ML18038B688

LER 96-002-00:on 960323 & 24,main Steam Isolation Valve Leak Rates Exceeded Acceptance Criteria.Caused by Excessive Leakage on a & C Inboard MSIV Line Valves.Leaking Valves investigated.W/960506 Ltr

ML18038B688
Person / Time
Site:	Browns Ferry
Issue date:	05/06/1996
From:	Machon R, Jay Wallace TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-96-002-03, LER-96-2-3, NUDOCS 9605090176
Download: ML18038B688 (18)

Text

CATEGORY 3g

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,REGULATORY INFORMATION DISTRIBUTION SYSTEM '(RIDS,)

ACCESSION 'NBR:9605090176 DOC.DATE: 96/05/06 NOTARIZED: NO DOCKET

'FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH. NAME AUTHOR AFFILIATION WALLACE,J.E. Tennessee Valley Authority MACHON,R.D. Tennessee Valley Authority RECIP.NAME ,REC1'PIENT AFFILIATION

SUBJECT:

LER 96-002-00:on 960323 !'4,main steam isolation valve leak rates exceeding acceptance. criteria occurred. Caused by e'xcessive leakage on 'A' 'C'nboard MSIV line valves. 'A Leaking valves investigated.W/960506 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED! LTR J ENCL I SIZE:

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

NOTES:

,,RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME 'LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3-PD 1 1 WILLIAMS,J. 1 1 INTERNAL: ACRS 1 1 AEOD/SPD/RAB 2 2 1" 1 AEOD/SPD/RRAB NRR/DE/ECGB

.NRR/DE/EMEB 1

1 1

1 1

RP'-D~

NRR/DRCH/HHFB 1

1 1

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 NRR/DSSA/SRXB 1' 1 1 RGN2 FILE 01 1 1 D RES/DSIR/EIB'XTERNAL:

L ST LOBBY WARD 1 1 I ITCO BRYCE,J H 2 2 NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 .1 N,

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP'S'O REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM'WFN 5D-5(EXT. 415-2083), TO ELIMINATE YOUR NAME FROM DISTRIBUTION'lSTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED:, LTTR 26 ENCL 26

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Og a May 6, 1996 U.S. Nuclear Regulatory Cc.ziimission 10 CFR 50.73 ATTN: Document Control Desk Washington, D.C. '20555

Dear Sir:

BROWNS FERRY NUCLEAR PLANT NOS.

(BFN) 'UNITS 50-259, 260, and 296 FACILITY'PERATING li 2i AND 3 - DOCKET LICENSEE EVENT REPORT (LER) 50 260/96002 LICENSEDPR 33i 52@ .AND68 The enclosed report provides details concerning the ma'in steam i'solation valve 'leak rates exceeding acceptance criteria. TVA is submitting this event as a voluntary,LER for information

.only.

'incerely, R. D. chon Enclosure cc: See page 2 O OP g;]0 9605090i76 960506 PDR ADOCK 05000260 8 PDR

4i i U.S. Nuclear Regulatory Commission Page 2 May 6, 1996 Enclosure cc (Enclosure)'r.

Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Region II Marietta Street, NW, Suite 2900 101 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama. 35611 Mr. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockvi lie, Maryland 20852

II+I NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150%104 (4.85) EXPIRES 04/30/88 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH lHIS MANDATORY INFORMATION COLLECTION REQUEST:

50.0 HRS. REPORTED LESSONS LEARNED ARE LZCENSEE EVENT REPORT (LER) INCORPORATED INTO THE LICENSING PROCESS ANO FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING (See reverse for required number of BURDEN ESTIMATE TO THE INFORMATIONAND RECORDS digits/characters for each block) MANAGEMENT BRANCH .IT% F33). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON DC 205554xx)t.

FAaLITY NAME II) OOCKST NURSER IT) PAOE IT) 05000260 1 OF 6 Browns Ferry Nuclear Plant (BFN) Unit 2 TITLE Ia)

Main Steam Isolation Valves Leak Rate Exceeded the Local Leak Rate Test Acceptance Criteria due to Internal Component Wear EVENT DATE (5) LER NUMBER 6 REPORT DATE OTHER FACILITIES INVOLVED 8)

FACILITYNAME YEAR SEQUENllAL REVISION MONTH DAY YEAR DAY 05000 NUMBER NUMBER NA FAaLITY NAME 23 96 96 002 00 96 NA 05000 OPERATING THIS REPORT IS SUBMITTED PURS UANT TO THE REQUIREMENTS OF 10 CFR: (Check one or moro) (11)

MODE (8) N 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii) 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

POWER 000 LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 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 Absl)act bekw or in NRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12 T ELEPIICNE NIIMBER (IooIuda Acaa coda)

James E. Wallace, Compliance Engineer (205) 729-7874 COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS B SB FCV A585 SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION X NO DATE (15)

(If yes, compkte EXPECTED SUBMISSION DATE).

