ML18039A603

LER 98-004-00:on 981007,primary Containment Allowable Leak Rate Was Exceeded.Caused by Failure of Check Valve to Fully Seat Following Shutdown of Rbbccw Sys for Llrt.Valve Was Retested During Next Refueling Outage.With 981104 Ltr

ML18039A603
Person / Time
Site:	Browns Ferry
Issue date:	11/04/1998
From:	Austin S, Sunger K TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-98-004-01, LER-98-4-1, NUDOCS 9811120108
Download: ML18039A603 (16)

Text

~ CATEGORY 3i REGULATORY INFORMATION DISTRIBUTION SYSTEM (RZDS)

ACCESSION NBR:9811120108 DOC.DATE:

FACZL:50-296 Browns Ferry Nuclear Power Station, 98/11/04 NOTARIZED: NO Unit 3, Tennessee DOCKET 05000296 I

AUTH. NAME ' AUTHOR AFFILIATION AUSTIN,S.W. Tennessee Valley Authority SUNGER,K.W. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 98-004-00:on '981007,primary containment allowable leak rate was exceeded. Caused by failure of check valve to fully seat following shutdown of RBBCCW sys for LLRT.Valve was retested during nest refueling outage. With 981104 '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'D COPIES ID CODE/NAME LTTR ENCL CODE/NAME LTTR ENCL PD2-3-PD 1 1 DEAGAZIO,A 1 1 INTERNAL: ACRS 1 1 AEOD/SP 2 2 AEOD/SPD/RRAB 1 1 1 1 NRR/DE/ECGB'RR/DE/EMEB 1 1 NRR/DE/EELB 1 1 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOHB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 RES/DET/EIB 1 1 RGN2 FILE 01 1 1 D

EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 1 1 NOAC POORE,W. 1 1 NOAC QUEENER,DS 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 E

N NOTE TO ALE. "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRZBUTZON LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 23 ENCL 23

Tennessee Valley Authority, Post Office Box 2000, Decatur, iftfabama 35609-2000 Kali W. Singer Vice President. Browns Feny Nuclear Rant November 4, 1998 U.S. Nuclear Regulatory Commission 10 CFR 50."73 ATTN: Document Control Desk Washington, D. C. 20555

Dear Sir:

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 3 DOCKET 50-296 FACILITY OPERATING LICENSE DPR 68 LICENSEE EVENT REPORT (LER) 50-296/1998004 The enclosed report provides details concerning the Unit 3 primary containment total leak rate exceeding allowable limits. During local leak rate testing activities, TVA determined that leakage through the Reactor Building Water system containment check valve was Closed'ooling approximately 2491 standard cubic feet per hour (SCFH). This exce'eded the maximum allowable containment leak rate of 655.9 SCFH.

TVA is reporting this event pursuant to 10 CFR 50.73(a) (2) (ii) (B), as a condition outside the design basis of

/f the plant. There are no commitments made in this report.

Sin erely, Karl W. Sin r cc: See page 2

'tf811120108 tf81i04 PDR ADQCK 05000296 S PDR

U.S. Nuclear Regulatory Commission Page 2'ovember 4, 1998 Enclosure cc (Enclosure):

Mr. A. W. De Agazio., Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. H. 0. Christensen, Branch Chief U.S. Nuclear Regulatory Commission

'egion II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 NRC Resident Inspector Browns -Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. L. Raghavan, Project Manager U.S.. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB No. 3150-0104 EKPmm (6. 1998) oeIsonoo I Estimated burden per response to comply with this mandatory kdonnation codection request; 50 hm. Reported lessons learned are ncorporated into LICENSEE EVENT REPORT (LER) the licensing process and fed back to Industry. Forward comments regarding burden estimate to the Records Management Branch (T F33).

