ML20247G403

Cycle 3 Startup Rept. W/

ML20247G403
Person / Time
Site:	Catawba
Issue date:	08/31/1989
From:	Tucker H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8909190060
Download: ML20247G403 (55)

Text

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T.p 6 Duxn POWER GourAwv P.O. Box 33189 L' CHARLOTTE, N.C. 28242 v- HALH. TUCKER ren.ewsoxe

{ ,

vine emenmen mmm. .o (704) 073-4 Sat September 7, 1989 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 CJcle 3 Startup Report Gentlemen:

Pursuant to Catawba Nuclear Station Technical Specification 6.9.1.3, please find attached the Catawba Unit 2 Cycle 3 Startup Report. The cycle 3 core includes three demonstration assemblies manufactured by Babcock and Wilcox Fuel Company.

Very truly yours, d.k w Hal B. Tucker Attachment JGT/4/U2C3 STAR xt: Mr. S. D. Ebneter Regional Administrator, Region II U. S. Nuclear Regulatory Commission 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station 8909190060 890$$/ I i PDR ADOCK 0500v413 1 p FDC

d4 .

DUKE P0k'ER COMPANY CATAk'BA NUCLEAR STATION UNIT 2 CYCLE 3 STARTUP REPORT AUGUST, 1989

, o, TABLE OF CONTENTS Page List of, Tables ............... ................................. iii List of Figures ................................................ iv 1.0 Introduction .............................................. I 2.0 Precritical Testing .................. .................... 3 2.1 Total Core Reloading - PT/2/A/4150/22 ............... 4 2.2 Refueling ENB Calibration - PT/2/A/4600/05E ......... 6 2.3 1/M Approach To Criticality - PT/2/A/4150/19 ........ 9 3.0 Zero Power Phj sics Testing ........................... .... 12 3.1 Source Range / Intermediate Range Overlap Data -

PT/2/A/4150/21 ...................................... 14 3.2 Point of Nuclear llent Addition - PT/2/A/4150/21 ..... 16 3.3 Reactivity Computer Checkout - PT/2/A/4150/21 ....... 18 3.4 ARO Boron Endpoint Measurement - PT/2/A/4150/10 .... 20 3.5 ARO Isothermal Temperature Coefficient Measurement -

PT/2/A/4150/12A ..................................... 21 3.6 Reference Bank Worth Measurement by Dilution PT/2/A/4150/11A .................................... 23 3.7 Reference Bank in Boron Endpoint Measurement -

PT/2/A/4150/10 ...................................... 25 3.8 Differential Boron Worth Determination -

PT/2/A/4150/21 ............. ........................ 26 3.9 Control Rod Worth Measurement by Rod Swap -

PT/2/A/4150/11B ... .......... ...................... 27 4.0 Power Escalation Testing .................................. 29 4.1 Core Power Distribution - PT/2/A/4150/05 ............ 30 4.2 Interim Incore/Excore Calibration - PT/2/A/4600/05D . 35 4.3 Delta Temperature Extrapolation - PT/2/A/4150/16 .... 37 4.4 Incore/Excore Calibration - PT/2/A/4600/05G ... ... 39 i

9 TABLE OF CONTENTS (Continued)

Page l 4.5 Hot Full Power Critical Boron Concentration -

PT/2/A/4150/04 ............. .................-...... 42 4.6 Incore/Excore Calibration Check - PT/2/A/4600/05B ... 43 4.7 Calorimetric Reactor Coolant Flow Measurement -

PT/2/A/4150/13B ....... ............................. 46 4.8 Transient Test for Steam Generator Level Tap Modification - TT/2/A/9200/58 .... .................. 48 l

l l

l 11

23 .-

i -*-

LIST OF TABLES Page

1. 17 x 17. Hybrid'B C RCCA Dimensional Informat' ion 2 4

2. Preliminary Calibration Data for Excore Instrumentation' 7

3. Summary cf ZPPT Results 13

4. Source Range / Intermediate Range Overlap Data 15

5. Nuclear Heat Determination 17

6. Reactivity Computer Checkout 19

7. ITC Measurement Results 22

8. Control Rod Worth Measurement Data 28

9. Core Power Distribution Results, 30% Power 31

'10 Core Power Distribution Results, 80% Power 32

'11. Core Power Distribution Results, 98% Power 33

12. Core Power Distribution Results, 100% Power 34

13. Interim Incore/Excore Calibration Results 36

14. AT Extrapolation Data 38

15. Incore/Excore Calibration Data 40

16. Incore/Excore Calibration Results 41

17. Incore/Excore Calibrat cn Check Results 8 44

18. Full Power Interim Inc.re/Excore Calibration Results 45

19. Calorimetric Reactor Coolant Flow Measurement Data 47 iti

_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . -- _ _ _ . _ _ _ _ I

LIST OF FIGURES

, Page I l

l 1. Core Loading Pattern, Catawba Unit 2 Cycle 3 5

2. Assemblies Selected For Preliminary Excore Calibration 8

3. 1/M Approach to Criticality - ICRR vs. Water Addition 10

4. 1/M Approach to Criticality - ICRR vs Control Bank Worth 11'

