ML20128G806
| ML20128G806 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/22/1985 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20128G812 | List: |
| References | |
| NUDOCS 8505300318 | |
| Download: ML20128G806 (7) | |
Text
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UNITED STATES -
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g NUCLEAR REGULATORY COMMISSION
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'p; WASHINGTO N, D. C. 20555
^g-.....J GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION AMENOVENT TO PROVISIONAL OPERATING LICENSE Amendment No. 82 License No. DPR-16 1.
The Nuclear Regulatory Comission (the Ccmmission) has found that:
A.
The application for amendment by GPU Nuclear Corporation and Jersey Central Power and Light Company (the licensees) dated June 8, 1984, which supersedes the application dated December 11, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Connission'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; 0.
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 Connissinn's regulations and all applicable requirements have been satisfied.
8505300318 850522 PDR ADOCK050g9 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 Provisional Operating License No. DPR-16 is hereby amended to read as follows:
- (2) Technical Sct'rifications The Technical Specifications cnntained in Appendices A and B as revised through Amendment No. 82, 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 its issuance.
FOR T NUCLEAR REGULATORY COMMISSION
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John
. Zwolinski, Chief Opera ing Reactors Branch #5.
Divisi n of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 22, 1985.-
i
c ATTACHMENT T LICENSE AMENDMENT NO. 82 PROVISI0f!AL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Revise Appendix A Technical Specific'ations by renoving the pages identified below and insertire the enclosed pages. The revised pages are identified by the captioned amendrent number ' nd contain vi:rtical lines indiceting the area a
o#. change;
. REMOVE INSEP.T 4.3-1 4.3-1 4.3-la 4.3-2 4.3-2' 4.3-3 4.3-3 4.3-4 4.3-5'
'4.3-6 o
4.3-7 4.3-8 4.3-9 4.3-4
w l,
t 4.3 P.EACTOR COOLANT 4.3-1 Applicability:
Applies to the surveillance requirements for the reactor coolant system.
Ob,1ective:
To determine the condition of the reactor coolant system-and the operation of the safety devices related t; it.
Specification:
A.
Neutron flux monitors shall be installed in the reactor vessel ed.iacent to the vessel wall at the core midplane level. The monitors.shall be removed and tested at the first refueling outage to-experimentally verify the calculated values of integrated neutron flux that are used to determine the NDTT fron Figure 3.3.1.
B.
Inservice inspection of ASME Code Class 1, Class 2 and Class:3 systems and components shall be performed l
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 pumos 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 10 CFR, Section 50.55a(g)(6)(1 D..
A visual examination for leaks shall be made with l-the reactor coolant system at pressure during each scheduled refueling outaoe or after major repairs have been made to the-reactor coolant system. The requirements of specification 3.3.A shall be met during the. test.
E.
Each replacement safety valve or valve that has
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been repaired.shall be bench checked for proper set point. A minimum of 5 of the valves shall be bench = checked or replaced with a bench checked valve each refueling outage such.that all valves are checked in three successive refueling outages, i
to ensure set points are as follows:
~ Number of Valves Set Point (psio) 4 1212 12 4
1221 12 4
1230 12 4
1239 : 12 Amendment No. 82
4.3-2 l
'F.
A sample of reactor coolant shall be analyzed at least
-l 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.
- G.
Primary Coo
- ant System Pressure Isolation Valves l
Specification:
1.
Periodic leakage testing (a)on each valve listed in table 4.3.1 shall be accomplished l
prior to exceeding 600 psio reactor pressure every time the plant is placed on the cold shutdown condition for refueling, each tire the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 nonths, and prior. to returnino the valve to service after maintenance, repair or replacement work is performed.
i (a)
To satisfy ALARA reouirenents, leakage may be measured indirectly (as from the performance of pressure indicatorsi if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
I
- NRC Order dated April 20, 1981 Amendment No. 82
4.3-3 Bases:
Numerous data are avaijable relating integrated flux and the change in Nil-Ductility Transitien Temperature (NDTT) in various steels.
The base
-retal has been demonstrated to be relatively insensitive to neutron irradiation (see expected NDT changes in FDSAR Table IV-1-1, and Figures IV-2-9 and IV-2-10). The most conservative data has been used in Specification 3.3.
The integrated flux at the vessel wall is calculated from core physics data and will be measured using flux monitors installed inside the vessel.
The measurements of the neutron flux at the vessel wall will be used to check and if necessary correct, the calculated data to determine an accurate flux.
From this a conservative NDT temperature-cag be determined..Since no shift will occur until an -integrated flux of 10'7 nyt is reached, the confirmation can be made lona before an NDTT shift would occur.
The inspection orogram will reveal problem areas should they occur, before a leak develops.
In addition, extensive visual ~ inspection for leaks will be made on critical systems.
Oyster Creek was designed and constructed prior to the existence of ASME Section XI.
For this reason, the degree of access required by ASME Section XI is not generally available and will be addressed as " requests for relief" in accordance with 10 CFR 50.55a(g).
Experience in safety valve operation shows that a check of approximately 1/3 of the safety valves per year is adequate to detect failures or deterioration. The tolerance value is specified in Section I of the ASME Code at +1% of design pressure. An analysis has been performed which shows that with all safety valves set 12 psig higher the safety limit of 1375 psig is not exceeded.
Conductivity _ instruments continuously monitor the reactor coolant.
Experience indicates that a check of the conductivity instrumentation at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is adequate to ensure accurate readings. The reactor water sample will also be used to deternine the chloride ion content to assure that the limits of 3.3.E are not exceeded. The chloride ion content will not change rapidly over a period of several days; therefore, the sampling frequency is adequate.
Amendment No. 82 i
A 5, i:
4.3-4 T_A9LE 4.3.1 PRIt'APV C0QLANT' SYSTEM PRESSUPE ISOLATION VALVES Maximum (a)
System
. Valve No.
Allowable Leakage Core Spray System 1 NZO2A 5.0 GPM NZ02C 5.0 GPM Core Spray System 2 NZO2B 5.0 GPM NZO2D 5.0 GPM v
Footnote:
(a)-
1.
Leakage rates less than or equal to 1.0 gpm are considered acceptable.
2.
Leakage rates greater than 1.0 ppm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
- 3. -Leakage rates greater than 1.0 gom but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an anount that reduces the margin between measured leakage rate and the maximum pennissible rate of 5.0 gpm by 50% or greater.
4.
Leakage rates greater than 5.0 gpm are considered unacceptable.
5.
Test differential pressure shall not be less than 150 psid.
NRC Order dated April 20, 1981 Amendment No. 82
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