(See reverse for required number of U.S. Nudear Regulatory Commission. Washington, OC 20M54001, and to the Papenvork Reduction Project (31500104), Ofdce of Management and digits/characters for each block) Budget. Washington. OC 20503. If an hformation collection does not display a currently valid OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to, the infonnabon coIIectkxL FACIUTY NAME (I I DOCKET NVMSER I21 PAOE Is)

Browns Ferry Nuclear Plant - Unit 3 05000296 1 of 5 TITLE (41 Primary'ontainment Allowable Leak Rate Exceeded EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (BI MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITYNAME DOCKET NVMBER NUMBER NVMBER NA 05000 DOCKET NUMBER 10 07 1998 1998 - 004 - 00 04 1998 NA 05000 oPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (chock one or more) (11)

MODE (9) 20. 2201 (b) 20.2203(a) (2) (v) 50 73(a)(2)(6 50.73(a) (2) (vlil

)

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

LEVEL (10) 000 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)((v) oTHER 20.2203(a)(2) (iii) 50.36(c)(1) 50.73 (a) (2) (v) 6 P ecif Y In Abstract below or In NRC Form 366A

20. 2203(a) (2) (Iv) 50.36(c) (2) 50.73(a)(2) (vli)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Ares Code)

Steven W Austin, Senior Project Manager (256) 729-2070 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

,I.",e<'."~v:>:;:

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO SYSTEM COMPONENT MANUFACTURER REPORTABLE NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH OAY YEAR YES X No SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). DATE (15)

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

On October 7, 1998, at approximately 1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br />, Central Daylight Time (CDT), during performance of the surveillance instruction that quantifies the leakage through a Reactor Building Closed Cooling Water (RBCCW) system dlywell inlet header primary containment supply check valve, it was determined that the leak rate of the valve was approximately 2491 standard cubic feet per hour (SCFH). Hence, the leak rate of the valve exceeded, the maximum allowable containment leak rate of 655.9 SCFH. Following the initial test, the RBCCW system was returned to service to provide cooling for personnel working in containment. The valve was inspected for abnormalities that could lead to the failure. The inspection did not reveal any problems. The valve was reassembled, and then successfully retested with an as found leak rate of 1.0546 SCFH. The most probable root cause of this event is failure of the check valve,to fully seat following shutdown of the RBCCW system for the LLRT. During the Unit 3 cycle 7 outage the check valve was a good performer with an as-found leakage of 0.3574 SCFH. Based on operating history of this valve, and observed condition of the valve during the inspection, additional corrective actions are not warranted. The valve is scheduled to be retested during the next refueling outage in accordance with the LLRT program requirements.

NRC FORM 3665 (6-1999)

II NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6'1998I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME' DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISIO NUMBER 2 of 5 Browns Ferry Nuclear Plant - Unit 3 05000 1998 - 004 00 TEXT (lf more spaceis required, use additional copies of NRC Form 366Ai (17)

PLANT CONDITIONIS)

At the time of the event, Unit 3 was shutdown in Mode 5 in the scheduled cycle 8 refueling outage. Unit 2 was in Mode 1 at 100 percent reactor power and approximately 3293 megawatts thermal. Unit 1 was shutdown and defueled.

II. DESCRIPTION OF EVENT A. Event:

On October 7, 1998, at approximately 1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br />, Central Daylight Time (CDT),,the Unit 3 primary containment total allowable leak rate was determined to have been exceeded. During performance of the surveillance instruction that quantifies the leakage through the Reactor Building Closed Cooling Water (RBCCW) [CC] system drywell [JM] inlet header primary containment supply check valve [ISV], it was determined that the leak rate of the valve was approximately 2491 standard cubic feet per,hour (SCFH). Hence, the leak rate of the valve exceeded the maximum allowable containment leak rate of 655.9 SCFH.

At approximately 1516 hours0.0175 days <br />0.421 hours <br />0.00251 weeks <br />5.76838e-4 months <br />, maintenance personnel [utility, non-licensed] initiated due local leak rate test (LLRT) of the RBCCW system penetrations. After establishing test conditions; it was determined that the required volume. test pressure of 50.6 psig could not be obtained on the inlet header primary containment supply check valve. The maximum obtainable pressure in the volume was 38 psig. It was subsequently determined that the calculated leak rate through the penetration was 2491 SCFH.

