ML17262A482

LER 91-005-00:on 910414,steam Generator Tube Degradation Occurred Due to Iga/Scc & Primary Water Stress Corrosion Cracking Attack on Steam Generator Tubing.Affected Tubes Sleeved or plugged.W/910513 Ltr

ML17262A482
Person / Time
Site:	Ginna
Issue date:	05/13/1991
From:	Gorski P, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-005, LER-91-5, NUDOCS 9105210249
Download: ML17262A482 (20)

Text

ACCELERATED DILUTION DEMONS~TION SYSTEM iP REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9105210249 DOC.DATE: 91/05/13 NOTARIZED: NO DOCKET g FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME . AUTHOR AFFILIATION GORSKI,P. Rochester Gas & Electric Corp.

MECREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME 'ECIPIENT AFFILIATION R

SUBJECT:

LER 91-005-00:on 910414,steam generator tube degradation.

Caused by "A" & "B" S/GS tube degradation in excess of Ginna QAM reportability limits. Tubes repaired using Combustion Engineering welded sleeve.W/910513 ltr. . S DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event, Report (LER), Incident Rpt, etc.

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 1 PD1-3 PD 1 1 JOHNSON,A 1 1 INTERNAL: ACNW 2 2 AEOD/DOA 1 1 AEOD/DS P/TPAB 1 1 AEOD/ROAB/DSP 2 ' 2 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 NRR/DLPQ/LHFB11 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB SD 1 1 NRR/DST/SICB 7E 1 1 Nj%/BS~LBSD1 1 1 NRR/DST/SRXB SE 1 1 RE~IL~ FILE 02 01 1

1 1

1 RES/DSIR/EIB 1 1 EXTERNAL: EG&G BRYCE, J. H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 D

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NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED'TTR 31 ENCL 31

ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER N.Y. 14649-0001+,

ROBERT C. MECREDY TELEPHONE Vice President AREA CODE 716 546 2700 Cinna Nuclear Production May 13, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 91-005, Steam Generator Tube Degradation Due to IGA/SCC Causes Q.A. Manual Reportable Limits to be Reached R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (Other), and the Ginna Station Quality Assurance Manual Appendix B, which requires that, "If the number of tubes in a generator falling into categories a or b below exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73," the attached Licensee Event Report LER 91-005 is hereby submitted.

This event has in no way affected the public's health and safety.

Ver truly yours, Robert C. Me redy XC: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector 9105210249 9i05i3 PDFl ADQCK 0 000244 WG &a s FDR ~II

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IVSSLCNINTAL IICSOAT CLSCCTCO IIONTII CAY YCA1 CISCCTCO WIINCC ION OATI 11 ~ I Ylt Oy y<<. CXSCCTCO CVINIC>>ON OAT>> NO AICTAACTILJevc 0 I Axl eeeee. IA, eeeyeeeeeMy NINee eeeseeeeeee Iyftsrlllee <<eeeI IW During the 1991 Annual Refueling and Maintenance Outage subsequent to the eddy current examination performed on both the "A" and "B" Westinghouse series 44 Steam Generator (S/G), 116 tubes in the "A" S/G and 117 tubes in the "B" S/G required corrective action due to tube degradation.

The immediate cause of the event was that the "A" and "B" S/Gs tube degradation was in excess of the Ginna Quality Assurance Manual reportability limits.

The underlying cause of the tube degradation is a common S/G problem of a partially rolled tube sheet crevice with recurring Intergranular Attack/Stress Corrosion Cracking (ZGA/SCC) and Primary Water Stress Corrosion Cracking (PWSCC) attack on S/G tubing.

Corrective action taken was to ei.ther sleeve or plug the affected tubes with accepted industry repair methods.

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~ LICENSEE EVENT REPORT ILERI TEXT CONTINUATION II.5. IIIICLtA1 154MLATOAY COQMINIISI r

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~ ~ ~. ~ I@AC /4PIII ~tl IITI o s o o o 24 491 005 00 02 OF PRE-%TENT PLANT'ONDITIONS The unit was in the Cold/Refueling Shutdown condition for

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the Annual Refueling and Maintenance Outage. Reactor Coolant System (RCS) pressure was zero (0) psig and RCS temperature was approximately 68 F. Steam Generator eddy current inspection was in progress.

DESCRIPTION OP ~9FZ A. DATES AND APPROXIMATE TIMES OP MAZOR OCCtJREUBTCES:

o April 14, 1991, 1600 EDST: Event date and time.

o April 14, 1991, 1600 EDST: Discovery date and time.

o April 16, 1991, 1430 EDST: Oral notification made to the NRC office of Nuclear Reactor Regulation (NRR).

o April 22, 1991, 0400 EDST: Steam Generator repairs completed.

o April 26, 1991, Follow-up written report sent to miseries NRC Office of NRR. r B. &TENT:

During the 1991 Annual Refueling and Maintenance Outage, an eddy current examination .was performed in both the "A" and "B" Westinghouse 44 iDesign recirculating steam generators.

