IR 05000369/1989021: Difference between revisions

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{{Adams
{{Adams
| number = ML20246N653
| number = ML19351A355
| issue date = 08/16/1989
| issue date = 09/29/1989
| title = Insp Repts 50-369/89-21 & 50-370/89-21 on 890707-18. Violations Noted.Major Areas Inspected:Operation of Unit 1 Following Improperly Performed Heat Balance & Concomitant Nonconservative Calibr of Power Range Nuclear Instruments
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-369/89-21 & 50-370/89-21
| author name = Belisle G, Burnett P, Vandoorn P
| author name = Julian C
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| addressee name =  
| addressee name = Tucker H
| addressee affiliation =  
| addressee affiliation = DUKE POWER CO.
| docket = 05000369, 05000370
| docket = 05000369, 05000370
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-369-89-21, 50-370-89-21, NUDOCS 8909080193
| document report number = NUDOCS 8910200110
| package number = ML20246N646
| title reference date = 09-20-1989
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| page count = 17
| page count = 2
}}
}}


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UNITED STATES s jj l
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' NUCLEAR REGULATORY COMMISSION l
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. 101 MARIETTA STREET,N.W.'
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ATLANTA, GEORGIA 30323 ~
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Repob, Nos.': 50-369/89-21'and~50-370/8P21'
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Licensee:' Duke. Power Company-
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422 South Church Street
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Docket Nos.:
50-369 and 50-370-
' License ~Nos.:
NPF-9 and NPF-17
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Facility Name: McGuire 1 and 2 f
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September 29, 1989
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Inspection Conducted: J'
7 - 18, 1989 j
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-Inspector:
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Duke Power Company E
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7:-T. 'Bu'rNett,' Re' actor: Inspector, TPS,
.Date Signed
 
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Inspectori
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k P.K.,VanDoorn,' Senior'ResidenyInspector DatelSigned Approved by:
CM ' C
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G. A Belisle, Chief-Date/ Sisned-
' Test Programs Section
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ATTN: 1 Mr. H. B. Tucker, Vice President Nuclear Production Department 422 South Church Street y.


-SUMMARY l
Charlotte, North Carolina 28242 Gentlemen:-
Scope:
SUBJECT:
REPORT NOS. 50-369/89-21 AND 50-370/89-21
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'Thank you for your response of September 20, 1989, to our Notice of Violation issued on August 21, 1989, concerning activities conducted at your McGuire
~ This reactive,. unannounced inspection addressed the operation of Unit-1 l
. facility.
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on. July' 5,r1989 following an improperly performed ' reactor heat balance and the concomitant. non-conservative calibration of the, power ' range nuclear instruments.
 
Results:
d Unit I? was found to have operated in excess of 101% of rated thermal l
power for; a period of nearly three hours.
 
Furthermore, the unit operated in.
 
I excess of 102% of rated thermal power for a period of less than 10 minutes.
 
It.


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We have evaluated your response and found that it meets the-requirements of 10 CFR 2.201.
:was determined that the ; licensee had an opportunity-to identify the - heat l
balance error several ^ hours before it was identified and before rated thermal l
power was exceeded.


Consequently, the overpower operation was identified as a l
We will examine the implem(ntation of your L.
violatien paragraph.4.


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. corrective actions during future inspections, i ::,.
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The miscalibration of the nuclear instruments did not lead to operation with the high' flux' trip setpoint greater than that used in the safety analyses of
We appreciate your cooperation in this matter.
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reactivity. transients.


The overpower-delta-temperature trip was functional l
Sincerely,
throughout the event.
 
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.g 0909080193 890821
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PDR ADOCK 0$000369
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REPORT DETAILS
original signed by
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1.
Persons Contacted Licensee Employees
*N. Atherton, Compliance
*D. Baumgardner, Unit 1 Operations Manager
*D. Baxter, Operations Support Manager
*D. Bradshaw, Operations, General Offices
*K. Carmley, Operations Training
*S. Copp, Planning and Materials
*D. Ethington, Compliance Engineer
*G. Gilbert, Superintendent of Technical Services
*T. Hammond, Engineer, Instrt. mentation and Electrical
*G. Hart, General Supervisor, Instrumentation and Electrical
*E. Hite, Maintenance Engineer, General Offices
*R. Isenhour, Jr., Quality Assurance
*T. Kibler, Engineer, Performance
*M. Kitlan,:Jr., Reactor Engineer
*M. Mallard, Operations, General Offices
*T. McConnell, Station Manager
*J.- Neel, Supervisor, Instrumentation and Electrical E. Owens, Maintenance Engineer
*B. Pitsea, Operations Engineer
*C. Roberson, Engineer, Performance
*J. Rowe, NPD Engineer, General Offices
*M. Sample, Maintenance Superintendent R. Sharp, Compliance Manager
*G. Small, Safety Review Group
*D. Smith,-Test Engineer, Performance
*J. Snyder, Performance Engineer
*W. Suslick, Engineer, Test Group R. Travis, Operations Superintendent Other licensee employees contacted included engineers, technicians, operators, mechanics, and office personnel.
* Attended exit interview on July 10, 1989.
Acronyms and initialisms used throughort this report are listed in the last paragraph.
2.
Overpower Operation of McGuire Unit 1 (61706)
Unit I started up for operating cycle 6 in January 1989.
On January 17, 1989, PT/0/A/4150/03, Thermal Power Output Measurement, was completed successfully.
The three acceptance criteria for the test were:
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Caudie A. Julian Caudie A. Julian, Chief
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Step 11.1 The primary core power level derived from the primary heat balance calculated hy the OAC (Enclosure 13.2 of the PT) and that obtained by the off line computer calculation ( Enclosure 13.6 of the PT and computer program MNSPHB) shall agree within 2% F.P. (absolute difference).
 
Step 11A2 The secondary core power level derived from the secondary heat balance calculated by the OAC (Enclosure 13.2 of the PT) and that obtained by the off line computer calculation ( Enclosure 13.6 of the PT and computer program MNSSHI) shall agres within i 2% F.P. (absaiute difference).
 
Step 11.3 The Best Estimate Thermal Power calculated by the OAC (Enclosure 13.2 of the PT) shall' agree with the Best Estimate Thermal Power obtaintd on Enclosure 13.3 (of the PT) within 12% F.P. (absolute difference).
 
The Best Estimate Thermal Power is defined as:
Q = ALPHA * (secondary power) + (1-ALPHA) * (primary power).
 
For secondary power < 20% RTP, ALPHA = 0.0 For secondary power > 50% RTP, ALPHA = 1.0 Otherwise, ALPHA = -2/3 + (secondary power (%)/30)
At low power and low feedwater flow rates, there is considerable variation in the indicated feedwater flow and, hence, in the calculation of secondary side power.
 
Use of the Best Estimate Thermal Power is an attempt to provide the operators with a stable and reasonable display of thermal power over the entire operating range.
 
It can be seen in the above equation that the inherently more accurate secondary side power measurement is the sole term above 50% RTP.
 
The results of this PT in percent of RTP were:
98.27 = primary power level from MNSPHB 100.05 = sseondary power level from MNSSH1 99.79 = a.1 rage secondary power level from 0AC/ TOP 98.01 = average primary power level from 0AC/ TOP It can be seen that the differences between pairs of primary and secondary power measurements were less than 2%.
- IP/0/A/3007/17, NIS Power Range Calibration to Best Estimate Thermal Power, is performed by IAE technicians upon demand by operations person-nel.
 
The demand is generated by a difference of 2% RTP between PRNIs and thermal power in steady-state conditions or a real or anticipated
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difference of 5%.when power level is being changed.
 
Step 10.1.3 requires the following computer points be recorded in the procedure:
P1385: BEST ESTIMATE THERMAL POWER, P1445: SECONDARY THERMAL OUTPUT, and P1447: PRIMARY THERMAL OUTPUT.
 
Step 10.1.4 and 10.1.5 require that these points be verified to be within 2% RTP by the technician, and, if not within this limit, a reactor group engineer is to be contacted to determine which OAC point is to be used in the calibration.
 
On July 1,1989, with the primary thermal output indicating 34.7% 2TP and secondary thermal output indicating 31.9% RTP, a reactor group engineer selected P1445 as the basis for the PRNI calibrations.
 
Later that day, all three points read within the 2% allowance, and the technician chose to calibrate PRNIs against best estimate thermal power at 49.7%.
On July 2, 1989, with Unit 1 at about 55% RTP, work request 139042 OPS was issued for the repair of the IC steam generator control level gauge. An IAE technician was assigned to perform the repair in accordance with IP/0/A/3001/01C, Main Steam Flow Calibration, Loop C, Channel I.
 
Step 10.1.4 of that procedure required that the following computer points be locked out:
A1072: STEAM GENERATOR C MAIN STEAM FLOW - CHANNEL I, A0867: STEAM GENERATOR C FEEDWATER FLOW - CHANNEL I, and A1119: STEAM GENERATOR C MAIN STEAM PRESSURE - CHANNEL. I.
 
To lockout a point means that the OAC does not read the instrument source, but uses a substitute value entered at lockout for all calculations using the point.
 
Step 10.6.5 requires, as part of system restoration, that the same computer points be unlocked.
 
That step was not performed.
 
Hence, the OAC continued to use the substitute values in calculations rather than the actual values of these variables.
 
Computer point A0867 is one of two analog measurements of differential pressure across a calibrated flow venturi.
 
Each point provides an independent measurement of feedwater flow to steam generator C.
 
That direct measurement is the primary variable input to the OAC program FLO, which converts it to units of millions of pounds mass per hour.
 
The converted measurement is available to other applications, including other 0AC programs, at computer points P1416 and P1095.
 
The later is a two-minute average of results calculated with ten-second periodicity.
 
P1095 is one of two. measurements of feedwater flow to steam generator C, which are averaged and used by TOP, the OAC program for calculating secondary side thermal power.


The other point P1096 was not affected by the procedural error.
A1119 is one of three channels of analog input of steam pressure for each generator.
In TOP, they are averaged, converted to units of psia, and the
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T. L. McConnell Station Manager I '
result used in determining the thermodynamic properties of the steam.
-McGuire Wuclear Station l
 
P. O. Box 488-
Two of the three channels were unaffected by the error.
 
No use is made of A1072, the direct measurement of steam flow, in TOP.
 
Absent the measurement error introduced by the procedural error of not restoring the computer points, feedwater flow measurements are inherently 1:
more accurate than -the direct steam flow measurements.
 
Hence, for heat balance calculations, equating steam flow to feedwater flow less blowdown flow is more accurate.
 
On July 5,1989, IP/0/A/3007/17 was performed at over 80% RTP.
 
The OAC power points indicated 81.1%, 81.1%, and 85.8% RTP, with primary power indicating the highest.
 
The four as-found PRNI readings ranged from 86.6%
to 87.2%.
The reactor group engineer on duty was contacted, and ha selected the best estimate thermal power point, the same as secondary power, as the basis for the PRNI calibrations.
 
He later stated that, when contacted, he was heavily involved with a test on Unit 2, and the over four percent difference in power indications'did not register in his mind.
 
His reason for selecting secondary power as the basis for calibration was that secondary power is inherently more accurate than primary power above 50% RTP.
 
This was the same engineer that had performed PT/0/A/4150/03 for Unit 1, cycle 6 in January 1989.
 
As a result of the recalibrations, the PRNI indications were reduced to a range of 80.9% to 81.1% with secondary power indicating 80.8% RTP and primary thermal power at about 85.5% RTP.
 
This recalibration was completed at about 8:30 a.m.
 
l By about 11i30 a.m. on July 5, power had been increased to approximately 95%
RTP as indicated by best estimate / secondary power.
 
Power was held at that level by the operators until about 1:30 p.m. ; since plant electrical output had reached previous 100% power levels and to investigate low suction pressure on the condensate booster pumps.
 
The operators discussed the anomalous relationship between thermal and electrical power, but ascribed it to lower lake temperature, lower turbine back pressure, and reduced use E
of auxiliary steam.
 
Secondary power was then increased to an indicated 96% RTP, but further increase was halted because of continued low suction l
pressure.
 
The operators stated that power was not limited by condenser i
performance, but that they wanted it performing to expectations before
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increasing power further.
 
At about 5:50 p.m. on July 5, another reactor group engineer was consulted about the discrepancy among thermal power indicators.
 
He initiated a Thermal Output Calculation Dump from the OAC, and that printout clearly indicated the three locked-out sensors.
 
Upon removing the lock outs, thermal power indication increased to over 100% RTP.
 
The operators immediately reduced power to 98.2% as indicated by best estimate / secondary thermal power and 98.7% as indicated by primary thermal power.
 
At that time all PRNIs indir.ated 94%.
The PRNIs were then recalibrates to best estimate thermal power using IP/0/A/3007/17.
 
Once discovered, the overpower operation was reported to the NRC promptly, and the licensee initiated a broad, interdepartmental review of the event,
 
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' which culminated in an abnormal piant event' meeting on July 10, 1989.
Nuclear Production Department L
 
Duke Power Company P. 0.' Box 33189
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This ' meeting' was ' attended by. the inspectors.
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Charlotte, NC 28242
Each involved department -
g operations, IAE, and performance / reactor group.- appeared to have made a thorough evaluation of its' performance and had proposed corrective action.
 
The licensee's evaluation of this event is continuing and will be reported
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The inspectors' evaluation is addressed in.the following paragraphs:
c Paragraph 3 establishes the validity of both the licensee's computer
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. analyzing plant data to determine plant thermal power for both
.. steady-state and slowly changing power levels.
 
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Paragraph 4' addresses the application of TPDWR2 to historical plant data for the period of 11:00 am to 5:30 pm on July 5,- 1989 and-the conclusions drawn.from that analysis.
 
Paragraph 5 addresses other observations and findings pertinent to the July 5, 1989 event.
 
3.
 
Independent Analysis of Thermal Power (61706)
a.
 
References (1) NUREG-1167, TPDWR2: Thermal Power Determination for Westinghouse Reactors, Version 2, (2) McGuire Nuclear Station FSAR, Chapter 5, (3) Westinghouse. Technical Manual 1440-C247, Pressurizer Instre;-
tions...,
(4) Westinghouse Technical Manual 1440-C250, Vertical Steam Genera-tor Instructions...,
(5) 0AC Manual:
:(a) Thermal Outputs Calculation, Section 3.2.10, and (b) Thermal Outputs Calculation Dump, Section 3.2.14.
 
b.
 
Parameter and Data Acquisition The micro computer program, TPDWR2, developed by the NRC's Indepen-dent Measurements Program for analysis of licensee thermal power data is described in Reference (1).
 
In order to customize the program for use at McGuire, plant specific physical and performance parameters were obtained from' references (2) to (5),
Those parameters are given on page 1 of. Attachment 1 alonc with typical input data for the calculations described below.
 
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On July 7, 1989, the_ inspector used the OAC to log the input data for TPDWR2.
. A. V. Carr, Esq.


Computer logging provides better numerical resolution and contemporaneousness of the data than manual collection from MCB indicators could provide.
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Duke Power Company
All of the necessary data were obtained using edits from four computer point identification tables estab-lished with the help of a licensee engineer.
 
The points were logged
~at five minute intervals for eight hours, and the tables were printed out after all data had been logged.
 
For the first 33 intervals the reactor was at a nominal 100% RTP.
 
Over the next three hours, the unit under went a slow power reduction to about 64% RTP.
 
By the end of data collection, the unit had recovered to 70% RTP Most of the data were not in a form that could be used directly for input to TPDWR2.
 
Data sources, with the loop A computer points used in the examples of loop-specific parameters, and the required manipulations are described below:
S/G pressure = pressure (psig, A1107) + atmospheric pressure (P0117)
FW flow (Mlb/hr) = average (P1412, P1413)
FW temperature = A0454 (no manipulation necessary)
BD flow (gpm, S/G conditions) = A0652(lbfir) * 0.002514 S/G 1evel(inches) = A1059(%) * 2.33 + 394 LD flow = A0764 (no manipulation necessary)
LD temperature = A1088 (loop C cold leg)
CHG flow (gpm) = A0758 - 32gpm (flow to the seals does not return enthalpy from the regenerative heat exchanger)
CHG temperature = A0758 (regenerative heat exchanger outlet temperature)
PZR pressure (psia) = P1389 PZR level (inches) = A0976(%) * 5.205 + 25.75 NC average temperature = P1461 (no manipulation required)
NC average cold leg temperature = average (A1064, A1076, A1088, A1100)
A SUPERCALC3 spreadsheet was used to perform all of the necessary calculations and to organize the results in an order best suited for input to TPDWR2.
 
