ML18142A536

Summary of 850610 Meeting W/Utils & Westinghouse in Bethesda,Md Re Repair of Steam Generator Transition Zone Upper Girth Weld.List of Attendees & Meeting Handouts Encl

ML18142A536
Person / Time
Site:	Surry
Issue date:	06/20/1985
From:	Neighbors J Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8507030088
Download: ML18142A536 (23)

Text

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June 20, 1985

• Docket No. 50-281 LICENSEE:

FACILITY:

SUBJECT:

VIRGINIA ELECTRIC AND POWER COMPANY (VEPCO)

SURRY POWER STATION, UNIT NO. 2

SUMMARY

OF MEETING HELD ON JUNE 10, 1985, TO DISCUSS STEAM GENERATOR WELD REPAIR The subject meeting was held in Bethesda, Maryland, to discuss the repair of the steam generators* transition zone upper girth weld repair. A list of attendees is enclosed.

The enclosed presentation describes the discovery, inspections, repairs and analyses of the cracking in the welds of the three steam generators.

The cracks in the welds in all steam generators have been repaired by contoured grinding.

The NRC, Mr. Hazelton, said that he had no concerns with what VEPCO has done and startup of Unit 2 would not be impacted.

However, he could not comment on the long term aspects until NRC has reviewed VEPC0 1s final report which will be a fracture mechanics analysis. This report should be submitted about July 1985.

Enclosure:

As stated ORB#l:DL~

CParrish 06/9"{¥85 ORB#l:DL DNeiQhbors/ts 06~c)/85 8507030088-850620-

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PDR ADOCK 05000281 p

PDR I

JDNeighbors eph D. Neighbors, Project Manager Operating Reactors Branch #1 Division of Licensing

r UNITED STATES e

rJUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 June 20, 1985 Docket No. 50-281 LICENSEE:

FACILITY:

SUBJECT:

VIRGINIA ELECTRIC AND POWER COMPANY (VEPCO)

SURRY POWER STATION, UNIT NO. 2

SUMMARY

OF MEETING HELD ON JUNE 10, 1985, TO DISCUSS STEAM GENERATOR WELD REPAIR The subjiect meeting was held in Bethesda, Maryland, to discuss the repair of the steam generators' transition zone upper girth weld repair. A list of attendees is enclosed.

The enclosed presentation describes the discovery, inspections, repairs and analyses of the cracking in the welds of the three steam generators.

The cracks in the welds in all steam generators have been repaired by contoured grindin9.

The NRC, Mr. Hazelton, said that he had no concerns with what VEPCO has done and startup of Unit 2 would not be impacted.

However, he could not comment on the long term aspects until NRC has reviewed VEPCO's final report which will be a fracture mechanics analysis. This report should be submitted about July 1985.

Encl osU1re:

As stat1~d as h~~ct Operating Reactors Branch #1 Division of Licensing Manager

NRC D. Neighbors W. Hazelton D. Smith J. Henderson K. Johnston VEPCO J. McAvoy H. Miller D. Fortin J. Hegner WESTINGHOUSE R. Rishel A. Vaia W. Banford J. Crane e

LIST OF ATTENDEES JUNE 10, 1985 MEETING POWER AUTHORITY OF THE STATE OF NEW YORK P. Kokolakis L. Hi 11 e

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e UT Indication

Background

e UT iriidication identified August 1983 o H-1

• Initiiil evaluation by Virginia Power was contour

• NRC iquestioned evaluation

• NRC evaluated indications as ei,ther contour or cracking Virgi111ia Power agreed to MT inspect in 198~5

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"2-I11dications

• Pi,ts over general area/visual

• Linear crack-like at toe of crown of weld

1. 6/vi'sual 1 Ox MT, UT

• Sub surf ace in region of OD weld root/UT

e R'epair of Welds by Grinding

• All three stea1JJ generators S11rf ace flaws only

• -- Use oJf detailed repair procedures Grincling in increments of 1/16" or 1/8"

• MT irtspection after each increment. of grintiing Final surf ace MT clear

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Surf ace Flaws

• All surf ace jla,ws removed by 0.500" grinding tiepth in '~" SG

• All surf ace flaws removed by 0.31 "gtindi11g tlepth in "B" and "C' SG

A" SG contour ground more deeply than 1zecessary at several locations

Sub surf ace Flaws

• Locateti and sized by UT

• Analyzed by Westinghouse

• Analysis technique L4 W ASME Section XL Apperidix A methodology OH - 5""""°

e Stress Analysis

• Stresses result from Design Loads

- -Faulted Loads Test Loads

• Repair configurations with a 1/2" deep and 1" deep groove meet Primary allowables

• Repair configuration modeled by finite elements

• Finite elements used to calculate pressure and thennal stresses o H - b

e Fatigue Analysis

• Peak stress conditions during two enveloping transients are used for fatigue evaluation

• Plant operating data for Suny 1 and 2 used for projected number of occurrences of eachtransient for 40-year period Stress ranges due to pressure and temperature combined uith number of transient cycles for fatigue analysis For deepest grindout (0.85") stress ranges have been ampUfied by use of stress concentration factor for fatigue ana/JJSis

• Reference repair and local overgrind have fatigue lives in excess of 40 years

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e Mt:~chanisni of Fla.w Formation and Growth

