ML18038B151

LER 94-008-01:on 941002,containment Penetration & MSIV Leak Rates Exceeded TS Limits.Caused by Abnormal Rib Guide Wear. MSIVs Repaired & Retested.Maint Instructions Will Be Written for Valves to Address Proper Shimming Requirements

ML18038B151
Person / Time
Site:	Browns Ferry
Issue date:	02/27/1995
From:	WALLACE J E TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18038B150	List: ML18038B149 ML18038B151
References
LER-94-008, LER-94-8, NUDOCS 9503060060
Download: ML18038B151 (18)

Text

NRC FORH,366 (5-92)U.S.IN)CLEAR REGULATORY CQHISSION APPIIOVED BY (NRI NO 3150-0104 EXPIRES 5/31/95 LICENSEE'EVENT REPORT (LER)(See reverse for required nunber of digits/characters for each block)ESTIHATED BURDEN PER RESPONSE TO COHPLY MITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORNARD COHMEHTS REGARDIHG'BURDEN ESTIMATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (MNBB 7714), U.ST NUCLEAR REGULATORY COHHISSION, UASHIHGTON, DC 20555-0001 AND TO THE PAPERNORK REDUCTION PROJECT (3140-0104), OFFICE OF HANAGEHENT AND BUDGET UASHINGTON OC 20503.FACILITY NAME (1)Browns Ferr Nuclear Plant BFN Unit 2 DOCKET IRlSER (2)05000260 PAGE (3)1 OF 9 TITLE (4)Contaireent penetration and main steam isolation valve leek rates exceeded Technical Specification limits.EVENT DATE 5 MONTH DAY YEAR YEAR LER INNER 6 SEQUENTIAL NUMBER REVISION NUHBER REP(NIT DATE 7 HONTH DAY FACILITY HAHE NA YEAR DOCKET NUHBER OTHER FACILITIES INVOLVED 8 10 02 94 94 008 01 02 27 FACILITY HAHE HA DOCKET NUHBER OPERATING NX)E (9)N S REP(NIT IS SIÃNITTED PINISUANT 20.402(b)TO THE REQUIRBKNTS OF 10 CFR=Check one or mor e 11 73.71(b)50.73(a)(2)(iv) 20.405(c)PONER LEVEL (10)000 20.405(a)(1)(i) 20.405(a)(1)(ii) 20.405(a)(1)(iii) 20.405(a)('1)(iv) 20.405(a)(1)(v) 50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(i)(B) 50.73(a)(2)(ii) 50'3(a)(2)(iii)

LICENSEE CONTACT FOR THIS LER 12 50.73(a)(2)(v) 50.73(a)(2)(vii) 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(x) 73.71(c)OTHER (Specify in Abstract below and in Text, NRC Form 366A NAME James E.Wallace, Compliance Licensing Engineer TELEPHONE NUHBER (Include Area Code)(205)729-7874 CHC>LETE ONE LINE FOR EACH CADENT FAILURE DESCRIBED IN THIS REP(NIT 13 CAUSE SYSTEH COHPONENT'HANUFACTURER REPORTABLE TO HPRDS CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO HPRDS BF FCV iF049 SB FCV A585 RPPLEHENTAL REPORT EXPECTED 14 YES (If yes, complete EXPECTED SUBHISSION DATE).X NO EXPECTED SISNI SS I ON DATE (15)HOHTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On October 2, 1994, local leak rate testing was being performed during the BFN Unit.2 refueling outage.At 1240 hours0.0144 days <br />0.344 hours <br />0.00205 weeks <br />4.7182e-4 months <br />, primary containment ventilation penetrations X-25 and 205 (2809.2329 SCFH)exceeded the Technical Specification (TS)limit of 655.9 SCFH.Subsequently, at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, the outboard'D'ain steam isolation valve (MSIV)had leakage (60.5745 SCFH)which exceeded the TS limit of 11.5 SCFH.Therefore, these events are reportable in accordance with 10 CFR 50.73 (a)(2)(ii).

The cause of X-25 and 205 penetration leakages resulted from a lack of adequate procedural guidance as to the proper shimming of seismic mounting brackets.The cause of the outboard'D'SIV leakage was abnormal rib guide wear.Corrective actions taken for the X-25 and 205 penetration leakages were to properly.shim seismic mounting brackets and to successfully retest the penetrations.

A maintenance instruction will be written for these and similar type of valves to address the proper shimming requirements.

