IR 05000296/1978021

Supplementary Reportable Occurrence Rept#BFRO-50-296/7821 on 780817.Following Reactor Scram,Cooldown Rate Was Exceeded Due to Safety Valves Actuating & Not Completely Closing. Caused by Excessive Pilot Leakage

ML18024A523
Person / Time
Site:	Browns Ferry
Issue date:	11/08/1978
From:	Fox H TENNESSEE VALLEY AUTHORITY
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
NUDOCS 7811140168
Download: ML18024A523 (6)

Text

REGULATORY FORMATXON DXSTRXBUTION EM DOCKET NBR: 50-296 BROWNS FERR RECXPIENT:

O'REILLY J.P.ORXGXNATOR:

FOX H.S.COMPANY NAME: TN VALLEY AUTH SUBJECT: LTR SXZE: 3 ENCL DOC DATE: ACCESSION NBR: COPIES RECEXVED: Forwards Sup l Reportable Occurrence Re t SBFRO-SO-296 782 Following reactor scram, cooldown rate was exceeded due to s actuating 6 not completely closin.Caused b excessive ilot leaka e.DISTRIBUTION CODE!'002 DISTRIBUTION TITLE)INCIDENT REPORTS NOTARIZED NAME NR~nR I L MIPC L C SYSTEMS BR NOYAK/CHECK EEB AD FOR ENG PLANT SYSTEMS BR , HANAUFR AD FOR PLANT SYSTEMS AD FOR SYS II PROJ REACTOR SAFETY BR ENGINEERING BR KREGER/J.COLLINS (PQMEQ SYS BR E.Jordan LPDR NSIC I ACRS ENCL'M/g ENCL W/ENCL H/ENCL H/2 ENCL H/3 ENCL'H/ENCL'H/ENCL W/ENCL W/ENCL~/ENCL W/ENCL w/EhlCL W/ENCL W/ENCL 4l/ENCL W/ENCL 8/ENCL N/ENCL W/ENCL'N/ENCL 8/16 ENCL'OR ACTION k~~ORB 3 BC TOTAL NUMBER OF COPIES REQUIRED!I.TR ENCL ttoV l-G~~@NOTES:

November S, 1978 TENNESSEE VALLEY AUTHORITY CHATTANOOGA.

TENNESSEE 37401 Mr.Jam':s P.O'Reilly, Director U.S.Nuclear Regulatory Commission Office of Inspection and Enforcement Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303

Dear Yu.O'Reilly:

TENNESSEE VALLEY AUTHORITY-BROWNS FERRY NUCLEAR PIJQG'NIT 3-DOCKET NO.50-296-FACILITY OPERATING LICENSE DPR-68-REPORTABLE OCCURRENCE REPORT BFRO-50-296/7821 This refers to my letter dated August 30, 1978.Enclosed is supplemental reportable occurrence report BFRO-50-296/7821 which provides details concerning main steam relief valves which failed to reseat on two occasions after a unit scram, causing the reactor to exceed the cooldown rate.This report is submitted in accordance with Browns Ferry unit 3 technical specification 6.7,2.A(2)

.Very truly yours, TENNESSEE VALLEY AUTHORITY H.ST Fox Director of Power Production Enclosure (3)cc (Enclosure):

Director (3)Office of Management Info U,S.Nuclear Regulatory C Washington, D.C.20555 ation and Program Control mmission Director (40)Office of Inspection and Enforcement U.S.Nuclear Regulatory Commission Washington, D.C.20555 EQ gt P~I 7 I'!u~p egg An Equal Opportunity Employer

LER SUPPLEMENTAL INFORMATION BFRO-50-296/7821 Technical Specification Involved: 3.6.A.1 Reported under Technical Specification:

6.7.2.a.(2)

Date of Occurrence:

August 17, 1978 Time of Occurrence:

10:37 p.m.Unit: 3 Identification and Descri tion of Occurrence The cooldown rate specified in Technical Specification 3.6.A.l was exceeded following a reactor scram, and again 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 43 minutes later, due to safety relief valves actuating and not completely reclosing.

On the first occurrence, six valves actuated following the reactor scram.MSRV 1-41 failed to reseat causing an excessive cooldown rate.The second occurrence resulted when MSRV 1-41 reopened at a reactor pressure of approximately 840 1bs.During the blowdown, MSRV 1-30 opened at approximately 580 lbs.and MSRV 1-34 opened at approximately 300 lbs.This second occurrence also resulted in an excessive cooldown rate.In both cases, the torus tempera-ture reached approximately 118 F.Conditions Prior to First Occurrence Reactor at 940 MWe.A arent Cause of Occurrence Excessive pilot leakage.On all three valves that were replaced, steam cuts were found on the pilot seat and disc.On valve in position 1-30, the piston rings of the second stage were found to be worn.The second-stage disc and seat of all three indicated leakage but excessive wear or steam cutting was not present.The conditions found in these valves were similar to conditions found in other valves that have experienced similar failures.

r~r~-2-LER SUPPLEMENTAL INFORMATION BFRO-50-296/7821 Anal sis of Occurrence It was determined that the transients were less severe than the design blowdown transients.

For conservatism, the transients were considered as two complete blowdowns.

The fatigue usage factors are as follows: Feedwater nozzles Rx vessel at waterline Closure studs 0.00182 0 F 00007 0.00133 Corrective Action Valves 1-'41, 1-30, and 1-34 were replaced prior to return to service of the unit.The blowdowns were analyzed in accordance with standard heatup/cooldown analysis techniques and fatigue usage factors assigned.Long-range corrective action consists of increasing the simmer margin of the valves.Additionally, 2-stage topworks have been procured and will be installed subsequent to evaluation by the Division of Engineering Design.A recent revision to the technical specifi-cations which lowered the setpoint for the react r water level isolation of the main steam lines should minimize the operation of the safety relief valves.Failure Data 296/788 JGD: Jhll 10/4/78