5. Roference Bank Integral Rod Worth Measured by Dilution 24

6. Stea o...erator Response to Transient Test 49 j i

iv

- _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ . _ - - _ - _ _ _ _ - _ - - - - _ _ . - - _ _ _ . _ _ _ - _ __-__.____m _ --__---_ _-- - - - __ __ - _ _ . _ _ - _ _ _ _ . _ _ _

4

• Page 1 2

i:

)

1.0 INTRODUCTION

l Catawba Unit 2 Cycle 3 core loading began at 1310 on April 26, 1989 and l concluded at 0830 on April 30, 1989. The core loading includes three Rod Cluster Control Assemblies (RCCAs) manufacturer by Babcock and Wilcox Fuel Company (BWFC). These three " hardened" RCCAs (2 Armaloy Plated and one Chromium Carbide Coated) are part of a demonstration program that will demonstrate the interface compatibility and successful operation l experience with BWFC supplied 17 x-17 RCCAs. The program will also l define and compare the operational wear characteristics of wear resistant RCCA coatings and conventional RCCA clad materials. Table 1 shows dimensional information of the BWFC counted RCCAs and a standard Westinghouse RCCA.

Criticality was achieved at 1729 on June 6, 1989. Zero Power Physics Testing (ZPPT) was completed at 2300 on June 7, 1989. Power Escalation Testing up to full power was completed by 0300 on June 15, 1989. Full power physics testing was performed on June 16, 1989. All power escalation testing was complete by June 22, 1989.

1

Page 2 s

TABLE 1 17 x 17 HYBRID B C4RCCA DIMENSIONAL INFORMATION i

B&W Chromium B6V Armaloy Carbid Coated Westinghouse Description Plated RCCA RCCA RCCA Overall Assembly Length 161.07 in. 161,07 in. 160.95 in.

Rod Length - Tip of Bottom 150.68 in. '150.68 in. 150.58 in.

of Spring Retainer Cladaing Material 304 SS Inconel 625 304 SS Cladding 0.D. .381 in. .376% in. .381 in.

Cladding I.D. .304 in. .304 in. .304 in.

Absorber Stack Length 142.0 in. 142 in. 142 in.

B C Pellet Stack Length 4 102.0 in. 102 in. 102 in.

Ag-In-Cd 40.0 in. 40.0 in. 40.0 in.

Absorber Diameter B C Pellet .294 in. .294 in. .294 in.

4 Ag-In-Cd .300 in. .300 in. .301 in.

(28 in. length) (28 in. length) (40 in. length)

.294 in. .294 in.

(12 in. length) (12 in. length)

Coating or Plating Thickness .0005 in. .002 in. N/A RCCA Weight (Calculated) 91 lbs. 93 lbs. 94 lbs.

Spring Retainer Travel .880 in. .880 in. .880 in.

~

Spring Preload 340 lbs. 340 lbs. 352 lbs.

RCCA Spring Rate 706 lbs./in. 706 lbs./in. 706 lbs./in.

r

. '. Page 3 4

i 2.0 PRECRITICAL TESTING Precritical testing includes the procedures used for:

• loading the Cycle 3 core,

• determining initial calibration data for excore power range and intermediate range datectors, and

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

, '. Page 4 2.1 '{otal Core Reloading - PT/2/A/4150/22 l The Cycle 3 core was loaded under the direction of PT/2/A/4150/22, Total Core Reload. Plots of Inverse Count N te Ratio (ICRR) versus

^

number of fuel assemblies loaded were kept for each source range channel.

Core loading began at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br /> on April 26, 1989 and concluded at 0830 on April 30, 1989. The core loading was verified by PT/2/A/4550/03C., Core Verification, which was completed at 1500 on April 30, 1989.

