ML17263A627

LER 94-006-00:on 940323,SG Tube Degradation Due to Iga/Scc, Caused QA Manual Reportable Limits to Be Reached.Caused by Pwscc.Tubes repaired.W/940422 Ltr

ML17263A627
Person / Time
Site:	Ginna
Issue date:	04/22/1994
From:	Martin J, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-006, LER-94-6, NUDOCS 9404280277
Download: ML17263A627 (12)

Text

ACCELERATl DOCUMENT DISTRIBUTION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9404280277 DOC.DATE: 94/04/22 NOTARIZED: NO DOCKET g FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 R AUTH. NAME AUTHOR AFFILIATION t MARTIN,J.T. Rochester Gas & Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION D

SUBJECT:

LER 94-006-00:on 940323,both "A" & "B" steam Westinghouse Series 44 steam generators, required corrective action due to tube degradation. Cause was PWS.Corrective action: tubes were re P aired.W 940422 ltr.

/

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

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

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 D

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD1-3 PD 1 1 JOHNSON,A 1 1 INTERNAL: 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 N /DSSAQSPLB 1 1 G~~~

NRR/DSSA/SRXB 1 1 02 1 1 RES/DSIR/EIB 1 1 RGN1 FILE 01 1 1 EXTERNAL: EG&G BRYCE,J.H 2 2 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POOREiW 1 1 NUDOCS FULL TXT 1 1 D

D D

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 27 ENCL 27

ITlrlrlr'PIISI1'lR,'I'," 4~

IIC ~

~ ~ ~ I Ir 'oars scarc ROCHFSTER GAS AND ELECTRIC CORPORATION 4 89 EAST AVENUE, ROCHESTER N. K 14649-0001 ROBERT C MECREDY TELEPHONE Vice Presideni AREA COOE i16 546 2700 Cinna Nuclear Producrion April 22, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555

Subject:

LER 94-006, 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 I

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 94-006 is hereby submitted.

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

Very t ly yours, Robert C. Mecredy XC: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector 9404280277 qe0422

/7 PDR ADOCK 0 500024+ PDP 8

NRC FORM 366 U.S NUCLEAR REGULATORY CQIIISSI ON APPROVED BY (IHI NO. 3150-0104 5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY 'WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE IHFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION/

(See reverse for required nunber of digits/characters for each block) WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140.0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.

FAGILITY NAIK (1) R. E ~ Ginna Nuclear Power Plant DOCKET NQQIER (2) PAGE (3)

I 05000244 1OF8 TITLE (6) Steam Generator Tube Degradation Due to IGA/SCC, Causes Quality Assurance Manual Reportable Limits to be Reached EVENT DATE 5 LER NWBER 6 REPORT DATE 7 OTHER FACILITIES INVOLVED 0 SEQUENTIAL REVI S ION FACILITY NAME DOCKET NUMBER HONTH DAY YEAR YEAR MONTH DAY YEAR HUMBER NUMBER 03 23 94 94 --006-- 00 04 22 94 FACILITY NAME DOCKET NUMBER GRATIN G THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR  : Check one or mor e 11 NX)E (9)

N 20.402(b) 20.405(c) 50 '3(a)(2)(iv) 73.71(b)

RRKR 20 '05(a )(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) 000 LEVEL (10) 20.405(a)(1)(ii) 50 '6(c)(2) 50 '3(a)(2)(vii) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Specify in 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) Abstract and in Text, below 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) HRC Form 366A LICENSEE CONTACT FOR THIS LER 12 NAME John T. St. Martin . Director, Operating Experience TELEPHONE NUMBER (Include Area Code)

(315) 524-4446 C(NPLETE ONE LINE FOR EACH C(NPOHENT FAILURE DESCRIBED IN THIS REPORT 13 REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO HPRDS TO HPRDS B AB TBG H314 SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR YES SUBMISSION (If yes, complete EXPECTED SUBMISSIOH DATE). X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines), (16)

During the 1994 Annual Refueling and Maintenance Outage, subsequent to the eddy current examination performed on both the "A" and "B" Westinghouse Series 44 steam generators, 164 tubes in the "A" steam generator and 134 tubes in the "B" steam generator required corrective action due to tube degradation.

