ML19344D076
| ML19344D076 | |
| Person / Time | |
|---|---|
| Site: | Bailly |
| Issue date: | 02/14/1980 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Shorb E NORTHERN INDIANA PUBLIC SERVICE CO. |
| References | |
| NUDOCS 8003110186 | |
| Download: ML19344D076 (1) | |
Text
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b UNITED STATES fib E '
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'n NUCLEAR REGULATORY COMMISSION 9
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GLEN ELLYN ILLINOIS 60137 g4.....
Docket No. 50-367 I
- lNU Northern Indiana Public Service Company ATTN:
Mr. Eugene M. Shorb Senior Vice President 5265 Hohman Avenue Hammond, IN 46325 Gentlemen:
Our letter dated February 8, 1980, regarding IE Bulletin No. 80-04 was in error.
It should have stated for informat 6i only.
No written response is required.
Sincerely,
~ bb %
~
ames G. Keppler Director
Enclosure:
IE Bulletin No. 80-04 cc w/ encl:
Central Files Director, NRR/DPM Director, NRR/ DOR PDR Local PDR NSIC TIC Mr. Dean Hansell, Office of Assistant Attorney General 800S110 k L
o 9
UNITED STATES SSINS No.
6820 NUCLEAR REGULATORY COMMISSION Accessions No.-
0FFICE 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 Nuciaar Regulatory Commission dated September 7, 1979 that identified a deficic in the origiral 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 line 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 shows the opening of the startup valve and associated high feedwater addition to the l
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 reac reactors listed in Enclosure 1:
1.
Review the containment pressure DUPLICATE DOCUMENT potential for containment overp Entire document previously entered into system under:
ANO W
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