ML20045A436
| ML20045A436 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/07/1993 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co |
| Shared Package | |
| ML20045A437 | List: |
| References | |
| DPR-16-A-164 NUDOCS 9306100257 | |
| Download: ML20045A436 (5) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO, 50-219 0YSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.164 License No. DPR-16 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.,
(the licensee), dated March 3, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission'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 Comission; 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 Comission'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.
9306100257 930607 PDR ADOCK 05000219 P
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. 2.
Accordingly, the license is amenued ~.,y changes to the Technical-Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-16 is hereby amended to read as follows:
(2) Technical Soecifications.
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.164, are hereby incorporated in the -
license. GPU Nuclear Corporation shall operate the facility.in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance, to be implemented during the 15R refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION W
8kS-John F. Stolz, Director Project Directorate I-4 Division of Reactor Projects -.I/II Office of Nuclear Reactor Regulation
Attachment:
I I
Changes to the Technical Specifications Date of Issuance: June 7, 1993 1
2
ATTACHMENT TO LICENSE AMENDMENT NO. 164 l
FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 2.3-2 2.3-2 4.3-1 4.3-1
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' FUNCTION LIMITING SAFETY SYSTEM SETTINGS B.
Neutron Flux, Control Rod Block The Rod Block setting shall be FRP S 5 [(0.90 x 10') W + 53.1) [MFLPD]
with a maximum setpoint'of 108% for core flow equal to 61 x 10' lb/hr and greater.
The definitions of S, W, FRP and MFLPD used above for.the APRM scram trip apply.
The ratio of FRP to MFLPD shall be set equal to 1.'O unless the actual operating value is less than 1.0, in which case the actual operating value will be used.
This adjustment may be accomplished by increasing the APRM gain and thus reducing the flow referenced APRM rod block curve by the reciprocal of the APRM gain change.
1 C.
Reactor High, 51060 psig Pressure, Scram t
D.
Reactor High Pressure, 2 e 5 1070 psig Relief Valves Initiation
. 3 e 5 1090 psig E.
Reactor High Pressure, 51060 psig with time delay Isolation Condenser 53 seconds Initiation F.
Reactor High Pressure, 4 e 1212 psig 12 psi Safety Valve Initiation 5 e 1221 psig.
!12 psi l
G.
Low Pressure Main Steam 1825 psig (initiated in IRM Line, MSIV Closure
- range 10)
H.
Main Steam Line Isolation 510%-Valve Closure from
. Valve Closure, Scram full open I.
Reactor Low Water Level, 111'5" above the top of the Scram active fuel as indicated under normal operating conditions J.
Reactor Low-Low Water 17'2" above the top of the Level, Main Steam Line active fuel as indicated under Isolation Valve Closure normal operating conditions OYSTER CREEK 2.3-2 Amendment No. 73, 75,111,15a164
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3Il 4.3 REACTOR COOLANT Acolicability:
Applies to the surveillance requirements for the reactor coolant system.
Obiective:
To determine the condition of the reactor coolant system and the operation of the safety devices related to it.
Soecification:
A.
Materials surveillance specimens and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core.
Specimens and monitors shall be periodically removed, tested, and evaluated to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials.
The results of these evaluations shall be used to assess the adequacy of the P-T curves (a),(b) and (c) in Figure 3.3.1.
New curves shall be generated as required.
B.
Inservice inspection of ASME Code Class 1, Class 2 and Class 3 systems and components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(1).
C.
Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to
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10 CFR, Section 50.55a(g)(6)(i).
D.
A visual examination for leaks shall be made with the reactor coolant system at pressure during each scheduled refueling outage or after major repairs have been made to the reactor coolant system in accordance with Article 5000,Section XI. The requirements of specification 3.3.A shall be met during the test.
E.
Each replacement safety valve or valve that has been repaired shall be tested in accordance with subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code.
Setpoints shall be as follows:
Number of Valves Set Points (esial 4
1212 2 12 i
i 5
1221 1 12 1
F.
A sample of reactor coolant shall be analyzed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the purpose of determining the content of chloride ion and to check the conductivity.
OYSTER CREEK 4.3-1 Amendment No.: 82, 90, 120, 150, 151,164