IR 05000413/1987016
| ML20215A848 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/03/1987 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20215A831 | List: |
| References | |
| 50-413-87-16, 50-414-87-16, NUDOCS 8706170111 | |
| Download: ML20215A848 (19) | |
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[># REGg UNITED STATES
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D NUCLEAR REGULATORY COMMISSIOM.
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REGION 18 n
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,j 101 MARIETTA STRE ET, N.W.
ATL ANTA, GEORGI A 3o323
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Report Nos.:
50-413/87-16 and 50-414/87-16'
Licensee: -Duke Power Company 422 South Church Street Charlotte, NC 28242 Docket Nos.:
50-413 and 50-414 License Nos.:
NPF-35 and NPF-52 Facility Name:
Catawba 1 and 2 Inspection Conducted: May 11-15, 1987 Inspector:
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A Z-P~. r Burnett C0 ate Signed-l
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Approved by:
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F. Jape, Sectlon Chief V
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Date Sign &d-Engineering Branch Divistori of Reactor St.fety SUMMARY Scope:
This routine, unannounced inspection addressed the areas of startup test results (Unit 2), core performance surveillance (Unit 2), and thermal power monitoring (both units).
Results: No violations or deviations were identified.
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8706170111 870605 PDR ADOCK 05000413 G
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REPORT DETAILS i
1.
Persons Contacted Licensee Employees
- H. B. Barron, Operations Superintendant R. G._ Blessing, Engineer, Reactor Group S.~W. Brown, Reactor Engineer
- M. A. Cote, Licensing Specialist S. L. Cox, Training Specialist
- T. E. Crawford, Integrated Scheduling Superintendent
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- C. L. Hartzell, Compliance Supervisor
- M. W. Hawes, Engineer, Reactor Group
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- W. H. Miller, Planning L.gineer, Maintenance
- Z. L. Taylor, Test Engineer
- R. F. Wardell, Technical Services Superintendent
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D. A. We11baum, Engineer, Reactor Group Other licensee personnel included engineers and office personnel.
NRC Resident Inspector
- M. S. Lesser, Resident Inspector Other Organizations (by telephone)
K. N. Jabbour, NRR/PADA T. G. Dunning, NRR/HFTS
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on May 15, 1987, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings.
No dissenting comments were received from the licensee. Thi; licensee did not identify as propriatary any. of the materials provided to or reviewed by the inspectors durt,g this _ inspection.
Management made the commitment stated below:
f Inspector follnwup item: Complete startup test procedures TP/2/A/2650/06 and TP/2/A/2100/01 by May 31. 1987 - paragraph 5.
3.
Licensee Action on Previous Enforcement Matters I
This subject was not addressed in the inspectio I t
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4.
Unresolved Items No unresolved items were identified.
5.
Unit 2 Startup Tests at 100% Power (72624)
The following completed or ir progress startup test procedures were reviewed:
a.
TP/2/A/2100/01, Controlling Procedure for Power Escalation, had as its last entry date August 14, 1987, although steps 12.7.3.5, steps 12.3.8 to 12.3.13, step 12.7.4, and step 12.8 have not been signed off.
The large load reduction test will be removed f rom the test program, which will eliminate all or part of two steps.
b.
TP/2/A/2600/14, Nuclear Instrumentation System Initial Calibration, was performed at all power levels and was completed at 100% power on August 14, 1986.
All acceptance criteria were satisfied.
The completed procedure was approved by management on October 25, 1986.
For the power range nuclear instruments, the inspector independently analy:ed the total current linearity with power for each channel using a least-squares fit spreadsheet with the SUPERCALC 3 microcompJter program. In each case, the correlation coefficient was greater than 0.999 and the two-sigma uncertainty in slope was less i
than 2.5%.
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c.
PT/2/A/4150/05, Core Power Distribution, was performed at 89% RTP on August 12, 1986, at 98.8% RTP on August 14, 1986, and at 99.9% RTP on August 27, 1986. In all cases, the maximum enthalpy rise hot channel factor, the heat flux hot channel factor, incore-measured quadrant power tilt ratio, and relative assembly power deviations satisfied the acceptnce criteria and technical specifications.
The last
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measurement listed was performed at 31.9 effective full power days (EFPD).
d.
PT/2/8/4150/03B, Thermal Outputs Reliability Check, was performed to assure the inputs to the operator aid computer (OAC) thermal outputs program were providing reasonable values.
It is a prerequisite to PT/2/A/4150/03A, NSSS Thermal Outputs, and was performed with acceptable frequency throughout the power escalation program.
The determination of acceptability is made by internal comparison of values when there is a multiplicity of parameter measurements as with feedwater temperature and pressure.
For single point measurements, the comparison is with the control room instrument.
There is no direct traceability to the output of the calibrated sensor.
Confirmation of the calibration of the computer points, traceability, will be addressed in a future inspection.
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TP/2/A/2650/06, Unit Loss of Electrical Load, was performed on March 23, 1987, and most data reduction m # ompleted by March 24th.
However, at the time of this inspection, steps 12.13, 12.15, 12.16, and 12.17 had not been signed off, primarily because final review of the data had not been completed. The reactor did trip, although the system design basis is that there be no trip en loss of load.
Discussions with licensee staff revealed that the turbine emergency trip system (ETS) gave a false indication of a turbine trip, which led to a reactor trip because power was still above 69% of rated thermal power (RTP). Later in the transient, the turbine did trip, possibly btcause of the rear. tor trip. All safety acceptance criteria were satisfied.
At the exit interview, the licensee made a commitment to complete procedures TP/2/A/2650/06 and TP/2/A/2100/01 by May 31, 1987. (Inspector followup item:
Complete startup test procedures TP/2/A/2650/06 and TP/2/A/2100/01 by May 31, 1987).
No violations or deviations were identified.
6.
Unit 2 Core Performance Surveillance (61702, 61707)
The following completed surveillance test procedures were reviewed:
a.
PT/2/A/4150/04, Reactivity Anomaly Calculation, must be performed every 31 EFPD to satisfy lechnical Specification 4.1.1.1.2.
A review of the completed procedures provided the following results:
i DATE BURNUP ANOMALY j
PERFORM'!D (EFPD)
(pcm)
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08/27/86 29.6
+32.9 12/04/86 44.7-55.4 12/16/86 55.2-11.4
01/14/86 83.5-92.1
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02/13/87 107.6-194.8 03/19/L 136.8-225.6 Linear extrapolation of these results indicates that the limit of 1000 pcm difference between measured and predicted core reactivity will not be reached before the end of the cycie.
However, a continued linear relationship is not assured.
b.
PT/2/A/4150/05, Core Power Distribution, was performed for routine surveillance on December 9, 1986 at 99.5% RTP (45 EFPD), on January 9, 1987 at 88% RTP (78EFPD), on February 5, 1987 at 99.6% RTP (99.9 EFPD), on March 10, 1987 at 100% RTP (128 EFPD), and on April 20,1987 at 79.8% RTP (149 EFPD).
In all cases, the maximum enthalpy rise hot channel factor, the heat flux hot channel factor,
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the incore-measured quadrant power tilt ratio, and relative assembly f
powers satisfied the acceptance criteria and appropriate technical j
specifications.
The licensee has improved the surveillance by comparing fuel assembly enthalpies between successive power maps.
This provides a basis for a qualitative review' of changes-in power i
distribution.
For example, this review determined that between the-maps at 99.9 and 128 EFPD the maximum chaage in assembly output was a 5.9% increase for the bundle in core position N-2.
During that same j
period, most of the fuel assemblies near the center of the core decreased in output from 1 to 3%.
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PT/2/A/4150/08, Target Flux Difference Calculation, was performed by
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measurement of equilibrium axial flux difference on July 21, 1986,
August 15, 1986, August 22, 1986, December 4,1986, December 30, 1986,-
January 20, 1987, and April 24, 1987. On the February 27, 1987, the procedure was completed by interpolation. The surveillance frequency intervals of Technical - Specifications 4.2.1.3 and 4.2.1.4 were satisfied.
No violations or deviations were identified.
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7.
Thermal Power Determination, Units 1 and 2 (61706)
j The NRC independent measurement program for' determination of reactor thermal power is described in NUREG-1167, TPDWR2: Thermal Power Determina-tion for Westinghouse Reactors, Version 2.
To obtain data for use.with the microcomputer program TPDWR2, the inspector was instructed by licensee j
personnel in the use of a remote terminal of the OAC. Three outputs from
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the Thermal Outputs Dump program (NUCLEAR 28) were obtained on the line printer for each unit. The the time between outputs was 20 to 25 minutes for each unit. Immediately after calling up NUCLEAR 28, a scan was made'
using progran. TECH SPEC 9 to obtain steam generator and pressurizer levels existing at the time of the dump. The level data were recorded by hand.