Figure 1 shows the core loading for Cycle 3.

l l

. J l 1 l

1 l

Page 5 FIGURE 1 CORE LOADING PATTERN CATAWBA UNIT 2 CYCLE 3 R P N M L K J H G F E D C B A

!P12 !G06 !P44 'E30 !P11 !G35 !P24 i 1  !  ! I I  ! I f I 1 1258KT !!44KT 154KT !!53KT !!46KT !121KT !211ti i

!P61 !P48 !G53 IN37 1955 !P5B !G03 !N52 !G67 !P41 !P55 1 2  !  ! i  ! l  !  ! 1 1  !  !  ! 2

!253KT tR54 1138KT !R62 !!2P78K!R83 112P70K!R65 1166KT IR55 !228KT I

!P57 !G13 !P40 !N35 1G46 !N53 !P49 lN49 !Q48 !N46 !PO4 IG40 !P60 1 3 1 1  ! I i i  !  !  !  ! I I I I 3

!234kT !!68KT !!54KT !R93  !!2P80K!R200 10554K 1R201 112P95K!D7B !49KT !!97ET !201tT I

!P22 !P25 !P39 fG44 !N19 !G29 !N39 !E32 !N1B !G01 !P32 !P30 !P21 1 4  ! t i 1 1  !  ! 1  !  !  ! I I I 4

!R38 1145KT !R57 112P79K!183KT !12P88K!R92 !!2P94K!!49KT !!2P74KtR60 1109KT IR$9 I

!P09 !G52 IN63 !G27 !N59 !G15 !N0B !G61 !N64 !G59 !N51 !G50 !N36 19T2 !P3B i 9i  !  !  ! I I i 1  !  !  ! I  ! I i !5

!242KT 1102KT !R70 112P93Kl200KT 112P66K!53KT 18P113Kl51KT 112P96K1104KT 112P92K!R96 1116KT 121BKT I

!G49 !N24 !G45 !N02 !QS7 !P01 !P23 !N10 !P46 IP33 IR14 !N32 1863 IN20 !Q12 1 6i  !  !  ! t i 1  !  !  !  !  ! I i i I6

!274KT !R67 112P89Kl173KT 112P65K!R105 !152KT IR84 !237KT !R106 112P87Kl180KT 112PB4KIR71 1172KT I

!P14 !G31 !N26 !R$6 !N2B !P36 !N43 1938 !N34 IP34 IN54 !G60 IN27 IG62 !P05 I 7!  !  ! I  !  !  !  !  !  ! 1  !  ! I i 17 156KT !!2P10L R75 112P69K!155KT 1163KT !67KT !BP116K!170KT 1107KT 1110KT 112P99K!R76 112PB3K!!25KT I

!Q?1 !P53 !P59 !N50 !G07 !N13 !GTO IH61 1000 IN22 !G22 IN14 !PS2 IP62 !Q11 1 8!  !  !  !  !  !  ! I l  ! I i  !  ! I 18 1181KT !RB5 1268KT !R94 !8P114K!RBB !BP111KIR56 !8P115K!R98 !8P110K!R102 !254KT !R87 1177KT I IP27 !E23 !N29 !G37 !N62 !P26 !N09 !Q36 !NSB !P35 !N21 !Q16 !N44 !Q54 !$29 !

9!  !  !  !  !  !  !  !  !  !  !  ! I f I l9

!!26KT !!2P68K!R77  !!2P67Kl127K1 !!64KT t150KT !8P112K!52KT !215KY 1176KT 112P81K!R86 !12P105!132KT I IG42 !N40 !G64 !N57 !E39 IP42 !P16 !N12 !P10 !P47 IG17 !N23 1928 !N41 !E10 1 10 !  !  !  !  ! I I  !  !  ! I I I I f 110 1143KT IR72 112P91K!174KT !!2P82K!R103 !62KT !R89 1187KT !R104 112P90K!184KT 112P101!R74 1175KT !

!P2B !G09 !N48 !G19 !N33 !E51 !NC2 !G65 !N25 (G20 IN15 !G04 !N16 1902 !P31 1 11 1  !  ! 1  !  !  ! I i  !  ! I I I I til 1202KT 1148KT !R97 112P85K!!05KT !!2P71K!!35KT !8P109Kl159KT t12P72Kt158KT !!2P98K!R100 1151KT 1239KT I

!P45 IP06 !P07 !E26 !N47 !G69 !N38 !G66 IN05 !G68 IP1B !P13 !P08 l 12 I I i  !  !  !  ! I I  !  !  ! I i 12

!R61 !50KT !R68 !!2P103!!47KT !!2PB6K!R95 112P77El156KT 112P75KIR80 1178KT !R63 1

!P64 !G25 !P19 !N56 !B24 !N06 !P54 !N45 !G47 IN17 !P15 !G43 !P50 1

!  ! I t  ! I I I  ! i  ! 13 13 1 1  !

