indicatad in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:
I 9909300198 990921 PDR ADOCK 05000354 P
2-
-(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.123, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license. PSE&G shall operate the
' facility in accordance with the Technical Specifications and the EnvTonmental Protection Plan.
3.
The license amendment is effective as of its date of issuance, and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION Y.
Ja es W. Clifford, Chief, Section 2 Project Directorate I.
Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical
- Specifications Date of Issuance:
September 21, 1999 9
ATTACHMENT TO LICENSE AMENDMENT NO.123 l
FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 i
i Replace the following pages of the Appendix "A" Technical Specifications with the attached revised pages. The revised pages are identified by Amendment number and contain marginal i
lines indicating the areas of change.
i Remove insert iv IV vi vi vil vii X
X Xiii Xiii XIV XIV xviii xviii XIX XIX i
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i 1
6 i
i INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE l
I
'1 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow........
2-1 THERMAL POWER, High Pressure and High Flow 2-1 3
Reactor Coolant System Pressure 2-1 Reactor Vessel Water Level..............
2-2 I
I 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints..........
2-3 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints..........
2-4 BASES 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...........
B 2-1 l
i THERMAL POWER, High Pressure and High Flow............
B 2-2 Table B2.1.2-1 Uncertainties Used in the Determination of the Fuel Cladding Safety Limit...........
B 2-3 Reactor Coolant System Pressure.............................
B 2-5 Reactor Vessel Water Level.....................
B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......
B 2-6 s
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HOPE CREEK iv Amendment No. 123 l
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1 s'
INDEX gMITINGCONDITIONSFOR'OPERATIONANDSURVEILLANCEREQUIREMENTS SECTION PAGE 3/4 2.2 APRM SETPOINTS..........................................
3/4 2-2
.3/4.2.3 MINIMUti CRITICAL POWER RATIO............................
3/4 2-3
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3/4.2.4 LINEAR HEAT GENERATION RATE.......
3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...............
3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation.........................
3/4 3-2 Figure 4.3.1.1-1 Reactor Protection System Surveillance Requirements............
3/4 3-7 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.....................
3/4 3-9 Table 3.3.2-1 Isolation Actuation Instrumentation..
3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints................................
3/4-3-22 Table 4.3.2.1-3 Isolation Actuation Instrumentation Surveillance Requirements..............
3/4 3-28 HOPE CREEK vi Amendment No.123 l
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE
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3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.
3/4 3-32 Table 3.3.3-1 Emergency Core Cooling System Actuation Instrumentation.
3/4 3-33 1
1 Table 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation Setpoints........
3/4 3-36
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I Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements.
3/4 3-39 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION l
ATWS Recirculation Pump Trip System Instrumentation......
3/4 3-41
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Table 3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation..
3/4 3-42 Table 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints.
3/4 3-43 Table 4.3.4.1-1 ATWS Recirculation Pump Trip Actuation
{
Instrumentation Surveillance Requirements....
3/4 3-44 End-of-Cycle Recirculation Pump Trip System
{
Instrumentation......
3/4 3-45 j
Table 3.3.4.2-1 End-of-Cycle Recirculation Pump Trip System Instrumentation...............
3/4 3-47 Table 3.3.4.2-2 End-of-Cycle Recirculation Pump Trip Setpoints......................
3/4 3-48 j
Table 3.3.4.2-3 End-of-Cycle Recirculation Pump Trip System Response Time.
3/4 3-49 Table 4.3.4.2.1-1 End-of-Cycle Recirculation Pump Trip System Surveillance Requirements.....
3/4 3-50 3/*.a.a REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION..
3/4 3-51 i
HOPE CREEK vii Amendment No. 123 l
6 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumentation Setpoints...
3/4 3-107 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation Surveillance Requirement.........
................. 3/4 3-108 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops........
3/4 4-1 Figure 3.4.1.1-1
% Rated Thermal Power Versus Core Flow....
3/4 4-3 Je t Pumps...........
3/4 4-4 Recirculation Loop Flow.
3/4 4-5 Idle Recirculation Loop Startup.............
3/4 4-6 3/4.-4.2 SAFETY / RELIEF VALVES Safety / Relief Valves.........
3/4 4-7 Safety / Relief Valves Low-Low Set Function.........
3/4 4-9 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...
3/4 4-10 Operational Leakage........................
3/4 4-11 Table 3.4.3.2-1 Reactor Coolant System Pressure l
Isolation Valves.......
3/4 4-13 Table 3.4.3.2-2 Reactor Coolant System Interface j
Valves Leakage Pressure Monitors......
3/4 4-14 i
3/4.4'4 CHEMISTRY...................
3/4 4-15 Table 3.4.4-1 Reactor Coolant System Chemistry Limits.......
3/4 4-17 3/4.4.5 SPECIFIC ACTIVITY.........................
3/4 4-18 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program..
