Text

Cycle 2 Startup Rept
	ML20150E058
Person / Time
Site:	Catawba
Issue date:	06/30/1988
From:	Tucker H; DUKE POWER CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	NUDOCS 8807140302
	Download: ML20150E058 (60)
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DUKE POWER COMPANY CATAWBA NUCLEAR STATION UNIT 2 CYCLE 2

, STARTUP REPORT JUNE, 1988 8807140302 ADOCKBB Og g4 .

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l TABLE OF CONTENTS T-9

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List of Tables ................................................. 11 List of Figures ................................................ iii 1.0 Introduction .............................................. 1 2.0 Precritical Testing ....................................... 2 2.1 Total Core Reloading - PT/2/A/4150/22 ............... 3 2.2 Post-Refueling NIS Realignment -

PT/2/A/4600/05E .... 5 2.3 1/M Approach To Criticality - PT/2/A/4150/19 . . . . . . . . 7 3.0 Zero Power Physics Testing ................................ 11 l 3.1 Boron Endpoint Measurement - PT/2/A/4150/10 ......... 15 -

1 3.2 Isothermal Temperature Coefficient of Reactivity j Measurement - PT/ 2/A/4150/12A . . . . . . . . . . . . . . . . . . . . . . . 16 3.3 Control Rod Worth Measurement by Boration/ Dilution -

PT/ 2 / A / 415 0/11 A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.4 Control Rod Worth Measurement by Rod Swap -

PT/2/A/4150/11B ..................................... 24 i

4.0 Power Escalation Testing .................................. 26 1

4.1 Post Refueling: Incore and NIS Recalibration -

PT/2/A/4600/05F ..................................... 38 4.2 NSSS Thermal Outputs - PT/2/A/4150/03A .............. 41 4.3 Reactivity Anomaly Calculation - PT/2/A/4150/04 ..... 42 4.4 Target Flux Difference Calculation - FT/2/A/4150/08 . 43 4.5 Core Power Distribution - PT/2/ A/4150/05 . . . . . . . . . . . . 44 4.6 Incore and Nuclear Instrumentation System Calibration Check - PT/2/ A/4600/05 B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 4.7 Incore and NI System Correlation Check -

PT/2/A/4600/05A ..................................... 50 4.8 Calorimetric Reactor Coolant Flow Measurement -

PT/2/A/4150/13B ..................................... 53 i

LIST OF TABLES Page

1. Source Range / Intermediate Range Overlap Data 12 )

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2. Nuclear Heat Determination 13 j

3. Reactivity Computer Checkout 14

4. Isothermal Temperature Coefficient Measurement Results 17 l

5. Control Rod Worth Measurement by Rod Swap Data 25 l

6. Core Pow'er Distribution Results, 30% Full Power 28 l

7. Core Power Distribution Results, 80% Full Power 32

8. Intermediate Range / Power Range NIS Overlap Data 36

9. Extrapolated Delta Temperatures to 100% F.P. 37 j

10. Quarter Lore Flux Map Data For Post Refueling: -

Incore and NIS Recalibration 39

11. Core Pcver Distribution Results, 100% Full Power 46

12. Flux Map Data For Incore and NI System Correlation Check 51

13. Calculation of Average NC Elbow Tap Flow Coefficients 54 e

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LIST OF FIGURES

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1. Core Ioading Pattern, Catawba Unit 2 Cycle 2 4

2. Assemblies Selected For Calculating NIS Preliminary Setpoints 6

3. 1/M Approach to Criticality - ICRR vs Control Bank Position 9 1

4. Isothermal Temperature Coefficient of Reactivity 1 Measurement Data 18 1

5. Control Bank C Integral Rod Worth curve, Measured by i Dilution 22

6. Control Bank C Differential Rod Worth Curve, Measured by Dilution 23

7. Power Distribution Factors and Comparison to Tech Specs <

for Fq - 30% F.P. 29 ;

8. Measured Fuel Assembly F - 30% F.P. 30 AH ;

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9. Relative Errors in Assembly F - 30% F.P. 31 AH

10. Power Distribution Fac. tors and Comparison to Tech Specs ,

for Fq - 80% F.P. 33

11. Measured Fuel Assembly F - 80% F.P. 34 AH

12. Relative Errors in Assembly F - 80% F.P. 35 AH

13. RPECALIB Output - NIS Recalibration Data, from 40 PT/2/A/4600/05F l

14. Power Distribution Factors and Comparison to Tech Specs for F q- 100% F.P. 46

15. Measured Fuel Assembly F - 100% F.P. 47 AH

16. Relative Errors in Assembly F AH - 100% F.E 48

17. RPECALIB Output - NIS Recalibration Data, from PT/2/A/4600/05A 52 111

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1.0 INTRODUCTION

Catawba Unit 2 Cycle 2 core loading began at 0550 hours0.00637 days 0.153 hours 9.093915e-4 weeks 2.09275e-4 months on January 26, 1988 and concluded at 1700 hours0.0197 days 0.472 hours 0.00281 weeks 6.4685e-4 months on January 30, 1988. The i core loading includes 256 Vet Annular Burnable Absorber (WABA) rods. !

Cycle 1 used borosilicate glass absorber rods. I Criticality was achieved at 0545 on March 7, 1988. Zero Power Physics Testing (ZPPT) was completed at 1320 on March 8, 1988. Power Escalation Testing through 20% Full Power was completed by 1740 on March 9, 1988.

The unit then entered an outage to resolve Auxiliary Feedwater System (CA) operability concerns. The Power Escalation Test Program was re-entered at 1547 on March 19, 1988 at 20% Full Power. The unit reached 98% full power, where the full power testing was conducted at 0100 on March 24, 1988. Essentially, all testing was complete by March 26, 1988.

All power escalation testing was complete by April 1, 1988.

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1 o e 2.0 PRECRITICAL TESTING w "

Procriti testing includes the procedures used for:

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1) loading the Cycle 2 core,

2) determining initial calibration data for Excore Power Range and Intermediate Range detectors, and

3) the approach to criticality Sections 2.1 through 2.3 describe these procedures and results for Catawba 2 Cycle 2.

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2.1 Total Core Reloading - PT/2/A/4150/22 ,

Cycle 2corewasloadedunderthedirectionofPT/2/A/4150/22, Toe %1 Core Reload. Plots of Inverse Count Rate Ratio (ICRR) versus number of fuel assemblies loaded were kept for cach source range channel. 1 i

Core loading began at 0550 hours0.00637 days 0.153 hours 9.093915e-4 weeks 2.09275e-4 months on January 26, 1988 and concluded at 0807 hours0.00934 days 0.224 hours 0.00133 weeks 3.070635e-4 months on January 30, 1988. The core loading was verified by PT/2/A/4550/03C, Core Verification, which was completed at 1700 hours0.0197 days 0.472 hours 0.00281 weeks 6.4685e-4 months on January 30, 1988.

Figuro 1 shows the core loading for Cycle 2.

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FIGURE 1

.- CORE LOADING PATTERN CATAWBA UNIT 2 CYCLE 2 R P N M L I J H G F E D C 8 A IM35 IP51 IP03 !P54 IP17 IP56 IM26 i 1 1 ! I i ! ! I t t 150ti 1241tf I238tf 1237tT 1265tf !240tT !62ti !

IMot IP20 IP34 IN54 IN19 IN51 IN18 lN28 lP36 1P92 IM23 l 2 l ! I I I ! ! I I I I I 2 1160tf IRS 6 l#31t IR61 !!83tf !R76 1149tf IR63 !#26t IR57 1104tf I IM02 IP62 IN36 fM56 lP45 IM40 IN49 IM64 IPt8 IM45 IN63 IP63 1833 I 3 I I I I ! ! l ! ! I I I ! ! 3 1206tf 1181ti 1172tf 1994 18P62t !R72 10553E IR74 !8P&9K IR87 1148tf 1217ti !!05tf I fp29 IN56 IP30 IM50 !N29 IM53 4 I i

!M38 IN44 IM4 lP25 !M1 IN17 IP27 I 1 1 ! I i ! ! I I I I I 4

tR60 !!43tf IR96 112P62tl162tf 1151tf !R103 1163tf 1110tf 112963tlR97 1161ti IR42 IM09 IP16 IM19 1P19 IM24 1P01 IM46 IN39 IM47 IP33 IM59 1P15 5t i I I IM43 IP10 iM20 1 I I ! ! I I 156tf !#32t !R88 I I I I i 1. 5

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112P60ti139tf !BP48E 151ti 1138tT 1157tT !8P71t 1127ti 112959t!R95 !4P25E 1132tf i f 1P50 IN42 !P41 IM43 lP39 IM18 6!

!N20 !N30 !N24 !M45 lP32 lM34 IP48 !N25 !P64 l-

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!212tf IR64 IB?601 1107tf 18P58t !R77 I!87tf IRS 4 1178tf 1978 18P66E 163ET 18963t 1R46 !!89tf l IP31 !N02 lM21 IN49 IM28 IN07 !N11

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1171tf 153ti !8P72t 1144tf 112958t!R89 l#28E 1125tf f IP05 158 IMS8 IP13 IM12 IN26 IM11 IN27 IMt6 IP06 fM32 12 I i 1 1 1 IN46 !P14 1 1 1 ! l ! ! I t ! 12 IR45 I N T 11160 !!!964til16tf 1141tf IR106 1152tf 175tf 112961t!R101 1179tf IR67 I IM13 IPS8 IN16 IR37 IP22 IM22 IN34 IM16 IP21 IM14 IN48 13 I I I I I IPS3 IM13 1 i 1 1 I I I I I i 13 154tT 1274tf 1175tf IR91 189448 1R84 10554t 1985 18P70t 1R102 1177tf 1230tf 1219tf f M fP11 IP35 IN21 IN47 IN33 IN05 fM62 IP26 IP44 14 !

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IM2 IP55 !P24 IP49 IP12 IP61 IM3 !

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l 2.2 Post-Refuelina NIS Realignment - PT/2/A/4600/05E j

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This test was performed on February 18, 1988, for the purpose of I calculating preliminary calibration data for the Power Pange and l Intermediate Range detectors following refueling. I The Cycle 2 preliminary calibration data was determined by taking I the calibration data at the end of Cycle 1 and adjusting it by ;

the ratio of the sum of the predicted assembly powers for Cycle 2 ,

core loading (from Westinghouse predictions) to the sum of the I measured assembly powers from the last Cycle 1 calibration. The core locations selected for the calculation of the ratios required j

to adjust the Power Range and Intermediate Range detectors are i shown on Figure 2. 1 The Beginning of Cycle (BOC) 2 - to - End of Cycle (EOC) I ratios for the Intermediate Range Channels were determined to be 1.070 and 1.097 for N35 and N36 respectively. The BOC 2 - tc - EOC 1 ratios for the four Power Range Channels were calculated to be 0.591, 0.576, 0.584, and 0.58& for N41 through N44 respectively.

These ratios were used to adjust the End-of-Cycle 1 Intermediate Range Rod Stop (20%) and Low Power Trip (25%) setpoints and the Power Range Full Power 0% axial offset calibration currents.

Recalibration by I&E using the adjusted values determined under this procedure was performed on February 19, 1988.

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. ASSEMBLIES SELECTED FOR CALCULATING NIS PRELIMINARY SETP0INTS 1R 56 ,

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l 2.3 1/M Approach To Criticality - PT/2/A/4150/19 l

b On February 24,1988, at 2024 hours0.0234 days 0.562 hours 0.00335 weeks 7.70132e-4 months , boron samples of the Reactor i Co61 ant System, Pressurizer, and Volume Control Tank were obtained i by Chemistry personnel in preparation for initial approach to criticality on Catawba 2, Cycle 2. The results of these samples were 2006 ppmB, 2009 ppmB, and 2009 ppmB respectively. This data was used to predict the Reactor Coolant System dilution required to )

achieve criticality at the desired Control Bank D position. The required dilution was calculated to be 27,641 gallons of demin water yielding a desired Reactor Coolant System boron concentration '

of 1295 ppmB. l 1

At 2130 hours0.0247 days 0.592 hours 0.00352 weeks 8.10465e-4 months on February 24, 1988, with all Control and Shutdown Banks inserted, a controlled Reactor Coolant System dilution of )

~ 66 gpm was commenced. Dilution was halted at 0430 hours0.00498 days 0.119 hours 7.109788e-4 weeks 1.63615e-4 months on l February 25, 1988 after the required 27,641 gallons of water had I been added. The hourly NC System boron samples, commenced with the I initiation of system dilution, were continued until it was noted that the Reactor Coolant System was sufficiently mixed at a concentration of 1290 ppmB at 0830 hours0.00961 days 0.231 hours 0.00137 weeks 3.15815e-4 months on February 25, 1988.

Mode change requirements prevented the unit from going into Mode 2 '

so the 1/M Approach to Criticality procedure was put on hold. At 1720 hours0.0199 days 0.478 hours 0.00284 weeks 6.5446e-4 months on February 26, 1988, the unit was forced to begin cooldown to Mode 4, as required by Technical Specifications, to perform repairs to the Auxiliary Feedwater pump turbine. Mode 4 was entered at 2247 hours0.026 days 0.624 hours 0.00372 weeks 8.549835e-4 months on February 26, 1988. As a result of going to Mode 4, the Reactor Coolant System boron concentration was ,

increased to meet shutdown margin requirements.

On March 3, 1988, at 2002 hours0.0232 days 0.556 hours 0.00331 weeks 7.61761e-4 months , boron samples of the Reactor Coolant System, Pressurizer, and Volume Control Tank were obtained by Chemistry personnel in preparation for initial approach to criticality on Catawba 2, Cycle 2. The result of these samples were 1386 ppmB, 1415 ppmB, and 1387 ppmB respectively. This data was used to pradict the recend dilution of the Reactor Coolant System required to achieve criticality at the desired Control D position. The required dilution was calculated to be 4199 gallons of domin water yielding a desired Reactor Coolant System boron concentration of 1295 ppmB.

Atr0440 hours on March 4,1988, with all Control and Shutdown Banks inserted, a controlled Reactor Coolant System dilution of ~ 70 gpm was commenced. Dilution was halted at 0540 hours0.00625 days 0.15 hours 8.928571e-4 weeks 2.0547e-4 months on March 4, 1988 after 4200 gallons of demin water had been added. The hourly Reactor Coolant System boron samples, commenced with the initiation of system dilution, were continued until the system was sufficiently mixed at a concentration of 1300 ppmB at 0910 hours0.0105 days 0.253 hours 0.0015 weeks 3.46255e-4 months on March 4, 1988.

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l 1/M Approach to Criticality procedure was again put on hold as a result of Auxiliary Feedwater System problems which were required to be fixed before ettering Mode 2. Unit cooled down to 400*F in order to repair 2SA-2 valve. Boron concentrations were not significantly changed having met Shutdown Margin Requirements for that temperature.

Following repairs to the Auxiliary Feedwater System problems, Reactor Coolant System, Pressurizer, and Volume Control Tank were obtained by Chemistry personnel at 2126 hours0.0246 days 0.591 hours 0.00352 weeks 8.08943e-4 months on March 6, 1988.

The results of these samples were 1310 ppmB,1325 ppmB, and 1314 ppmB, respectively. Following the acquisition of baseline count rates from the Source Range NIS channels, sequential withdrawal of the Shutdown Banks was commenced (with S/D Bank A starting at 0205 hours0.00237 days 0.0569 hours 3.38955e-4 weeks 7.80025e-5 months on March 7, 1988). Withdrawal of the Shutdown Banks w'as completed at 0252 hours0.00292 days 0.07 hours 4.166667e-4 weeks 9.5886e-5 months , with S/D Bank E at 228 steps withdrawn.

Control rod withdrawal resumed with the Control Banks in overlap.

ICRR plots were maintained throughout rod withdrawal (see Figure 3) and criticality was achieved at 0545 hours0.00631 days 0.151 hours 9.011243e-4 weeks 2.073725e-4 months on March 7,1988 with Control Bank D at 140 steps withdrawn.

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3.0 ZERO POWER PHYSICS TESTING Zero Power Physics Testing (ZPPT) for Catawba 2 Cycle 2 was controlled by PT/2/4/4150/21, Post-Refueling Controlling Procedure for Startup Testing. ZPPT began at 0205 hours0.00237 days 0.0569 hours 3.38955e-4 weeks 7.80025e-5 months on March 7, 1988 and concluded at 1320 hours0.0153 days 0.367 hours 0.00218 weeks 5.0226e-4 months on March 8, 1988. The output from Power Range detector N41 was used as input to the Westinghouse Analog Reactivity Computer, an IBM 9000 (digital) Reactivity Computer, and a Westinghouse Digital Reactivity Computer. The Westinghouse Digital Reactivity Computer was new to Catawba Nuclear Station, and was only connected to N41 for benchmarking purposes. The Westinghouse Analog Reactivity computer was the official source for reactivity measurements. Core design data for ZPPT were supplied by Duke Power Nuclear Design. All acceptance criteria associated with ZPPT were met.

The required minimum of one decade of overlap between the Source Range and Intermediate Range channels was verified at 0600 on March 7,1988.

Table 1 summarizes the results.

The point of nuclear heat addition was determined at 0815 on March 7, 1988. This determination was made by the observation of increasing Reactor Coolant System average temperature and Pressurizer i Level in conjunction with a change in the reactivity trace during a slow '

positive startup rate. Table 2 summarizes the results.

r A checkout of the reactivity computers was completed at 0900 on March 7, 1988. This checkout was done by inserting (withdrawing) Control Bank D = 25 pcm and measuring the halving (doubling) time (time required for flux signal to halve or double). From the halving (doubling time), ,

the theoretical reactivity was found using predicted data, and was compared to the measured reactivity. The test was repeated for reactivity insertions of of = 50 pcm. Table 3 summarizes the results. '

Sections 3.1 through 3.4 describe the other procedures used for ZPPT.

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TABLE 1

< a ISCURCE RANGE /I!CERMEDIATE RANGE OVTRLAP DATA -

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SOURCE RANGE INTERMEDIATE RANGE (CPS) (AMPS)

N31 N32 N35 N36

INITIAL DATA NIS Cabinet 1500 2000 3.0 x 10'11 1.5 x 10~81 OAC 1870 2104 2.2 x 10'11 2.2 x 10~11

FIRST DECADE NIS Cabinet 10,000 15,000 2.0 x 10~1' 1.0 x 10~1' OAC 11,870 13,230 1.2 x 10'.1' 1.2 x 10

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TABLE 2

,f NUCLEAR HEAT DETERMINATION

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REAL"TIVITY COMPtTTER INTERMEDI!TE RANGE (AMPS) (AMPS)

N41 N35 N56 RUN #1 2.5 x 10 s 1.720 x 10 1.774 x 10

4 RUN #2 2.2 x 10 1.659 x 10 1.652 x 10

Upper limit of test band is minimum reading at point of adding heat divided by 2 or:

2.2 x 10 amps + 2 = 1.1 x 10~8 amps s

ZPPT TEST BAND: 10 to 10 amps on N41 $

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TABLE 3 s REACTIVITY COMPlTTER CHECK 0UT -

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9 WESTINGHOUSE REACTIVITY COMPlTTER HALVING REACTIVITY PERIOD (DOUBLING) COMPlTTER THEORETICAL PERCENT DIFFERENCE (sec) TIME (sec) Ap (pca) REACTIVITY ERROR

Ap-ApDT

&nDT (pcm) (%) (pcm)

-333.71 231.31 -25.0 -25.10 0.40 0.10 284.51 (197.21) +24.0 +23.43 2.43 0.57

-220.43 152.79 -40.0 -40.91 2.22 0.91 132.73 (92.00) +46.0 +45,24 1.68 0.76 Ap -ApDT .

PERCENT ERROR = ApDT x 100 ;

ACCEPTANCE CRITERIA: PERCENT ERROR < 4.0% or 1 pcm difference between Ap and [pDT, whichever is greater.

