05000277/LER-1982-036, Forwards LER 82-036/01T-0.Detailed Event Analysis Submitted

Forwards LER 82-036/01T-0.Detailed Event Analysis Submitted

ML20066J766
Person / Time
Site:	Peach Bottom
Issue date:	11/08/1982
From:	Cooney M PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20066J775	List: 05000277/LER-1982-036, Forwards LER 82-036/01T-0.Detailed Event Analysis Submitted
References
NUDOCS 8211290300
Download: ML20066J766 (3)

LER-2082-036, Forwards LER 82-036/01T-0.Detailed Event Analysis Submitted

Event date:
Report date:
2772082036R00 - NRC Website

text

. L PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 -

PHILADELPHI A. PA.19101 I 12151841 4000 November 8, 1981 Mr. R. C. Haynes, Administrator U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406

SUBJECT:

LICENSEE EVENT REPORT NARRATIVE DESCRIPTION

Dear Mr. Haynes:

The following occurrence was reported to Mr. R. Blough, Region I, United States Nuclear Regulatory Commission on October 24, 1982.

Reference:

Docket No. 50-277 Report No.: 2-82-36/lf Event Date: November 5, 1982 Report Date: October 24, 1982 Facility: Peach Bottom Atomic Power Station RD #1, Delta, PA 17314 This LER is submitted in accordance with the Peach Bottom l Technical Specification and a letter from D. G. Eisenhut, NRC, dated May 7, 1980.

Technical Specification

Reference:

Technical Specification 3.7.A.l.b requires that "at any time the nuclear system is pressurized above atmospheric pressure or work is being done which has the potential to drain the vessel, the pressure suppression pool water volume and temperature shall be i

maintained within the following limits except as specified in 3.7.A.2.

a. Minimum Water Volume - 122,900 ft.

I

b. Maximum Water Volume - 127,300 ft.

8211290300 821108 PDR ADOCK 05000277 S PDR 1

6 L Mr . R. C. Haynes Page 2 In the letter from Mr. Eisenhut, dated May 7, 1980, a staff position was stated as, " Future failures of a relief valve to close should be reported promptly to the NRC".

Description of the Event During start-up of Unit 2, following a maintenance shutdown on October 23, 1982, Unit 2 was being started up using normal start-up procedures. Primary coolant pressure was being increased, when at approximately 3:07 p.m. on October 24, with primary coolant pressure at 832 psig, the 71J relief valve opened.

Reactor power at the time was approximately 1%. Opening of the relief valve immediately reduced pressure resulting in swell of reactor water level which '; ripped the operating reactor feed pump. Before the reactor feed pump could be recovered, reactor level reduced to approximately zero inches and the reactor scrammed. The HPCI system was manually started, before it reached its automatic start level of minus 48 inches, to recover reactor level to normal. Minimum level reached during this transient was approximately minus 10 inches.

The relief valve remained open until primary coolant pressure reduced to 80 psig. At 3:58 p.m. on October 24, 1982. During this transient, the torus volume ir:reased less than 1% above the maximum Technical Specification limit of 127,300 cubic feet.

Following closure of the relief valve, steps were taken to pump down the torus to normal volume.

In accordance with the site emergency plan, an unusual event was declared and proper notifications were made. The unusual event was terminated when the relief valve reseated.

Probable Consequences of the Event During the above transient, all ECCS systems operated properly and all parameters responded as expected. No release of activity occurred during this event. Safety significance is therefore minimal.

Cause of the Event Unknown at this time. The relief value will be sent to a vendor to determine the cause of this spurious opening.

s  %

Mr. R. C. Haynes Page 3 Immediate Corrective Action The 71J relief valve was replaced. An inspection of the downstream piping was conducted and no discrepancies identified.

Preliminary data indicated that the 71J valve had operated numerous times. A further investigation of the data indicated that the relief valve did not cycle repeatedly, but actually remained open until approximately 80 psig. Fatigue analysis was performed by Bechtel and a determination made that the downstream piping was not overstressed. Tlie vacuum breaker on the downcomer from this valve and a second valve which had been operated manually during the transient were inspected. Both values showed some binding on the 1.inge pin such that normal spring pressure was insufficient to close the valve. The two vacuum breakern were replaced. Corrective actions have therefore been completed with the exception of our continued investigation into the original cause of-the valve opening.

Very truly yours, M.

bb

. Cooney S erintendent G neration Division - Nuclear cc: Mr. Victor Stello, Director Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. Norman M. Haller, Director R. Blough Office of Management & Program Analysis Site Inspector U.S. Nuclear Regulatory Commission P.O. Box 399 Washington, DC 20555 Delta, PA 17314-0399