ML18038A997

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LER 94-008-00:on 941002,primary Containment Penetrations & Main Steam Isolation Valves Leak Rates Exceeding TS Limits Occurred.Cause of Leakage Initially Investigated.Valves Will Be repaired.W/941031 Ltr
ML18038A997
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 10/31/1994
From: Machon R, Jay Wallace
TENNESSEE VALLEY AUTHORITY
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-008-01, LER-94-8-1, NUDOCS 9411090286
Download: ML18038A997 (18)
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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSlON NBR:9411090286 DOC.DATE: 94/10/31 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 P RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 94-008-00:on 941002,primary containment penetrations &

main steam isolation valves leak rates exceeding TS limits occurred. Cause of leakage initial'ly investigated. Valves will

~ 'be repaired.W/941031 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR / ENCL / 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 PD2-4-PD 1 1 WILLIAMS,J. 1 1 INTERNAL: ACRS 1 1 AEOD/ROA'B/DSg 2 2 AEOD/SPD/RRAB 1 1 ~F ILE CENTER~% 2 1 1 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DORS/OEAB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRSS/PRPB 2 2 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NRR/PMAS/IRCB-E 1 1 RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 D

EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H .2 2 NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 0 NRC PDR 1 1 NUDOCS FULL TXT .1 1 U

YOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO RFDUCE 4VKSTE! CONTACT'I'IIE IMCL'ifEiTCONTROL DESk. ROOXI PI -37 I EXT. 504. 083 ) TO F Lf if ICE 'jOI:R iAifL'ROif DISTRI8U'I IOi LIS'I'S FOR DOCI'h,fl'.i'I'S 5'Of.'Oi "Il.'I:.I)!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 27

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Tennessee Vatiey Authority.: Post Otfce Box 2000. Decatur. Alabama 35609.2000 R. D. (Rick) Machon Vce President, Brovrns Ferry t4uotear Ftant October 31, 1994 U.S., Nuclear Regulatory Commission 10 CFR'50.73 ATTN: Document Control Desk Washington, D.C. 20555

Dear Sir.:

BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1i 2i AND 3 - DOCKET NOS 50-259i 50-260i AND 296 FACILITY OPERATING'ICENSE DPR 33i 52'ND 68 ,LICENSEE EVENT REPORT 50 260/94008 The enclosed .report provides details concerning primary containment penetrations and main steam isolation valves leak rates exceeding their Technical Specification limits., A supplemental report will be submitted after troubleshooting, rework, and retesting of the affected components for the two cited events have been performed.

This report is submitted in accordance with 10 CFR 50.73 (a)(2)(ii.). Section VIZ of the report describes the commitment provided in this report.

Sincerely, glzg<~

R. D. hon Site Vice President Enclosure cc: See page 2 Q Q(3~M~ /'p, 941'109028b, 'st41031 050002b0 PDR, ADOCK PDR

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U.S. Nuclear Regulatory Commission Page 2 October 31, 1994 cc (Enclosure):

INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Paul Krippner American Nuclear Insurers Town Center, Suite 300S 29 South Main Street West Hartford, Connecticut 06107 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35611 Regional Administrator U.S. Nuclear Regulatory Commission Region 101 II Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

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NRC FORH 366 U.S. NUCLEAR REGULATORY COMMISSION PROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY 'WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.

IiZCENSEE EVENT REPORT (LER) FORWARD COHHENTS REGARDING BURDEN ESTIHATE TO THE INFORMATION AND RECORDS HANAGEHENT BRANCH (HNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, (See reverse for required number of digits/characters for each block) WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NiNBER (2) PAGE (3)

Browns Ferr Nuclear Plant BFN Unit 2 05000260 1 OF 6 TITLE (4) Containment penetration and main steam isolation valve leak rates exceeded Technical Specification limits.

EVENT DATE 5 LER NUMBER 6 REPORT DATE 7 OTHER FACILITIES INVOLVED B SEQUENTIAL REVISION FACILITY NAME NA DOCKET NUHBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUHBER NUHBER FACILITY NAHE NA DOCKET NUMBER 10 02 94 94 008 00 I OPERATING THIS REPORT IS SUBHITTED PURSUANT TO THE REQUIREHEHTS OF 10 CFR: Check one or more 'l1 MODE (9) N 20.402(b) 20 '05(c) 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a)(1)(i) 50.36(c)(1) 50.73(s)(2)(v) 73.71(c) 000 50.73(s)(2)(vii)

LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i)(B) 50.73(a)(2)(viii)(A) (Specify in 20.405(s)(1)(iv) )('0.73(a)(2)(ii) 50.73(s)(2)(viii)(B) Abstract and in Text, below 20.405(a)(1)(v) 50 '3(s)(2)(iii) 50.73(a)(2)(x) NRC Form 366A LICENSEE CONTACT FOR THIS LER 12 NAME TELEPHONE NUMBER (Include Area Code)

James E. Wallace, Compliance Licensing Engineer. (205)729-7874 COMPLETE ONE LINE FOR EACH C(NPONENT FAILURE DESCRIBED IN THIS REPORT 13 REPORTABLE REPORTABLE CAUSE SYSTEH COHPONENT MANUFACTURER CAUSE SYSTEH COMPONENT MANUFACTURER TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED HONTH DAY YEAR SUBlilSS I OH X YES NO 'DATE (15) 12 15 94 (If. yes, complete. EXPECTED SUBHISSION DATE) ~

ABSTRACT (Limit to 1400 spaces, i.e., approximstety 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 /> the leak rate at the primary containment ventilation penetrations X-25 and 205 (2809.2329 SCFH) exceeded the primary containment penetration 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 />, it was determined that the 'D'ain steam isolation valves had leakage (60.5745 SCFH) which exceeded the TS limit of 11.5 SCFH. The excessive leakages were measured by testing between the inboard and outboard isolation valves. Therefore, these events are reportable in accordance with 10 CFR 50.73 (a)(2)(ii).

The cause of the excessive leakages will not be positively determined until repairs and retests have been completed. Troubleshooting for the high leak rate for the X-25 and 205 penetrations has been scheduled. The cause of the leakage for the 'D'ain steam isolation penetration was- initially investigated'. After manipulating the 'D'ain steam outboard isolation valve, 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 never altered.

NRC FORM 366A U S. NUCLEAR REGULATORY C(%MISSION APPROVED BY (NGI NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS. FORIIARD COMMENTS REGARDING BURDEN EST IHATE TO THE INFORMATION AND RECORDS MANAGEHENT LICENSEE EVENT REPORT BRANCH (HNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION/

TEXT CONTINUATION llASHINGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150 0104), OFFICE OF HANAGEHENT AND BUDGET, HASHINGTON, DC 20503 FACILITY NAME (1) DOCKET NWBER (2) LER NINBER (6) PAGE (3)

YEAR SEQUEN'IIAL REVISION NUMBER NUHBER Browns Ferry Unit 2 05000260 94 008 00 2 of 6 TEXT If more s ce is r uired use additional co ies of NRC Form 366A (17)

PLANT CONDITIONS Unit 2 was shutdown for a scheduled refueling outage. Units 1 and 3 were shutdown and defueled.

DESCRIPTION OF EVENT Ao Event:

On October 2, 1994, during performance of local leak rate testing, 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:

Penetration X-25 and 205 Primar Containment Ventilation 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 (SI) (2-SI-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.

'D'ain Steam Isolation outboard Valve Subsequently, at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> during the performance of an SI (2-SI-4.7.A.2.i-3/1d) the leak rate between the 'D'ain steam [SB]

isolation valves (MSIVs) [ISV] was calculated to be 60.5745 SCFH. This calculated value exceeded the Technical Specifications limit of 11.5 SCFH. The outboard valve was initially investigated and preliminary commenced. After manipulating only the outboard troubleshooting'rocedures 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.

These events are reportable in accordance with 10 CFR 50.73 (a)(2)(ii) as a condition in a nuclear plant, including its principle safety barriers, being seriously degraded.

4l NRC FORH 366A U.S. NUCLEAR REGULATORY CQSIISSION APPROVED BY (NB NO. 3150-0104 (5-92) EXPIRES 5/31/95

'ESTIMATED BURDEN PER RESPONSE TO COHPLY IJITH THIS INFORHATION COLLECTION REQUEST: 50. 0 HRS. FORIIARD COHHENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT LICENSEE EVENT REPORT BRANCH (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSION, TEXT CONTINUATION NASHINGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT . (3150-0104), OFFICE OF, HANAGEMENT AND BUDGET, WASHINGTON DC 20503 FACILITY NAHE (1) DOCKET HIRIBER (2) LER HUNGER (6) PAGE (3)

YEAR ,

SEQUENTIAL REVISION NUHBER NUHBER Browns Ferry Unit 2 05000260 94 008 OO 3 of 6 TEXT If more s ace is re uired use additional co ies of NRC Form 366A <<17)

B. Ino erable Structures Com onents o'r S stems that Contributed to the Events None.

C~ Dates and A roximate Times of Ma'or Occurrences:

October 2, 1994 at 0430 CST Penetrations (2-X-25 and 205) SI began after the Unit 2 shutdown for a scheduled refueling outage.

at 1240 CST The leakage for these penetrations is determined to have exceeded the acceptance criteria.

at 1408 CST TVA made a four-hour notification to the NRC pursuant to 10 CFR 50.72(b)(2)(i).

at 1612 CST 'D'ain steam isolation valves SI began.

at 2000 CST It was determined that the leak rate excessive.

was at 2115 CST TVA made a four-hour notification to the NRC pursuant to 10 CFR 50.72(b)'(2)(i).

D. Other S stems or Seconda Functions Affected:

None.

E~ Method of Discove The leakage was determined to be unacceptable for both conditions by the performance of each SI in accordance with the BFN local leak rate testing program.

F. 0 erator Actions:

None.

t 0 i+i 4

NRC FORM 366A U.S. NUCLEAR REGULATORY C(WHISSION APPROVED BY Q(B NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY MITH THIS INFORHATION COLLECTION REQUEST. 50."0 HRS. FORllARD COHHENTS REGARDING BURDEN EST IHATE TO .THE INFORHATION AND RECORDS HANAGEHENT LICENSEE EVENT REPORT BRANCH (MNBB 77'14), U.S. NUCLEAR REGULATORY COMHISSION, TEXT CONTINUATION llASHINGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFF ICE OF HANAGEHENT AND BUDGET, NASHINGTON, DC 20503 FACILITY NAME (1) DOCKET NWER (2) LER N NSER PAGE (3)

(6)'EAR SEQUENTIAL REVISION NUHBER NUHBER Browns Ferry Unit 2 05000260 94 008 00 4 of 6 TEXT If more s ce is re ired use additional co ies of NRC Form 366A (17)

G. Safet S stem Res onses:

None.

III'AUSE OF THE EVENT Immediate Causes The immediate cause of these events was: (1) gross leakage from the components for primary containment ventilation penetrations 2-X-25 and 205, and (2) excessive leakage from the 'D'utboard MSIV.

B, Root Cause:

The cause of the excessive leakages will not be determined until repairs have been completed. Each high leak rate will be quantified when TVA begins additional investigations by blanking off valves and determining inboard and outboard leakages.

Troubleshooting for the high leak rate for the X-25 and 205 penetrations has been scheduled. The 'D'utboard MSIV 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.

IVo 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 'D'SIV leakages did not comply with design requirements, they were tested to accident pressure during the local leak rate testing prior to Unit 2 restart on May 25, 1993.

Penetrations X-25 and 205 The components that were responsible for the excessive leakage have not yet been determined'. Further analysis of this event for these penetrations will be provided in the supplemental report.

i' NRC 'FORM 366A U.S. NUCLEAR REGULATORY COltlISSI ON APPROVED BY (HtB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPOHSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS ~ FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AHD RECORDS MANAGEMENT LICENSEE EVENT REPORT BRANCH (MHBB 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 HISSER (2) LER NQlBER (6) PAGE (3)

TEAR SEQUENTIAL REVISION NUMBER NUMBER Browns Ferry Unit 2 05000260 94 008 00 5 of 6 TEXT tf more s ace is r uired use additional co res of HRC Form 366A (17)

'D'ain steam outboard isolation valve Each main steam line has two isolation valves, one on the inside and one on the outside of primary containment. The isolation prevents radiation release in excess of 10 CFR 100 guidelines in the event of a steam line break outside primary containment. The valves also limit inventory losses during a loss of coolant accident. Technical Specifications require the MSIVs be tested during each refueling outage. If the leakage rate for any one MSIV exceeds 11.5 SCFH, TS require that the valve be repaired and retested. Furthermore, if the "as found" maximum path leak rate is greater than allowed, TS require that repairs are implemented and that local leakage meets acceptance criteria as demonstrated by retest. Preliminary investigation after the initiating event concluded that the maximum leak path was from the

'D'utboard MSIV. During this investigation, it also was concluded that leakage from the 'D'nboard MSIV was within the TS limit.

Consequently, it was concluded'hat plant safety was not adversely affected, and that the safety of plant personnel and the public was not compromised.

V. CORRECTIVE ACTIONS Immediate Corrective Actions:

Corrective actions are underway and have not been completed at this time. However, the 'D'utboard MSIV 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.

B. Corrective Actions to Prevent Recurrence:

Further investigations are scheduled to be performed.

Corrective actions will be determined. Additionally, the valves which resulted in the excessive leakages will be repaired and retested. The results of the investigation and the retests will be provided in the supplemental report.

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NRC, FORM 3664, U.S. NUCLEAR REGULATORY CQOIISSION APPROVED BY (NB NO. 3150-0104 (5-92) EXPIRES 5/3'I/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECT IOH REOUEST: 50.0 HRS ~ FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT LICENSEE EVENT REPORT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION WASHINGTON, DC 20555-000'I, 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) PACE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Browns Ferry Unit 2 05000260 94 008 00 6 of 6 TEXT If more s ace is re uired Use additional co ies of NRC Form 366A (17)

VIE ADDITIONAL INFORMATION Failed Components:

The supplement report necessary.

will provide further details, if B~ Previous 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). 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 prior to the implementation of the TVA MSIV upgrade program.

Prior to implementing the TVA MSZV upgrade program, MSZV leakages would have revealed that the inboard and outboard valves both had excessive leakages. The conclusion in this LER (260/94008) was that the inboard valve minimally contributed to the noted 'D'SIV leakage. TVA believes that adequate previous corrective actions have been taken to reduce the number of failures.

VIZ ~ Commitments A supplemented report is expected to be'ubmitted by December 15, 1994.

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

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