ML17263A283

LER 93-002-00:on 930404,during 1993 SG Eddy Current Exam, Determined 1% of Total Tubes in SG a & B Degraded.Caused by Iga & IGSCC within Tube Sheet Crevice Region.Tubes Welded Using Tube Sheet sleeve.W/930504 Ltr

ML17263A283
Person / Time
Site:	Ginna
Issue date:	05/04/1993
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-002, LER-93-2, NUDOCS 9305140080
Download: ML17263A283 (22)

Text

ACCELERATED DOCUMENT DISTRISUT105 SYS'j.'j 'M REGULRT% INFORMATION DISTRIBUTIO&STEN (RIDE)

ACCESSION NBR:9305140080 DOC.DATE: 93/05/04 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION BACKUSFW.H. Rochester Gas & Electric Rochester Gas & Electric Corp. Corp.'ECREDY,R.C.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 93-002-00:on 930404,during 1993 SG eddy current exam, d determined 1% of total tubes in SG A & B degraded. Caused by IGA & IGSCC within tube sheet crevice region. Tubes welded using tube sheet sleeve.W/930504 DISTRIBUTION CODE: IE22T NOTES:License Exp date COPIES RECEIVED:LTR ltr.

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

in accordance with 10CFR2,2.109(9/19/72).

l lENCL SIZE: /G 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 JOHNSONFA 1 1 INTERNAL: ACNW 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DORS/OEAB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRIL/RPEB 1 1 NRR/DRSS/PRPB 2 2 1 1 NRR/DSSA/SRXB 1 1 RE FIL 02 1 1 RES/DSIR/EIB 1 1 RG ILE 01 1 1 EXTERNAL: EG&G BRYCEFJ.H '

2 2 L ST LOBBY WARD 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POOREFW. 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTEI CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 30 ENCL 30

'l ROCHESTER GAS AND ELECTRIC CORPORATION 4 89 EAST AVENUE, ROCH STER N.Y.; ""649.0001 ROBERT C. MCCREDY TEr.ErssiONE Vrre Presidenr AREA coDE7)8 546 2700 Cinna Nuclear Produerion May 4, 1993 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 93-002, Steam Generator Tube Degradation Due To IGA/SCC, Causes Quality Assurance 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 93-002 is hereby submitted.

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

Very ruly yours, Robert C. Mecredy xco U.S. Nuclear Regulatory Commission Region 475 I

Allendale Road King 'of Prussia, PA 19406 Ginna USNRC. Senior Resident Inspector 4 OAArg.

9305140080 930504 PDR ADQCK 05000244 J'p JD lpP

i NRC FORA 366 U.S. NUCLEAR REGULATORY COMMISSION

~ (6$ 9) APPROVEO OMB NO. 31504)104 EXPIRES: 4130l92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 50.0 HRS, FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F630), V.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (11 DOCKET NUMBER (2I PAGE 3 R.E. Ginna Nuclear Power Plant 0 5 0 0 0 24 4 1 OFO 9 Steam Generator Tube Degradation Due To 'lGA/SCC, Causes. Quality Assurance Manual Reportable Limits to be Reached EVENT DATE IS) LER NUMBER (6) R EPO RT DATE (7) OTHER FACILITIES INVOLVED (6)

MONTH DAY YEAR YEAR ctsS BEGUBNTIAL REVISION MONTH DAY YEAR FACILITYNAMES DOCKET NUMBER(S)

NUMBER iNS NUMBER 0 5 0 0 0 0 4 4 9 393 0 0 2 0 0 0 504 9 3 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE RLGUIREMENTS OF 10 CF R ('): (Cnecti one or more of tne followinpl (11)

OPERATING MODE (9) 20A02(B) 20A05(c) 60.73(o l (2)(iv) 73.71(B)

POWER 20A06( ~ )(1)(i) 60.36(c) (I ) 60.73( ~ )(2)(v) 73.71(cl LEYEL 0 0 0 20.405(o) l1) liil 50.36(c) l2) 60.73(o) (2)(vii) X OTHER ISpeciyyin AOttrect farrow end ln Teat. HRC Form 20A05( ~ llllliiil 60.7 3( ~ ) (2)(il 60,73(ol(2) (viiil(A) 366A) 20A05(e)(1)(Ivl 60.73(o)(2) liil 50,73(ol(2) (villi(BI 20.405( ~ )(1)(v) 60.73(ol(2) (ill) 50.73( ~ ) l2) l a)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER Wesley H. Backus.- AREA CODE Technical Assistant to the Operations Hanager 3 35 524 -44 46 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC.

