ML20076G431
| ML20076G431 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 10/13/1994 |
| From: | Rehn D DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9410190321 | |
| Download: ML20076G431 (28) | |
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' l DukePbwerCornpany D.L Rtw L
'l Catauba,hclear Generation Department Voce hesident
'4800 Concord Road (803)8313205 Office
' York,SC29745 '
(803)8313426 Fax
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,' ~ DUKEPOWER.
October 13,1994 l
i U. S. Nuclear Regulatory Commission Attention: Document Control Desk.
Washington, D. C. 20555 i
Subject:
Catawba Nuclear Station, Unit 2 Docket No. 50-414 j
Special Report l
Unit 2 Cycle 7 Startup Report
' Pursuant to Catawba Technical Specification 6.9.1 please find attached the Stanup Report l
for Unit 2 Cycle 7 core design.
i i
t Very truly yours, i
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/g D. L. Rehn J
i DT/
Attachment j
xc: Mr. S. D. Ebneter Regional Administrator,RegionII i
Mr. Robert E. Manin, Project Manager, ONRR Mr. R. J. Freudenberger NRC Resident Inspector Catawba Nuclear Station i
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.i U. S. Nuclear Regulatory Commission -
- j October 13,1994 Page 2
.i bxc: Z. L. Taylor J.R. Fox-NSRB Staff A. V. Carr J. E. Snyder J. E. Burchfield NCMPA-1 l
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Duke Power Company Catawba Nuclear Station Unit 2 Cycle 7 STARTUP REPORT October,1994
~
TABLE OF CONTENTS ER99 List of Tables li List of Figures il 1
1.0 '
Introduction...
2.0 Precritical Testing
.2 2.1 Total Core Reloading..
2 2.2 Preliminary NIS Calibration
.2 2.3 Reactor Coolant System Dilution
.2 3.0 Zero Power Physics Testing
.6 3.1 1/M Approach to Criticality
.6 3.2 Source Range / intermediate Range Overlap Data 9
3.3 Point of Nuclear Heat Addition 9
3.4 Reactivity Computer Cbeckout 10 3.5 ARO Boron Endpoint Measurement.
10 3.6 ARO lsothermal Temperature Coefficient Measurement.
11 3.7 Reference Bank Worth Measurement by Dilution.
. 11 3.8 Reference Bank in Boron Endpoint Measurement 13 3.9 Differential Boron Worth Determination..
.. 13 3.10 Control Rod Worth Measurement by Rod Swap
. 13 4.0 Power Escalation Testing
. 15 4.1 Core Power Distribution 15 4.2 One-Point incore/Excore Calibration.
19 4.3 Reactor Coolant Delta Temperature Measurement
. 20 4.4 Hot Full Power Critical Boron Concentration Measurement 21 4.5 Calorimetric Reactor Coolant Flow Measurement 21 4.6 Incore/Excore Calibration 23 1
i i
e LIST OF TABLES Paae 1.
Core Design Data 1
2.
Preliminary NIS Calibration Data
.4 3.
Summary of Zero Power Physics Testing Results..
.7 4.
Source Range / Intermediate Range Overlap Data 9
5.
Nuclear Heat Determination.
10 6.
Reactivity Computer Checkout 10 7.
ITC Measurement Results 11 8.
Control Rod Worth Measurement Data
.. 14 9.
Core Power Distribution Results. 30% Power
. 16 10.
Core Power Distribution Results,75% Power 17 11.
Core Power Distribution Results,100% Power 18 12.
One-Point incore/Excore Calibration Results 19 13.
Reactor Coolant Detta Temperature Data.
20 14.
Calorimetric Reactor Coolant Flow Measurement Data 22 15 incore/Excore Calibration Results 23 LIST OF FIGURES Page 1.
Core Loading Pattem
.3 2.
ICRR vs. Demin Water Added During Reactor Coolant System Dilution 5
3.
Inverse Count Rate Ratio vs. Control Rod Worth During Approach to Criticality 8
4.
Integral and Differential Worth of Reference Bank.
12 ii i
4 Page 1 of 23
1.0 INTRODUCTION
Catawba Unit Two Cycle 7 includes a feed batch of 88 MkBW fue: assemblies manufactured by B&W Fuel Company (BWFC). The f eed batch includes 48 f uel assemblies with axial blankets that were manuf actured by BWFC for Trojan. The blankets consist of 0.71% w/o enrichment in the top and bottom 6 inches of the fuel rods. Of these 48 fuel assemblies 40 were enriched to 4.00% (w/o) and 8 to 3.60 % (w/o). The remaining 40 f uel assemblies were un-blanketed MkBW f uel assemblies enriched to 3.50% (w/o). Bumable poison rod assemblies used in the feed batch were also manufactured by BWFC.
1 Catawba Unit Two Cycle 7 core loading began at 1257 on May 28,1994 and ended at 0237 on May 31, 1994. Initial criticality for cycle 7 occurred at 2359 on July 2,1994, Zero Power Physics Testing ended j
at 2255 on March 31,1993. The unit reached full power at 1005 on July 10,1994. Power escalation testing, including testing at full power, was completed by July 21,1994.
Table 1 contains some important characteristics of the Catawba 2 Cycle 7 core design.
TABLE 1 C2C7 CORE DESIGN DATA 1.
C2C6 end of cycle bumup: 380 EFPD 2.
C2C7 design length: 430 10 EFPD Region Fuel Type Number of Enrichment,
- Loading, Cycles Assemblies w/o U
- MTU" Bumed 7A (S)
OFA 29 3.75 12.2815 2
8A (T)
MkBW 76 3.75 34.6788 1
9A (U)
MkBW 40 4.00/0.71*
18.2520 0
t 9B(U)
MkBW 8
3.60/0.71*
3.6504 0
9C(U)
MkBW 40 3.50 18.2520 0
Totals 193 87.1147 Natural U blanketed fuel assemblies (0.71 w/o enrichment - 6 inches top and bottom)
Design MTU loadings which were used in all design calculaations.
}
Page 2 of 23 2.0 PRECRITICAL TESTING Precritical testing includes:
+ core loading
+ preliminary calibration of nuclear instrumentation
+ dilution of reactor coolant system to estimated critical boron concentration Sections 2.1 through 2.3 describe results of precritical testing for Cetawba 2 Cycle 7.
2.1 Total Core Reloading The cycle 7 core was loaded under the direction of PT/0/N4150/22, Total Core Reload. Plots of Inverse Count Rate Ratio (ICRR) versus number of fuel assemblies loaded were kept for each applicable source range and boron dilution mitigation system (BDMS) channel.
Core loading began at 1257 on May 28,1994 and concluded at 0237 on May 31,1994. Core loading was verified by PT/0/N4150/03C, Core Verification, which was completed by 0830 on May 31,1994.
Figure 1 shows the core loading pattem for Catawba 2 Cycle 7.
2.2 Preliminary NIS Calibration Periodic test procedure PT/0/N4600/05E, Preliminary NIS Calibration, is performed before initial criticality for each new f uel cycle. Intermediate range reactor trip and rod stop setpoints are adjusted using measured power distribution f rom the previous fuel cycle and predicted power distribution for the upcoming f uel cycle.
Power range full power currents are similarly adjusted. Intermediate range calibration data is checked and revised as necessary during power escalation insting.
Table 2 shows the calibration data calculated by PT/0/N4600/05E. Calculations were performed on June 13,1994 and June 14,1994. Calibrations were complete by July 2,1994.
2.3 Reactor Coolant System Dilution The reactor coolant system boron concentration was diluted from the ref ueling boron concentration to the estimated critical boron concentration per PT/0/N4150/19A, NCS Dilution with Shutdown Banks Inserted.
Inverse Count Rate Ratio (ICRR) was plotted versus gallons of domineralized water added.
Initial reactor coolant boron concentration was 2201 ppmB. The estimated critical boron concentration was calculated to be 1710 ppmB. The calculated volume of demineralized water required was 15340 gallons.
This change in boron concentration was expected to decrease ICRR from 1.0 to 0.6.
Reactor coolant system dilution began at 1120 on July 2,1994 and concluded at 1712 on July 2,1K4.
The final reactor coolant system boron concentration, after allowing system to mix, was 1697 ppmB.
Dilution was temporarily suspended after the addition of 10313 gallor.s demineralized water when the ICRR went outside (higher than predicted) the predicted ICRR behavior band. This ICRR behavior was conservative since it indicated that the reactivity of the core was being changed less than expected per gallon of domineralized water being added. Dilution was resumed after a procedure change was made to allow dilution to continue if the ICRR is more conservative than the predicted ICRR band. Figure 2 shows ICRR versus volume of water used.
1 Page 3 of 24 FIGURE 1 North CORE LOADING PATTERN, CATAWBA UNIT 2 CYCLE 7 A
I 180 FUELTRANSFER CANAL -->
T64 US42 S10 US1 S29 US44 T62 g____________
PD PD PD PD PD PD PD S49 T70 U63 T21 U66 T31 U76 T42 U77 T34 S72 2------
PD RCCA BP RCCA BP RCCA BP RCCA BP RCCA PD 3----
S27 T59 U50 T33 U85 T71 U56 TSO U71 T30 U72 T25 S68 PD PD BP RCCA BP RCCA BP RCCA BP RCCA BP PD PD T69 U68 T61 US03 T66 US15 S16 US18 T14 US28 T17 U69 T20 4----
RCCA BP RCCA BP PD BP RCCA BP PD BP RCCA BP RCCA 5-T60 U62 T38 US29 S59 USO2 T04 US37 T73 US08 S25 US23 T44 U65 T12 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP SS US45 T08 U49 T37 US24 T53 US20 T75 US21 TOS US14 T56 U75 T39 US46 g_
PD RCCA BP PD BP RCCA BP RCCA BP RCCA BP PD BP RCCA PD 7-S02 U61 T41 US17 T01 US01 S63 US27 S01 US13 T46 US10 T54 U58 S20 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA 12 BP PD 8-U82 T51 U80 S24 US34 T22 US32 S36 US26 T11 US40 S33 U74 T68 U88 ED PD RCCA BP RCCA BP RCCA BP RCCA BP RCCA BP RCCA BP RCCA PD 270 g_
576 U73 T72 US25 T45 US22 S26 US31 S44 USO9 T26 US36 T35 U70 S62 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 10 -
US47 T67 U79 T65 US07 T4B US16 T16.
US30 T47 US19 T07 U57 T23 US41 PD RCCA BP PD BP RCCA BP RCCA BP RCCA BP PD BP RCCA PD 11 -
T29 U84 T63 US33 S06 US35 T15 US39 T52 USO4 S40 US38 T74 USS T19 SS BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 12 - - - -
T76 U64 T32 USOS T09 US11 S75 US06 T10 US12 T13 U78 T28 I
I RCCA BP RCCA BP PD BP RCCA BP BP BP RCCA BP RCCA i
13 - - - -
SS1 T06 U67 T49 U81 T18 UB3 T43 U87 T24 U52 T40 S64 I
I PD PD BP RCCA BP RCCA BP RCCA BP RCCA CP PD PD 1
14 -- - - - - -
SS4 T03 U86 T55 U53 T58 U59 T57 U54 T36 S30 l
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i PD RCCA BP RCCA BP RCCA BP RCCA BP RCCA PD 1
1 15-----------
T02 US48 S12 U60 S66 US43 T27 I
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Page 4 of 23 TABLE 2 PRELIMINARY NIS CALIBRATION DATA Intermediate Range Ratio Cycle 6 Cycle 6 BOC 7 BOC 7 Channel (BOC 7 +
Reactor Trip Rod Stop Reactor Trip Rod Stop Cycle 6)
- Setpoint, Setpoint,
- Setpoint, Setpoint, pAmps pAmps pAmps pAmps N35 1.009 81 65 82 66 N36 1.008 70 56 71 57 1
i Power Range Ratio Axial Cycle 6 Full Power Cur-BOC 7 Full Power Cur-Channel (BOC 7 +
Offset, %
rent, pAmps rent, pAmps Cycle 6)
Upper Lower Upper Lower
+20 386.6 294.4 367.7 280.0 N41 0.951 0
335.3 344.6 318.9 327.7
-20 284.1 394.6 270.2 375.2
+20 304.4 237.8 293.4 229.2 N42 0.964 0
265.8 278.6 256.2 268.5
-20 227.3 319.2 219.1 307.6
+20 349.9 2C9.5 327.4 252.2 N43 0.936 0
300.6 313.9 281.3 293.7
-20 251.5 358.3 235.3 335.3
+20 335 6 262.3 313.0 244.6 N44 0.933 0
291.0 305.0 271.4 284.4
-20 246.2 347.7 229.6 324.2
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Page 6 of 23 3.0 ZERO POWER PHYSICS TESTING Zero Power Physics Testing (ZPPT) is performed at the beginning of each cycle and is controlled by PT/0/A/4150/01, Contro ling Procedure for Startup Physics Testing. Test measurements are made below the point of nuclear heat using the output of one power range detector connected to a reactivity computer.
Measurements are compared to predicted data to verify core design. The following tests / measurements are included in the ZPPT program:
+1/M Approach to Criticality
+ Measurement of point of adding heat
+ Reactivity computer checkout
+All Rods Out critical boron concentration measurement
+All Rods Out isothermal temperature coefficient measurement
. Measurement of reference bank worth by dilution
+ Reference bank in critical boron concentration measurement
+Ditferential boron worth determination
+ Control rod worths measurement by Rod Swap Zero power physics testing for Catawba 2 Cycle 7 began at 2148 on July 2,1994 with the beginning of rod withdrawal for approach to criticality. ZPPT ended at 0030 on July 4,1994 following analysis of rod swap data. Table 3 summarizes results from ZPPT. All acceptance criteria were met.
Sections 3.1 through 3.10 describe ZPPT measurements and results.
3.1 1/M Approach to Criticality initial criticality for Catawba 2 Cycle 7 was achieved per PT/0/A/4150/19,1/M Approach to Criticality. In this procedure, Estimated Critical Rod Position (ECP) is calculated based on latest available reactor coolant boron concentration. Control rods, beginning with shutdown banks in norrnal sequence, are withdrawn until Boron Dilution Mitigation System (BDMS) count rate doubles. Inverse Count Rate Ratio (ICRR) is plotted for each source range and BDMS channel. ICRR data is used to project critical rod position. If projected critical rod position is acceptable, rod withdrawal may continue.
Rod withdrawal for the approach to criticality began at 2148 on July 2,1994. Criticality was achieved at 0015 on July 3,1994 with Control Bank D at 117 steps withdrawn.
Figure 3 shows the ICRR plots that were used during the approach to criticality. All acceptance criteria of PT/0/A/4150/19 were met.
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Page 7 of 23 TABLE 3
SUMMARY
OF ZPPT RESULTS PREDICTED VAL-PARAMETER MEASURED VALUE UE/ ACCEPTANCE CRITERIA Nuclear Heat 3.0 x 10# amps (N44)
N/A.
ZPPT Test Band 10* to 10# amps (N44)
N/A ARO Critical Boron 1749 ppmB 1776 2 50 ppmB AROITC
-3.07 pcm/ F
-3.36 2 pcm/*F ARO MTC
-1.181 pcm/*F
-1.49 pcm/'F Reference Bank (Control Bank 1009 pcm 9511143 pcm C) Worth Ref. Bank in Crltical Boron 1619 ppmB 1645 ppmB Differential Boron Worth
-7.76 pcm/ppmB
-7.2611.27 pcm/ppmB Control Bank D Worth 602 pcm 560 200 pcm Control Bank B Worth 717 pcm 645 200 pcm Control Bank A Worth 366 pcm 3731200 pcm Shutdown Bank E Worth 466 pcm 442 200 pcm Shutdown Bank D Worth 464 pcm 430 2 200 pcm Shutdown Bank C Worth 459 pcm 425 200 pcm 3
Shutdown Bank B Worth 1001 pcm 917 t 301 pcm Shutdown Bank A Worth 315 pcra 280 200 pcm Total Rod Worth 5401 pcm 5027
- 503 pcm i
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Page 9 of 23 3.2 Source Range / intermediate Range Overlap Data During the initial approach to criticality, source range and intermediate range data was obtained to verify that at least one decade of overlap existed. If one decade of overlap did not exist, intermediate range componsation voltage would have been adjusted to provide the overlap.
Overlap data for Cycle 7 was obtained per PT/0,A/4150/01, Controlling procedure for Startup Physics Testing, on July 3,1994. Table 4 contains the overlap data. The acceptance criterion was met.
TABLE 4 SOURCE RANGE / INTERMEDIATE RANGE OVERLAP DATA SOURCE RANGE INTERMEDIATE RANGE N31, cps N32, cps N35, amps N36, amps INITIAL DATA:
NIS Meters 220 190 1 x 10 '"
1 x 10 "
OAC 216 173 1.092 x 10~"
1.065 x 10'"
FINAL DATA:
NIS Meters 11,000 8,000 1 x 10 "*
1 x 10 "'
OAC 12,600 9,896 1.055 x 10 "
1.065 x 10 "
3.3 Point of Nuclear Heat Addition The point of nuclear heat addition is measured by trending reactor coolant system temperature, pressurizer level, flux level, and reactivity while slowly increasing reactor power. A slow, constant startup rate is initiated by rod withdrawal. An increase in reactor coolant system temperature and/or pressurizer level accompanied by a change in reactivity and/or rate of flux increase indicates the addition of nuclear heat.
The measurement is repeated to ensure confidence in results.
For Cycle 7, the point of nuclear heat addition was determined per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing, on July 3,1994. Table 5 summarizes the data obtained.
The zero power physics test band was set at 10~8 to 10# amps on power range channel N41 (connected to reactivity computer). This test band provided more than a factor of two margin to nuclear heat for zero power physics testing. Acceptance criterion was satisfied.
l Page 10 of 23 TABLE 5 NUCLEAR HEAT DETERMINATION Reactivity Computer Intermediate Range Intermediate Range (N44), amps Channel N35, amps Channel N36, amps RUN #1 4.7 x 10~7 1.384 x 10~7 1.236 x 10~7 RUN #2 3.0 x 10 3.458 x 10
3.381 x 10~7 7
1 3.4 Reactivity Computer Checkout The reactivity computer checkout was performed per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing, to verify that the power range channel connected to the reactivity computer can provide reliable reactivity data. Reactivity insertions of approximately +25,-30 and +40 pcm are made. The period is measured and used to determine the theoretical reactivity. The measured reactivity for each case is compared to the theoretical reactivity and verified to be within 4.0%.
The checkout was performed for Cycle 7 on July 3,1994. The +40 pcm test was repeated since the reactivity addition was needed to build up flux. Table 6 lists the results of the 4 reactivity insertions. The acceptance criterion was met in all 4 cases.
TABLE 6 REACTIVITY COMPUTER CHECKOUT Period, seconds Theoretical Reac-Measured Reac-Absolute Error, Percent Error,%
tivity, pcm tivity, pcm pcm 154.39 40.73 39.69 1.04 2.56 309.57 22.15 21.64 0.51 2.30
-292.71
-29.81
-29.34 0.47 1.58 162.31 39.05 38.65 0.40 1.02 3.5 ARO Boron Endpoint Measurement This test is performed at the beginning of each cycle to verify that measured and predicted total core reactivity are consistent. The test is performed near the all rods out (ARO) configuration. Reactor coolant system boron samples are obtained while control bank D is pulled to the fully withdrawn position. The reactivity difference from criticality to the ARO configuration is measured and converted to an equivalent boron worth using the predicted differential boron worth. The average measured boron concentration is adjusted accordingly to obtain the ARO critical boron concentration.
i The Cycle 7 beginning of cycle, hot zero power, all rods out, critical boron concentration was measured on July 3,1994 per PT/0/A/4150/10, Boron Endpoint measurement. The measured boron concentration (average of 5 samples) was 1747 ppmB. This value was adjusted by 2 ppmB to yield an ARO concentration of 1749 ppmB. Predicted ARO critical boron concentration was 1776 ppmB. The acceptance criterion, measured boron within 50 ppmB of predicted, was met.
