ML19291B946

Tech Spec Change Request 76 Re Surveillance Requirements in Section 4.3 Re Revised Inservice Insp Program.Certificate of Svc Encl

ML19291B946
Person / Time
Site:	Oyster Creek
Issue date:	12/11/1979
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML19291B945	List: ML19291B944 ML19291B946
References
NUDOCS 7912140419
Download: ML19291B946 (2)

Text

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JERSU.Y CENTRAL POWER 6 LIGilT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION Provisional Operating License No. DPR-16 Technical Specification Change Request No. 76 Docket No. 50-219 Applicant submits, by this Technical Specification Change Request No. 76 to the Oyster Creek Nuclear Generating Station Technical Specifications, changes to the surveillance requirements in Section 4.3.

JERSEY CENTRAL POWER 6 LIGHT COMPANY BY: A f Vice P ide STATE OF NEW JERSEY COUNTY OF MORRIS Sworn and subscribed to before me this .- / day of - <. , 1979.

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/, a- /u Notary Public

, m. , a :s. !D.::S bcTARY pbtu:: of htw Jctsty lAt Ccmra:aion bpires Aug l$,1934

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UNITED STATES OF AMERICA NUCLEAR REGULATORY C050!ISSION IN THE MATTER OF )

) DOCKET NO. 50-219 JERSEY CENTRAL POWER 4 LIGHT COMPANT)

CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 76 for the Oyster Creek Nuclear Generating Station Technical Specifications, filed with the U. S. Nucicar Regulatory Commission on December 11, 1979 has this lith day of December, 1979 been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail addressed as follows:

The Honorable Marylou Smith Mayor of Lacey Township P. O. Box 47S Forked River, New Jersey 08731 JERSEY CENTRAL POWER 6 LIGHT COMPANY BY: Off ,

Vice Tes{ent DATED: December 11, 1979 if09 2

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Jersey Central Power & Light Cornpany Cf C' b, p" Madison Avenue at Punch Bow! Road Morratown New Jersey 07960 (201)455-8200 December 11, 1979 i

The Honorabic bhrylou Smith Mayor of Lacey Township P. O. Box 47S Forked River, New Jersey 08731

Dear bhyor Smith:

Enclosed herewith is one copy of Technical Specification Change Request No. 76 for the Oyster Creek Nuc1 car Generating Station Technical Specifications.

These documents were filed with the U. S. Nuclear Regulatory Commission on December 11, 1979.

Very truly yours, Y

Ivan R. Fin ock Jr.

Vice President la Enclosure ,

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Jerse' Central Power & Licht Compan is a Member of the General P . 4 ft<!:t,e S m

JERSEY CENTRAL POWER 6 LIGHT COMPANY OYSTER CPIEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE DPR-16 (DOCKET NO. 50-219).

Applicant hereby requests the Commission to change Appendix A to the above captioned license as follows:

1. Section to be changed:

Section 4.3

2. Extent of Change:

Change Section 4.3 to remove inservice inspection requirements from this section.

3. Changes Requested:

Replace page 4.3-1 with the attached page 4.3-1.

Replace page 4.3-2 with the attached page 4.3-2.

Eliminate page 4.3-3 through page 4.3-8.

4. Discussion: l In accordance with 10 CFR 50.55a, paragraph (g), a revised inservice inspection program which will be separate from the Oyster Creek Station Technical Specifications will be implemented on December 8, 1979. It is therefore necessary to remove inservice '

inspection requirements from Section 4.3 of the Technical Specifications, i

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" 3-1

.: . 3 REACTOR COOLA.T N

Applicability: Applies to the surveillance requirements for the reactor l coolant system.

Objective: To determine the condition of the reactor coolant system and the operation of the safety ' devices related to it.

Specification: A. Neutron flux monitors shall be installed in the reactor vessel adjacent 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 from Figure 3.3.1. -

B. Full visual inspections of the accessible surfaces and welds of both core spray spargers and the repair assembly on core spray sparger no.2 shall be carried out during each of the next five refueling outages beginning in 1979, subsequent inspections will be conducted at 5 year intervals.

C. 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.

The requirements of specification 3.3. A shall be met during the test.

D. Each replacement safety valve or valve that has been repaired shall be bench checked for the proper set point. A mirimum 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, to insure set i points are as follows I

Number of Valves Set Point (psig)  ;

t 4 1212 + 12 ,

4 1221 + 12  ;

4 1230[12  ;

4 1239 .+._12 l

E. 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 con-tent of chloride ion and to check the conductivity.

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-. . . 4.3-2 BASIS: Numerous data are available relating inte' grated flux and the change in Nil-Ductility Transition Temperature (NDTT) in various steels. The base metal 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 can be determined. Since no shift will occur until an integrated flux of 10 17 nyt is reached, the confirmation can be made long before an NDTT shift would occur.

Extensive visual inspection for leaks will be made on critical systems. The inspection period is based on the observed rate of growth of defects from fatigue studies sponsored by the AEC. These stuides show that it requires thousands of stress cycles, at stresses beyond any conceived in a reactor system to propagate a crack and it is thus concluded that the frequency is adequate.

Experience i.n 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 l be used to determine 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. i i

1 1501 340 I

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