ML19338G566

Feedwater Nozzle Cracking,Brunswick Steam Electric Plant Unit 1, Technical Evaluation Rept

ML19338G566
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-C5257-077, TER-C5257-77, NUDOCS 8010310169
Download: ML19338G566 (4)

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TECHNICAL EVALUATION REPCRT FEECWATER N0ZZLE CRACKING: BRUNSWICK STEAM ELECTRIC PLANT UNIT 1 FRC TASK NO. 77 NRC TAC NO. 08E39 Prepared by: J. E. Prior Performing Organization Franklin Research Center The Parkway at Twentieth Street FRC Project No.

Philadelphia, PA 19103 C5257 Seonsoring Agency Nuclear Regulatory Co==ission NRC Contract No.

'a'ashingto n, D.C. 20555 NRC-03-79-118 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, =akes any warranty, expressed or 1: plied, or assumes any legal liability or responsibility f or any third party's use, or the results of such use, of any infor=ation, 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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1 TECHNICAL EVALUATION REPORT i

l UNIT: BRUNSWICK STEAM ELECTRIC LICENSEE: CAROLINA POWER

PLANT, UNIT N0. 1 AND LIGHT COMPANY DOCKET NO.50-32S TAC NO. 08839 I

l 1.

SUMMARY

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j During the 1979 refueling outage, the accessible areas of the feedwater j no:zles were dye-penetrant (PT) inspected. No cracks were found. The con-

] trol red drive (CRD) return line t.o::le was also PT inspected. Several cracks l vere found and ground out. The ther=al sleeve was cracked and was not

! replaced. The return line was valved out, eliminating the cold flow and sub-i

sequent ther=al cycling. It is urged that inspection programs be devised to f

monitor cracking in the apron area as well as to inspect the stainless steel return line for possible stress corrosion cracking.

j 2. INTRODUCTION z

in a letter (E. E. Utley to T. A. Ippolito, dated 0:tober 2',1978) a surnary of the proposed inspec ticn progra: for the feedwater no: les and the

! CRD return line no::le was submitted to the NRC for review. S ubs eq uen tly ,

6 l in a letter (E. E. Utley to T. A. Ippolito, dated June l' ,1979) the results of the inspection were reported. The object of this review is to ensure that the actions taken by the Licensee will be technically adequate. Accordingly, the program was evaluated following the guidelines established in the appli-cable document, NUREG 0312, confining the review to the issue of the pressure l boundary integrity as influenced by nozzle cracking.

3. BACKGROUND i

Field experience has shown that fatigue cracks can be expected in feed-water nozzles and CRD return line nozzles in Boiling Water Reactor (BWR)

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pressure vessels. Ob se rva ticas , theoretical explanations, and reco= mended ,

remedial measures are discussed in NUREG 0312.

. The principal f actors respcnsible for nozzle crack initiation and growth l are thermal cycling and the stresses from the dif ferential ther=al expansion between the weld-deposited stainless steel cladding and the forged, low-alloy steel nozzle.

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4. TECHNICAL EVALUATION i 4.1 FEEDWATER N0ZZLES l During the 1979 refueling outage the accessible areas of the nozzles were a

PT inspected with the welded thermal sleeves in place. The penetrant tests l revealed no indications of cracks.

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Several operating and design factors were considered in this evaluation, including (1) the welded ther=al sleeve, (2) better control of water che istry, and (3) a reduced number of reactor scrass. It is recognized that properly l welded ther:a1 sleeves should prevent any feedwater leakage and thereby alle-viate thermal cycling and subsequent fatigue cracking. However, the welded sleeve does prevent a thorough inspection of the bore and inner blend radius of the nozrle. Because practical considerations do not permit re oval of the welded sleeve, the planned inspection pregram of the nozzles is acceptable and satisfies the requirements of NUREG 0312. Also, as discussed in another Technical Evaluation Report, the possibility of stress corrosion cracking in

the attachment weld is an area of concern.

l' 4.2 CONTROL R00 DRIVE RETUP.N LINE N0ZZLE j

f The CRD return line nozzle was PT exa=ined with the ther=al sleeve removed.

I Several indications of cracks were found in the nozzle cladding and were removed by grinding. One of these cracks extended 3/16 inches into the base

! metal. Several through-wall cracks were found on the thermal sleeve, which was not replaced. The CRD return line war valved out. This should elicinate one source of ther=al cycling. l I

Two areas of concern exist:

1. Crack'. ng has been observed in 3*=~R reactors in the apron area 4

directly below the no sie a year af ter the return line was "MYNs2

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l valved out. Therefore, an appropriate inspection progras

! must be i=plemented during the next scheduled refueling j

outage to ensure that cracks have not developed that might

impair the safety of the reactor.

l l 2. Because of the susceptibility of stagnant stainless steel i lines to stress corrosion cracking, an appropriate inspection program =ust be developed to exa=ine the sv.inless steel welds on the reactor vessel side of the valve used for isolation.

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5. CONCLUSION The inspection progran for the feedwater nozzles and the CFO return line I nozzle is acceptable. Although valving out the return line effectively elimi-

{ nates the cold flow and subsequent thermal cycling, concern still exists f or j possible cracking in the apron area below the no::le. An adequate inspection program must be instituted to =enitor this area. An additional inspection a program must be developed to monitor the possible initiation of stress corro-t sion cracking in the stagnant stainless steel return line.

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