05000237/LER-1983-018, Forwards LER 83-018/01T-0.Detailed Event Analysis Encl

Forwards LER 83-018/01T-0.Detailed Event Analysis Encl

ML20072Q312
Person / Time
Site:	Dresden
Issue date:	03/28/1983
From:	Scott D COMMONWEALTH EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20072Q315	List: 05000237/LER-1983-018, Forwards LER 83-018/01T-0.Detailed Event Analysis Encl 05000237/LER-1983-018-01, /01T-0:on 830315,during Local Leak Rate Testing, 763 Std Cubic Ft Per Hour Leakage Measured Through Valve A0-1601-24.Caused by Valve Seat Surface Aging.Valve Replaced & Vol Retested
References
83-335, NUDOCS 8304040528
Download: ML20072Q312 (3)

LER-1983-018, Forwards LER 83-018/01T-0.Detailed Event Analysis Encl

Event date:
Report date:
2371983018R00 - NRC Website

text

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• O Dresden Nucleir Power St Commonwe::lth Edison R.R. #1 Morris, Illinois 60450 Telephone 815/942-2920 March 28, 1983 DJS Ltr. #83-335 James G. Keppler, Regional Administrator Region III i

U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 Reportable Occurrence Report #83-18/0lT-0, Docket #050-237 is being submitted to your office in accordance with Dresden Nuclear Power Station Technical Specification 6.6.B.l. (c), abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

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D. J. Scott Station Superintendent Dresden Nuclear Power Station DJS/nh Enclosure cc/ Director of Inspection & Enforcement Director of Management Information & Program Control U.S. NRC, Document Management Branch File /NRC 8304040528 830328 MAR 2 91983 PDR ADOCK 05000237

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STA UNIT YEAR NO.

l TITLE OF CEVI ATION OCCURRED PART 1 3-15-81

'.i nn Excessive Leakage Thru Valve A0-2-1601-24 om v,

SYSTEM AFFECTED 1600 PLANT STATUS AT TIME CF EVENT TESTING Primary Isolation Moct Refuel

, PWR(MNT) 0 LOAD (MWE) 0 y 3 OESCRIPTION OF EVENT While nerforming a local leak rate test between valves 2-1601 61, 62, 63, 23 and 24, leakage in excess of 763 SCFH was measured.

All leakane was determined to be thru leakage passing through valve A0-2-1601-24.

The tech spec limit is 493.116 SCFH total thru leakage for all penetrations combined.

10 CFR50.72 NRC RED PHCNE (Section 3.7.A.2.b.(2)(a))

NOTIFICATION MADE YES NO EOUlPMENT FAILURE 26471 OVES NO 40RK REQUEST NO.

RESPONSIBLE SUPERVISOR Robert Stachniak OATE 3-16-83 P ant 2 l cPERATING ENGINEER'S COMMENTS The A0-2-1601-24 vnive will be renaired and retested to verf #y confennce to tech spec leakage limits prior to start up.

EVENT OF PUBLIC INTEREST 24 HOUR NRC NOTIFICATION REQ'O TELEPH TECH. SPEC. VIOLATION REGION lli DATE TIME D NON REPORTABLE OCCURRENCE TELEGM/TELECOPY

@ 14 DAY REPORTABLE /T.S.6.6.3.1.C REGION lil DATE TIME O 30 DAY REPORTABLE /T.S.

CECO CORPORATE NOTIFICATION MADE ANNUAL /SPECL REPORT REQ'O IF ABOVE NOTIFICATION IS PER 10CFR21 A.I.R.

• 12-82-13 S-DAY #RI TTEN REPORT RE0'O PER 10CFR21 L.E.R. = R3-1R/nTT-0 Telecopy Dennis P. Galle 1 16 A'

CECO CORPORATE OFFICER DATE TIME A

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ATTACHMENT To LICENSEE EVENT REPORT #83-18/01T-0 COM40NWEALTH EDISON COMPANY (CWE)

DRESDEN UNIT 2 (ILDRS-2)

DOCKET #050-237 During the scheduled refueling outage, while local leak rate testing, leakage in excess of 9,975 SCFH was measured from the piping volume bounded by valves A0-2-1601-23, 24, 60, 61, 62 & 63.

During subsequent leak rate tests conducted on the line, valves A0-2-1601-23, 60 & 61 were determined to be the cause of leakage and were replaced. Valves A0-2-1601-23, 60, 61 & 62 constitute the first primary containment isolation boundary and valves A0-2-1601-24 & 63 constitute the second isolation boundary.

Since all leakage was determined to have passed through only the first isolation boundary, none of the initial leakage measured was determined to be through leakage.

When the valves were repaired, the line was retested and leakage in excess of 763 SCFH was measured.

All measured leakage was determined to be passing through valve A0-2-1601-24.

Since the first primary isolation boundary had initially been leaking in excess of 9,975 SCFH, all 763 SCFH passing through valve 1601-24 was determined to be through leakage.

The Tech. Spec. limit (3.7. A.2.b. (2)(a)) for all combined leakage is 493.116 SCFH.

There was minimal effect on public health and safety since continuous moni-toring of the Reactor Building ventilation showed no abnormal releases during operation. The last occurrence of this type was reported by R.O.

82-04 on Docket #50-249.

Aging of the valve seating material is believed to be the cause of failure.

The valve was replaced with a different type design valve.containing replaceable seats. After replacement, the volume was successfully -retested.

The station will conduct an investigation to verify the failure mechanism and determine a solution to prevent recurrence.

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