TVRER REPORTABLE R%6+3:44Ry'N@'S CAUSE SYSTEM COMPONENT MANUFAC.

TVRER EPORTABI.E TO NPRDS Yg~E)IIT~ne" I, 4

AB TB G H 1 H SUPPLEMENTAL REPORT EXPECTED (14) MONTH kwM!DAY YEAR EXPECTED SUBMISSION DATE (15)

YES (/I yeA complete EXPECTED SVBMISSIDH DATEI X NO ABSTRACT (Limit to tetXJ tpecet, ie., epproalmetely fifteen tinple.rpece rypewritten lineri (16)

During the 1993 Annual Refueling and Maintenance Outage, subsequent to the eddy current examination performed 122 tubes in the on both the "A" and "B" "A"

Westinghouse Series 44 Steam Generators, steam generator and 171 tubes in the "BLS steam generator required corrective action due to tube degradation.

"AL) and "B" steam The immediate cause of the event was that the generator tube degradation was in excess of the Ginna Quality Assurance Manual Reportability Limits.

The underlying cause of the tube degradation is a common steam generator problem of a partially rolled tube sheet crevice with recurring intergranular attack/stress corrosion cracking (IGA/SCC) attack on and Primary Water Stress Corrosion Cracking (PWSCC) steam generator tubing. (This event is NUREG-1022 (X) cause code)

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

NRC Form 366 (6$ 9)

l LJ

NRC FOAM 366A US. NUCLEAA REGULATORY COMMISSION (669) APPROVED 0MB NO. 31500104 EXPIRES: 4/30/92 LICENSEE EYE REPORT ILER) E ATEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATIOIV COLLECTION REOUESTI 500 HRS. FORWARD TEXT CONTINUATION COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BAANCH (F430), U.S. NUCLFAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (315041041. OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 20503.

FACILITY NAME (1) DOCKET NUMBER (21 LEA NUMBER (6) PAGE (3)

YEAR SEOUBNTIAL 'PPI REVISION NUMBER ~SI NUMBBR R.E. Ginna Nuclear Power Plant TEXT IIImove s/rsce Is nq)rr/rer/, Iree edv//I/arrsl NAC Farm 3664's/ (12) 05000244 .9 3 0 0 2 00 02 oF0 9 PRE-EVENT P COND TIONS The plant was in the cold/refueling shutdown condition for the Annual Refueling and Maintenance Outage. Reactor Coolant System (RCS) was depressurized and RCS temperature was approximately 644F. Steam Generator (S/G) eddy current examination was in progress.

DESCRIPTION OP EVENT A. DATES AND APPROXIMATE TIMES OF MAZOR OCCURR19lCES:

o April 4, 1993, 1800 EDST: Event date and time.

o April 4, 1993, 1800 EDST: Discovery date and t.ime.

o April 6, 1993, 1300 EDST: Oral notification made to the NRC Office of Nuclear Reactor Regulation (NRR).

o April 7, 1993, 2128 EDST: Steam Generator repairs completed.

o April 19, 1993: A Special Report was sent to the USNRC.

B. LRGFNT During the 1993 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the "A" and "B" Restinghouse 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 the 1992 examin-ation.

