ML20134P918

Forwards Unit 1 Cycle 6 Midcycle Outage,Summarizing SG Inservice Insp Rept

ML20134P918
Person / Time
Site:	Braidwood
Issue date:	11/19/1996
From:	Tulon T COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9611290377
Download: ML20134P918 (5)

Text

Commonweahh Edison Osmpany liraidwood Generating $tation

• Route 81, ik)x M i

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• liraceville, II. 60 407.%I9 Tel Hl5-4 SR2801 November 19,19%

U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Comed Braidwood Station Unit 1 Cycle 6 Midcycle Outage

' Steam Generator Insenice Inspection Report Docket No. STN 50-456

References:

(1) NUREG-1276, Technical Specifications, Braidwood Station, Unit Nos. I and 2 Specification 4.4.5.5.c of reference (1) requires that results of steam generator (SG) tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days, and prior to resumption of plant operation. This report shall proside a description of investigations conducted to determine the cause of tube degradation and corrective measures taken to prevent recurrence.

The initial sample inspection resulted in the following SGs being classified into Category C-3 based on the number of defective tubes found during the Cycle 6 Midcycle outage (AIP02):

1A,IB,1D On October 24,1996 at 20:45 hours 1C On October 26,19% at 18:30 hours Notification per Technical Specification Table 4.4 2 pursuant to 10 CFR 50.72 (b) (2) (i) for steam generators being classified in Category C-3 was initiated. The enclored summarizes the inspection results.

Please direct any questions regarding this submittal to Douglas Huston, Braidwood Licensing .

Supenisor, (815) 458-2801, extension 2511. l l

Very truly yours, )

l y J.Tulon

~

tion Manager Braidwood Station l l

TJT/fb/= ew l 1

Enclosure:

Steam Generator C-3 Report ]

Attachment , l cc: Senior Resident inspector - Braidwood Braidwood Project Manager, NRR

[ J Regional Administrator, Region 111 C00098 9611290377 961119 PDR ADOCK 05000456 G PDR A t!nicom Company

4 Braidwood Station Unit 1 Cycle 6 Midcycle Outage alp 02 Steam Generator C-3 Report On October 15,1996, a Steam Generator (SG) Tube Inservice Inspection was initiated on 4 Braidwood Unit I as a result of a commitment to the NRC for a midcycle inspection.

Technical Specification Surveillance Requirement (TSSR) 4.4.5.2.e requires that the results of each sample inspection be classified into one of three categories. A SG will be

classified in Category C-3 if more than 10% of the total tubes inspected are degraded or more than 1% of the inspected tubes are defective. A SG tube is considered degraded ifit has an imperfection of greater than or equal to 20% of the nominal tube wall thickness. A SG tube is considered defective ifit has an imperfection of greater than or equal to 40% of ,

- the nominal tube wall thickness. .

t j An initial sample size of 100% of all available hot-leg top-of-tubesheet roll transitions was selected. The inspection was conducted using the 3-coil Plus Point probe consisting of a  !

Plus Point coil, a 0.080 inch Pancake coil, and a 0.115 inch Pancake coil. The eddy .

current data was analyzed using the EddyNet 95 software. The initial sample inspection i

resulted in the following SGs being classified into Category C-3 based on the following j l reasons: ,

3 IA On October 24,1996 at 20:45 hours, greater than 1% of the 4029 inservice  !

tubes being defective.

IB On October 24,1996 at 20:45 hours, greater than 1% of the 4450 inservice  !

tubes being defective.

1C On October 26,1996 at 18:30 hours, greater than 1% of the 3835 inservice tubes being defective.

l' ID On October 24,1996 at 20:45 hours, greater than 1% of the 4181 inservice tubes being defective.

Notification per Technical Specification Table 4.4-2 pursuant to 10 CFR 50.72 (b) (2) (i) for steam generators being classified in Category C-3 was initiated within four hours of the affected SGs being evaluated as C-3.

i Table 1 provides, by SG, the number of tubes plugged during this outage, the number of ) '

tubes sleeved during this outage as well as the number of tubes plugged in previous outages.

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l Table 1

Braidwood Unit 1 A1P02 Inspection / Repair Overview

}Mication 1 A SG 1B SG IC SG 1D SG Total ',

Circumferential 221 276 610 293 1400 Axial 1 20 3 8 32 Mixed Mode 2 3 6 3 14 TotalIndications 224 299 619 304 1446 j Total Raamirable Tubes

• 224 295 618 299 1436 4 Total Tubes Sleeved 181 0 445 271 897 Iubes Stabili>*A and Plunned~

43 295 173 28- 539 AIP02 Equiv. Plugged Tubes 53.5 2% 198.9 43.8 592.2 Tubes Previously Plugged 549 128 743 397 1817 '

Total Tubes Pinooed -

602.5 424 941.9 440. 2409.2 Total Eauivalent Plu_nn_ing (%) 13.2 9.3 20.6 9.6 13.2  !

• Ten tubes had more than 1 indication. {

All repairable indications identified during this inspection have been repaired using the

Westinghouse Laser Welded Sleeve or have been removed from service by plugging.

