ML18038B612

LER 95-008-00:on 951228,core Thermal Power Exceeded Operating License Maximum Power Level Due to Drifting Temp Transmitter.Reduced Reactor power.W/960208 Ltr

ML18038B612
Person / Time
Site:	Browns Ferry
Issue date:	02/08/1996
From:	Hseih C, Machon R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-95-008-02, LER-95-8-2, NUDOCS 9602150048
Download: ML18038B612 (16)

Text

GA.J. 8 VV6.L J.

REUULRTOINPORNATION DISTRIBUTION I+TEN (RIDE)

! ACCESSION NBR:9602150048 DOC.DATE: 96/02/08 NOTARIZED:,NO DOCKET FACIL.:,50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION HSEIHPC.S. Tennessee Valley Authority MACHON,R.D. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 95-008-00:on 951228,core thermal power exceeded operating license maximum, power level due to drifting temp transmitter. Reduced reactor power.W/960208 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR i ENCL ~ SIZE:

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

NOTES:

COPIES RECIPIENT COPIES E'ECIPIENT ID CODE/NAME ..LTTR ENCL ID CODE/NAME ,LTTR ENCL PD2-3-PD 1 1 WILLIAMS,J. 1 1 INTERNAL: ACRS 1 1 B 2 2 AEOD/SPD/RRAB 1 1 E CENTER 1 1 NRR/DE/ECGB 1 1 NRR DE EE B 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 1 1 LITCO BRYCE I J H 2 NOAC MURPHYPG.A 1 1 NOAC POOREPW. 1 0 NRC'DR 1 1 NUDOCS FUL'L TXT 1 U

N NOTE TO ALL "RIDS" RECIPIENTS:

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

PULL TEXT CONVERSION REQU:RED TOTAL NUMBER'F COPIES REQUIRED: LTTR 26 ENCL 26

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Choice Tennessee Va ~ Au!honty, Post Box 2000. Decatur, ~~35609.2000 February 8, 1996 R. D. (Rick) Machon Vice Prestdent, Brow'eny Nuclear Plant U:.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Dear Sir:

BROGANS FERRY NUCLEAR 'PLANT (BFN) UNITS 1', 2 r AND 3 DOCKET NOS. 50-259, 260, and 296 FACILITY OPERATING LICENSE DPR 33r 52r AND 68 VOLUNTARY LICENSEE EVENT REPORT 50-296/95008 The enclosed report provides details concerning an event during which the core thermal power exceeded the operating license's maximum power level of 3293 Megawatts thermal (MWt). The 8-hour average reactor power was calculated to be about 3306 MWt with an estimated peak power of 3328 MWt.

This is a voluntary report.

Sincerely, R. . Machon cc: See page 2 J 44 ~ts ~rp 9602i50048 960208 tII PDR ADOCK 05000296 S PDR'

tl U. S. Nuclear Regulatory Commission Page 2 February 8, 1996 Enclosure

.cc Enclosure):

Mr. Mark S. Less'er, Branch Chief U.S. Nuclear Regulatory Commission Region II Suite 101 Marietta Street, NW',

Georgia 30323 2900'tlanta, NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw, Alabama 35611 Road'thens, Mr. J. F. Willi:ams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville. Pike Rockville, Maryland 20852

41 NRC FORM 366 . NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150%104 (4-95) EXPIRES 04/30NS ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST:

60.0 HRS. REPORTED LESSONS LEARNED ARE LICENSEE EVENT REPORT (LER) INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING (Sae reverse for required number of BURDEN ESTIMATE TO THE INFORMATIONAND RECORDS digits/characters for each block) MANAGEMENT BRANCH (TED F33), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20%554)001.

FACILITY NAME II) OOCKE'T NUMBER (2) PAOE IT)

Browns Ferry Nuclear Plant (BFN) Unit 3 05000296 1OF5 TITLE Ie)

Core Thermal Power Exceeded Operating License's,Maximum Power Level due to a Drifting Temperature Transmitter EVENT DATE (5) LER NUMBER 6) REPORT DATE P) OTHER FACILITIES INVOLVED (6)

FACIUTY NAME SEQUENTIAL REVISION MONTH DAY YEAR MONTH DAY YEAR 05000 NUMBER NUMBER NA FACIUTY NAME DOCKET NUMBER 12 28 95 95 008 00 02 08 96 NA 05000 OPERATING THIS REPORT IS SUBMITTED PURS UANT TO THE REQUIREMENTS OF 10 CFR: (Check one or more) (11)

MODE (6) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii) 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)(4) 50.73(a)(2)(iv) X OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Spec'n Abstract below or In FtRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12) rIAME TELEPHQNE NUMBER IIoetvde Area code)

Clare S. Hsieh, Compliance Engineer (205) 729-2635 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS X IM R369 SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE).

