ML18038B700

LER 96-003-00:on 960501,unit 3 Scrammed on Low Reactor Water Level.Caused by Failure of Steam Packing Exhauster by Pass Flow Control Valve.Valve replaced.W/960530 Ltr

ML18038B700
Person / Time
Site:	Browns Ferry
Issue date:	05/30/1996
From:	Austin S, Machon R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-96-003-04, LER-96-3-4, NUDOCS 9606070228
Download: ML18038B700 (18)

Text

CATEGORY 1i REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9606070228 DOC.DATE: 96/05/30 NOTARIZED: NO DOCKET FACIL:50-296 Bz'.owns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION AUSTIN,S. Tennessee Valley Authority MACHON,R.'D. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 96-003-00:on 960501,unit 3 scrammed on low reactor water C level. Caused by failure of steam packing exhauster by pass flow control valve. Valve repl'aced.W/960530 ltr. A DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR ( ENCI 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 INTERNAI: ACR S 1 1 AEO~+RD/~B 2 2 AEOD/SPD/RRAB 1 1 1 NRR/DE/ECGB 1 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 NRR/DSSA/SRXB 1 1 RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 D EXTERNAL: L ST LOBBY WARD NOAC 'MURPHY,G.A 1

1 1

1 'OAC LITCO BRYCEIJ POOREPW.

H 2,2 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

4 NOTE TO ALL "RIDS" RECIPIENTS:

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

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

4l li Tennessee vatey Authonty. post crace Box 2000. Decatur. A'aoanta 3MG9 2000 R. D. (Rick) Machon Vce Presxtent. Browns Ferry Nuctear Ptant May 30, 1996 U.S. Nuclear Regulatory Commission 10 CFR 50. 73 ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of Tennessee Valley Authority BROWNS FERRY 'NUCLEAR PLANT - UNITS 1 2 AND 3 - DOCKET NOS. '50-259~ '260~ AND 296 - (BFN) g ~

FACILITY LICENSE DPR-33'g 52'ND 68 - LICENSEE EVENT REPORT 50-296/96003 The enclosed report provides details concerning a reactor scram on low reactor water level following the failure of the Steam Packing Exhauster Condensate Bypass Valve.

The shaft connecting the valve disk to the operator. failed resulting in momentary closure of the valve. This action was followed by tripping of two of the condensate booster pumps and one feedwater pump thus causing a low reactor water level.

TVA has determined that this valve is a single point failure with high scram potential. Therefore, TVA is evaluating the design of the Steam Seal Condenser and Steam Packing Exhauster Condensate Bypass Valve to determine if design enhancements can be made to prevent recurrence of this type of event.

This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv) as a condition that resulted in automatic actuation of any engineered safety feature including the, reactor protection system.

ttt606070228 'tt60530 PDR ADQCK 05000296 S PDR

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Oi U.S. Nuclear Regulatory Commission Page 2 May 30, 1996 There are no commitments made in this submittal. If you have any questions regarding this, please contact Pedro Salas at (205) 729-2636.

Sincerely, R..D., chon Enclosure cc (Enclosure):

Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Region 101 II Marietta Street, NW, Suite 2900 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 Rockville, Maryland 20852

41 <Qi NRC FORM 366. U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 31504104 (4.06) EXPIRES 04/30/88 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORYINFORMATlON COLLECDON REQUEST:

60.0 HRS. REPORTED LESSONS LEARNED ARE

.LICENSEE EVENT REPORT (LER) INCORPORATED INTO THE UCENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING (See reverse for required number of BURDEN ESllMATE TO THE INFORMATIONAND RECORDS digits/characters for. each bfock) MANAGEMENT BRANCH (T4 F33), UN. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 205564001 FACIUTY HANK LI) OOCKKT NUNBKIII1) FAOK IT)

Browns Ferry Nuclear (BFN) Plant Unit 3 05000296 1OF 6 TITLK Ie)

Unit 3 Scram On Low Reactor Water Level Due To Failure Of The. Steam Packing Exhauster Bypass Flow Control Valve EVENT DATE 5 LER NUMBER 6 REPORT DATE OTHER FACILITIES INVOLVED 8 FACIUTY NAM SEQUENTTAL REVISION DAY YEAR DAY YEAR 05000 NUMBER NUMBER NA FACIUTY NAM 05 01 96 96 003 00 05 30 96 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR: (Check one or more 11 MODE (8) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(vill) 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

POWER LEVEL (10) 100 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)(2)(iii) 20.2203(a)(4) 50.36(c)(1)

)( 50.73(a)(2)(iv) 50.73(a)(2)(v) 6 'n OTHER Abstract below or in)tRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

UCENSEE CONTACT FOR THIS LER 12 NAME TELEPNoNK NUMBER Lrnrrrrde Area code)

Steve Austin, Ucensing Engineer (205) 729-2070 COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORTi 13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPROS TO NPRDS SJ FCV F127 SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yea, complete EXPECTED SUBMISSION DATE).

X NO DATE (15)

BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten linea) (16) 0n May 1, 1996, at approximately 1110 hours0.0128 days <br />0.308 hours <br />0.00184 weeks <br />4.22355e-4 months <br />, a full reactor scram on low reactor water level occurred. At 1109 hours0.0128 days <br />0.308 hours <br />0.00183 weeks <br />4.219745e-4 months <br />, a "Condensate Booster Pump Suction Pressure Low" alarm was received in the Unit 3 Main Control Room."A" This alarm re suited from the closure of the Steam Packing Exhauster Condensate Bypass Valve. At 1110 hours0.0128 days <br />0.308 hours <br />0.00184 weeks <br />4.22355e-4 months <br />, the reactor fe edwater pump tripped. The reactor automatically scrammed when the vessel level reached +11.2 inches. At Q5 in ches the High Pressure Coolant Injection and Reactor Core Isolation Cooling systems auto initiated and injected into th e vessel. The automatic Engineered Safeguard Features and automatic isolation or actuations occurred as expected.

The cause of the valve failure was a material defect in the valve shaft. Specifically, the shaft contained a material defect in a notch sensitive area when subjected to transients that caused rapid valve position changes, resulted in high st resses being placed on the shaft, causing the shaft to fail. Additionally, personnel involved in disposition of a crack in th e shaft inappropriately determined the condition acceptable.

The valve and pneumatic. operator was replaced with a manual operator. This operator will remain in place until disposition of the system design. TVA is evaluating the design of the Steam Packing Exhauster Condensate Bypass Valve to see if design enhancements can be made that could prevent recurrence of this type of event. The Engineering individuals involved in the event received personnel corrective actions. TVA will provide information to the applicable pe rsonnel on management's expectations for evaluating plant conditions.

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NRC FORM 866A U.S. NUCLEAR REGULATORY COMMISSION

(~

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET NUMBER NUMBER Browns Ferry Unit 3 05000296 3 of 6 96 -- 003 00 l110fe SPQCO CO/Vlf ~ Vie 11NAQ COPNS orm 1 )

The reactor scram,was reset by 1120 hours0.013 days <br />0.311 hours <br />0.00185 weeks <br />4.2616e-4 months <br /> and the affected systems returned to service by 1141 hours0.0132 days <br />0.317 hours <br />0.00189 weeks <br />4.341505e-4 months <br />. A Reactor Feed Pump was placed into service to maintain reactor level and RCIC was returned to standby readiness by 1148 hours0.0133 days <br />0.319 hours <br />0.0019 weeks <br />4.36814e-4 months <br />. All safety systems responded's expected during the reactor scram.

This event is reportable in accordance with 10 CFR 50.73 (a)(2)(iv), as any event or condition that resulted in manual or automati'c actuation of, any engineered safety feature, including the reactor protection system.

B. Zno rable Structures, Co onents, or 8 stems that Contributed to e Event:

None.

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

May 1, 1996, at 1110 CDT The Unit 3 reactor received a full scram due to low reactor water level.

May 1, 1996, at 1120 CDT The scram was reset.

May 1, 1996,'t 1203 CDT TVA made a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification to NRC in accordance with 10 CFR 50.72 (b)(1)(iv) for HPCI injection.

Other S stems or Seconda Functions Affected:

Following this event, the UO attempted to restart the Reactor Recirculation [AD] pumps. Pump 3A failed to initially start.

However, it was eventually returned to service.

successfully returned to service.

Pump 3B was All three of the Low Pressure Feedwater Heater strings isolatedthe as required during the event. However, due to problems with valve operators they could not be immediately returned to service. TVA has concluded the isolation valve operators experienced thermal overload or torque switch problems.