On October 8, 1998, at 0524 CDT,,the RBCCW system was placed back in service to provide cooling for personnel in containment. Subsequently on October 8, 1998, at 1907 CDT, the system was removed from service and the valve was inspected. Inspection of the valve did not identify any problems. On October.9, 1998, at approximately 0229 hours0.00265 days <br />0.0636 hours <br />3.786376e-4 weeks <br />8.71345e-5 months <br />, the valve was successfully retested. The as-left leak rate was determined to be 1.0546 SCFH.

As a result of the primary containment total allowable leak rate being exceeded, TVA is submitting this report in accordance with 10 CFR 50.73'(a) (2) (ii) (B), as a condition outside the design basis of the plant.

B. Ino erable Structures Com onents or S stems that Contributed to the Event:

None.

C. Dates and A roximate Times of Ma or Occurrences:

October 7, 1998, at 1825 CDT Primary containment total leak rate is determined to have exceeded the maximum allowable containment leak rate.

October 7, 1998, at 2130 CDT TVA made a four-hour non-emergency notification to NRC in accordance with NRC FORM 366 I6-199BI

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-I 998I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVIBIO NUMBER 3of5 Browns Ferry Nuclear Plant - Unit 3 05000 1998 - 004 - 00 TEXT llfmore space is required, use additional copies of NRC Form 366Al {17}

10 CFR 50.72(b)(2)(i).

October 9, 1998, at 0229 CDT The valve was successfully retested.

D. Other S stems or Seconda Functions Affected None.

E. Method of Discove The leak rate was determined to be unacceptable during the performance of LLRT, during cycle 8 refueling. outage activities.

F. 0 erator Actions None.

G. Safet S stem Res onses None.

III. CAUSE OF THE EVENT A. Immediate Cause The immediate cause of the event was that the RBCCW dlywell inlet header primary containment supply check valve failed the LLRT.

B. Root Cause The most probable root cause of this event is failure of the check valve to fully seat following shutdown of the RBCCW system for the LLRT. Inspection of the valve internals did not provide any indications of a cause for the LLRT failure. The valve seat internals were checked for corrosion, missing parts, or unusual wear. The valve disk was verified to operate freely from full open to full closed position.

Although the symptoms indicated that small particles under the valve seat could have caused the failure, no foreign material was located, consequently, this could not be substantiated as a root cause.

C. Contributin Factors Since the recovery of Unit 3 in November of 1995, no work has been performed on the RBCCW system which required breaching the system piping, so it is unlikely that any maintenance performed resulted in foreign material being introduced into the system piping.

The drywell chiller was used during this outage for the first time since recovery of Unit 3. The piping loop associated with the drywell chiller had been drained since that time. It is possible that some small particles of debris could have migrated from the chiller package into the main loop of RBCCW while the drywell chiller was in service. However, there was no conclusive evidence of an articles in the s stem.

NRC FORM 366 I6-1998)

(k NRC FORM 366A 0 u.s. NLICLEAR REGULAToRY CoMMIssIoN 16.1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL NUMBER 4of5 Browns Ferry Nuclear Plant - Unit 3 05000 1998 - 004 - 00 TEXT (ll more space is required. use additional copies ol iYRC Form 366Al {17)

IV. ANALYSIS OF THE EVENT The RBCCW system is a closed loop system consisting of pumps, heat exchangers, and necessary controls and support equipment that provide a continuous supply, of cooling water to certain equipment located inside primary containment. The portion of the RBCCW system inside primary containment and outside primary containment up to and including the containment isolation valves is considered a part of the primary containment boundary designed seismic class I. This ensures that the primary containment function is not compromised following a seismic event. Containment isolation. is provided by a motor operated valve controlled from the control room for the RBCCW discharge line, and an in-line check valve for the RBCCW supply line. The valve discussed in this report, the inline check valve, has no active automatic function during a postulated loss-of-coolant accident.

