ML20076F864

Telecopy Ro:On 830815,indications Identified on Welds DCS-1-7,DSRWC-1-3,DCS-1-8 & DSRWC-1-2.Weld Insp Results Encl

ML20076F864
Person / Time
Site:	Browns Ferry
Issue date:	08/15/1983
From:	Coffey J TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
NUDOCS 8308260180
Download: ML20076F864 (2)

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TRtRCOPY TO NRC b?,' l A T[' .~,

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- v i? Gli U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW g

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Suite 3100 Atlanta, Georgia 30303 BFRO-50-259/83049R2 Reported under Technical Specification 6.7.2.A.(3)

Telecooy Date 8/15/83 Time 1000 Date of Occurrence: 8/15/83 Time of occurrence: 1430 Unit 1 Technical Specification Involved: 3 6.0 Conditions Prior to Occurrence Unit 1 in refueling outage Unit 2 at 3139 MWt Unit 3 at 3081 MWt Identification and Descriotion of Occurrence Weld DCS-1-7 (core spray) has an intermittent indication on the pipe side of the weld that is 360-degrees in length and approximately 41% through wall depth. This is a 12-inch, 304 stainless steel, schedule 80 pipe. The cracking is between the manual isolation valve and the testable check valve (inboard containment isolation valve).

Weld DSRWC-1-3 (reactor water cleanup) has two indications that are approx-imately 5" long between 11:30 and 2:30 and 1-1/2" long between 8:30 and 9:30 with a through-wall depth of greater than 80%. This is a 6-inch, 304 stainless steel, schedule 304 pipe. The cracking is between the inboard and outboard containment isolation valves.

Weld DCS-1-8 (core spray) has two indications that are approximately 9" long between 12:30 and 3:30 and 3" long between 9:30 and 10:30 with a through-wall depth of approximately 30%. This is a 12-inch, 304 stainless steel, schedule 80 pipe. The cracking is on the reactor vessel side of the pipe-to-valve weld on valve HCV-75-27 (manual isolation valve).

Weld DSRWC-1-2 (reactor water cleanup) has a indication that is approximately 3" long between 12:00 and 1:00 with a throu6h -wall depth of greater than 50%. This is a 6-inch, 304 stainless steel, schedule 80 pipe.

The cracking is between the inboard and outboard containment isolation coo valves.

ES g Uodate Item c3n O To date the attached are our inspection results.

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1. Acting Power Plant Superintendent Browns Ferry Nuclear Plant

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ATTACHMENT - BFRO 50-254/89049R1 ,'

Total Inspection Inspection Inspection Welds with Under Reauired Comolete Satisfactorr Cracks Evaluation Core Spray 27 20 18 2 0 RWCU 5 2 0 2 0 RHR (Head Spray) 21 15 15 0 0 SATISFACTORY Core Spray DCS 1-6, 9, 14, 15,18, 16, 17, 5 DSCS 1-4, 5, 6, 8, 9, 10, 11, 12, 13, 3 RHR (Head Spray) DHS 1-2, 3 DSHS 1-1A, 3, 4, 5, 6, 7, 8, 9, to, 11, 12, 13, 14 UNDER EVALUATION Core Spray None RWCU NoneC 1-2 RHR (Head Spray) None This is an update of the event first reported previously on 8/11/83, and includes all the results through 1430 on 8/14/83 This information was provided verbally to Dave Verelli on 8/13 and 14.

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