ML17262A863

LER 92-005-00:on 920420,determined That 242 Tubes in a SG & 216 Tubes in B SG Required Corrective Action Due to Tube Degradation.Affected Tubes Either Sleeved or Plugged.W/ 920520 Ltr

ML17262A863
Person / Time
Site:	Ginna
Issue date:	05/20/1992
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-005, LER-92-5, NUDOCS 9205280129
Download: ML17262A863 (22)

Text

ACCELERATED DIS~BUTION DEMONS+ATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9205280129 DOC.DATE: 92/05/20 NOTARIZED: NO DOCKET.iz FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION BACKUS,W.H. Rochester Gas & Electric Corp.

MECREDY,R.C. Rochester Gas &'lectric Corp.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 92-005-00:on 920420,annual maintenance exam performed on both S/G "A" & "B." 242 tubes in "A" & 216 tubes in "B" S/G D required C/A due to tube degradation. Caused by recurring IGA/SCC & PWSCC.S/G return to normal.W/920520 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR l. ENCL Q 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 'A D

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD1-3 LA 1 1 PD1-3 PD 1 1 JOHNSON,A 1 1 S,

INTERNAL: ACNW 2 2 AEOD/DOA AEOD/DS P/TPAB 1 . 1 AEOD/ROAB/ DS P NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFB10 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D NRR/DST/SICB8H3 1 1 NRR DST/SPLB8D1 NRR/DST/SRXB 8E 1 1 R & 02 RES/DSIR/EIB 1 . 1 RGN1 FI E 01 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD NRC PDR 1 1 NSIC MURPHY,G.A NSIC POOREiW ~ 1 . 1 NUDOCS FULL TXT R D

A D

D NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOiv1 P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

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

t I i  ; coen aeaee

'OCHESTER GAS AND ELECTRIC CORPORATION ~ e9 EAST AVENUE, ROCHESTER N.K 14649-0001 a

ROBERTC MECREDY TECEPHOkE Vice President AREA CODE 716 546 2700 Cinna Nuclear PsoducRon May 20, 1992 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 92-005, 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 92-005'is hereby submitted.

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

Very truly yours, Robert C. Me red XC'.S. Nuclear Region I Regulatory Commission 475 Allendale Road King of Prussia PA 19406 Ginna USNRC Senior Resident Inspector

~@PE,<7~E W/

9205280129 920520 PDR ADOCK 05000244

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l NAC FORM 368 (64)9) APPROVED OMB NO. 31504)104 EXPIRES: 4(30(92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LERI 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.

NAME (1I DOCKET NUMBER (2) PA E 3

'ACILITY R.E. Ginna

"""" Nuclear Power Plant 050002441OF09 Steam Generator Tube Degradation Due To 1GA/SCC Causes Q.A. Manual Reportable Limits To Be Reached EVENT DATE IS) LER NUMBER (6I >> REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR BEOUENTIAL erg REYIBON MONTH OAY YEAR FACILITYNAMES DOCKET NUMBER(S)

NUMBER NUMBER 0 5 0 0 0 042 0 9 2 9 2 0 0 5 000 5 2 0 9 2 0 5 0 0 0 THIS REPORT IS SUBMITTFO PURSUANT T 0 THE RLQVIREMENTS OF 10 CFR ()I ICnec>> One or more ol the lollowinPI (11 OPERATING MODE (9) 20A02(B) 20A05(c) 50.73(s) (2) (irl 73.71(BI POWER 20.405 (s I (I) ill 50.38(cl(1) 50.73(sl(2)(vl 73.7((cl LEVEL (10) 20.405 (s IllI (III 50.38(cl(2) 50,73(sl(2)(vill X OTHER ISpeclly In Aottrect Below end ln Tenet, HRC Form 20.405(el(1llill) 50.73(s I l2)(l) 50,73(s I (ri II I (Al 366A)

@ '. I (2 P)CP@&m(r

)P a 20A05 (s I (1) (lv) 50.73(el(2)(li) 50,73(s) (2)(rilll(B) 20.405(s) (I I(vl 50.73(s l(2) Bill 50,73(sl(2)(sl LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER Wesley H. Backus AREA CODE