X NO DATE (15)

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

On December 28, 1995, manual heat balance calculations of the Unit 3 reactor thermal power indicated that the average core thermal power for an 8-hour period had exceeded the operating license's maximum power level of 3293 Megawatts thermal. At approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> on December 27, 1995, a feedwater temperature point on the plant process computer started to gradually increase until it reached 391 degrees F at approximately 1852 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.04686e-4 months <br />. To compensate for the slight decrease in indicated core thermal power, the Unit Operator (UO) increased power. At approximately 0045 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> on December 28, 1995, the feedwater temperature indication suddenly increased to 402 degrees F with a corresponding decrease in indicated thermal power. The UO noticed the power decrease without a corresponding decrease in electrical power. Subsequently, the UO noticed the abnormally high feedwater temperature indication, and reduced reactor power. The event was caused by the gradual failure of a feedwater temperature instrument. This provided, inaccurate input to the nuclear heat balance resulting in an inaccurate indication of reactor power. As a result, the operator responded to the observed plant indication (i.e., decrease in power level) and proceeded to adjust recirculation flow. The immediate correction action was to reduce reactor power. TVA is investigating the root cause of the failed instrument. Other corrective actions include briefing personnel on the lessons learned from this event and revising the recirculation system operating instruction to ensure it contains guidance for review of heat balance inputs.

This is a voluntary report.

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

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TEXT CONTINUATION FACILITYNAME 1 OOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER NUMBER Browns Ferry Unit 3 05000296 2 OF 5 95 008 00 TEXT (lfmore space is required, use additional copies of NRC Form 366A) (17)

I. PLANT CONDITIONS At, the time of this event, Unit 3 and Unit 2 were operating at approximately 100 percent power. Unit 1 was shutdown and defueled.

DESCRIPTION OF EVENT A. Event:

On December 28, 1995, at approximately 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> Central Standard Time (CST), manual heat balance calculations of the Unit 3 reactor thermal power indicated that the average core thermal power for an 8-hour period ending approximately 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on December 28 had exceeded the operating license's maximum power level of 3293 Megawatts thermal (MWt). The actual 8-hour average reactor power was later calculated to be approximately 3306 MWt with an estimated peak power of 3328 MWt.

On. December 27, 1995, at approximately 1455 hours0.0168 days <br />0.404 hours <br />0.00241 weeks <br />5.536275e-4 months <br />, a temperature transmitter (reactor feedwater tSJ] inlet A1 temperature transmitter 3-TT-3-48A tTT]) which had previously been deleted from computer processing due to its wide swings in readings was recalibrated by maintenance personnel [utility, nonlicensed] and reentered'into the plant computer for feedwater temperature point 348A. At approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, feedwater temperature point 3<8A started to gradually increase until it reached 391 degrees F at approximately 1852 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.04686e-4 months <br />.

During this time period, the Integrated Computer System (ICS) heat balance display indicated a slight reduction in core thermal power. To compensate for the slight power reduction and to maintain an indication of near full power on the ICS display,'the Unit Operator (UO) futility, licensed] increased reactor recirculation [AD] pump speed to maintain indicated average core thermal power between 3289 and 3293 MWt.

Feedwater temperature point 3-48A remained at about 391 degrees F until approximately 0045 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> on December 28, 1995 when it suddenly increased to 402 degrees F. This also resulted in a sudden decrease in core thermal power on the ICS heat balance display. The UO noticed this sudden decrease in core thermal power (approximately 20 MWt) with no corresponding change in generator output (MWe). The operator investigated this change and noted that the indicated feedwater temperature point was abnormally high. As a conservative action, reactor power was.reduced approximately 15 MWt. Subsequent troubleshooting on feedwater temperature point 348A determined that transmitter 3-Tl-348A had drifted high resulting in the increase in temperature readings. Foilowing discovery of the drifting transmitter, at approximately 0112 hours0.0013 days <br />0.0311 hours <br />1.851852e-4 weeks <br />4.2616e-5 months <br /> temperature point 3<8A was deleted from processing and indicated thermal power immediately increased above 3293 MWt. Power was again reduced to less than 3293 Mwt.

TVA is reporting this event as a voluntary report.

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

None.

