05000324/LER-1993-008

LER 93-008-00:on 930714,core Rated Thermal Power Exceeded Allowable Amount Due to Feedwater Flow Inaccuracy.Cause Believed to Be Due to Erosion/Corrosion of FW Flow Elements. FW Instrumentation recalibr.W/930812 Ltr

ML20056D573
Person / Time
Site:	Brunswick
Issue date:	08/13/1993
From:	Jonathan Brown, Mcgowan J CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BSEP-93-0117, BSEP-93-117, LER-93-008-02, LER-93-8-2, NUDOCS 9308170089
Download: ML20056D573 (5)

LER-2093-008,

Event date:
Report date:
3242093008R00 - NRC Website

text

V ,

• Carolina Power & Light Company

""emrtmavs4Lva?atsc3 Brunswick Nuclear' Plant  !

P. C. Box 10429 Southport, N.C. 28461-0429 August 12, 1993 i

FILE: B09-13510C 10CFR50.73 ,  ;

SERIAL: BSEP 93-0117 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555  !

I BRUNSWICK NUCLEAR PLANT UNIT 2 l DOCKET NO. 50-324 .l LICENSE NO. DRP-62 ,

LICENSEE EVENT REPORT 2-93-008 I

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Gentlemen: j In accordance with Title 10 of the Code of Federal Regulations, the enclosed j

-Licensee Event Report is submitted. This report fulfills the requirement for a  ;

written report within thirty (30) days of a reportable ' occurrence and is submitted in accordance with the format set forth in NUREG-1022,' September 1983.

?

Very truly yours, Nh

. M. Brown, Plant' Manager - Unit 1 'j Brunswick Nuclear Plant t

JFM/jfm  ;

1 Enclosure f 1

cc: Mr. S. D. Ebneter Mr. P. D. Milano BSEP NRC Resident Office i

i 170070 9300170089 930813 [c // .)

PDR S

ADOCK 05000324 PDR I ['h R

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t NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 '

(5/92) . EXPIRES: 5/31/95 E STIM ATED BURDEN PER RC SPONSE TO COMPLY WITH TH!S INf ORM ATION cot LE CTION FtE QUEST: 50.0 HRS. FORWARD LICENSEE EVENT RL.,/ ORT (LER) COMMEyTs RtoARo,Na ,URDE N E,T,M,1t yo THE ,NEcRMA1,oN AND RECORDS MANAGEMENT BRANCH (MNBB 771al, U.S. NUCLE AR REGULATORY COMMISSION. WASHINGTON. DC 20H5-DD01. AND TO THE PAPERWDRK REDUCTION PROJECT (3150-0104). OFhCE OF

' MANAGEMENT AND BUDGET. WASH!NGTON, DC 20S03.

FACILIT'Y NAME (1) DOCKET NUMBER {2) PAGE (3)

Brunswick Steam Electric Plant, Unit 2 05000324 1 of 4 i

TITLE 141 CORE RATED THERMAL POWER EXCEEDED THE ALLOWABLE AMOUNT DUE TO FEEDWATER FLOW INACCURACY EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH Day YEAR YEAR t/ONTH CAY YEAR NuMetR NUMBER BSEP 1 05000325 7 14 93 93 - 008 - 0 8 13 93 ' ACurY NAMt OccKtT NuMetR 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 C"R 8: (Check one or more of the followingM11) 1 MODE $) 20 402ft:) 20.40!.(c) 50 73(aH2Hiv) 73.71(b) 20 405taHini) 50.36(cH1) 50.73laH2Hvi 73.71(c)

PMR (EVEL (10! 100 20 405(aH1Hii) 50 3f4(cH2n 50 73(aH2)M4 OTHER 20 405(aH1Hiul X 50.73taH2Hi) 50.73(eH2)lviiiMA) (Spenfy m Abstract and Text)  ;