.Technical Assistant to -the Operations Manager 31 552 4-4 446 COMPLETE ONE LINE FOA EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC. EPORTABLE MANUFAC EPORTABLE TURER TO NPRDS SYSTEM COMPONENT TURER TO NPRDS X A B T BGH 31 4 IR4$e($ .k45%4:

BW@~NIINA!r SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED SU 6 M I SS IO N DATE 115)

YES III yer, complete EXPECTED S(IB4IISSION DATEI NO ABSTRACT ILlmlt to 1400 rpecsL i,e., epproslmerely Irfteen tinple spree typewritten linn) (18)

During the 1992 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)., 242 tubes in the "A" S/G and 2'16 tubes in the "B" 'S/G required corrective action due to tube degradation. -

Two additional tubes in each S/G were stabilized and plugged as preventive action, due to anti-vibration bar (AVB) concerns.

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 of a partially rolled tube sheet crevice with recurring S/G'roblem intergranular attack/stress corrosion cracking (IGA/SCC) and Primary Water- Stress Corrosion Cracking (PWSCC) attack. on S/G tubing.

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

NRC Form 366 (64)9)

t I NRC FORM 38SA (589) t LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION US. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150010O EXPIRES: O/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUESTI 50/) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION. PROJECT (315041(H), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1I DOCKET NUMBER l2) LER NUMBER (5) PAGE (3)

SEQUENTIAL YEAR g) NUMBER REVISION NUM854 R.E. Ginna Nuclear Power Plant o s o o o 2 4 4 9 2 0 0 5 0 0 02 o" 0 9 TEXT ///II>>>> <<>>cob JoqokaL Uoo d/I/ooo/HRC FeIIR 3SSAB/(17)

PRE-EVENT PLANT CONDITIONS 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 65 F. Steam Generator (S/G) Eddy Current Inspection was'in progress.

DESCRIPTION OR EVENT A. DATES AND APPROXIMATE TIMES OR MAZOR OCCURRENCES:

0 April 20, 1992, 1200 EDST: Event Date and Time 0 April 20, 1992, 1200 EDST: Discovery Date and Time 0 April 21, 1992, 1330 EDST: Oral Notification made to the NRC Office of Nuclear Reactor Regulation (NRR)..

0 April 26, 1992, 1742 EDST: Steam Generator repairs completed.

0 May 5, 1992: Follow-up written report sent to NRC Office of NRR.

B. EVENT During the 1992 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 the 1991 examination.

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.

NRC Form 355A (609)

I NRC FORM 388A (IS$ 9)

LICENSEE EVENT REPORT ILER)

TEXT CONTINUATION US. NUCLEAR REGULATORY COMMISSION t APPROVED OMB NO. 3) 600104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS .

INFORMATION COLLECTION REQUEST: 503) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS 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 (2) LER NUMBER (6) PAGE (3)

YEAR $ g SEDUENTIAL @> REVISION NUMBER .x: NUM ER R.E. Ginna Nuclear Power Plant 0 s 0 0 o 2 4 4 9 2 005 0 0 0 3 cF 09 TEXT (ffmore Soeoe le rfo/reer/, Iree /frfooo/HRC %%drrrr 36543/(12)

The initial Eddy Current Examination of the "AE( and "B" Steam Generators was performed" utilizing the Zetec 3-Coil Motorized Rotating Pancake Coil (MRPC) probe and the Zetec MIZ-18 Digital Data Acquisition System. The'frequencies selected were 400, 300, 100, and 25 KHZ.

Additional Eddy Current Examinations of the "A"'nd "B" Steam Generators were performed utilizing the.

standard 0.740" or 0.720" O.D. Bobbin Coil Probe. 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 top of the tube sheet with MRPC. Twenty percent of the tubes were selected and examined for their full length (20%

random sample as recommended in the Electric Power Research Institute (EPRI) guidelines) with a bobbin coil. In addition, 204 of .each type of sleeve was examined and the remainder (unsleeved portion) of the tube was examined full length. All Row 1 and Row 2 U-bend regions selected as part of the 204 random sample were examined with the MRPC between the g6 Tube Support Plate Hot Side (TSPH) and the g6 Tube Support Plate Cold Side (TSPC) from the Cold Leg Side.