NRC Farm 366A (64)9)

1 NRC FORM 366A U S. NUCLEAR REGULATORY COMMISSION 164)9) APPROVED OMB NO. 31500104 EXPIRES: 4/30/92 ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT ILER) E INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IF@30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT 131500104), OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER 12) LER NUMBER )6) PAGE 13)

YEAR gq>> SEOVENTIAL NUM>>44 ~Ay REVISION x.% NUM>>84 R.E. Ginna Nuclear Power Plant o s o o o 24 4 / 3 002 0 0 3 OF 0 9 TEXT ///14>>i>> <<>>c>> /4 nqvkaf, >>4>> atdtdon>>/NRC %%dnn 3664'4/ ()7)

The examination was performed by personnel from Rochester Gas and Electric (RG&E) and Allen Nuclear Associates, Inc. (ANA). All personnel were trained and qualified in the eddy current examination method and have been certified to a minimum of Level I for data acquisition and Level II for data analysis.

The initial eddy current examination "B" steam generators of the "A" and utilizing a was performed standard bobbin coil technique with data acquisition being performed with the EDDYNET Acquisition System.

The frequencies selected were 400, 200, 100, and 25 KHz.

Additional eddy current examinations of the "A" and "B" steam generators were performed utilizing the Zetec 3-coil Motorized Rotating Pancake Coil (MRPC) probe to examine the roll transition region, selected crevices and support plates. The frequencies used for these examinations were 400, 300, 100, and 25 KHz.

The inlet or hot leg examination program plan was generated to provide the examination of 100% of each open unsleeved steam generator tube from the tube end through the first tube support plate, along with 20%

of these tubes being selected and examined for their full length (20% random sample as recommended in the Electric Power Research Institute (EPRI) guidelines) with the bobbin coil. In addition, 20% of each type of sleeve was examined and the remaining tube examined full length. All Row 1 and Row 2 U-Bend regions were Coil examined with the Motorized Rotating Pancake (MRPC) between the g6 tube support plate hot side and the g6 tube support plate cold side from the cold leg side.

NRC Form 366A )))69)

NRC FOAM 366A U.S, NUCLEAR REGULATORY COMMISSION (64)9) APPROVED 0 M 6 NO. 31504)(04 EXPIRES: 4/30/92 E ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (6) PAGE (3)

REVISION YEAR <5+j SEQUENTIAL NUMBER NUMBER R.E. Ginna Nuclear Power Plant o s o o o 2 4 4 9 3 002 0 0 0 4 QF 0 9 TEXT Ilfmore Epeoe lr rer)eked, Iree er/I/lr/one/NRC %%dnrr 3664'4/ (12)

Results of the above examinations indicated that 122 tubes in the "AEE steam generator required action (i,.e. 121 tubes that were found to have "new" tubesheet crevice indications, and one tube that was obstructed by a foreign object.) 171 tubes in the "B" steam generator required action (i.e. 123 new repairs, plus 48 previously plugged tubes.) Corrective actions were therefore taken for 122 tubes in the "A" steam generator, and for 171 tubes in the "B" steam genera-tor.

On April 4, 1993 at approximately 1800 EDST, with the RCS depressurized and temperature at approximately 644F, final review of the 1993 Steam Generator eddy current examination results was completed. Results of this review indicated that more than one percent of the total tubes inspected are degraded (i.e.

imperfections greater than the repair limit).

Because of the above, the r'esults 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 6, 1993, at approximately 1300 EDST oral notification was made"B" to the NRC Office of NRR pursuant to Appendix of the Ginna Station Quality Assurance Manual.

On April 19, 1993, a Special Report listing the number of tubes required to be plugged or sleeved in each Steam Generator, was reported to the NRC, pursuant to Appendix "B" of the Ginna Station Quality Assurance Manual.

C 'NOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRZBUTED TO THE EVENT:

None.

NRC Form 366A (64)9)

NRC FORM 3SSA US. NUCLEAR REGULATORY COMMISSION (669) APPROVED OMB NO. 3150d')04 EXPIRES: 4/30/92 LICENSEE EVE REPORT (LER) E ATED BURDEN PER AESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION AEOUESTI SOA) HRS. FOAWARD TEXT CONTINUATION REGARDING BURDEN ESTIMATE TO THE RECORDS 'OMMENTS ANO REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PAO/ECT (3)50d)04). OFFICE OF MANAGEMENTAND BUDGET, IVASHINGTON,OC 20503.

