ML17264A606

LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr

ML17264A606
Person / Time
Site:	Ginna
Issue date:	09/19/1996
From:	Mecredy R, St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:	Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-96-009, LER-96-9, NUDOCS 9609270259
Download: ML17264A606 (13)

Text

4 CATEGORY j REGULATE Y INFORMATION DISTRIBUTIO dYSTEM (RIDS)

ACCESSION NBR:9609270259 DOC.DATE: 96/09/19 NOTARIZED: NO Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G DOCKETACIL:50-244 05000244'UTH.

NAME AUTHOR AFFILIATION ST MARTIN,J.T. Rochester Gas a Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION VISSINGiG.S.

SUBJECT:

LER 96-009-01:on 960723,leakage outside containment occurred,due to weld defect, resulting leak rate greater than program limit. Source of leakage was isolated from RWST by freeze seal, allowing exit from ITS LCO 3.0.3. W/960919 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR U 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 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 VISSINGiG. 1 1 INTERNAL: 1 1 AEOD/SPD/RRAB 1 1 ILE CENTER 1 1 NRR/DE/ECGB 1 1 NR D EELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 D NRR/DSSA/SRXB 1 1 RES/DSIR/EIB 1 1 RGN1 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCEgJ H 1 1 NOAC MURPHYiG.A 1 1 NOAC POORE,W. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

NOTE TO ALL "RIDS" RECZPIENTS:

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

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 23 ENCL 23

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ROCHESTER GAS AND ELECTRIC CORPORATION ~ B9 EASTAVENIJE, ROCHESTER, N. Y Id6d9.060I AREA CODE 716 5'-2700 ROBERT C. MECREDY Vice Ptetdttent Nvcfeor Operotdoos September 19, 1996 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

'd

Subject:

LER 96-009, Leak Outside Containment, Due to Weld Defect, Results in Leak Rate Greater Than Program Limit R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (i) (B), "Any operation or condition prohibited by the plant's Technical Specifications", the attached v~

'sujyleiiieilt::,';to Nk Licensee Event Report LER 96-009 is hereby submitted. Entry into Limiting Condition for Operation 3.0.3 of Technical Specifications was 6ise4:;:,.upo@:,Gie'i'a>84'0'ko7i XCC C~a':.:. eC CRU.OvN$5':S".'2'-:.

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~ ~ "vvv'.: 'd (wNv'v Mv'dv((xc'i'i'v'c ., xv~viNv(((((ddvw~cvMvwVAYwvÃ4xvde This event has in no way affected the public's health and safety.

Ve truly yours, Robert C. Mecre y xc: Mr. Guy S; Vissing (Mail Stop 14C7)

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 Senior Resident Inspector 9609270259 960919 PDR ADOCK 05000244 S PDR

4 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB NO. 3150%104 (4.95) 'XPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK To INDUSTRY FORWARD COMMENTS REGARDING BURDEN ESTIMATE To THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-B F33),

digits/characters for each block) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. Dc 20555-0001 AND To THE PAPERWORK REDUCTION PROJECT FAG IUTY RAME Ill oocKET rrUMSDI I2I PAOE Isl R.E. Ginna Nuclear Power Plant 05000244 1 OFR TITLE I4)

Leak Outside Containment, Due to Weld Defect, Results in Leak Rate Greater Than Program Limit EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION FACIUTY IrAME OOCKETkUMSER MONTH DAY YEAR NUMBER NUMBER DAY YEAR FACILity NAME DOCKETNVMBER 07 23 96, 96 009 01 09 19 96 OPERATING THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR E: (Check one or moro) (11)

MODE (9) 20.2201(b) 20.2203(a)(2) (v) X 50.73(a)(2)(i) 50.73(a) (2) (viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a) (2)(x)

LEVEL (10) 100 20.2203(a)(2)(i) 20.2203(a)(3) (n) 50.73(a) (2)(iii) 73.71 20.2203(a) (2) (ii) 20.2203(a) (4) 50.73(a)(2) (iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Specify in Abstract below or In NRC Form 3BBA 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Iooiuda Area Coda)

John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN THIS REPORT (13)

SYSTEM REPORTABLE REPORTABLE CAUSE COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS To NPRDS BE PSF SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). X NO DATE (15)

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

On July 23, 1996, at approximately 0857 EDST, it was determined that the leak rate (from a leak on a piping system outside Containment) was greater than the analyzed value of two gallons per hour for a primary coolant source outside Containment.

Immediate corrective action was to conservatively enter Technical Specification Limiting Condition for Operation 3.0.3 and initiate a plant shutdown, and to initiate actions to isolate the leak using a freeze seal Once the leak ~

was isolated, the plant exited Technical Specification 3.0.3 and the shutdown was stopped. The leaking pipe was cut out and replaced.

The underlying cause of the leak was a weld defect which dates back to original construction in the 1960's.

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

Corrective action to prevent recurrence is outlined in Section V.B. Additional corrective actions w((I-be EIr6 identified in this supplement. te-t)t(e-GER NRC FORM 3BB (4-B5)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4-95) ,

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION r'ACILITY NAME (1) DOCKET LER NUMBER (6) PAGE I3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 96 009 01 2OFN TEXT llfmore spaceis required, use additional copies of NRC Form 366AI (17)

I. PRE-EVENT PLANT CONDITIONS:

On July 22, 1996, the plant was in Mode 1 at approximately 100% steady state reactor power. At approximately 1130 EDST, a Radiation Protection (RP) technician, who was performing a routine weekly survey of the Auxiliary Building, observed water dripping in an area behind the Refueling Water Storage Tank (RWST). The RP technician notified the Control Room operators. Efforts to locate the source of the dripping water and to quantify the teak rate were initiated by Operations and Nuclear Engineering Services (NES). The source of water was determined to be a small leak on a test line for the Containment Spray (CS) system in the Auxiliary Building, which is outside the Containment (CNMT). The leak rate was estimated to be approximately one (1) gallon per hour (GPH).

The Control Room operators did not identify any requirement that would limit plant operations. The Shift Supervisor notified Operations management, maintenance management, and plant and NES staff of the leak. All reached a similar conclusion. The staff worked on developing and implementing an action plan to address the problem. The issues of line integrity and RWST operability were investigated by NES and Laboratory Inspection Services (LIS) personnel. LIS personnel measured the amount of pipe cross-section and weld metal remaining, and NES personnel determined that there was ample metal in the affected area to provide system integrity.

The source of the leak was identified as a small pinhole leak on a pipe-to-90 degree forged socket elbow weld on a two (2) inch nominal pipe size Schedule 10 stainless steel test line for the CS system. The leak was in a part of the CS system that could not be isolated from the RWST by valve manipulations without making both trains of the emergency core cooling system (ECCS) inoperable. The leak was monitored throughout the remainder of the day and night.

An operability assessment had been requested by the Shift Supervisor on July 22 V'iheo':th'e,'edst'soGe:-ofg tha jja'aTth3 6rlst bIaaidtirdttfjid. Qn the morning of July 23, 1996, this assessment was reviewed by the Plant Operations Review Committee iPORCi. This assessment, which was based on conservative assumptions, concluded that the RWST and ECCS were still operable with the unisolated leak. This conclusion was supported by the fact that there was still ample metal area in the cross-section that could provide integrity of the pipe during a seismic event and that the total combined stress was less than the yield strength of the material.

PORC requested a more accurate estimate of the leak rate. The leak rate was measured and estimated to be Q-.S 2.:8 GPH. This leak rate is in excess of two (2) GPH, which is specified in Surveillance Test Procedure PT-39, "Leakage Evaluation of Primary Coolant Sources Outside Containment", as the program limit for the maximum integrated leak rate outside CNMT from these sources. The leak rate exceeded the requirements of procedure PT-39, and it was initially judged to be a violation of the "Primary Coolant Sources Outside Containment Program", as specified in the Ginna Station Improved Technical Specifications (ITS) Section 5.5.2.

Operation (LCO) 3.0.3, based on the COI1Ser'Vat)vd requirements of procedure PT-39 and Administrative Procedure A-52A, "Control of Limiting Conditions for Operating Equipment", even though the plant was in compliance with all ITS LCOs.

NRC FORM 366A I4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 96 NUMBER 009 NUMBER 01 3OFI TEXT Iifmore speceis required, use edditionel copies of NRC Form 366AI (17);

II. DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES' July 22, 1996, 1130 EDST: A leak is identified on a line outside of Containment.

o July 23, 1996, 0857 EDST: Event Date and Time and Discovery Date and Time.

o July 23, 1996, 0900 EDST: Load reduction is started.

o July 23, 1996, 1440 EDST: Integrated leakage outside Containment is determined to be less than two GPH. Load reduction is stopped.

v B EVENT'n July 23, 1996, at approximately 0857 EDST, with the plant in Mode 1 at approximately 100%

steady state reactor power, the Plant Operations Review Committee (PORC) conservatively directed the Shift Supervisor to enter ITS LCO 3.0.3. The Shift Supervisor directed the Control Room operators to initiate a plant shutdown. At approximately 0900 EDST, the Control Room operators initiated a plant shutdown per Normal Operating Procedure 0-2.1, "Normal Shutdown to Hot Shutdown".