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.

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) and Primary Water Stress Corrosion Cracking (PWSCC) attack on steam generator tubing.

This event is NRC Performance Indicator System Cause Code 5.8.4.3 and NUREG-1022 Cause Code (B).

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

HRC FORM 366 (5.92)

HRC FORH 366A U S. NUCLEAR REGULATORY C(NIIISSIOH APPROVED BY (NIB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPOHSE TO COHPLY WITH THIS IHFORHATIOH COLLECTION REQUEST: 50.0 HRS.

FORWARD COHHENTS REGARDING BURDEN ESTIHATE TO L'ICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEHEHT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE 1 DOCKET NNBER 2 LER NNBER 6 PAGE 3 SEQUEHTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 94 -- 006-- 00 2 OF 8 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

PRE-EVENT PLANT CONDITIONS i

The plant was in the cold/refueling shutdown condition for the 1994 Annual Refueling and Maintenance Outage. The Reactor Coolant System (RCS) was depressurized and RCS temperature was approximately 66 degrees F. Steam Generator (S/G) eddy current examination was in progress.

II. DESCRIPTION OF EVENT A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

March 23, 1994, 1500 EST: The number of degraded S/G tubes was known to exceed Ginna Station Quality Assurance (QA)

Manual reportable limits, based on completion of "A>> and "B" S/G hot leg and "A" S/G cold leg inspection program. Event.

date and time.

f March 23, 1994, 1500 EST: Discovery date and time.

March 24, 1994, 1300 EST: Oral notification made to the NRC Office of Nuclear Reactor Regulation (NRR) that the number of degraded S/G tubes exceeded QA Manual reportable limits.

March 26, 1994, 1800 EST: All eddy current programs completed, and the evaluation of the 1994 inservice inspection of S/G tubes completed.

March 27, 1994, 2235 EST: S/G repairs completed.

April 8, 1994: A Special Report was sent to the USNRC, reporting the number of tubes plugged or sleeved in each S/G.

HRC FORH 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY CQIIISS ION APPROVED BY (NB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS IHFORHATIOH COLLECTION REQUEST: 50.0 HRS.

FORllARD CONHEHTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE IHFORNATIOH AND RECORDS HANAGEHEHT BRANCH TEXT CONTINUATION (HHBB 7714), U.S. NUCLEAR REGULATORY COHNISSIOH, UASHIHGTOH, DC 20555-0001 AND TO THE PAPERIIORK REDUCTION PROJECT (3140-0104), OFFICE OF HANAGEHENT AND BUDGET llASHIHGTON DC 20503.

FACILITY NANE 1 DOCKET IRNBER 2 LER NINBER 6 PAGE 3 YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 94 -- 006-- 00 3 OF 8 TEXl'If more space is required, use additional copies of NRC Form 366A) (17)

B. EVENT:

During the 1994 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the >>A>> (EMSOlA) and

>>B>> (EMSOlB) 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 1993 examination.

The examination was performed by personnel from Rochester Gas and Electric (RG&E) and ABB Combustion Engineering. 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 examinations of the >>A>> and >>B>> S/Gs were performed utilizing a 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>> S/Gs 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. I, The inlet or hot leg examination program plan was generated to provide the examination of 100% of each open unsleeved S/G 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 (204 random sample as recommended in the Electric Power Research Institute .(EPRI) guidelines] with the bobbin coil. In addition, 204 of each type of sleeve was examined and the remaining tube was examined full length. All Row 1 and Row 2 U-Bend regions were examined with the MRPC between the g6 tube support plate hot side and the g6 tube support plate cold side from the cold leg side.

