ML18005B000

LER 89-011-00:on 890607,inaccurate Reading from Resistance Temp Detector (RTD) Used to Measure Thermocouple Ref Junction Box Temp Reported.Caused by Corrosion of RTD Terminal Contacts.Connection repaired.W/890719 Ltr

ML18005B000
Person / Time
Site:	Harris
Issue date:	07/19/1989
From:	Lew G, Richey R CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HO-890076-(O), LER-89-011-01, LER-89-11-1, NUDOCS 8907250137
Download: ML18005B000 (7)

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8907250137 DOC.DATE: 89/07/19 NOTARIZED: NO DOCKET N FACIL:50-400 Shearon Harris Nucl'e'ar. Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION LEW,G.'T Carolina Power.& Light Co.

RICHEY,R.B. . Carolina'ower, &,Light. Co..

RECIP.NAME -."'ECIPIENT

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AFFILIATION'. ~ ,

SUBJECT:

LER 89-011-00:on 890619,reactor power slightly exceeded 1004 on several occassions due to erroneous FW temp reading.

W/8 ltr.

S DISTRIEUTION-CODE: IE22T COPIES RECEIVED: LTR:, ( ENCL . (',- SIR+: -.-

TITLE:,. 50.73/50.9 Licensee*Event'Report'LER) Indident Rpt, etc-.'-

NOTES:Application for permit renewal filed. 05000400 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/N'AME LTTR ENCL PD2-1 LA 1 1 PD2-1 PD 1 1 D BECKER,D 1 1 S

INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 ACRS WYLIE 1 1 'EOD/DOA 1 1 AEOD/DS P/TPAB :1 ~ 1 . AEOD/ROAB/DS P 2 ~ ~

2 DEDRO 1 1 'RM/DCTS/DAB 1 1 NRR/DEST/ADE SH 1 1 NRR/DEST/ADS /E 1 0 NRR/DEST/CEB SH 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1

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EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY,A 1 1 L ST LOBBY WARD 1 1 LPDR 1 1 D NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHY,G.A 1 1 S D

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FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 43 ENCL 42

Carolina Power & Light Company HARRIS NUCLEAR PROJECT P.O. Box 165 New Hill, NC 27562 JUL I e ]989 File Number.'SHF/10-13510C Letter Number: HO-890076 (0)

U.S. Nuclear Regulatory Commission ATTN: NRC Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 89-011-00 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is, submitted. This report fulfills the requirement for a written report within thirty (30) days of a reportable occurr'ence and is in accordance with the format set forth in NUREG-1022, September 1983.

Very truly yours, R. B. Richey, Man ger Harris Nuclear Pr ject RBR:sbg Enclosure cc: Mr. R. A. Becker (NRR)

Mr. W. H.-Bradford (NRC SHNPP)

Mr. S. D. Ebneter (NRC - RII)

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P> ~DOC>: OSOOOqOP PDC MEM/LER-89-011/1/OS1

NRC Fotm 356 US. NUCLEAR REOULATORY COMMISSION (943 I APPROVED OMS NO, 3150010I EXPIRES; 6/3(/SS LICENSEE EVENT REPORT ILERI FACILITYNAME (I) DOCKS'7 NUMSE 5 (2)

SHEARON, HARRIS NUCLEAR POWER PLANT UNIT 1 050004001OF04 7'7'E"'REACTOR POWER SLIGHTLY EXCEEDED lOOX ON SEVERAL OCCASIONS DUE TO AN ERRONEOUS PEEDWATER TEMPERATURE READING CAUSED BY A CORRODED CONNECTION.

EVENT DATE (5) LER NVMSER (6) REPORT DATE (7) O'THER FACILITIES INVOLVED (5)

MONTH DAY YEAR YEAR SSOUSNTrAL RSVO~ MON'TH DAY YEAR FACILITYNAMES DOCKET NUMSER(SI HUMSER rrUMSSR 0 5 0 0 0 0 5 0 0 0 OPERATINO THIS REPORT IS SVSMITTED PURSUANT TO THE REOUIREMENTS OF 10 CFR (Ir (Crena onc or mnc of tne followinFI l(ll MODE (5) 20.$ 02(S) 20AOS(cl 50.73(e) (2)(rrl 73.71(SI POWER 20AOS(eHIHO 60.35(cl(ll 50.73(el(2)(v) 73.71(cl LEYEL 1 0 0 20 ASS( ~ ) (I I (SI 50.35(cl(2) 60.73(e)(2) (rSI OTHER ISpcclfy In Aoroect Snow end ln Tert. HIIC Form 20AOS(el((l(ISI 50.73( ~ ) (2)(rl 50.73(el(2)(r51)(A) 3FFAI 20AOS(e) (I ) (Irl 50.73(e I (2)(S) 50.73(e) (2l (r(5)(5) 20AOS ( ~ ) (I ) (v) 50.73(e) (2) (IU) 50.73(e) (2l(c)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMSER AREA CODE G. T. LEW PROJECT ENGINEER REGULATORY COMPLIANCE 919 362 203 COMPLETE ONE LINE FOR EACH COMPONENT FAILVRE OESCRISED IN THIS REPORT l(3)