The data obtained, although sufficient for use in TPDWR2, were not.in the order or, in all cases, in the units required for input to that program. A SUPERCALC 3 spreadsheet was created to facilitate ordering and conversion of the data for input to TPDWR2.
The parameters and data used in TPDWR2 calculations of thermal power are shown in attachment 1 for Unit 1 and attachment 2 for Unit 2.. Typical.
results of the calculations are given in attachment 3 and 4 for Units 1 and 2 respectively.
The differences between the results for the ' programs on the operator -
assist computers (OAC) and TPDWR2 are consistent by unit (see attachment.
5). The Unit 1 typical difference of 0.5% is acceptable. The agreement for Unit 2 is even better, and may stem from one or two sources: There was no blowdown durirg Unit 2 data collection, and there was no need to account for flow through the auxiliary feed water nozzles since that flow
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does pass through the feedwater flow venturis.
It was necessary to correct the feedwater flow input for TPOWR2 to include the flow through the Unit 1 auxiliary feedwater nozzles.
That flow was assumed to be at feedwater conditions.
For both units, insulation surface area and heat transfer coefficient were adjusted to agree with the constant value used in the licensee's calculations.
The licensee's calculation includes an active calculation of pump heat by monitoring individual pump current and bus voltage.
Those data are available from the thermal outputs dump, and were averaged for each unit to obtain a constant pump heat contribution fcr use in TPDWR2.
No violations or deviations were identified.
8.
Non-Conservative High Power Trip Setpoint - Unit 2(92705)
On April 28, 1987, tempering flow to the steam generator auxiliary feedwater nozzles was not taken into account when performing a Unit 2 heat balance, and power was underestimated by about 2% of RTp when the power range nuclear instruments were calibrated against the heat balance.
Consequently, the high power trip setpoints for all four channels exceeded 111% RTP, with the highest setpoint, N44, at 112.3%.
In the Technical Specification equation 2.2-1, where Z + R + S </= TA, the licensee increased Z from the Table 2.2-1 value of 4.56 to 5.96 to account for up to a 5% power mismatch, and still remained under the TA limit of 7.5.
Hence, the licensee concluded that the power range nuclear instruments were OPERABLE throughout the brief (about two hours) event, and that the event was not reportable.
Subsequent to the inspection, the inspector contacted members of the NRR staf f by telephone to obtain their interpretation of equation 2.2-1, and their guidance on the reportability of the event.
They did not agree with or identify a basis for the licensee's increase of the Z term.
In their judgement, Z should remain at the table value.
It was their opinion that the sensor error value, S, given as 0 in table 2.2-1, should be increased by the percent of span the system was in error (112.3 - 109)/1.2 = 2.75).
Then the terms of the equation become:
4.56 + 0.1 + 2.75 = 7.41, which is less than the allowable 7.5. in the worst case.
i No violations or deviations were identified.
ATTACHMENTS:
1.
Heat Balance Data Catawba 1 l
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Heat Balance Data Catawba 2 3.
Heat Galance Data Catawba 1 4.
Heat Balance Data Catawba 2 5.
Catawba Thermal Power Measurements (5/13/87)
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Attachment 1 M Ai SAL M C Dr.TA CATAHA 1 5-13-E7 PLMT WMETEFh FEACTOR LOOLM:T SYSTEM EEFLEC!iVE IN30LAT10N Fm Pc*er (N ea:n)
5.0 Inside Surface Area ts: it)
E2,bl9 Fao Efficien:y (U 95.B hat less Ceef fittent (Blus/br s:; f t)
31 0')
Pressurizer inside Diaeeter linches)
94.0 ND'GEFLECTIVE IN30LAT10N STEi.M SELEFAf0;3 Insice Br face Area.fsa ft)
C lae lest/ 'ineter nr:bes)
163.50 1hickr.ess (in;hes)
0.0 Fiser OLtnce Diante- (te;hes)
20.00 Thereal Ccr.dectivit) (9T&s/hr it F)
0.000 Naber of hisers
a histur e Cury-over m in A 0.350 LICEN5ED THERMAL F0WER id t)
5411 i
histure Ce ry-o.er !D in B 0. 35')
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10 in C 0.350 fain ure Cur rcve r histure Cery-wer (D in 0 0.350 MTA:
EET 1 EET 2 3ET 1 SET 2 I
IldE 1552 1616 TIME 1552 1616 STE M 6ENERATUC A STEM BENERATOR F Sten Fressure !; star 1019.!
1019.5 Steae tresscre (es a)
1021.4 1021.3 f eedwater Flen (E6 lbitr)
3.269 3.E67 Feedwater Flew (E6 lb/hd 3.763 3.779 Fee: mater Tes;erature (F)
438.3 4:3.3 Feedwater Temperature IF)
437.4 437.4 h f ate 51cadm (pt)
0.0 0.:
Erf ace Blenke (pm)
0.0 0.0 httct Bl ad: e S W 108.,
107.6 htten Biovdce (pe)
155.0
!!7.0 hier Level tintnest 474.5 475.0 Water Level tinches)
475.4 475.7 5tF M EENEFA!0; C STEM GENERAT0F D Sten Fressee tusia)
!029.0 1020.0 Eteae Fressare ipsia)
1019.6 10!i.i reedwater F!cw 'E6 lb hr>
3.B:5 3,23E Feed *ater Flea (E6 lb/hr)
3.72)
3.724 Feedaner Te c eratcre (9 05.7 422.5 Feedsater Tncerature (F)
439.1 437.1 Erf a:e Hcudce !:W 0.0 0.*
h<f ate flowho (pe)
0.0 0.0 btt u Elcadce 4;s:
115.0 114.0 httce Bluke (po 116.0 116.0 oter Level Lire es)
473.6 473 :
hter Level bnthes)
474.3 A73.6 LETECW L!hi CHAF6i% LlHE R oa (pt)
76.0 76.o Fitw Qca)
95.1 97,7 Tenerature (F)
56J.4 550.3 Tecerature (F)
57: 0 50s.6 FRES30hIEF EEACTOR'
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Fressure tpsia!
22t5.0 2255.0 1 n e (F)
537,3 559.3 hter Level hnches)
352.0 3hi.:
T tole iF)
560.9 560.8 J
- 1 jnQ 08%g UNITED $TATES
,oo, NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET, N.W., SUITE 2000
ATLANTA, GEORGIA 30323 o,s,
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Attachment 2.
1 WEAT BALMG LATA q
CATA S 2 i
5-13-B7 FLMT EMeETER3:
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REACT 09 CLOL M! 5 HTE'!
REFLECTIVE IGJLATIO1 En: f txer (N eado 5.0 Inside hrface Area (sa 't)
82,618
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i Pm Ef hciency (7.)
90.0 Peat less Ccethcient GTUs/tr sa f t)
35.0)
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Fressct:er Inside DiautE'r (Inches)
64.0-l i
N09 EFLECl!VE ! GJ.ATION ETEM GESEEATO S Irside krface Area (sq ft)
One Inside 0:neter tintnes)
163,5)
hickness (inches)
0.0 lbser Datstde Dineter (aces)
20.00 Then al Con hctivity (BTLs/hr ft F)
0.000
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N A er of Risen
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kistcre Carryear Oi in 4 L 250 LICENEEE TFERMAL FCE; tht)
3411 Mcisture Carry-ever 10 in B 0.250
t'aistere Carry-w er :1) in C 9.250 Mcisture.arr ro er (7.) in 2 0.230 c
D14:
SET 1 EET 2 EET 1
$ET 2 TYE 1501 1620 TIME 1601 1620 ETEM EENEu!0R A STEM GENERATDE B 5te u Pressure estai 10X.2
'! 0( 0. D Sten Pressure tcsia)
1911.1 1010.6 Feedwater Flow tis Ib/rtr)
3.662 3.694 Feedsater Flcd (E$ lb/br>
3.795 3.329
Feecuater Tee:eratve (F)
G5 43.5 Feedwater Tetceraire (F)
433.9 433.5 l
Erf ace Elcadw: t;p!
9.0 0,0 hrf1:e El u b e (g m 0.0 0.0 Scita 91ew ban i;:e:
0.0
.o b itte Blowdce (;u)
0,0 0.0 hater Len! HMeu 317.3 511.!
Water Level Unches)
515.7 516.1 ETE.M :ENERATER C STEM SENE9109 0 sim Fressve tula)
10:8.1 16.1 Sten f ressure esia)
5H.5 996.2 Feednate P!te (E$ Itthr)
- 716 3.7!3 Feedwater Fic= (Et It!hr)
3.751 3.772 feednater Te m rait e (C
- 3.1 4:7,9 Feed 4ater T m erttu*e (F)
437.9 437.9 Su f a:e Ei:+1er icx!
0(
kriace Eltwdce (geel 0.0 0.0
a b! a !!ame hm
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httos Elenke (gn)
0.0 0.0 6tr tevel honesi 517.0 517.0 Water Level br. chest 515.4 516.2
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LETEC+1 LINE 3AR5 h3 L!bE Fl:o (;n!
78.7 72.2 Flu yW 93.4 E9.4 Tegeratun (F)
561.0 561,0 Tunrature (F)
519.2 521.1 NEE %RIZE; REACTCR Dressere asie.
2253.0 2253.b I ave (F)
SEi.1 5s9,0 kater Len! Or.chasi 375.3 375.0 T ccid (F)
561.5 561.5 I
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UNITED STATES 4y
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NUCLEAR REGULATORY COMMISSION
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Attachment 3 I
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HEAT BALANCE CATAWBA 1 5-13-87 DATA SET 1 OF 2 ENTHALPY FLOW POWER POWER 1552 hours0.018 days <br />0.431 hours <br />0.00257 weeks <br />5.90536e-4 months <br /> (BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
LMWL)
i STEAM GENERATOR A Steam 1189.9 3.826 4.552 Feedwater 417.6-3.869-1.616 Surface Blowdown 545,6 0.00000 0.00000 Bottom Blowdown 479.1 0.04291 0.02056
Power Dissipated 2.9571 866.1 l
STEAM GENERATOR B Steam-1189.9 3.717 4.423 Feedwater 416.6-3.763-1.568 Surface Blowdown 545.8 0.00000 0.00000 Bottom Blowdown 478.7 0.04571 0.02188
Power Dissipated 2.6768 842.6 STEAM GENERATOR C Steam 1189.9 3.789 4.509 Feedwater 418.0-3.835-1.603 Surface B3 owdown 545.6 0.00000 0.00000 Bottom Blowdown 479.4 0.04568 0.021'90
Power Dissipated 2.9278 857.5 STEAM GENERATOR D Steam 1189.9 3.675 4.373 F eed we '. er 418.5-3.720-1.557 Surface Blowdown 545.6 0.00000 0.00000 Bottom Blowdown 479.6 0.04607 0.02210
Power Dissipated 2.8382 831.2 OTHER COMPONENTS Letdown Line 560.2 0.0281e 0.01577 Charging Line 496.8-0.03764-0.01870-Pressurizer 703.1-0.00006-0.00004 Pumps-0.05879'
Insulation Losses 0.00289
Power Dissipated-0.05887-17.2-3
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REACTOR POWER 3380.1 l
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UNITED STATES
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Attachment'3
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HEAT BALANCE CATAWBA 1 5-13-87
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DATA SET 2 OF 2 ENTHALPY FLOW POWER POWER 1616 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.14888e-4 months <br /> (BTUs/lb)'
(E6 lb/hr)
(E9 BTUs/hr)
(MWt)
STEAM GENERATOR A Steam 1189.9 3.824 4.551 Feedwater 417.6-3.867-1.615 Surface Blowdown 545.5 0.00000 0.00000 Bottom Bl owdown 479.1 0.04252 0.02037
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Power Dissipated 2.9559 865.7 j
i STEAM GENERATOR B Steam 1189.9 3.732 4.441 Feedwater 416.6-3.779-1.575 Surface Blowdown.
545.8 0.00000 0.00000 Bottom Blowdown 478.7 0.04651 0.02226
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j Fower Dissipated 2.8890 846.1 FTEAM GENERATOR C Steam 1189.9 3.793 4.513 Feedwater 417.8-3.838-1.604 l
Surface Blowdown 545.6 0.00000 0.00000 Bottom Blowdown 479.3 0.04529 0.02171
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Power Dissipated 2.9314 858.5
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Steam 1189.9 3.679 4.378 Feedwater 418.5-3.724-1.558 Surface Blowdown 545.6 0.00000 0.00000 Bottom Blowdown 479.6 0.04607 0.02210
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Power Dissipated 2.8413 832.1 OTHER COMPONENTS Letdown Line 560.1 0.02816 0.01577 Chargina Line 495.1-0.03873-0.01918 Pressurizer 703.1-0.00006-0.00004 Pumps-0.05879 Insulation Losses 0.00289
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F"wer Dissipated-0.05935-17.4
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REACTOR POWER 338..
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NUCLEAR REGULATORY COMMISSION '
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Attachment 4 HEAT BALANCE CATAWBA 2 5-13-87 DATA SET 1 OF 2 ENTHALPY FLOW POWER POWER 1601 hours0.0185 days <br />0.445 hours <br />0.00265 weeks <br />6.091805e-4 months <br /> (BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
(MWt)
STEAM GENERATOR A Steam 1191.3 3.662 4.363 Feedwater 415.6-3.662-1.522 Surface Blowdown 542.6 0.00000 0.00000 Bottom Blowdown 476.7 0.00000 0.00000
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Power Dissipated 2.8408 832.0 STEAM GENERATOR B Steam 1190.9 3.795 4.519 Feedwater 414.8-3.795-1.574 Surface Blowdown 544.3 0.00000 0.00000 Bottom Blowdown 477.1 0.00000 0.00000
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Power Dissipated 2.9445 862.4
STEAM GENERATOR C i
Steam 1191.3 3.716 4.427 Feedwater 417.4-3.716-1.551'
)
Surface Bl owdown 542.6 0.00000 0.00000 i
Bottom Blowdown 477.6 0.00000 0.00000
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Power Dissioated 2.8762 842.4-
)
i STEAM GENERATOR D i
Steam 1191.4 3.780 4.504
)
Feedwater 417.1-3,781-1.577 Surface Blowdown 542.0 0.00000 0.00000 Battam Blowdown 477.3 0.00000 0.00000
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Power Dissipated 2.9263 857.0 OTHER COMPONENTS Letdown Line 561.0 0.02913 0.01634
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Charging Line 509.9-0.03650-0.01861 Pressurizer 701.7-0.00011-0.00008 Pumps-0.06109 Insulation Losses 0.00289
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Power Dissipated-0.06055-17.7
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REACTOR POWER 3376.0 l
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-Attachment 4
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HEAT BALANCE CATAWBA 2 5-13-87 FA SET 2 OF 2 ENTHALPY FLOW POWER POWER i
lo2O hours (BTUs/lb)
(E6 lb/hr)
(E9 BTUs/hr)
(MWt)
STEAM GENERATOR A Steam 1191.3 3.694 4.401 Feedwater 415.6-3.694-1.535 Surface Blowdown 542.6 0.00000 0.00000 Bottom Blowdown 476.7 0.00000 0.00000
_______
I Power Disuipated 2.8659 839.3 STEAM GENERATOR B Steam 1190.9 3.829 4.559 L
I Feedwater 414.5-3.829-1.587 Surface Blowdown 544.2 0.00000 0.00000 Bottom B1owdown 476.9 O.00000 O.00000
_______
Power ~ Dissipated 2.9722 870.5 STEAM GENERATOR C Steam 1191.3 3.713 4.423 j
Feedwater 417.1-3.713-1.549 i
Surface Blowdown 542.6 0.00000 0.00000 Bottom Blowdown 477.5 0.00000 0.00000
-.______
Power Dissipated 2.8742 841.8 STEAM GENERAlOR D Steam 1191.4 3.777 4.500 Feedwater 417.1-3.778-1.576 Surface B1owdown 542.O O.00000 O.00000 Bottom Bl owdown 477.2 0.00000 0.00000
_______
Power Di ssi pated 2.9240 856.4 i
OTHER COMPONENTS Letdown Line 561.0 0.02894 0.01623 Charging Line 512.1-0.03486-0.01785 Pressurizer 701.7-0.00011-0.00008 Pumps-0.06109
Insul a ti on Losses 0.00289
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i Power Dissipated-0.05990-17.5 l
______
j REACTOR POWER 3390.4 i
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L Attachment 5, Report 413/414/87-16 CATAWBA THERMAL PCWER MEASUREMENTS 5-13-87
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RESULTS (MWth)
Difference
UNIT TIME DAC*
TPDWR2 (%)
3o
_________________________________________________
1552 3397.8 3380.1
.52
1616 3402.8 3384.9xx
.53
<
1635 3401.2 3383.4
.52
2 1601 3380.6 3376.0
.14
1620 3394.8 3391.2**
.1 1,
,
1639 3383.2 3380.1
.09
- OAC = Operator Assist Computer l
- Average of two calculations.
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