1189tT !179KT 163KT IR101 !!2P104!R202 10553K tR90  !!2P100!R99 1141KT !!B2KT !212KT I 8PS6 !P37 !G34 !N07 !Q18 !P63 !G05 !N60 !G58 !P43 !P51  !

14 I  !  !  !  !  ! I I  !  !  !  ! 14

!240KT !R64 8217KT !RB1 !12P73X!R91  !!2P76K!RB2 !!61KT !R66 1241tf 1

!P17 !G41 !PC2 !G33 !P20 !G21 !P03 !

15  !  !  !  !  !  !  !  ! 15 l  !?65KT !75KT !157KT f137KT 172KT 1230KT 123BRT !

R P f M L K J H C F E D C B A l

l l *** THE CONTROL COMPONENT ID FOR ALL BURNABLE POISON ROD ASSEMBLIES (BPRA'51 5HOULD END WITH THE ***

tst LETTER 'K'. DUE 10 SPACE LIMITATIONS ON THIS MAP, DNLT 6 CHARACTERS OF THE CONTROL COMPONENT ***

• • • ID ARE ABLE TO BE IRINTED -- NOTE THAT 8.LL BP ID'S SHOULD END WITH A 'K' ***

L_ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J

'i Page 6 2.2 Refueling ENB Calibration - PT/2/A/4600/05E This procedure.is performed for each cycle to ensure that Nuclear Instrumentation System (ENB) setpoints and power indications are t

valid for cycle startup. Final adjustments are made using measured

. data as part of power escalation testing (See Section 4.0).

' . Calibration data was calculated by obtaining the last calibration data from Cycle 2 and the. fuel assembly powers for selected core locations from the corresponding incore flux measurements. The, beginning of cycle (BOC) 3 predicted powers for these locations were obtained from Duke Design Engineering's "Startup and Operational Report". The ratio of predicted BOC 3 power to measured Cycle 2 power was used to adjust the Cycle 2 calibration. data.

Table 2 lists-the ratios and calibration data calculated by this procedure. Figure 2 shows the core locations used to calculate the ration. Calculations were performed on May 18, 1989. Calibrations were complete before entering Mode 3 (Hot Standby) on June 2, 1989.

___________-___-__--x.-_,-_ - - - - - , - - - - - - - _ _ - - - - - - - - - - - - . - - - - _ , - - - - - - - - - . - - _ _ - - - - - - - - - - - - - - - - - - . - - ~ - - - - - - - - - - _ _ - - - - - . _ - -

= *' Page 7 TABLE 2 PRELIMINARY CALIBRATION DATA FOR EXCORE INSTRUMENTATION Intermediate Range Ratio- Cycle 2 Trip BOC 3 Trip Channel (BOC 3/ Cycle 2) Setpoint, pA Setpoint, VA N35 0.9549 99 95 N36 0.9629 '92 89 I

Power Range Ratio Axial Cycle 2 Full Power BOC3 Full Power Channel (BOC3/ Cycle 2) Offset, % Currents, pA Currents, pA Upper Lower Upper Lower

+20 482.0 379.3 551.5 434.0 y!41 1.1442 0 420.8 437.9 481.5 501.0

-20 359.6 496.5 411.5 568.1

+20 388.0 328.7 438.6 371.6 N42 1.1304 0 340.0 380.5 384.3 430.1

-20 292.1 432.4 330.2 488.8

+20 398.1- 334.3 436.1 366.2 N43 1.0954 0 344.7 387.5 377.6 424.5

-20 291.2 440.8 319.0 482.9

+20 452.4 393.3 509.9 443.3 N44 1.1272 0 393.0 450.1 443.3 507.4

-20 333.6 506.8 376.0 571.3 E

f

. FIGURE 2 Pegs 8

, . ASSEMBLIES SELECTED FOR PRELIMINARY CALIBRATION OF EXCORE INSTRUMENTATION I/R 36 G P W M L K J H G F E D C B A 1

F/R P/R l N 444 -

_ . . . . N42

?  !  !  !  !  !  !  !  !  ! I i 2 1  ?  !  !  !  ! 1 1 1  !  !  ! 2 i  !  !  !  !  ! I f 1 1  ! l 1 i  !  !  !  !  !  !  ! 8 f  !  ! t 3 f  !  !  !  !  !  !  !  !  ! I 8 1  ! 3 8 ' t  !  !  !  !  !  !  !  ! l 8  !

!  !  !  !  !  !  ! I i  ! t I  !  !

