ML20011F674

LER 90-002-00:on 900131,errors Identified in Calculation Used to Establish Calibr Tables for Steam Generator Water Level Transmitters.Caused by Incorrect Static Pressure Assumption.Trip Setpoint Bistable increased.W/900302 Ltr

ML20011F674
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/02/1990
From:	Ewing E, Millar D ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2CAN039003, 2CAN39003, LER-90-002, LER-90-2, NUDOCS 9003070125
Download: ML20011F674 (4)

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f Arkansas Power & Ught Company 42 We st aMol j Lee Rock AR 7??D3 Tel 501377 4:00  !

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March 2, 1990 1

2CAN039003 U. S. Nuclear Regulatory Commission Doctment Control Desk Mail Station PI-137

Washington, D. C. 20555

SUBJECT:

Arkansas Nuclear One - Unit 2

_ Docket No. 50-368 License No. NPF-6 Licensee Event Report No. 50-368/90-002-00 Gentlemen

In accordance with 10CFR50.73(a)(2)(1)(B), attached is the subject report concerning low Steam Generator (SG) water level trip values being less than allowed by Technical Specifications due to errors in calculations used to '

establish the calibration data for the SG 1evel transmitters.

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Very truly.yours,

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E. C. Ewing General Manager, Technical Support and Assessment ECE/DM/sgw attachment cc: Regional Administrator Region IV q' U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INP0 Records Center 1500 Circle 75 Parkway Atlanta, GA 30339-3064

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.9003070125 900302 h k PDR ADOCK 05000368 R. PDC #(?

Fore 1062.01A NRC Fom 366 U.$. Nuclear Regulatory Commission (9 83) Approved OMB No. 3150-0104 Expires: 8/31/85 LICENSEE EVENT REPORT (L E R)

EACILITY NAME (1) Arkansas Nuclear One, Unit Two l DOCKET NUMBER (2) lPAGE (3) 10151010101 31 61 81110Fl013 TITLE (4) Low 5 team Generator Water Level Trip Values Less than Allowed by Technical Specifications due to Errors in Calculations Used to Established the Calibration Data for the $ team Generator Level Transmitters '

_ EVEC DATE (5) LER NUBBER (6) REPORT DATE (7) OTHER FAtlLITIES INVOLVED (B)

Day l l il - 5eouentiali (Revision i i i Month Year Year Number Number Month Day lYear Facility Names Docket Number (s1 1 0 5 0 0 0

_011131l '9'0 91 Ol** O I 01 2 *-

'I010 Of 3 of 21 91 0 1 0 5 0 0 0 OPERATING THM5 REPORT 15 50BMITTI;D PUR5UANT O THE REQUIREMENTS OF 10 CFR 5:

MODE (9) I1 1 (Check one or more of the followino) (11)

. PDWERI i 20.402(b) l~ l 20.405(c) l 50.73(a)(2)(iv)

LEVEll 1 l l 50.36(c)(1) l ~ll 50.73(a)(2)(v) i[I73.71(b) l ' 73.71(c) 110)111010[I20.405(a)(1)(1) 1 20.405(a)(1)(ii) ll 50.36(c)(2) l 20.405(a)(1)(111) 3l50.73(a)(2)(1) 1 13150.73(a)(2)(vii)l l

50.73(a)(2)(viii)(A)I_

Abstract below andl Other (Specif l~ll 20.405(a)(1)(iv) _l 50.73(a)(2)(ti) l_l 50.73(a)(2)(viii)(B)] in Text. NRC Fom i I 20.405(a)(1)(v) l 50.73:a)(2)(111) l _ I 50.73:o)(2)(x) l 366A)

LICENSEE CONTACT FOR THI5 LER (1J)

Name l Telophone Number lares ,

Dana Millar Nuclear Safety and Licensin0 Specialist l Code l

'5l0111916141 13111010 ,

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THlb REPCRT (L3) l 1 l Heportablel l l , lReportableu -

Cause Systee Component IManufacturer, to NPROS l Cause Systee Component IManufacturer to NPR05 1 ,

I l l l l I l l I l l I I I l 1 I l l I i l I l l l l I I I I I I I 1 i i l i l l I I SUPPLEMENT REPORT EXPECTED (14) EAPECTED Month Day Year l $UBM15$10N l l~l Yes (if yes complete Expected Subsission Date) til No 1 0 ATE (15) l I 1 I I i ABSTRACl (Limit to 1400 spaces, i.e., approximately fif teen single-space typewritten lines) (16)

On January 31,3990, several errors were identified in the calculation used to establish the calibration tables for the Steam Generator ($G) water level transmitters. A preliminary evaluation identified an incorrect assumption for the effect of static pressure on the span of the level transmitters. This resulted in the actual SG water level being less than the minimum allowable value required by Technical Specifications for a low SG water level reactor trip. To compensate for this error, the reactor trip setpoint bistable for low SG water level in the Plant Protective System was increased. This provided

• an assurance that the reactor would trip when actual SG water level was greater than the minimum allowable value of Technical $pecifications. After a thorough evaluation of the calculation was completed, two additional errors were identified. The result of the combined errors was that the actual SG water level was .92 percent less than the indicated water level, above the minimum allowable value of Technical specifications. The safety functions provided by the low SG water level reactor trip were therefore not challenged and no safety concerns existed. The root cause of this condition was personnel error. The errors in the original calculation had not been identified. The calculation ,

errors were corrected and the calibration procedures for the SG water level transmitters revised, l 4

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'. s WRC Fcro M6A Fom 1062.01B U.S. Nuclear Regulatory Commission (9-83) Approved DMB No. 3150-0104 LICEN5EE EVENT REPORT (LER) TEXT CONTINUATION FACI'ITY NAME (1) IDDCLET NJEER (2) l l.ER Nd4BER (6)

Arkansas Nuclear One, Unit Two l PAGE (3)

I i j inequentiell 1 Revision l l l Year Number Number l 1015101010[ 31 61 81 91 0 --

0 I 01 2 --

010 101210Fl013 ItKT (If more space is required, use additional NRC f om 366A's) (17)

A. Plant Status At the time of discovery of this condition, Arkansas Nuclear One Unit Two (AND-2) was at 100 percent of rated thermal power operating in Mode 1 (Power Operation). Reactor Coolant System (RCS) [AB) pressure was approximately 2250 psia and RCS temperature about 580 degrees Fahrenheit.

B. Event Description As a result of identifying an assumption error associated with static pressure shift in a calculation used to establish the calibration tables for a High Pressure Injection (HPI) [BJ) flow transmitter on AND-1, a review of the calculations used to establish calibration tables for other safety related transmitters was performed. It was ioentified that the calculation associated with the calibration tables used for the Steam Generator (SG) [$G) water level transmitters were in error.

The SG water level transmitters provide input into the Plant Protective System (PPS) [JC). There are four channels of PPS, each receiving input from two SG water level transeitters, one from each

$G. A reactor trip and an Emergency Feedwater Actuation Signal ((FAS) are generated by the PP5 when SG water level reaches a preselected trip bistable setpoint of 23 percent. A reactor trip is also generated when a SG water level of 93.7 percent is reached.

At approximately 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br /> on January 31, 1990, a preliminary evaluation of the calculation was completed. it was concluded that a compensation factor for static pressure effects of the fluid in the SGs at normal plant operating conditions had not been correctly included in the calculation.

This resulted in incorrectly calibrating the transmitters and an inaccurate SG water level measurement.

ANO detemined that with an indicated level of 23.25 percent, actual SG water level could be 21.8 percent which is less than the allowable value stated in Technical Specifications. The four PPS low level $G channels were declared inoperable and Technical Specification 3.0.3 entered et 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br />. Technical Specification 3.0.3 allows one hour to satisfy the requirements of the actions associated with the Limiting Condition for Operation. At 2006 hours0.0232 days <br />0.557 hours <br />0.00332 weeks <br />7.63283e-4 months <br />, a power reduction was commenced to comply with Technical Specifications. As required by Station Emergency Plan procedures.

4 Notification of Unusual Event (NUE) was declared.

In order to compensate for the calculation errors a decision was made to conservatively increase the PPS trip setpoints to approximately 25 percent by adjusting the low SG 1evel trip bistable in each PPS channel. This provided assurance that a reactor trip would be generated when actual water level in the SGs was greater than or equal to 23 percent. By approximately 1129 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.295845e-4 months <br />, the trip setpoint in each PPS channel had been increased to 25 percent to ensure compliance with Technical Specifications. The Technical Specification action statements were exited and the NUE ,

i terminated.

In the afternoon of January 31, 1990, a comprehensive review of the calculation used to establish .

the calibration tables for the SG water level transmitters was completed. Two additional input assumptions, other than the effects of static pressure, were identified. The original calculation had assumed the water in the SG 1evel reference leg was in a saturated condition, the compressed water tables should have been used. The other error was an incorrect interpolation included in the calculation. The hw level trip setpoint in the PPS is set at 23.25 percent by .ocedure.

