ML19290F072
| ML19290F072 | |
| Person / Time | |
|---|---|
| Site: | Perkins, Cherokee  |
| Issue date: | 02/08/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Dail L DUKE POWER CO. |
| References | |
| NUDOCS 8003180049 | |
| Download: ML19290F072 (2) | |
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o, UNITED STATES y y, ([',g NUCLEAR REGULATORY COT.*f'ISSION 8
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In Reply Refer To:
FEB S 1980 RII:JP0 d 6 91,0 024921
,_50-493, 50 488 50-489, 50-490 Duke Power Company Attn:
L. C. Dail, Vice President Design Engineering P. O. Box 33189 Charlotte, NC 28242 Gentlemen:
The enclosed IE Bulletin No. 80-04, is forwarded to you for information.
No written response is required. If you desire additional information regard-ing this matter, please contact this office.
Sincerely, so"
/
James P. O Re lly Director
Enclosures:
1.
Most Recently Issued 4
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l UNITED STATES SSINS No.:
6820 NUCLEAR REGULATORY COMMISSION Accessions No..
OFFICE OF INSPECTION AND ENFORCEMENT 7910250504 WASHINGTON, D.C.
20555 February 8,1980 IE Bulletin No. 80-04 ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION Description of Circumstances:
Virginia Electric and Power Co. submitted a report to the Nuclear Regulatory Commission dated September 7, 1979 that identified a deficiency in the original analysis of containment pressurization as a result of reanalysis of steam line break for North Anna Power Station, Units 3 and 4.
Stone and Webster Engineering Corporation performed a reanalysis of containment pressure following a main steam line break and determined that, if the auxiliary feedwater system continued to supply feedwater at runout conditions to the steam generator that had experienced the steam line break, containment design pressure would be exceeded in approximately 10 minutes. The long term blowdown of the water supplied under runout conditions by the auxiliary feedwater system had not been considered in the earlier analysis.
On October 1, 1979, the foregoing information was provided to all holders of operating licenses and construction permits in IE Information Notice No. 79-24.
The Palisades facility did an accident analysis review pursuant to the information in the notice and discovered that with offsite power available, the condensate pumps would feed the affected generator at an excessive rate.
This excessive feed was not considered in the analysis for the steam lire break accident.
On January 30, 1980, Maine Yankee Atomic Power Company informed the NRC of an error in the main steam line break analysis for the Maine Yankee plant.
During a review of the main steam line break analysis, for zero or low power at the end of core life, the licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during the transient.
In reality, the startup feedwater control valves will ramp to 80% full open due to an override signal resulting from the low steam generator pressure reactor trip signal. Reanalysis of the event shews the opening of the startup valve and associated high feedwater addition to the affected steam generator would cause a rapid reactor cooldown and resultant return-to power, a condition outside the plant design basis.
Actions to be Taken by the Licensee:
For all pressurized water power react reacters listed in Enclosure 1:
DUPLICATE DOCUMENT 1.
Reviev the containment pressure i Potent 2 e1 for containment overpre Entire doctunent previously entered into system under:
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No. of pages: