ML20030C457
| ML20030C457 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/11/1981 |
| From: | Papay L SOUTHERN CALIFORNIA EDISON CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| 10CFR-050.55E, 10CFR-50.55E, NUDOCS 8108260125 | |
| Download: ML20030C457 (4) | |
Text
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50.55(e) Report
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C*C Southem Califomia Edison Company g6 RO. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770
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- i3 572 1474 August 11, 1981 Mr. R. H. Engelken, Director
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Office of Inspection and Enforcement V
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U. S. Nuclear Regulatory Commission g (hU g
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- Suite 202, Walnut Creek Plaza co,-
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1990 North California Boulevard Walnut Creek, California 94596 es
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Dear Mr. Engelken:
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Subject:
Docket Nos. 50-361 and 50-362 San Onofre Nuclear Generating Station, Units 2 and 3 In a letter to your office dated July 15, 1981 we identified a condition which we considered reportable in accordance with 10CFR50.55(e).
The condition involves the Unit 2 steam generators.
Enclosed in accordance with 10CFR50.55(e) are twenty-five (25) copies of a Final Report entitled, " Final Report on Steam Generator Feedring Deformation, San Onofre Nuclear Gen-erating Station, Units 2 and 3."
If you have any questions regarding this report, we would be pleased to discuss them with you at your convenience.
Very truly yours, Enclosures cc:
Victor Stello (NRC, Director I&E)
TO'l R. J. Pate (NRC, San Onofre, Units 2 and 3) y sh 81082o0125 810811 PDR A00CK 05000361 S
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J FINAL REPORT ON STEAM GENERATOR FEEDRING DEFORMATION San Onofre Nuclear Generating Station, Units 2 and 3 i
INTRODUCTION This report is submitted pursuant to 10CFR50.55(e).
It describes a condition involving deformation of the feedring in one steam l
generator of Unit 2.
The deformation is presumed to have occurred during the secondary feedwater system waterhammer test performed l
per FSAR Paragraph 14.2.12.72t.3, as part of the pre-core hot functional test on March 30, 1981.
The damage, however, was not noted at that time but was discovered during a subsequent ateam generator inspection on July 14, 1981.
This report includes a description of the deficiency, an analysis of the safety implications of the condition, and a summary of the corrective actions being taken.
By letter dated July 15, 1981 Southern California Edison confirmed notification to the NRC of this condition which was considered reportable in accordance with 10CFR50. 55(e).
I BACKGROUND l
i The condition which is reported here was discovered subsequent to the completion of the pre-core hot functional test.
The water-hammer test on the secondary feedwater system was performed on l
March 30, 1981 and no anomalies were detected.
This test involved shutting off feedwater to one steam generator in order to demonstrate the absence of any significant waterhammer in the feedwater piping during reinitiation of feedwater following the exposure of the feed-l ring (sparger) to a steam environment.
The steam generator water i
level was lowered to below the feedring with the use of the secondary side blowdown.
The steam generator water level was maintained below the feedring for about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (no feedwater was introduced into the steam generator through the feedring during this period).
At the end of this period maximum auxiliary feedwater flow was manually initiated to refill the steam generator.
No significant noise or vibrations were noted during the test (other than a low volume " thud" which was attributable to check valve i
slamming) and visual inspection indicated the.t the integrity of the feedwater piping and supportt aad not been violated.
The hot func-tional tests continued for another week and no anomalies were noted.
The required feedwater flow rates were comparatively low during this final period of hot functional testing.
The inside of the steam generator was inspected on July 14, 1981 and the feedring damage was noted.
l.
FINAL REPORT ON STEAM GENERATOR FEEDRING DEFORMATION Page Two San Onofre Nuclear Generating Station, Units 2 and 3 1
j The feedwater (main and auxiliary) enters the steam generator through the Feedwater nozzle which contains a thermal sleeve to reduce the severity of thermal transients.
The feedwater then enters a distribution box and the flow splits into two directions as it enters the 12 inch diameter schedule 40 pipe feedring sec-tions.
The water is discharged into the steam generator through 76 "J" tubes made of 1-1/2 inch pipe and located at the top of the 12" horizontal feedring.
During the inspection this feedring was i
found to be partially collapsed starting a short distance from i
either side of the distribution box and extending to about 900 from the distribution box in both directions.
"J" tubes in the j
collapsed area were still capable of discharging fluid.
Flow paths existed through the partially collapsed area to channel feedwater to the undamaged portions of the feedring.
The upper support bracLat of the distribution box was bent and the bolts l
of the lower support were sheared, allowing the distribution box to move toward the steam generator wall and rotate slightly.
No l
damage to the steam generator pressure boundary was observed.
i DISCUSSION I
The following discussion is responsive to 10CFR50.55(e)(3).
l Description of Deficiency I
Analysis of the condition of the steam generator feedring and the j
waterhammer test conditions supports the conclusion that the cause of failure was a suddenly applied pressure differential which 4
occurred when the auxiliary feedwater flow was initiated.
The waterhammer test required the water level in the steam generator to be lowered to below the feedring.
With no feedwater entering the feedring for a period of about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the water content in j
the feedring was allowed to drain out through the clearance between the thermal sleeve and the distribution box.
At the time the auxiliary feedwater flow was initiated (i.e., at the end of the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> drain period) the feedring was in a steam environment internally and externally.
The steam generator pressure was about l
1000 psia and the temperature abour '45 F.
The rapidly initiated cold feedwater flow caused the steam inside the feedring to condense very rapidly.
This created a temporary pressure differential between the outside and the inside of the feedring.
Apparently, the internal depressurization was so rapid that the pressure could i
not be equalized through the "J" tubes.
l FINAL REPORT ON STEAM GENERATOR FEEDRING DEFORMATION Page 'Ihree San Onofre Nuclear Generating Station, Units 2 and 3 Analysis of Safety Implications Deformation of the feedring in the manner observed does not pose a safety problem.
The test, as conducted, clearly indicated the j
ability of the external piping to survive a very severe condition, well beyond expected operating conditions.
The feedring was, at all times, capable of admitting full auxiliary feed flow.
- Thus, the plant could be safely shut down and decay heat removed.
j l
During actual plant operations, the damage would have been evident
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at a relatively low power level.
Although auxiliary feedwater flow was not affected, in achieving the required main feedwater flow l
during power ascension an inability to maintain water level in the steam generator would be noted and the plant could be shut down safely.
Corrective Action l
The necessary corrective action to restore fu13 feedring function l
and assure maximum plant availability is still unoer review.
Al-l ternatives being considered are:
I l
1.
Replacing damaged portion of the feedring with increased vent area and strengthened supports.
l 2.
Replacing entire feedring with heavier schedule pipe.
3.
Reducing significantly the leakage around the thermal sleeve.
4.
Provide level alarm indication to advise operator when the steam' generator water l'evel is below'the feedring.
5.
Appropriate operator guidance will be added to plant operating procedures to control the manner of initiation of auxiliary feedwater input and to minimize the possibility of uncovering the feedring during operation.
I
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