Considering the effects of the three errors identified in the calculation, the total errors were approximately .92 percent. Therefore, when indicated water level was 23.25 percent, actual water level could be approximately 22.33 percent, which is above the minimum value allowed by Technical Specifications.

The high $G water level trip setpoints were evaluated in regard to the calculational errors and found to be conservative. Therefore, no adjustments were made to the high SG water level trip setpoints.

C. Safety Significance The SG Iow water level reactor trip provides protection against a loss of feedwater flow incident and assures that the design pressure of the RCS will not be exceeded due to loss of SG heat sink.

The specified setpoint ensures sufficient water inventory in the SG at the time of the reactor trip generation to provide margin before emergency feedwater is required. The specified set' point also functions to initiate an EFAS to automatically provide emergency feedwater flow to the SG.

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• Fore 1062.018 NRC Fem 366A U.S. Nuclear Regulatory Commission (D 83) Approved OMB No. 3150 0104 LICEN$[E EVENT REPORT (LER) TEXT CONTINUAT101;  !

IACILITY M4ML (1) (DOCKET NLDSER (2) l LER NilMBER {6) l PAGE (3)

Arkansas Nuclear One, Unit Two l l 15eguensielu l Revision l '

l l Year Nuuter l Number l ,

10151010101 31 61 81 91 0 --

0 1 01 2 --

010 101310F1013 ,

ILU (If more space is required, use additional NRC Form 366A's) (17) +

Afterathoroughreviewandevaluationofthecalchletionusedtoestablishthecalibrationtables for the 50 water level transmitters was completed, it was determined that the total effect of the errors resulted in an indioted water level of 22.33 percent, a .92 percent error in 5G water level indication. The Technical Specification allowable value for low $G water level is 22.111 percent.

Therefore, the actual water level was not less than the Technical specification allowable value and ,

the safety functions provided by the low $G water level reactor trip were not challenged. No safety '

concerns existed. '

O. Root Cause The root cause of this event was personnel error. The individual who perfomed the original calculation did not correctly account for the effects of static pressure on the transmitter output. It was assumed that the effect of static pressure on the tren shift would effect only the  ;

upper end of the measurement band. The level transmitter is calibrated at an ambient condition. i When the transmitter is placed in service at an operating static pressure of approximately 900 psia, a span shift in the transmitter output actually occurs at both ends of P.he measurement band.

With the transmitter calibrated to account for a span shift only at the upper measurement band, the actual SG water level measurement was inaccurate when the transmitter was placed in service at normal SG operating conditions.

E. Basis for Reportability The four Low SG Water Level PPS channels were declared inoperable and Technical Specification 3.0.3 was entered. This condition is, therefore, reportable pursuert to 10CFR50.73(a)(2)(1)(B),

operation prohibited by Technical Specifications. This condition is also reportable pursuant to 10CFR50.73(a)(2)(v11), where a single cause resulted in four independent channels to become inoperable in a multiple channel system designed to shut down the reactor and maintain it in a safe shutdown condition.

This event was reported to the NRC Operations Center via the Emergency Notification System pursuant to 10CFR50.72(a)(1)(1) and 10CFR50.72(b)(1)(1)(A) at approximately 1021 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.884905e-4 months <br /> at January 31, 1990.

F. Corrective Actions l The calculation was corrected considering the effect of the static pressure of the fluid in the l SGs and the other identified errors. The result after a thorough review was that the actual water level in the SGs was 22.33 percent when the indicated water level and the PPS bistable values were at 23.25 percent.

The calibration tables in the calibration procedures for the $G 1evel transmitters have been revised to correct the calculational errors which were identified. The SG 1evel transmitters will be recalibrated using the revised procedures during a forced outage or refueling outage, whichever is more appropriate. After the level transmitters are recalibrated the PPS low $G 1evel trip cetpoint bistables will be returned to approximately 23 percent as allowed by Technical Specification.

The calculations associated with the calibration of other similar transmitters have been reviewed.

No errors were identified in other calculations.

The SG 1evel transmitters have been calibrated using incorrect calculational assumptions since 1979 when the transmitters were originally installed in the plant. A cursory review of the results of previous calibrations was performed. Assuming corrections for the calculational errors, only two times were identified when the transmitters were calibrated and left at values less than the allowable value estabitshed by Technical Specifications.

G. Additional Information The $G 1cvel transmitters are model 1153DA manufactured by Rosemount (R369).

There have been no previously reported similar events.

Energy Industry Identification System (E!!S) codes are identified in the text as (XX).

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