• Mechianism still under investigation

• Records under review Mciterial certs Raidiographs of welds Weld records Pll7HT weld 6 a11d u,eld 11 Wt:zter chemistry op,erational history

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Possible Scenario Surf ace pitted due to oxygen, chloride, and copp,er ingres~ due to leaky condensers

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Corrc,sion pits acted as nucleation sites for stress co11osion or co11osion - fatigue

-- "'cracks CracA~s propagated by combination of factors Corrosion Sttitic stress ( operational and residual)

-Fatigue

e Cl.1en1istry Considerations Secondt:lry chemistry problems prior to SG/cofzdenser replacement O~J,,gert often greater than 25 PPB Chloricle often in PPM range occasionally 300 tl> 400 PPM Copper alloy condensers and f eedwater heaters Since 5IG}ondenser replacement chemistry has been very good O',,. - '0

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Construction Period Welding Considerations

• ID1 preheat temperature 180 - 185° F

01) preheat temperature 210 - 220°F

/£J welded completed be/ ore OD

• PlVHT was accomplished at 1000° F to 1' l00°F 0 f-/ - J /

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e FacttJrs Mitigating Re-Initiation of Cracking

• Much b 1etter water chemistry

• Lou,er residual stress due to PWHT of weld 6 u.1it/J weld 11 at 1150 to 1180°P

• Remov.~l of all linear indications o H -

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e Results of Analysis of Subsurface UT Indications OH -

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'~" ~5G No unacceptable indications per ASME Xl, IWB 3511 "C" SG

/\\lo unacceptable indications per ASME J{J, IWB 3511 "B" SG

_s, unacceptable. indications per ASME XI, llVB 3511

.1tll unacceptable are sub surf ace

~411 indications acceptable by ASME XI,.

.1.4Ppendix A Analysis

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Future Inspections

• Inspectfons above Code /SI requirments

• . Surface inpections during 1986 and 1988 refi1eling outages

'~" SG MT inspection at 4 locations of highest stress concentration in weld no. 6 Each, location 2' minimum length Tota:/ minimum length inspected 8' ALAlM considerations for stay time in generators

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e Future Inspections

. Inspections above Code /SI requirments

• Subsu~face inspections during 1988 refuel-ing oi,tage "B" 5'G UT inspection of all flaws accepted by A.SME XI Appendix A FM analysis

- If n(J growth is found, further inspection will beper ASME XI for flaws accepted by ana'./ysis ot-1-1~

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e MAGNIFIED

-,~ REGION I

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I FIGURE lA FIGURE 1A:

~u~R~I~t~~fs~~E~~~~~io~F CONE AND UPPER SHELL.

FIGURE 18:

TRANSITION CONE AND UPPER SHELL BEFORE GRINDING FIGURE lC:

TRANSITION CONE AND UPPER SHELL AFTER GR!NDING e

FIGURE lB FIGURE 1 C

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RESULTS OF FLAW.EVALUATION

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.4 LOCATION a.*

Q.+(30 YEARS)

KICMAX)

KI( ALLOWABLE)

,.L 8" 3/4"

.290 0.3i37

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OU u.-, I-'

10'9-1/2" - 11'2-1/2"

.235 0.2853 60 63.3 14'0" - 14'4"

.145 0.1463 27 63.3 16'8-3/8" - 16'9"

.175 0.1755

20.

63.3 19'9-1/4" - 19'10-1/2"

.130 0.1304 21 63.3 20'8-3/4" ~ 20'9-1/4"

.175 0.1832 44 63.3 27'10-15/16" - 28'3-1/2"

.130 0.1306 21.5 63.3 45'7-1/4" - O'

.230 0.2586 58 63.3 45'9" - 46'0"

.175 0.1846 46 63.3

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TABLE 1.1 INDICATIONS IN STEAM GENERATOR B (UNACCEP:TABLE PER ASHE SECTION XI IWB-3511)

Minimum Half Distance Flaw Flew from Depth Length Surface LOCATION a Cin) l Cin) s Cin) lie alt olt an 3/4"

.290 2.750

.,s 9.5

.0817

.132 10'9-1/2" - 11 12-1/2"

.235 5.000

.18 21.3

.0662

.117 14'0" - 14'4"

.145 4.000

.54 27.6

.0408

.193 16 18-3/8" - 16'9"

.175

.625 1.20 3.5

.0493

.387 19'9-1/4" - 19'10-1/2"

.130 1.250 l.00 9.6

.0366

.318 20 18-3/4" - 20 19-1/4"

.175

.500

.13 2.8

.0493

.086 27'10-15/16" - 28 13-1/2"

.130 4.562

.67 35.l

.0366

.225 45'7-1/4" - 01

.230

~-750

.14 20.1

.0648

.104 45'9" - 46'0"

.175 3.000

.18 17.l

.0493

.100

e MEETING

SUMMARY

DISTRIBUTION OPERATING REACTORS BRANCH NO. 1 Docket or Central file NRC PDR Local PDR ORB#l ROG J. Partlow (Emergency Preparedness only)

Steve Varga Project Manager OELD E. Jordan P. McKee ACRS (10)

NSIC Gray file Plant Service List C. Parrish J. Partlow NRC Participants

~J. Hazelton D. Smith J. Henderson K. Johnston