ML18039A802

LER 99-002-00:on 990501,SRs for Single CR Withdrawal During Cold SD Were Not Adequately Implemented.Caused by Procedural Inadequacy.Revised Applicable Plant Surveillances.With 990614 Ltr

ML18039A802
Person / Time
Site:	Browns Ferry
Issue date:	06/14/1999
From:	Charlton E, Singer K TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-99-002-02, LER-99-2-2, NUDOCS 9906210092
Download: ML18039A802 (16)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM '(RIDS)

ACCESSION NBR:9906210092 DOC.DATE: 99/06/14'OTARIZED: NO DOCKET'.5 FACIL:50-260'rowns Ferry Nuclear Power .Station, Uni't 2,, Tennessee 05000260 AUTH. NAME AUTHOR AFFILIATION CHARLTON,E.D. Tennessee Valley Authority SINGER,K.W. . Tennessee Valley Authority RECZP.NAME . RECIPIENT AFFILIATION

SUBJECT:

LER 99-002-00:on 990501,SRs for single CR withdrawal during cold SD were not adequately implemented.'Caused. by procedural inadequacy. Revised applicable plant surveillances.With 990614 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 COPXES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL LPD2-2 PD 1 1 'DEAGAZIO',A 1 INTERNAL: AC 1 1. AEOD/SPD/RRAB 1 ILE CENTER 1 1 NRR/DIPM/IOLB 1 NRR'IP QMB 1 1 NRR'/DRIP/REXB'ES/DET/EIB 1

NRR/DSSA/SPLB 1 1 1 RGN2 FILE 01 1 1 EXTERNAL: L 'ST LOBBY WARD 1 1 LMITCO MARSHALL NOAC POORE,W. '1 1 NOAC QUEENER,DS NRC PDR 1 1 NUDOCS'ULL TXT NOTE TO ALL "RZDS" RECIPIENTS':

PLEASE'HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LI.'R REDUCE THE NUMBER OF COPIESiRECEIVED BY YOU OR YOUR ORGANIZATION'ONTACTTHE DOCUMENT CONTF DESK (DCD) ON 'EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 17 ENCL 17'

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Tennessee Valley Authority; Post Office Box 2000, Decatur, Alabama 35609.-2000 Karl,W. Singer Vice President, Browns Ferry Nuclear Plant'une 14, 1999 U.S. Nuclear Regulatory Commission 1'0 CFR 50.73 ATTN: Document Control Desk

'Washington, D. C. 20555

Dear Sir:

TENNESSEE VALLEY AUTHORITY 'BROWNS FERRY NUCLEAR PL'ANT (BFN)

UNIT 2 DOCKET 50-260 FACILITY OPERATING LICENSE 'DPR-52 LICENSEE EVENT REPORT (LER) 50-260/1999002 The enclosed report provides details concerning an event involving inadequate implementation of Technical .Specification surveillance requi'rements.

This condition is reportable in accordance with 10 CFR'0.73 (a) (2) (i) (B) as a condition prohibited by, the plant' technical specifications.

Sincerely, Karl W.. Sin r cc: See page 2 9't t06210092 9906k 4 PDR ,ADQCK 05000260 PDR 8

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U.S. Nuclear Regulatory Commission Page 2 June 1'4, 1999 Enclosure cc (Enclosure):

Mr. William O. Long, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville,,Maryland 20852-2739 Mr. Paul E. Frederickson, Branch Chief U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 NRC Resident Inspector Browns Ferry Nuclear Plant 10833'haw Road Athens, Alabama 35611

t'ai NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EKFIREs norse)soot (6-1996) Estimated burden per response to comply w)th this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into the ))cerning process and fed back to hdustry. Forward conunents regardhng LICENSEE EVENT REPORT (LER) burden estimate to the Records Management Branch (TA F33), U.S.

Nuclear Regulatory Commission, Washington. DC 205554001, and to the Paperwork Reduction Project (31500104). Olrce of Management and (See reverse for required number of Budget, Washington, OC 20503. If an information collection does not digits/characters for each b(ock) display a currently valid OMB control number. the NRC mey not conduct or sponsor, and a person is not required to respond to. the information collection.