Corrective actions taken to address the excessive leakage for outboard'D'SIV were to repair and successfully retest the MSIV.9503060060 950227 PDR ADOCK 05000260 S PDR 0 gl NRC FORH 366A'(5-92)U S.NICLEAR REGULAT(NIY CONIISSION LZCENSEE EVENT REPORT TEXT CONTZNUATZON APPIIOVED BY (BNI NO 3150-0104.EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY llITH THIS IHFORliATION COLLECTION REQUEST.50 0 HRS.FORllARD COHHENTS REGARDING BURDEN'ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (HNBB 7714), U.ST NUCLEAR REGULATORY COHHISSION, llASHINGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET, llASHINGTON, DC 20503 FACILITY NAIL (1)'rowns Ferry Unit 2 DOCKET IRNRER (2)05000260 LER IRNNKR (6).TEAR SEQUENTIAL REVI SION NUHBER NUHBER 01 008 94 PAGE (3)2 of 9 TEXT If more s ce is r ired use additional co ies of NRC Form 366A (17)Zo PLANT CONDZTZONS At the time, this event was initiated Unit 2 was shutdown for the scheduled Unit 2, Cycle 7 refueling outage.Units 1 and 3 were shutdown and'efueled.

ZZ~DESCRZPTZON OF EVENT A..Event: On October 2, 1994, during performance of local leak rate testing (LLRT), two events were identified involving primary containment leakage in excess of allowed limits.The excessive leakages were measured by testing between the inboard and outboard isolation valves.Further details of these events are provided below: Dr ell Penetrations X-25 and 205'At 1240 hours0.0144 days <br />0.344 hours <br />0.00205 weeks <br />4.7182e-4 months <br /> during the performance of a Surveillance Instruction (SZ)(2-SZ-4.7.A.2.g-3/64a), attempts to achieve the required pressurization (51.0 psid)and stabilization were unsuccessful due to gross leakage.The components affected were flow control valves[ISV](2-FCV-64-17, 18, 19 and 2-FCV-76-24).

A leak rate calculation for the penetration

[PEN]was performed and documented to be 2809.2329 SCFH.However, the allowable total leak rate for primary containment penetrations

[BF]was 655.9;SCFH.Additional LLRT performed during the Unit 2, Cycle 7 refueling outage determined that valves 2-FCV-64-18 and 19 were the major contributors for the 2809.2329 SCFH'eakage.

Outboard'D'ain Steam Isolation Valve At 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> on October 2, 1994, during the performance of an SI (2-SI-4.7.A.2.1-3/ld) the leak rate between the'D'ain steam[SB]isolation valves (MSZVs)[ZSV]was calculated to be 60.5745 SCFH.This calculated value exceeded the Technical Specification (TS)limit of 11.5 SCFH.The outboard valve was initially investigated and preliminary troubleshooting procedures commenced.

After manipulating only the outboard valve, a significantly lower leak rate was observed.Therefore, it was concluded that the excessive leakage was through the outboard valve since the position of the inboard valve was not altered.

0 0 V f NRC FORJI 366A (5-92)U.S.JRJCLEAR REGULATORY (XWIISSI(HI LICENSEE EVENT REPORT TEXT CONTINUATION'PPROVED BY HR HO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPOHSE TO COMPLY IJITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 MRS.FORllARD COMMENTS REGARDING-BURDEH ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, lJASHIHGTOH, DC 20555-0001, AND TO THE PAPERMORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AMD BUDGET, lJASMINGTOM DC 20503 FACILITY WVK (1)DOCKET HWER (2)YEAR LER IRMRJER (6)SEQUENTIAL NUMBER REVISION NUMBER PAGE (3)Browns Ferry Unit 2 05000260 94 008 01'of 9 TEXT If more s ce is r ired use additional co ies of MRC Farm 366A (17)These events are reportable in accordance with 10 CFR 50.73 (a)(2)(ii) as a condition in a nuclear plant, including its principal safety barriers, being seriously degraded.BE Ino erable Structures Com nents or S stems that Contributed to the Event: None.C~Dates and A roximate Times of Ma or Occurrencest October 2, 1994 at'0430 CST at 1240 CST at 1408 CST at 1612 CST at 2000'ST Penetrations (2-X-25 and 205)SI began after the Unit 2 shutdown for a scheduled refueling outage.The leakage for these penetrations was determined to have exceeded the acceptance criteria.TVA made a four-hour notification to the NRC pursuant to 10 CFR 50.72(b)(2)(i).

'D'ain steam, isolation valves'SI began.It was determined that the leak rate was excessive.

at 2115 CST October 25, 1994 November 2, 1994 November 21, 1994 at 0429 November 28, 1994 TVA made a four-hour notification to the NRC pursuant to 10 CFR 50.72(b)(2)(i).

SI for Penetrations X-25 and 205 was completed and was successfully retested.SI for outboard'D'SIV was successfully retested.Reactor mode switch was positioned to startup During an investigation for the recycling of the drywell differential pressure air compressor, valves 2-64-FCV-18 and 19 were identified to be

NRC FOHN 366A (5-92)U.S IRICLEAR REGULAT(HIY CQSIISSION LICENSEE EVENT REPORT TEXT CONTINUATION APPROVED BY (HEI NO.3150-0104 EXPIRES 5/31/95'STIHATED BURDEN PER RESPONSE TO COHPLY MITH THIS INFORHATIOH COLLECTION REQUEST: 50.0 HRS.FORMARD COHHEHTS REGARDING BURDEN ESTIHATE TO THE IHFORHATIOH AND RECORDS HANAGEKENT BRANCH (NHBB 7714), U.S.NUCLEAR REGULATOR'Y COHHISSIOH, MASHINGTON, DC 20555-0001, AND TO THE PAPERMORK REDUCTION PROJECT (3150-0104), OFFICE OF NANAGENENT AND BUDGET, MASHINGTOH DC 20503 FACILITY NA%(1)Browns Ferry Unit 2 DOCKET WSER (2)05000260 YEAR 94 LER'RlSER (6)SEQUENTIAL NUHBER 008 REVI S I OH NUHBER 01 PAGE (3)4 of 9 TEXT If more s ce is r ired use edditioneI co ies of NRC Form 366A (17)leaking after being exercised to support Unit 2, Cycle 7 startup.November 30, 1994 The valves were repaired, and an SI for penetrations X-25 and 205 was satisfactorily performed.

D~Other'stems or Seconda Functions Affected: None.Method of Di.scove I The leakages were determined to be unacceptable for both conditions by the performance of each SI in accordance with the BFN LLRT program.F.0 erator ActionsI None.G.Safet S stem Res nsesI None.CAUSE OF THE'VENT A.Immediate Cause: The immediate causes of these events weres Dr ell Penetrations X-25 and 205 Gross leakage from the components for primary containment ventilation penetrat."lone 2-X-25 and 205 was due to improper shimming of the seismic mounting brackets.Outboard'D'ain Steam Isolation Valve Excessive leakage from the outboard'D'SIV was due to abnormal ri.b guide wear.

0~I NRC FORM 366A (5-92)U.S I)CLEAR REGULATORY CNHIISSION APPROVED BY (HEI NO.3150-0104 EXPIRES 5/31/95 FACILITY Nba (1)DOCKET NEER (2)LICENSEE EVENT REPORT'TEXT CONTINUATION LER IRBRIER (6)PAGE (3)REVI SION NUMBER SEQUENTIAL NUHBER YEAR ESTIHATED BURDEN PER RESPOHSE TO COMPLY IJITH THIS IHFORHATIOH COLLECTION REQUEST: 50.0 HRS.FORllARD COHMENTS REGARDING BURDEN ESTIHATE TO THE IHFORHATIOH AND RECORDS HANAGEMEHT BRANCH'HNBB 7714), U.S., NUCLEAR REGULATORY COHHISSIOH, llASHIHGTOH, DC 20555-0001, AHD TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF HAHAGEMEHT AND BUDGET, llASHINGTON DC 20503 Browns Ferry Unit 2 05000260 94 008 01 5 of 9 TEXT tf more s ce is r ired use additional co ies of NRC Form 366A (1/)Bi Root Causes The root cause of these events were: Dr ell Penetrations X-25 and 205 X-25 and 205 penetration leakages resulted'rimarily from a lack of adequate, procedural guidance as to the proper shimming of seismic mounting brackets.Outboard'D'ain Steam Isolation Valve The reason for abnormal rib guide wear is unknown.ZV.ANALYSIS OF THE EVENT ,At the time of discovery of the two leak paths, Unit 2 was.shutdown and was in a scheduled refueling outage.Primary containment was not required to be maintained.

