ML20044B013

LER 90-018-00:on 900612,reactor Protective Sys (RPS) Trip Units for Axial Power Distribution Determined to Be Inoperable.Caused by Procedural Deficiencies.Procedure Revised & RPS Surveillance Tests reviewed.W/900712 Ltr

ML20044B013
Person / Time
Site:	Fort Calhoun
Issue date:	07/12/1990
From:	Adams J, Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-018, LER-90-18, LIC-90-0560, LIC-90-560, NUDOCS 9007170185
Download: ML20044B013 (5)

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Omaha Public Power District

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' July 12,-1990, LIC-90-0560

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U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137

, Washington, DC-20555

Reference:

Docket No. 50-285 Gentlemen ,

Subject:

Licensee Event Report 90-18 for the Fort Calhoun Station W :Please find atta:hed Licensee Event Report 90-18 dated July 12, 1990.

=This. report is:being submitted pursuant to requirements of 10 CFR 50.73(a)(2)(v).-

If you'should have any questions, please contact me.

. Sincerely,.

Al - 5.&

W.'G. Gates:

Division Manager Nuclear Operations WGG/ tem Attachment-c: : R. D. Martin, NRC Regional Administrator A. Bournia, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector.

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On May M.1990 et 1240 hours0.0144 days <br />0.344 hours <br />0.00205 weeks <br />4.7182e-4 months <br />, the Channel "C" Reactor Protective System (RPS) trip units for Axial Power Distribution (APD) and Thermal Margin / Low Pressure '

l= (TM/LP)weredeterminedtobeinoperable. The inoperability resulted from procedural deficiencies for channel calibration in surveillance test RE-ST-NI-0001. Channel "C" was subsequently correctly calibrated and returned ,

L to operability. Further evaluation of this event on June 12, 1990, determined  ;

that the procedural deficiencies could have led to ino)erability of the APD and TM/LP trip functions for all four RPS channels, even t1ough only the "C" channel operability was.actually affected. This event is therefore reportable l pursuantto10CFR50.73(a)(2)(v). l Corrective actions include revising the procedure to eliminate the  ! i deficiencies, reviewing other RPS surveillance tests to ensure similar i deficiencies do not exist, and implementing any relevant improvements in the l l procedure upgrade program. l

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01 0 0l 2 JF Ql4 text ru . - e r ,asenn Periodic checks are performed at Fort Calhoun Station when the plant is in Mode 1 to ensure that the excore Nuclear Instrumentation (NI), which supplies inputs to the Reactor Protective System (RPS), is properly calibrated to provide accur. ate indication of power level and axial )ower shape. Each of four safety-channels of NI consists of upper and lower su) channels which are used to calculate an Axial Sha)e Index (ASI). This ASI value is used as an input to theAxialPowerDistri)ution(APD)tripunitandtheThermalMargin/ Low Pressure (TM/LP) calculator. ASI can also be determined by using incore instrumentation in conjunction with Combustion Engineering's CECOR computer code. This computer generated value has been shown to be more accurate than the excore value and more reflective of actual power distribution.

The periodic checks referenced above are accomplished by the monthly performance of surveillance test RE-ST-NI-0001. This procedure compares ASI as calculated from the excore NI's with the ASI determined by the CECOR code.

Also, the various subchannel powers of NI's are compared to an average to determine if any'particular subchannel deviates excessively. If the CECOR and excore detector calculated ASI values differ from each other by more than 0.01 ASI units or if any subchannel power differs from the average by more than 0.5%, calibration of the subchannels is required. The same criteria are used-as acceptance criteria once that calibration has been completed. Procedure RE-ST-NI-0001 was recently rewritten under a major upgra'de effort to improve the cuality of Fort Calhoun Station procedures, especially from a human factors stanc point.

On May 30. 1990, with the plant at approximately 30%-power, RE-ST-NI-0001 was performrd "or the first time since it had been rewritten. After Instrument and Control O&C) Technicians had completed the calibration portion of the procedure, the Shift Technical Advisor (STA) completed a post calibration test reccrd and discovered that the CECOR ASI and the excore ASI did not meet the acceptance criteria for the test, although they were considerably closer to being in agreement than they were prior to calibration. (Excore ASI is'a computer calculated value which uses all four channels of the RPS. The fact that the acceptance criteria is not met does not necessarily mean that operability of the RPS is in question. It simply indicates that the individual c.Sannels should be checked for problems.)

