ML17263B054

LER 95-004-00:on 950407,SG Tube Degradation Occurred Due to Iga/Scc That Caused QA Manual Reportable Limits to Be Reached.Sleeved or Plugged Affected Tubes W/Accepted Industry Repair Methods

ML17263B054
Person / Time
Site:	Ginna
Issue date:	05/08/1995
From:	St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263B053	List: ML17263B052 ML17263B054
References
LER-95-004, LER-95-4, NUDOCS 9505180167
Download: ML17263B054 (12)

Text

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150 ~ 0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE IHFORHATION AND RECORDS MANAGEHENT 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 (3150-0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) R.E. Ginna Nuclear Power Plant DOCKET NUMBER (2) PAGE'3) 05000244 'I OF 12 TITLE (4) Steam Generator Tube Degradation Due to IGA/SCC, Causes Quality Assurance Manual Reportable Limits to be Reached EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER HONTH DAY YEAR YEAR HONTH DAY YEAR NUHBER NUMBER 04 07 95 95 --004-- 00 05 08 95 FACILITY NAME DOCKET NUHBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE (9) 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73. 7'I (b)

POWER 20.405(a )(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) 000 LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) X Oj'HER 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 - Technical Assistant TELEPHONE NUMBER (Include Area Code)

(315) 524-4446 COHPLETE ONE LINE FOR EACH COHPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEH COHPONENT HANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS TBG H314 SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED HOHTH DAY YEAR YES SUBMISSION (If yes, complete EXPECTED SUBHISSION DATE).

X NO DATE (15)

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

During the 1995 Annual Refueling and Maintenance Outage, subsequent to the eddy current examination performed on both the "A" and "B" Westinghouse Series 44 steam generators, 88 tubes in the "A" steam generator and 122 tubes in the "B" steam generator required corrective action due to tube, sleeve, or plug 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.

NRC FORM 366 (5-92) 9505180167 950508 PDR ADOCK 05000244 8 PDR

f NRC FORM 3666 U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50 ' HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTONR DC 20555 0001 AND TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 N M 2 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

PRE-EVENT PLANT CONDITIONS:

The plant was in the cold/refueling shutdown condition for the 1995 Annual Refueling and Maintenance Outage. The Reactor Coolant System (RCS) was depressurized and RCS temperature was approximately 100 degrees F ~ Steam Generator (S/G) eddy current examination was in progress'I.

DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

April 7, 1995, 1700 EDST: The number of degraded S/G tubes was known to exceed Ginna Station Quality Assurance (QA)

Manual reportable limits. Event date and time.

April 7, 1995, 1700 EDST: Discovery date and time.

April 10, 1995, 1430 EDST: 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.

April 13, 1995, 1200 EDST: All eddy current programs completed, and the evaluation of the 1995 inservice inspection of S/G tubes completed.

April 15, 1995, 2100 EDST: S/G repairs completed.

April 24, 1995: A Special Report was sent to the USNRC, reporting the number of tubes plugged or sleeved in each S/G.

NRC FORH 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 NRS.

FORWARD COMHENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HAHAGEHENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MAHAGEMENT AND BUDGET WASHINGTON DC 20503.

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

YEAR Q TIAL REVIS' R.E. Ginna Nuclear Power Plant 05000244 3 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

B. EVENT:

During the 1995 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the "A" (EMS01A) and "B" (EMS01B) 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 1994 examination.

The examination was performed by personnel from Rochester Gas and Electric (RG&E) and Asea Brown Boveri - Combustion Engineering (ABB-CE). All personnel were trained and qualified in the eddy current examination method and have been certif ied 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.

Sleeves were inspected using the Zetec "Plus Point" probe, which allows for improved inspection capability of the parent tube behind the sleeve. Since this advanced probe is more sensitive, it also can identify volumetric indications on the sleeve inside surface.

NRC FORH 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVEO BY OHB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COHHENTS REGARDING'URDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSIOH, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUHBER (2) LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M

4 OF 95 004 00 12'EXT (If more space is required, use additional copies of NRC Form 366A) (17)

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 [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 examined with the MRPC between the N6 tube support plate hot side and the N6 tube support plate cold side from the cold leg side.

