ML20085M138
| ML20085M138 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  (DPR-77-A-204, DPR-79-A-194) |
| Issue date: | 06/14/1995 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20085M141 | List: |
| References | |
| NUDOCS 9506290059 | |
| Download: ML20085M138 (16) | |
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UNITED STATES j-j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20666-0001
'+9 *ss e* g TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT. UNIT 1
' AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 204 License No. DPR-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 6, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),r j
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 (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 conducted in compliance with the C0mmission'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.
9506290059 950614 PDR ADOCK 05000327 P
PDR e._
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- Accordingly, the license is amended by changes to the Technical l
Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby.
-amended to read as follows.
j (2) Technical Specifications The Technical. Specifications contained in Appendices A and B, as revised through Amendment No. 204, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of its date of issuance, to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
)-
A Frederick J. Hebdon, Director r
Project Directorate II-3 i
Division of Reactor Projects - I/II.
j Office of Nuclear Reactor Regulation.
j
Attachment:
Changes to the Technical Specifications Date of Issuance: June 14, 1995 I
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ATTACHMENT TO LICENSE AMENDMENT NO. 204 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the' captioned amendment number and contain marginal lines indicating the area of change.
I REMOVE INSERT Index X Index X Index XIV Index XIV 3/4 9-7 3/4 9-7 3/4 9-7a 3/4 9-7a B3/4 9-2 B3/4 9-2 r
a
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION Egg Motor Operated Valves Thermal Overload Protection.........
3/4 8-34 Isolation Dev1ces.........................................
3/4 8-37 l
3/4.9 REFUELING OPERATIONS q
3/4.9.1 BORON CONCENTRATION.......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................
3/4 9-2 3/4.9.3 DECAY TIME................................................
3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.........................
3/4 9-4 I
3/4.9.5 COMMUNICATIONS............................................
3/4 9-5 3/4.9.6 MANIPULATOR CRANE.........................................
3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED)..............
3/4 9 7 7
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION i
All Water Levels..........................................
3/4 9-8 Low Water Level...........................................
3/4 9-8a 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM..................
3/4 9-9 l
3/4.9.10 WATER LEVEL - REACTOR VESSEL..............................
3/4 9-10 3/4.9.11 WATER LEVEL - SPENT FUEL PIT..............................
3/4 9-11 3/4 9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM...................
3/4 9-12 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN...........................................
3/4 10-1 j
3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.....
3/4 10-2 3/4.10.3 PHYSICS TESTS.............................................
3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS.....................................
3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.....................
3/4 10-5 SEQUOYAH - UNIT 1 X
Amendment No. 61, 204
INDEX BASES SECTION Eagg 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM........................
B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK.........................................
B 3/4 7-4 3/4.7.6 FLOOD PR0TECTION...........................................
B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM..................
B 3/4 7-4 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM....................
B 3/4 7-5 3/4.7.9 SNUBBERS...................................................
B 3/4 7-5 3/4.7.10 SEALED SOURCE CONTAMINATION................................
B 3/4 7-6 3/4.7.11 FIRE SUPPRESSION SYSTEMS...................................
B 3/4 7-7 3/4.7.12 FIRE BARRIER PENETRATIONS..................................
B 3/4 7-7 I
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS.............................................................
B 3/4 8sl 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES....................
B 3/4 8-2 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION........................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION............................................
B 3/4 9-1 3/4.9.3 DECAY TIME.................................................
B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS..........................
B 3/4 9-1 3/4.9.5 COMMUNICATIONS.............................................
B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE..........................................
B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED)...............
B 3/4 9-2 l
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION..............
B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM...................
B 3/4 9-2 SEQUOYAH - UNIT 1 XIV Amendment No. 157, 204
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REFUELING OPERATIONS i
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3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA i
LIMITING CONDITION FOR OPERATION 3.9.7 This specification is deleted.
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SEQUOYAH - UNIT 1 3/4 9-7 Amendment No. 91, 167, 194, 204 l
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r This page intentionally deleted SEQUOYAH - UNIT 1 3/4 9-7a Amendment No. 167, 204
REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:
- 1) manipulator cranes will be used for movement of drive rods and fuel assem-blies, 2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and 3) t,he core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
i 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA This specification is deleted.
l 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that; 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING MODE, and 2) sufficient coolant circulation is '
maintained through the reactor core to minimize the effects of a boron dilutidr.
incident and prevent boron stratification. The minimum required flow rate of '
2000 gpm ensures decay heat removal, minimizes the probability of losing an RHR pump by air-entrainment from pump vortexing, and minimizes the potential for valve damage due to cavitation or chatter.
Losing an RHR pump is a particular concern during reduced RCS inventory operation. The 2000 gpm value is limited by the potential for cavitation in the control valve and chattering in the 10-inch check valve.
i The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of resid-ual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the Core.