1
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1 Page 11 of 23 3.6 ARO Isothermal Temperature Coefficient Measurement The all rods out (ARO) isothermal temperature coefficient (lTC) is measured at the beginning of each cycle l
to verify consistency with predicted value. In addition, the moderator temperature coefficient (MTC) is obtained by subtracting the doppler temperature coefficient from the ITC. The MTC is used to ensure compliance with Technical Specification limits.
To measure the ITC, a slow (< 20*F/ hour) reactor coolant system cooldown is initiated while trending reactivity versus temperature on an X-Y plotter. When sufficient data (approximately 5 'F) is obtained, a heatup is performed while again trending reactivity versus temperaturef The slopes of the Reactivity versus temperature lines are used to determine the ITC. The cooldown/heatup cycle is repeated if additional data' is required.
l The beginning of cycle 7 ITC was measured por PT/0/A/4150/12A, Isothermal Coefficient of Reactivity Measurement, on July 3,1994. No additional cooldown/heatup cycles were required because of good agreement between the heatup and cooldown results. Table 7 summarizes the data obtained during the measurement.
l Average ITC was -3.07 pcm/*F. Predicted ITC was -3.36 pcm/*F. Measured ITC was within acceptance criterion cf predicted ITC 2 pcm/'F.
The MTC was determined to be -1.18 pcm/*F. This value was used with procedure PT/0/A/4150/21,.
Temporary Rod Withdrawal Limits Determination, to ensure that MTC would remain within Technical Specification limits at all power levels. No rod withdrawal limits were required.
TABLE 7 ITC MEASUREMENT RESULTS ITC, pcm/*F j
AT,'F Ap,pcm T,, ' F 557 'F Cooldown
-5
+13.7 554.5
-2.74 -
-3.04 -
Heatup
+5
-14.0 554.5
-2.80
-3.10 l
Average:
-3.07 I
3.7 Reference Bank Worth Measurement by Dilution i
The control rod bank predicted to have the highest worth is designated the reference bank and is measured by inserting the bank (with all other rod banks fully withdrawn) in discrete steps while slowly diluting the reactor coolant boron concentration. The reactivity worths of the discrete steps of rod insertion are
~
measured using the reactivity computer and summed to obtain the integral worth of the reference bank.
The beginning of cycle 7 reference bank (Control Bank C) worth was measured on July 3,1994 per i
PT/0/A/4150/11 A, Control Rod Worth Measurement by Boration/ Dilution. Figure 4 shows integral worth of reference bank versus bank position. The reference bank was measured to be worth 1009 pcm; predicted worth was 951 pcm. The acceptance criterion, measured worth within
- 15% of predicted, was met.
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Page 12 of 24 FIGURE 4 INTEGRAL AND DIFFERENTIAL WORTH OF REFERENCE BANK 1000 10 900 9
800 8
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Page 13 of 23 3.8 Reference Bank in Boron Endpoint Measurement i
This test is performed at the beginning of each cycle to measure the critical boron concentration with the reference bank fully inserted and all other control rod banks fully withdrawn. The measured boron concentration is used with the measured ARO critical boron concentration and the measured worth of the reference bank to calculate the differential boron worth. Reactor coolant system boron samples are obtained while control rods are inserted or withdrawn to the " Reference Bank in" configuration. The j
reactivity difference from criticality to the " Reference Bank in" configuration is measured and converted to an equivalent boron worth using the predicted differential boron worth. The average measured boron concentration is adjusted accordingly to obtain the "Aeference Bank in" critical boron concentration.
The Cycle 7 beginning of cycle, hot zero power, reference bank in, critical bcron concentration was measured on July 3,1994 per PT/0/A/4150/10, Boron Endpoint measurement. The measured boron concentration (average of 5 samples) was 1617 ppmB. This value was adjusted by 2 ppmB to yield a
" Reference Bank in" concentration of 1619 ppmB. Predicted " Reference Bank in" critical bcron concentration was 1645 ppmB. There is no quantitative acceptance criteria directly associated with this test.
3.9 Differential Boron Worth Determination The differential boron worth is calculated from the measured ARO critical boron concentration, Reference Bank in critical boron concentration, and total reactivity worth of reference bank. The calculated value is compared to predicted value to verify consistency. This calculation also provides an indirect check of measured reference bank worth and of the boron endpoint measurements.
The beginning of Cycle 7, hot zero power differential boron worth was calculated to be 7.76 pcm/ppmB per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. The predicted value was -7.26 pcm/ppmB. The acceptance criterion (measured within i 15% of predicted), was met.
3.10 Control Rod Worth Measurement by Rod Swap The worths of all control rod banks except the reference bank are measured by inserting each bank while withdrawing the ref erence bank and/or previously measured bank to maintain near critical conditions. When the bank being measured is fully inserted, the ref erence bank is positioned to achieve critical conditions with all other rod banks fully withdrawn. The worth of the fully inserted bank is determined from the critical position of the reference bank. The measured worth is compared to predicted worth to verify consistency.
The sum of the worths of all banks, including tha referer.ce bank, is also compared to predicted total.
The beginning of cycle 7 rod worth nreasurement by rod swap was performed on July 3,1994 per PT/0/A/4150/11B, Control Rod Worth Measurement by Rod Swap. Table 8 summarizes the results. All acceptance criteria were met.
TABLE 8 CONTROL ROD WORTH MEASUREMENT DATA Adjusted Cntcal Remahng Difference
% Diff. (Pred -
Reference Positon of Worth of Ref.
Measured Predeted (Predcted -
Meas)/Pred x Acceptance Bank Bank Worth Ref. Bank Bank Alpha Worth, pcm Worth, pcm Measured) 100 Cntena Control C 951 2 143
('*f l**)
N/A N/A N/A N/A 1009 951
-58.0
-6.1 (15%)
284 200 Shutdown A 1027.0 90 643 1.108 314.6 284
-30.6
-10.8 pcm I
3731200 Contr x A 1026.3 77 734 0.899 366.4 373 6.6 1.8 pcm 4251200 Shutdown C 1025.6 108 520 1.090 459.0 425
-34.0
-8.0 pcm 430t200 Shutdown D 1024.9 109 513 1.094 463.6 430
-33.6
-7.8 pcm 4421200 Shutdown E 1024.1 97 594 0.940 465.5 442
-23.5
-5.3 pcm 560t200 Control D 1023.4 129 382 1.103 602.1 560
-42.1
-7.5 pcm 6451200 Control B 1022.7 144 294 1.042 716.9 645
-71.9
-11.1 pcm 917 275 Shutdown B 1022.0 209 22 0.835 1003.9 917
-86.9
-9.5 (30%)
>5027-Total 5401.1 5027
-374.1
-7.4 503 pcm
9
~
l Page 15 of 23 l
4.0 POWER ESCALATION TESTING Power escalation testing is performed during the initial power increase to f ull power for each cyc!9 and is controlled by PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. Tests are perf ormed f rom 0% through 100% power with major testing plateaus at 30%,65%, and 100% power.
Significant tests performed during power escalation are:
+ Core Power Distribution (at 30%,65%, and 100% power)
- One-Point incore/Excore Calibration (at 30% power)
+ Reactor Coolant Delta Temperature Measurement (at 65% and 100% power)
+ Hot Full Power Critical Boron Concentration Measurement (at 100% power)
+ Reactor Coolant Flow Measurement by Precision Calorimetric (at 100% power)
+1ncore/Excore Calibration (at 100% power) in addition to the tests listed abovs, PT/0/A/4150/01 performs checks on the incore detector system, on-line thermal power program, intermediate range setpoints, etc. The results of these checks are not included in this report.
Power escalation testing for Catawba 2 Cycle 7 began on July 4,1994. Full power was reached on July 9,1994. On July 9,1994 the Reactor was manually tripped at 55% F. P. after a runback from 100% due to loss of CFPT 2A.. Full power was again achieved on July 15,1994. Full power testing was completed on July 21,1994.
Sections 4.1 through 4.6 describe the significant tests performed during power escalation and their results.
4.1 Core Power Distribution Core power distribution measurements are performed during power escalation at low power (approximately 30%), intermediate power (approximately 65%), and full power. Measurements are made to verify flux symmetry and to verify core peaking factors are within limits. Data obtained during this test are also used to check calibration of power range channels and to calibrate them if required (see sections 4.2 and 4.6).
Measurements are made using the moveable incore detector system and analyzed using Duke Power's CORE and MONITOR codes (adapted from Shangstrom Nuclear Associates' CORE package and BWFC's MONITOR code, respectively).
The Catawba 2 Cycle 7 core power distnbution measurements were performed on July 5, 1994 (30 %
power), July 7,1994 (65% power), and July 17,1994 (100% power). Table 9 through 11 summarize the results. All acceptance criteria were met.
