ML20045F228

Cycle 6 Startup Rept. W/

ML20045F228
Person / Time
Site:	Catawba
Issue date:	06/30/1993
From:	Rehn D DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9307070113
Download: ML20045F228 (28)

Text

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- Duke Fou er CompuV D L Rott Catretc Suclear Generanon Department Dce President Gf> ConcordRoad (gg3g,33g5 og;(,

Yo-k. SC 29 6 y93;gy;.3g5y, i

DUKEPOWER June 30,1993 j

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 t

Subject:

Catawba Nuclear Station, Unit 2 Docket No. 50-414 Startup Report, Unit 2 Cycle 6 -

In accordance with Section 6.9.1 of the Catawba Nuclear Station Technical Specifications, find attached the Unit 2 Startup Report for Cycle 6 core design.

Any questions concerning this report may be directed to Kay Nicholson at (803) 831-3237.

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Very truly yours, D. L. Rehn t

KEN /U2C6.SR l

Attachment i

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S. D. Ebneter Regional Administrator, Region II R. E. Martin, ONRR R. J. Freudenberger Senior Resident inspector i

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Catawba Nuclear Station Unit 2 Cycle 6 -

STARTUP REPORT l

i June,1993 l

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TABLE OF CONTENTS Pace ii List of Tables il List of Figures 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 Criticaldy 6

3.2 Source Range / Intermediate Range Overlap Data.

10 3.3 Point of Nuclear Heat Addition 10 3.4 Reactivity Computer Checkout

. 11 3.5 ARO Boron Endpoint Measurement 11 3.6 ARO Isothermal Temperature Coefficient Measurement 12 3.7 Reference Bank Worth Measurement by Dilution

. 12 3.8

. Reference Bank in Boron Endpoint Measurement 14 3.9 Differential Boron Worth Determinat'on 14 3.10 Control Rod Worth Measurement by Rod Swap.

14 4.0 Power Escalation Testing.

16 4.1 Core Power Distribution 16 4.2 One-Point incore/Excore Calibration..

.. 20 4.3 Reactor Coolant Delta Temperature Measurement.

.. 21 4.4 Hot Full Power Critical Boron Concentration Measurement.

22 4.5 Calorimetric Reactor Coolant Flow Measurement 22 4.6 incore/Excore Calibration 224 i

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LIST OF TABLES Page 1

I 1.

Core Design Data 2.

Preliminary NIS Calibration Data -

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3.

Summary of Zero Power Physics Testing Results'

.7 i

4.

Source Range / intermediate Range Overlap Data

. 10 11 5.

Nuclear Heat Determination 6.

Reactivity Computer Checkout 11 7.

ITC Measurement Results

. 12 15 i

B.

Control Rod Worth Measurement Data

l 9.

Core Power Distribution Results. 30% Power

. 17 10.

Core Power Distribution Results. 75% Power

. 18 l

11.

Core Power Distribution Results.100% Power

.. 19 f

12.

One-Point incore/Excore Calibration Results 20

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13.

Reactor Coolant Delta Temperature Data 21

+

14.

Calorimetric Reactor Coolant Flow Measurement Data.

23

. 24 15 incore/Excore Calibration Results l

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LIST OF FIGURES l

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1.

Core Loading Pattern

. 3

.2.

ICRR vs. Demin Water Added During Reactor Coolant System Dilution

.5

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Inverse Count Rate Ratio vs. Control Rod Worth During Approach to Criticality..

. 8'

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4.

Renormalized ICRR vs. Control Rod Worth During Approach to Criticality 9

5.

Integral and Differential Worth of Reference Bank 13.

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Page 1 of 24 j

1.0 INTRODUCTION

Catawba Unit Two Cycle 6 includes a feed batch of 76 MkBW fuel assemblies manufactured by B&W Fuel Company (BWFC). All fuel used in previous cycles were optimized fuel assemblies manufactured by Westinghouse. Burnable poison rod assemblies used in the feed batch were also manufactured by BWFC.

The Catawba Unit Two Cycle 6 feed batch is the sixth batch provided by BWFC for Duke's Westinghouse units. Catawba Unit Two is the last Westinghouse reactor on the Duke Power system to use BWFC fuel.

t McGuire Units One and Two and Catawba Unit One have used MkBW fuel for at least one fuel cycle.

i Catawba Unit Two Cycle 6 core loading began at 2020 on March 1,1993 and ended at 1505 on March i

4,1993. Initial criticality tor cycle 6 occurred at 1855 on March 30,1993. Zero Power Physics Testing ended at 2255 on March 31,1993. The unit reached full power at 1320 on April 5,1993. Power escalation testing. including testing at full power, was completed by April 9,1993.

i Table 1 contains some important characteristics of the Catawba 2 Cycle 6 core design.

