ML17250B160

LER 90-004-00:on 900416,steam Generator Tube Degradation Noted on 75 Steam Generator a Tubes & 211 Steam Generator B Tubes.Caused by Intergranular Attack/Stress Corrosion Cracking.Tubes Sleeved or plugged.W/900510 Ltr

ML17250B160
Person / Time
Site:	Ginna
Issue date:	05/10/1990
From:	Gorski P, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-004, LER-90-4, NUDOCS 9005160332
Download: ML17250B160 (13)

Text

ACCELERATED DISJBUTION DEMONS~ION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9005160332 DOC.DATE: 90/05/10 NOTARIZED: NO DOCKET 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 RECIPIENT AFFILIATION

SUBJECT:

LER 90-004-00:on 900416,steam generator tube degradation due to IGA/SCC causes QA manual reportable limits to be reached.

W/9 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR 2 ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

L SIZE:

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 PDl-3 LA 1 "

1 PD1-3 PD 1 1 JOHNSON,A 1 1 INTERNAL: ACNW 2 2 AEOD/DOA ,1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP '2 2 DEDRO 1 1 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB9H3 1 1 NRR/DLPQ/LHFB11 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB11 1 1 ~

NRR/DREP/PRPB11 2 2 NRR/DST/SELB SD 1 1 NRR/DST/SICB 7E 1 1 NRRQBSQ'LBSD1 1, 1 NRR/DST/SRXB SE 1 .1 ,

EG FILE 02 1 1 RES/DSIR/EIB 1 1 R E 01 1 1 EXTERNAL: EG&G STUART,V.A 4 4 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHYiG A 1 1 NUDOCS FULL TXT 1 1 P>govs 7MB 1

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: LTTR 34 ENCL 34

Q sCe a ~ 5. ~ r ~

15I+A 55ACC ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 TCLCa~OhC,

• RCA COOt. 1III 54G.2700 May 10, 1990 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 90-004, 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,.

Appendix B; which requires that, "If the number of tubes in Assurance'anual 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 attached Licensee Event Report LER 90-004 is hereby submitted.

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

Ve truly yours, o ert C. M redy Division Manager Nuclear Production xco U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector yA555 pgiP

NAC tea )00 VS NUCLCAA AIOVLATOARCOLNIIHION flOII APPAOVIO OLII NO. IIIO OIOH LICENSEE EVENT REPORT {LER) cxf)At~ ~ IIIIII PACILITY NAME lll OOCXI'1 NUMIIANl R.E. Ginna Nuclear Power Plant

"'"'" Steam o 6 o o 024 co(07 Generator Tube Degradation Due To IGA/SCC Causes Q.A. Manual Reportable Limits To Be Reached EVINT OATI III LtA NVMICA (4 AttOAT OATt ltl OTHER IACILITIIIINVOLVEO III IEOUINZIAL " AIvrjON MONTH OAY YIAA PACILITYH*ULI OOCXtT NVUIIAI4 UOHTN OAY YCAA YEAH NUUII A sf HUU ~ ~ A 0 5 0 0 0 IN)(HI 0 4 9 0 9 0 004 00 05 090 0 s 0 0 0 Otl A ATI HO THII ACPOAT II IUILIITTCOtUAIVANT1 0 THC AIOVIAIMENTIOt 10CPA $: IOUct Art W re tt IAP IUNrUPI 111 MOO C Ili II.LOIILI 10.MN(h IO.) 14

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LICCNIII CONTACT POA TH)I LIA l11)

HALIC TELEPHONE HUMICA Paul Gorski *AIACOOt Mechanical Maintenance Manager 1552 -4446 COMtLCTI ONt LINt fOA EACH COLIPOHtNT I AILUAC OIICAIItO IN THII AIPOAT (11(

MAHUfAC MAHVfAC EPOATAIL CAUlt IYCTIM COMPO H I HT TVAEA COMPOHCHT TVAEA TO HtADI l~4+8ggjr~eq,Sg It~I> '+%/K<"'

l(4 MONTH CAY YCAA IVttLCMlHTALAltOAT CxtlCTCO txtCCTtO lUILNCCIOH OATI (I ~ I

'YCI l)t tw. ~N CIPICTCO IUIUIIJIOHOATCI X HO LAITAACT I@cut IP I t00 NWN, I I, tyyntrr lelt AlrrL alLPtmCt nttrrINA VM (14 During the 1990 Annual Refueling and Maintenance Outage subsequent to the eddy current examination performed on both the "A" and "B" Westinghouse Series 44 Steam Generators (S/G), 75 tubes in the "A" S/G and 211 tubes in the "B" S/G required corrective action due to tube degradation. This defect population includes 28 tubes in the "B" S/G that had known defects plugged in prior outages. These tubes were unplugged for full length eddy current examination and were returned to service with a sleeve repair in the degraded region.