FACILITY NAME (1) DOCKET NUMBER (2)

LER NUMBER (6) PAGE (3)

YEAR @~ SEQUENTIAL NUMSER r,.~IS REVISION NUMSER R.E. Ginna Nuclear Power Plant o s o o o 2 44 9 3 002 0 0 0 5 OF 0 9 TEXT /// moro EPEco /I ror/o)od, IIEP odd/dooo/ HRC Form 30549/ (17)

D OTHER SYSTEMS OR SECONDARY FONCTIONS AFFECTED:

None.

METHOD OF DISCOVERY The event was apparent after the final review of the "A" and "B" steam generator eddy current examination results.

OPERATOR ACTION Control Room operators completed the notifications and evaluations required by the A-25.1 (Ginna Station Event Report), submitted for the event by the Steam Generator examination and repair supervision.

SAFETY SYSTEM RESPONSES:

None.

III CAUSE 0 EVI2ÃT 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 Inter-granular Attack (IGA) and Intergranular Stress Corrosion Cracking (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 103 new crevice indications reported in 1993. In the "A" steam generator, 41 new crevice indications were reported in 1993.

NRC Form 368A (04)9)

NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31500104 EXPIRES: 4/30/92 LICENSEE EVE REPORT ILER) E ATEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (3(504)(04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (l) DOCKET NUMBER (2)

LER NUMBER (6) PAGE (3)

~<Q7j SEQUENTIAL ,"SN'EVISION NUMSER N% NUMSER R.E. Gonna Nuclear Power Plant 0 s 2 4 4 0 0 2 0 0 60F 09 0 0 o TEXT /// mare g>>ce /4 mr/Idred. o>> edd/done/HRC Fomr 35649/ l)7)

In 1992, 118 new crevice indications were reported in the "B(l steam generator, and 34 new crevice indications were reported in the "A" steam generator. Comparison of 1992 and 1993 results does not suggest any signi-ficant change in the rate of tube degradation due to IGA/SCC.

The majority of the inlet tubesheet crevice corrosion indications are IGA/SCC of the Mill Annealed Inconel 600 tube material. This form of corrosion is believed to be the result of an alkaline environment forming in the tubesheet crevices. This environment has developed over the years as deposits and active species, such as sodium and phosphate, have reacted, changing a neutral or inhibited crevice into the aggressive environment that presently exists.

Along with IGA/SCC in the crevices, Primary Water Stress Corrosion Cracking (PWSCC) at the roll transi-tion continued to be active during the last operating cycle. This mechanism was first addressed in 1989 and this year there were 20 roll transition (PWSCC) indications in the "Bsl steam generator and 80 roll transition (PWSCC) indications in the "A" steam generator. These numbers include tubes that may have PWSCC in combination with IGA or SCC in the crevice.

Comparing the number of roll transition indications reported in 1992 with the number of these indications reported in 1993, results reveal that significantly fewer roll transition indications were reported in 1993. However, the number of these indications reported in 1992 was unusually high, and represents a data anomoly due to the first-time use of the MRPC technique for examining 1004 of the roll transition and tubesheet crevice region.

large number of pre-existing roll It is believed that a transition indica-tions were first detected by MRPC in 1992, and had not been detected by previous standard bobbin coil techniques. The use of the MRPC probe for examining the roll transition region was continued in 1993.

