ML20042F768

LER 90-002-01:on 900131,errors Identified in Calculation Used to Establish Calibr Tables for Steam Generator Water Level Transmitters.Errors in Original Calculation Not Identified.Calibr Procedures revised.W/900501 Ltr

ML20042F768
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/01/1990
From:	Ewing E, Millar D ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2CAN059005, 2CAN59005, LER-90-002, LER-90-2, NUDOCS 9005090369
Download: ML20042F768 (5)

Text

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-, Arkansas P:wer & Lignt Company

.= AP8L ==r Hussellvilie. AR 72801 Tel 501964 3100 l-I l

May 1, 1990 2CAN059005 U. S. Nuclear Regulatory Commission Document Control Desk '

Mail Station PI-137 Washington, D. C. 20555

$UBJECT: Arkansas Nuclear One --Unit 2 Docket No. 50-368 License No. NPF-6 Licensee Event Report No. 50-368/90-002-01 Gentlemen:

In accordance with 10CFR50.73(a)(2)(i)(B) and 10CFR50.73(a)(2)(vii), 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.

4 This report is being supplemented to provide a correction to the original issued report.

Very truly yours, E. . Ewing

' General Manager, Technical Support and Assessment ECE/DM/sgw attachment cc: Regional Administrator Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 l INPO Records Center 1500 Circle 75 Parkway Atlanta, GA 30339-3064 g 9005090369 900501 g i

PDR ADOCK 05000368 (

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l . Form 1062.01A t

NRC Fom 366 U.S. Nuclear Regulatory Commission (9 83) Approved OMB Ho. 3150-0104 Expires: 8/31/85 LICENSEE EVENT REPORT (L E R)

IACILITY hAME (1) Arkansas huclear One, Unit Two lDOCAET NUMBER (2) lFAGE (3) 10151010101 31 61 Bill 0Fl014 .

TITLE (4) Low Steam Generator Water Level Trip Values Less than Allowed by Technical Specifications I due to Errors in Calculations Used to Established the Calibration Data for the Steam Generator Level Transmitters EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACIL]T]E5 INVOLVED (8) l l5equentiall l Revision ,

l Month Day Year l Year i Number ' Number IMonth Day lYear Facility Names Docket Number (s) l l l l 0 5 0 0 0 01 1 31 1 9 01 91 Ol--I O I 01 2 l -i0l1 1015101Il910l 0 5 0 0 0 OP E RAING THIS REPORT 15 5UBMITTLD PUR5UAN1 10 THE REQUIREMENTS OF 10 Cf R b:

M00E C9) 1 (Check one or more of the following) (II)

POWER "- 20.402(b) - 50.73(a)(2)(iv) - 73.71(b)

LEVEL 20.405(a)(1)(1) l l 50.36(c)(1) l[l20.405(c) 50.73(a)(2)(v) 73.71(c)

(10) 1 11010 - 20.405(a)(1)(ii) l 17 50.73(a)(2)(vii)  ?"~l Other (Specify in i 20.40$(a)(1)(iii) l]I l50.36(c)(2) 50.73(a)(2)(1) l~~ 50.73(a)(2)(viii)(A)l- Abstract below and in Text, NRC Form l 20.405(a)(1)(iv) l_I 50.73(a (2)(11)

I l 20.405(a)(1)(v) I l 50.73(a (2)(iii) l-l 50.73(a)(2)(viii)(B)l l 50.73(a)(2)(x) l 366A)

LICEN5EE CONTACT FOR THIS LER (12) home l Telephone Number IArea l Dana Millar, Nuclear Safety and Licensing Specialist (Code l 1510111916l41-13111010 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN T"15 REPORT (13) l l l Reportable. ~ l l l l Reportable l Cause System Component Manufacturer to NPR05 Causelsysteml Component ]Manufacturerl to NPR05 l 1 I I I I I I l l t i I I I I I I I I I I I I i I I I l l 1 l l I ' i f f I I I I l l  !

5UPPLEMENT REPORT EkPECTtD (14) EXPECTED Month Day Year SUBMISSION l l~l Yes (If yes, complete Expected Submission Date) III No o DATE (15) 1 I '

I I AB5 TRACT (Limit to 1400 spaces, i.e. , approximately fif teen single-space typewritten lines) (16)

On January 31, 1990, several errors were identified in tne calculation used to establish the calibration tables for the Steam Generator (SG) 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 l an assurance that the reactor would trip when actual SG water level was greater than the minimum allowable value of Technical Specifications. After a thorough evaluation of the calculation was completed, two additional errors were identift d. 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 l 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 nct been identified. The calculation errors were corrected and the calibration procedures for the SG water level transmitters revised.  !

