ML18039A695

LER 99-002-00:on 990122,LCO Was Not Entered During Calibration Testing of 3D 480 Volt Rmov Board.Caused by Personnel Error.Tva Has Briefed Operations Personnel to Preclude Recurrence of Event.With 990219 Ltr

ML18039A695
Person / Time
Site:	Browns Ferry
Issue date:	02/19/1999
From:	Rogers A, Singer K TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-99-002, LER-99-2, NUDOCS 9903020320
Download: ML18039A695 (20)

Text

CATEGORY 1i REGULATORY INFORMATION DISTRIBUTION SYSTEM (RXDS) j ACCESSION NBR:9903020320 DOC.DATE: 99/02/19 NOTARIZED: NO DOCKET ZACIL:50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTHNANT AUTHOR AFFILZAT1ON ROGERS,A.T. Tennessee Valley Authority SINGER,K W. Tennessee Valley Authority RECXP.?CAME RECXPZENT AFFZLZATION

SUBJECT:

LER 99-002-00:on 990122,LCO was not entered during calibration testing of 3D 480 volt RMOV board. Caused by personnel error.TVA has briefed operations personnel to preclude recurrence of event. With 990219 ltr.

DISTRI'BUTTON CODE: IE22T COPIES RECEIVED:LTR I ENCL I SIZE:

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

NOTES:

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Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama'35609-2000 February 19, 1999 Karl W. Singer Vice President, Browns Feny Nucfear Rant U.S. Nuclear Regulatory Commission 10 CFR 50.73 ATTN: Document Control Desk Washington, D. C. 20555

Dear Sir:

BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 3 DOCKET NO. 50-296 FACILITY OPERATING LICENSE DPR-68 LICENSEE EVENT REPORT (LER) 50-296/1999002 The enclosed report provides details concerning failure to enter a Limiting Condition for Operation (LCO) for an inoperable Unit 3 Reactor Motor Operated Valve Board due to personnel error.

This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's Technical Specifications.

Sincerely, Karl W. S ger cc: See page 2 9903020320 9902i9 PDR ADQCK OM00296 S PDR

I U.S. Nuclear Regulatory Commission Page 2 February 19, 1999 Enclosure cc (Enclosure):

Mr. Paul Fredrickson, 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 Shaw. Road Athens, Alabama 35611 Mr. L. Raghavan, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pi.'ke Rockville, Maryland 20852-2739

0 0 I

0 !I NRC FORM 366 U.S. UCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EKPIREs I6-1996) oeldonoos Estimated burden per response to comply with th)s mandatory informagon codecoon request: 50 hnL Reported lessons learned are 'ncorporated into LICENSEE EVENT REPORT (LER) the rKens)ng process and fed back to hdusby. Fowsrd comments reganrIng burden estimate to the Records Management Bnmch (TA F33). U.S.

Nudear Reguhtory Comm)sslon. Washington, DC 20555000), and to the (See reverse for required number of Paperwork Reducdon pro)set (31500104), Oidce of Management and digits/characters for each block) Budget, Washington, DC 20503. If an infcrmsdon codecgon does not disp)ey a currently vadd 0MB contre) number, the NRC may not conduct or sponsor. and a person is not required to respond to. the Infer)nation codectkxL FAOIUTY NAME 11) OOCKEr NUMBER I2) PAOE (3)

Browns Ferry Nuclear Plant Unit 3 05000296 1of5 TITLE I4)

LCO Not Entered During Calibration Testing Of 3D 480 Volt RMOV Board EVENT DATE l5) LER NUMBER (6) REPORT DATE (7) 0 ER C S 0 ED 6)

MONTH OAY YEAR SEQVE)mAL REVISION A ILI OOCKET NUMBER NUMBER NUMBER NA OOCKET NUMBER 01 22 99 1999 002 00 02 19 99 NA OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or mora) (11)

MODE (9) 20.2201(b) 20. 2203(a) (2) (v) 50.73(a) l2)(B(B) 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) 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) (4) 50.73(a)(2)(iv) OTHER 20.2203(a) l2) (ul) 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)

TELEPHONE NUMBER Seclude Ares Cede)

Anthony T. Rogers, Senior, Ucensing Project Manager (256) 729-2977 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO SYSTEM COMPONENT MANVFACTURDI REPORTABLE NPROS TO NPROS NA SUPPLEMENTAL REPORT EXPECTED l14) EXPECTED MONTH DAY YES NO SUBMISSION X DATE (15)

(If yes, complete EXPECTED SUBMISSION DATE).

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

On January 4, 1999; a relay calibration on 3D 480 volt Reactor Motor Operated Valve (RMOV) Board was performed which disabled the auto-transfer logic of the board. The plant's Technical Specifications (TS) Bases state that the 3D and 3E 480 volt RMOV Boards are to be considered inoperable if the auto-transfer capability between the normal and alternate power supply for these boards is inoperable for any reason. TS 3.8.7 Required Action C.1 requires that the respective Residual Heat Removal (RHR) subsystem supported by the board must be declared inoperable for Low Pressure Coolant Injection (LPCI) immediately. However, the appropriate LCO had not been entered and,TS 3.8.7 Required Action C.1 had not been satisfied by declaring the affected RHR subsystem immediately inoperable.

This condition was discovered on January 22, 1999, when Operations personnel conducted a review of past performances of this calibration procedure and determined that on January 4, 1999, the proper LCO had not been entered.

The root cause of this event was personnel (utility-licensed, operator) error resulting from failure to recognize the LCO condition. TVA has briefed Operations personnel to preclude recurrence of this event.

This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's TS.

NRC FORM 366B (6-1999)

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

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEGUENTIAL NUMBER 2 of 5 Browns Ferry Nuclear Plant - Unit 3 05000296 1999 002 pp TEXT fffmore speceis required, use eddidonel copies of NRC Form 366Ai I17)

I. PLANT CONDITION(S)

At the time of the discovery of this condition, Unit 2 and Unit 3 were operating at 100 percent power and Unit 1 was shutdown and defueled.

II. DESCRIPTION OF EVENT A. Event:

On January 4, 1999, a relay calibration on 3D 480 volt Reactor Motor Operated Valve (RMOV) Board was performed which disabled the auto-transfer logic of the board. The plant's Technical Specifications (TS) Bases state that the 3D and 3E 480 volt RMOV Boards are to be considered inoperable if the auto-transfer capability between the normal and alternate power supply for, these boards is inoperable for any reason. TS 3.8.7 Required Action C.1 requires that the respective Residual Heat Removal (RHR) subsystem supported by the board must be declared inoperable for Low Pressure Coolant Injection (LPCI) immediately. However, the appropriate LCO had not been entered and TS 3.8.7 Required Action C.1 had not been satisfied by declaring the affected RHR subsystem immediately inoperable.

This condition was discovered on January 22, 1999, when Operations personnel conducted a review of past performances of this procedure and determined that on January 4,,1999, the proper LCO had not been entered.

The root cause of this event was personnel (utility-licensed operator) error resulting from failure to recognize the LCO condition. TVA has briefed Operations personnel.to preclude recurrence of this event.

As a result of not entering the LCO on January 4, 1999, this report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's TS.

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

None.

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

, January 4, 1999, 0918 hours0.0106 days <br />0.255 hours <br />0.00152 weeks <br />3.49299e-4 months <br /> CST Unit 3 Control Room Unit Supervisor approved performance of relay calibration on 3D 480V RMOV Board.

JanuaIy 4, 1999, 1340 hours0.0155 days <br />0.372 hours <br />0.00222 weeks <br />5.0987e-4 months <br /> CST Relay calibration was completed on 3D 480V RMOV Board.

January 22, 1999, 1626 hours0.0188 days <br />0.452 hours <br />0.00269 weeks <br />6.18693e-4 months <br /> CST As a result of review of past performances of the relay calibration, Operations personnel determined the LCO was not entered'in accordance with the plant's TS for the relay calibration on 3D 480V RMOV Board which occurred on Janua 4, 1999.

NRC FORM 366 {6-1998)

0 II NRC FORM 366A U.S.'UCLEAR REGULATORY COMMISSION I6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL NUMBER 3 of 5 Browns Ferry Nuclear Plant - Unit 3 05000296 1999 " 002 - 00 TEXT (/f more space is required, use additional copies of IVRC Form 366A J I17)

D. OtherS stems or Seconda Functions Affected None.

E. Method of Discove This event was discovered when Operations personnel reviewed the Unit 2 and Unit 3 narrative logs for prior performances of the relay calibration for the D and E 480 volt RMOV Boards.

F. 0 erator Actions None.

G. Safe S stem Res onses None.

III. CAUSE OF THE'EVENT A. Immediate'Cause The'immediate cause of this event was failure to enter an LCO as specified by the plant's TS.

B. Root Cause The root cause of this event was personnel (utility-licensed operator) error resulting from failure to recognize an LCO condition existed. This LCO is based on a requirement specified by Improved TS which were implemented July 27, 1998. Custom TS did not require an LCO entry for this condition.

IV. ANALYSIS OF THE EVENT I ASSESSMENT OF SAFETY CONSEQUENCES Performance of 3-ETU-SMI 3-.48RMOVDE, Procedure for Making 18 Month Relay Calibrations on 480 Volt Reactor MOV Boards 3D and 3E, temporarily disables the auto-transfer scheme for the 3D or 3E RMOV Board (whichever is being tested). TS Bases 3.8.7 Section C.1 states that 480 volt RMOV Board 3D or 3E must be considered inoperable if the auto-transfer capability between the normal and alternate power supply (LPCI Motor Generator sets) is inoperable for any reason. The LPCI Motor Generator sets are fed from 480 volt Shutdown Boards which are fed from 4160 volt Shutdown Boards which have emergency power supplied from their respective diesel generators.

The purpose of the auto-transfer capability is to ensure that RMOV Boards D and E which supply power for the RHR Loop I and Loop II LPCI subsystem inboard injection valves have alternate electric power available to operate following loss of normal power. The ability to provide alternate power maintains redundancy for the LPCI function for certain postulated failures of upstream power supplies. Therefore, inoperability of the auto-transfer capability is not allowed for extended periods of time. Accordingly, the TS Bases provide that the board be considered inoperable if the auto-transfer feature is inoperable. This, in turn, results in the NRC FORM 366 I6-1998)

0 ri

NRC FORM 366A U.S.'UCLEAR REGULATORY COMMISSION Ie-I 998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL NUMBER 4 of 5 Browns Ferry Nuclear Plant - Unit 3 05000296 1999 002 00 TEXT (lfmore space is required, use addi donal copies of NRC Form 366A/ l17)

IV. ANALYSIS OF THE EVENT/ASSESSMENT OF SAFETY CONSEQUENCES (continued) required entry into a seven day LCO. TS 3.8.7 Required Action C.1 requires declaring the affected RHR subsystem inoperable immediately. However, during the January 4, 1999, performance of the relay calibration for 3D 480 volt RMOV Board, the Unit 3 Control Room Unit Supervisor (utility-licensed operator) did not correctly recognize that performance of 3-ETU-SMI 3-.48RMOVDE would render the 3D 480 volt RMOV Board inoperable. As a result, RHR Loop I (LPCI mode) was not declared inoperable immediately as required by TS 3.8.7 Action C.1. The relay calibration was completed the same shift and the 3D 480 volt RMOV Board was returned to an operable condition. Consequently, this condition (inoperable LPCI subsystem) lasted only about four and one half hours. TS permit seven days to restore operability.

No other Emergency Core Cooling Systems (ECCS) were inoperable during the'time 3D 480 volt RMOV Board auto-transfer capability was defeated. Therefore, while in this condition, the plant was operated within the requirements of the TS and remaining operable ECCS subsystems provided adequate core cooling capability during a loss of coolant accident.

Accordingly, the safety of the plant, its personnel, and the public was not compromised.

V. CORRECTIVE ACTIONS A. Immediate Corrective Actions None.

B. Corrective Actions to Prevent Recurrence Shift Managers briefed crews concerning this event prior to their next assumption of shift.

During a two week period, the affected Unit Supervisor will have a peer check performed for work activity authorizations. Any deficiencies will be corrected and reported to the Operations Superintendent.'ll shift crews will review and discuss this event during shift turnover for heightened awareness and to ensure management expectations are clear on this issue.'hift Managers will conduct one-on-one counseling with each Unit Supervisor to stress. the expectations of holding a pre-job brief, exhibiting the proper questioning attitude, and personnel accountability.

'VA does not consider this corrective action a regulatory commitment. The completion of this item will be tracked in TVA's Corrective Action Program NRc FoRM 366 l6-1998)

ik NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (8-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL NUMBER 5 of 5 Browns Ferry Nuclear Plant - Unit 3 05000296 1999 002 00 TEXT (If more speceis required, use edditionel copies ol NRC Form 366A/ l17)

VI ADDITIONALINFORMATION

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A. Failed Com onents None.

B. Previous LERS on Similar Events LER 296/97005 describes an event where an LCO was not entered for a malfunctioning valve in which the root cause was that Operations personnel lacked a questioning attitude and had a predetermined mindset regarding the failure mechanism. The event described in LER 296/97005 was different circumstances'however similar to this event in that an LCO was not entered and one of the contributing factors was lack of a questioning attitude. The corrective actions for LER 296/97005 included counseling of the personnel involved, crew briefings on the event, and'sensitivity training for Operations personnel involving LCO entry.

No other LERs were identified where an LCO was not entered when required.

C. Additional Information None.

VII. COMMITMENTS None.

NRC FORM 366 l6-1998)

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