ML20064M929
| ML20064M929 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/17/1994 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20064M932 | List: |
| References | |
| NUDOCS 9403290089 | |
| Download: ML20064M929 (10) | |
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UNITED STATES l
NUCLEAR REGULATORY COMMISSION gs.....,/
WASHINGTON D.C. 20555-0001 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION HUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.132 License No. NPF-5
' 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 1,1993, as revised January 6,1994, and supplementeri February 3,1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as 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 Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
9403290089 940317 PDH ADOCK 05000366 P
i I 2.
Accordingly, the license is hereby amended by page 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-5 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 132, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B.
atthews, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation-
Attachment:
Technical Specification Changes Date of Issuance:
March 17,1994
2 ATTACHMENT TO LICENSE AMENDMENT NO.tas FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and
'contain vertical lines indicating the areas of change.
Remove Pacei Insert Paaes VII VII XII XII 3/4 6-3 3/4 6-3 3/4 6-4 3/4 6-4 3/4 6-7 3/4 6-7 3/4 6-24 3/4 6-24 8 3/4 6-2 8 3/4 6-2 i
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS EASE SECTION 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT 3/4 6-1 Primary Containment Integrity............................
3/4 6-3 Primary Containment Leakage..............................
3/4 6-6 Primary Containment Air Lock.............................
l Deleted Primary Containment Structural Integrity.................
3/4 6-8 3/4 6-9 Primary Containment Internal Pressure....................
3/4 6-10 Drywell Average Ai r Temperature..........................
3/4.6.2 DEPRESSURIZATION SYSTEMS 3/4 6-11 Suppression Chamber......................................
3/4 6-14 Suppre s s i on Pool Cool i ng.................................
3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.....................
3/4 6-15 i
3/4.6.4 VACUUM RELIEF Suppression Chamber - Drywell Vacuum Breakers............
3/4 6-33 Reactor Building - Suppression Chamber Vacuum 3/4 6-35 Breakers...............................................
3/4.6.5 SECONDARY CONTAINMENT 3/4 6-36 f
Secondary Containment Integrity..........................
Secondary Containment Automatic Isolation Dampers........
3/4 6-37 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL 3/4 6-40 Standby Gas Treatment System.............................
3/4 6-43 Primary Containment Hydrogen Recombiner Systems..........
3/4 6-44 Primary Containment Hydrogen Mixing System...............
HATCH - UNIT 2 VII Amendment No. 132
a INDEX BASES PAGE SECTION REACTOR COOLANT SYSTEM (Continued) 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 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM B 3/4 5-1 3/4.5.3 LOW PRESSURE CORE COOLING SYSTEMS Core Spray System B 3/4 5-2 Low Pressure Coolant Injection System B 3/4 5-3 3/4.5.4 SUPPRESSION CHAMBER B 3/4 5-3 3/4.6 CONTAINMENT SYSTEM!!
3/4.6.1 PRIMARY CONTAINMENT INTEGRITY Primary Containment Integrity B 3/4 6-1 Primary Containment Leakage B 3/4 6-1 Primary Containment Air Lock B 3/4 6-1 I
Deleted Primary Containment Structural Integrity B 3/4 6-2 Primary Containment Internal Pressure B 3/4 6-2 Drywell Average Air Temperature 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 6-4b l
3/4.6.4 VACUUM RELIEF B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT B 3/4 6-5 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL B 3/4 6-5 HATCH - UNIT 2 XII Amendment No.132
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:
a.
An overall integrated leakage rate of:
1.
s L, 1.2 percent by weight of the containment air per 24 a
hours at P,, 57.5 psig, or 2.
s L, 0.849 percent by weight of the containment air per t
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a reduced pressure of P, 28.8 psig.
t b.
