ML19344D056
| ML19344D056 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/08/1980 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| References | |
| IEB-80-04, IEB-80-4, TAC-46851, TAC-46852, NUDOCS 8003110150 | |
| Download: ML19344D056 (1) | |
Text
>' no UNITED STATES P
NUCLEAR REGULATORY COMMISSION IIC
^
n E
REGION lli b[
e 799 ROOSEVELT ROAD j
o GLEN ELLYN. ILLINOls 60137 Docket No. 50-266 Docket No. 50-301 Wisconsin Electric Power Company ATTN:
Mr. Sol Burstein Executive Vice President Power Plants 231 West Michigan Milwaukee, WI 53201 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,
\\, m S N g-'
James G. Keppler Director
Enclosure:
IE Bulletin No. 80-04 cc w/ encl:
Mr. G. A. Reed, Plant Manager Cet. ral Files Director, NRR/DPM Director, NRR/ DOR C. M. Tramtrell, ORB /NRR PDR Local PDR NSIC TIC Sandra A. Bast, Lakeshore Citizens for Safe Energy Mr. John J. Duffy, Chief Boiler Inspector, Department of Industry, Labor and Human Relations 800311oMo
UNITED STATES SSINS No.:
6820 NUCLEAR REGULATORY COMMISSION Accessions No.:
0FFICE OF INSPECTION AND EEORCE!1ENT 7910250504 WASHINGTON, D.C.
20555
]D " ]D PD' " i Jg[]g,1p' February 8, 1980 J.
ANALYSIS OF A PWR liAIN STEA!; LINE BREAK UITH CONTINUED FEEDUATER ADDITION Description of Circunstances:
Virginia Electric and Power Co. submitted a recort to the Nuclear Regulatory l
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.
i Stone and Webster Engineering Cornoration performed a reanalysis of centainment pressure following a main steam line break and detemined 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 tern 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.
i On January 30, 1980, fiaine Yankee Atomic Power Company inforned 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 oostulation that the startup feedwater control valves would remain positioned "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. leanalysis of the event shows the i
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:
71/0250EdV h
No. of es: