ML20043A513

LER 90-007-00:on 900419,determined That Errors in Calculations for Measuring Feedwater Flow Resulted in Nonconservative Calibr of Flow Transmitters.Caused by Error in Design Calculation.Transmitters recalibr.W/900514 Ltr

ML20043A513
Person / Time
Site:	Hatch
Issue date:	05/14/1990
From:	Hairston W, Tipps S GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-1096, LER-90-007-01, LER-90-7-1, NUDOCS 9005220178
Download: ML20043A513 (9)

Text

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Ooorga Poe Company

• 333 Piedenoot Avenue Atlanta, Georgia 30308 hphone 404 $20 3105 Maang Adorcer

• 40 inverness Center Pa kway 1- Post Offeo Box 1295 B;nningham A;abama 35201 k!cphone 205 068 550 t the sw., rrr vitt tre systern W. G. Hairston, til Son:or Vce President Nuclear Opera $ons HL-1096 000525 May 14. 1990 U.S. Nuclear Regulatory Commission -

ATTN: Document Control Desk Washington, D.C. 20555 ,

.1 PLANT HATCH - UNIT 1 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 LICENSEE EVENT REPORT ERRORS IN FEEDWATER FLOW DP TRANSMITTER CALCULATIONS RESULTS IN 1% THERMAL OVERPOWER  !

Gentlemen:

Georgia Power Company is submitting the enclosed voluntary Licensee ,

Event Report (LER) because of the potential interest in the calibration of i BWR feedwater flow instrumentation and its effect on calculated thermal.

power output. This event occurred at Plant Hatch - Units 1 and -2.

Sincerely,

& .k. (kS W. G. Hairston,,III SWR /GKM/eb

Enclosure:

LER 50-321/1990-007 c: (See next page.)

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. ( pCNyia l'0Wei J L U.S. Nuclear Regulatory Commission l May 14, 1990  :

1 Page Two c: Georaia Power' Company Mr.. H. C. Nix, General Manager - Nuclear _ Plant-  ;

Mr. J. D. Heidt, Manager Engineering-and Licensing - Hatch GO-NORMS U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. L. P. Crocker, Licensing Project Manager - Hatch- '

U.S. Nuclear Reaulatory Commission. Reaion II  :

!- Mr..S. D. Ebneter, Regional Administrator Senior Resident Inspector - Hatch 1

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NAME TE LEPHONs NUMBEM A811& CQut Steven B. Tipps, Manager Nuclear Safety and Compliance, Hatch 911 R 316 67 l- 17 8 1511 COMPLETE ONE LINE FOR E ACM COMPONE NT F A'LURs Ot3CRIBED IN THit RtPORT H3)

MA C ntFDHTA LE CAust SvsitM CovPONENT 5] PORTABLE p c Aygg gygygg coy,ggg 7 C I I I I I I 1. I i i f 1 I I l l l l I l l l l l i 1 l 1 SUPPLEMENT AL REPORY E XPECitD H4i MON 1H Dat vtAR v t s !!* ves comorere f o?CTED sugMiss>ON DA Til ko l l l l A32TR ACT tt mur to rop .peces , e . eopros,merely r>rreen e,op,e wue ryotentwa Inaest H 6' On 4/19/90 at approximately 1520 CDT, Unit 1 was in the Refuel mode with the core completely loaded and Unit 2 was in the Run mode at an approximate power level of 2380 CMWT (approximately 98% rated thermal power). It was determined that errors in the Unit 1 and Unit 2 calculations for measuring feedwater flow resulted in non-conservative calibration of the flow transmitters. This, in turn, resulted in an indicated feedwater flow of approximately 1% less than actual and a calculated thermal power of approximately 1% less than actual. Consequently, reactor power operation at a calculated 100% rated thermal power would have resulted in operation at an actual 101% rated thermal power. This condition apparently has existed since August 1978 on Unit 1 and July 1979 on Unit 2. Operation slightly above rated thermal power would not have impaired the ability of the plant to achieve and maintain a safe shutdown condition or violated applicable safety analyses.

The cause of this event is a design calculation error by the Nuclear Steam Supply System (NSSS) vendor. Calculations supplied by the NSSS vendor via letters dated 9/9/75 (Unit 1) and 4/5/78 (Unit 2) were incorrect in that they used the wrong area thermal expansion factor for the feedwater flow elements and did not include the span correction factor as recommended by the transmitters' vendor. These errors resulted in non-conservative calculations of transmitter full scale flow values and, consequently, non-conservative calibration of the feedwater flow differential pressure transmitters.

Corrective actions for this event included administratively de-rating Unit 2 to 98% -

rated thermal power until flow transmitters 2C32-N002A and B could be recalibrated.

The Unit I flow transmitters will be recalibrated before startup from the current maintenance / refueling outage.

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General Electric - Boiling Water Reactor 1 Energy Industry Identification System codes are identified in the text as (EIIS Code XX).

SUMMARY

0F EVENT On 4/19/90 at approximately 1520 CDT, Unit 1 was in the Refuel mode with the core-completely loaded and Unit 2 was in the Run mode at an approximate power level of 2380 CMWT (approximately 98% rated thermal power). It was determined that errors in the Unit 1 and Unit 2 calculations for measuring feedwater flow resulted'in. ,

non-conservative calibration of the flow transmitters. This, in turn, resulted in an indicated feedwater flow of approximately 1% less than actual and a calculated thermal power of approximately l% less than actual. Consequently, reactor power operation at a calculated 100% rated thermal power would have resulted in~ operation at an: actual 101% rated thermal power. This condition apparently has existed since f August 1978 on Unit 1 and sluly 1979 on Unit 2. Operation slightly above rated .

thermal power would not have impaired the-ability of the plant to achieve and '

maintain a safe shutdown condition or violated applicable safety analyses.

The cause of this event is a design calculation error by the Nuclear Steam Supply System (NSSS) vendor. Calculations supplied by the NSSS vendor via letters dated '

9/9/75 (Unit 1) and 4/5/78 (Unit 2) were incorrect in that they used the wrong area thermal expansion factor for the feedwater flow elements and did not include the  ;

span correction factor as rec'ommended by the. transmitters' vendor. These errors

• resulted in non-conservative calculations' of full scale flow values and, consequently, non-conservative calibration of the feedwater flow differential pressure transmitters, i Corrective actions for this event included administrative 1y de-rating Unit 2- to 98%

• rated thermal ;:ower; calculating the' correct calibration factors using~ the guidance -

of Service Information Letter 452, Supplement 1; revising procedure 57CP-CAL-069-25,

• Rosemount Model ll51AP, OP and GP Transmitters," to incorporate the corrected calibration factors; recalibrating flow transmitters 2C32-N002A and B; and updating plant documents to reflect-the corrected calculations. Similar actions will be taken on Unit 1 before startup from its current maintenance / refueling outage.  ;

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DESCRIPTION OF EVENT In October 1988, Hatch received Rapid Information Communication Service Information' '

Letter (RICSIL) No. 030, "Feedwater Flow Element Transmitter Calibration Error,"

dated 10/4/88. The RICSIL stated an error was discovered at an operating Boiling Water Reactor in the calculations used to determine feedwater flow from a ,

differential pressure (dp) transmitter. The material expansion factor for the feedwater flow element had been omitted from the calculation for the transmitter's full scale flow value. This error resulted in feedwater flow indicating lower than actual .

Based on this information, plant Engineering Support personnel reviewed the feedwater flow dp transmitter calculations for Units 1 and 2 to determine if the .

flow element's material expansion factor had been omitted from the calculations as

• described in the RICSIL. Hand-written calculations for Unit 1 were found. They had-included a material expansion factor. The Unit 2 calculations could not be found.

• Engineering Support personnel then contacted the plant's Architect / Engineers (A/Es) and General Electric in an attempt to locate the Unit 2 feedwater flow dp ,

transmitter calculations. The calculations could not be found, but General Electric informed Engineering Support persronnel the problem described in the RICSIL probably did not exist on Unit 2 because it was newer vintage than th_e plant which was the

! subject of the RICSIL. (The Unit 2 calculations were found in July 1989 by Engineering Support personnel and they did include the material expansion factor.)

Therefore, the event described in the RICSIL was thought not to apply-to Hatch and 1 there was no reason to suspect other errors in the Hatch calculations.

In November 1988, Hatch received Service Information Letter-(SIL) 452, Supplement.1, "Feedwater Flow Element Transmitter Calibration," dated 11/18/88. Engineering.

Support personnel, although they had no reason to suspect any problems, began a detailed review of the calculations for Units 1 and 2 in light of the SIL recommendations. This review identified some discrepancies in the documentation available. This documentation was needed to verify the then-current method of

, determining feedwater flow. Engineering Support personnel contacted General l- Electric to address the items which had been identified as deficient or inadequate. <

In a memo dated 3/31/89, Engineering Support personnel notified their-supervision of the plan of action, status of review, and problems with documentation in regards to SIL 452, Supplement 1. At that time, the Unit 2 calculations still had not been located.

In July 1989, Engineering Support personnel, in their continuing review of feedwater '

flow dp calculations, located the Unit 1 and 2 feedwater-flow dp transmitter calculations in the Hatch Nuclear Plant Document Control Storage Vault among startup.

testing documents. - The calculations found were the official calculations, including i

the material expansion factor, flow element data, plant data,. viscosity versus water temperature curve,- and equations used. These data and calculations had been transmitted to the site by General Electric in letters dated 9/9/75 (Unit 1) and 4/5/78 (Unit 2).

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0l0l7 - oF 0l 7 Per their plan of action, Engineering Support personnel requested General Electric to determine if the calculations discovered in the Document Control Vault were acceptable. It was determined that General Electric instead would provide new Unit 2 full scale flow values calculated per SIL 452, Supplement 1. This was done and the new calculations transmitted to the site via a General Electric letter dated 11/6/89.

Upon receipt of the new calculations, Engineering Support personnel discovered several errors in the new calculations. Therefore, General Electric was requested to correct the errors and again perform the calculations. The corrected calculations were sent to the site via a letter dated 12/15/89.

Upon receipt of the revised calculations, Engineering Support personnel, in light of the errors found in the calculations supplied 11/6/89, proceeded to conduct a '

thorough, line-by-line review of the calculations. The programming of a Personal Computer to perform the calculations, the gathering of the applicable reference material, and the actual verification of both sets of calculations took until approximately 4/18/90. This effort included a check of the site's calculation results to ensure no errors existed in the programmed equations or the values used in the equations.

On 4/18/90, Engineering Support personnel, once they were confident their feeowater flow dp transmitter calculations were accurate, reported to their supervisor that calibration values used in the Unit 2 flow dp transmitter calibration procedure were in error and resulted in an indicated feedwater flow, and thus, a calculated thermal power of approximately one percent less than actual. (At this time, Unit 1 was in a maintenance / refueling outage; therefore, efforts were concentrated on Unit 2.)

Engineering Support personnel determined that the calibration values were in error because the original calculations had omitted the span correction factor for the dp transmitters and had used the incorrect area thermal expansion coefficient for the feedwater flow elements. The new calculations showed these errors combined to result l in non-conservative feedwater flows and thermal power outputs of approximately one l percent.

Investigation revealed the incorrect calculations had been used to revise the Unit 1 feedwater flow dp transmitter calibration procedure on 8/14/78 and the Unit 2 calibration procedure on 7/18/79. ' Additional document searches disclosed the Unit 1 flow dp transmitters had been calibrated using incorrect calibration values on 8/5/78 and the Unit 2 flow dp transmitters on 7/18/79. Consequently, as of 8/5/78 on Unit 1 and 7/18/79 on Unit 2, when the feedwater flow dp transmitters were calibrated based on the incorrect calculations, actual thermal power was approximately one percent higher than that calculated by each unit's process computer. Therefore, operation at a calculated 100% rated thermal power would have  !

l resulted in operation at an actual 101% rated thermal power.

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PLANT HATCH, UNIT 1 TEXT f# more apsee 4 m n o ls lo lo lo l3 l 2l1 9l0 0l0l7 0l 0 0l5 OF 0l7 On 4/19/90, site management received final confirmation of the accuracy of the revised feedwater flow dp transmitter calculations (generated by plant Engineering Support personnel) from one of the plant's A/Es, Southern Company Services. At that point, the site's General Manager - Nuclear Plant ordered Unit 2 administratively de-rated to 98%

rated thermal power until the correct transmitter calibration values could be incorporated into the calibration procedura and the traasmitters recalibrated. Power was reduced to approximately 98% rated thermal power at 1315 CDT. Procedure 57CP-CAL-069-2S and NSSS vendor supplied Instrument Data Sheets contained in plant document SX-29453 were revised to incorporate the correct calibration values for dp transmitters 2C32-N002A and B.

On 4/21/90 at approximately 1322 CDT, plant Instrument and Controls personnel began the recalibration of flow dp transmitters 2C32-N002A and B using the corrected calibration values. At approximately 1500 CDT, dp transmitter 2C32-N002A was calibrated and returned to service. A process computer calculation of thermal power performed at approximately 1505 CDT showed an increase of about 0.45% rated thermal power (11 CMWT ) . Actual power was unchanged, but calculated power increased due to the recalibration. At approximately 1630 COT, dp transmitter 2C32-N002B was calibrated and returned to service. Another calculation of thermal power performed at approximately '

1647 CDT showed an increase of about 0.5% rated thermal power (12 CMWT). Total increase in calculated power resulting from the re-calibration of the feedwater flow transmitters 2C32-N002A and B was approximately 0.95% (23 CMWT).

On 4/21/90 at approximately 1850 CST, Unit 2 was returned to 100% rated tharmal power '

following approtal of the site's General Manager - Nuclear Plant, j CAUSE OF THE EVENT The cause of this event is a design calculation error by NSSS vendor (General j Electric). In letters dated 9/9/75 and 4/5/78, incorrect full scale flow differential  !

pressure values for Units 1 and 2, respectively, were supplied to the. site. Site l personnel incorporated these incorrect values into the Unit 1 and 2 calibration procedures on 8/14/78 and 7/18/79, respectively. The feedwater flow dp transmitters {'

were calibrated using the incorrect values on 8/5/78 (Unit 1) and 7/18/79 (Unit 2). I This resulted in an indicated thermal power approximately one percent lower than actual.

The NSSS vendor-supplied full scale flow values were in error because the incorrect  :

area thermal expansion factor was used for the feedwater flow elements. This factor accounts for thermal expansion of the flow element as it is heated from the temperature at which each flow element is tested in the lab (68'F) to rated feedwater temperature (420*F). The vendor-supplied flow values were derived using the thermal expansion factor for carbon steel and not stainless steel, the material for Hatch's flow l

eleaehts. Calculations for the Unit 2 flow elements indicated the use of the wrong area thermal expansion factor resulted in a non-conservative error of approximately j 0.17% in indicated feedwater flow and, hence, calculated thermal power, j i

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PLANT HATCH, UNIT 1 o ;5 ;o ga ;o ;3 ; 2;l 9; O 0l0 l7 ._ Og 0,0 ;6 or 0l7 ine u ~,. . o . m ma mm The full scale flow values also were in error because the transmitters' span correction factor was not included in the calculations as recommended by their vendor. This factor accounts for span and zero shifts which result from static pressurization of the transmitter. The calculations were not adjusted by the span correction factor.

Calculations indicated not using the correction factor resulted in a non-conservative error of approximately 0.8%.

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This report is being submitted voluntarily because of the potential interest _in the calibration of BWR feodwater flow instrumentation and its affect upon calculated thermal power output.

Feedwater flow to the reactor pressure vessel is determined by measuring the change in pressure as the feedwater flows through a venturi (flow element) in each of two feedwater lines. The change in pressure is measured by differential pressure 1 transmitters (flow dp transmitters) manufactured by Rosemount, one for each flow element. Electrical circuitry converts these differential pressure signals to flow signals. The two flow signals are then combined into a total feedwater flow signal which is used by the process computer to calculate the unit's thermal power output.