CAUSE SYSTFM COMPONENT MANUFAC. EPORTASLE COMPONENT MANUFA(r EPORTASLE TVRER TO NPRDS TVRER TO NPRDS B ID CON X99 9 N gklh+I(II@~W

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MSr>>?x'Pv'A",'UPPLEMENTAL REPORT EXPECTED (ICI MONTH DAY YEAR EXPECTED SVSMISSION DATE (16)

YES Ilfyn, corr Pic re EXPECTED SIISMISSIOH DA TEI NO AbsTRAcT ILimit to ic(ro roccn, I c., coororlmetcry liltccn einpic rpecc typewritten linnl (15)

On June 7, 1989 an inaccurate reading from a Resistance Temperature Detector (RTD) used to measure a thermocouple reference junction box temperature was.

reported to the Plant Reactor engineers. A corroded connection caused the temperature detectors associated with an instrumentation multiplexing cabinet to indicate a higher than actual temperature. The condition affected the accuracy of a feedwater temperature data poi'nt used in the determination of the daily reactor power calculation (calorimetric).

The connection was repaired and an analysis of the impact on plant performance initiated on June 9. On June 19, it was concluded that reactor power exceeded 100K thermal power by a small amount on several 'ccasions. The review.

indicated that 100X thermal power was exceeded on 20 days between May 1 and June 6, 1989. The highest indicated power reache 1 was 100.32K on May 257 1989. ,On the remaining days, power levels ranged from 100.06X to 100.29X.

This event is reported as a violation of Section 2.C. (1) of the 'operating license.

NRC Form 3$ 6 TS 53)

NRC fw>> DDDA VS. NUCLEAR REGULATORY COMMISSION (882)

LICENSEE EVENT REPORT ILER).TEXT CONTINUATION APPROVEO OM8 NO. 2(9)MIOI EXPIRES: SID(IDS fACILITYNAME (ll COCKET NUMSER 12) LER NUMSER (8) PAGE (2)

YEAR '.'.P'<

.8 DIOVENTIAI F'~. A(VISION NVM ER .,YF NVM EA SHEARON HARRIS NUCLEAR POWER PLANT-UNIT l.

TEXT Cf n>>fP N>>Ce H Ifqvf>>E I>>f DESCRIPTION:

~ lf)(C fcffn IDEA'fl(12) o s o o o 4O 089 011 00 2 OF 0 On May 23, 1989 an unexplained high "B"- train condensate temperature was observed. Troubleshooting of the problem led to the discovery (on May 25),

and repair (on June 6) of a corroded connection on a Resistance Temperature Detector (RTD) used to measure thermocouple reference box temperature.

Separately another engineer, reviewing feedwater temperature data, noted high temperature readings for data point TFW-2000B (feedwater temperature to steam Igenerator "B"). The engineer was not aware of the condensate temperature issue. .The engineer checked on the status of the feedwater temperature problem the following day and observed that the temperature had returned to normal. The engineer continued to investigate the problem and discovered that the RTD terminal connection had been repaired. At this time, the engineer recognized that the RTD repair had resolved both the feedwater and condensate temperature issues. The engineer recognized that this feedwater temperature was used in the plant heat balance and immediately informed the reactor Engineering Section and Plant Management of a possible discrepancy in feedwater temperature data.

Reactor engineering was notified of the feedwater temperature data problem on June 7, 1989 and began an investigation of the problem. By June 9, it was determined that the corroded connection would cause an erroneous high temperature reading. Data was collected for actual indicated reactor power and revised calorimetric results were -prepared. On June 14, management was notified of the possibility. that operation above 100X thermal power had taken place. By June 19, 1989, analysis confirmed that the reactor was operated slightly above 100% thermal power.