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Form 1062.018

NRC Form 366A U.S. Nuclear Regulatory Commission (9-83) Approved OMB No. 3150-0104 Expires

8/31/85 LICENSEE Et'ENT REPORT (LER) TEXT CONTINUAT]ON FACILITY NAME (1) l DOCKET NUMBER (2) l LER NUMBER (6) l FAGE (3)

Arkansas Nuclear One, Unit Two I l l 15equentiall lRevisioni l l_ Year Number Number l 1015l010101 31 61 81 9l 0 --

01 01 2 -

Di 'Il01210Fl0J4, TEM (If more space is required, use additional hRC Form 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 (HP]) (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 identified that the calculation associated with the calibration tables used for the Steam Generator (SG) (SG) 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 transmitters, one from each SG. A reactor trip and an Emergency Feedwater Actuation Signal (EFAS) are generated by the PPS 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 *he 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 determined 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 LG channels were declared inoperable and Technical Specification 3.0.3 entered at 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. As 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, a 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 tu 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. Ey 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 l- 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 level reference leg was in a Saturated condition, the compressed water tables should have been used. The other error was an incorrect interpolaticn included in the calculation. The low level trip setpoint in the PPS is set at 23.25 percent by procedure.

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 ailowed by Technical Specifications. i The high SG water level trip setpoints were evaluated in regard to the calculational errors and 1

found to be conservative. Therefore, no adjustments were made to the high SG water level trip setpoints.

l 1 C. Safety Significance The SG low water level reactor trip provides protection against a loss of feedwater flow incident l

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 wate" inventory in the SG at the time of the reactor trip generation to provide margin before emergency feedwater is required. The specified setpoint also functions to initiate an EFAS to automatically provide emergency feedwater flow to the SG.

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' .; ,3-Form 1062.01B NAC Fcrm 366A U.S. Nuclear Regulatory Commission (9 83) Approved DNB No. 3150-0104 Expires: 8/31/85 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY hAME (1) IDDCkET NUMBER (2) l LER NJMBER (6) l PAGE (3)

I i l l5equentiall l Revision l Arkansas Nuclear One, Unit Two l l Year Number Number ]

10l$1010101 3I 61 81 91 0 -

01 Ol 2 --

Of II01310F1014 TEXT (If more space is required, use additional NRL Form 3fi6A's) (17)

After a thorough review and evaluation of the calculation used to establish the calibration tables for the SG water level transritters was completed, it was determined that the total effect of the errors resulted in an indicated water level of 22.33 percent, a .92 percent error in SG water level I indication. The Technical Specification allowable value for low SG 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 SG water level reactor trip were not challenged. No safety concerns existed.

D. Root Caust -

The root cause of this event was personnel error. The individual who performed 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 span shift would effect only the upper end of the.me#surement band. The level transmitter is calibrated at an ambient condition.

When the transmitter is placed in service at en operating static pressure of approximately 900 psia, a span shift in the transmitter output actually becurs at both ends of the measurement band.

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

E. Basis for Reportability The feur Low SG Water Level PPS channels were declared inoperable and Technical Specification 3.0.3 was entered. This condition 16.. therefore, reportable pursuant to 10CFR50.73(a)(2)(1)(B),

operation prohibited by Technical $pecifications. This condition is also reportable pursuant to 100FR50.73(a)(2)(vii), 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 Oaerations Center via the Emergency Notification $ystem  ;

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 The calculation was corrected considering the effect of the static pressure of the fluid in the 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 SG level transmitters have been revised to correct the calculational errors which were identified. The SG level 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 SG level trip setpoint bistables will be returned to approximately 23 percent as allowed by Technical Specification.

A review of other Rosemount "Q" differential pressure loops was completed on January 31, 1990.

The calibration procedures for the "Q" Rosemount differential pressure transmitters which would be exposed to pressures greater than 200 psig were reviewed to determine if static pressure effects had been considered. Each of the procedures reviewed appeared to consider static pressure '

effects with the exception of four, 2PDT-4602, 2 POT 4603 ( AND-2 differential pressure transmitters which are not used during power operation and provide no safety related indications or interlocks) t and PDT-2400 and PDT-2401 (ANO, Unit One indications and alarm for the discharge flow for the Reactor Building Spray pumps). Each of the transmitters is classified as "Q" only to insure RCS pressure boundary integrity. The ANO-2 transmitters do not provide safety related indications or ,

interlocks. The AND-1 transmitters provide information to indicate the operation of an individual safety system. Since the Reactor Building Spray system is redundant (single failure proof), gross

, performance diagnosticf. are all that is recessary tc maintain operator awareness of system status.

l Therefore, the failure to incorporate static pressure and span correction factors i~to n the calibration of these loops does not affect the system capability to provide its function and/or maintain design basis conditions.

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Foo 1062.01B -'

- NRC Form 366A U.S. Nuclear Regulatory Commission (6-89) Approved OMB No. 3150 0104 LICENSEE EVENT REPORT (LER) TEXT CONTINVATION Ni (1) lDOCAET NUMBER (2) l LER NUMBER (61 l PAGL (3) l l l l5equential .

Revisioni Arkansas Nucitrar One. Unit Two l l Year Number Number l 10l51010101 31 61 81 91 0 --

01 01 2 --

01 Il01410Fl014 TEXT (If more space it, required, use additional NRC Form 366A's) (17)

Four calibration procedures were verified for the proper application of the static pressure effecto. Of these, one was identified to have used an incorrect correction factor for static pressure. AND concluded that the discovered error was not significant.

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 established by Technical Specifications.

6. Additional Information The SG' level transmitters are model 11530A manufactured by Rosemount [R369).

There have been no previously reported similar events.

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

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