ML20084Q333
| ML20084Q333 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/29/1973 |
| From: | Stephenson B COMMONWEALTH EDISON CO. |
| To: | Oleary J US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20084Q329 | List: |
| References | |
| BBS-73-258, NUDOCS 8306130214 | |
| Download: ML20084Q333 (3) | |
Text
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g r rTT2 November 29, 1973 John F. O' Leary, Director Directorate of Licensing Regulation U. S. Atomic Energy Commission Washington, D. C. 20545
Reference:
Quad-Cities Nuclear Power Station, Unit One Docket No. 50-254, DPR-29 Appendix A, Sections 1.0.A.2, 3.6.D.1, and 6.6.B
Dear Mr. O' Leary:
The purpose of this letter is to inform you of the details concerning the abnormal occurrence which took place on November 20, 1973, whereby reactor coolant leakage into the primary containment exceeded five (5) gpm.
This ab-normal occurrence was previously reported to you by tele-phone and telegraph on November 20, 1973 PROBLEM AND INVESTIGATION On November 18, at about 1600, bank 2, point sixteen (16) on the Valve Leak Detection System temperature recorder, corresponding to MO l-0202-4A Recirculation pump 1A suction valve, was observed to be reading 260 F.
This relatively high temperature was duly noted and a work request written to check this valve for packing leakage when the next out-age occurred.
The Unit 1 Drywell Continuous Air Monitor (C. A.M) reading for the day was 13k c.p.m.
total for halo-gens and particulates.
At 0001 on November 19, 1973, the integrated flow reading for the Drywell Floor Drain Sump during the previous eight hour period was 900 gallcns.
This reading was substantially higher than the 110 gallons calculated for the 0800-1600 shift on the 18th.
Drywell Floor Drain Sump flow deter-minations, at an interval of every two hours, were begun at 0200.
Drywell Equipment Drain Sump readings, at the same interval, were begun at 0930 on the 19th.
The Drywell C. A.M. reading vias 15k c.p.m. for the 19th.
8306130214 740208 s
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Mr. J. F. O ' Le ary November 29, 1973 At 0830 on November 20, 1973, reactor coolant leakage into the primary containment was determined to be slightly in excess of five (5) pgm. (Tech Spec 3.6.D.1)
An orderly shutdown of Unit 1 reactor was initiated at that time from 722 MWe.
The Drywell C.A.M. reading on the 20th was 39k c.p.m.
Upon drywell entry at about 0400, the source of the excessive coolant leakage was confirmed to be the 202-4A valve packing.
In addition, the sight flow indicating glass in the valve stem leak-off line was found to be broken which explained why the leakage was detected in the floor drain sump.
Repairs were made by tightening the packing and repairing the broken sight glass.
A reactor start-up was initiated at 1650 on November 21, 1973 following the repairs.
EVALUATIONS AND CORRECTIVE ACTIONS SAFETY IMPLICATIONS Considering the magnitude and source of this leak, the safety implications from this abnormal occurrence are minimal.
The Limiting Condition for Operation of Technical Specification 3.6.D.1 restricts unidentified reactor coolant leakage to five gpm with the total reactor coolant leakage into the primary containment limited to a maximum of 25 gpm.
Total unidentified reactor coolant leakage into the primary con-tainment for this occurrence was determined to be about 5 3 gpm.
The leak detection systems proved to be effective in identifying this leak.
DETERMINATION OF CAUSE AND CORRECTIVE ACTION The 202-4A valve has a double, independently adjustable packing arrangement with a leak-off connection between the packing stages.
Failure of the first stage packing is not abnormal.
A shutdown commenced immediately and the reactor was in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by Spec-ification 3.6.D.3 The corrective action consisted of tightening the packing, with packing replacement being de-ferred until a scheduled outage when an attempt will be made to backseat the valve.
As stated above, the sight glass was also replaced.
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o Mr. J. F. O' Leary 3-November 29, 1973 EVALUATION OF CUMULATIVE EXPERIENCE FOR SAFETY IMPLICATIONS No previous occurrence of this type has been experienced at Quad-Cities Nuclear Power Station.
Very truly yours,
COMMONWEALTH EDISON COMPANY QUAD CITIES NUCLEAR POWER STATION Ashcp4vM s
B. B. Ste nson Station S erintendent BBS/lk cc:
Regional Director Directorate of Regulatory Operations-Region III h
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