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Inithe points. identified above, the A prefix refers to a directly read analog point, an unmanipulated value.. The P prefix refers to a-calculated. point.
 
In the case of: feedwater. flow, program FLO takes-
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the - square root of the basic ' analog measurement of' differential pressure in inches of water, multipliek the root by orifice calibra-
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i 422 South Church Street Charlotte, NC 28242
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auxiliary feedwater nozzles.. The resulting flow rate 'in mass per-unit ~ time is displayed at the appropriate P point. The feedwater
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flows, in units of millions.of pounds 1 mass per hour,.used in these -
calculations' were.the' instantaneous' values - calculated by the OACL program FLO every thirty seconds.
 
The licensee's TOP program uses two minute averages'of the FLO output.
 
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TPDWR2 calculational. Results TPDWR2 can analyze. single or paired sets of data.
In the paired-set-mode, it can account _ for energy stored"in cr transferred-from the pressurizer and-steam generators from the net change in mais invento-ry.. The inspector selected seven paired sets of data (- 14 power
, sets) for analysis with the pairs separated by 15 to 25 minutes.-
The comparison between TPDWR2 and the -licensee's calculations of power was very good, as shown in the following table, which was arranged from the highest to the lowest power calculated-with TPDWR2.
McGUIRE 1: Heat Balance Comparison for 7 July 1989 THERMAL POWER TOP.
TPDWR2 S/G-C TOP-TPD TIME (Mwt)
(%)
(Mwt)
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(Mwt)
(%)
(Mwt) (%)
1755 3416.0 100.15 3422.0 100.32 860.1 25.13-6.0
.18 1640 3401.1 99.71 3405.4 99.84 851.0 24.99-4.3
.13 1655 3405.0 99.83 3402.4 99.75 860.4 25.29 2.6.08 1540 3396.3 99.57 3382.9 99.18 843.8 24.94 13.4.39 1525 3407.3 99.89 3374.1 98.92 839.1 24.87 33.2.97 1820 3353.9. 98.33 3347.4 98.14 839.1 25.07 6.5.19 1920 2703.5 79.26 2677.0 78.48 670.3 25.04 26.5.98 1935 2534.1 74.29 2522.0 73.94 641."
25.42 12.1.48 2320 2397.8 70.30- 2392.9 70.15 601.0 25.12 4.9.21 2305 2320.0 68.02 2306.9 67.63 585.0 25.36 13.1.57 2050 2203.5 64.60 2200.0 64.50 560.1 25.46 3.5.16 2150 2191.8 64.26 2200.0 64.50 560.5 25.48-8.2
.37 2035 2192.3 64.27 2189.9 64.20 555.1 25.35 2.4.11 2205 2202.2 64.56 2189.0 64.17 552.7 25.25 13.2.60 AVERAGE =
25.20 The mean absolute difference between results was 0.39% of the TOP value.
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The goodf agreement between TOP and TPDWR2 calculations over a range of powers from 64% to 100% RTP, confirms that. TOP is properly sam-pling-the, process - variables and ~ performing correct calculations of the thermodynamic properties of water and steam.
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Neither the differences in magnitudei nor percent of reference (the TOP. result)
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Similarly,. steam.
. generator C made.a consistent contribution to total power, an average
'of 25.2%,.regardless.of-power.
 
Most of the random variations between.
 
-calculations probably come-from the input' values of feedwater flow.
 
The. 0AC samples all variables with-ten-second periodicity, and.the inputs to.TPDWR2 were' all from the ten-second snapshots recorded at five minute intervals.
 
The feedwater flow input to TOP is a two-min-uteiaverage of the ten-second snapshots.
 
This smooths the basically noisy flow measurement.
 
A small consistent difference between results. may come from the calculation of blowdown enthalpy.
 
TPDWR2 uses an average of steam generator saturation conditions and feed-water ' conditions to calculate enthalpy for bottom BD flow. ' TOP uses saturation conditions-in the steam generator.
 
Typical results for TPDWR2 are given on pages 2 and 3 of Attachment 1.
 
The licensee's calculational method is acceptable as programmed.
 
The differences noted reflect reasonable variations in engineering judgement.
 
4.
 
Analysis of. Historical Data for July 5,1989 Selected plant data from those monitored by the OAC with ten-second periodicity are transferred with five minute periodicity, on a snapshot basis without averaging, to a remote minicomputer.
 
Those data are then retained for one ween.
 
The licensee recovered the Unit I data for the period of potential over power operation for their analysis, and provided the inspectors copies of the recovered data in the form of five ASCII
- files on computer disks.
 
The files included three steam pressure measure-ments per steam generator, two feedwater temperatures per generator, one feedwater flow for generators A and B, and two feedwater flows for genera-tors C and D.
 
The feedwater flows recorded were the two minute averages.
 
-Of course, for steam generator C one steam pressure and one feedwater flow were invalid-because of the locked-out points.
 
Other recorded data pertinent. to the calculation of thermal power included: one charging temperature, one charging flow, one cold leg (letdown) temperature, letdown flow, and barometric pressure.
 
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To make a rapid first assessment of this mass of data, a correlation was made between total feedwater flow and thermal power.
 
The data for the correlation were obtained from the thermal power analysis performed on Unit I using the plant data obtained on July 7,1989 and the results of the corresponding analyses using TPDWR2.
 