HRC FORH 366A (5.92)

NRC FORM 366A U.S. NUCLEAR REGULATORY CQIIISSION APPROVED SY Q(B NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDIHG BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRAHCH TEXT CONTINUATION (MHBB 7714), U.ST NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAJK 1 DOCKET lRNBER 2 LER NINBER 6 PAGE 3 YEAR SEQUENT IAL REVISION R.E. Ginna Nuclear Power Plant M 05000244 94 006-- 00 4 OF 8 EXT (If more space is required, use additional copies of HRC Form 366A) (17)

Results of the above examinations indicated that 164 tubes in the >>A>> S/G required repair (21 new repairs by plugging and 143 new repairs by sleeving). 134 tubes in the >>B>> S/G required repair (31 new repairs by plugging and 103 new repairs by sleeving). Corrective actions were therefore taken for 164 tubes in the >>A>> S/G, and for 134 tubes in the >>B>> S/G.

On March 26, 1994, at approximately 1800 EST, with the RCS depressurized and temperature at approximately 66 degrees F, final review of the 1994 S/G eddy current examination results was completed. Prior to completion of this review (on March 23, 1994, at approximately 1500 EST),

more than one percent of the total tubes inspected were degraded it was evident that (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 QA Manual.

On March 24, 1994, at approximately 1300 EST, oral notification was made to the NRC Office of Nuclear Reactor Regulation pursuant to Appendix B of the QA Manual.

On April 8, 1994, a Special Report listing the number of tubes required to be plugged or sleeved in each S/G, was reported to the NRC, pursuant to Appendix B of the QA Manual.

C~ INOPERABLE STRUCTURES i COMPONENTS I OR SYSTEMS THAT CONTRI BUTED TO THE EVENT:

None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None E. METHOD OF DISCOVERY:

This event was apparent during the review of the >>A>> and >>B>> S/G eddy current examination results.

NRC FORM 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY C(IIIISSI(NI APPROVED BY IGNI NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRAHCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140.0104), OFFICE OF.

MANAGEMENT AHD BUDGET WASHIHGTOH DC 20503.

FACILITY NAME 1 DOCKET NEER 2 LER NIMBER 6 PAGE 3 SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant p5ppp244 94 -- 006-- 00 5 OF 8 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

F. OPERATOR ACTION:

The Control Room operators were notified that the number of degraded tubes exceeded the reportable limits of the QA Manual, and that the NRC (NRR) had already been notified by corporate staff. The 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 examination and repair supervision.

G. SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT A. IMMEDIATE CAUSE:

The immediate cause of the event was the "A" and "B" S/G tube degradation was in excess of the QA Manual reportable limits.

B. ROOT CAUSE:

The results of the examination indicate that Intergranular Attack (IGA) and Intergranular Stress Corrosion Cracking (IGSCC or SCC) continue to be active within the tubesheet crevice region on the inlet side of each S/G. As in the past, IGA/SCC is much more prevalent in the "B" S/G with 74 new crevice indications reported in 1994. In the "A" S/G, 42 new crevice indications were reported in 1994.

In 1993, 41 new crevice indications were reported in the "A" S/G, and 103 new crevice indications were reported in the >>B" S/G. In 1994, 42 new crevice indications were reported in the "A" S/G, and 74 new crevice indications were reported in the "B" S/G. Comparison of 1993 and 1994 results does not suggest any significant change in the rate of tube degradation due to IGA/SCC.

HRC FORM 366A (5-92)

HRC FORM 366A U.S. NUCLEAR REGULATORY CQHISSIOH APPROVED BY W NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY llITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.

FORNARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AHD RECORDS MAHAGEMEHT BRANCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, MASHIHGTOH, DC 20555-0001 AHD TO THE PAPERNORK REDUCTION PROJECT (3140 0104), OFFICE OF MANAGEHEHT AND BUDGET l!ASHIHGTON DC 20503.

FACILITY IUNE 1 DOCKET HINBER 2 LER IWIER 6 PAGE 3 SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 -- 006-- 00 6 OF 8 TEXT (If more spece is reqUIred, use edditionel copies of HRC Form 366A) (17)

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 environment into the aggressive environment that presently exists.