Results of the above inspections indicated that 244 tubes in the "A" Steam Generator (i.e. 226- new repairs, plus 16 previously plugged tubes, plus 2 tubes stabilized for Anti-Vibration Bar (AVB) concerns) and 218 tubes in the "B" Steam Generator (i.e. 183 new repairs, plus 3 B&W explosive plugs, stabilized plus 30 previously plugged tubes, plus 2 tubes for AVB concerns) required action. Corrective"A" actions were therefore taken for 242 tubes in the S/G, and for 216 tubes in the "B" S/G. Preventive actions (for AVB concerns) were taken for 2 tubes in each S/G.

NRC Form 368A (689)

NRC FORM 366A (669)

LICENSEE EVENT REPORT (LERI U.S. NUCLEAR REGULATORY COMMISSION e APPAOVED 0MB NO. 3)500)OE E XPIR E 5 I 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST.'00 HRS. FORWARD COMMENTS AEGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR AEGULATOAYCOMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3)500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

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

YEAR SEOVENTIAL REVISION NVM ER @r NVM ER R.E. Ginna Nuclear Power Plant 0 6 0 0 0 2 4 4 9 2 0 0 5 00 04 QF 0 9 TEXT ///mom EPEEE /E/PEVked, IIEP J/(/BRE/NRC %%dnR 35663) ((7)

On April 20, 1992 at approximately 1200 EDST, with the RCS depressurized and temperature at approximately 65 F, final review of the 1992 S/G Inspection Eddy Current results was completed. Results of this review indicated that more than 10 percent of the total tubes inspected are degraded (i.e. imperfections greater than 20 percent of the nominal wall thickness) and more than one percent of the total tubes inspected are 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 21, 1992 at approximately 1330 EDST, oral notification was made to the NRC Office of NRR pursuant to Appendix B of the Ginna,Station Quality Assurance Manual.

On May 5, 1992 a follow-up written report of the Steam Generators Inspection and Repairs was sent to the NRC pursuant to'Appendix B of the Ginna Station Quality Assurance Manual.

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

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED None.

E. METHOD OP DISCOVERY:

The event was apparent after the "A" and "B" Steam Generators Eddy final review of the Current Examination results.

NRC PARR 366A (669)

NRC FORM 366A (889)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION (LS. NUCLEAR REGULATORY COMMISSION t APPROVED OMB NO. 3(500)04 EXPIRES: 4(30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUESTI 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150010E), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

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

YEAR SEQUENTIAL REVISION oyer NUMSER NUMBER R.EP Ginna Nuclear Power Plant 0 6 0 0 0 2 4 4 9 20 0 5 0 005 QF0 9 TEXT (I/more epece Je eePJPPIL Iree aB(ro'encl HRC Fcnn 36848) ()7)

P. 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 S/G supervision.

G. SAFETY SYSTEM RESPONSES.

None.

III CAUSE OP EVENT A. IMMEDIATE CAUSE:

The immediate cause of the event was that the "ALI and "BLI 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 Intergranular 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 118 new crevice indications reported. In the "A" Steam Generator, 34 new crevice indications were reported.

The majority of the inlet tubesheet crevice corrosion indications are IGA/SCC of the Mill Annealed Inconel 600 tube material. This form of corrosion's 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.

NRC Form 366A (6J)9)

NRC FORM 386A (BJIS) i LICENSEE EVENT REPORT ILER)

TEXT CONTINUATION (LS. NUCLEAR AEGULATORYCOMMISSION t APPROVED OMB NO. 31500104 EXPIRES'/30/$ 2 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFOAMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (PJ)30), US. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1'HE PAPERWORK REDUCTION PROJECT (3)500104), OFF ICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3)

YEAR PR SEQUENTIAL Sa~r REVISION

..6% NUM ER  % o NVMSER R.E. Ginna Nuclear Power Plant TEXT ////Sore 8/reoe /e /Fr/rer/, Iree ///0'ooe/HRC ForIII 38//AS/(IT) 24 492 00 5 000 OF Along with IGA/SCC in'the crevices, there appears to have been a slight increase in "Primary Water Stress Corrosion Cracking (PWSCC) at the roll transition during the last operating cycle. This mechanism was first addressed in 1989 and this year there were 63 roll transition PWSCC indications in "B" Steam Generator and 189 roll transition (PWSCC) indications in "A" Steam Generator. These numbers include tubes that may have PWSCC in combination with IGA and SCC in the crevice.