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

LICENSEE &TENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER NUMBER Browns Ferry Unit 3 05000296 3 OF 5

, 95 008 00 TEXT (lfmore spaceis required, use additional copies of NRC Form 366A) (17)

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

December 27, 1995 at 1455 hours0.0168 days <br />0.404 hours <br />0.00241 weeks <br />5.536275e-4 months <br /> CST Transmitter 3-TT-3-48A recalibrated and reentered into the plant computer for feedwater temperature. point 3<8A.

at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> CST Feedwater temperature point 3-48A on the computer began to gradually increase until it reached 391 degrees F at 1852 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.04686e-4 months <br />.

Recirculation flow was increased to compensate for the indicated decrease in core thermal power.

December 28, 1995 at 0045 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> CST Feedwater temperature point 3-48A suddenly increased to 402 degrees F. Indicated reactor power dropped approximately 20 MWt with no change in MWe. UO reduced core thermal power.

at 0112 hours0.0013 days <br />0.0311 hours <br />1.851852e-4 weeks <br />4.2616e-5 months <br /> CST Feedwater temperature point 3-48A deleted from computer processing. Reactor power reduced to less than 3293 MWt.

at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> CST 8-hour average reactor thermal power for period ending approximately 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> calculated to be above 3293 MWt.

D. Other S stems or Seconda Functions Affected:

None.

Method of Discove The. UO observed a sudden decrease in indicated reactor thermal power without a corresponding, decrease in electrical power. The UO then. discovered the abnormally high indication of feedwater temperature point 3<8A on the plant process computer.

F. 0 erator Actions:

Between approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> and 1852 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.04686e-4 months <br />. on December 27, 1995, the UO adjusted recirculation flow to maintain near full power indication on the ICS heat balance display.

Following the sudden decrease in indicated thermal power, the UO discovered the abnormal feedwater temperature point indication. Subsequently, reactor power was reduced. After feedwater temperature point 3-48A was deleted from processing,.reactor power was reduced to less than 3293 MWt.

G. Safe S stem Res onses:

None.

I' NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4-96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 1 DOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVlslON NUMBER NUMBER Browns Ferry Unit 3 05000296 4 OF 5 95 008 00 TEXT (Ifmore space is required, usa additional copies of NRC Form 366A) (17) lll. CAUSE OF THE EVENT This event was caused by the gradual failure of a feedwater temperature instrument. This provided inaccurate input to the nuclear heat balance resulting in an inaccurate indication of reactor power. As a result, the operator responded to the observed plant indication (i.e., decrease in power level) and proceeded to adjust recirculation flow.

IV. ANALYSIS OF THE EVENT Both the design basis Loss of Coolant Accident (LOCA) and the design basis containment accident analyses assume 102 percent of rated core thermal power. These analyses demonstrate that the emergency core cooling acceptance criteria of 10 CFR 50.46 would be met in the event of a design basis accident at 102 percent of rated core thermal power. Since the reactor in this event operated at a maximum 101.06 percent of rated core thermal power, it is within the bounds of the design basis LOCA and containment analyses. Therefore, this event did not adversely impact plant and public safety.

V. CORRECTION ACTIONS A. Immediate Corrective Actions:

The immediate corrective actions were to reduce reactor power and delete the feedwater temperature point from processing.

B. Corrective Actions to Prevent Recurrence:

TVA will perform troubleshooting on the resistance temperature detector (RTD) [DETj loop of the failed transmitter to identify the root cause of the failure. Although the operator responded as trained and in accordance with procedures, TVA has taken this event as an opportunity to improve future performance. As a result, TVA-'has clearly communicated to operations personnel management expectations for the review of reactor heat balance plant parameters during and after power adjustments. TVA also plans to revise the operator training program to review these expectations. Additionally, TVA will'brief operations shift personnel on the lessons learned from this event. Finally, TVA plans to review the recirculation system operating instruction and revise this instruction as necessary to ensure it includes appropriate guidance for, review of heat balance inputs.

Vl. ADDITIONALINFORIIATION A. Failed Com onents:

The biasing RTD fDETj (Rosemount Model 414M linear bridge) to feedwater temperature transmitter 3-TT-348A P7] .

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NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4.95)

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TEXT CONTINUATION FACILITYNAME 1 OOCKET LER NUMBER 6 PAGE 3 YEAR SEQUENTIAL REVISION NUMBER NUMBER Srowns Ferry Unit 3 05000296 5 OF 5 95 008 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

B. Previous LERS on Similar Events:

None.

VIL CQIVlMITMENTS None.

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

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