ML19290E434
| ML19290E434 | |
| Person / Time | |
|---|---|
| Site: | Bailly |
| Issue date: | 02/08/1980 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Shorb E NORTHERN INDIANA PUBLIC SERVICE CO. |
| References | |
| NUDOCS 8003110134 | |
| Download: ML19290E434 (1) | |
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'o UNITED STATES h
I
~,j NUCLEAR REGULATORY COMMISSION REGION lil 3
O 799 ROOSEVELT ROAD o
GLEN ELLYN, ILLINOIS 60137 FEB ebgg Docket No. 50-367 Northern Indiana Public Service Company ATTN:
Mr. Eugene M. Shorb Senior Vice President 5265 Hohman Avenue Hammond, IN 46325 Gentlemen:
The enclosed IE Bulletin No. 80-04, is forwarded for action. A written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, b)$t-W James G. Kepp er Director
Enclosure:
IE Bulletin No. 80-04 cc w/ enc 1:
Central Files Director, NRR/DPM Director, NRR/ DOR PDR Local PDR NSIC TIC Mr. Dean Hansell, Office of Assistant Attorney General 8002110 13'
d UtlITED STATES SSINS No.:
6820 NUCLEAR REGULATORY COMMISSION Accessions No.:
0FFICE OF INSPECTION AND Ei! FORCE!!ENT 7910250504 WASHINGTON, D.C.
20555 February 8, 1980 IE Bulletin No. 80-04 ANALYSIS OF A PWR MAIN STEAM LINE BREAK UITH CONTINUED FEED'IATER ADDITION Description of Circunstances:
Virginia Electric and Power Co. submitted a recort to the Nuclear Regulatory Comission dated September 7,1979 that identified a deficiency in the original analysis of containment pressurization as a result of reanalysis of stean line break for North Anna Power Station, Units 3 and 4.
Stone and Webster Engineering Corooration performed a reanalysis of centainment pressure following a main steam line break and determined that, if the auxiliary
% the steam feedwater system continued to supply feedwater ai. runout conditione generator that had experienced the steam line break, containment de.
oressure would be exceeded in approximately 10 minutes.
The long tern blowcown 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 infornation was provided to all holders of operating licenses and construction permits in IE Information tiotice 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 line break accident.
On January 30, 1980, flaine Yankee Atomic Power Company inforned the NRC of an error in the main steam line break analysis for the !1aine 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 oostulation that the startup feedwater control valves would remain oositioned "as is" during the transient.
In reality, the startup feedwater control valves will rano to 80% full open due to an override signal resulting from the low steam generator pressure reactor trip signal. Reanalysis of the event shows 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 pow reactors listed in Enclosure 1.
Review the containment p potential for containmen Entire document previously entered into system under:
- 71/0250 Eof h
No. of pages:
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