ML20028H684

LER 90-021-00:on 901222,potential RCS Leak Noted in Area of Pressurizer Upper Level Instrument Nozzle.Caused by Pure Water Stress Corrosion Cracking.New Nozzle Installed Into Penetration from Shell OD.W/910121 Ltr

ML20028H684
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/21/1991
From:	James Fisicaro, Scheide R ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN019112, 1CAN19112, LER-90-21, LER-91-021, NUDOCS 9101280081
Download: ML20028H684 (5)

Text

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Operations .;a, u n m x n, Januar) 21, 1991 1CAN019112 U. S. Nucinar Reguintory Commission Document Control Desk Mail Station PI-137 Washirig;.on, D. C. 20555

SUBJECT:

Arksnsas Nuc1 car Ono - Unit 1 Docke.c No. 50-313 Licenne No. DPR-51 Licenio Event Report 50-313/90-021-00 Gentlemen:

In accordance with 10CFR50.73(a)(2)(1)( A), attached is the subject report concerning a reactor shutdown required by Technical Specifications due to an unisolable leak in a pressurizer nozzle which was caused by purn water stress corrosion cracking.

Very truly yours,

.Ex Jamt. J. Fisicato Man ger, Licensing JJF/RilS/mmg i

Attachment cc: Regional Administrator Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INPO Records Center Suite 1500 1100 Circle, 75 Parkway Atlanta, GA 30339-3064 9101280081 910121 PDR S ADOCK 05000313 /'

PDR C u p V O '.?

NIC forn 366 U.S. Nocimr Regulatory Comissim 16-89)* Arptwn101B No. 3150-0104 Expires: 4/30/92 1lCENSEE EVENT REPORT (I,E R)

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On December 22, 1990, maintenance personnel identified a potential Reactor Coolant System leak in the area of a pressurizer upper level instrumentation nozzle. An

. inspection was ccnducted which verified the existence of a very small leak at the nozzle. A Notification of Unusual Event was declared at 1011, and the plant was taken to cold shutdown. Subsequent inspection using Nondestructive Examinatton methods confirmed the existence of a small axial crack in the nozzle inner surface which extended to the annulus between the nozzle and the pressurizer shell and breached the outside diameter (00) of the nozzle at the toe of the nozzle to vessel weld. Based on the location and orientation of the flaw, and industry experience, the most probable root cause was :letermined to be Pure Water Stress Corrosion Cracking. A temporary repair was cornpleted which consisted of establishing the nozzle pressure boundary at the outside surface of the pressurizer and installing a new nozzle into the penetration from the shell OD. A Design Change Package will be developed and implemented during the next refueling outage (IR10) to provide a permanent repair to the nozzle.

NRC Fonn 366A U. S. Nuclear Reilatory C mnlasicri ,

(fr89) Appmved O!B No. 3150-0104 l

• Expires: 6/30/92 1 L10NTE IMNT REITT (ITR) 1DT QNf1NUATICN j TblLITY NME (1) RDIfNUMBER(2)' IFR NUMBER (6) IWIE (3)

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A. Plant Status i

At the time of this event, Arkansas Nuc1 car One, Unit One (ANO-1) was at a power Icvol of approximately 10-8 amps (intermediate range). Reactor Coolant System  ;

(RCS) [AB) temperature was approximately 532 degrees and RCS pressure was 2150 psig. Low power physics testing was in progress.

B. Event Description On December 22, 1990, after repairing a leak on a pressurizer level instrumentation isolation valve (RC-1002A), maintenance personnel identified a potential RCS leak in the area of the pressurizer upper level instrumentation nozzle. Although no visual indications were apparent, the noise level in the area indicated the presence of a leak. An inspection was conducted which verified the existence of a very small leak at the nozzle. Since this condition constituted an unisolable RCS pressure boundary leak, a Notification of Unusut1 Event (NUE) was declared at 1011, and a plant cooldown was initiated in accordance with Technical Specifications requirements. At 1949, the plant reached cold shutdown and the NUE was terminated.