The purpose of the eddy current examination was to assess any corrosion or mechanical damage that may have occurred during the cycle since the 1990 examin-ation.

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I NAC term ESSA I~ I LICENSEE EVENT REPORT ILERI TEXT CONTINUATION U.S. NUCEtAN AtoULATOAVCOMMISOON

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Aee1OVEO OMS NO SISO&IOe EXPIIIES. SIS14S DOCKET NUMStll Itl Itl

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EEII NUMOSN ~ AOE ISI stoUENnAI. " etv4ON N M 1 M A R.E. Ginna Nuclear Plant 244 91 005 00 030F TExT /1I'vee NMee e ~. eee eeeaonM ITIIC JKI'el I ITI o s o o o 0 8 The examination was performed by personnel from Rochester Gas and Electric (RG&E) and Allen Nuclear .

Associates, Inc. (ANA). .All p'ersonnel were trained and qualified in the eddy current examination method and had been certified to a minimum af Level I for data acquisition and Level II for data analysis.

The eddy current examination of the "A" and "B" steam generators was performed utilizing the Zetec Miz-18 Digital Data Acquisition system. The frequencies selected were 400, 200, 100 and 25 KHZ.

The inlet or hot leg examination program plan included the examination of 100% of each open unsleeved steam generator tube from the tube end to the first tube support. 20% of the hotleg tubes were selected for examination for their full length (20% random sample as recommended in the Electric Power Research Institute (EPRI) guidelines.) In addition, 20% of each type of sleeve was examined and the remaining tube examined full length. All previous tubes with indications greater than 20% through wall (TW) depth were examined, as a minimum, to the location of their degradation.

All Row 1 and Row 2 U-Bend regions selected as part of the 20% random sample were examined with the Motorized Rotating Pancake Coil (MRPC) between the g6 Tube Support Plate Hot (TSPH) and the g6 Tube Support Plate Cold (TSPC) from the cold leg side.

Results of the above inspections indicated that 116 tubes in the "A" steam generator (i.e. 91 new repairs

-plus 1 pulled tube plus 24 previously plugged tubes) and 117 tubes in the "B" steam generator (i.e. 98 new

- mepairs plus 1 obstructed sleeved tube plus 16 previously plugged tubes plus 2 plugs exhibiting PWSCC) required corrective action.

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R.E. Ginna Nuclear Power Plant tIIIC Iktttk ~'Sl IITI 2 449 l 0 05 00 04 Oi0 8 On April 14, 1991 at approximately 1600 EDST, the reactor was in the Cold/Refueling Shutdown condition with RCS Temperature and pressure at approximately 68 F and zero (0) psig respectively. At this time a final review of the eddy current data was completed and results indicated that more than 1 percent of the total tubes inspected were degraded (i.e. imperfections greater than the repair limit). Because of the above, the results of the inspection are considered a reportable event pursuant to 10 CFR 50.73 per Appendix B of the Ginna Station Quality Assurance Manual.

On April 16, 1991 at approximately 1430 EDST Oral Notification was made to the NRC 'office of NRR pursuant to Appendix B of the Ginna Station Quality Assurance Manual.

On April 26, 1991, a follow-up written report of the steam generators inspection and repairs was sent to the NRC Office of NRR pursuant to Appendix B of the Ginna Station Quality Assurance Manual:

C INOPERABLE STRUCTURESt COMPONENTSt OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED-k None.

E. METHOD OF DISCOVERY:

The event was apparent after the review of the "A" and "B" Steam Generators eddy current examination results.

F. OPERATOR ACTION:

None.

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None.

III. CAUSE OP EVE2FZ A. IMMEDIATE CAUSE:

The Immediate Cause of the event was that the "A" and

~ "B" steam generator tube degradation was in excess of the Ginna Station Quality Assurance Manual Reportable limits.

B. ROOT CAUSE:

The results of the examination indicate that IGA and IGSCC continue to be active within the tubesheet crevice region on the inlet side of each steam generator. As in the past, IGA/SCC is much more prevalent in the "B" steam generator with 42 new IGA indication and 37 new IGSCC indications reported. In the "A" steam generator, 14 new IGA indications and 16 new IGSCC indications were reported.

The majority of the inlet tubesheet crevice corrosion

. indications are IGA/SCC of the Mil-annealed Inconel 600 Tube Material. This form of corrosion is believed to be the result of the tubesheet crevices forming an alkaline environment. This environment has developed over the years as deposits and active species like sodium and phosphates, have reacted, changing a neutral or inhibited crevice into the aggressive environment that presently exists.