Results of the above examinations indicated that 88 tubes in the "A" S/G required action (13 new repairs by plugging and 75 new repairs by sleeving). 122 tubes in the "B" S/G required action (31 new repairs by plugging and 91 new repairs by sleeving). Corrective actions were therefore taken for 88 tubes in the "A" S/G and for 122 tubes in the Il ll B S/G On April 7, 1995, at approximately 1700 EDST, with the RCS depressurized and temperature at approximately 100 degrees F, final review of the 1995 S/G eddy current examination results was completed. More than one percent of the total tubes inspected were degraded (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 April 10, 1995, at approximately 1430 EDST, oral notification was made to the NRC Office of Nuclear Reactor Regulation pursuant to Appendix B of the QA Manual.

On April 24, 1995, 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.

NRC FORH 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB 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 REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSION, WASHINGTON, OC 20555-0001, ANO TO THE PAPERWORK REDUCTIOH PROJECT (3150-0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Po~er Plant 05000244 NUM 5 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED 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.

F. OPERATOR ACTION:

The Control Room operators were notified that the number of degraded tubes exceeded the reportable limits of the QA Manual.

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.

NRC FORH 366A (5-92)

I

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

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

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

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 6 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

B. ROOT CAUSE:

1. TUBE DEGRADATION 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 80 new crevice indications reported in 1995. In the "A" S/G, 35 new crevice indications were reported in 1995.

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 1994 and 1995 results does not suggest any significant 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. 'his 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 transition continued to be active during the last operating cycle. This mechanism was first addressed in 1989 and this year there were 60 roll transition (PWSCC) indications in the "A" S/G and 32 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." The tube degradation does not meet the NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", definition of a "Maintenance Preventable Functional Failure".

HRC FORM 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS.

FORWARD COMMENTS REGARDIHG BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSIONS WASHINGTON, DC 20555-0001,'ND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MAHAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)

YEAR L ON R.E. Ginna Nuclear Power Plant 05000244 7 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

2. SLEEVE INDICATIONS Eddy current examination of 515 ABB-CE welded sleeves was performed using a "Plus Point" probe. This examination identified eleven (11) sleeves with inside surface (ID) volumetric indications at the upper weld location.

Subsequent visual examination (VT) of the upper weld joint for these 11 sleeves identified five (5) sleeves with weld "tail-off" indications, three (3) sleeves with pinholes in the weld, two (2) sleeves with blowholes in the weld, and one (1) sleeve with no weld in the upper weld zone.

The eight (8) sleeves with weld tail-off and pinhole indications were left in service since these conditions did not affect either the structural integrity or the leak tightness of the upper weld. One of the sleeves with a blowhole indication was also left in service. The blowhole was located in the upper portion of the weld. Sufficient weld material existed below the blowhole location to provide both sleeve/tube structural integrity and a leak tight weld.

The second blowhole was located in the lower portion of the sleeve weld. Sufficient fusion existed for weld structural integrity. However, the possibility existed for a leak path to develop from the sleeve to the secondary side of the S/G.

Although the sleeve could have been accepted as a leak-limiting sleeve, as a precautionary measure, the sleeve was repaired by plugging.

The one sleeve with no upper weld was identified as being a curved sleeve originally installed in 1990. In discussion with ABB-CE it was concluded'hat the lack of any weld in the sleeve was a result of an equipment problem with the flexible welding tool used to weld curved sleeves. Consequently, all installed curved sleeves were examined either by a review of the Plus Point data or by performing a VT inspection to verify the presence of an upper weld on the sleeve ID. This re-examination process discovered a second curved sleeve with no upper sleeve weld.

NRC FORH 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OHB NO. 3150.0104 (5 92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COHMENTS REGARDIHG BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEHENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

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

YEAR SEQUENTIAL REVISION R.E. Qinna Nuclear Power Plant 05000244 8 OF 12 95 004 00 TEXT (If more space is required, use additions( copies of NRC Form 366A) (17)

Both curved sleeves with missing upper welds were installed in 1990, and both welds were examined and accepted by UT to confirm the existence of weld fusion. The UT examinations were performed by the same UT inspector. As a result of the failure of the UT examination to discover the lack of an upper weld, all of the 113 sleeves examined by the one Level II UT inspector (in 1990) were re-examined by a different ABB-CE Level III UT inspector examination process discovered an additional six (6) sleeves in 1995. This UT re-that had inadequate weld fusion. All eight of the sleeves discovered with either missing or inadequate weld fusion during the 1995 outage were repaired by plugging.