The correlation was performed using a least-squares spreadsheet and SUPERCALC3.
 
Expressed algebraical-i ly, the correlation was:
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Power (Mwth) = 238 Mwth + 207.9 * (Total feedwater flow (Mlbm/hr)).
The correlation coefficient was 1.00.
Another spreadsheet was set up to apply this correlation to the sum of all valid.feedwater flow measurements for all five-minute-interval data captured between 1:55 p.m. and 6:10 p.m. on July 5, 1989.
This analysis identified one time, 2:20 pm, at which thermal power exceeded 102% of RTP.
TPDWR2 was then used to analyze the data for that time.
Since historical records of blowdown flow rates, steam generator and pressurizer levels, and NC temperatures were not available to the inspectors; nominal values of these parameters, from the analysis for 4:55 p.m. on July 7, 1989, were used.
With the exception of-the assumptions of blowdown flow rates, use of these nominal values for a single set of performance data has no effect on the results from TPDWR2.
Since blowdown flow has only a small effect on the calculation; the differences between actual and assumed blowdown flow rates are expected to introduce negligible error into the calcula-tion.
The input and output for the 2:20 p.m. calculation are give in Attachment 2.
In that calculation, core thermal power was determined to be 102.3% of RTP, and the contribution of steam generator C to the total was 25.2%, which was in go.od agreement with the results of the calcula-tions using plant data collected on July 7, 1989.
Analysis of the thermal power calculations from the power-to-flow correla-tion showed that during the period of 1:55 p.m. to 4:45 p.m. Unit 1 averaged over 101% RTP for the entire period.
Extended operation in excess of 101%
RTP as well as any operation in excess 102% RTP is considered to be a violation of the license limit.
Although the event was identified by the licensee, the identification was not made at the earliest opportunity.
That opportunity came when the reactor group engineer was requested to evaluate a difference greater than 2% between primary and secondary thermal power calculations on the OAC with the reactor in the 80% power range early on July 5,1989.
Identification of the locked out values in the OAC at that time would have precluded any overpower operation from the failure to properly complete IP/0/A/3001/01C on July 2,1989.
Hence this event has been identified as a violation of the license limit of a maximum core power of 3411 thermal megawatts given in License NPF-9, Paragraph 2.C(1) (VIO 50-369/89-21-01).
The licensee does not have as-found values for the high flux trip set-points on July 5,1989.
A licensee engineer stated that setpoint drift from the 109% calibration value required by Technical Specification 2.2 is uncommon.
Using the the PRNI readings of S4% and the best estimate thermal power of 98.2% RTP observed at 5:50 pm on July 5,1989, a the high flux trip would have occurred at (98.2/94.0) * 109% = 114% RTP.
This is less than the maximum overpower trip setpoint of 118% used in the FSAR Chapter 15 safety analyses of reactivity transients.
(See FSAR Table 15.1.4-1.)
The effect of the PRNI calibration error on positive and negative flux rate trips was judged by the inspector to be too small to require quantitative analysis.
For slow power transients, the ove. -
power-delta-temperature trip would have functioned at 108.8% RTP according I
. _ _ _ _ - _ _. _ _ _ _ _. _. _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _. _ _.. _ _ _ _ _ _ _ _ _ _.. _ _ _.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _.
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,
E J. Michael McGarry, III, Esq.


That trip function was'not affected by either the thermal power or the PRNI calibration errors.
[
 
Bishop, Cook, Purcell and Reynolds
No additional violations or deviations were identified in this inspection
,
. area.
 
.5.
 
Other-Inspection Findings and 0 observations-During NRC Inspection 50-370/87-42, TPDWR2 was used with data obtained from Unit 2.
 
One observation reported was that the reactor coolant pump efficiency used in TOP, at' that time, was' the lowest -the inspector had ever observed and.that the licensee might be. incurring a power production penalty - from an over-conservative calculation of core thermal power. The licensee subsequently responded by telephone that an error in pump effi-ciency did' exist 'in TOP and that four units at McGuire and Catawba had-each.been penalized about 1 Mwe.
 
At the start of this inspection, the inspector requested an up-to-date copy of the TOP program description.
 
The copy provided had the revised pump power calculation on page 3.2.10.5, but the latest revision date shown on the page was 4/1/85.
 
Review of the OAC Manual in the ' computer room revealed that page 3.2.10.5 had not been updated and showed the old pump heat calculation.
 
The licensee stated that TOP had actually been revised, and that is substantiated by the agreement between TOP and TPDWR2 results for July 7,1989.
 
The TOP program description is well written and the flow of ' operations from data input' to analysis to output is relatively easy to follow. The same observation is not true of the FLO program description in any aspect.
 
Both programs are,' effectively, part of required surveillance procedures, but FLO is not auditable.
 
Licensee control and documentation of computer programs used in the performance of required surveillance will be addressed in a later inspection.
 
6.
 
Followup of Previous Violations (92702)
(Closed) Violation 50-369/87-42-01:
Failure to make a required report of overpower operation within the required time.
 
On January 5,1988, the licensee submitted LER' 369/87-35, which was a complete and adequate-description of-the event.
 
The licensee's response to the violation, dated February. 15, 1988, was reviewed in the Region II office and found acceptable.
 
The licensee acknowledged a need to be more thorough in their
- evaluations of potentially reportable events, but no programmatic changes were identified.
 
7.
 
Exit Interview (30703)
The inspection scope and findings were summarized on July 10, 1989, with those persons indicated in paragraph 1 above.
 
