ML18283A995

LER 1978-005-00 for Browns Ferry Nuclear Plant, Unit 2, Unidentified Coolant Leakage in Drywall Was Found to Be 9.5 Gpm & Exceeded 5 Gpm Limit of Technical Specification 3.6.C.1. During Normal Operation

ML18283A995
Person / Time
Site:	Browns Ferry
Issue date:	03/29/1978
From:	Fox H Tennessee Valley Authority
To:	O'Reilly J NRC/IE, NRC/RGN-II
References
LER 1978-005-00
Download: ML18283A995 (8)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

DISTRIBUTION FOR INCOMING MATERIAL 50-260 REC: NRC ORG: FOX H S DOCDATE:- 03/29l78 NRC TN VALLEY AUTH DATE RCVD: 04l03/78 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED

SUBJECT:

LTR i ENCL i LICENSEE EVENT REPT (RO 50-260/78-005) ON 03/i8/78 CONCERNING UNIDENTIFIED

'OOLANT LEAKAGE IN DRYWELL FOUND TO EXCEED TECH SPEC 3. 6. C. i... W/ENCL.

PLANT NAME:BROWNS FERRY UNIT 2 REVIEWER INITIAL: X JM DISTRIBUTOR INITIAL:

DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS INCIDENT REPORTS (DISTRIBUTION CODE A002)

FOR ACTION: BR CH AR+> 4 ENCL INTERNAL: REG FILE++M/ENC NRC PDR++W/ENCL MIPC++W/3 ENCL SCHROEDER/IPPOLI TO++M/ENCL HOUSTON44W/ENCL NOVAK/CHECK++W/ENCL I GR MES< 4W/ENCL KNIGHT++W/ENCL BUTLER+4W/ENCL HANAUER44W/ENCL TEDESCO~~W/ENCL E ISENHUT+~W/ENCL BAER++W/ENCL SHA0~4W/ENCL VOLLMER/BUNCH4+W/ENCL KREGER/J., COLLINS++W/ENCL ROSA+~W/ENCL K SEYFRIT/IE++W/ENCL 1 EXTERNAL: LPDR S ATHENS'L+~W/ENCL T I C+4 W/ENCL NSIC++W/ENCL ACRS CA'f B+<W/16 ENCL PER COPIES gOT SUBMITTED REGULATORY GUIDE 10.1 DISTRIBUTION: LTR 45 ENCL 45 CONTROL NBR: 780950029 SIZE: i P+ iP+ i P THE END

i' 7" NESSEE VALLEY AUTHORIT CHATTANOOGA. TENNESSEE 37401 MAR 39 1978 Nr. James P. O'Railly, Mrector U.S, Nuclear Ra latory Comnission ~

4

~

Cflt Office of Insp ction and Enforcement r Region II gC.

230 Peachtreo Street, NW., Suite 1217 Atlanta, Geo gia 30303 Deax'fr. 0'illy:

TENNESSEE V AUTHORITY BROHNS PERRY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-260 - PACILITY OPERATINC LICENSE DPR-52 - REPORTABLE OCCURRENCE REPORT BPRO-50-260/785 The enclosed report provides details concerning unidentified coolant leakage in tha dry+all which was found to bo 9.5 gpa and exceeded tha 5 gpm limit of Technical Specification 3.6.C.1. duxing norxml operation.

This report is submitted in accordance ~~ith Browne Perry unit 2 Technical Specifications, Section 6.7.2.B.(2) .

'ary truly youxs, TENNESSEE VALLEY AUTHORITY H. S. Pox Director of Power Production Enclosure cc (Enclosure):

Director (3)

Office of Hanagament Xnformagion and Progran Control U.S. Nuclear Regulatory C ssion Washington, DC 20555 Director (40)

Office of Inspection and Enfor ment.

U.S. Nuclear Regulatory Commiss on Washington, DC 20555 7SOWSOO29 An Equal Opportunity Employer

1 ll hf I~ /lt II fl O'Pi h P v'.Vfil ) 4 Sf'l" )) dhh ~ f II f -ff)Idden.y-fq>> y ~I f;, g S,;)

hfdf If i~ 1 - 'f):

.)<i~dr)<<S) g<<-Pg h,",

f: P 1 dd ~,' h Z ~

~

g~ l'f f fdi'I lf I r S, gl lh lj'lh ) R ff)>h l,h

'I' ff I

~ f ); "

l '

'~,4,

~ l (~

II fhfl g,) g gf hf'I I r)hfh LVSIr h

'f $ "g f "V ~> fi" hhSV' gpss,fy ~ ') p~ ,) ~ ve,f.- P )gag )~g f'<<rS: I f rf] g~ J hfj,<<h f S) ) Vl)f I)

J pl'"' fI'O'I, f *] I, f>>p 4, j d Cl jh

~ fB l'9)'LQ) t,Iif ~ I g'p>f f g, g gled

') ff r) ~ )T ~did) Il

) LSC + lf f 'S) I'S) ')S) 'h'd f ')' $ g

~ dfgy il)) Stir )')$ L '~'l",'I " 't&v"d)

.-'.:,l> ",t""")ls..').l

)de SP rh p p 41Pp d ) \ vf, f, V

')9",<"X ', 'I'.'I .h 'N),,) ~'<i' "P )l)i!6"')H "