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

~

On March 23 and 24, 1996, during the BFN Unit 2 refueling outage with the reactor in cold shutdown, the

'A'nd main steam isolation valves (MSIV) had a leakage that exceeded the local leak rate test (LLRT) acceptance

'C'nboard criteria of 11.5 standard cubic feet per hour (SCFH). The as-found leakage was 18.7 SCFH and 32.0 SCFH respectively. The cause of the excess leakage resulted from a misalignment between the valve mating seats because of internal component wear. Corrective actions taken,to address the excessive leakage for the

'A'nd 'C'nboard MSIVs were to replace internal components, to reassemble MSIVs, and to successfully retest the valves. The results of the as-left leakage'rate tests were 6.1 SCFH and 1.5 SCFH respectively. There were previous LERs (260/94008, 260/93002,.259/85039, and 296/84007) that also resulted from abnormal internal wear of MSIVs; however, only LER 260/93002 involved one of the failed valves in this LER. That valve ('C'nboard valve) had an acceptable leakage rate during the 1994 Unit 2 refueling outage. Consequently, corrective action taken in LER 260/93002 should have precluded this Lg RT failure. This LER is being submitted as a voluntary LER for information. only.

~i i NRC FORM 666A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACZLZTZ NAME DOCKET LER NUMBER NUMBER NUMBER Browns Ferry Unit 2 05000260 2 of 6 96 -- 002 -- 00 more space rs reqolr, use a coca copres o olm PLANT CONDZTZONS At the time of this event, Unit 2 was in cold shutdown for the scheduled refueling outage. Unit 3 was,operating at 100 percent power. Unit 1 was shutdown and defueled.

ZZ. DESCRZPTZON OF EVENT A. Event On March 23, 1996, at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />, during the performance of a local leak rate test (LLRT) surveillance instruction (SI)(2-SI-4.7.A.2.i-3/1c), the Unit 2 'C'ain steam isolation valve (MSIV)

[SB] line was tested between flow control valves (FCV)[V)(2-FCV-1-37 and 2-FCV-1-38). The result of the LLRT identified an as-found leak rate of 32.0 standard cubic feet per hour (SCFH).

This leakage exceeded the LLRT acceptance criteria of 11.5 SCFH.

Additionally, on March 24, 1996, at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, during the'A'SIV performance of LLRT SI (2-SI-4.7.A.2.i-3/la), the Unit 2 was also identifi:ed as exceeding (18.7'CFH) the LLRT acceptance criteria. This MSIV line was also measured by testing between the inboard (2-FCV-1-14) and outboard (2-FCV-1-15) isolation valves.

On March 24, 1996, at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br />, after testing was completed on the four Unit 2 MSIV lines ('A' 18.7 SCFH, 'B' 8.2 SCFH, 'C' 32.0 SCFH, 'D' 11.4 SCFH), TVA made a 4-hour nonemergency notification to NRC in accordance with 10 CFR 50.72(b)(2)(i) for the excessive leakage on 'A'ND 'C'SIV lines.

On April 2, 1996, after MSIV 'A'nboard valve poppet was replaced, the 'A'SIV line was retested, and the as-left leakage was 6.15 SCFH. On April 9, 1996, after MSIV 'C'nboard valve poppet .was replaced, the 'C'SIV line was retested, and the as-left leakage was 1.51 SCFH.

On April 19, 1996, after a reportability evaluation concluded that this event was not reportable, the 10 CFR 50.72(b)(2)(i) notification was retracted. Therefore, this LER is being submitted as a voluntary LER for information only.

Zno rable Structures, Co onents, or S stems 'that Contributed to e Event:

The faulty valves (model number 20851-H-26) were manufactured by Atwood and Morrill Company.

NRC FORM 366A (4-95)

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACZLZTZ NAME DOCKET NUMBER NUMBER

'Browns Ferry Unit .2 05000260 3 of 6 96 -- 002 -- 00 EXT more spree rs requir, use a rrrona copes orm (17)

C. Dates and roximate Times of,Ha or Occurrences:

March 23, 1996, at 2300 It was determined

'C'SIV leakage was the CST.'arch excessive.

24, 1996, at 0300 CST It was determined the

'A'SIV leakage was excessive.

March 24, 1996, at 1430 CST TVA made a 4-hour nonemergency notification to NRC in accordance with 10 CFR 50.72(b)(2)(i) for both events.

April 2, 1996 The 'A'SIV line was successfully retested (6.15 SCFH).

April '9, 1996 The 'C'SIV line was successfully retested (1.51 SCFH)..

April 18, 1996 After a reportability evaluation was performed, the 10 CFR 50.72(b)(2)(i) notification was retracted.

Other S stems or Seconda Functions Affected:

None.

Hethod of Discove The leakage, was determined to be unacceptable for the 'A'nd

'C'nboard MSIV line valves by the performance of approved SIs in accordance with the .Browns Ferry Nuclear .Plant (BFN) LLRT program.

rator Actions:

'None.