The inspector described the areas inspected and discussed in detail the inspection findings.
 
The-licensee was informed at that time that no decision had been reached with
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respect to-the issuance of a notice of violation..No. dissenting comments l
were received from the licensee.
 
The licensee did not identify as propri-etary any of the materials provided to or reviewed by the-inspectors during this inspection.
 
The licensee was informed on August 14, 1989, that a decision had been made by Region II management to issue a violation for overpewer operation.
 
8.
 
Initialisms and Acronyms Used in This Report ASCII American Standard Code for Information Interchange
--
BD
-
blowdown CHG charging
-
FLO.
 
0AC program for calculating feedwater flow
-
FSAR
-
Final Safety Analysis Report FW feedwater
-
gpm gallons per minute
-
IAE Instrumentation and Electrica', Department
-
IP
-
instrument procedure LD
-
letdown LER
-
licensee event report MCB
-
main control board Mlbm/hr
-
million pounds mass per hour Mwe megawatts electrical
-
Mwth megawatts thermal
-
NC nuclear coolant system (reactor coolant system)
-
NIS-
-
nuclear instrument system 0AC operator assist computer, the plant computer
-
PRNI power range nuclear instrument
-
psia pounds per square inch gauge
-
psig
-
pounds per square inch absolute PT periodic test
-
PZR pressurizer
-
RTP
-
rated thermal power TOP thermal outputs program
-
TPS
-
Test Programs Section
- Attachments:
1.
 
Typical Output from TPDWR2 2.
 
Analysis of Overpower Operation on July 5, 1989
___-_-____-_____a
 
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ATTACHMENT 1
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mt umCt un I
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f kEsin 1 7-7-89
1400 L Street, NW j'.
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Mf!Pitstfits:
Washington, D. C.
LBCf0t000laffSYSfit ttfl20fift IISpufl01 i
 
Pu p Pou r ( W uen)
20005
5.2 InsideSerfaceAru(git)
''
15,958 i
Dayne H. Brown, Chief t
Pasp Ifficioner (1)
Radiation Protection Branch
.
92.7 hat Lou Cafficient (IfDsAr og it)
55.00
)
Presseriser Insida Dianeter (inches)
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501HFIEflU1550Mfl0E
I Division of facility Services N. C. Department of Human Resources 701 Barbour. Drive
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Raleigh, NC. 27603-2008
11,575
!
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l.
hoeinsideDiameter(incks)
168.50 Dichess (ineks)
4.0 l
>
linerOstsideDiameter(iseks)
21.00 thereal Coadsetivity (ITUsAr ft f)
0.035
,
. Imbr of liners


bistareCarrr-ever(I)is&
i County Manager of Mecklenburg County
0.070 LICDSD NIBtL POSE (pt)
''
Mll Ecistare Carry-over (I) is B 0.070
.
BeistareCarry-over(1)isC 0.070 bistureCarry-over(1)isD 0.070 NfA:
Sif1 Sif2 Sif1 Sif2 fib 1755 1820
!IS 1755 1820 SftM G B M f0t i Sfte Gtimf0! 8 Steas Presare (pia)
995.2 1004.6 Stas Pnsare (pia)
1002.1 1012.0 feedsater fles (86 lbAr)
3.797 3.670 hedsater flw (16 lbAr)
3.882 3.640 bedsater imprature (f)
438.5 4M.1 fuduter fespratare (f)
(M.6 (M.2 Sarface lloedon (gn)
0.0 0.0 Serface floodas (gn)
0.0 0.0 bite Bloedon (gy)
118.5 120.6 httas Bladoes (gp)
1M.1 133.6 Rater level (iseks)
548.0 548.0 hier Intel (iseks)
545.9 545.2 STIM GIItuf0t C STLE GRtuf0t D StusPressare(psia)
1002.4 1012.7 Stas Presan (pia)
992.6 1002.9 feedsater flos (16 lbAr)
3.822 3.721 feedsater flw (16 lbAr)
3.811 - 3.716 feedsster fesprature (f)
4M.6 (M.4 feeduter fesperatare (f)
437.5 435.4 Surface lloedon (gp)
0.0 0.0 Surface lloedon (gy)
0.0 0.0 l
lottos Bloodon (gp)
110.6 113.6 httos lloedon (gy)
131.4 142.1 hter!stel(inches)
551.3 551.0 later Level (inenes)
545.2 545.9
'
;
i LtD05 LIII CutGIIGLIR flos (gp)
101.6 101.6 flw (gp)
66.7 62.5 fesprature(f)
559.7 559.8 fesperstare(f)
'489.1 499.9 PRISSUt!!B tuCf02 Pressure (psia)
2280.0 2275.0 fare (T)
587.8 587.8
'
'
Nater level (inches)
M3.3 Mt.2 fcold(T)
558.9 559.5
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720 East fourth Street
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Page 2'of 3?
Charlotte, NC 28202
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Dr. John M. Barry i
<* 4
Department of Environmental Health
,?3-DATA GET 10F_2 NfHAIM FIDi POWER POWER
<%
"if 1755 lunars (BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
;(tttt)
STRAtf GENERATOR A i
 