Subsequent inspection using nondestructive examination (NDE) methods confirmed the existence of an exial crack in the nozzle Inner surface starting about 0.2 inch from the inner end and extending for approximately 0.4 inch. The leak path was apparently through this crack, which is believed to extend to the annulus betwoon the nozzle and the pressurizer shell, breaching the outside diameter (OD) of the nozzle at the toe of the nozzle to vessel weld. The indication on the OD of the nozzle was extremely small and closed up during cooldown, making it extremely difficult to locate.

C. Root Cause Based on the location of the nozzle flav, its axial oric+1tation, similar indications at other nuclear utilities and information supplied by the pressurizer vendor (Babcock and Wilcox), it was determined that the most l probable cause of the crack was Pure Water Stress Corrosion Cracking (PWSCC).

PWSCC refers to intergranular stress corrosion cracking in the primary water environment of pressurized water reactors (PWR). Laboratory and service experience indicates that this cracking can be hastened at elevated i temperatures, which is believed to be the reason the majority of the PWR nozzle failures have occurred in pressurizers. The evidence also suggests that certain of the product forms of Inconel Alloy 600, of which the ANO-1 pressurizer nozzles are made, are susceptible to PWSCC. A conclusive determination of the root cause could not be completed at this time because the portion of the nozzle containing the crack remains in the pressurizer and Is unavailable for analysis.

NRC Fonn 366A U. S. Nuclear Rg;ulatory Otmodssim

, , (6489), A;preved WB No. 3150-0104 Txpires: 4/30/92

, LICfNSff. IMhT REITT (112) TEXT CENrlNUAT1(N FACILITY NNE (1) ImIf NltiBER (2) 11R NltiBER (6) pKE (3)

Sntuantial Revisim Arkansas Nucicar Om, Unit Die Year Number NumLE_

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D. Corrective Actions A temporary repair was completed which consisted of establishing the level instrumentation aozzle pressure boundary at the outside surface of the pressurizer shell by depositing a wold pad on the shell OD around the nozzle penetration. A partial penetration weld preparation was formed in the pad and a new nozzle was installed into the penetration from the shell OD. A portion of the original nozzle will remain in place until a permanent repair is completed.

Evaluations were completed by the pressurizer vendor which concluded that the structural integrity of the pressurizer will not be jeopardized by the temporary nozzle repair for at 1 cast one fuel cycle.

A visuni inspection of the repaired pressurizer nozzle and other nozzles on the vessel was conducted during hot shutdown conditions prior to startup. No leakage was observed. Additionally, these nozzles will again be visually inspected for degradation at hot shutdown conditions prior to cooldown for the next scheduled shutdown.

A Design Change package will be developed and implemented during the next refueling outage (1R10) to provMe a permanent repair to the pressurizer nozzle.

If the portion of the nozzle remaining in the pressurizer which contains the crack can be saved at the Limo permanent repairs are made, it will be analyzed in an attempt to conclusively determine the root cause of the failure.

E. Safet.v Significance The safety significance of this condition is lessened by the fact that the unisolable RCS leak which resulted from the crack in the pressurizer nozzle was extremely small and did not cause a noticeable degradation of RCS pressure or result in any significant loss of inventory from th9 RCS.

Industry experience documents that failure of Alloy 600 components due to PWSCC occurs as a result of the propagation of axial cracks and that no such failures have been attributed to circumfenntini crack propagation. Therefore, considering tbc inherent toughness of Alloy 600 and the location of the nozzle within the pressurizer shell, a catastrophic failure of the nozzle was not likely to occur if ANO-1 had returned to power without identifying the leak.

The " leak before break" mode of failure which is characteristic of PWSCC would have facilitated detection of the leak prior to its becoming a significant safety concern.

NRC Fom 366A U. S. Nuclonr Regulatory Ctruissicu

.. (6'89), Arpmval WB No. 3150-0104

. Expires: 4/30/92 LICDMI E;BT RFRET (I1R) 'ITXT 0[NTINUATIW l

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F. Basis For Reportability This event is considered reportabic pursuant to 10CFR50.73(a)(2)(1)(A) because the identification of an unisolable RCS Icak necessitated the initiation and completion of a shutdown as required by the plants Technical Specifications.

G. Additional Information A similar event in which stress corrosion cracking resulted in an unisolable RCS leak was reported in LER 50-368/87 003-00.

Energy Industry Identification System (EIIS) codes are identified in the text as

[XX].

.