Along with IGA/SCC in the crevices, Primary Water Stress Corrosion Cracking (PWSCC) at the roll transition continued to be active during the last operating cycle. This mechanism was first addressed in 1989, and this year there were 120 roll transition (PWSCC) indications in the "A" S/G and 66 roll transition (PWSCC) indications in the "B" S/G.. These numbers include tubes that may have PWSCC in combination with IGA or SCC in the crevice.

This event is NRC Performance Indicator System Cause Code 5.8.4.3, "Maintenance Equipment Failure", and NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction/Installation."

IV. ANALYSIS OF EVENT This event is reportable in accordance with 10 CFR .50.73, Licensee Event Report System, item (Other) and the QA 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 tube degradation in the "A" and "B" S/Gs exceeded the criterion of (b) which states, "More than 1 percent of the total tubes inspected are degraded (imperfections greater than the repair limit)".

An assessment was performed considering both the safety consequences and implications of this event with the 'following results and conclusions:

There were no operational or safety consequences resulting from the S/G tube degradation in excess of the QA Manual reportable limits because:

HRC FORM 366A (5-92)

Ci NRC FORH 366A U.S. NUCLEAR REGULATORY CQIIISSION APPROVED BY QIS NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COHHENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORNATION AND RECORDS HANAGEKEHT BRANCH TEXT CONTINUATION (HHBB 7714), U.S. NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001 AND'O"THE PAPERWORK REDUCTION PROJECT (3110-0104), OFF ICE OF NANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NANE 1 DOCKET NWBER 2 LER NNNIER 6 PAGE 3 YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 94 -- 006-- 00 7 OF 8 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

~ The degraded tubes were identified and repaired prior to any significant leakage or S/G tube rupture occurring.

~ Even assuming a complete severance of a S/G 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 was assured'at all times.

V. CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS Of the 164 tubes repaired in the "A" S/G, 143 tubes were repaired using a 20 3/4 inch Babcock and Wilcox kinetically welded tubesheet sleeve in the hot leg. All of these 143 tubes will remain in service. The remaining 21 tubes were removed from service by plugging both the hot and cold leg tube ends. A total of 215 tubes in the "A" S/G are currently plugged and 811 tubes are sleeved.

Of the 134 tubes repaired in the "B" S/G, 103 tubes were repaired using a 20 3/4 inch Babcock and Wilcox kinetically welded tubesheet sleeve in the hot leg. All of these 103 tubes will remain in service. 30 tubes were removed from service by plugging both the hot and cold leg tube ends.

The remaining tube was previously plugged and exhibited a crack indication in the plug. The subject tube plug was removed and a new plug installed. A total of 315 tubes in the "B" S/G are currently plugged and 1390 tubes are sleeved.

All the above repairs on the "A" and "B" S/Gs were completed on March 27, 1994, at approximately 2235 EST.

HRC FORH 366A (5-92)

HRC FORM 366A U.S. NUCLEAR REGULATORY CQIIISSIOH APPROVED BY QQI NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDIHG BURDEN ESTIMATE TO ZiICENSEE EVENT REPORT (LER) THE INFORMATION AMD RECORDS MANAGEMENT BRANCH (HNBB 771C), U.S. NUCLEAR REGULATORY CONHISSION, TEXT CONTINUATION WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET 'WASHINGTOM DC 20503.

FACILITY NAME 1 MxxET HIM(BER 2 LER NWBER 6 PAGE 3 YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 94 006-- 00 8 OF 8 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE-The occurrence/presence of IGA, SCC, and PWSCC is a PWR S/G problem. Utilities with susceptible tubing and partially rolled crevices must deal with this recurring attack on S/G tubing.

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 S/G 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. The 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, AO-75-07, R0-75-013, and LERs76-008, 77 008'8003I 79 006'9022I 80003g81009g82003I 82022 83-013,89-001, 90-004,91-005, 92-005, and 93-002.

C. SPECIAL COMMENTS:

A more in-depth final report will be submitted to the NRC, as required by the Ginna QA Manual.

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

HRC FORH 366A (5 92)

I