i j
l 1
f Page 16 of 23 TABLE 9 CORE POWER DISTRIBUTION RESULTS 30% POWER Plant Data Map ID:
FCM/2/07/01 Date of Map:
July 5,1994 Cycle Bumup:
0.166 EFPD Power Level:
29.64 %
Control Rod Position:
Control Bank D at 218 steps withdrawn Reactor Coolant System Boron Concentration:
1550 ppmB ~
i CORE Results
-l Core Average Axial Offset:
12.504 %
Tilt Ratios for Entire Core Height: Quadrant 1:
0.99657 Quadrant 2:
1.00250 i
Quadrant 3:
0.99176 Quadrant 4:
1.00916 Maximum Fo (nuclear):
1.987 f
Maximum F,(nuclear):
1.481 I
Maximum Error between Pred. and Meas F,:
7.57 %
Average Error between Pred, and Meas. F,:
2.67 %
Maximum Error between Expected and 5.80 %
t Measured Detector Response:
[
RMS of Errors between Expected and Measured 3.4%
Detector Response:
MONITOR Results Minimum F Operational Margin:
71.26 %
o Minimum Fo RPS Margin:
25.64 %
Minimum F LCO Margin:
53.42 %
[
o t
Minimum F, Surveillance Margin:
42.41 %
Minimum F, LCO Margin:
20.29 %
i
4 Page 17 of 23 TABLE 10 CORE POWER DISTRIBUYlON RESULTS 65% POWER Plant Data Map ID:
FCM/2/07/02 Date of Map:
July 7,1994 Cycle Bumup:
0.963 EFPD Power Level:
63.67 %
r Control Rod Position:
Control Bank D at 215 steps withdrawn t
Reactor Coolant System Boron Concentration:
1380ppmB CORE Results Core Average Axial Offset:
4.272 %
Tilt Ratios for Entire Core Height: Quadrant 1:
1.00221 Quadrant 2:
0.99825
'f Quadrant 3:
0.99277 Quadrant 4:
1.00676 Maximum F (nuclear):
1.794 t
o Maximum F,(nuclear):
1.414 Maximum Error between Pred and Meas F,:
6.0%
Average Error between Pred. and Meas. F,:
1.40%
Maximum Error between Expected and 6.10%
{
Measured Detector Response:
RMS of Errors between Expected and Measured 2.00 %
Detector Response:
MONITOR Results Minimum Fo Operational Margin:
32.30%
Minimum F RPS Margin:
24.29 %
o Minimum Fo LCO Margin:
46.73 %
l Minimum F, Surveillance Margin:
35.33 %
Minimum F, LCO Margin:
20.04 %
i l
Page 18 of 23 TABLE 11 CORE POWER DISTRIBUTION RESULTS 100% POWER Plant Data l
Map ID:
FCM/2/07/03 Date of Map:
July 17,1994 Cycle Bumup:
7.852 EFPD Power Level:
99.79%
Control Rod Position:
Control Bank D at 212 steps withdrawn Reacior Coolant System Boron Concentration:
1149 ppmB CORE Results Core Average Axial Offset:
-3.616 %
Tilt Ratios for Entire Core Height: Quadrant 1:
0.99972 Quadrant 2:
1.00379 Quadrant 3:
0.98695 Quadrant 4:
1.00954 Maximum Fo (nuclear):
1.791 Maximum F g (nuclear):
1.416 o
Maximum Error between Pred and Meas F,s:
3.MY.
Average Error between Pred. and Meas. Fw 1.13%
Maximum Error between Expected and 3.4%
Measured Detector Response:
RMS of Errors between Expected and Measured 1.5%
Detector Response:
MONITOR Results l
Minimum F Operational Margin:
2.66 %
o Minimum Fo RPS Margin:
22.20 %
Minimum Fa LCO Margin:
16.69%
Minimum F s Surveillance Margin:
5.21 %
3 Minimum F s WO Margin:
M. UY.
3
Page 19 of 23 4.2 One-Point incora/Excore Calibration PT/0/A/4600/05D, One-Point incore/Excore Calibration, is performed using results of power range data taken at 30% power and the incore axial offset measured at 30%. Power ranges are calibrated before exceeding 50% in order to have valid indications of axial flux difference and quadrant power tilt ratio for subsequent power increase. The calibration is checked at 65% power. If necessary, power ranges are calibrated again per PT/0/A/4600/05D or PT/0/A/4600/05A, incore/Excore Calibration.
Data for Catawba 2 Cycle 7 was obtained on July 5,1994 and all power range calibrations were completed on July 6,1994. Results are listed in Table 12. All acceptance criteria were met.
TABLE 12 ONE-POINT INCORE/EXCORE CAllBRATION RESULTS Reactor Pow 9r = 29.64%
Axial Offset = 12.504%
Measured Power Range Currents, pAmps N41 N42 N43 N44 Upper 80.7 65.3 74.8 68.0 Lower 72.8 61.0 70.2 63.5 Ratio, Extrapolated (from measured) Currents to " Expected" (from last calibration) Currents N41 N42 N43 N44 Upper 0.8456 0.8457 0.8391 0.8290 Lower 0.8900 0.8911 0.8919 0.8782 New Calibration Currents, pAmps Axial N41 N42 N43 N44
- Offset, Upper Lower Upper Lower Upper Lower Upper Lower
+20 286.5 230.9 231.3 193.4 266.4 223.1 241.5 201.9 0
248.5 270.2 202.0 226.5 228.9 259.8 209.4 234.8
-20 210.6 309.5 172.7 259.6 191.5 296.6 177.2 267.7
Page 20 of 23 4.3 Reactor Coolant Loop Delta Temperature Measurement Reactor coolant system (NC) hot leg and cold leg temperature data is obtained between 50% and 80%
power and at 100% power per PT/0/A/4600/26, NC Temperature Calibration, to ensure that f ull power delta temperature constants (ATo) are valid. AT is used in the overpower and overtemperature delta temperature o
reactor protection functions.
PT/0/A/4600/26 was performed at 64% power on 7/7/94 and at 100% power on 7/17/94 No adjustments were required based on results obtained at 64% power. Loops A, B, and D were calibrr4ed using f ull power results. Table 13 summarizes the test results.