NR C Form 366A (64)9)

NRC FORM 366A US;NUCLEAR REGULATORY COMMISSION (649) APPROVED OMB NO. 31504104 E X P I R ES: 4/30/92 E ATEO BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT (LER) INFORMATION COLLECTION REQUEST: 508) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR @i: SEQUENTIAL NUMBER

.~ REVISION

'&%5 NUMBER R.E. Ginna Nuclear Power Plant o s o o o 2 4 4 9 3 0 2 0 0 0 7 QF 0 9 TEXT /// Ruuu 4/>>ce /4 rapkaL u>> udder>>l//RC Form 35642/ (12)

ANALYSZS OP EVENT This 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, "Xf 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 reportablein event pursuant to 10 CFR 50.73." The tube degradation the "A" and "B" steam generators exceeded the criterion of (b) which states, "more than 1 percent of the total tubes inspected are degraded (imperfections greater than the repair limit)". This repair limit is defined as, than "Steam Generator tubes that have imperfections greater 40 percent 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 safety consequences or implications resulting from the steam generator tube degradation in excess of the Quality Assurance Manual Reportable Limits because:

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

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 health and safety was assured at all times.

can be concluded that the public's NRC Form 366A (669)

'I NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION

)689) APPROVED OMB NO. 31500104

~ EXPIRES: 4/30/92 LICENSEE EVE REPORT (LER) ATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD TEXT CONTINUATION COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IF@30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PRO/ECT 131504)104), OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON, DC20503.

FACILITYNAME n) DOCKET NUMBER 12)

LER NUMBER (6) PAGE 131 YEAR SEQUENTIAL . REVISION NVMSER NUMBER R.E. Gonna Nuclear Power Plant TEXT illmare <<wee /4 ra)rdred, o s o o o 2 4 4 3 002 0 0 0 8 QF 09 Iree eddldaae/kRC Farm 3664'4/ I IT)

V. CO CTXVE ACTION A ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS o Of the 122 tubes repaired in the "A" steam generator, 51 tubes were repaired using a Combustion Engineering 27" welded sleeve in the hot leg, plus 62 tubes were repaired using a Babcock and Wilcox explosively welded tubesheet sleeve in the hot leg. All of the above tubes will remain in'ervice. The remaining 9 tubes were removed from service by plugging both the hot and cold leg tube ends. A total of 194 tubes in the "A" steam generator are currently plugged and 668 tubes are sleeved.

0 Of the 171 tubes repaired in the "B" steam generator, 153 tubes were repaired using a Babcock and Wilcox explosively welded tube sheet sleeve in the hot leg. All of the above tubes will remain in service. The remaining 18 tubes were removed from service by plugging both the hot and cold leg tube ends. A tota'l of 284 tubes in the "B" steam generator are currently plugged and 1286 tubes are sleeved.

All the above repairs on the "ALE and "B" steam generators were completed on April 7, 1993 at approxi-mately 2128 EDST.

B . ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

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

NRC Form 366A (PIB)

NRC FORM 366A UA. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31504)104 EXPIRES: 4/30/92 LICENSEE EVE REPORT (LER) E ATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504)(04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2)

LER NUMBER (6) PAGE (3)

YEAR PI:y, SEOUENTIAL @X REVISION NUMBER R.E. Ginna Nuclear Power Plane o s o o o 2 4 4 002 0 9 or- 0 9 TEXT llfmare 4/>>ae /4 Iertu/rer/ u>> er/d/I/ane/ JVRC Farm 36643) (17)

R.E. Ginna Nuclear Power Plant will continue, careful monitoring of both primary RCS and secondary side .

water chemistry parameters.

These water chemistry parameters will continue to be evaluated against accepted industry guidelines in order to minimize harmful primary and/or secondary side environments.

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

VI. ADDITIONAL INFORMATION A. FAILED COMPONENTS:

The degraded components are: Inconel 600 Mill Annealed U-Bend tubes having an outside diameter of 0.875 inches and a nominal wall thickness of 0.050 inches. These tubes were manufactured by Huntington Alloy Company.

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, 19-006, 79 022P 80 0036 81 009P 82 003l 82 022/ 83 013/ 89 001,90-004, 91-005, and 92-005.

C. SPECIAL COMMENTS For a more indepth report, refer to the Special Report "Summary Examination Report for the 1993 Steam Generator Eddy Current Inspection at R.E. Ginna Nuclear Power Station", Revision 1, dated April 20, 1993.

As a note of interest, RG&E has ordered new steam generators for R.E. Ginna Nuclear Power Plant to be installed in 1996.

NRC Fomr 366A (64)9)

1 '