View source for 05000254/LER-1976-002, Updated LER 76-002/03L Re Excessive Leakage from Double Gasketed Seal X-4.Initially Reported on 760202.Caused by Equipment Failure & Insufficient Compression.Hatch Bolts Tightened

{{Adams
| number = ML20084S487
| issue date = 03/03/1976
| title = Updated LER 76-002/03L Re Excessive Leakage from Double Gasketed Seal X-4.Initially Reported on 760202.Caused by Equipment Failure & Insufficient Compression.Hatch Bolts Tightened
| author name = Hooverson O, Kalivianakis N
| author affiliation = COMMONWEALTH EDISON CO.
| addressee name = Keppler J
| addressee affiliation = NRC OFFICE OF INSPECTION &amp; ENFORCEMENT (IE REGION III)
| docket = 05000254
| license number = 
| contact person = 
| document report number = LER-76-002-03L, LER-76-2-3L, NJK-76-81, NUDOCS 8306170057
| package number = ML20084S475
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 4
}}
{{LER
| Title = Updated LER 76-002/03L Re Excessive Leakage from Double Gasketed Seal X-4.Initially Reported on 760202.Caused by Equipment Failure & Insufficient Compression.Hatch Bolts Tightened
| Plant = 
| Reporting criterion = 
| Power level = 
| Mode = 
| Docket = 05000254
| LER year = 2076
| LER number = 2
| LER revision = 
| Event date = 
| Report date = 
| ENS = 
| abstract = 
}}

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March 3, 1976                                                                                                          l
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J. Keppler, Regional Director                                                                                    '
Office of Inspection and Enforcement                                                     -
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Glen Ellyn, Illinois 60137                                                       ,.
Re fe rence :   Quad-Cities Nuclear Power Station kD,d.
Nu-Docket No. 50-254, DPR-29, Uni t One Appendix A, Section 6.6.B.2.b Enclosed please find a supplement to Reportable Occurrence Report Number R0 50-254/76-2 for Quad-Ci ties Nuclear Power Station. This report is sub-mitted to you in accordance with the requirements of Technical Specification 6.6.B.2.
Following the completion of all initial local leak rate tests pertaining to primary containment isolation valves and testable penetrations on January 30, 1976, double gasketed s2a1 X-4 (drywell head access hatch) was found to have excessive leakage.
Initial reporting of RO 50-254/76-2 was submi tted to you on February 2, 1976.
Details of the X-4 leakage are described herein. A further supplemental report regarding the details of all repairs necessary to resolve the pre-viously reported items will be sent upon thei r completion.
Very truly yours, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION r Pr                 -
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N. J. Kalivianakis Station Superintendent NJK/L FG/l k 2489 cc:      G. A. Ab re ' '
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REPORT NUMBER:       RO 50-254/76-2 REPORT DATE: March 3, 1976 i
,          OCCURRENCE DATE:      February 2,1976 FACI LITY : quad-Cities Nuclear Power Station Cordova, Illinois 61242 l           lDENTIFICATION OF OCCURRENCE:
Excessive leakage measured from the Uni t One drywell head access hatch.
CONDITIONS PRIOR TO OCCURRENCE:
Unit One was in the REFUEL mode for a scheduled refueling outage.
DESCRIPTION OF OCCURRENCE:
At 2:30 p.m. on Feb ruary 2, 1976, a local leak rate test was performed on the drywell head access hatch, designated X-4. The measured leakage was 39 2 SCFH, which was in excess of the Technical Specification limit. During the performance of the test, the hatch exterior and associated fittings were tested for leakage with soap-bubble solution. No bubbles were observed
         -and it was postulated that the excessive leakage was through the inner seal.
I          Work Request number 375-76 was issued to repair the access hatch.
DESIGNATION OF APPARENT CAUSE OF OCCURRENCE:
Equipment Fai lu re The apparent cause of this occurrence is designated as equipment failure.
Upon opening the hatch, maintenance personnel determined that there was insufficient compression of the Inner gasket by the inner knife edge of the hatch door. The outer gasket appeared to be in satisfactory condition.
The re fo re , it was concluded that the leakage path was out through the inner gasket seal.
l          ANALYSIS OF OCCURRENCE:
Since the excessive leakage f rom the drywell head access hatch was determined to be only through the inner seal, the safety implications of- this occurrence are minimal . Primary containment was maintained through the outer seal of
;        . thi s penetration. The re fo re , reactor safety and the health and safety of the public were not affected by this ' occurrence.
t CORRECTIVE ACTION:
The drywell head access hatch bolts were tightened by maintenance -personnel following closure. A subsequent leak rate test on the hatch resulted in a leakage of 17 66 SCFH, which is within the 11'm itation given by the Technical Speci ficat ions .

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FAILURE DATA:
This is the fi rst occasion whereby the drywell head access hatch on either unl t failed to meet the acceptance criteria during local leak rate testing.
t'                    Therefore, there is no cumulative experlence related to this occurrence.
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