The immediate cause of the event was that the "A" and "B" S/G 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 Zntergranular Attack/Stress Corrosion Cracking (EGA/SCC) and Primary Water Stress Corrosion Cracking (PWSCC) attack on S/G tubing.

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

NAC ters III IIIII

NRC Form 944A U.S, NUCLEAR REOULATORY COMMISSION l943I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OME NO 3140&l04 EXPIRES'ISI/SS FACILITY NAME>>I OOCKET NUMEER Ql LER NUMEER (41 ~ AOE IS) wE::, SEOVENTIAL AEVOIOH NVM EA MVM ER R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 90 0 04 000 2 OF 0 7 TEXT IIF uuuu <<MPI <<newt'. uup <<PPppA<< IYIIC Avw SOSASI II TI PRE-EVENT PLANT CONDITIONS The unit was in cold/refueling shutdown for the Annual Refueling Maintenance Outage. All fuel had been removed from the Reactor Vessel. Steam Generator eddy current inspection was in progress.

DESCRIPTION OP EVENT A. DATES AND APPROXIMATE TIMES FOR MAJOR OCCURRENCES:

o April 16, 1990, 1711 EDST: Event date and time. ~ t>

o .':pril 16, 1990, 1711 EDST: Discovery date and

".;.me.

o  :.

made ril to18, the1990, NRC 1000 EDST: Oral Office of Nuclear Reactor notification

."Regulation (NRR).

o A--.ril 21, 1990, 1800 EDST: Steam Generator rap 'rs completed.

o May 1, 1990: Followup report sent to NRC Office of Nuclear Reactor Regulation (NRR).

B. EVENT:

During the 1990 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the >>A>> and >>B>> Westinghouse Series 44 Design 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 May 1989.

'AC FORM 944A 1943 I

NRC Perm SSBA V.S. NUCLEAR REOULATORY COMMISSION (9451 LICENSEE EVENT REPORT (LERI TEXTCONTINUATION APPROVED OMB NO 5150&(95 ExPIRES: BISI/BS PACILITY NAME (11 DOCKET NUMBER (l( LER NUMBER (51 ~ AOE (SI 55OVENTIAL REVISION NVM I4vM eo TEXT R.E..Ginna Nuclear Power Plant lllmore 5oece r reeeeerf. Pee ~ Arr(C forrrr JRLI Sl (Ill o s o o o 2 4 4 90 004 00 03 oF 07 The examination was performed by personnel from Rochester Gas and Electric Corporation (RG&E)f and Allen Nuclear Associates. All personnel had been trained and qualified in the eddy current examination method and had been certified to a minimum of 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 was generated to provide the examination of 1004 of each open (not sleeved, or plugged) steam generator tube from the tube end to the first tube support. Zn addition, 204 of these tubes were selected and examined for their full length as recommended in the EPRZ guidelines. All tubes with previous indications greater than 20% through wall (TW) depth were examined at a minimum to the location of their degradation.

Approximately 20% of all open Row 1 U-bend regions were examined with the Motorized Rotating Pancake Coil (MRPC) between the g6 Tube Support Plate Hot (TSPH) and the 56 Tube Support Plant Cold (TSPC) from the cold leg side.

Results of the above inspections indicated that 75 tubes in the "A" steam generator and 211 tubes in the "B" steam generator (183 new repairs plus 28 previously plugged tubes) required corrective action.

C. INOPERABLE STRUCTURES f COMPONENTS f OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None.

E. METHOD OF DISCOVERY:

The event was apparent after the review of the eddy current examination results.

NRC POIIM 555A (9451

D NRC Form 844A U.S. NUCLEAR REOULATORY COMMISSION 19') 5 I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OM8 NO S)50&104 EXPIRES. 8181/85 FACILITY NAME 111 DOCKET NUMSER Ill I.ER NUMSER 14) PACE (8) 54OVSNTIAL 45. EVIS o NVMSSR -. I" MVM R.E. Ginna Nuclear Power Plant o 5 o o o 244 90 004 00 4 oF 0 7 TEXT lllmcpp N>>ce N>>EMcpd. p>> cHtcNanV HRC %%dmc JR)A 9) I IT)

F. OPERATOR ACTION:

None.

G. SAFETY SYSTEM RESPONSES:

None.