FACIU'rY NAME (11 DOCKET NUMSER 12) PAOE 13)

Browns Ferry Nuclear Plant - Unit 2 05000260 1 of5 TITLE f4)

Surveillance Requirements For Single Control Rod Withdrawal During Cold Shutdown Not Adequately Implemented EVENT DATE (5) 'LER NUMBER (6l REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACIUTY NAME DOCKET NUMBER NUMBER NUMBER 05000 DOCKET NUMBER 05 01 1999 1999 002 000 06 14 1999 NA 05000 OPERATING THIS, REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR'll: (Check one or more) (11)

MQDE (9) 20. 2201 (b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)

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

LEVEL (10) 0 20.2203(a) (2) (i) 20.2203(a) (3) (ii) 50.73(a)(2)(nl) 73.71 20.2203(a) (2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv) QTHER 20.2203(a) (2)(iii) 50.36(c) (1) 50.73(a)(2)(v) Specify in Abstract below 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)

NAME TELEPHONE NUMBER gnclude Area Coda)

E. Dorris Charlton, Industry Affairs Specialist 256.729.7533 COMPLETE ONE LINE FOR EACH COMPONENT. FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABI.E NFRDS TO NFRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH 'AY YEAR YES No SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). DATE (15)

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

On May 14, 1999, an, inadequate implementation of a Technical Specification surveillance requirement (SR) was discovered in the performance of 2-SR-3.10.4, Verification of Surveillance Requirements for Single Control Rod Withdrawal - Cold Shutdown. During control rod drive (CRD) system testing, the SR for verification that all control rods, other than the control rod being withdrawn, are fully inserted was not performed at. the required frequency. The verification was performed at the beginning of the CRD testing evolution. However, it was not subsequently performed at the required 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency during the evolution. Although the plant surveillance instruction was written to apply to the withdrawal of one single control rod, the instruction was improperly used to document multiple withdrawals of single control rods during the testing evolution. The root cause of this event is procedural inadequacy which lead to a misinterpretation of the plant surveillance instruction 2-SR-3.10.4. Recurrence control will include revising the, applicable plant surveillances to address multiple individual control rod withdrawals for testing and maintenance.

This condition is reportable in accordance with 10 CFR 50.73(a)(2)(l)(B) as a condition prohibited by the plant's Technical Specifications.

NRC FORM 366B (6-1998)

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

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER 20f5 Browns Ferry Nuclear Plant - Unit 2 05000260 1999 002 000 TEXT /Ifmore space is required, use addirional copies of NRC Form 366AJ l17l I. PLANT CONDITION(S)

At the time the event occurred, Unit 2 was in mode 4, during the Unit 2 cycle 10 refueling outage, Unit 3 was operating at 100 percent power, and Unit 1 was shutdown and defueled.

II. DESCRIPTION OF EVENT A. Event:

During the performance of routine NRC inspection activities, an example of inadequate implementation of a Technical Specification (TS) surveillance requirement (SR) was discovered. Specifically, plant surveillance 2-SR-3.10.4, Verification of Surveillance Requirements for Single Control Rod Withdrawal-Cold Shutdown, was not performed at the intended frequency.

On May 12, 1999, a NRC Resident Inspector noted during review of surveillance 2-SR-3.10.4, that the verification of all control rods, other than the control rod being withdrawn, are fully inserted had not been completed at. the required 24'hour frequency during control rod drive (CRD) system [AA] movemefit.

Subsequent investigation completed by TVA'on May 14, 1999, confirmed that the SR was not adequately implemented from April 29, 1999, through May 1, 1999. Further investigation identified an additional example of inadequate implementation of this SR on May 3 and 4, 1999.

On April 29, 1999, at approximately 2258 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.59169e-4 months <br /> CDT, in preparation for the continuation of the ongoing CRD maintenance activities following entry into Mode 4 operation, plant operators began implementation of surveillance 2-SR-3.10.4; Verification of Surveillance Requirements for Single Control Rod Withdrawal-Cold Shutdown. On April 30, 1999, 0634 hours0.00734 days <br />0.176 hours <br />0.00105 weeks <br />2.41237e-4 months <br /> CDT, the Unit Operators verified conditions required by TS 3.10.4 and resumed CRD activities, and on May 01, 1999, at approximately 1459 hours0.0169 days <br />0.405 hours <br />0.00241 weeks <br />5.551495e-4 months <br /> CDT 2-SR-3.10;4 was completed.