The condition of these 8 ABB-CE sleeves was not identified during the original installation inspection because of the inexperience of the Level II UT inspector who performed the examination. The lack of fusion indications at the weld location resulted in the UT inspector mistaking the sleeve expansion region for a weld.

NRC FORM 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COHHISSION APPROVED BY OHB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS ~

FORWARD COMMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB /714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

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

SEOUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M 9 OF 12 95 004 00 TDT (If more space is required, use additional copies of NRC Form 366A) (17)

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:

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 RE 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.

Sleeve indications do not present any operational safety consequences because there was no major defect in the design, construction, or installation of the ABB-CE welded sleeve which would have resulted in a structural failure of the installed sleeve. Additionally, the lack of adequate fusion during the installation process does not prevent the installed sleeve from functioning as a leak-limiting sleeve.

The parent tube would remain constrained by the tubesheet and the installed sleeve. Therefore, RCS pressure boundary integrity has been maintained.

NRC FORM 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB 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 REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEMENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON OC 20503.

FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)

SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 UMB 10 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

~ No significant increase in radiation exposure or release to the general public would have occurred. Based on the maximum calculated leakage where a sleeve has a completely missing weld, the resulting primary to secondary leakage is well below the bounding leakage for the existing Ginna S/G tube rupture accident itallcantimes.

analysis'ased on the above, be concluded that the public's health and safety was assured at CORRECTIVE ACTION:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

Of the 88 tubes repaired in the "A" S/G, 75 tubes were repaired using a 20 3/4 inch Babcock and Wilcox kinetically welded tubesheet sleeve in the hot leg. All of these 75 tubes will remain in service. The remaining 13 tubes were removed from service by plugging both the hot and cold leg tube ends. A total of 228 tubes in the "A" S/G are currently plugged and 885 tubes are sleeved.

Of the 122 tubes repaired in the "B" S/G, 91 tubes were repaired using a 20 3/4 inch Babcock and Wilcox kinetically welded tubesheet sleeve in the hot leg. All of these 91 tubes will remain in service. Of the remaining 31 tubes, 29 tubes were removed from service by plugging both the hot and cold leg tube ends. The other 2 tubes were previously plugged and exhibited minor leakage indications. For these 2 tubes, the tube plugs were removed and new plugs installed.

A total of 343 tubes in the "B" S/G are currently plugged and 1464 tubes are sleeved.

All sleeves that were inspected by the inexperienced Level II UT inspector in 1990 were re-examined in 1995 by a different ABB-CE Level III UT inspector, to verify acceptable weld fusion between the sleeve and the parent tube.

NRC FORH 366A (5-92)

NRC FORH 366A U.S. NUCLEAR REGULATORY COMHISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714)i U S NUCLEAR REGULATORY COMMISSIONS WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

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

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 11 OF 12 95 004 00 TEXT (If more space is required, use additional copies of NRC 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.

To ensure that this lack of experience was limited to the 1990,installation of ABB-CE welded sleeves, the records all of the lead UT inspectors used by ABB-CE from 1986 of to 1993 at Ginna Station were reviewed. This review determined that, excluding the one Level II UT inspector used in 1990, all of the other UT inspectors used at Ginna Station had prior experience with the performance of UT examination of these sleeves' review of ABB-CE's complete sleeve installation and inspection program since 1984 has shown that this Level II UT inspector was not employed by ABB-CE for any other sleeving program at any other utility.

NRC FORM 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVEO BY OMB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS. INFORHATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COHMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 N M 12 OF 12 95 -- 004 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

VI. ADDITIONAL INFORMATION:

A. FAILED COMPONENTS:

The degraded tubes 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 AO-74-02, A0-75-07, RO-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, 91-005,92-005, 93-002, and 94-006.

C. SPECIAL COMMENTS:

A more detailed 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.

NRC FORH 366A (5-92)