The flow dp transmitters are calibrated periodically to ensure they are providing an accurate indication of feedwater flow. The calibration is performed by checking that the dp transmitter's output matches a known input within certain tolerances. The .

expected output, in inches of water of differential pressure, for a given input, in  !

pounds of water per hour, is determined by calculation. Equations are used to determine the dp transmitter's expected output for various flows. The equations take into account such factors as the flow element's characteristics as determined from laboratory testing, changes in the flow element's environment from the laboratory to 1 the field, the static pressure span shift for the flow transmitter, and plant unique '

data. The results of the calculations using these equations are used in plant  !

calibration procedures for expected transmitter output values at various known inputs. i" The flow dp transmitters are adjusted so their outputs match the calculated value at each calibration point.

In this event, the expected values given in the Unit 1 and Unit 2 calibration procedures for the feedwater flow dp transmitters were incorrect by approximately one percent in the non-conservative direction. This was because some of the factors used in the equations to determine these values were either wrong or not included. Because i

these values were incorrect, the output of the flow dp transmitters was incorrect, leading to thermal power calculations which also were incorrect by approximately one percent in the non-conservative direction.

Operation of Plant Hatch slightly above rated thermal power did not impair the ability i of either Unit 1 or Unit 2 to achieve and maintain a safe shutdown condition, or violate any applicable safety analyses. The applicable safety analyses include fuel thermal limits, loss-of-coolant accidents (LOCA), and containment response, i

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PLANT HATCH, UNIT 1 itKT IW more space a requeest, use ed* hone! NRC form 388A s)(11) o ls lo lo lo l3 l 2l1 9l 0 -

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0l 0 0 l7 0F 0l7 At -least a two percent allowance for power uncertainty.is included in all fuel safety -

analyses. These analyses are used to stipulate the operating limit minimum critical power ratio (MCPR) and'to determine linear'. heat generation rate (LHGR) limits and }

- maximum average planer linear heat generation rate (MAPLGHR) limits for the fuel.

Loss-of-Coolant Accident analyses can also set fuel MAPLGHR limits. These calculations are performed at no less than 102% thermal power, per the requirements of 10 CFR 50, Appendix K. Containment response analyses to postulated accidents also have sufficient margin to assure applicable limits were not violated by operation at approximately 101%  ;

rated thermal power.  :

Based on the above' assessment, it is concluded that no FSAR analyses safety limits or Technical Specifications operating limits have been exceeded as a result of errors in the feedwater flow dp. transmitter calibration. Therefore, there was no impact on the health and safety of the public as a result of this condition.

CORRECTIVE ACTIONS i Upon confirmation of a' non-conservative error in the 'feedwater flow dp transmitters' full scale flow calculations, plant management ordered Unit 2, the only unit operating- '

at the time, reduced to 98% rated thermal' power.' Southern Company Services, one of the i plant's A/Es, calculated correct calibration facts -t Y+ 2 flow dp transmitters 2C32-N002A and B using the guidance of SIL 452. . otMF ' The corrected calibration factors were incorporated .into prou e" d Tr v d69-2S and plant document SX-29453 and, on 4/21/90, the Unit 2 flow dp trar v ttm wm calibrated using the corrected factors. The unit was returned to 1C pcwer N11mng the flow dp transmitters' calibration.

Corrected Unit 1 feedwater flow dp transmitter calibration factors will be calculated and Unit 1 procedure 57CP-CAL-069-1S, "Rosemount Model 1151AP, DP and GP Transmitters,"

will be revised. Feedwater flow dp transmitters 1C32-N002A and B will be calibrated using the corrected calibration factors. These actions will be-completed prior to Unit 1 startup from its current maintenance / refueling outage. >

Unit 1 and Unit 2 documents will be updated, as appropriate, to reflect the corrected feedwater flow dp transmitter calculations.

ADDITIONAL INFORMATION No similar events in which calculational errors resulted in either unit operating above l

its rated thermal power level were noted.  !

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