On May 03, 1999, at approximately 0355 hours0.00411 days <br />0.0986 hours <br />5.869709e-4 weeks <br />1.350775e-4 months <br /> CDT surveillance 2-SR-3.10.4 was again performed to support additional CRD movement in Mode 4. The surveillance was completed on May 04, 1999, at approximately 1205 hours0.0139 days <br />0.335 hours <br />0.00199 weeks <br />4.585025e-4 months <br /> CDT.

In both instances the surveillance was performed in support of CRD'system timing adjustments which involved multiple single control rod withdrawals. The withdrawals continued over several 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time periods. The verification that all control rods were inserted was performed and documented at the start of the surveillance. The operators who performed the surveillance misinterpreted the applicability of the plant surveillance instruction applying it to multiple iterations of single control rod withdrawal. Since no single control rod was withdrawn for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the verifications required at the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency were not performed. Also, the operators did not document that all control rods were inserted prior to each control rod manipulation The failure to perform the SR in the intended frequency is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by the plant's Technical Specifications.

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

None NRC FORM 366 I6-1998)

i+i 0 NRC FORM 366A u.s. NucLEAR REGULATORY COMMissiON (6-19981 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1) DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL NUMBER 3of5 8rowns Ferry Nuclear Plant - Unit 2 05000260 1999 002 000 TEXT (lf more space is required, use addirional copies of NRC Form 366A/ l17)

Dates and A roximate Times of Ma or Occurrences:

May 12, 1999 NRC Resident Inspector Identified potential inadequate implementation of SR 3.10.4.

May 14, 1999, 1115 hours0.0129 days <br />0.31 hours <br />0.00184 weeks <br />4.242575e-4 months <br /> CDT TVA confirmed the inadequate SR implementation.

D. Other S stems or Seconda Functions Affected None E. Method of Discove This condition was discovered by a NRC Resident Inspector's during review of completed surveillance 2-SR-3.10.4.

F. 0 erator Actions None.

G., Safet S stem Res onses None III. CAUSE OF THE EVENT A. Immediate Cause Operations personnel failed to verify the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR of TS.

B. Root Cause The root cause of this event. is procedural inadequacy which lead to a misinterpretation of the plant surveillance instruction 2-SR-3.10.4. Specifically, the operators performing the surveillance incorrectly

. applied the surveillance to multiple iterations of single control rod withdrawal. The intent of the procedure was to be only applicable to the withdrawal of one single control rod.

Contributin Factors The layout of surveillance 2-SR-3.10.4 data sheets in a matrix format was misleading, suggesting that multiple iterations of single control rod withdrawal could be documented.

This special operations Limiting Condition for Operation (LCO) had been entered only once previously. On Unit 3 in October, 1998, the LCO was entered for similar activities related to the Unit 3 refueling outage.

The surveillance was used to document multiple'iterations of single control rod withdrawal, however, the required 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> verification was performed.

NRC FORM 366 (6-1998l

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME' DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REvlslON NUMBER 4 of 5 Browns Ferry Nuclear Plant - Unit 2 05000260 1999 002 - 000 TEXT illmore spaceis required, use addi rional copies of NRC Form 366A/ I17)

IV. ANALYSIS OF THE EVENT On April 29, 1999, at approximately 2258 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.59169e-4 months <br /> CDT, Technical Instruction O-TI-20, Control Rod Drive (CRD)

System Testing and Troubleshooting, was in progress with the plant in Mode 5. Concurrently, the reactor vessel head bolts were being tensioned to return to Mode 4 (Cold Shutdown). Plant surveillance instruction 2-SR-3.10.4 was initiated to allow continued CRD testing and be in compliance with TS 3.10.4, Single Control Rod

,Withdrawal - Cold Shutdown, after entering Mode 4. Subsequently, Mode 4.was entered at 0627 hours0.00726 days <br />0.174 hours <br />0.00104 weeks <br />2.385735e-4 months <br /> on April 30, 1999, when the reactor vessel head bolts were. fully tensioned.

Limiting Condition for Operation (LCO) 3;10.4 allows the reactor mode switch to be moved to the refuel position when in Mode 4 to allow withdrawal of a single control,rod, provided all other control rods are fully inserted and the requirements of specified LCOs (or other alternate actions) are met. Similarly, TS'SRs require periodic performance of the applicable SRs (or verification of the alternate actions) and verification that all other control rods are fully inserted. Likewise, Plant surveillance 2-SR-3.10.4 requires periodic verification that SRs for the specified LCOs have been performed within their periodicity (or alternate actions taken) and that all control rods, other than the one being withdrawn, are fully inserted.

