ML17263B105

LER 95-005-00:on 950607,FW Isolation on High SG Level Occurred.Caused by Decrease in Instrument Air Pressure Due to an Air Leak in Containment.Fw Flow Switched to Manual control.W/950707 Ltr

ML17263B105
Person / Time
Site:	Ginna
Issue date:	07/07/1995
From:	Mecredy R, St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
LER-95-005, LER-95-5, NUDOCS 9507120033
Download: ML17263B105 (10)

Text

yPRIORI TY 1 ~

(ACCELERATED RIDS PROCESSING)

I

.REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9507120033 DOC.DATE: 95/07/07 NOTARIZED: NO DOCKET g FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 P AUTH. NAME AUTHOR AFFILIATION ST MARTIN,J.T. Rochester Gas & Electric Corp.

MECREDY,R.C. Rochester Gas & Electric Corp. R RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Project Directorate I-1 (PDl-1) (Post 941001)

I

SUBJECT:

LER 95-005-00:on 950607,FW isolation on high SG level 0 occurred. Caused by decrease in instrument air pressure due to an air leak in containment.FW flow switched to manual control.W/950707 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED LTR ENCL 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 T RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 JOHNSON,A 1 1 INTERNAL: AEOD S B 2 2 AEOD/SPD/RRAB 1 1 1 1 NRR/DE/ECGB 1 1 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DISP/PIPB 1 1 NRR/DOPS/OECB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DSSA/SPLB '1 1 NRR/DSSA/SPS B/B 1 1 NRR/DSSA/SRXB 1 1 RES/DS IR/EIB 1 1 RGN1 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LI TCO BRYCE i J H 2 2 NOAC MURPHY i G A 1 1 NOAC POORE i W 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

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PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D8 (415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

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eGCHE51FR G~s~wo ElEcTec coepcw "crI ~ 8w ~Sr ~vEriuE eGc~EsTFe. MY ~sd~a CATE July 7, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson PWR Project Directorate I-1 Washington, D.C. 20555

Subject:

LER 95-005, Instrument Air Leak in Containment Causes Feedwater Isolation R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (iv), which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS) ", the attached Licensee Event Report LER 95-005 is hereby submitted.

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

Very truly yours, Robert C. Mecredy xc: U.S. Nuclear Regulatory Commission Mr. Allen R. Johnson (Mail Stop 14B2)

PWR Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector

'1 P r 9507l20033 950707 PDR ADOCK 05000244 S ma

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OHB NO. 3150-0'104 (5-92) EXPIRES 5/3'I/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY llITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORllARD COMMENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (HNBB 7714), U.S. NUCLEAR REGULATORY CCHHISSION, for required nunbcr of digits/characters for each block) MASNINGTON, DC 20555-0001 AND TO THE PAPERWORK (See reverse REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET NASHINGTON DC 20503.

FAGILITY NAHE (1) R . E . G irma Nuclear Power P lant DOCKET NUHBER 05000244 (2) PAGE (3) 10F8 TITLE (4> Instrument Air Leak in Containment Causes Feedwater Isolation EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR HOHTN DAY YEAR NUMBER NUMBER 06 07 95 95 --005-- 00 07 07 FACILITY HAHE DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREHENTS OF 10 CFR 5: (Check one or mor e) (11)

MODE (9)

N 20.402(b) 20.405(c> 50.73(a)(2)(iv) 73.71(b) 20.405(a )(1)(i) 50.36(c)(1) 50 '3(a)(2)(v) 73.71(c)

P(NER 097 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER LEVEL (10) 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Spec'I fy In 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) Abstract below and in Text, 20.405(a)(1)(v) 50 '3(a)(2)(iii) 50 '3(a)(2)(x) NRC Form 366A)

LICENSEE CONTACT FOR THIS LER (12)

NAME John T. St. Hartin - Tcchnical Assistant TELEPHONE NUHBER (Include Area Code)

(716) 77'1-3641 COHPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IH THIS REPORT ('13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT HANUFACTURER CAUSE SYSTEM COMPONEHI'ANUFACTURER TO NPRDS TO NPRDS B LD PSF 0000 N SUPPLEHENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUSHI SS I ON X HO (If yes, complete EXPECTED SUBMISSION DATE). DATE (15)

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

On June 7, 1995 at approximately 1905 EDST, with the plant at approximately 97% steady state reactor power and the Instrument Air system isolated to Containment due to an air leak, Feedwater Isolation on high Steam Generator level occurred when levels went above 67% narrow range level in the Steam Generators.

Immediate corrective action was to manually control feedwater flow until levels in the Steam Generators were restored to their normal operating band.

The underlying cause of the inability to control Steam Generator levels was a decrease in Instrument Air pressure due to an Instrument Air leak in Containment, followed by restoration of air pressure with a demand signal to fully open main feedwater regulating valves.

This event is NUREG-1022 Cause Code (B).

Corrective action to preclude repetition is outlined in Section V.B.

NRC FORM 366 (5-92>

NRC FORH 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 INFORHATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEHENT BRANCH (MNBB 1714), U.S. NUCLEAR REGULATORY COHHISSION, TEXT CONTINUATION WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3180-0104), OFFICE OF MANAGEMENT AND BUDGET 'WASHINGTON DC 20503.

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

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 NUHBER NUMBER 95 -- 005-- 00 2 OF 8 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

PRE-EVENT PLANT CONDITIONS:

The plant was at approximately 97% steady state reactor power with no significant activities in progress. A soldered joint connection in the Instrument Air system in Containment failed, causing a decrease in Instrument Air pressure, and loss of control air to air-operated components, including the main feedwater regulating valves.

II. DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

o June 7, 1995, 1856 EDST: Instrument Air system leak occurs in Containment.

o June 7, 1995, 1902 EDST: Instrument Air to Containment is isolated, restoring normal air pressure to air-operated components outside Containment.

o June 7, 1995, 1905 EDST: Steam Generator (S/G) levels increase above 67%. Event date and time.

o June 7, 1995, 1905 EDST: Discovery date and time.

June 7, 1995, 1910 EDST: "A" and "B" S/G levels restored to pre-event normal operating band.

B. EVENT:

On June 7, 1995, at approximately 1856 EDST, with the plant at approximately 978 steady state reactor power, a soldered joint connection on a two inch Instrument Air (IA) line in Containment (CNMT) failed, resulting in leakage from the IA system and decrease in IA pressure. This decrease in IA pressure resulted in loss of control air to air-operated components, with valves beginning to travel to their respective "fail" positions. Among these components were the two main feedwater regulating valves (MFRV) (fail closed) which drifted towards the closed position as IA pressure at the valve actuator decreased.

NRC FORH 366A (5-92)

366A U.S. NUCLEAR REGULATORY COHNISSIOH APPROVED BY OHB NO. 3150-0104

.NRC FORH (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATIOH COLLECTION REQUEST: 50.0 HRS.

FORWARD COHHENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEHEHT BRANCH (HNBB 7714), U.S. HUCLEAR REGULATORY COHHISSION, TEXT CONTINUATION WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3180-0104), OFFICE OF HAHAGEHEHT AND BUDGET WASHINGTON DC 20503.

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

SEQUENTIAL REVI SION YEAR R.E. Ginna Nuclear Power Plant 05000244 NUHBER NUHBER 3 OF 8 95 -- 005-- 00 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

At approximately 1902 EDST, the Control Room operators had diagnosed the probable location of the IA leak, and closed the IA CNMT Isolation valve, AOV-5392. With the closure of AOV-5392, the leak was isolated, and normal IA pressure was restored to components outside CNMT.

Due to the MFRVs drifting closed, feedwater (FW) flows and Steam Generator (S/G) levels decreased, resulting in an increasing "demand" signal to the MFRVs. Isolation of the IA leak resulted in restoration of IA pressure, and the MFRVs opened fully, responding to the increased demand signal. At the time the MFRVs went to the full open position, level was approximately 25% in the "A" S/G and 40% in the "B" S/G. The increase in FW flow resulted in increasing level in the "A" and "B" S/Gs.