~ lf if ~

17 $ P

(". > .gg + ad,~ <<

'I

)Qf fhl p) jgfl()f)f I ff )*)fgpl" f ))grfhh'S,d Span,,jfg'f4(d i

,,(Ilkf'<<5 ) I f cf) 3 <> )SW>)r'k>~'9 ~4<)) I f ~

("C~) '.~-aO>t)'V'

'hf) g()) jA'A 4 'rg<h )Xd,fd)IS 4r VV) <<hfh .3Q II

~ "444)SI BFA85 10/5/77 NRC FORM 366 LL S. NUCLEAR REGULATORY COMMISSION fl ill E./(HI 0 IT A LICENSEE EVENT REPORT CONTROL OLOCKI Ql (PLEASE PRINT OA TYPE ALLREQUIRED INFOAMATIONf

,6 Pap R 0 0 0 0 0 0 - 0 0 E 4 1 1 1 1QE~QL 9 LICKNSIK CODE 14 15 LICENSE NIJMGKR 26 LICENSE TYPK 40 Sl CAI SS CO'N'T

~

~O RKPDIIT

~g 0 5 0 0 6 0 8 0 32 8 78 Qo 7 60 61 DOCKET NUMBER IJ4 67 EVENT OATS 14 l5 REPORT DA'IE EVENT DESCIIIPTION AND PR08AOLE CONSEQUENCES Qlo Durin normal o eration unidentified coolant leaka e in the dr ell was I

~o 3 found to be 9.5 m exceedin the 5 limit of Technical S ecification

~o 6 of the reactor was be within 24 hours. There were no effects u on

~OG ublic health or safet redundanc was not a licable. A revious

~8

~o 7 occurrence was re orted on BFRO-50<<296/7714. (See attached).

~OG 50 7 4 9

~i8 LJS SYST'M CAUSE CAUSE COMP. VALVf COMPONE'NT CODK

~

CODE CODE SUSCOOK SUPCOOK SUSCOOK

~O9 9 9 10 'I l LJS 12 12 16 L'JO 'WS 19 20 7

SEOUENTIAL, OCCURRENCE AEPCRT REVISION

~If ~ ~

~8 EVENT YEAR RKPOAT NO. COO E TYPE NO.

LfR/RO QEE RE IOII I

~00 84 5

26

~W 27

~01 29 29

~T 20 Sl

~0 22 21 22 22 ACTION fUTURK EFFECT SHUTOOY/N ATTACHMENT KPR~ PRIME COMP. COMPONENT TAKEN ACTIC'I ON PLANT METHOD HOURS +22 SUBLUTTEO FO>MSU0. SUPPLIER EEEEANUFACTURKR LJSLJS 24 LEES 2'5 LJS 26 Sl 40 LJS LJO LJS 41 42 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS Q27

~10 e occu red f m the cki f fl c ntr 1 valve FCV-68-33 22&nch Darlin Valve Co. No. S150IR)D'ocated on the recirculation loo cross-i

~8~

on the valve "be erf rmed 4

7 ~ 9 FACILITY METHOD

'%POWER OTHER S'TATUS QSD OISCOVEAYOI'TATUS DISCOVERY OESCAIPTION Q32

'E00 LJO 7 9 9 10 12 17 44 LJS 45 46 ACTIVITY C ONTENT 7

6 9

~

RKLEASKD OF RELfASK 9

Q33 ~gQ3 10 PEASONNfL EXPOSURES I)

AMOUNTOF ACTIVITYQ36 A

41 LOCA'SION OF AELEASE Q3 NUMSER TYPE DESCRIPT)ON Q38 EE 7 9 LOOOOJEQL~JS 9 11 12 12

~ ERSONNEI. INJURIES EEI 4

~oo 8 9

NUMBER 11 OESCRIPTEONQ41 12 N A LOSS OF OA DAMAGE TO FACILITY TYPE DESCRIPTION Q43 9 ~ZQ42 7 6 9 10 O

PUSLI CITY ISSUED DKSCREPTION ~ NAC USE ONLY 7 4 9 I lp 66 69 NAME OF PREPARER PHONE:

• Revision

0 BFRO-50-260/785 Report Number  : BFRO-50-260/785 Report Date  : March 28, 1978 Occurrence Date: 'arch 18, 1978 Facility Browns Ferry Nuclear Plant Unit 2 Identification of Occurrence Drywell unidentified leakage exceeded Technical Specification value of 5 gpm by 4.5 gpm.

Conditions Prior to Occurrence Unit at approximately 90-percent power.

Descri tion of Occurrence Drywell unidentified leakage was determined to exceed 5 gpm during surveillance testing.

Desi ation of A arent Cause of Occurrence Packing leak from reactor recirculation loop crosstie valve FCV-68-33.

Anal sis of Occurrence The unit was removed from service, and an inspection of the drywell revealed the leakage from FCV-68-33 as the major contributor to the unidentified leakage. No effects existed upon the public health or safety; redundant systems are not applicable.

Corrective Action The unit remained shutdown to begin the refuel outage. FCV-68-33 will be repacked with a different type packing.

JGD:SGS 3/27/78