05000254/LER-1994-004, Submits Reproduction of Original Text of LER 94-004 & Suppl Info

Submits Reproduction of Original Text of LER 94-004 & Suppl Info

ML20095K427
Person / Time
Site:	Quad Cities
Issue date:	12/05/1995
From:	Pearce L COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20095K432	List: 05000254/LER-1994-004, Submits Reproduction of Original Text of LER 94-004 & Suppl Info
References
LWP-95-111, NUDOCS 9512290009
Download: ML20095K427 (5)

LER-2094-004, Submits Reproduction of Original Text of LER 94-004 & Suppl Info

Event date:
Report date:
2542094004R00 - NRC Website

text

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Quad Ofics Generating MJtuni 22'10 JtKHh Asenue North

. Otrdos a. lL 61242 9'40 Tel 4094422 41 LWP 95-ll!

December 5,1995 U. S. Nuclear Regulatory Commission

. Document Control Desk Washington, D.C. 20555

Reference:

Quad Cities Nuclear Station Docket Number 50-254, DPR-29, Unit One

Subject:

Licensee Event Report (LER) 254/94-004 Supplemental Information.

i

, As stated in LER 254/94-004, supplemental information is being provided and is

enclosed as Attachment 1. This information constitutes revision 01 to the original LER documentation.

1 Attachment 2 is a reproduction of the original text of LER 254/94-004.

This report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73(a)(2)(ii)(B), "Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded or that resulted in the nuclear plant being in a condition that was outside the design basis of the plant." l If there are any questions or comments concerning this letter, please refer them to Nick Chrissotimos, Regulatory Assurance Administrator at 309-654-2241, ext. 3100.  :

1 Respectfully, l

CDMMO TH EDISON COMPANY QUAD CITIIS NUCLEAR STATION i [M Q L.W. Pearce Station Manager Attachment 1- LER Supplemental Information Attachment 2- LER 254/94-004 (Copy) cc: J. Schrage C. Miller INPO Records Center NRC Region III 9512290009 951205 J PDR ADOCK 05000254 ,3 i

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- () mad Osies Nmeiser JWrer 3sasies Syssau KV- W Desaudent To L. W. Pearce Prepared By R. E. Dvorak N Frems D. Craddick Reviewed By: 5. Davis Dates 12/18/95 l

ATTACHMENT 1 l This supplemental LER report is being submitted to inform the Nuclear Regulatory Commission of Quad Cities Stations' i actions to address GE SIL 452 Supplemental 1 Revision 1 in l accordance with LER 94-004. GE SIL 452 Supplemental 1, Revision 1 addressed concerns with the calibration of feedwater flow element transmitters.

The following actions have been taken in accordance with the recommendations of SIL 452 Supplemental 1, Rev 1:

Recommendation 1:

Inspect the feedwater flow nozzles at periodic intervals to maintain the calibration basis. On-line methods of discharge coefficient verification (ie: ultrasonics, tracer tests) may be utilized as a guide for inspection frequency.

The feedwater flow nozzles were calibrated in 1994 utilizing tracer methodology under Tests 1-176, 1-178, 2-110, and 2-112. The nozzles were not individually calibrated prior to original installation. The feedwater flow nozzles are inspected every other outage via an inspection port just upstream of the nozzle to maintain this tracer calibration. This activity is maintained within the station surveillance tracking (GSRV) program to ensure its frequency of completion.

In addition, the station has begun testing on the measurement of feedwater flow utilizing ultrasonic techniques. This method of feedwater flow measurement has not been permanently installed.

Recommendation 2:

Include calibration of the feedwater flow transmitters in instrument maintenance planning for each refuel outage.

The feedwater flow transmitters are calibrated once every refuel outage in accordance with the GSRV program.

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l assommendation 3:

Review procedures for calibrating feedwater flow i transmitters to eliminate potential biases. Also, review the full scale differential pressure calculation for

potential errors.

! A procedure for calibrating the feedwater flow

transmitters has been issued effective 12/26/95. The j procedure addresses zero and span shifts to ensure  !

l these biases do not affect flow measurement. The problem associated with transmitter calibration was  ;

j also acknowledged in the response to NRC NOV 94-005.

! The resistors used for deriving the feedwater flow fron l l the output of the transmitters were verified to be j accurate in Nuclear Work Requests Q15256 and Q15257.

The full span pressure drop calculation was reviewed for correctness and applied to the results of Tests 1- i 176, 1-178, 2-110, and 2-112 for the new nozzle i coefficient calculation. The correct thermal expansion factor was being used in the calculation. An error was

, found in the original span calculation which was i corrected in the subsequent calculation, documented j with Problem Identification Form 94-039, and placed on i the Nuclear Network.

i Recommendation 4:

, Determine whether the appropriate stainless steel area factor was included in the feedwater temperature j compensation process computer calculation.

l ~The process computer feedwater temperature compensation

calculation was reviewed with General Electric

assistance and found to be acceptable.

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Chin N hwer Snanien Synta E l-M Dunaminal I Recommendation 5:

l Verify that the measured feedwater temperature is corrected I

for any factor which may change the temperature from which j

is measured at the nozzle.

The core thermal power affect as described in the SIL from the Reactor Water Cleanup system does not apply j due to the Quad cities plant configuration.

Stratification of the final feedwater temperature has been identified as a factor affecting the process

! computer feedwater flow temperature correction. This

affect was documented in PIF 95-0111 and may occur only i when one of the three final feedwater heaters is

) removed from service. A temperature stratification may l occur which causes the final feedwater temperature to l be erroneously indicated. The affect on feedwater flow and core thermal power was evaluated in PIF 95-0111 response and was found to be bounded by the existing j uncertainty analysis. Further, plant procedures

require the unit to be derated when any final feedwater heater is removed form service.

i j Thus, factors which affect the feedwater temperature 5

have been evaluated and determined not to have a j detrimental affect on the feedwater flow measurement.

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UNIT 1(2)

REVISION O ATTACHMENT A (Page 1 of 1)

OFFSITE REVIEW AND INVESTIGATIVE FUNCTION TRAN5MiTTAL i Quad Cities Nuclear Power Station

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j FOR REVIEW:

1. Safety EvaluationsE involving an unreviewed safety question as defined in 10CFR50.59 l for:

a. Changes to procedures as described in the Safety Analyss Report.

b. Changes to equipment or systems as described in the Safety Analysis Report.

c. Tests or experiments E described in the Safety Analysis Report.

2. Proposed changes which involve an unreviewed safety question as defined in 10CFR50.50. /

a. Procedure changes.

1 I b. Equipment or system changes.

c. Tests or experiments l

j 3. Proposed changes to the Technical Speelnestions or Operating Ucenes i 4. Noncompliance with codes, regulations, orders, Technical Specifications, license j requirements, or lidemal procedures or instructions having nuclear safety significance.

5. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affects nuclear safety,

6. M REPORTABLE EVEffrS (LERs only). l i

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l 7. M recogntred indications of an unanticipated deficiency in design or operation of safety-related structures, systems, or components.

I 8. M changes to the Station Emergency Plan prior to implementation.  ;

i 1 9. M ltems referred by the Systems Engineering Supervisor, Statior Manager, Site Vice President, and General Manager of Qualty Programs and Assessments. ,

3 FOR INFORMATION:t 1  ;

i

10. Other OSR ltems/DocumentsE addressed above.

j; This Transmittal is being made in accordance with Quad Cities Nuclear Power Station ,

Technical Speelnestions 6.1.G.2.d(1) for information only. No specEc action is required i unless deemed necessary by Othdte Review and investigative Function.

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