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i TABLE 1 C2C6 CORE DESIGN DATA 1.

C2C5 end of cycle bumup: 371.3 EFPD

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C2C6 design length: 380 10 EFPD l

l Region Fuel Type Number of Enrichment, Loading, MTU Cycles Assemblies w/o U "

Burned l

2 5 (O)

OFA B

3.613 3.403 3

I 6(R)

OFA 33 3 507 14.014 2

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7 (S)

OFA 76 3.758 32.279 1

J 8 (T)

MkBW 76 3.747 34.646 0

r Totals 193 84.342

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2.0 PRECRITICAL TESTING Precritical testing includes:

+ core loading

• preliminary calibration of nuclear instrumentation

+ dilution of reactor coolant. system to estimated critical boron concentration r

Sections 2.1 through 2.3 describe results of precritical testing for Catawba 2 Cycle 6.

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2.1 Total Core Reloading r

The cycle 6 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 2020 on March 1,1993 and conc 8uded at 1505 on March 4,1993. Core loading was verified by PT/0/N4150/03C, Core Verification, which was completed by 2300 on March 4,1993.

Figure 1 shows the core loading for Catawba 2 Cycle 6.

2.2 Preliminary NIS Calibration l

Periodic test procedure PT/0/N4600/05E, Preliminary Nls Calibration, is performed before initial criticality for each new f"el cycle. Intermediate range reactor trip and rod stop setpoints are adjusted using measured pv..e distribution from the previous fuel cycle and predicted power distribution for the upcoming fuel cycle. Power range full power currents are similarly adjusted. Intermediate range calibration data is j

t checked and revised as necessary during power escalation testing.

-l Table 2 shows the calibration data calculated by PT/0/N4600/05E. Calculations were performed on March 10,1993. Calibrations were complete by March 29,1993.

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2.3 Reactor Coolant System Dilution l

)

The reactor coolant system boron concentration was diluted from the refueling 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 demineralized water added.

l Initial reactor coolant boron concentration was 2111 ppmB. The estimated critical boron concentration was j

. calculated to be 1655 ppmB. The calculated volume of demineralized water required was 14790 gallons.

This change in boron concentration was expected to decrease ICRR from 1.0 to 0.6.

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Reactor coolant system dilution began at 0457 on March 30,1993 and concluded at 0801 on March 30,

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393. The final reactor coolant system boron concentration, after allowing system to mix, was 1656 ppmB.

l ICRR behavior during dilution was as expected; final ICRR was within 0.1 of expected value of 0.6. Figure '

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' 2 shows ICRR versus volume of water used.

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t Page 3 of 24 FIGURE 1 North CORE LOADING PATTERN, C ATAWBA UNIT 2 CYCLE 6 4

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180 FUELTRANSFER CANAL ->

1 - - - - - - - - - - - - -

S33 T06 S76 S25 S62 T31 S01 PD PD PD PD PD PD PD 2------

S30 518 T36 S34 T55 SS6 T57 S21 T03 SOS S54 PD RCCA 4 BP RCCA 12 BP RCCA 12 BP RCCA 4 BP RCCA PD 3----

S64 552 T71 S47 T13 S61 558 S35 T48 S15 T50 S23 SS1 PD PD 4 BP RCCA 16 BP RCCA PD RCCA 16 BP RCCA 4 BP PD PD' I

4____

$32 T41 R48 T46 R33 T30 R18 T33 R06 T01 R24 T54 S14 RCCA 4 BP RCCA 12 BP PD 12 BP RCCA 12 BP PD 12 BP RCCA 4 BP RCCA 7

5-S63 T28 S43 T15 032 T64 R02 S70 R04 T62 Q16 T52 S42 T76 S75 j

PD 4 BP RCC.A 12 BP PD 8 BP PD PD PD 8 BP PD 12 BP RCCA 4 BP PD.

