ML20107B159

Requests Authorization to Conduct Low Power Analyses Tests W/Reactor Head Removed & Power Less than 5 Mwt

ML20107B159
Person / Time
Site:	Oyster Creek
Issue date:	05/15/1973
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML18039A986	List: IR 05000219/1970007 IR 05000219/1971001 IR 05000219/1971002 IR 05000219/1971003 IR 05000219/1971004 IR 05000219/1971005 IR 05000219/1972001 IR 05000219/1972002 IR 05000219/1972003 IR 05000219/1972004 IR 05000219/1972005 IR 05000219/1972013 IR 05000219/1972025 IR 05000219/1973003 IR 05000219/1973004 IR 05000219/1973006 IR 05000219/1973007 IR 05000219/1973008 IR 05000219/1973009 IR 05000219/1973010 IR 05000219/1973011 IR 05000219/1973012 IR 05000219/1973013 IR 05000219/1973014 IR 05000219/1973015 IR 05000219/1973016 IR 05000219/1973017 IR 05000219/1973019 IR 05000219/1973020 IR 05000219/1973021 IR 05000219/1973022 IR 05000219/1973023 IR 05000219/1974001 IR 05000219/1974002 IR 05000219/1974003 IR 05000219/1974004 IR 05000219/1974005 IR 05000219/1974006 IR 05000219/1974007 IR 05000219/1974008 IR 05000219/1974009 IR 05000219/1974010 IR 05000219/1974011 IR 05000219/1974012 IR 05000219/1974013 IR 05000219/1974014 IR 05000219/1974015 IR 05000219/1974016 IR 05000219/1974017 IR 05000219/1974018 ... further results
References
FOIA-95-258 NUDOCS 9604150398
Download: ML20107B159 (1)

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In ymW lettue to Jordcy Central Atted Aortl. 16, 1973, n o th o ei r.t r ic<.t to (md roo hem.'ot (not in the Oyater Creek R.: actor was Atauted, but in I.hu abdea.',a u( noce dc t:1t h'd litt'otr.ctrto't, not hoci mt ion to w: rat.u. the reactor vu vlt.hh61d. Ilu havn ucce.r.ulu ted t hu r!.'gefu ted infoi,? tr.ico nad 5

havu filed it.

llouever, in the r"cantlN, we requent authorir,ation to coneluct luv po'on phyoico teato with tha reacter br.td renoved und pcuw lesa r.han :i Wt.

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this authorization will pernit work to continua at tha plant whilo your utaf f reviews tho additicual inforention vc have t'iled.

The addttlonal informtlon you requested, has no bearing on operations at this lov pn ter levol ',ilth the,hnad cor.oved, nnd is only required to satisfy your concetus about operattou at the aorntL cperattug power levais near 1930 Wr..

74 Your continuod cooporation in this entter is appreciated.

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n Jersey Central Po.Weer & Light Company N\\VM[ \\l M AolsoN AVENUE AT PUNCH Bowl Ro Ao e MoRRisToWN, N.J. 07960 e 539 6111 May 15, 1973 Mr. A. Giambusso Deputy Director for Reactor Projects Directorate of Licensing United States Atomic Energy Commission Washington, D. C. 20515

Subject:

Oyster Creek Station Docket No. 50-219 Failure of Torus to Reactor Building Vacuum Relief Valve V-26-18 This event is considered to be an abnormal occurrence as defined in the Technical Specifics.tions, Paragraph 1.15.F..

Notifica-tion of this event, as required by the Technical Specifications, Para-graph 6.6.B was made t6 AEC Region I, Directorate of Regulatory Opera-tions, on Friday, May 4, 1973.

hhile performing local leak rate testing, it was found that one of the torus to reactor building vacuum breaker lines would not hold pressure between the isolation valves (V-26-17 6 18). The lack of air leakage through check valve (V-26-17) indicated that the leakage was through the butterfly valve (V-26-18).

Details of the valve are as follows:

Air Operated Butterfly, Rockwell Si c - 20" Rating - 150 psi Operator - Conoflow Corp.

Inspection of V-26-18 showed that the valve was 0.010 inches off the seat, indicating that the linkage on the valve arm required adjustment.

The valve was inspected.nnd the boot seat and butterfly disc were found in good condition. The valve was found to be 0.010 inches off the scat. The boot seat and butterfly disc were cleaned and the valve b

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linkage was adjusted to position the valve disc properly on:the seat..

The adjustment consisted of increasing the stroke in the close direction.

The line between the isolation valves was tnen pressurized and the leakage was calculated to be 0.492 SCFH well within the technical specifi cation limit. The cause of the change in linkage adjustment has not been determined..

There was no safety significance associated with this occurrene since the redundant component (check valve V-26-17) was shown to be

- Icakage tight. This is implied by the fact that the check valve was not moved during the testing period, so that the maximum possible leakage through the check valve was 0.492 SCFH.

To prevent a reoccurrence of this type problem, indicating mark uere placed on the disc shaft.

Several operating tests will be performed prior to plant startup to verify repeatability of linkage and valve position based on these markings.

Also, additional leak rate measurement will be perfo' med prior to plant startup to verify the adequacy of re-r lying on these new markings to ensure the valve has closed properly following future operability surveillance tests.

Very truly yours,

N x

Donald A. Ross Manager, Nuclear Generating Stations Cs Enclosures (40)

Mr. J. P. O'Reilly, Director

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Directorate of Regulatory Operations, Region 1 4

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