>U Steam 1192.7-
'3.749 4.472
' Feedwater -
417.8'
-3.797-1.586
.I
 
_
W Surface Blair3oun 541.8 0.00000 0.00000-Botta Blowdown 477.5 0.04715 0.02251
,,.
Power Dissipated 2.9078 851.6
' STEAtt GENERATOR B
.p Staae 1192.4 3.828 4.565
: Feedwater
.415.7-3.882-1.614
. Surface Blowdoun 542.9'
O.00000 0.00000 Bottom Bicmdown 476.9 0.05418 0.02584 Pbuer Dissipated 2.9768 871.8-STEati GENERATOR C Steam 1192.4 3.778 4.505
"
Feedsater 415.7-3.822-1.589 Surface Blowdown 542.9 0.00000 0.00000
'
'
. 90ttom Bloudoun 476.9 0.04403 0.02100 Pbuer Dissipated 2.9369 860.1 STEAtt GENERATOR D Steam 1192.8 3.757 4.482
Mecklenburg County 1200 Blythe Boulevard
.Feedsater 416.7-3.811-1.588
,
Surface Blowdown 541.4 0.00000 0.00000 Bottom Blowdown 476.7 0.05231 0.02494 Pbuer Dissipated 2.9186 854.8 ODER OCMPONENTS Intdoun Line 558.6 0.03772 0.02107 M arging Line 475.0-0.02893
--0.01279 Pressurizer 704.8-0.00021-0.00015 Pumps-0.06568 Insulation losses 0.00147 Pbuer Dissipated-0.05608-16.4
,
REACIDR IDER 3422.0
.
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-ATTACHMENT 1 Page 3 of'3
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HEAT BMAum
Charlotte, NC 28203
.
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r Karen E. Long
McGuire 1 7-7-89
!
- DhTA SET 2 OF 2.
Assistant Attorney General
 
;
ENTHALPT FILM RMER PGfER 1820 hours (BTUs/lb)
N. C. Department of Justice
(E6 lb/hr)
(E9 BTUs/hr)
(19tt)
STEAlf GEIERA2OR A
' Steam _
1192.3 3.621 4.318 Feeduater 415.2-3.670-1.524 Surface Bloudoun 543.3 0.00000 0.00000 Botton Bloudoun 476.8 0.04802 0.02289 Power Dissipated 2.8171 825.1 STEAlf GEIERATOR B Steam
'1192.0 3.787 4.514=
Feeduater 413.1-3.840-1.586 Surface Sloudoun 544.4 0.0GD00 0.00000
' Bottom Bloudoun 476.2 0.05323 0.02535 Pouer Discipated 2.9535 865.0 STEAlf GENERATOR C Steam 1192.0 3.676 4.381 Feeduater 413.3-3.721-1.538 Surface Bloudoun 544.5 0.00000 0.00000 Botton Blowdown 476.4 0.04525 0.02156
,
,
Power Dissipated 2.8651 839.1 STEAM GEIERATOR D Steam 1192.4 3.658 4.382 Feeduater 414.4-3.716-1.540 Surface Bloudoun 543.0 0.00000 0.00000 Bottom Bloudoun 476.3 0.05661 0.02896 Power Dissipated 2.8490 834.4 l'
P. 0. Box 629 i
011ER OCRE01ENTS Ietdown Line 559.4 0.03768 0.02106 Qiarging Line 487.4-0.02496-0.01216 Pressurizer 704.2-0.00021-0.00015 Pumps-0.08568 Insulation losses 0.00147 Power Dissipated-0.05544-16.2
Raleigh, NC 27602 i
,
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REACIUR IDER 3347.4
State of North Carolina
.
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D. Hood, NRR Document Control Desk NRC Resident Inspector U.S. Nuclear Regulatory Commission 12700 Hagers Ferry Rd.
.
 
___
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_ _ _.
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ATTACHMENT.2
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184f BALilCI MfA leGnire1
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7-5-89 l
PL&lfPAl&ElftBS:
IIACTORC00LAlfSTSill IITLICfiffIISULAfl01 Pasp Poser ( H each)
5.2 1:sideSurfaceArea(sqft)
15,958
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92.7
:leatLossCoefficient(Bffs/hrsqft)
55.88 PospIfficieser(1)
PressuriserInsideDiase'.er(inches)
84.8 IDillfLICfif t IISULATIDI SftAlCIllilf0tS'
InsideSurfaceAreatsqft)
11,575 DoseInsideDisseter(inches)
168.58 thickness (inches)
4.8 IlserOutsideDiaseter(inches)
21.88 TherealConductivity(BfDs/hrftf)
8.835 Insberoflisers
. 12
 
BolstureCarry-over(1)inA 8.878 LICIISID fill 5AL POWII (But)
3411 HolstoreCarrrover(1)inB 8.878 HolstoreCarrr-over(1)inC 8.878 HolstureCarry-over(1)inD 8.878
. DATA:
-fllt 1428 fil!
1428 SflalGilitif01A Sft&B GIltlAf01 B SteasPressure(psia)
981.9 SteasPressure(pala)
988.8 feedeaterFlos(16lbihr)
3.848 feediaterflos(16It/hr)
3.966 feedsaterfesperature(f)
439.2 feediaterfesperature(F)
437.5 SurfaceBlondova(gps)
8.8 SurfaceBloedorn(spa)
8.8 BottonBlondesn(gps)
111.7 BottonBloedoen(gps)
136.5 NaterLevel(inches)
548.8 NaterLevel(inches)
544.8 I
.
Sf!ARGilllATORC Sf!AH GiftlATOR D SteasPressureipsia)
Hl.5 SteasPressure(psia)
978.3 feedsaterFlor(16lb/hr)
3.895 feedsaterFlos(E6lb/hr)
3.983 feedsaterfesperature(F)
437.5 feedsaterfesperatore(F)
438.2 SurfaceBlondoen(gps)
8.8 SurfaceBloedorn(gps)
8.8
$citosBloedoen(gps)
113.3 BottesBlosdorn(gps)
137.3 L
NaterLevel(inches).


