ML19338G568

Feedwater Nozzle Cracking,Brunswick Steam Electric Plant Unit 2, Technical Evaluation Rept

ML19338G568
Person / Time
Site:	Brunswick
Issue date:	10/09/1980
From:	Prior J FRANKLIN INSTITUTE
To:
Shared Package
ML19338G553	List: ML19338G566 ML19338G568
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-078, TER-C5257-78, NUDOCS 8010310179
Download: ML19338G568 (4)

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TECHNICAL EVALUATION REPORT FEEDWATER N0ZZLE CRACKING: BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 FRC TASK NO. 78 I NRC TAC NO. C8840 Prepared by: J. E. Prior Performing Organization i Franklin Research Center The Parkway at Twentieth Street FRC Project No.

Philadelphia, PA 19103 C5257 Sponsoring Agency Nuclear Regulatory Co= mission NRC Contract No.

j Washington, D.C. 20555 KRC-03-79-118 1

i This report was prepared as an account of work sponsored by an agency of the '.'nited States Govern =ent. Neither the United States Government nor any agency thereof, or any of their employees, =akes any warranty, expressed or i= plied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process dis-closed in this report, or represents that its use by such third party would not infringe privately owned rights.

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s i TECHNICAL EVALUATION REPORT U'IT: BRUNSWICK STEAM ELECTRIC LICENSEE: CAROLINA POWER PLANT, UNIT N0. 2 AND LIGHT COMPANY DOCKET NO. 50-324 TAC N0. 08E40 l

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SUMMARY

The planned inspection progras for the feedwater and control rod drive return line nozzles during the 1979 fuel outage is acceptable. Accessible

! portions of all four feedwater nozzles were dye-penetrant (?T) inspected j

with the spargers in place. Two small crack indications were found and were a

ground out. The CRD return line nozzle was PT examined with the ther=al sleeve renoved. 50 indications were noted on either the no:zle or the thermal sleeve. It was suggested that inspe: tion programs be developed to monitor two areas of concern in the CRD return line nozzle assembly.

2. INTRODUCTION i In a letter (E. E. Utley to T. A. Ippolito, dated October 24, 1978), a i

su= mary of the proposed inspection program for the feedwater nozzles and the ERD return line nozzle was submitted to the NRC for review. Subsequently, in a letter (E. E. Utley to T. A. Ippolito, dated June 14.-1979), the results of the inspection were outlined. The object of this review is to ensure that the ac: ions taken by the Licensee will be technically adequate. The progra:

i was evaluated according to the guidelines established in the applicable docu-men:, NUREG 0312. The review was confined to the issue of the pressure boundary integrity as influenced by nozzle cracking.

3. BACKGROUND i

Field experience has shown that f atigue cracks can be expected in feed-water nozzles and CRD return line nozzles in 3 oiling Water Reactor (3*.*R) pressure

.efgg, vessels. Observations, theoretical explanations, and recc= mended remedial

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l measures are discussed in NCREG 0312.

The principal factors responsible for no::le crack initiation and growth are thermal cycling and the stresses f rom the dif f erential ther=al expansion l

l between :he weld-deposited stainless steel cladding and the forged, icw-alloy steel nozzle.

[ 4. TECHNICAL EVALUATION i

4.1 FEEDWATER N0ZZLES i

lt was originally planned to re=ove one sparger and ?! inspect the entire nozzle. Accessible areas of the three remaining nozzles were also scheduled 1

for PT examination. Because extreme difficulty was encountered in attemp:ing to remove the sparger, the accessible areas of all four nozzles were PT exam-ined with the spargers in place. This examination was done with the concur-i rence of the NRC staff. Two superficial indications, identified as an arc-strike indication on the 65 nozzle and a weld splat:er on the 225 nozzle, were found and ground out by removing less than 1/16 inch of cladding.

4.2 CONTROL ROD DRIVE RETURN LINE N0ZZLE i.

1 A PT inspection was performed on the nozzle with the thermal sleeve recoved. No indications were found. A PT inspection was also perforced on the thermal sleeve with nc indications showing. The return line was ef fec-1 tively valved out, eli=inating the cold flow and alleviating the thernal cycling. The thermal sleeve was not replaced. Two cissing washers were t ot i

found following these actions, but it has been determined that these co=po-nents will not interfere with nor=al operation. The inspection progra for the CRD return line nozzle is considered to be acceptable with two areas of concern:

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l. Cracking has been observed in some 3WR reactors in the apron area directly below the no::le a year af ter the return line was valved out. Therefore, an appropriate inspection program must be implemented during the next regueling outage to ensure that cracks have not develcpt< which =ight impair the safety of the reactor.

2. Because of the suse ,p '.t Js . of stay. ant stainless steel lines

' to stress corrosic, cane an appropriate inspec:icn program

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=ust be develeped te exa= toe the stainless steel welds on tha reactor vessel side of the valve used for isolation.

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5. CONCLUSION i"

The inspection program for the feedwater nozzles and the CRD return line nozzle is acceptable. Modification of the CRD return line by valving out the line ef fectively eliminated the cold flow and subsequent thermal cycling.

l However, concern still exists for possible cracking in the apron area below 1

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the nozzle. An adequate inspection progras cust be developed to ensure that

!- cracting does not occur in this area. An inspection program also =ust be developed to monitor the possible initiation of stress corrosion cracking in the stagnant stainless s: eel return line.

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