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T51 S07 T47 R16 T60

-R44 T14 060 T66 R21 T12 R60 T32 S39' T59 PD RCCA 16 BP PD 8 BP RCCA 8 BP RCCA 8 BP RCCA 8 BP PD 16 BP RCCA PD S29' T39 S45 T63 R35 T65 R34 T75 R23 T07 R11 T74 S69 T08 S10

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PD 12 BP RCCA 12 BP PD 8 BP PD 8 BP PD 8 BP PD 12 BP RCCA 12 BP PD I

S59 S11 S38 Roa S19 O29 T22 R68 T11 066 S60 R56 574 S22 S40 g_

PD ACCA

.SS RCCA PD RCCA 8 BP RCCA 8BP RCCA PD RCCA SS RCCA PD 270 l

90 S66 T23 S28-T38 RS3 T37 R03 T16 R59 T56 R12 T44 S46 T67 P12 9_

PD 12 BP RCCA 12 BP PD B BP PD 8 BP PD 8 BP PD 12 BP RCCA 12 BP PD t

ido 567 T17 R46 T29 R36 T10 037 T09 RS2 T19 R14 T53 S37 T68 to _

PD RCCA 16 BP PD 8 BP RCCA 8 BP RCCA 8 BP RCCA 8 BP PD 16 BP RCCA PD S16 T20 S55 T04 Oss T02 R28 S36 R41

. T27 069 T73 S71 T69 S44 11 -

PD 4 BP RCCA 12 BP PD 8 BP PD PD PD 8 BP PD 12 BP RCCA 4 BP PD 12 - - - -

SO9 T72 RS3 T26 RDS T24 R29

. T49 R32 Td5

-R49 T35 S17 I

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RCCA 4 BP RCCA 12 BP PD 12 BP RCCA 12 E P PD 12 BP RCCA 4 BP RCCA i

13 - - - -

S68 SOB T18 S73 TOS S41 S31 S13 T61 S53 T43 SO4 S27 I

t PD PD 4 BP RCCA 16 BP RCCA PD RCCA 16 BP RCCA 4 BP PD PD 1

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S72 S48 T34 S57 T21 S65 T42 S03 T70 S50 S49

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RCCA 4 BP RCCA PD I

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S26 T58 S02 SOS

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TABLE 2 PRELIMINARY NIS CALIBRATION DATA Intermediate Range Ratio Cycle 5 Cycle 5 BOC 6 BOC 6 Channel (BOC 6 +

Reactor Trip Rod Stop Reactor Trip Rod Stop Cycle 5)

Setpoint, Setpoint,

Setpoint, Setpoint, pAmps pAmps pAmps pAmps N35 0.8657 93 77 81 65 N36 0.8691 80 67 70 56

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Ratio Axial Cycle 5 Full Power BOC 6 Full Power

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Channel (BOC 6 +

Offset, %

Current, pAmps Current, pAmps Cycle 5)

Upper Lower Upper Lower I

+20 309.0 235.4 366.0 278.9 N41 1.1846 0

270.4 273.8 320.3 324.3'

-20 231.8 312.3 274.6 370.0

+20 239.4 188.6 ~

293.4 231.2 f

N42 12257 0

209.8

_219.3

.220.9 306.5 257.1 268.8

-20 180.2 250.1 6

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+20 269.7 212.7 325.8 256.9 i

N43 1.2079 0

234.0 247.0-282.7 298.4

-20 198.3 281.2 239.5

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+20 276.6 221.1 331.6 265.1 i

N44 1.1989 0

241.1 255.6

~ 289.1 306.4

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-20 205.5 '

290.0 246.4 347.7 h

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l Page 6 of 24 3.0 ZERO POWER PHYSICS TESTING I

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Zero Power Physics Testing (ZPPT) is performed at the beginning of each cycle and is controlled by

-l PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. Test measurements are made below 6

the point of nuclear heat using the output of one power range detector connected to a reactivity computer.

j Measurements are compared to predicted data to verify core design. The following tests / measurements _

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are included in the ZPPT program:

+1/M Approach to Criticality

. Measurement of point of adding heat

. Reactivity computer checkout E

.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 j

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.Ditterential boron worth determination l

. Control rod worths measurement by Rod Swap

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Zero power physics testing for Catawba 2 Cycle 6 began at 1527 on March 30,1993 with the beginning l

of rod withdrawal for approach to criticality. ZPPT ended at 2255 on March 31,1993 following analysis of 1

rod swap data. Table 3 summarizes results from ZPPT. All acceptance criteria'were met.

i Sections 3.1 through 3.10 describe ZPPT measurernents and results.

l 3.1 1/M Approach to Criticality l

I initial criticality for Catawba 2 Cycle 6 was achieved per PT/0/A/4150/19,1/M Approach to Criticality. In j

this procedure, Estimated Critical Rod Position (ECP) is calculated based on latest available reactor coolant l

boron concentration. Control rods, beginning with shutdown banks in normal sequence, are withdrawn until i

Boron Dilution Mitigation System (BDMS) count rate doubles. Inverse Count Rate Ratio (ICRR) is plotted -

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for each source range and BDMS channel. ICRR data is used to project critical rod position. If projected

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critical rod position is acceptable, rod withdrawal may continue.