Therefore, the UO controlled vessel level utilizing RCIC until the. low pressure feedwater heater strings could be unisolated and the feedwater system returned to service. TVA is tracking this issue through the corrective action program.

method of Discove The event was immediately known to the Unit 3 Main Control Room Operators upon receiving a "Condensate Booster Pump Suction Pressure Low" alarm. This was followed by the reactor scram and subsequent ESF actuations.

NRC FORM 366A (4-95)

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET DER NUMBER NUMBER NUMBER Browns Ferry Unit 3 05000296 4 of 6 96 -- 003 -- 00 X more space rs rsqm, use rtsrea copes orm 1 )

ratoz Actions:

Operator actions taken during this event were in accordance with applicable procedures. At the onset of 'the event, the UO attempted to reduce the reactor power to return the reactor water level to normal. At +12 inches, the Assistant Shift Operations Supervisor (Utilityr Licensed) directed the reactor to be manually scrammed. However, before .the manual scram could be completed, the reactor received an automatic scram at +11.2 inches (low-water level). The reactor operator then performed actions described by Abnormal Operating Instruction "Reactor Scram," bringing the reactor to hot standby condition.

G. Safet S stem Res onses:

The safety systems listed in Section II.A of this report responded to the reactor scram as designed.

ZZZ. CAUSE OF THE EVENT Zmmediate Cause:

The immediate cause of the reactor scram was the failure of the shaft connecting the valve disk to the actuator resulting in a loss of control of the disk. This was followed by trip of two of the condensate booster pumps and one reactor feedwater pump on low suction pressure.

B. The Cause Of The Event:

The, root cause of the valve failure was a material defect in the valve shaft. Specifically, the shaft contained a material defect in a notch sensitive area when subjected to transients that caused rapid valve position changes, resulted in high stresses

.being placed on the shaft, causing the shaft to fail.

Metallurgical analysis of .the shaft supports the conclusion that failure was related to high'tress impacts, resulting in complete torsional failure of the shaft. BFN was experiencing erratic valve operation, which on occasion, the valve would slam closed then reopen. This erratic operation resulted in cracks in the operator [FCO) mounting bracket and the valve shaft.

TVA had an opportunity to prevent the reactor scram. A crack on the valve stem was detected by craft personnel during maintenance activities on the valve. Engineering evaluation and acceptance of the condition were not adequate. The personnel involved in the disposition of the crack inappropriately determined the condition of the shaft acceptable.

NRC FORM 366A (4-95)

~I Ii NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION ZaER NUMBKR NUMBER NUMBER Browns Ferry Unit 3 05000296 5 of 6 96 003 -- 00 X more space rs reqlrrr; use rrsrna copes Ollll 4 ZV. ANALYSZS OF THE EVENT Plant safety systems and'ssociated components performed as designed during the event. Operations personnel stabilized the reactor in accordance with applicable plant procedures.

There were no operator actions that could have prevented this event.

The Steam Packing Exhauster Condensate Bypass Valve is a single point failure with high scram potential. Closure of the valve at '100 percent reactor power has in the past lead to a low-water level reactor, scram. The failure of the valve resulted in the trip of two condensate booster pumps and one reactor feedwater pump on low suction pressure; thereby, resulting in a low-reactor water level scram.

This event was categorized as a partial loss of reactor feedwater.

Full loss of feedwater is an analyzed plant transient and bounds the circumstances associated with this event. Therefore, the event did not affect the health and safety of plant personnel or the public.

V. CORRECTIVE ACTZONS Znmediate Corrective Actions:

The affected systems were restored to operable status. The valve was replaced with a manual valve.

B. Corrective Actions to Prevent Recurrence:

The original operator '(an air operator) was replaced with a manual operator. This operator will remain in place until disposition of the system design can be finalized.

2. The Engineering personnel involved in the event received personnel corrective action in accordance with TVA policy.

3. In order to address the single point failure aspect of the valve, TVA is evaluating, the design of the Steam Seal Condenser [COND] and'team Packing Exhauster Condensate Bypass Valve to determine if design enhancements could be made to prevent recurrence of this type of event.

TVA will provide the applicable personnel information regarding management'.s expectations for evaluating plant conditions.'VA does not consider these actions Regulatory Commitments. That is, they are not actions required to restore 'compliance with obligations. Obligation means an action that is a legally binding requirement imposed through applicable rules, regulations, orders, and licenses.

The TVA corrective action program. will track completion of the corrective actions 3 and 4.

NRC FORM 366A (4-95)

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