The following morning, the Reactor Engineer was reviewing the results of the test and noticed the out of tolerance values. He also noticed that "C" RPS channel indicated an ASI value which deviated from the other RPS channels and the CECOR value'by approximately 0.04 ASI units, a condition whic" was abnormal after calibration.

The Reactor Engineer contacted the RPS System Engineer who immediately began to search for a reason for the abnormality. The System Engineer determined within a few hours that the reason that "C" channel differed from the other channels was that a sign error had been introduced into the equations for calculating l the voltage of the upper subchannels. This error was introduced during the i procedure upgrade process and did not exist in the old revision of the l procedure which was used in previous performances of the test.

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0 j0 q3 or 0 l4 rixt in . - , u,,c ,-., nn Further investigation revealed that "A", "B" and "0" RPS channels were calibrated by an experienced Senior I&C Technician while'"C" channel was calibrated by e less experienced Technician who was not intimately familiar with the equations. The Senior Technician had performed the calibration many times using the old revision of the procedure and was so familiar with the proper equation.that he did not notice the error in the new revision. The less experienced Technician had relied on the new procedure and had therefore used the incorrect equation. The result was that "C" upper subchannel was calibrated incorrectly.

It should be noted that the summed value of the upper and lower subchannels is used for NI power indication. This signal is adjusted separately to agree with the secondary calorimetric and was not affected by the error in the-subchannel calibration.

The effect of the sign error in the calibration equation was to create a difference between "C" channel ASI and CECOR ASI of approximately 0.04 ASI units. When the error was discovered, the Reactor Engineer contacted the Supervisor - Reactor Performance Analysis to determine operability.of "C" channel. It was concluded that it would be conservative to declare the channel inoperable because of the effect on the APD and TM/LP trips, make a procedure change to correct the sign error, and recalibrate."C" upper subchannel. At 1240 on May 31, 1990, "C" RPS channel was logged as inoperable for the APD and TM/LP trip functions in accordance with Technical Specification 2.15.

Recalibration was performed the same afternoon and the LC0 was cleared at 1440 on May 31, 1990.

At no time during'this event was more than one channel of-the RPS inoperable.

-However, it is possible to envision a scenario in which all four channels of the RPS could have been adversely affected for the APD and TM/LP trips. That is, if the erroneous equation had been used to calibrate all four NI channels, all the resultant ASI values would have been erroneous. This postulated situation would have been detected by the post-calibration operability check, but not before all four channels were affected. Thus, the error in the procedure resulted in a condition reportable pursuant to 10CFR50.73(a)(2)(v).

This was determined on June 12, 1990 during further evaluation of the "C" channel inoperability.

If the scenario in which all four channels were inoperable had occurred, Technical Specifications would have required that the plant be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Inoperability of all four APD and TM/LP trip channels would put the plant outside its design basis as specified in the setpoint analysis.

A reactor trip could be delayed causing a potential violation of the Specified Acceptable Fuel Design Limit.

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010 0l4 0F 0 l4 rixT ~ . --c ,-., im There was little actual safety significance since only one channel was made inoperable, and the requirements of the Technical Specification Limiting Condition for Operation were followed as soon as that one channel was determined to be inoperable. The total time during which ope.wbility was affected was less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This is less than the allowabic time for a single channel to be inoperable. Also, the amount by which the ASt value was in error for "C" channel was only slightly outside the margin assumeo in the setpoint analysis.

A root cause investigation for this event was performed. The root cause was

-determined to be a-procedure deficiency which permitted all four channels of the RPS to be calibrated without verifying that each channel was acceptable

before proceeding to the next. A contributing cause was the sign error in the

-calibration equation introduced during +he procedure upgrade.

The following corrective actions will prevent this event from occurring in the future:

1. A procedure change to correct the sign error in the calibration.

equation was incorporated into RE-ST-NI-0001.

2. A review of applicable RPS surveillance tests has been performed to determine if_any other procedures would permit calibration of RPS channels without checking the operability of previously-calibrated.

channels. No other RPS surveillance. tests have been identified with this deficiency.

3. RE-ST-NI-0001 will be revised to require a check for operability for each channel before calibrating the next channel. .The revision will also include a requirement to notify the Reactor Engineer immediately when acceptance criteria are-not met. This change will be.

incorporated by July 31, 1990.

4. The procedure upgrade program, including the verification and validation process, will be reviewed and revised, if necessary, to include specific guidance aimed at ensuring the accuracy of equations transferred from previous documents or references. These actions will be completed by August 14, 1990.

This is the first event in which a part of the Reactor Protective System has become inoperable as a result of procedural deficiencies.

- NRC Perm Je6A (6491