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3.1 Baron Endpoint Measurement - FT/2/A/4150/10 2%:I- l test was first performed on March 7,1988, to measure the l Al Out Boron Endpoint. Control Bank D was initially at 214 l st withdrawn. The Reactor Coolant System boron concentration was 1346 ppmB. j Control Bank D was pulled to the All Rods Out (ARO) configuration and the resulting reactivity change was converted to equivalent boron using the differential boron worth. This was performed two times to ensure repeatability.

The results of these reactivity changes were added to the initial Reactor Coolant System boron concentration and the values averaged to give a Final ARO Boron Endpoint of 1349 ppmB. This met the acceptance criterion of 1370 150 ppmB for the ARO Boron Concentration.

This test was again performed on March 8,1988, during Rod Worth Measurements to obtain the Boron Endpoint with the Reference Bank fully inserted. Following insertion of the Reference Bank Control Bank C, during its measurement by dilution, the Reactor Coolant boron concentration was determined to be 1259 ppmB with Control ,'

Bank C at 18 steps withdrawn.

Insertion of Control Bank C to its fully inserted position and adjustment of the critical boron concentration with the reactivity associated with this maneuver yielded a Reference Bank In Boron Endpoint of 1258 ppmB. .

This measurement was used in conjunction with the ARO Boron Endpoint and the predicted Differential Boron Worth to calculate the worth of the Reference Bank inferred. (See Section 3.4) t' 1

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3.2 Isothermal Temperature Coefficient of Reactivity Measurement -

Pf/tfA/4150/12A .

Thitstest was performed on March 7,1988. The test measures the Isotternal Temperature Coefficient (ITC) by noting indicated reactivity changes versus Average Reactor Coolant System Temperature changes. The Moderator Temperature Coefficient (MTC) is found using the relationship as follows:

MTC (pcm/*F) = ITC - Doppler Temperature Coefficient The acceptance criteria on the ARO ITC was +0.65 i 3 pcm/'F. The predicted Doppler Temperature Coefficient was -1.29 pcm/'F. The average measured ARO ITC (for 3 heatups and 2 cooldowns) was +1.65 pcm/*F. The ARO KIC was therefore +2.94 pcm/*F. The MTC limit between 0% and 70% RTP is +7 pcm/*F. The results are summarized on Table 4. The actual measurement traces are shown on Figure 4. The data obtained from the first cooldown was not used to obtain an average MTC due to difficulty in analyzing the X-Y Plotter Trace (created by a scaling problem).

Using data from thic test, PT/2/A/4150/20 Temporary Rod Withdrawal i Limit Dete rmination, was performed on March 8,1988, to ensure that the MTC was within the limits of Technical Specification {

3.1.1.3. It was detarmined that the MTC was within the limits of -

TerAnical Specification 3.1.1.3, and that no rod withdrawal limits were required.

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. +

,; TABLE 4

'IT.MPERARIRE COEFFICIENT MEASUREMENT RESULTS.

.; .i !

y

c. ,g.'y AT AP ITC

(*F) (pcm) (pcm/*F)

+First Ccoldown N/A N/A N/A First Heatup +4.1 +7.0 +1.7073 Second Cooldown -6.4 -9.5 +1.4844 Second Heatup +4.7 +7.75 +1.6489 Third Cooldown -4.7 -8.0 +1.7021 Third Heatup +4.4 +7.5 +1.7045 AVERAGE: +1,6494 MTC = +2.939 pcm/'F '

MTC = Average ITC - (BOL, HZP Doppler Coefficient) where BOL, HZP Doppler Coefficient = -1.29 pea /*F

+ NOTE: First cooldown ineasurement invalid due to improper scaling of X-Y ,

Plotter. This measurement is not used in calculation of Average MTC.

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FIGURE 4 ISOTHERMAL TEMPERATURE COEFFICIENT OF REACTIVITY UREMENT DATA FROM HEATUP fl MARCH 7,1988 at 1356 HOURS

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FIGURE 4

ISOTilERMAL TEMPERATURE COEFFICIENT OF REACTIVITY MEASUREMENT DATA FROM COOLWWN 82 AND llEATUP #2 MARCll 7, 1988 AT 1626/1738 Il0URS '

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FIGURE 4 ISOTIIERMAL TEMPERATURE COEFFICIENT OF REACTIVITY ~

MEASUREMENT DATA FROM C00LDOWN #3 AND HEATUP #3 MARCII 7, 1988 AT 1830/1907 HOURS

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3.3 Control Rod Worth Measurement By Boration/ Dilution - IT/2/A/4150/11A On Ehech 8,1988, Control Bank C was measured using the dilution metEsd. There were no other rods in the core at the time. Control Bank C was predicted to be the "heaviest" bank and was measured using this method in order for it to serve as the Reference Bank for Control Rod Measurement By Rod Swap.

The performance of this procedure resulted in a measured worth of Control Bank C of 988 pcm. The predicted worth was 955 pcm.

This represented an error of 33 pcm (+3.46%) and was within the accep*.ance criteria of 955 143 pcm (* 15%).

The measured integral and differential rod worths for Control Bank C are shown in Figures 5 and 6.

l 21

he a h 5 5 FIGURE 5 CONTROL BANK C INTEGRAL R0D WORTH MEASURED BY DILUTION e

I

/ !8 X

/ gi

/ B X

/ g "E M

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FIGURE 6 CONTROL BANK C DIFFERENTIAL R00 WORTH l MEASURED BY DILUTION l

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3.4 Control Rod Worth Measurement By Rod Swap - PT/2/A/4150/llB r, .

On Mae'h c 8, 1988, the Rod Swap method of control rod worth measurement was performed. Control Bank C was used as the refe'rence bank and its worth was measured by the Boration/ Dilution method (see Section 3.3).

The Reference Bank worth obtained by the Boration/ Dilution measurement technique was checked against an inferred worth obtained using the results of the two Boron Endpoint Measurements and a predicted Reference Dif ferential Baron Worth of -10.49 pcm/ppmB. Multiplying the difference between the Boron Endpoints by the Differential Boron Worth yielded an inferred worth of 954.6 pcm. This value deviated from the measured worth of 988 pcm by only 3.38%, well within the desired i 15%.

Starting with the reference bank deeply inserted, and the reactor just critical, each shutdown and control bank was swapped into the core for the reference bank. The integral worth of each test bank was inferred from the critical position of the Reference Bank with the test bank fully inserted in the core. The measured worths were ccmpared with predicted worths, and all the test acceptance i criteria were sat'isfied. ,'

The results from this test are shown on Table 5.