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Rod withdrawal for the approach to criticality began at 1527 on March 30,1993. Rod withdrawal was interrupted briefly at 1723 due to " Control Rod Bank Lo Limit" annunciator not clearing properly. Criticality was achieved at 1855 on March 30,1993 with Control Bank D at 30 steps withdrawn.

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Figure 3 shows the ICRR plots that were used during the approach to criticality. Figure 4 is ICRR data

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renormalized to 1 after the first three doublings. All acceptance criteria of PT/0/N4150/19 were met.

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1 Page 7 of 24 TABLE 3

SUMMARY

OF ZPPT RESULTS PREDICTED VAL-PARAMETER MEASURED VALUE UE/ ACCEPTANCE CRITERIA l

Nuclear Heat 6.6 x 10-7 amps (N44)

N/A I

ZPPT Test Band 10' to 10~7 amps (N44)

N/A ARO Critical Boron 1747 ppmB 1715 50 ppmB AROITC

+0.42 pcm/ F

+0.77 2 pcm/'F ARO MTC

+1.81 pcm/*F

+2.16 pcm'*F Reference Bank (Shutdown 975.5 pcm 988 148 pcm Bank B) Worth Ref. Bank in Critical Boron 1626 ppmB 1598 ppm 5 Differential Boron Worth

-8.06 pcm/ppmB

-8.44 1.27 pcm/ppmB Control Bank D Worth 467 pcm 485 200 pcm

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Control Bank C Worth 798 pcm 846 254 pcm -

Control Bank B Worth 778 pcm 781 234 pcm Control Bank A Worth 258 pcm 285 200 pcm i

Shutdown Bank E Worth 356 pcm 3811200 pcm Shutdown Bank D Worth 447 pcm 466 200 pcm Shutdown Bank C Worth 441 pcm 466 200 pcm Shutdown Bank A Worth 285.5 pcm 311 200 pcm j

Total Rod Wor 1h 4807 pcm 5009 501 pcm t

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Page 8 of 24 FIGURE 3 ICRR vs. CONTROL ROD WORTH DURING APPROACH TO CRITICALITY 1

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10 20 30 40 50 60 70 80 90 100 Rod Worth, % Withdrawn BDMS A BDMS B o

N31 N32 ECP k

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rod insertion limit u-m+

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Page 9 of 24 FIGURE 4 RENORMAllZED ICRR vs. CONTROL ROD WORTH DURING APPROACH TO CRITICAllTY 1

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0 65 70 75 80 85 90 95 100 Rod Worth, % Withdrawn BMDS A BDMS B A

N31 N32 ECP M

lower limit M

upper limit K

rod insertion limit

Page 10 of 24 3.2 Source Range / intermediate Range Overlap Data During the initial approach to criticality, source range and intermediate tange data was obtained to verify that at least one decade of overlap existed. If one decade of overlap did not exist, intermediate range compensation voltage would have been adjusted to provide the overlap.

6 Overlap data for Cycle 6 was obtained per PT/0/A/4150/01, Controlling procedure for Startup Physics i

Testing, on March 30,1993. Table 4 contains the overlap data. The acceptance criterion was met.

TABLE 4 SOURCE RANGE / INTERMEDIATE RANGE OVERLAP DATA SOURCE RANGE INTERMEDIATE RANGE 4

N31, cps N32, cps N35, amps N36, amps INITIAL DATA:

NIS Meters 300 150 1 x 10 '"

1 x 10 '"

OAC 361 174 1.007 x 10'"

1.050.x 10'"

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FINAL DATA:

NIS Meters 10,500 10,000 1 x 10

• 1 x 10 "

OAC 10.520 11,960 1.076 x 10

• 1,087 x 10 "

t 3.3 Point of Nuclear Heat Addition The point of nuclear heat addition is measured by trending reactor coolant system temperature, pressurizer level, llux level, and reactivity while slowly increasing reactor power. A slow, constant startup rate is initiated by rod withdrawal. An increase in reactor coolant systern 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 6, the point of nuclear heat addition was determined per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing, on March 30,1993. Table 5 summarizes the data obtained.

7 The zero power physics test band was set at 10+ to 10 amps on power range channel N44 (connected to ieactivity computer). This test band provided more than a factor of two margin to nuclear heat for zero -

j power physics testing. Acceptance criterion was satisfied.