TABLE 13 REACTOR COOLANT DELTA TEMPERATURE DATA Reactor Power = 63.96%
Loop A Loop B Loop C Loop D Meas. T y, *F 594.5 596.0 595.1 596.0 m
Meas. Tcan, *F 556.2 556.8 556.6 557.2 Calc. Ah, BTU /lb 50.50 51.87 50.91 51.39 Calc. Aho, BTU /lb 78.96 81.09 79.59 80.35 Calc. ATo, *F 57.0 58.4 57.4 57.9 Current ATo, 'F 56.8 58.8 57,9 56.6 Diff erence, 'F
+0.2
-0.4
-0.5
+ 1.3 Reactor Power = 99.89%
Loop A Loop B Loop C Loop D Meas. T y, *F 617.9 620.6 618.6 620.6 m
Meas. Temn, *F 560.1 560.8 560.6 561.6 Calc. Ah, BTU /lb 79.94 83.26 80.42 82.19 Calc. Aho, BTU /lb 80.03 83.35 80.51 82.28 Calc. ATo. *F 57.9 59.9 58.1 59.0 Current ATo, *F 56.8 58.8 57.9 56.6 Ditference, *F
+1.1
+ 1.1
+0.2
+2.4
1 Page 21 of 23 4.4 Hot Full Power Critical Boron Concentration Measurement The hot full power critical boron concentration is measured using PT/0/A/4150/04, Reactivity Anomaly Calculation. Reactor coolant boron concentration is measured (average of three samples) with reactor at essentially all rods out, hot full power, equilibrium xenon conditions. The measured boron is corrected for any off-reference condition (e.g. inserted rod worth, temperature error, difference from equilibrium xenon) and compared to predicted value.
For the purposes of Stanup Physics testing, the predicted critical boron concentration is adjusted for the difference between predicted and measured critical boron concentration measured at zero power. The difference between measured boron concentration and adjusted predicted value is used to compare to acceptance criterion (tSO ppmB).
For Catawba 2 Cycle 7, the hot full power critical boron concentration was measured on July 20,1994. The measured critical boron concentration was 1152.8 ppmB. Predicted critical boron concentration was 1158.4 ppmB; when adjusted for difference at zero power, the adjusted predicted critical boron concentration was 1131 ppmB. The difference between measured and adjusted predicted critical boron concentration was -
21.7 ppmB, which met the acceptance criterion.
4.5 Calorimetric Reactor Coolant Flow Measurement Reactor coolant flow is measured using a precision calorimetric based on secondary side parameters (f eedwater flow, feedwater temperature, steam pressure) with reactor coolant temperature data. Pressure drop data for each of the reactor coolant elbow taps is also obtained. Measured reactor coolant flow is used with pressure drop data to obtain a correction factor for each elbow tap to convert pressure drop to mass and volumetric flow rate. Reactor coolant flow as measured by elbow taps is used to perform Technical Specification surveillances on reactor coolant flow and in the primary side calculation of thermal l
power.
For Catawba 2 Cycle 7, the calorimetric flow measurement was performed on July 20,1994, per PT/2/A/4150/13B, Calorimetric Reactor Coolant Flow Measurement. Three test runs were performed; the average of the results was used for comparison to Technical Specification flow limit and for elbow tap correction factors. Table 14 summarizes the results. All acceptance criteria were met, i
I
\\
Page 22 of 23 TABLE 14 CALORIMETRIC REACTOR COOLANT FLOW MEASUREMENT DATA Test Run Reactor Calculated Reactor Coolant Flow Rate, gpm
% of Tech Loop A Loop B Loop C Loop D Total 1
99.460 95,031 96,843 98,677 98,413 388,963 101.024 2
99.104 95,018 96,824 98,648 98,372 388,862 101.003 3
99.005 95,037 96,788 98,611 98,414 388,850 101.000 Average of total reactor coolant flow for the throe test runs is 388,892 gpm.
Reactor Coolant Elbow Tap Correction Factor Flow Channel Loop A Loop B Loop C Loop D 1
0.298220 0.297966 0.309416 0.296364 2
0.286113 0.281457 0.291395 0.294453 3
0.294320 0.297773 0.298145 0.295479
a Pags 23 of 23 4.6 incore/Excore Calibration Excore power range channels are calibrated at f ull power per PT/0/A/4600/05A, incore/Excore Calibration.
Incore data (flux maps) and power range currents are obtained at various axial power distribution. A least squares fit of the output of each detector (upper and lower chambers) as a function of measured incore axial offset is determined. The slopes and intercepts of the fit for the upper and lower chamber for each channel are used to determine calibration data for that channel.
This test was performed for Catawba 2 Cycle 7 on July 18 and July 19,1994. All power range calibrations were completed on July 21,1994. Eight flux maps, with axial offset ranging from -10.333% to +1.726%
were used. Table 15 summarizes the results. All acceptance criteria were met.
TABLE 15 INCORE/EXCORE CALIBRATION RESULTS Full Power Currents, Microamps Axial N41 N42 N43 N44
- Offset, Upper Lower Upper Lower Upper Lower Upper Lower r
+20%
331.5 259.2 268.4 217.5 313.2 251.3 282.0 227.8 0%
291.5 305.2 236.8 253.8 271.5 292.7 246.2 264.8
-20%
251.5 351.1 205.3 290.0 229.7 334.1 210.4 301.8 Correction (M;) Factors N41 N42 N43 N44 1.389 1.450 1.355 1.403 4