558.6 NaterLevel(inches)
Huntersville, NC 28078-8985-RII: RS II:0RSp RI p
545.5 LifDONILIII CHAIGIIGLIII flos(gps)
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182.4 flov(gps)
57!
fesperature(F)
558.5 fesperature(F)
03.8
. PRIS$H!!Il
!! ACTOR
'
Pressure (psia)
2282.8 fare (f)
588.3 NaterLevel(inches)
343.8
? cold (f)
559.6
.
*
______________________________._____w


, _ _ -,
89 09{ /89
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4-ATTACllMENT 2 Page 2 of 2
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HEAT BALANCE i
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McGuire 1 7-5-89 DATA SET 1 0F 1 ENTHALPY FLOW POWER POWED 1420 hours (BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
(MWt STEAM GENERATOR A Steam 1193.1 3.803 4.538 Feedwater 418.6-3.848-1.610 Surface Blowdown 539.8 0.00000 0.00000 Bottom Blowdown 477.0 0.04446 0.02120
_______
Power Dissipated 2.9484 863.5 STEAM GENERATOR B-i Steam 1192.9 3.911 4.666 i
Feedwater 416.7'
-3.966-1.652 Surface Blowdown 540.7 0.00000 0.00000 Bottom Blowdown 476.4 0.05436 0.02590 y,-
_______
Power Dissipated 3.0393 890.1 STEAM GENERATOR C Steam 1193.2 3.850 4.593 Feedwater 416.7-3.895-1.623 Surface Blowdown 539.7 0.00000 0.00000 Bottom Blowdown 476.0 0.04514 0.02148
_______
Power Dissipated 2.9918 876.2 STEAM GENERATOR D
'
Steam 1193.3 3.849 4.593 Feedwater 417.5-3.903-1.629 Surface Blowdown 539.2 0.00000 0.00000 Bottom Blowdown 476.1 0.05469 0.02604
_______
Power Dissipated 2.9891 875.4 OTHER COMPONENTS Letdown Line 557.8 0.03807 0.02123 Charging Line 479.5-0.02295-0.01100 Pumps-0.06568 Insulation Losses 0.00147
_______
Power Dissipated-0.05398-15.8
______
'
REACTOR POWER 3489.5
.
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}}
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Latest revision as of 18:45, 23 December 2024

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-369/89-21 & 50-370/89-21
ML19351A355
Person / Time
Site: McGuire, Mcguire  Duke Energy icon.png
Issue date: 09/29/1989
From: Julian C
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To: Tucker H
DUKE POWER CO.
References
NUDOCS 8910200110
Download: ML19351A355 (2)
v • d • e


Text

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?

September 29, 1989

1 l

,

).

Duke Power Company E

!

ATTN: 1 Mr. H. B. Tucker, Vice President Nuclear Production Department 422 South Church Street y.

Charlotte, North Carolina 28242 Gentlemen:-

SUBJECT:

REPORT NOS. 50-369/89-21 AND 50-370/89-21

,

'Thank you for your response of September 20, 1989, to our Notice of Violation issued on August 21, 1989, concerning activities conducted at your McGuire

. facility.

We have evaluated your response and found that it meets the-requirements of 10 CFR 2.201.

We will examine the implem(ntation of your L.

. corrective actions during future inspections, i ::,.

ll.

We appreciate your cooperation in this matter.

Sincerely,

.

original signed by

-

'

Caudie A. Julian Caudie A. Julian, Chief

<

'

Engineering Branch Division of Reactor Safety.

,

!=

' cc:

T. L. McConnell Station Manager I '

-McGuire Wuclear Station l

P. O. Box 488-

..

Cornelius, NC 28031 L

J. S. Warren

!-

Nuclear Production Department L

Duke Power Company P. 0.' Box 33189

'

[;

Charlotte, NC 28242

,

L.

. A. V. Carr, Esq.

"

Duke Power Company

<

i 422 South Church Street Charlotte, NC 28242

'

(cc cont'd - see page 2)

.

l 8910200110 890929

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h fdC$

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C S:ptember 29, 1989

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  • D' uke Powet Company

.

,

,

(cc cont'd)

,

E J. Michael McGarry, III, Esq.

[

Bishop, Cook, Purcell and Reynolds

,

,

.

1400 L Street, NW j'.

'

Washington, D. C.

20005

Dayne H. Brown, Chief t

Radiation Protection Branch

'

I Division of facility Services N. C. Department of Human Resources 701 Barbour. Drive

!

!

Raleigh, NC. 27603-2008

!

l.

i County Manager of Mecklenburg County

'

,

720 East fourth Street

'

Charlotte, NC 28202

-

..

S. S. Kilborn, Area Mana er

!

Mid-SouthAreaESSDProfect

!

Westinghouse Electric Corporation

!

MNC West Tower - Bey 239 i

P. O. Box 355

Pittsburgh, PA' 15230 i

Dr. John M. Barry i

Department of Environmental Health

'

Mecklenburg County 1200 Blythe Boulevard

'

Charlotte, NC 28203

.

r Karen E. Long

!

Assistant Attorney General

N. C. Department of Justice

,

P. 0. Box 629 i

Raleigh, NC 27602 i

!

State of North Carolina

!

bec:

D. Hood, NRR Document Control Desk NRC Resident Inspector U.S. Nuclear Regulatory Commission 12700 Hagers Ferry Rd.

Huntersville, NC 28078-8985-RII: RS II:0RSp RI p

er le Sh lock 09$89

89 09{ /89

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