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TABLE 5 i

NUCLEAR HEAT DETERMINATION Reactivity Computer Intermediate Range Intermediate Range (N44), amps Channel N35, amps Channel N36, amps RUN #1 6.6 x 10~7 6.792 x 10~7 5.971 x 10~7 RUN #2 6.8 x 10~7 6.699 x 10 5.998 x 10'7 7

3.4 Reactivity Computer Checkout t

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. Re,tivity insertions of approximately +25, -25, +40, 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 venfied to be within 4.0%

The checkout was performed for Cycle 6 on March 30,1993. 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 Measured Absolute Error, Percent Error,%

Reactivity, pcm Reactivity, pcm pcm 206.58 31.77 31.24 0.53 1.668

-341.83

-24.97

-24.40 0.57 2.283 135.15 45.53 44.76 0.77 1.691

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-242 99

-37.18

-36_.36 0.82 2.205_

i 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 i

system boron samples are obtained while control bank D is pulled to the fu!!y withdrawn position. The l

reactivity difference from criticality to the ARO configuration is measured and conver1ed 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.

l The Cycle 6 beginning of cycle, hot zero power, all rods out, critical boron concentration was measured on March 31,1993 per PT/0/A/4150/10, Boron Endpoint measurement. The measured boron concentration (average of 5 samples) was 1745 ppmB. This value was adjusted by 2 ppmB to yield an ARO concentration of 1747 ppmB. Predicted ARO critical boron concentration was 1715 ppmB. The acceptance 7

criterion, measured boron within 50 ppmB of predicted, was met.

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i Page 12 of 24 3.6 ARO Isothermal Temperature Coefficient Me2.cmement i

1 The all rods out (ARO) isothermal temperature coetticient (lTC) is measured at the beginning of each cycle to venty consistency wrth predicted value. In addition, the moderator temperature coefficient (MTC) is obtained by subtracting the doppler ternperature coefficient from the ITC. The MTC is used to ensure compliance with Technical Specification limits.

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i To measure the ITC, a slow (< 20cF/ 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 j

heatup is performed while again trending reactivity versus temperature. The slopes of the Reactivity versus temperature lines are used to determine the ITC. The cooldown/heatup cycle is repeated if additional data r

is required l

1 The beginning of cycle 6 ITC was measured per PT/0/A/4150/12A, isothermal Coefficient of Reactivity I

Measurement, on March 31,1993. 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,

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Average ITC was +0.42 pcm/'F. Predicted ITC was +0.77 pcmPF. Measured ITC was within acceptance f

cnterion of predicted ITC 12 pcnf F.

i The MTC was determined to be +1.81 pen #F. This value was used with procedure PT/0/A/4150/21, l

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.

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TABLE 7 ITC MEASUREMENT RESULTS I

ITC, pcm/ F AT,<F Ap. pcm T ",'F Uncorrected Corrected to 557'F i

Cooldown

-4

-2.62 554.0 0.66 0.37

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Heatup

+5 43.55 554.5 0.71 0.47

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Average:

0.42 1

3.7 Reference Bank Worth Measurement by Dilution The control rod bank predicted to have the highest worth is designated the reference bank and is measured f

by inserting the bank (with all other rod banks fully withdrawn) :n 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.

i The beginning of cycle 6 reference bank (Shutdown Bank B) worth was measured on March 31,1993 per PT/WA/4150/11 A, Control Rod Worth Measurement by Boration/ Dilution. Figure 5 shows integral worth of reference bank versus bank position. The reference bank was measured to be worth 975.5 pcm; predicted l

worth was 988 pcm The acceptance criterion, measured worth within 15% of predicted, was met.

Page 13 of 24 '

l FIGURE 5 l

INTEGRAL AND DIFFERENTIAL WORTH OF REFERENCE BANK 1000 10 900 9

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0 50 100 150 200 250 Reference Bank Position, steps withdrawn Integral Worth Predicted Worth Differentlol Worth Predicted Diff. Worth

_ _ _ _-___ - _-_-______- - - _ ___ _-_-=

Page 14 of 24 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 fuliy 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 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 3

concentration is adjusted accordingly to obtain the " Reference Bank in" critical boron concentration.

The Cycle 6 beginning of cycle, hot zero power, reference bank in, critical boron concentration was measured on March 31.1993 per PT/0/A/4150/10, Boron Endpoint measurement. The measured boron concentration (average of 5 samples) was 1631 ppmB. This value was adjusted by -5 ppmB to yield a

" Reference Bank in" concentration of 1626 ppmB. Predicted " Reference Bank in" critical boron concentration was 1598 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 entical boron concentration, and total reactivity worth of reference bank. The calculated value is compared to predicted value to venty consistency. This calculation also provides an indirect check of measured reference bank worth and of the boron endpoint measurements.

The beginning of Cycle 6, hot zero power differential boron worth was calculated to be -8.06 pcm/ppmB per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. The predicted value was -8.44 pcm/ppmB. The acceptance criterion (measured within 15% of predicted), was met.