Following the initial test, the RBCCW system was returned to service. Subsequent inspection of the valve did not reveal any abnormalities that could be associated with the leakage. The valve was subsequently reassembled without installation. of replacement parts, and then successfully retested with an-as-found leak rate of 1.0546 SCFH.

During the Unit 3 cycle 7 outage the check valve was a good leak rate performer with an as-found leak rate of 0.3574 SCFH. History of the previous four LLRTs on the equivalent Unit 2.valve indicates an average leak rate of 0;045 SCFH. A review of the RBCCW system operation and methods used to perform the leak rate testing did not identify changes that would have any impact on the test results.

V. ASSESSMENT OF THE SAFETY CONSEQUENCES The purpose of primary containment is to ensure that the release of radioactive materials from the containment atmosphere will be restricted to.those leakage paths and associated leak rates assumed in the BFN accident analysis. This restriction, in conjunction with the leak rate limitation, will limit the site boundary dose limits to within those set forth in 10 CFR,part 100 during accident conditions. Additionally, the limitation of primary containment leakage rates ensures that the total containment leakage will not exceed the value assumed in the accident analysis at the peak accident containment pressure.

As discussed in section 5.2.3.5 of the BFN Updated Final Safety Analysis Report, lines that penetrate the primary containment, and which neither open into the primary system or open into primary containment, are provided with at least one isolation valve which may be located outside primary containment. Since the RBCCW system neither connects to the reactor primary system nor is open into primaly containment, a design basis accident within primary containment would not result in excessive leakage'through the RBCCW system.

Also, internal containment missile generation was evaluated under the BFN Design Baseline and Verification Program for the RBCCW system. It was determined that there are no credible missile hazards in the containment which could impact and damage the RBCCW piping and impact isolation capability.

As previously discussed, the RBCCW system piping within the primary containment neither connects to, nor is open to primary containment. The portion of the system inside containment is considered a part of the containment boundary and is seismic class 1. Also, there is no credible missile event which could impact and damage the piping. Because of these design features, failure of the check valve to close would not lead to a release of radioactive material. Therefore, TVA concludes that safet of the lant ersonnel and NRC FORM 366 (6-1998>

II NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION ia 199al LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISIO NUMBER 5of5 Browns Ferry Nuclear Plant - Unit 3 05000 1998 - 004 . 00 TEXT (If more speoe is required, use addi tionel copies of fVRC Form 366A/ (17) public was not compromised.

VI.'ORRECTIVE ACTIONS A. Immediate Corrective Actions The check valve was disassembled and inspected for indications of failure. No abnormal wear, missing parts, or other potential causes for failure were observed. The valve was subsequently reassembled and retested satisfactorily.

B. Corrective Actions to Prevent Recurrence The valve involved in this event is scheduled to be retested during the next refueling outage in accordance with the LLRT program requirements'. Based on operating history of this valve, the valve that performs the same function on Unit 2, and the observed condition of the valve during the inspection,,no additional corrective actions for the valve are warranted.

In order to.rule out the possibility of introduction of foreign material from the drywell chiller system, TVA plans to evaluate the drywell chiller system for corrosion products or foreign material prior to connection to.the RBCCW system in. future outages'.

VII. ADDITIONAL INFORMATION'.

Failed Com onents None.

B. Previous LERs on Similar Events This is the first report issued on the failure of this check valve. However, previous LERS 260/94008, 260/93002, 259/85039, and 296/84007 documented isolation valves that failed to pass LLRT. In the four previous reports, main steam isolation valve [SB] (MSIV) leak rates exceeded the technical specification limit. As required by Technical Specifications, the valves were repaired prior to being returned to service. Corrective actions associated with these pervious LERs would not have precluded this event.

Vill. COMMITMENTS None.

Energy industry Identification System (EEIS) system and component codes are Identified in the text with brackets (e.g., IXXI).

1 TVA does not consider these corrective actions regulatory commitments. TVA's Corrective Action Program will track completion of these actions.

NRC FORM 366 i6-1996)

II