A combined leakage rate of:
1.
s 0.60 L, for all penetrations and valves, except for main steam isolation valves, subject to Type B and C tests when pressurized to P,, and 2.
s 0.009 L for the following penetrations *:
a (a) Main steam condensate drain, penetration 8; (b) Deleted (c)
Reactor water cleanup, penetration 14-(d)
Equipment drain sump discharge, penetration 18; (e)
Floor drain sump discharge, penetration 19; and (f)
Chemical drain sump discharge, penetration 55; (g)
Deleted c.
When tested at 28.8 psig**, 100 scf per hour for any one main steam isolation valve and a combined maximum pathway leakage rate of 250 scf per hour for all four main steam lines.
1 APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1.
- Potential bypass leakage paths.
- Exemption to Appendix J of 10 CFR 50.
j HATCH - UNIT 2 3/4 6-3 Amendment No. 132
CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
With:
a.
the measured overall integrated containment leakage rate exceeding 0.75 L, or 0.75 L,, as applicable, or b.
the measured combined leakage rate for all penetrations and valves, except main steam isolation valves, subject to Type B and C tests exceeding 0.60 L or with the measured combined leakage rate for all specifi,d potential bypass leakage path e
penetrations exceeding 0.009 L, or c.
the main steam isolation valve measured leak rate exceeding 100 scf per hour for any one MSIV or a total maximum pathway leakage rate of > 250 scf per hour for all four main steam lines, Restore:
the overall integrated leakage rate (s) to < 0.75 L, or < 0.75 a.
L, as applicable, and b.
the combined leakage rate for all penetrations and valves, except main steam isolation valves, subject to Type B and C tests to s 0.60 L,l bypass leakage path penetrations toand the combined lea specified potentia s 0.009 L,,
and c.
the leakage rate to s 11.5 scf per hour for any main steam isolation valve that exceeds 100 scf per hour, and restore the combined maximum pathway leakage rate to s 250 scf per hour, Prior to increasing the reactor coolant temperature above 212 F.
SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4 - (1972):
Three Type A tests (Overall Integrated Containment Leakage a.
Rate) shall be conducted at 40 10 month intervals during shutdown at either P 57.5 each 10-year service,, period.psig or at P, 28.8 psig duringThe third \\est of each set be conducted during the shutdown for the 10-year plant inservice inspection.
1 HATCH - UNIT 2 3/4 6-4 Amendment No. 132
a CONTAINMENT SYSTEMS MSIV LEAKAGE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION J
3.6.1.4 Deleted 4
r HATCH - UNIT 2 3/4 6-7 Amendment No. 132 i
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T ABt E 3.6.3-1 (Cont:nueJ
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PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER zy B.
MANUAL ISOL ATION VALVES *I I
1.
Deleted 2.
RHR return to recirculation loop isolation valves 2E11-F015A. B 3.
LOCA H, recombiner isoletion velves 2T49-F002 A. B 2T49-FOO4 A. B 4.
Coes sprey isolation velves 2E21-F005 A. B 5.
Service air isolation velves d
2P51-F651 2P51-F513 m4 6.
RBCCW supply end return isolation valves A
2P42-F051 2P42-FOS 2 i
O 3
Q.
3 M
a W
N Idtncludes power cperated valves which do not isolete automatically.
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[0NTAINMENT SYSTEMS BASES 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Deleted 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the primary containment steel vessel will be maintained comparable to the original design standards for the life of the unit.
Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 57.5 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3/4.6.1.6 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations on primary containment internal pressure ensure that the containment peak pressure of 57.5 psig does not exceed the' maximum allowable internal pressure of 62 psig during LOCA conditions or that the external pressure does not exceed the design maximum external pressure of 2 psig. The limit of 0.75 psig for initial positive contain-ment pressure will limit the total pressure to 57.5 psig which is less than the maximum allowable internal pressure and is consistent with the accident analysis.
3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340 F during LOCA conditions and is consistent with the accident analysis.
HATCH - UNIT 2 B 3/4 6-2 Amendment no. 132
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