20 405f aH1Hiv) 50 73iaH2Hiii 50.73(aH2)(voiHE4 20 405(aH1)M 50 73(aH2Heid 50.73(aH2Hul LICENSEE CONTACT FOR THIS LER (12)

AAME TELEPHONE NUVfflR Jeanne F. McGowan, Regulatory Compliance Specialist (919) 457-2136 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE Sr$ TIM C OMDONE NT M Af UF ACTURER C A USE SYSTE M COMPONENT MANUF ACTURE R p {

B SJ NZL Y SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED

" U N'" D'* "*" ,

g3 g SUBMISSION of m. cmem Enf cTn susussicN C ATu DATE (15)

ABSTRACT (limit to 1400 spaces, i e. approwmateN fifteen single space typewritten kneal (161 f

On July 14, 1993, Unit 2 was cperating at an indicated power of 100t. A Sodium tracer injection test had been performed on July 2, 1993, to determine the calibration constants for the main steam flow elements and as part of the Brunswick power up-rate project. The >

V preliminary results of the test, received on July 14, 1993, indicated a non-conservative inaccuracy in feedwater flow of 1.094t. This inaccuracy in feedwater flow corresponds to an increase in core tnermal power of approximately 1%. At 1621 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.167905e-4 months <br /> on July 14, 1993, Unit 2 reduced reactor power to an indicated 2405 MWth Iwhich corresponds to an actual '

reactor power of c100t) Administrative controls were established to ensure core thermal power would not be exceeded. The administrative controls will remain in effect until feedwater instrumentation can be recalibrated.

The cause of the event is believed to be due to erosion / corrosion of the feedwater flow elements.

"Ine safety significance of the event is minimal. Reviews of thermal limit calculations determined that an overpcwer of approximately 11 would have a negligible impact on fuel integrity. No unsafe cperating ccnditions cr transient consequencec would have occurred as a result of the feedwater flow inaccuracy.

The cause classification for thie event per the criteria of NUREG-1022 is Cther.

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NRC FORM,366A U. S. NUCLEAR REcVLATORY COMMISSION APPROVED OMB NO. 3130-0104 j (5/p2) EXPIRES: 5/31/95 ,

, E STIM ATED BURDEN PER RE SPONSE TO COMrY WITH THis INFORMATION COLLECTION REQUEST: 60.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMEu1S u cAnomunotuSmut To rye ,N,oRMuioN AND TEXT CONTINUATION RE CORDS MANAGEMNT MANCH (MNbB 771dl U.S. NUCLE AR REGULATORY COMMISSION. WASHINGTON, DC 205Ek0001, AND TO 5 THE PAPE RWORK REDUCTION PROJECT (3150-0104L OFFfCE OF MANAGFMENT AND BUDGET WASHINGTON DC 205C3.

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FACitJTY NAME 0) DOCKET NUMBER Q) LER NUMBER (6) . PAGE {3)  !

SE QUENTIAL REVISION Brunswick Steam Electric Plant "#" "#"

05000324 2 of 4 f Unit 2 93 - 008 - 0 TEKT Uf more space is required. use additionalNRC Form 366A'al(17)

TITLE CORE RATED THERMAL POWER EXCEEDED THE ALLOWABLE AMOUNT DUE TO FEEDWATER FLOW INACCURACY INITIAL CONDITIONS On July 14, 1993, Unit 2 was operating at an indicated power of 100%. A Sodium tracer injection test had been conducted on July 2, 1993 to determine the calibration constants for the main steam flow elements and as part of the Brunswick power up-rate project. The results f rom the Sodium tracer test , received on July 14, 1993, indicated a non-conservative inaccuracy of 1.094%. At 1621 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.167905e-4 months <br /> on July 14, 1993, Unit 2 reduced reactor power to an indicated 2405 MWth (which corresponds to an actual reactor power of <100%)

EVENT NARRATIVE On October 4, 1989, an Engineering Work Request (EWR) was initiated which identified a mismatch on Brunswick Nuclear Plant (ENP) Unit 1 between the indicated feedwater flow ,

and steam flow. The EWR was initiated to have the calibration data for the Main Steam flow elements reviewed to determine if calibrations needed to be performed. The work l was targeted for March, 1991.

i In March, 1991, the EWR was revised to propose that a Sodium tracer test be performed for Units 1 and 2. The meter differential for the main steam flow elements for both units had not been revised since initial calibration. That value was believed to be inaccurate due to wear on the nozzle plate over the last 15 years. The Sodium tracer ,

test was requested to determine the feedwater flow rate, thereby determining the calibration constants for the main steam flow elements. The main steam flow elements could then be calibrated to ensure the feedwater flow indication and the main steam flow indication are in agreement. The Fodium tracer test was scheduled to be performed in 1992 for Units 1 and 2. Both Units were shutdown in early 1992 and the tests were rescheduled to be performed after each Unit's startup.

Measurement of the flow rate in each of the two feedwater loops on Unit 2 was made by two Sodium tracer tests, on June 30 and July 2, 1993. The data from the test conducted on June 30 was discarded due to weighing errors in the preparation of tracer standards.

The final results of the flow test, released by General Electric (GE) on August 11, l 1993, indicated an actual feedwater flow inaccuracy of 1.094%. An increase of 1.094%  ;

in feedwater flow corresponds to an increase in core thermal power of approximately 1% i (based on a manual heat balance calculation). l At 1621 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.167905e-4 months <br /> on July 14, 1993, Unit 2 reduced reactor power to an indicated 2405 MWth. ,

A Standing Instruction (SI 93-166) was issued to limit core thermal power to less than j or equal to 2405 MWth. Im indicated power of 2409.49 MWth corresponds to 100% power; however, to ensure that core thermal power would not be exceeded, an administrative limit of 2.405 MWth was established until the Process Computer could be changed to reflect the new information. On August 5, 1993, the necessary changes were made to the Procecs Computer and a new Standing Instruction (SI 93-172) was issued to limit core thermal power to less than or equal to 2409.49 MWth indicated. The administrative limit will be in place until the feedwater flow instrumentation can be recalibrated.

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i l NRC FORM 366A U, S. NUCLEAR REGULATORY COMMISSION APPROVED OMB No. 3150-0104 (5/92) EXPlRES: 5/31/95

, ESTIMATED BURDEN PER RE SPONSE TO COMPLY WITH THis

• INFORMATION CDLLE CTION REQUEST 2 $0.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)

COMME NTs REcARomo euRoEN ESTiu ATnO THE imORuATiOuNo  ;

TEXT CONTINUATION RE CORDS M ANA GEME NT ORANCH (MNBB 7714), U.S. NUCLE AR j REGULATORY COMMfSSION. WASHINGTON, DC 20565 0001, AND TO THE PAPERWORK REDUCTION PRO.!ECT {3150 0104L OFFICE OF MANAGEMENT AND BUDGET, W ASHINGTON, DC 20503, l FACluTY NAME (1) DOCEET NUMBER (2) LER NUMBER (61 PAGE (3) [

SEQUENTIAL REVISION Brunswick Steam Electric Plant """ """

05000324 3 of 4 i Unit 2 93 - 008 - 0 l

TEKT lit more space is requirect. use additicmst NRC form 36CA 's) (17) 1 l CAUSE OF EVENT The cause of the event is believed to be eresion/ corrosion of the feedwater flow elements. This is based on operating experience at three other plants which have also experienced non-conservative errors in feedwater flow measurement. Visual inspections at two of the plants revealed erosion of the carbon steel piping around the high pressure taps. Flow element wall loss resulted in the non-conservative deviation. The third plant was unclear as to whether this problem existed from construction or was due to nozzle erosion, improper tap location, or fouling.

I CORRECTIVE ACTIONS Corrective actions include the following:

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1. Maintain Unit 2 reactor power at less than or equal to 2409.49 MWth )

indicated until the feedwater instrumentation can be recalibrated. l

2. Recalibrate the Unit 2 feedwater instrumentation to the calculated flow coefficients based on the tracer test data.

3. Perform a _,imilar test on Unit 1 to determine actual feedwater flow.

1 SAFETY ASSESSMENT l An overpower of approximately 1% would non-conservatively affect the calculation of all three monitored thermal limits; Minimum Critical Power Ratio (MCPR) , Maximum Average Planer Linear Heat Generation Rate UMPLHGR), and Peak Linear Heat Generation Rate (PLHGR). The overpower of 1% would result in an increase in all thermal limits of approximately the same magnitude. Typically, Brunswick Nuclear Plant operates with j generous margin to thermal limits (on the order of 5-10%), however, it is possible during powP ascension / maneuvering, that thermal limits may approach to within 1% of l the Technical Specification limits for short periods of time. j The MCPR limit is used to prevent overheating of the fuel rod cladding due to inadequate cooling following postulated events. The derivation of the Safety Limit  !

MCPR values for GE fuel designs includes the effects of process monitoring uncertainties. One of the uncertainty inputs is feedwater flow, which is the largest component in reactor power uncertainty. The uncertainty used in feedwater flow is

1. 7 f> t , which bounds the detected feedwater flow miscalibration of 1.094t. Therefore, operating with a 1.094% feedwater flow miscalibration does not impact MCPR and the i operating MCPR limits remain valid. Therefore, there is no negative impact on fuel integrity.

The MAPLHGR limits ensure fuel integrity following loss of coolant accidents. The analysis of postulated loss of coolant accidents for Unit 2 assume core thermal power conditions which are approximately 10% greater than the current rated core thermal power of Unit 2. Thus, fuel integrity follcwing a loss of coolant accident would not be affected by an increase in the actual core thermal power of 1%.

The PLHGR limit applies to long term steady state operation and is established to prevent fuel clad cracking due to differential expansion of the fuel pellet. PLHGR is

NRC FCRM 366A U. S. NUCtEAR REGULATORV COMMISSION APPROVED OMB NO. 3150-0104

'(5/921 EXPIRES: 5/31/95

1. E STIM ATED BURDEN PER RE SPONSE TO COMPLY WITH THIS INFORMATION CDU.E CTION REQUEST: 60.0 HRS. F ORW ARD  !

LICENSEE EVENT REPORT (LER) COMMEuu RtoARo NO BURotN ES1,M,TE TO wE ,N,ORM A1,0N ANo ;

. TEXT CONTINUATION RE CORDS M ANAGfME NT BRANCH (MNBB 77141, U.S. NUCLE AR REGULATORY CDMM6SION WASHINGTON, DC 2055frOOO1. AND TO THE FAPERWORK FIEDUCTION PROJECT (3150-0104). OFFICE OF  ;

MANAGEMENT AND BUDGET. WASHINGTON. DC 20503.

FACILfTY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) i SE QUE NTIAL REVISION Brunswick Steam Electric Plant """" """"'"

Unit 2 05000324 4 of 4 93 - 008 - 0  !

TEKT Uf more spece is reovired. use additional NRC form 36'6A's) (17) i redundant to MAPLHGR and is no longer monitored as a Technical Specification thermal limit. However, a review of the steady state operating histol,/ for Unit 2, Cycle 10 indicates that significant margin has existed due to prolonged reduced power operation. l Therefore, fuel integrity has not been impacted. l In conclusion, no unsafe operating conditicns or transient consequences would have occurred as a result of the f eedwater flow tr.iscalibration. This determination is based upon assessment of conservatisms contained within the supporting basis of the fuel thermal limit calculatiens.

PREVTOUS SIMILAR EVENTS None.

i EIIS COMPONENT IDENTIFICATION  !

i System /Cornonent EIIS Code )

l Feedwater System SJ j r

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