Plant surveillance 2-SR-3.10.4 was originally written to apply to the withdrawal of a single control rod and included provisions for rod withdrawal for an extended period. of time. The data sheets for this surveillance were arranged in a matrix format to allow for multiple periodic verification of each of the supporting SRs (or alternate actions) and that all other control rods were fully inserted. The operators misinterpreted the matrix format of the data sheets to allow multiple iterations of single control rod withdrawal.

The surveillance provided documentation of initial verification that all control rods, other than the. one being withdrawn, were fully inserted into the core. The matrix, as stated above, was intended to allow for subsequent verifications at the required 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency'if the single control rod remained withdrawn for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When the matrix format was used to document multiple iterations of a single control rod withdrawal, operators placed an N/A in the space for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> check and did not perform the required verification since the individual control rods were not withdrawn for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Also, the operators did not document that all control rods were inserted prior to each control rod manipulation.

Due to inadequate SR, the required 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> verification of all control rods inserted was not performed by 2258

. hours on April 30, 1999, (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the start time of the SR), nor.by 0458 hours0.0053 days <br />0.127 hours <br />7.572751e-4 weeks <br />1.74269e-4 months <br /> on May 1, 1999 (the maximum allowable extension time of 25 percent of the frequency). The same misinterpretation and failure to perform the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> verification of all control rods inserted occurred during subsequent control rod movement on May 3-4, 1999, after CRD system valve maintenance.-

V. ASSESSMENT OF SAFETY CONSEQUENCES In addition to the requirement that all other control rods are fully inserted, TS 3.10.4 requires verification that LCOs are met for Refuel Position One-Rod-Out Interlock (LCO 3.9.2) and Control Rod Position Indication (LCO 3.9.4) or, alternately requires insertion of a contr'ol rod withdrawal, block. Finally, TS 3.10.4 requires verification that LCOs for Reactor Protection System (RPS) Instrumentation - Mode 5 requirements (LCO 3.3.1.1) and Control Rod OPERABILITY- Refueling (L'CO 3.9.5), or alternately, disarming of all other control rods in a five by five array centered on the control rod being withdrawn.

Durin the CRDtestin,theo eratorsverifiedthe erformanceof.theSRswhichsu ortLCOs3.9.2and3.9.4to NRC FORM 366 I6-1998)

igi 0 II NRC FORM 366A (6.1998) 0 U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER 5 of 5 Browns Ferry Nuclear Plant - Unit 2 05000260 1999 002 000 TEXT (If more space is required, use addi rional copies of hfRC Form 366A/ f17) ensure operability of the one-rod-out-interlock and LCOs 3.3.1.1 and 3.9.5 to ensure operability of,RPS instrumentation and control rods. During the period that SR 3.10.4 was applicable, with the exception of one and one half hours when the plant computer was unavailable, the plant computer showed that no more than one control rod was withdrawn at any time. Additionally, operator training emphasizes that only one control rod may be withdrawn;at any given time while in Mode 4 operation. Therefore, because there only one control rod withdrawn during the time that SR 3.10.4 was applicable, there was no actual or potential safety consequences as a result of this event, and this event did not adversely affect the safety of plant personnel or the public.

VI. CORRECTIVE ACTIONS A.,Immediate Corrective Actions Upon notification of NRC's concern regarding the adequacy of implementation of the SR, Operations personnel confirmed the. condition.

B. Corrective Actions to Prevent Recurrence Applicable plant sufveillances will be revised to address multiple individual control rod withdrawals for testing and maintenance. The verification for all other control rods fully inserted will be required prior to each control rod withdrawal and every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter if the control rod remains withdrawn.'The appropriate operations personnel will review the circumstances surrounding this event.'II.

ADDITIONALINFORMATION A. Failed'Com onents:

None.

B. Previous LERs on Similar Events:

None.

C. Safet S stem Functional Failure:

This event did not result in a safety system functional failure in accordance with NEI 99 02.

VIII. COMMITMENTS None.

Energy Industry Identification System (EIIS) system and component codes are identified in the text by brackets (e,g., POq).

'VAdoes not consider these corrective actions r ulato commitments. TVA's Corrective Action Pr ram viilltrack corn letion of these actions.

NRC FORM'366 l6-1996)

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