NRC FORM 366A (4.95)

41 ii 4

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACZLZTX NAME'OCKET NUMBER NUMBER Browns Ferry Unit 2 05000260 4 of 6 96 -- 002 -- 00 X more space rs requir, use a rtioiia copes oim 1 G., Safet S stem Res onses:

None.

ZZZ. CAUSE OF 'THE EVENT A. Zmmediate Cause:

The immediate cause was the excessive leakage on 'A'nd MSIV line valves. 'C'nboard B. Root Cause:

The root causes were the misalignment of the mating seats on the

'A'nd 'C'nboard MSIVs. The reason for the misalignment on the valve mating, seats was a result of internal component wear.

ZV. ANALYSZS OF THE EVENT At the time of the discovery of the two leak paths, Unit 2 was in cold shutdown and in a scheduled refueling outage. Primary containment was not required to be maintained. The 'A'nd 'C'SIV line leakage did not comply with design requirements. They were tested acceptable to the Technical Specifications (TS) pressure of 25 psig during the LLRT prior to Unit 2 restart in October 1994.

Each MSIV line has two isolation valves, one inside and one outside primary containment. 'The isolation valves prevent a radiation release in excess of 10 CFR 100 guideline during a steam break outside primary containment. The valves also limit inventory losses during a loss of coolant acci:dent. BFN TS require that the MSIVs be tested during each refueling outage. If the leakage for any MSIV exceeds 11.5 SCFH. The BFN TS only require that the valve be repaired and retested until the valve leakage meets the LLRT acceptance criteria.

In this event, no valves failed open or failed to close. The maximum total leakage through the four steam lines was s 35.2 SCFH', which is significantly less than the 655.9 SCFH limit for primary containment.

Furthermore, even if the current maximum pathway leakage of primary containment prior to shutdown for this refueling outage '(approximately 180 SCFH) is added,to the current MSIV leakage, the total primary containment leakage is still less than the 655.9 SCFH limit.

'he maximum total leakage is calculated by, adding the total leakage from the four MSIV lines and dividing by 2 (i.e., [18.7 + 8.2 + 32.0 + 11.4]/2). In all likelihood, the actual total leakage would be S35.2 SCFH.

NRC FORM 366A (+66)

ii, 9

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NRC'FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME DOCKET NUMBER NUMBER Br owns Fe r ry Unit 2 05000260 5 of 6 96 -- 002 -- 00 TEXT ( more space is requir, use a iriooa copies orm (1 Therefore, TVA concluded the safety of the plant personnel and the public was not comprised.

V. CORRECTIVE ACTIONS A. Zaanediate Corrective Actions:

The leaking valves were investigated for a failure mechanism,,

scheduled for repair, and retested.

B. Corrective Actions to Prevent Recurrence:

The poppet on 'A'nboard MSIV was changed to a long-nosed poppet'to provide a guidance mechanism to improve the alignment on the mating seats. This action was proven effective in that the as-left leakage of 6.1 SCFH was below the allowable leakage of 11.5 SCFH.

Although some signs of wear and oxidation were observed on the

'C'nboard MSIV, there was no obvious sign of rib guide wear.

Since only one long-nose poppet was available, the internal mating seats of the 'C'SIV were cleaned and lapped. Based on the retesting leakage of the 'C'SIV line (1.51 SCFH), the performed corrective actions appear to be effective.

VZ. ADDZTZONAL ZNFORHATZON Failed Co onents:

The faulty valves (model number 20851-H-26) were manufactured by Atwood and Morrill Company.

B. Previous LERs on Similar Events:

The latest LER (260/94008) resulted from abnormal rib guide wear on the 'D'nboard MSIV. This valve was repaired and retest"9 and did not experience excessive leakage during this LLRT, In this LER (260/96002) the failed components were not the MSIV nor was excessive rib guide wear observed. 'D'nboard Therefore, the corrective actions taken in the LER (260/94008) would not have precluded this event.

Additi'onally, other previous LERs (260/93002, 259/85039, and 296/84007) resulted from MSIV abnormal internal wear; however, only 260/93002 involved one of the valves in this LER (260/96002). The 'C'nboard valve had an acceptable leakage during the 1994 Unit 2 refueling. outage. Consequently, 1993 corrective actions for the 'O'SIV inboard valve appear to be acceptable and should have precluded this LLRT fai'lure.

NRC FORM 366A (4-95)

IOI 4

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACZLZTX NAME DOCKET NUMBER NUMBER Browns Ferry Unit 2 05000260 6 of 6 96 -- :002 -- 00 XT ( more space rs requrr, use a nrorra copes err'1 )

VZZ. CObRCITMENTS None.

Energy Industry Identification System (EIIS) system and component codes are identified in the text with brackets (e.g., [XX]).

NRC FORM 366A (4.95)

(Qi L