3/4 4-20 l
1 HOPE CREEK x
Amendment No.123 l
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INDEX LIMITING CONDITIONS FOR OPERATION AND' SURVEILLANCE REQUIREMENTS i
l SECTION PAGE l
3/4.7.3 FLOOD PROTECTION........
i 3/4 7-9 Table 3.7.3-1 Perimeter Flood Doors...............
3/4 7-10 s
3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM..............
3/4 7-11 3/4.7.5 SNUBBERS...................
3/4 7-13 Table 4.7.5-1 Snubber Visual Inspection Interval....
3/4 7-17a Figure 4.7.5-1 Sample Plan 2) for Snubber Functional Test..
3/4 7-18 3/4.7.6 SEALED SOURCE CONTAMINATION..
3/4 7-19 3/4.7.7 MAIN TURBINE BYPASS SYSTEM....
3/4 7-21 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C.
SOURCES A.C.
Sources-Operating............
3/4 8-1 Table 4.8.1.1.2-1 Diesel Generator Test Schedule...
3/4 8-10 A.C.
Sources-Shutdown.......
3/4 8-11 3/4.b.2 D.C.
SOURCES D.C.
Sources-Operating....
3/4 8-12 l
Table 4.8.2.1-1 Battery Surveillance Requirements..
3/4 8-15 D.C.
Sources-Shutdown...........................
3/4 8-17 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution - Operating..............
3/4 8-18 nistribution - Shutdown...
3/4 8-21 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary Containment Penetration Conductor Overcurrent Protective Devices.........
3/4 8-24 Table 3.8.4.1-1 Primary Containment Penetration Conductor Overcurrent Protective Devices...
3/4 8-26 Motor Operated valve Thermal Overload Protection (Bypassed)........................
3/4 8-30 HOPE CREEK xiii Amendment No. R3 l
a INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Motor Operated Valve Thermal Overload Protection (Not Bypassed)..................
3/4 0-38 Table 3.8.4.3-1 Motor Operated valves-Thermal Overload Protection (Not Bypassed).............
3/4 8-39 Reactor Protection System Electric Power Monitoring.....
3/4 8-40 Class 1E Isolation Breaker Overcurrent Protection Devices (Breaker Tripped by LOCA Signal)........
3/4 8-41 Table 3.6.4.5-1 Class 1E Isolation Breaker Overcurrent Protective Devices (Breaker Tripped by a LOCA Signal)............
3/4 0-42 Power Range Neutron Monitoring System Electric Power Monitoring.
3/4 8-44 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH..................
3/4 9-1 3/4.9.2 INSTRUMENTATION..............
3/4 9-3 3/4.9.3 CONTROL ROD POSITION.....
3/4 9-5 3/4.9.4 DECAY TIME............
3/4 9-6 3/4.9.5 COMMUNICATIONS.......
3/4 9-7 3/4.9.6 REFUELING PLATFORM.....
3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL.......
3/4 9-10 j
1 3/4.9.8 WATER LEVEL - REACTOR VESSEL..............
3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL...............
3/4 9-12 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal....
3/4 9-13 Multiple Control Rod Removal....
3/4 9-15 i
HOPE CREEK xiv Amendment No.123 l
I
r INDEX BASES W
SECTION
~PAGE INSTRUMENTATION' (Continued)
Remote Shutdown Monitoring 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 3-5
'3/4.3.8 DELETED.........
....B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION
.B 3/4 3-7 Figure B3/4 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 3/4.1.2 SAFETY / RELIEF VALVES...........
.....B 3/4 4-2 l
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 4-3 3/4.4.5 SPECIFIC ACTIVITY........
........B 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS............
................B 3/4 4-5 Table B3/4.4.6-1 Reactor Vessel Toughness.....
................B 3/4 4-7 Figure B3/4.4.6-1 Fast Neutron Fluence (E>1Mev) at (1/4lT as a Function of Service life................B 3/4 4-8 l
HOPE CREEK xviii Amendment No.123 l
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INDEX BASES SECTION PAGE 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...
B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY............
B 3/4 4-6 3/4.4.9 RESIDUAL HEAT REMOVAL.........
B 3/4 4-E 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1/2 ECCS - OPERATING and SHUTDOWN......
B 3/4 5-1 l
3/4.5.3 SUPPRESSION CRAMBER..........
B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMElff Primary Containment Integrity......
B 3/4 6-1 Pritaary Containment Leakage......
B 3/4 6-1 Primary Containment Air Locks.
B 3/4 6-1 MSIV Sealing System..........
B 3/4 6-1 Primary Containment Structural Integrity...
B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure B 3/4 6-2 Drywell Average Air Temperature....
B 3/4 6-2 Drywell and Suppression Chamber Purge System B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...........
B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.......
B 3/4 6-5 3/4.6.4 VACUUM RELIEP....
B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT...............
B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL..
B 3/4 6-6 HOPE CREEK xix Amendment No. 123 l