.,. we- . ,_

24

TABLE 5

CEBmt0L ROD WORTH MEASUREMENT BY R0D SWAP DATA ,

[Results Using Reference Bank Worth From Dilution Measurement on March 8, 1988 Predicted Worth Measured Worth Percent Difference +

Bank (pcm) (pce)

Control Bank C 955 988* 3.5 (Reference)

Shutdown A 339 369.1 8.9 Shutdown B 779 819.9 5.3 Shutdown C 345 323.3 -6.3 Shutdown D 348 332.6 -4.4 Shutdown E 421 436.5 3.7 Control A 360 399.8 11.1 Control B 701 787.5 12.3 Control D _457 443.8 -2.9 Total Rod Worth 4705 4900.5 4.2++

Measured by Boration/ Dilution - Dilution

+ Measured - Predicted x 100%

Predicted ,

++ Sum of Measured Worths - Sum of Pradict i Vorp x 100*.

Sum of Predicted Worth NOTE: This includes Reference Bank Worth.

6 25 o

. O i

4.0 POWER ESCALATION TESTING 1

Catawba Unit 2, Cycle 2 Power Escalation commenced at 1305 hours0.0151 days 0.363 hours 0.00216 weeks 4.965525e-4 months on I March 8,.1968. Mode 1 was initially entered at 1618 on March 8, 1988.

At 0510 am March 9, 1988, following Turbine Shell and Chest Warming at a power level of 10% F.P. , the Turbine / Generator was placed on line. The Turbine Overspeed Test was performed by Operations at a power level of 16% F.P. approximately six hours later.

The Turbine / Generator was placed on line at 1245 on March 9, 1988, and power escalation to full power was commenced. Power Escalation Testing through 20% F.P. was completed by 1740 on March 9, 1988. The unit then entered an outage to resolve Auxiliary Feedwater System (CA) operability concerns. The Power Escalation Test program was re-entered at 1547 on March 19, 1988 when power escalation commenced from 20% F.P. to 30% F.P.

at 2.5%/ hour.

The 30% F.P. Testing Plateau was the first place that data acquisition of important NSSS and Main Turbine parameters per PT/2/A/4150/16, Unit Load Steady State, was performed. Turbine Impulse Pressure readings at a lower power level had previously been obtained under.this procedure. In addition, baseline data from the Loose Parts Monitoring System i.as taped (

at 30% F.P. These testing activities were repeated at the 80% and 100% ;

testing plateaus. The most critical application of this data would be the extrapolation of core delta temperature for each Reactor Coolant '

Loop. This allowed the solid state protection systems process cards to be adjusted as necessary to align full power indicated Delta Temperature with 100% Reactor Thermal Power. See Table 9 for Extrapolated Delta Temperatures to 100% F.P. for each individual loop. .

A Full Core Flux Map was obtained at the 30% testir.g plateau per PT/2/A/4150/05, Core Power Distribution. This map demonstrated compliance with Tech Specs with respect to all core peaking factors, and incore tilts also met their acceptance criteria. All acceptance criteria associated with PT/2/A/4150/05 were satisfied. The results of this test are summarized on Table 6 and Figures 7, 8, and 9.

At 1130 hours0.0131 days 0.314 hours 0.00187 weeks 4.29965e-4 months on March 20, 1988, power escalation was resumed at 2.5%/ hours from 30% F.P. to 50% F.P. During the power increase (at 44%

F.P.) PT/2/A/4150/03A, NSSS Thermal Outputs, was performed. The purpose of this test was to qualify the Operator Aid Computer Thermal Output programs. See Section 4.2 for a summary of this testing.

At 0141 on March 21, 1988, Reactor Power began a 2.5%/ hour escalating per i PT/2/A/4600/05F, Post Refueling: Incore and NIS Recalibration. 80%

F.P. was reached at 1350 on March 22, 1988. See Section 4.1 for a summary of this testing.

At the 80% Testing Plateau, a number of activities were performed. A Full Core Flux Map was obtained per the Core Power Distribution Procedure l to ensure continued compliance with Tech Spec limits and evaluate the extrapolated full power Nuclear Heat Flux peaking factors to verify that operation in RAOC Mode on up to 100% F.P. was permissible. The results {

of this flux map indicated that all Tech Specs at 80% F.P. were satisfied ,

and that power escalation to full power could be performed without l invoking Base Load Mode of operation. The results of the 80% Flux Map !

are summarized on Table 7 and Figures 10, 11, and 12. l

- 26

The Fower Range NIS Calibration data obtained per the Post Refueling Incore and NIS Recalibration was incorporated by IAE to calibrate the Excore NES at 1440 on March 23, 1988. Also at the 80% F.P. Testing Plateau, Unit Load Steady State data was again obtained.

Power Escalation was resumed at 2.5%/ hour at 1500 on March 23, 1968. The unit reached 98% F.P., at 0100 on March 24, 1988. At this power level, following achievement of equilibrium core conditions, the testing described by Sections 4.3, 4.4, 4.5, and 4.6 were performed. Table 8 summarizes the NIS Power Range and Intermediate Range data obtained over the duration of power escalation.

Due to PT/2/A/4600/05B, Incore and Nuclear Instrumentation System Calibration Check not passing its acceptance criteria (see Section 4.6 for details) it was necessary to perform PT/2/A/4600/05A, Incore and NI System Correlation Check. This test collected and generated calibration data for the excore NIS. This test began at 1200 on March 26, 1988 and concluded data collection portion of the test at 1741 ou March 26, 1988.

On March 26,1988, PT/2/A/4150/13B, Calorim.tric Reactor Coolant Flow Measurement, was performed to validate the flow indication of elbow taps and to verify that the total Reactor Coolant Flow is greater than the i minimum Tech Spec required flow. j 4

l 1

l l

l d

27

TABLE 6 CORE POWER DISTRIlWIION RESULTS 30% FULL POWER

'y -

FLUX MAP FCM/2/02/001 N.

Date/Timeof$bp 03/20/88 at 0040 Reactor fower Level 28.86% F.P.

Cycle Burnup 0.306 EFPD NC System Boron Conc. 1198 ppmB Control Bank D Position 213 steps withdrawn Maximum Total F q 2.0871 at Axial Loc. 50 Horiz. Loc. A-08 Maximum F g 1.4108 at Axial Loc. 50 Maximum FXY (unexcluded) 1.6526 at Axial Loc. 21 Horiz. Loc. N-13 Maximum Total Fq /K(2) 2.1912 at Axial Loc. 50 Horiz. Loc. A-08 Minimum Margin to F Limits q -50.0011% at Axial Loc.}51 Maximum Reduction of AFD RAOC Wings 0%

Maximum Pin F 1.3945 at Horiz. Loc. R-08 aH Pin # 215 l Max. Assembly Error FAH (fr m predicted) 5.35% at Horiz. Loc. G-12 Average of Absolute Errors in F 1.45%

3g ,

Maximum Calculated "R" 0.7708 Reactor Coolant Flow (OAC Indicated) 401,968 GPM i Required Tech Spec Flow for Full l Power Operation 387,600 GPM '

Incore Axial Offsets:

Total Core + 19 ' ~1'2 7 %

Quadrant 1 +19.447%

Quadrant 2 , +18.369%

Quadrant 3*

~ +19.430t' q.

Quadrant 4 +19.250%

IncoreTilta,hhorsalized): I i Upper Core Lower Core Quadrant 1: 1.006 0.999 l Quadrant 2: 0.983 0.999 I Quadrant 3: 1.005 0.998 Quadrant 4: 1.006 1.003 Quadrant 5: 1.006 1.010 Quadrant 6: 0.986 0.987 Quadrant 7: 1.010 1.008 Quadrant 8: 0.999 0.995

NOTE: Axial Loc.1 is bottom of core, Axial Loc. 61 is top of core.

28

P0uEN DISTH I NT 10!. FACTCRS AND CCMPARISCN TO TLCifNICAL 5FECIFICATIONS FOR f SUD C AXIAL kic HOH.

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TH COMPONENTS F THE MA F SUB o ARF F SU 4 L.= 4 ,4 AND F SUD Z= 1.41os gay g )7 AL ss , v n, . , . . . .

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E CATAWBA Dt:TECTOR I!IIN Ulk 11 2 CYCLE 2) t.,oPHOL CET*2/02/001

,EUCLEAR PEAKING FKTOR5 FOR ENTHALPY RISE FCR AS5th0LAGES IN THE PitWER t10Rf4 AL I Z AT ION , ,__ (, y 4 NN* r ge - 04 05 04 07 08 09 10 11 12 13 14 15 0.409 0.894 1.011 1.0 79 1.017 0.907 0.417 ,,,

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30's F.P.

31

-~

~

)

TABLE 7 l

CQRE POWER DISTRIBLTTION RESULTS 80% FULL POWER FLUX MAP FCM/2/02/014 Date/ Time of 03/22/88 at 2330 Reactor Power Level 8 0. 3*. F . P .

Cycle Burnup 1.6 EFPD l NC System Boron Conc. 1020 ppmB Control Bank D Position 181 steps withdrawn Maximum Total F q 1.9063 at Axial Loc. 50 Horiz. Loc. F-05 Maximum F g 1.2586 at Axial Loc. 48 Maximum Fg (unexcluded) 1.5603 at Axial Loc. 18 Horiz. Loc. N-13 Maximum Total Fq /K(Z) 2.0018 at Axial Loc. 51 Horiz. Loc. F-05 Minimum Margin to F Limits q -24.7928 at Axial Loc. 40 Maximum Reduction of AFD RAOC Wings 0* '

l Maximum Pin F g 1.3788 at Horiz. Loc. R 08 j Pin # 217 !

Max. Assembly Error FAH (fr m Predicted) 12.25*. at Horiz. Loc. H-08 (uninstrumented) . !

Average of Absolute Errors in F 1.88*.

g Maximum Calculated "R" 0.8733 Reactor Coolant Flov (OAC Indicated) 400,405 GPM Required Tech Spec Flow for Full

( Power Operation 387,600 GPM Incore Axial Offsets:

Total Core +11.006*

Quadrant 1 +11.128*,

Quadrant 2 +10.924*,

Quadrant 3 +11.133*,

,_ Quadrant 4 +10.839*.

Incore Tilts permalized):

l% er Core Lower Core Quadrant 1: 1.002 1.000 Quadrant 2: 0.993 0.994 1 Quadrant 3: 1.003 1.001 l Quadrant 4: 1.002 1.005 l Quadrant 5: 1.005 1.003 i Quadrant 6: 0.991 0.989 Quadrant 7: 1.008 1.009 Quadrant 8: 0.996 0.999

NOTE: Axial Loc. 1 is bottom of core, Axial Loc. 61 is top of core.

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CATAttBA DETECTOR RUN (UNIT 2 CYCLE 2)

CAT 2/02/014 U150

/

NUCLEAB_PEAEIBC i FACTEB5 FOR ENTHALPY RISE FOR ASSEMBLAGES.IN THE_POWEH.!J03MALIZAIION __

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35

6 .

TABLE 8

-q... INTERMEDIATE RANGE / POWER RANGE NIS OVERLAP DATA,

. n: + ,

DATE ,

" TIME INTERMEDIATE RANGE (AMPS) THERMAL POWER

, N35 N36 BEST ESTIMATE (OAC) 03/08/88 1427 9.50E-6 8.86E-6 1.44% F.P.

03/09/88 0050 4.75E-5 4.43E-5 9.65% F.P.

03/09/88 1738 8.11E-5 7.50E-5 19.80% F.P.

03/19/88 1958 9.89E-5 9.22E-5 25.10% F.P.

03/20/&8 2257 2.15E-4 1.95E-4 48.87% F.P.

03/22/88 2045 3.70E-4 3.36E-4 80.80% F.P.

03/25/88 0843 4.58E-4 4.10E-4 98.40% F.P.

i E

A 0

36

_____________._______________________________._.______________________._______._________.._____.___________________________.______m _

1 I

i 1

TABLE 9 I EXTRAPOLATED DELTA TEMPERARTRES

_-- TO 100% F.P.

-s The following' data is from PT/2/A/4150/16, Unit Load Steady State performed at the 80% Test Plateau.

Loop Extrapolated Delta T (*F)

A 57.77 B 58.37 C 56.34 D 57.29 The following information was calculated using data from PT/2/A/4150,13B, Calorimetric Reactor 'aolant Flow M ae surement performed on March 29, 1988 at the I 100% Test Plateau. i I Loop Extrapolated Delta T (*F) '

A 57.19 B 58.31 C 55.92 D 57.02 i

i i

l I

l i

1 l

37

---,1 , , - , - . , , e ,--- , - ,-.,,.-r-.. - - - , , , - . - _ _ r - , .

- , , _ _ ~ , - , _ . - -

4.1 Post Refuelinz: Incore and NIS Recalibration - FT/2/A/4600/05F

,J. .

Thisjtest was conducted as Reactor Power was increased from 50%

F.P.~;to 80% F.P. The first Quarter Core Flux Map for this test was initiated at 0749 hours0.00867 days 0.208 hours 0.00124 weeks 2.849945e-4 months on March 21, 1988, and testing concluded with the recalibration of the NIS at 1440 on March 23, 1988.

The data acquisition portion of the test required flux Japping (using a Quarter Core Flux Map pattern qualified per Fr/0/A/4150/23) coincident with the recording of Power Range NIS currents at various axial flux dif ferences durint power escalation. The core axial offsets derived from the twelve Quarter Core Flux Maps and the associated NIS data were used by the RPECALIB off-line program to generate calibration data for the Excore NIS. The results of the flux maps are shown on Table 10.

The RPECALIB output (shown on Figure 13) was used by I&E personnel to set the NIS amplifier gains and the axial flux difference function of the Overpower AT setpoints in the SSPS. This correlated the excore axial offset indications to the "true" incore axial offsets.

Proper calibration of the affected instrumentation systems was -

verified per the test procedure and all acceptance criteria were l met.

l

)

D 38 l

l

-_ - -_ , _ = -- .

TABLE 10 QUARTER CORE FLUX MAP DATA FOR

. POST REFUELING: INCORE AND NIS RECALIBRATION MAP ID AVERAGE THERMAL POWER INCORE AXIAL OFFSET QCM/2/02/002 52.63% F.P. +14.981%

QCM/2/02/003 55.76% F.P. +12.823%

QCM/2/02/004 58.64% F.P. +10.062%

QCM/2/02/005 61.07% F.P. +9.658%

QCM/2/02/006 62.80% F.P. +8.026%

QCM/2/02/007 65.78% F.P. +6.155%

QCM/2/02/008 67.61% F.P. +4.253%

QCM/2/02/009 70.51% F.P. +1.910% .,'

QCM/2/02/010 72.59% F.P. +0.286%

QCM/2/02/011 75.35% F.P. -0.705%

I QCM/2/02/012 77.90% F.P. -0.477% ,

QCM/2/02/013 79.26% F.P. -0.370%

39

j .- o 1 .

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. . 1 4.2 M888 Thermal Output -

PT/2/A/4150/03A

-- l Thin [5est was used to verify that the Primary and Secondary Heat l Bale,qpe Programs on tF* plant computer were consistent with Primary I and Secondary Heat Balance Programs on an of fline computer. This test was successfully performed on March 20, 1988,. at 44% Rated Thermal Power.

The acceptance criteria of i0.1% for the deviation between offline and plant computer secondary heat balance indications was met for the performance of this test. The acceptance criterion of i 0.5%

)

for the deviation between offline and plant computer primary heat '

balance indications was also met for this test. As a prerequisite to performing this test, PT/2/A/4150/03B, Thermal Outputs Inputs ,

Reliability Check was performed to verify the inputs to the OAC {

Thermal Output Program are reasonable.

The results of the NSSS Thermal Outputs test are summarized below:

Plant Computer Offline Computer

% F.P. % F.P.

4 Primary Heat Balance 44.4238% 44.36% 8 Secondary Heat Balance 43.6460% 43.64%

4 l

l 1

l l

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9 41

d 4.3 Reactivity Anomaly Calculation - PT/2/A/4150/04

.~T :

This, test was performed on March 25, 1988 for the purpose of comparing actual core reactivity to Theoretical HFP, ARO Critical Boron i Condentration prediction. The test method involved the correction of l Reactor Coolant boron concentration (measured per Chemistry sample) for i the presence of control rods and off-equilibrium Xenon and Samarium l worths so that it could be compared to the predicted Full Power, All ,

Roos Out, Equilibrium Xenon / Samarium boron concentration provided by !

the core design parameters report. Since the unit was at 97.79% F.P. j (instead of 100% F.P.) an adjustment for power level was required. '

The measured boron concentration obtained was 942.8 ppmB. This translated to an error of +107.0 pcm when compared to the Theoretical HFP, ARO Critical Boron Concentration value of 931.9 ppmB, well within the test acceptance criterion of i 1000 pcm. Continued surveillance per this proceoure will be performed for renormalization of the design HFP Boron Letdown Curve prior to 60 EFPD per Tech Spec 4.1.1.1.2.

As required by PT/2/A/4150/21, Post Refueling Controlling Procedure for Startup Testing, the above measured value of 942.8 ppmB was adjusted by the error between the Measured and Theoretical HZP, ARO Critical Boron ConcentrationsobtainedduringZeroPowerPhysicsTesting(seeSectfyn 3.1 - Boron Endpoint Measuremer.t) . Since this error was -21 ppmB, adjustment of the measured HFP, ARO Critical Boron Concentration was' performed by subtracting this value from 942.8 ppmB.

The adjusted value of 963.8 ppmB was determined to be within the required 50 ppmB of the Theoretical KFP, ARO, Critical Boron ,

Concentration of 931.9 ppmB, the actual error being 31.9 ppmB.

l l

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42

. --c. , .-, _ . . . - . -- , , , , -

l 4.4 Termet Flux Difference Calculation - PT/2/A/4150/08 htwasperformedonMarch 25, 1988, for the purpose of ,

as ishing optimum operating targets for core axial flux i difi%rence with Control Bank D in the desired position for full i power operation. With Control Bank D at 205 steps withdrawn the l following AFDs were noted and incorporated as target values for continuing operation:

Quadrant 1 (N-42) -------

+8.18%

Quadrant 2 (N-43) -------

+8.30%

Quadrant 3 (N-41) -------

+8.31%

Quadrant 4 (N-44) -------

+8.53%

This test was again performed on March 29, 1988, due to the recalibration of the Power Range Detectors performed on March 28, 1988. With Control Bank D at 206 steps withdrawn the following AFDs were noted and incorporated as target values for continuing operation:

Quadrant 1 (N-42) -------

+5.10%

i l

Quadrant 2 (N-43) -------

+5.12% l Quadrant 3 (N-41) -------

+5.42%

Quadrant 4 (N-44) -------

+5.60% ,

$' l 43

4.5 Core Power Distribution - PT/2/A/4150/05 A en11 core flux map was obtained under this test on Marhh 25, 1988, forithe purpose of demonstrating compliance with all Power Distribution TecF Specs at full power operation. Table 11 and Figures 14, 15, and 16 summarize the results of this test. All acceptance criteria were met and continued operation in Relaxed Axial Offset Control (RACC) Mode was shown to be permissible from Fqanalysis, with adequate margin to the Tech Spec limit shown on Figura 14.

9 I

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44 1

1 i

TABLE 11 CORE POWER DISTRIBITI' ION RESULTS

,,. 100% FULL POWER

. 5 FLUX MAP FCM/2/02/015 Date/ Time of f 03/25/88 at 1059 Reactor Power Level _ 97.95 % F.P.

Cycle Burnup 4.18 EFPD NC System Boron Conc. 942 ppmB Control Bank D Position 205 steps withdrawn Maximum Total F q 1.7868 at Axial Loc. 49 Moriz. Loc. F-05 Maximum F 3

1.1984 at Axial Loc. 48 Maximum Fyy (unexcluded) __ 1.5092 at Axial Loc. 51 Horiz. Loc. L-08 Maximum Total Fq /K(Z)_ 1.8734 at Axial Loc. 50 Horiz. Loc. F-05 Minimum Margin to F Limits q -4'.7302 at Axial Loc. 14 Maximum Reduction of AFD RAOC Wings 0% ',

Maximum Pin F gg 1.3474 at Horiz. Loc. F-05 Pin # 341 Max. Assembly Error FAH (fr m Predicted) 5.01% at Horiz. Loc. A-11 Average of Absolute Errors in F l' AH ,

Maximum Calculated "R" 0.8983 Reactor Coolant Flow (OAC Indicated) 400,117 GPM Required Tech Spec Flow for Full Power Operation ,

387,600 GPM Incore Axial Offsets:

Total Core +5.343%

Quadrant 1 +5.399%

Quadrant 2 +5.343%

Quadrant 3 +5.378%

Quadrant 4 +5.252%

Incore Tilts-(Normalized):

Upper Core Lower Core Quadrant 1: 1.003 1.002 Quadrant 2: 0.995 0.995 Quadrant 3: 1.000 1.000 Quadrant 4: 1.002 1.003 Quadrant 5: 1.013 1.008 Quadrant 6: 0.987 0.987 Quadrant 7: 1.006 1.006 Quadrant 8: 0.994 1.000

NOTE: Axial Loc.1 is bottom of core, Axial Loc 61 is top of core.

. 45

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4.6 Incore and Nuclear Instrumentation System Calibration Check -

FT/2/A/4600/05B i

This' test was performed on March 25, 1988, in order to compare the valdes of the axial flux difference obtained from incore and excore instrumentation to determine if calibration of the excore detectors is necessary. This test used the incore total core axial offset from flux map FCM/2/02/015 to calculate the incore total core AFD which is compared to individual quadrants excore AFD. If the absolute difference between the incore total core AFD and any of the quadrants excore AFD is more than 3%, then PT/2/A/4600/05A, Incore and Nuclear Instrumentation System Correlation must be performed in order to recalibrate the excore detectors. As seen below, quadrant 4 showed an absolute difference greater than the acceptable limit. The recalibration of the excore power range detectors was completed on March 29, 1988.

Absolute difference between incore total core AFD and excore AFD:

Quadrant 1 (N-42) -----

-2.75%

Quadrant 2 (N-43) -----

-2.82%

Quadrant 3 (N-41) -----

-2.90% .

Quadrant 4 (N-44-) -----

-3.05% ;

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49

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4. 7- Incore and NI System Correlation Check - FT/2/A/4600/05A Th t was conducted when Reactor Power was 98% F.P. at 1200 on 6, 1988.

'h.

The data acquisition portion of the test required flux mapping (using a Quarter Core Flux Map pattern qualified per FT/0/A/4150/23) coincicent with the recording of Power Range NIS currents at various axial offset. The core axial offsets derived from the seven Flux Maps (one Full Cora Map taken for the Core Power Distribution Test, and six Quarter Core Flux Maps taken during this test) and the associated NIS data were used by the RPECALIB off-line program to generate calibration data for the excore NIS. The results of the flux maps are shown on Table 12.

The RPECALIB output (shown on Figure 17) was used by IAE personnel to set the NIS amplifior gains and the axial flux difference function of the Overpower AT setpoints in the SSPS. This correlated the excore axial offset indications to the "true" incore axial offsets.

Proper calibration of the affected instrumentation systems has ;

verified per the test procedure and all acceptance criteria were ;

met.

93 l

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.. TABLE 12

/g' FLUX MAP DATA FOR .

Ap- INCORE AND NI SYSTEM CORRELATION CHECK

$^

MAP ID ~ AVERAGE THERMAL POWER INCORE AXIAL GFFSET FCM/2/02/015 97.95% F.P. +5.343 QCM/2/02/016 97.85% F.P. +3.333 QCM/2/02/017 97.47% F.P. +1.658 QCM/2/02/018 97.53% F.P. -0.699 QCM/2/02/019 97.85% F.P. -2.434 QCM/2/02/020 97.64% F.P. -5.184 QCM/2/02/021 97.78% F.P. -6.620 I

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4.8 Celerimetric Reactor Coolant Flow Measurement - PT/2/A/4150/13B On /88, PT/2/A/4150/13B, Calorimetric Reactor Coolant Flow t, was performed to accurately determine the total flow of the al tor Coolant Loops and normalize the elbow tap instrumentation.

Due~to instrumentation problems, the test had to be rerun on 3/29/88.

Results of the 3/29/88 test are summarized below cnd on Table 13:

RUN #1 RUN #2 RUN #3 AVERAGE

)

POWER LEVEL 95.194% 95.088% 95.094% 95.125%

]

LOOP A FLOW 96,899 GPM 96, 883 GPM 96,942 GPM 96,908 GPM LOOP B FLOW 98,028 GPM 97,917 GPM 97,834 GPM 97,926 GPM LOOP C FLOW 100,638 GPM 100,572 GPM 100,604 GPM 100,605 GPM LOOP D FLOW 100,783 GPM 100,773 GPM 100,731 GPM 100,762 GPM I TOTAL NC FLOW 396,349 GPM 396,146 GPM 396,111 GPM 396,202 GPM

% TECH ,

SPEC FLOW 102.257% 102.205% 102.196% 102.219%

A l'

53 O

7 TABLE 13 CALCULATION OF AVERAGE NC ELBOW TAP FLOW COEFFICIENTS ,

RUN # !LAI EA_2 j@l M EQ2 kDJ ffd HJ;;g g Egl gpg g 1 0.303'484 0.291033 0.299462 0.299592 0.282133 0.303f31 0.315954 0.293661 0.303800 0.300473 0.300879 0.302613 4

2 0.303514 02.91101 0.299538 0.299325 0.281838 0.303211 0.315830 0.293580 0.303676 0.300471 0.300877 0.302564 ,

3 0.303692 0.291184 0. g 18 0.299019 0.281632 0.302899 0.315818 0.293582 0.303765 0.300381 0.3006 p308382

.t H i . .n AVG 0.303563 0.291106 0.299539 0.299312 0.281668 0.303180 0.315867 0.293608 0.303747 0.300442 0.300003 '0.843520 -

54

444 ' e.

DUKE POWER GOMPANY P.O.HOX 33180 CitARLOTTE. N.C. 28242

'IIAL H. TUCKER returnown turrareins=v. (704) 373-4538 wtuaan reonervion June 27, 1988 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555

Subject:

Catawba Nuclear Station, Unit 2 Docket No. 50-414 Cycle 2 Startup Report Gentlemen:

In accordance with Section 6.9.1. of the Catawba Nuclear Station Technical Specifications, please find attached the Unit 2 Cycle 2 Startup Report. This report is being submitted due to the installation of Wet Annular Burnable Absorber (WABA) rods for Unit 2, Cycle 2. Further information regarding WABA rods is provided in WCAP-10021 (Revision 1), Westinghouse Wet Annular Burnable Absorber Evaluation Report.

Very truly yours, 7

M 9~

llal B. Tucker PGL/35/sbn Attachment xc: Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. P. K. Van Doorn NRC Resident Inspector Catawba Nuclear Station (.

I k

. _ - -. . . - . _ - _. . ._. ._

ML20150E058

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