The Mechanical Support group initiated actions to provide a freeze seal to isolate the source of leakage from the RWST. A freeze seal was initiated between the leaking socket and the CS pump suction line from the RWST. At approximately 1440 EDST on July 23, 1996, the freeze seal had isolated the leak. ITS LCO 3.0.3 was exited and the load reduction was stopped. The affected pipe and socket welds were cut out and a new prefabricated spoolpiece was installed. The affected weld was retained for failure mode analysis.

A load increase to return the plant to full power was initiated at approximately 1530 EDST on July 23, 1996.

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

None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

Due to the location and size of the leak, an operability assessment determined that no emergency safeguards system function was made inoperable. However, the leak was not isolable from the RWST without valve manipulations that would have resulted in the inoperability of the RWST and both trains of the ECCS.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

~

(4-96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 96 009 01 Z

TEXT llfmore space is required, use eddi donel copies of NRC Form 366A/ (17)

METHOD OF DISCOVERY:

The leak was first discovered by an RP technician during a routine weekly survey of the Auxiliary Building. He notified the Control Room operators, and the precise location was confirmed by Operations and Engineering. Confirmation that the leak rate was above 2 GPH occurred after measurements performed at the direction of PORC.

OPERATOR ACTION:

When notified by PORC, the Shift Supervisor directed the Control Room operators to enter ITS LCO 3.0.3 and to initiate a plant shutdown. The Control Room operators initiated a plant shutdown per procedure 0-2.1.

The Shift Supervisor subsequently notified the NRC per 10 CFR 50.72 (b) (1) (i) (A), non-emergency one hour notification, at approximately 0951 EDST on July 23, 1996. When the teak was isolated, ITS LCO 3.0.3 was exited, and the plant was returned to full power.

G. SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT:

A. IMMEDIATECAUSE:

The immediate cause of the condition prohibited by Technical Specifications was the conservative determination by PORC to enter ITS LCO 3.0.3 due to the leak.

B. INTERMEDIATE CAUSE:

The intermediate cause of the leak was a pinhole leak in the weld attaching the pipe to the socket elbow.

C. ROOT CAUSE:

The affected pipe, socket, and fillet weld were analyzed for the failure mode. The pinhole was determined to be an original installation weld defect (dating from the 1960's) that existed over an area of slag inclusion in the weld. Almost all the slag had been leached from the pinhole channel over the past 25 years.

This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction / Installation".

NAC FOAM 366A (4-95)

r r

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

~ (4-95)

LICENSEE EVENT REPORT.(LER)

TEXT CONT(NUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 96 009 NUMBER NUMBER 01 5oFI TEXT llfmore spaceis required, use additional copies of NRC Form 366AJ (17)

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IV. ANALYSIS OF EVENT:

This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (i)

(B), "Any operation or condition prohibited by the plant's Technical Specifications". The integrated leak rate in excess of 2 GPH outside CNMT resulted in conservative entry into ITS LCO 3.0.3.

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NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 96 009 NUMBER NUMBER 01 80.F I TEXT llfmore spaceis required, use additional copies of NRC Form 386Al (17)

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

The source of the leakage was isolated from the RWST by freeze seal, permitting exit from ITS LCO 3.0.3.

0 The affected pipe and socket welds were cut out and replaced with a new prefabricated spoolpiece, and the freeze seal was removed.

ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

1 o Failure Analysis of the leak in the socket weld has been performed.

o ~ The adjacent original existing welds were cut out and replaced.

o Additional corrective actions vd(I(-be are identified in this supplement'te-th(e-GER Br(ed:::arevI v~hwwevw ef~w:

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NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

L1CEHSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 05000244 96 009 01 NOFN TEXT Iifmore spece is required, use edditional copies of NRC Form 366AI (17)

Vl. ADDITIONALINFORMATION:

A. FAILED COMPONENTS:

The failed component was a 2 inch nominal pipe size Schedule 10 pipe-to-90 degree forged socket elbow weld. The pipe and socket elbow are stainless steel, ASTM A312 Type 304 and A182 F304 (forged) respectively. The socket weld was an original installation weld made before 1970.

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 identified.

C. SPECIAL COMMENTS:

None NAC FORM 366A (4.96)