Along with IGA/SCC in the crevice, PWSCC at the roll transition appears to have a slight increase in growth during the last operating cycle. This mechanism was first addressed in 1989 and this year there were 19 PWSCC indications in "B" Steam Generator and 59 PWSCC indications in "A" Steam Generator.

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'4IOH R.E. Ginna Nuclear Power Plant 2 4 4 91 00 5 0 006 0FO 8 TEXT Ill' Vrv r teWeer. yv ereecveV r'IIC ~ ~'al I ITI o s o o o ANALYSIS OF EVENT The event is reportable in accordance with 10 CFR 50.73, Licensee Event Report item (Other) and the Ginna Station Quality Assurance Manual Appendix B which requires that, "If the number of tubes in a generator falling into categories a or b exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73." The, tube degradation in the "A" and "B" steam generators exceeded the criterion of (b) which states, "more than 1% of total tubes inspected are degraded, (imperfections greater than the repair limit)."

This repair limit is defined as, "steam generator tubes that have imperfections greater than 404 through wall, as indicated by eddy current, shall be repaired by plugging or sleeving."

An assessment was performed considering the safety con-sequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences or safety implications resulting from the steam generator tube degradation in excess of the Quality Assurance Manual reportable'imits because:

o The degraded tubes were identified and repaired prior to any significant leakage or steam generator tube rupture occurring.

o Even assuming a complete severance of a steam generator tube at full power, as stated in the R.E. Ginna Nuclear Power Plant updated Final ,Safety Analysis Report (Ginna/UFSAR) section 15.6.3, (Steam Generator Tube Rupture) the sequence of recovery actions ensures early 'ermination of primary to secondary leakage with or without offsite power available thus limiting offsite radiation doses to within the guidelines of 10 CFR 100.

Based on the above, it can be concluded that the public's health and safety were assured at all times.

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~s tA R.E. Ginna Nuclear Power Plant 449- 05 000 08 TExf illsssvY AsAcs 0 ~. ssv Atsssawss sYAc ssssssss ~ TO I I TI o s o o o 2 1 7 ov V. CORRECTIVE ACTION A. ACTION TAKEN TO RETURN APPECTED SYSTEMS TO PRE-E%9lT STATUS: 'ORMAL 0 Of the 116 degraded tubes in the "A" steam generator, 61 tubes were repaired using a Combustion Engineering'27" welded sleeve in the hot leg plus 41 tubes were repaired using a Combustion Engineering 30" welded sleeve in the hot leg and all of the above tubes will remain in service. The remaining 14 tubes were removed from service by plugging both the hot and cold leg tube ends. A total of 190 tubes in "A" @G are. currently plugged and 324 tubes are sleeved.

0 Of the 117 degraded tubes in the "B" steam generator, 80 tubes were repaired using a Combustion Engineering 27" welded sleeve in the hot leg plus 28 tubes were repaired using a Combustion Engineering 30" welded sleeve in the hot leg and all of the above tubes will remain

.in service. The remaining 9 tubes were removed from service by plugging both the hot and cold leg tube ends. A total of 326 tubes in the "B" S/G are currently plugged and 938 tubes are sleeved.

B. ACTION TAKEN OR PLANNED TO PE~TENT RECURMQTCE The occurrence/presence of IGA, SCC and PWSCC is a common PWR Steam Generator problem. Utilities with

~usceptible tubing and partially rolled crevices must deal with this recurring attack on steam generator

- -<ubing.

R.E. Ginna Station will continue careful monitoring of both primary RCS and secondary side water chemistry parameters. These water chemistry parameters will be evaluated against accepted industry guidelines in order to minimize harmful and/or secondary side environments.

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s>>sIAss snIns SACILITY NAIIC Ill DOC>>ST NU~SII ITI LSII NI>>AOSII III SAOS ISI ssQUANT>AL YNIOH R.E. Ginna Nuclear Power Plant o 5 o o o 24 491 00 5 0 0 08 OFO 8 TGCT IIIee>>e NNee ~ teeeec ceo eeeeeeeI JV>>C See>> ~'el I ITI Degraded Steam Generator tubes shall be sleeved or plugged in accordance with the inservice inspection program and accepted industry repair methods.

ADDITIONAL INFORMATION FAILED COMPONENTS.

The degraded components are: Inconel Grade 600 tubes having an outside diameter of 0.875 inches and a nominal wall thickness of 0.050 inches.

B. PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historical search was conducted with the following results: The crevice indications are similar to those reported in A0-74-02, A0-75-07, R0-75-013, and LERs76-008, 77-008,78-003, 79-006, 79 022s 80 003s 81 009s 82 003s 82 022s 83 013s 89 001, and 90-004.

C. SPECIAL COMMIBFXS:

For a more indepth report, refer to the, "1991 Steam Generator Eddy Current Examination Summary Report" sent to the NRC April 26, 1991.

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