Within three minutes narrow range level in the "B" S/G had increased to cause FW Isolation on high level "B" in the "B" S/G (S/G level >/= 67 0 narrow range level). The MFRV closed in response to this FW Isolation signal as designed, and reopened when level decreased below 67%. For the next ninety seconds, there were several occurrences of FW Isolation for the "B" S/G as level"A"cycled around 67%. During this time narrow range level in the S/G also increased to cause FW Isolation on high level in the "A" S/G. For approximately twenty seconds, there were occurrences of FW Isolation for the "A" S/G as level cycled around 67%. This short term S/G level transient continued until the Control Room operators took manual control to "A" restore S/G levels. At approximately 1910 EDST, levels in the and "B" S/Gs were restored to their normal operating band.

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

The decrease in IA pressure resulted in loss of control air to air-operated components.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

Due to the failed soldered joint connection in CNMT, and the subsequent isolation of IA to CNMT, air-operated components in CNMT failed to their respective "fail" positions. These included several valves and ventilation dampers. In addition, the Reactor Compartment Cooling (RCC) fan motor tripped when the associated dampers failed closed.

HRC FORH 366A (5-92)

JIRC 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 COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AHD RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION WASHINGTON, DC 20555-0001 AHD 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)

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 NUMBER HUMBER 95 -- 005-- 00 4 OF 8 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

For air-operated components outside the CNMT, there was a decrease in IA pressure throughout the plant for six minutes, until IA to CNMT was isolated. During this time, numerous air-operated components outside CNMT started to travel to their respective "fail" positions. With the exception of the MFRVs, this loss of control air did not adversely affect the ability of the Control Room operators to maintain plant conditions.

E. METHOD OF DISCOVERY:

This event was immediately apparent due to alarms and indications in the Control Room. In particular, Main Control Board annunciators C-17 (CONTAINMENT VENT SYSTEM) and H-8 (INSTRUMENT AIR LO PRESS 100 PSI) alarmed, indicating a problem with IA in CNMT.

OPERATOR ACTION:

The Control Room operators responded to Main Control Board annunciators C-17, H-8 and H-16 (INSTRUMENT AIR COMP), and referred to Alarm Response Procedures C-17, H-8 and H-16. They entered Abnormal Operating Procedure AP-IA.l (LOSS OF INSTRUMENT AIR). The Control Room operators requested that the auxiliary operator start the standby diesel-driven air compressor.

Following the steps of AP-IA.1 and with the knowledge that abnormal alarms were received on CNMT systems prior to those on secondary systems, the Control Room operators isolated IA to CNMT by closing the IA CNMT Isolation valve (AOV-5392). This action isolated the leak from the rest of the IA system, and IA pressure increased to normal pressure in the rest of the system.

After the FW Isolation, the Control Room operators transferred control of the MFRVs to "manual" to restore S/G levels to their normal operating band. When S/G levels and FW flows were stabilized, they transferred control of the MFRVs back to "automatic".

HRC FORM 366A (5-92)

JIRC FORA 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150 0104 (5-92) EXP I RES 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 (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION WASHINGTONt 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)

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 NUMBER NUMBER 5OF8 95 -- 005-- 00 EXT (If more space is required, use additional copies of NRC Form 366A) (17)

With loss of letdown flow, the operators manually decreased charging flow to minimum flow, and secured one charging pump.

The Shift Supervisor made a decision to initiate a power reduction until the leak was located, isolated, and IA pressure returned to normal throughout the system. At approximately 1908 EDST, a power reduction was started at one percent per minute, per Normal Operating Procedure 0-5.1 (LOAD REDUCTIONS).

Subsequently, the Control Room operators notified maintenance personnel and higher supervision.

An auxiliary operator and Radiation Protection technician conducted a CNMT entry, at power in an attempt to identify and isolate the leak. The leak was located on the main two inch IA header in CNMT. A temporary repair was made to the failed joint connection. This repair enabled the Control Room operators to restore some pressure to the IA system in CNMT, sufficient to allow operation of selected valves and ventilation dampers.