3.10 Control Rod Worth Measurement by Rod Swap The wc,rths of all control rod banks except the reference bank are measured by inserting each bank while withdrawing the reference bank and/or previously measured bank to maintain near critical conditions.

When the bank being measured is fully inserted, the reference 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 ta verify cons.istency. The sum of the worths of all banks, including the reference bank, is also compared to predicted total.

The beginning of cycle 6 rod worth measurement by rod swap was performed on March 31,1993 per PT/0/A/4150/11B, Cortrol Rod Worth Measurement by Rod Swap. Table 8 summarizes the results. All acceptance enteria were met.

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Page 15 of 30 TABLE 8 CONTROL ROD WORTH MEASUREMENT DATA Adjusted Cptical Remaining Difference S Diff. (Pred -

Reference Position of Worth of Ref.

Measured Predtcted (Predicted -

Meas}'Pred x Acceptance Bank Bank Worth Ref. Bank Bank A!rha Worth. ppm Werth. ogr1 Yoasured) 100 Catena Shutdown B 988 i 148 (ref. bank)

N/A N/A N/A N/A 975.5 988 12.5 1.3 (15%)

285k 200 Control A 936.5 91.5 637 1.065 258.0 285 27.0 9.5 pcm 311 i 200 Shutdown A 936.5 93 628 1.037 285.5 311 25.5 8.2 pcm 381 i 200 Shutdown E 936.5 89.5 650 0.894 355.7 381 25.3 6.6 pcm 1

466t200 Shutdown D 936.5 118 480 1.020 447.1 466 18.9 4.1 pcm 466 i 200 Shutdown C 936.5 117 485 1.020 441.4 466 24.6 5.3 pcm 485 1 200 Control D 936.5 127.5 428 1.098 466.9 485 18.1 3.7.

pcm -

781 i 234 Control B 936.5 176.5 186 0.848 778.4 781 2.6 0.3 (30%)

l 8461254 Control C 936.5 183 156 0.887 798.1 846 47.9 5.7 (30%)

>5009 -

Total 4806.6 5009 202.4 4.0 501 pcm

.. :.. = :.

.. -..... ~...

... ~.

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4.0 POWER ESCALATION TESTING

-[

Power escalation testing is performed during the initial power increase to full power for each cycle and is controlled by PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. Tests are performed from 0% through 100% power with major testing plateaus at 30%,75%, and 100% power.

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Significant tests performed during power escalation are:

• Core Power Distribution (at 30%,75%, and 100% power)

One-Point incore/Excore Calibration (at 30% power)

Reactor Coolant Delta Temperature Measurement (at 75% and 100% power)

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. 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) i In adddion 1o the tests listed above, 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 repor1.

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Power escalation testing for Catawba 2 Cycle 6 began on April 1,1993. Full power was reached on April'-

t 5,1993. On April 6,1993 power was reduced to 85% power for auxiliary feedwater system testing. The unit returned to ful! power on April 7,1993. Full power testing was completed on April 9,1993.

Sections 4.1 through 4.6 describe the significant tests performed during power escalation and their results.

j 4.1 Core Power Distribution Core power distribution measurements are performed during power escalation at low power (approximately -

30%), intermediate power (approximately 75%), 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).

j 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 j

MONITOR code, respectively).

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~ The Catawba 2 Cycle 6 core power distribution measurements were performed on April 2, 1993 (30%

power), April 4,1993 (75% power), and April 8,1993 (100% power). Table 9 through 11 summarize the j

results. All acceptance enteria were met.

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Page 17 of 24 TABLE 9 CORE POWER DISTRIBUTION RESULTS 30% POWER i

Plant Data i

Map ID:

FCM/2/06/01 Date of Map:

April 2,1993 j

I Cycle Bumup:

0.204 EFPD Power Level:

29.76 %

Control Rod Position:

Contro! Bank D at 215 steps withdrawn 4

Reactor Coolant System Boron Concentration:

1611 ppmB

\\

CORE Results Core Average Axial Offset:

17.364 %

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Tilt Ratios for Entire Core Height: Quadrant 1:

1.00882 Ouadrant 2:

1.02013 l

f Quadrant 3:

0.98635 Ouadrant 4:

0.98470 Maximum F (nuclear):

2.051 o

Maximum F,s (nuclear):

1.520 Maximum Error between Pred. and Meas Fw 7.92 %

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Average Error between Pred. and Meas. Fw 2.00%

Maximum Error between Expected and 7.80 %

i Measured Detector Response:

RMS of Errors between Expected and Measured 2.80 %

Detector Response:

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MONITOR Results

'l Minimum F Operational Margin:

25.12 %

a Minimum F RPS Margin:

11.58 %

a Minimum F LCO Margin:

51.24 %

a Minimum F Surveillance Margin:

35.83 %

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Minimum F LCO Margin:

19.17%

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Page 18 of 24 j

TABLE 10 CORE POWER DISTRIBUTION RESULTS 75% POWER i

Plant Data

~

Map ID:

FCW2/06/02 Date of Map:

April 4,1993 Cycle Burnup:

1.046 EFPD Power Level:

75.30 %

F Control Rod Position:

Control Bank D at 215 steps withdrawn Reactor Coolant System Boron Concentration:

1448 ppmB CORE Results Core Average Axial Offset:

3.240 %

Tilt Ratios for Entire Core Height: Quadrant 1:

1.01124 Quadrant 2:

1.01481 l

Quadrant 3:

0.98296 Quadrant 4:

099099 Maximum F (nuclear):

1.730 a

Maximum F,a (nuclear):

1.459 i

Maximum Error between Pred. and Meas F,g:

8.11 %

Average Error between Pred. and Meas. Fw 1.84 %

Maximum Error between Expected and 7.90%

Measured Detector Response:

RMS of Errors between Expected and Measured 2.60%

Detector Response:

MONITOR Results

-1 F

)

Minimum F Operational Margin:

12.75 %

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l Minimum F RPS Margin:

16.17%

a Minimum F LCO Margin:

39.15 %

o Minimurn F, Surveillance Margin:

17.60%

Minimum F,a LCO Margin:

15.68%

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Page 19 of 24 TABLE 11 CORE POWER DISTRIBUTION RESULTS 100% POWER Plant Data Map ID:

FCM/2/06/04 Date of Map:

April 8,1993 Cycle Burnup:

4.856 EFPD Power Level:

99.69 %

1 Control Rod Position:

Control Bank D at 202 steps withdrawn Reactor Coolant System Boron Concentration:

1223 ppmB CORE Results Core Average Axial Offset:

' 1.605 %

l Titt Ratios for Entire Core Height: Quadrant 1:

1.00668 Quadrant 2:

1.01553 Ouadrant 3:

0.98304 J

Quadrant 4:

0.99474 Maximum F (nuclear):

1.715 o

Maximum F,a (nuclear):

1,435 Maximum Error between Pred. and Meas Fs 5.60 %

Average Error between Pred. and Meas. Fs 1.56%

Maximum Error between Expected and 5.50%

Measured Detector Response:

1 RMS of Errors between Expected and Measured 2.20%

Detector Response-1 MONITOR Results Minimum F Operational Margin:

3.38 %

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Minimum F RPS Margin:

15.10%

o Minimum F LCO Margin:

20.57*'

o Minimurn F,,s Surveillance Margin:

4.43 %

Minimum F,y LCO Margin:

11.51 %

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4.2 One-Point incore/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 i

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 75% 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 6 was obtained on April 2,1993 and all power range calibrations were completed later that day. Results are listed in Table 12. All acceptance criteria were met.

TABLE 12 ONE-POINT INCORE/EXCORE CALIBRATION RESULTS

{

l Reactor Power = 29.76%

Axial Offset = 17.364%

Measured Power Range Currents, pAmps N41 N42 N43 N44 Upper 86.8 70.7 81.8 74.3 Lower 72.8 62.0 70.8 65.2 I

Ratio, Extrapolated (from measured) Currents to

• Expected" (from last calibration) Currents N41 N42 N43 N44

+

Upper 1.0253 1.0849 1.1103 0.9992 Lower 1.0780 1.1451 1.1515 1.0369 t

l New Calibration Currents, pAmps Offset..

Axial N41 N42 N43 N44 Upper Lower Upper Lower Upper Lower Upper Lower L

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+20 296S

' 239.5 241.5 204.1 279.8 232.9 254.0 214.6 i

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259.6 278.7 211.7 237.3 242.8 270.5 221.4 248.1 i

-20 222.6 317.8 181.8 270 6 205.7 307.9 188.7 281.5 s

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Page 21 of 24 I

4.3 Reactor Coolant Loop Delta Temperature Measurement Reactor coolant system (NC) hot leg and cold leg temperature data is obtained at 75% and 100% power j

per PT/0/A/4600/26, NC Temperature Calibration, to ensure that full power delta temperature constants 4

(AT ) are valid. AT is used in the overpower and overtemperature detta temperature reactor protection o

o functions.

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i PT/0/A/4600/26 was performed at 75% power on 4/4/93 and at 100% power on 4/6/93. Loops A and B were recalibrated based on results obtained at 75% power. All 4 loops were calibrated using full power results. Table 13 summarizes the test results.