Stabilizers were installed in the tubes that were plugged so that further degradation of the tubes can not result in a double ended break at the top of the tubesheet. With the

, additional 539 tubes plugged during this inspection and accounting for the 17.2:1 sleeves to plug ratio, Braidwood Unit I has 13.2% of the total SG tubes plugged with a single loop maximum of 20.6% plugged in the IC SG loop. These plugging levels are within the acceptable range to ensure the RCS total flowrate in Technical Specification 3.2.3 is ,

achieved. i i The Westinghouse Laser Welded Sleeving process and the mechanical SG tube plugging process used at Braidwood are approved methods of SG tube repair per the Braidwood 1 Technical Specifications.

Twenty-three tubes, with the largest indications, were insitu pressure and leak tested. .

3 None of the tubes burst at three times normal operating differential pressure l demonstrating the' structural integrity of the SGs prior to this inspection. Some of the j' tubes leaked under insitu pressure testing at Main Steam Line Break differential pressure.

The total leakage in the limiting SG was 2.15 gpm. Adding this to the predicted leakage i from the Tube Support Plate Interim Plugging Criteria (6.99 gpm) and potential leakage from the unfaulted SGs (0.3 gpm), the resultant leakage from the limiting SG is 9.44 gpm.

e This is less than the site allowable leakage of 26.8 gpm. Therefore, the accident dose rate 7.-

is below a small fraction of the 10CFR100 limits.

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v This report is being provided in accordance with Specification 4.4.5.5.c of NUREG 1276, Technical Specifications, Braidwood Station, Units Nos. I and 2.

I INVESTIGATION CONDUCTED TO DETERMINE THE CAUSE OF THE TUBE DEGRADATION During the Braidwood Unit I midcycle SG Tube inservice inspection, axial and  ;

circumferential indications were identified at the SG top-of-tubesheet roll transition  ;

I region. These top-of-tubesheet indications resulted in 539 tubes being removcd from service by plugging and 897 tubes being sleeved using the Westinghouse Laser Welded i

Sleeve. The top-of-tubesheet indications are a result of Outside Diameter Stress Corrosion Cracking (ODSCC). The cause of the indications is related to the  ;

f manufacturing precess of hard-rolling the SG tubes into the tubesheet. This process caused residual stresses in the SG tubes, thereby creating an environment for stress 1 corrosion cracking. A majority of the indications are located in the "T" slot of the tube  ;

bundle. Sludge accumulates in the area of the "T" slot. This suggests a possible link between sludge and the formation ofindications at the top-of-tubesheet roll transition region. The "T" slot is susceptible to dry-out which could also contribute to the number ofindications in this area.

l Twenty-three tubes with circumferential indications were insitu pressure and leak tested. l The insitu test program was discussed with NRR. All of the tests were conducted per the l agreements with NRR. All of the tubes were insitu tested to 3100 psi (above Main Steam l I

Line Break conditions when corrected for temperature and pressure). Ten of these tubes were tested to 5000 psi (greater than three times normal operating differential pressure -

Regulatory Guide 1.121). None of the tubes burst, therefore, the tests proved the ,

structural integrity of the tubes. Some of the tubes leaked under accident conditions. The i total leakage in the limiting SG was 2.15 gpm. Adding this to the predicted leakage from j the Tube Support Plate Interim Plugging Criteria (6.99 gpm) and potential leakage from {

the unfaulted SGs (0.3 gpm), the resultant leakage from the limiting SG is 9.44 gpm. This l l

is less than the site allowable leakage of 26.8 gpm.

Four of the ten tubes tested to 5000 psi were removed from the IB SG to determine the morphology of the indications, assess the percent degraded area and 100% throughwall area, and to assess the eddy current voltage sizing techniques.

CORRECTIVE MEASURES TO PREVENT RECURRENCE All of the defective tubes were repaired by sleeving or were removed from service by plugging. The Braidwood Unit 1 Steam Generators will be replaced during the AIR 07 refueling outage (Fall of 1998). The new Steam Generators will include a different tube material (Inconel 690) and the tubes will be hydraulically expanded into the tubesheet instead of hard-rolling. The Inconel 690, along with the hydraulic expansion of the tubes, will reduce the potential for stress corrosion cracking at the roll transition region.

The dominant form of corrosion identified during the AIP02 outage is circumferential )

ODSCC at the top-of-tubesheet roll transition regions. Braidwood Station implemented the following programs to mitigate the corrosive environment in the SGs which lead to  ;

ODSCC:

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- Use of advanced amines, such as methoxypropylamine (MPA), for secondary pH control to reduce the amount of corrosion products which enter the SG.

- Compliance with the EPRI Secondary Chemistry Guidelines.

- Maintain hotwell dissolved oxygta concentrations <3 ppb.

- Continue use of high hydrazine concentrations for maintaining reducing conditions in the SGs and passivation of piping systems and components.

i Analysis of samples of sludge from previous outages determined that Braidwood Unit 1 does not have significant levels oflead or copper in the sludge. Lead and copper have t

been identified as contributors to the formation of ODSCC.

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