05000278/LER-1983-006, Forwards LER 83-006/01T-0.Detailed Event Analysis Submitted

Forwards LER 83-006/01T-0.Detailed Event Analysis Submitted

ML20065A618
Person / Time
Site:	Peach Bottom
Issue date:	02/10/1983
From:	Cooney M PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20065A622	List: 05000278/LER-1983-006, Forwards LER 83-006/01T-0.Detailed Event Analysis Submitted 05000278/LER-1983-006-01, /01T-0:on 830131,during Local Leak Rate Testing of B Containment Atmosphere Dilution Nitrogen Injection Penetration,Determined Valve Leaking Excessively.Caused by Scored Seat & Foreign Matl.Bonnet,Stem & Plug Replaced
References
NUDOCS 8302220106
Download: ML20065A618 (3)

LER-1983-006, Forwards LER 83-006/01T-0.Detailed Event Analysis Submitted

Event date:
Report date:
2781983006R00 - NRC Website

text

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PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 (215)8414000 February 10, 1983 Docket Nos. 50-278 Mr.

R. C. Haynes, Administrator Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19046

SUBJECT:

Licensee Event Reoort Narrative Description

Dear Mr. Haynes:

The following occurrence was reported to Mr. A. R. Blough of Region I U.S. Nuclear Regulatory Commission on February 1, 1983.

Reference:

Docket No. 50-278 Report Number:

3-83-006/IT-0 Report Date:

February 10, 1983 Event Date:

January 31, 1983 Facility:

Peach Bottom Atomic Power Station RD 1, Delta, PA 17314 Technical Soecification Reference:

Technical Specification 3.7.D.2 requires that whenever a containment isolation valve becomes inoperable reactor power may continue provided that "at least one valve in each line having an inoperable valve shall be in the mode corresponding to an isolated condition."

10 CFR 50 Appendix J Section III.C.3 requires that the combined leakage for all penetrations and valves subject to Type B&C tests shall be less than 0.6 La.

8302220106 830210 PDR ADOCK 05000278 S

PDR g&

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Mr. R. C. Haynes Page 2

Descriotion of the Event:

During local leak rate testing of the B CAD nitrogen injection to torus penetration (N211 B), the test valve located between the inboard and outboard isolation valves was determined to be leaking in excess of 80 SCFH.

When this result was used to calculate total leakage, utilizing the conservative technique of ignoring the best series valve leakage for each penetration, the results exceeded the 0.6 La limit of 10 CFR 50 Appendix J.

Probable Consecuences of the Event:

The valve in question, a test connection, is an outboard containment isolation valve.

If the calculation were redone using the leakage rate of the inboard valve of this penetration (tested at 5000 sec/ min) the results would be only 41,000 sec/ min which is less than the 0.6 La limit of 71193 sec/ min.

In addition, threaded caps are normally used to seal the test connections between test cycles.

For the above cited reasons, it is believed that primary containment integrity was not lost and that the significance of this event is minimal.

Cause of the Event

The valve would not properly seat due to a build-up of foreign material and scored seating surfaces.

Immediate Corrective Action

Upon discovery of the excessive leakage, the in-series manual valve was closed.

Mr. R. C. Haynes Page 3

Corrective Action

The valve was disassembled, approximately one teaspoon of material (thought to be corrosion products from the carbon steel lines) was removed; the plug, stem and bonnet were replaced, the seats were lapped and the valve was satisfactorily retested on February 2, 1983.

Sincerely

{&f W M.

. Cooney O S

erintendent d neration Division-Nuclear cc:

Document Control Desk A.

R. Blough Site Inspector P.O. Box 399 Delta, PA 17314-0399