SEQUOYAH - UNIT I B 3/4 9-2 Amendment No. 134, 167, 204
pw 41 UNITED STATES i
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NUCLEAR REGULATORY COMMISSION C
WASHINGTON, D.C. 20556 4001
\\..... pf TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 i
SE000YAH NUCLEAR PLANT. UNIT 2
- AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 194 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 6,1995, 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; r
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 1
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 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.
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2.
Accordingly, the' license is amended by changes to the Technical ~
Specifications as indicated in the attachment to this license amendment i
and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby j
amended to read as follows:
l (2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 194, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of its date of issuance,' to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION i
?
et Frederick J. Hebaon, Director j
7 Project Directorate II-3 i
Division of Reactor Projects - I/II Office.of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 14, 1995 4
1 ATTACHMENT TO LICENSE AMENDMENT NO.194 i
FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the, captioned amendment number and contain marginal lines-indicating the area of change.
REMOVE INSERT Index X Index X Index XIV Index XIV 3/4 9-8 3/4 9-8 3/4 9-8a 3/4 9-8a l
B3/4 9-2 B3/4 9-2 a
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i INDEX LIMITING CONDITIONS FOR OPERATION ~AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices.................................................
3/4 8-16 Motor Operated Valves Thermal Overload Protection.........
3/4 8-33 Isolation Devices.........................................
3/4 8-36 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................
3/4 9-3 3/4.9.3 DECAY TIME................................................
3/4 9-4 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.........................
3/4 9-5 3/4.9.5 COMMUNICATIONS............................................
3/4 9-6 3/4.9.6 MANIPULATOR CRANE.........................................
3/49h 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED)..............
3/4 9-8 l
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels..........................................
3/4 9-9 Low Water Level...........................................
3/4 9-10 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM..................
3/4 9-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL..............................
3/4 9-12 3/4.9.11 WATER LEVEL - SPENT FUEL PIT..............................
3/4 9-13 3/4 9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM...................
3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN...........................................
3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.....
3/4 10-2 3/4.10.3 PHYSICS TESTS.............................................
3/4 10-3 SEQUOYAH - UNIT 2 X
Amendment No. 53, 194
o INDEX BASES i
SECTION pgd
[
3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM........................
B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK.........................................
B 3/4 7-4 i
3/4.7.6 FLOOD PROTECTION...........................................
B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM..................
B 3/4 7-4.
i 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM....................
B 3/4 7-5 3/4.7.9 SNUBBERS...................................................
B 3/4 7-5 3/4.7.10 SEALED SOURCE CONTAMINATION................................
B 3/4 7-6a 3/4.7.11 FIRE SUPPRESSION SYSTEMS...................................
B 3/4 7-7 j
3/4.7.12 FIRE BARRIER PENETRATIONS..................................
B 3/4 7-7 j
3/4.8 ELECTRICAL POWER SYSTEMS r
3/4.8.1 and 3/4.8.2 A.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS.............................................................
B 3/4 8-1 I
'3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES....................
B 3/4 8-2 l
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION........................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION............................................
B 3/4 9-1 3/4.9.3 DECAY TIME.................................................
B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS..........................
B 3/4 9-1 3/4.9.5 COMMUNICATIONS.............................................
B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE..........................................
B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED)...............
B 3/4 9-2 l
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION..............
B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM...................
B 3/4 9-2 SEQUOYAH - UNIT 2 XIV Amendment No. 194
REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA i
I LIMITING CONDITION FOR OPERATION 3.9.7 This specification is deleted.
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SEQUOYAH - UNIT 2 3/4 9-8 Amendment No. 81, 157, 185, 194
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l SEQUOYAH - UNIT 2 3/4 9-Ba Amendment No. 152 194 l
4 REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:
- 1) manipulator cranes will be used for movement of drive rods and fuel assen-blies, 2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and 3) t.he core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA This specification is deleted.
l 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that; 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and 2) sufficient coolant circulation ~1s maintained thru the reactor core to minimize the effects of a boron dilution '
incident and prevent boron stratification. The minimum required flow rate of 2000 gpm ensures decay heat removal, minimizes the probability of losing an RHR pump by air-entrainment from pump vortexing, and minimizes the potential for valve damage due to cavitation or chatter.
Losing an RHR pump is a particular concern during reduced RCS inventory operation. The 2000 gpm value is limited by the potential for cavitation in the control valve and chattering in the 10-inch check valve.
The requirement to have two RHR loops OPERABLE when there is less than i
23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operat-ing RHR loop, adequate time is provided to initiate emergency procedures to cool the core.
SEQUOYAH - UNIT 2 B 3/4 9-2 Amendment No. 121, 157, 194