The review of temperature data indicated that the temperature discrepancy began about May 1, and existed through repair of the corroded connection on June 6, 1989. The condition resulted in a maximum temperature error of 8' for one of the three feedwater lines. Normal feedwater temperature for 100X power, is on the order of 430'. Modifying the daily calorimetric calculations by substituting a lower value for TFW-2000B resulted in an error between 0.05 and 0.46X. For calorimetric purposes, feedwater temperatures are used to calculate feedwater enthalpy. Having an indicated temperature higher than the'ctual temperature translates into an error in feedwater enthalpy and a calculated power which is less than the real power. Adding the calculated error to the indicated power as determined by recorded highest indicated power level showed that the worst case true power condition, when averaged over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, slightly exceeded 100X for 20 of the 37 days. The maximum 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average was 100.32% on May 25, 1989. For minor power fluctuations, the plant policy is to use a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average power to determine compliance with the license limit.

NIIC f ORM DADA *V.S GPO.ID80&02A 088/ASS (882)

~ ~

NRC FSIM ASSR US. NUCLEAR REOVLATORY COMMISSION IS421 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION- APPROVED OMS NO. 3IEOWI04 EXPIRES: 5/Sl/IEI FACILITY NAME III DOCKET NVMSER 12) LER NVMSER ISI PACE IS)

IssII SEQUENTIAL j.y'i REVISION SHEARON HARRIS NUCLEAR POWER PLANT- NVM IN /4>> NVM SN UNIT 1 o 5 o'o o 40089 0 11 0 30FO 4 TEXT /F INN>> <<>>C>> /S />>SNSNE NS>> <<RR/Or<<//YIIC F>>N>> 2/ISAS/ OTI CAUSE%

The high temperature readings were caused by corrosion of RTD terminal contacts. The high resistance connection caused a high reference junction temperature reading. This caused the associated feedwater temperature to indicate a higher than actual temperature. The data point was not immediately flagged as an error because the difference between similar indicators was within 8'.

ANALYSIS:

The power range heat balance or calorimetric calculation is performed daily when operating above 15Z power by procedure OST-1004. Feedwater temperature, feedwater pressure and steam generator pressure are measured for each steam generator. The results are used to determine the enthalpy of the feedwater supplied to each steam generator and the enthalpy of the steam leaving each steam generator. The enthalpy rise across the steam generator is multiplied by the measured feedwater flow to determine the power produced by the steam generator. The values obtained for each steam generator, a value for reactor coolant pump heat and a value for thermal losses are summed to obtain net reactor output. The result is compared to -the licensed limit of 2775 MW to establish a percent of calculated power. This calculated result is compared to indicated power and used as a basis for adjustment of power range instrumentation.

Table 4.4.6-3 of the FSAR shows that feedwater temperature measurements are expected to be accurate to within 2'. In combination with the uncertainties of other measure plant parameters, the power (or flow) calculation is accurate to within 1.75X. For conservatism and compliance with 10CFR Appendix K, the accident analysis assumes that initial" power levels for transients beginning at 100X rated thermal power are analyzed assuming a minimum of 102X rated thermal power.

For this event, any errors introduced by the erroneous feedwater temperature measurement may be compensated for by opposing errors or by the accuracy of other measured parameters. These compensating effects were not quantified.

The slight temperature error has an insignificant effect on the consequences of any design basis accident.

This event is reported as a violation of Section 2.C.(1) of the operating license since the plant was operated at over 100X rated power (2775 MW).

There have been no other events of a similar nature for this plant.

NRC FORM SSSA >> U S OPO.10554.524 555/455 10 42l

N RC PerM 888A US. NUCLEAR REOUL*TORY COMMISSION (ddd>

LICENSEE EVENT REPORT HLER) TEXT CONTINUATION APPROVED OMS NO. 3(50&10(

EXPIRES: did>IN PACILITY NAME ((> DOCKET NUMSER (2l LER NUMBER (8> ~ AOE D>>

yd*N )jX.: e(QVENTIAI @(( 1(V<<ION NUM ee .i. NVM ee SHEARON HARRIS NUCLEAR POWER PLANT-UNIT 1 o so o o 4OO 89 0 11 0 4or0 4 TEXT N eeee <<Nce <<Nqeeed, eee eAA(<<ee(NRC Acne 885AEU(17>

-'ORRECTIVE ACTION:

1. The corroded connection was corrected as soon as the condition was noted. Other connections were inspected for corrosion and no problems were found.

2. The calorimetric procedure will be revised to define expected tolerances for data where multiple instruments are measuring the same pa'rameter.

ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES:

Com onent of S stem EIIS Code Computer System ID Condensate System SD Feedwater System SJ N<<c pollM cede e U.S GPO.(080-0.524 538/edd (Ddd>