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TABLE 13 REACTOR COOLANT DELTA TEMPERATURE DATA 1

I Reactor Power = 75.95%

1 i

Loop A Loop B Loop C Loop D l

Meas. T, "F 600.8 602.9 602.4 603.2 l

Meas, Tw w.'F 556.2 557.0 556.9 557.5 I

Calc. Ah, BTU /lb 59.46 61.46 60.87 61.27 Calc. Ah. BTU /lb 78.29 80.93 80.15 80.67 o

1 Calc. aT, 'F 56.59 58.30 57.79 58.13

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Current AT, 'F 58.08 59.30 57 43 57.77 l

s Difference, ' F

-1.49

-1.00

+0.36

+ 0.36 J

d Reactor Power = 99.59%

m-j Loop A Loop B Loop C Loop D i

Meas. Tc, "F 616.3 618.6 618.1 619.1 Meas. T

• F 558.9 560.0 559.8 560.7 mm, Calc. Ah, BTU /!b 78.95 81.11 80.58 81.00 Calc. ah, BTU /lb 79.28 81.45 80.91 81.34 3

Calc AT. ' F 57 61 58.82 58.51 58.62 o

Current AT. 'F 56.59 58.30 57.43 57.77 o

Diff erence, ' F

+ 1.02

+0.52

+ 1.08

+0.85 N

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=+

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Page 22 of 24 l

4.4 Hot Full Power Critical Boron Concentration Measurement The hot full power cntical boron concentration is measured using PT/0/A/4150/04, Reactivity Anomaly i

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 i

any off-reference condnion (e.g. inserted Tod worth, temperature error, difference from equilibrium xenon) l and compared to predicted value.

For the purposes of Startup Physics testing, the predicted critical boron concentration is adjusted for the

]

diNerence 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 j

acceptance criterion ( 50 ppmB).

l For Catawba 2 Cycle 6, the hot full power critical boron concentration was measured on April 8,1993. The measured critical boron concentration was 1225.0 ppmB. Predicted critical boron concentration,was 1228.7

-}

ppmB; when adjusted for difference at zero power, the adjusted predicted critical boron concentration was

.l 1260.7 ppmB. The ditference between measured and adjusted predicted critical boron concentration was 35.7 ppmB, which met the acceptance criterion.

.l 4.5 Calorimetric Reactor Coolant Flow Measurement I

Reactor coolant flow is measured using a precision calorimetric based on secondary side parameters i

(feedwater 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

.l Technical Specification surveillances on reactor coolant flow and in the primary side calculation of thermal j

power.

For Catawba 2 Cycle 6, the calorimetric flow measurement was performed on April 7,1993, per PT/2/A/4150/138. 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.

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Page 23 of 24 -

TABLE 14 i

CALORIMETRIC REACTOR COOLANT FLOW MEASUREMENT DATA l

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Test Run Reactor Calculated Reactor Coolant Flow Rate, gpm

% of Tech Number Power, %

Spec Flow Loop A Loop B Loop C Loop D Total 1

99.229 96,666 96,855 98.562 97,839 389,922 101.278 t

2 99.104 96,709 96,872 98,676 97,853 390,111 101.327 3

99.095 96,708 96,789 98,734 97,856 390,087 101.321 i

l Average of total reactor coolant flow for the three test runs is 390,040 gpm.

i Reactor Coolant Elbow Tap Correction Factor Flow Channel Loop A Loop B Loop C Loop D 1

0.304526 0.297966 0.310000 0.294938 j

2 0.292234 0.281457 0.291225 0.294235 i

3 0.300312 0.300864 0.298838 0.296512

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4.6 incore/Excore Calibration Excore power range channels are calibrated at full 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.

s This test was performed for Catawba 2 Cycle 6 on April 8 and April 9,1993. All power range calibrations were completed on April 9,1993. Nine flux maps, with axial offset ranging from -11.853% to +0.380% were used.' Table 15 summarizes the tesults. All acceptance criteria were met.

P TABLE 15 INCORE/EXCORE CAllBRATION RESULTS Full Power Currents, Microamps Axial N41 N42 N43 N44 I

Offset.

Upper Lower Upper Lower Upper Lower Upper Lower

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+20%

338.8 259.4 273.5 217.0 317.5 250.1 291.0 229.9 4

i 0%

293.9 303.6 238.8 254.2 272.8 291.3 252.3 267.4 20%

249.0 347.7 204.2 291.3 228.2 332.5 213.5 304.8 l

t Correction (M ) Factors 3

l N41 N42 N43 N44 1.341 1.374 1.311 1.362

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