While the X-25 and 205 penetrations and the outboard'D'SZV leakages did not comply with design requirements, they were tested to accident pressure during the LLRT prior to Unit 2 restart on May 25, 1993.'Dr ell Penetrations X-25 and 205 The components that were responsible for the excessive leakage for these penetrations have-been.determined.

An investigation of the X-25 and 205 penetrations determined that these penetrations did not pose a significant safety impact during Unit 2, Cycle 7 because the upstream valve (2-FCV-64-17) did not leak.The valves involved in the identified leakage were the 2-FCV-64-18 and-19 (see figure 1).These valves, as configured, result in a primary containment to primary containment leakage (drywell to torus).Therefore, based on the plant conditions and the plant configuration of the involved valves, plant safety was not adversely affected.Additionally, the safety of plant personnel and the public was hot compromised.

Outboard'D'ain steam isolation valve Each main steam line has two isolation valves, one'inside and one outside primary containment..

The isolation prevents radiation release in excess.of 10 CFR 100 guidelines during a steam-line break outside primary containment.

The valves also limit inventory losses during a loss of coolant accident.TS require the MSIVs be tested during each refueling outage.If the leakage rate for any one MSZV exceeds 11.5 SCFH, TS require that the valve be repaired and retested.

4k~l NRC FORM 366A (5-92)U.S.WCLEAR REGULATORY CQIIISSI(HI LICENSEE EVENT REPORT TEXT CONTINUATION APPROVED BY (HEI NO 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS IHFORHATIOH COLLECTION REQUEST: 50.0 MRS'ORIIARD COMMENTS REGARDIHG BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714)~U.S.NUCLEAR REGULATORY COWISSIONg NASHIHGTOM, DC 20555 0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AHD BUDGET, NASHINGTON DC 20503 FACILITY WUK (1)Browns Ferry Unit 2 DOCKET IMBIBER (2)05000260 YEAR 94 LER IRHRIER (6)SEQUENTIAL NUMBER 008 REVI SIOH NUMBER 01 PAGE (3)6 of 9 TEXT tf more s ce is r ired use sdditionei co ies of NRC Form 366A (17)Furthermore, if the"as found" maximum path leak rate is greater than allowed, TS require that repairs are completed and that local leakage meets acceptance criteria as proved by testing.Preliminary investigation after the initiating event identified that the maximum leak path was from the outboard'D'SZV.During this investigation, it also was determined that leakage from the'D'nboard MSZV was within the TS limit.Consequently, it was concluded that plant safety was not adversely affected, and that the safety of plant personnel and the public was not compromised.

V, CORRECTIVE ACTIONS Immediate Corrective Actionst'The leaking valves were investigated and were scheduled to be repaired and retested.Dr ell Penetrations X-25 and 205 For the two valves (2-FCV-64-18 and 19), corrosion product interferences were removed from the valve seat and the associated system piping.New butterfly valve seats were installed in the valves.Additionally, it was determined that the stem adaptor.key for 2-FCV-64-19 was longer than the manufacturer's recommendation and was subsequently machined to the manufacturer's recommendation.

Both valves were reassembled and a successful LLRT test was performed prior to the Unit 2 startup.During the Unit 2, Cycle 7 refueling outage startup,.excessive run times (cycling)were experienced on the drywell differential pressure air compressor.

An investigation followed, and valves 2-FCV-64-18 and 19 were again identified as leaking.The valves were disassembled, and it was determined that the shims between the valve flanges and the actuator mounting brackets were missing.These shims are required to ensure a level actuator and to eliminate a misalignment between the valve and actuator.This type of misalignment could intermittently bind the valve and prevent a seal.The two valves were shimmed, exercised several times, and retested.Both valves passed their LLRT testing with a total as-left maximum path leakage of 4.7960 SCFH..