XV ANALYSIS OP EVERT 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,

."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 "AL) and "B" Steam Generators exceeded the criterion of (a) which states, "more than 10 percent of the total tubes inspected are degraded (imperfections ,greater than 20 percent of the nominal wall thickness)", and 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, "Steam Generator tubes that have imperfections greater than 40 percent through wall, as indicated by Eddy Current, shall be repaired by plugging or sleeving."

NRC Form 386A (86$ )

. J NRC FOAM 366A (669)

LICENSEE EVENT REPORT (LER)

US. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO. 31500)04 EXPIRES: 6/30/92 ESTIMATED BURDEN PE4 RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50A) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE 4ECORDS TEXT CONTINUATION AND 4EPOATS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK 4EDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 20503.

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

YEAR SEOUENTIAL REVISION NUMBER gP NUMBER R.E.. Ginna Nuclear Power Plant 0 5 0 0 0 24- 4 9 2 0 0 5 00 07 OF 0 9 TEXT /I/more <<>>ce /e nrFI/rer/ Bee a///Noor>>l NRC Forrrl J(ISAB/((2)

An assessment was performed considering the safety

.consequences 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.

0 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, health and safety was assured at all times.

it can be concluded that the public's V. CORRECTIVE ACTION A.

o 'f ACTION TAKEN TO RETURN AFFECTED SYSTEMS. TO PRE-EVENT NORMAL STATUS:

the 244 tubes in the "A" Steam Generator, 36 tubes were repaired using a Combustion Engineering 27" welded sleeve in the Hot Leg, plus 194 tubes were repaired using a Babcock and Wilcox explosively welded tubesheet 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 185 tubes in the "A" Steam Generator are currently plugged and 555 tubes are sleeved.

NRC Form 366A (689)

I NRC FORM 3SSA (W)

LICENSEE EVENT REPORT ILER)

TEXT CONTINUATION US. NUCLEAR REGULATORY COMMISSION l APPROVEOOMB NO.3I50010O i EXPIRES: O/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (PJ)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACII.ITY NAME (1I DOCKET NUMBER (2) LER NUMBER (Sl PAGE (3)

YEAR SEOVENTIAL REVISION NVM ER NUMBER R.E. Ginna Nuclear Power Plant TEXT /I/more e/>>ce /or/q/rer/ Iree <<IChE/<<>>/ NRC %%de 3SBAB/ ( IT) 0 S 0 0 0 2 4 4 9 2 0 0 5 000 OF 0 Of the 218 tubes in the "B" Steam Generator, 186 tubes were repaired using a Combustion Engineering 27" welded sleeve in the Hot Leg, plus 9 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 23 tubes were removed from =service by plugging both the Hot and Cold Leg tube ends. A total of 313 tubes .in the "BL( Steam Generator are currently plugged and 1134 tubes are sleeved.

All the above repairs on the "A" and "BL( Steam Generators were completed on April 26, 1992 at approximately 1742 EDST.

B. ACTION TAKEN OR PLANNED TO PRVTENT 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.

R.E. Ginna Nuclear Power Plant will continue careful monitoring of both primary RCS and'econdary 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.

ADDITIONAL INFORMATION A. 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. These tubes were manufactured by Huntington Alloy Company.

'

NR C Form 3SSA (589)

NRC FORM 388A US. NUCLEAR REGULATORY COMMISSION (669) APPROVED OMB NO. 3)504104 E XP I R ES: E/30/IQ ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 505) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F830), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1)IE PAPERWORK REDUCTION PROJECT (3150010E). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, OC 20503.

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

YEAR S~ SEQVENTIAL j>gj NVM@ER REVISION

i. '/ NVMSER R.E. Ginna Nuclear Power Plant 0 6 0 0 0 2 4 4 9 0 0 5 0 0 0 9 "0 9 TEXT ///IIRvo EPooo JI ISA(RR/, vJo //I/ooo/NRC Form 36643/ (17)

B. PREVIOUS LERs OM 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, AO-75-07, R0-75-013, and LERs76-008, 77-008,78-003, 79-006, 79 022/ 80 003/ 81 009/ 82 003/ 82 022/ 83 013/ 89

=-

001,90-004, and 91-005.

C SPECIAL COMMEHTS:

For a more indepth report, refer to the "1992 Steam Generator Eddy Current Inspection" Summary Examination Report sent to the NRC May 5, 1992.

NRC Form 368A (BJ)9)

S ~ ~