III. CAUSE OF EVENT A. IMMEDIATE CAUSE:

The Immediate Cause of the event was that the <<A<< and steam generator tube degradation was in excess of 'B<<

the Ginna Quality Assurance Manual Reportable Limits.

B. ROOT CAUSE:

The results of the examination indicate that the ZGA and SCC continue to be active within the tube sheet crevice region on the inlet side of each steam generator. As in the past, the IGA/SCC is much more prevalent in the <<B<< steam generator with 108 ZGA indications and 49 SCC indications reported. Zn the

<<A<< steam generator 16 ZGA indications and 22 SCC indications were reported.

The majority of the inlet tube sheet 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 tube sheet crevices forming an alkaline environment. This environment has developed over the years as deposits and active species have reacted with sodium and phosphate, changing a neutral or inhibited crevice into the aggressive environment that presently exists.

In addition to the IGA/SCC in the crevices PWSCC indications continue to be found at the roll transi-tion.

This year there were 23 PWSCC indications in the <<B<<

steam generator and 37 PWSCC indications in the <<A<< .

steam generator.

NAC FOIIM 545A 19851

HRC form 559A I/.9. NI/CLEAR REGULATORY COMMIEEION (9451 UCENSEE EVENT REPORT ILERI TEXT CONTINUATION APPROVEO OME I/O 515OPOIOA EXP/RES: 9/91/85 PACILITY NAME 111 LER NUMEER 191 vEAR 9 9 O I/ C /ITI A L R9V IfIO Ir III/ II C R Ir I/II 9A R.E. Ginna Nuclear Power Plant o s o o o 2 4 4 9 0 0 0 4 000 5 OF 07 TEXT /llmIVO Opoco IP Olvrror/, vro ~/V/ICfo/IIIPS@'p/ IITI Ivo 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 (a) or (b) exceeds the criteria, then results 'ategories 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 the 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 40% through wall, as by eddy current, shall be repaired by -, @ging 'ndicated 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 im", . 'ations resulting from the steam generator tube dec; dation in excess of the Q.A. Manual reportable limits bee..use:

o The degraded tubes were identified and repaired prior to any significant leakage or S/G 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 termination 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.

NRC PORM POOA 19 EE I

NRC Form SSSA U.5. NUCLEAR REGULATORY COMMISSION 19431 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS HO 5150WI05 EXPIRES SISI/95 IEACILITYNAME III LER NUMSER IS) 55GUSNTrAL rr 5 Yl5 lO 1

~ rUM 51 ~ rUrr 51 R.E. Ginna Nuclear Power Plant o s o o o2 44 90 004 0 0 06oF 7 TEXT Illmore opec@ rp orrrrerF. 1>> poooronolHlIC Form PxWSI IITI CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

o Of the 75 degraded tubes in the "A" Steam Generator, 51 tubes were repaired using a Combustion Engineering welded 27" sleeve in the hot leg and these tubes will remain in service.

The remaining 24 tubes were removed from active service by plugging both the hot and cold leg tube ends.

o Of the 211 degraded tubes in the "B" steam 191 tubes were repaired using a 'enerator, Combustion Engineering welded 27" sleeve in the hot leg and these tubes will remain in service.

The remaining 20 tubes were removed Utile from active service by plugging both the hot and cold leg tube ends.

B. ACTION TAKEN OR PLANNED TO PREVENT RECUSANCE:

The occurrence/presence of IGA, SCC and PWSCC is a common PHR Steam Generator problem. "ies with susceptible tubing and partially rolled crevices must deal with this recurring attack on steam generator tubing.

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 primary and/or secondary side environments.

Degraded S/G tubes shall be sleeved or plugged in accordance with the inservice inspection program and accepted industry repair methods.

~ roc po1M 555k 19M I

NRC FoIm 344A V.S. NUCLEAR REGULATORY COMMISSION (943)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMB NO, 3)50M)04 EXPIRESI 813)185 FACILITY NAME 111 OOCKET NUMBER Ill LER NUMBER IS) ~ AGE IS)

YEAR 44OUENTIAL REVISION NUM SR NUM 4 II R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 9 p 004 0 0 7 QF p TEXT llfmCVP g>>44 N fPqviNL MW ~ff)ICAnn 3RLISl )IT)

ADDITIONAL INFORMATION A. FAILED COMPONENTS:

None.

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 LER's76-008, 77-008,78-003, 79-006,79-022, 80-003,81-009, 82-003,82-022, 83-013, and 89-001.

C. SPECIAL COMMENTS:

None.

NRC FORM 344A 1983)

0 E, J

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