The NRC Operations Center was notified at approximately 2211 EDST, per 10CFR50.72 (b) (2) (ii), non-emergency four hour notification.

G. SAFETY SYSTEM RESPONSES:

The MFRVs and MFRV bypass valves closed automatically as a result of the FW Isolation signals. Ventilation dampers for the Containment Recirculation Cooling fans and RCC fans failed to their respective safeguards positions. CNMT Isolation valves for charging and letdown also failed to their safeguards positions.,

III. CAUSE OF EVENT:

A. IMMEDIATE CAUSE:

The immediate cause of the FW Isolation was level in the S/Gs being )/= 67%. The high level was caused by increased FW flows when the MFRVs went full open in response to the valve demand signal. This situation resulted in overfeeding the S/Gs.

NRC FORM 366A (5.92)

JIRC 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 MITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORllARD COMHENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS HANAGEMEHT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, HASHINGTONt DC 20555 0001 AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.

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

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 HUMBER NUMBER GOF8 95 -- 005-- 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

B. INTERMEDIATE CAUSE:

The intermediate cause of the open valve demand signal for the MFRVs was decreased FW flows and S/G levels as the MFRVs drifted toward the closed position as IA pressure decreased.

C. ROOT CAUSE:

The underlying cause of the decrease in IA pressure was the failure of a soldered joint connection in a two inch IA line in CNMT. This was caused by insufficient insertion of the pipe into a fitting during original construction. This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction /

Installation". This event 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".

IV. ANALYSIS OF EVENT:

This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (iv), which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including "A" the reactor protection system (RPS)". The FW Isolation of the and "B" S/Gs was an automatic actuation of an ESF system.

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 or implications attributed to the FW isolations because:

The FW isolations occurred at the required S/G level.

o S/G levels were quickly stabilized and manual control of MFRVs was accomplished to mitigate any consequences of the event.

Based on the above, it at can be concluded all times.

that the public's health and safety was assured NRC FORM 366A (5-92)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 ARC FORM 366A EXP I RES 5/31/95 (5-92)

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

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

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

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 NUMBER NUMBER 95 -- 005-- 00 7 OF 8 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

V. CORRECTIVE ACTION:

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

o MFRVs were returned to automatic after S/G levels were restored to their pre-event normal operating band.

o A temporary repair was made to the failed joint connection in CNMT. This temporary repair enabled the Control Room operators to restore some pressure to the IA system in CNMT, sufficient to allow operation of selected valves and ventilation dampers.

After letdown flow was restored and the Control Room operators could control primary system volume, the load reduction was stopped.

Maintenance personnel installed a temporary modification designed by Engineering, which permitted isolation of the failed joint for permanent repair, while maintaining an air supply to the letdown valves. This allowed the operators to maintain letdown flow.

o Maintenance personnel performed the permanent repair by replacing the failed joint connection and adjacent pipe sections, and removed the temporary modification. At the completion of these activities, normal IA was restored to CNMT.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

A sample of joint connections in the IA system will be examined by non-destructive techniques to confirm adequate pipe insertion into fittings.

o This event will be evaluated and compared against Plant Simulator response under controlled conditions. Any lessons learned and enhancements to the control of primary system pressure will be identified, and procedures changed, as appropriate.

NRC FORM 366A (5-92)

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

FORUARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AHD RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION MASHINGTON, DC 20555-0001 AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET UASHINGTON DC 20503.

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

SEQUEHTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 NUMBER NUMBER 95 -- 005-- 00 8 OF 8 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

VI. ADDITIONAL INFORMATION:

A. FAILED COMPONENTS:

There were no component failures, in that the leak occurred when a soldered joint connection failed. This joint connected a two inch copper pipe to a two inch copper elbow fitting. The manufacturer of the pipe and fitting is not relevant, and the manufacturer of the solder is unknown.

B. PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Nuclear Power Plant could be iden-tified.

C. SPECIAL COMMENTS:

None NRC FORM 366A (5-92)