NRC FORH 366A (5-92)U.S.IN)CLEAR REGLRATORY CQBIISSIQI LICENSEE EVENT REPORT TEXT CONTINUATION APPROVED BY QRI NO.3150-0104 EXPIRES 5/31/95'ESTIHATED.BURDEN PER RESPONSE TO COHPLY.IIITH THIS IHFORHATION COLLECTIOH REQUEST: 50.0 HRS.FORIIARD COHHENTS REGARDING BURDEN EST IHATE TO THE'INFORHATI ON AND RECORDS HANAGEHEHT BRANCH (HNBB 7714), U.S.HUCLEAR REGULATORY CQOII SSIOH,'WASHINGTON, DC 20555 0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET, llASHINGTOH, DC 20503 FACILITY Nh%(1)Browns Ferry Unit 2 DOCKET IRHSER (2)05000260'YEAR 94 LER INNER (6)SEQUENTIAL HUHBER 008 REV IS I OH NUHBER 01 PAGE (3)7 of 9 TEXT If more s ce is r ired use edditionei co les of NRC Form 366A (17)The missing shims are believed to be the primary cause of the excessive leakages that was observed before and after Unit 2.startup.Outboard'D'ain steam isolation valve The, outboard'D'SZV was initially investigated and preliminary troubleshooting procedures commenced.

After manipulating only the'D'utboard MSIV, the boundary was retested, and.a significantly lower leak rate was observed.Therefore, it was concluded that the excessive leakage in the"as found" condition was through the outboard valve since the position of the inboard valve was not altered.Bo Corrective Actions to Prevent Recurrence:

Dr ell Penetrations X-25 and 205 A maintenance instruction will be written for these and similar type of valves to address the proper shimming requirements.

Outboard'D'ain steam isolation valve During the initial disassembling of the outboard MSIV, the sealing surface of the valve was cleaned and the MSIV was reassembled and tested.However, the valve did not pass the LLRT test.Consequently, the valve rib guide was grounded and the edge was tapered to blend with the cast steel part of the poppet ring.The valve was reassembled and passed its LLRT testing with a total as-left maximum path leakage of 1.6253 SCFH.'VZ~ADDITIONAL INFORMATION Ao Failed Componentsl The 2-FCV-64-18 is a 18-inch flow control valve manufactured by Flowseal, Model 18-1WA-121LGB-BXG.

The 2-FCV-64-19 is a 20-inch flow, control valve manufactured by Flowseal, Model 20-1WA-121LGB-BXG.

The outboard'D'SIV is manufactured by Atwood and Morrill, Model 20851-H-26.

0 lg NRC F(NDI 366A (5-92)U.S.IN)CLEAR REGULATORY DSNISSIOH LICENSEE EVENT REPORT TEXT CONTZNUATZON APPROVED BY (%HI NO 3150-0104 EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY IJITH THIS IHFORHATIOH COLLECTIOH REQUEST: 50.0 HRS.FORHARD COHHEHTS REGARDING BURDEN ESTIHATE TO THE IHFORHATIOH AHD RECORDS NANAGENEHT BRANCH (HHBB 7714), U.S.NUCLEAR REGULATORY COHHISSIOH, NASNINGTOH, DC 20555-0001, AND TO THE PAPERIJORK REDUCTION PROJECT (3150-0104)

~OFFICE OF HANAGEHEHT AND BUDGET, NASHINGTOH DC 20503 FACILITY NA%(1)Browns Ferry Unit 2 DOCKET IRHRIER (2)05000260 YEAR 94 LER IRlSER (6)SEQUENTIAL NUMBER 008 REVISION NUHBER 01 PAGE (3)8 of 9 TEXT tf more s ce is r ired use add tionst co les of HRC Form 366A (17)B.Pravious LERs.on Similar Eventst Previous Licensee Event.Reports were reviewed for exceeding local leak rate limits.LER 50-260/93002 described the event in which the'C'ain steam line inboard isolation valve exceeded its leak rate limit.However, the repair on the'C'nboard valve would not have precluded the event in this LER (260/94008).

The conclusion in this LER (260/94008) was that the inboard valve minimal'ly contributed to the noted'D'SIV leakage.Additionally, LER 296/84011 addressed the failure of a leak rate test for the residual heat removal testable check valves.However, the repairs on these valves would also not have precluded the event in this LER (260/94008).

Finally, LERs 259/85039 and 296/84007 were identified.

These LERs occurred before the implementation of the TVA MSZV upgrade program.Prior to implementing the TVA MSIV upgrade pxogram, MSZV leakages would.have revealed that the inboard and outboard valves both had excessive leakages.TVA believes that adequate previous corrective actions have been taken to reduce the number of failures.VZZ Commitment A maintenance instruction will be written for these and similar type of valves to, address the proper shimming requirements.

This instruction will be written by June 30, 1995.Energy Industry Identification System (EIZS)system and component codes, are identified in the text with brackets (e.g.,[XX)).

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