4  ! 8  !  !  !  !  !  !  !  !  !  !  !  ! 4

!  !  ! I t t I t i  !  ! l t

' t I  !  !  ! e 1 1 1 I t i  !  ! t 5!  ! 8 I f I i 1  !  !  !  !  !  !  ! !5

!  !  !  !  !  !  !  !  !  !  !  !  ! I I t

!  !  !  !  !  !  !  ! t I  !  !  !  !  !  !

6!  ! I  !  ! 1 1  !  ! I I  ! f f f !6

~

!  !  !  ! I  !  !  ! l I i 1 1 I i  !

!  !  !  ! I  ! l 8 i  !  !  !  !  ! l  ?

7!  !  ! I I I i  ! l I  !  !  !  ! I !7

!  !  !  ! I I i i i 1 1  !  !  !  !  !

! 1  !  !  ! I I i 1 1  !  ! I I f  !

8?  !  !  !  !  ! 1 1  !  !  ! I i  !  ! !8

!  !  !  !  !  !  !  ! I I I I  !  !  ! 1

!  ! I i f I  ! 1 1 I i  ! I i  !  !

91  !  !  !  !  !  ! i  !  !  ! I f I  ! l9 8

t  !  !  !  !  !  !  !  !  !  !  !  !  ! I i  !  !  ! I l  !  ! I t  !  !  !  !  !  !

10 9  !  ! I i  !  !  !  ! I f I  !  ! 1 110

? I I  ! I t  !  !  !  ! l  ! I f i I

!  !  !  !  !  !  ! I  ! I i  !  !  ! I i 11 ! I t I  !  !  !  ! I f I t I  !  ! 111 8  !  !  !  ! I i  !  !  !  !  !  !  ! I i 1  !  !  !  !  !  !  ! I  ! I t i 1 12 t  !  !  !  !  !  !  !  !  !  ! I t i 12

!  ! I f I  ! I f I  !  !  ! I t

!  !  !  !  !  !  !  !  !  !  ! I  ! 4 13  ! . !  !  !  !  !  !  !  !  !  !  ! I i 13

!  !  !  !  !  !  !  !  ! (  !  ! I '

!  !  !  ! 1  ! l I  !  !  !  !

14 I  ! ' I ' ' I  !  ! I  !  ! I4 P/R e i e i i t 8' e i e P/R

~......' ~......'

N41 N43

=

i i !O O Oi i i >>

0 P N M L K J H G F E D C B A Core Locations Used I/R Core Locations Used for 35 for Pwr Range Cals. Intermed. Range Cals.

". . Page 9

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2.3 1/M Approach'To Criticality - PT/2/A/4150/19 I The initial criticality for each' cycle is achieved under the direction of this procedure. The procedure consists of two major parts: 1) dilution to estimated critical boron concentration (ECB) and 2) rod withdrawal to criticality.

Dilution is performed in Mode 3 (Hot Standby) at essenticily Hot Zero Power (HZP) temperature and pressure with all shutdown and control banks inserted. Plots of inverse count rate ratio (ICRR) for each source range channel are maintained to ensure that adequate shutdown margin is maintained.

Rod withdrawal to criticality begins when unit is ready to enter Mode 2 (Startup). Zero power physics testing begins with rod withdrawal. An estimated critical. rod position is calculated based on latest boron concentration. ICRR is plotted for each source I range channel. Rods are withdrawn until source range count rate doubles, then ICRR data is obtained and used to project critical rod pos* tion. If projected position is acceptable, rod withdrawal continues. This process is repeated until criticality is achieved.

Dilution began at 1905 on June 4, 1989 and ended at 2354 following ,

the addition of 19,590 gallons of demineralized water. Initial 1 boron concentration was 1904 ppmB, the ECB was 1357 ppmB and the final boron concentration was 1373 ppmB. Figure 3 shows ICRR f j

versus volume of water added.

Rod withdrawal began at 1458 on June 6, 1989. Estimated critical I rod position was 135 steps on Control Bank D. Criticality was '

achieved at 1729 with Control Bank D.at 137 steps. Figurc 4 shows the ICRR plots maintained during rod withdrawal. All acceptance criteria were met.

)

,  ;. Pegs 10 FIGURE 3 ICCR VS. WATER ADDITION-fR.N-31ICRR(Co/C) 4 0.9 h 0 0

0.8 h 000 0 O.7 j_ 00 0 00 0 0

'O.6 -

00 0.5 l- N 0.4 h Louer 0.3 i Limit D.2 h Expected 2 ICRR' 0.1 -

O , , , , , , , ,, ,, , , ,,,, ,, 0 -S.R. N-31 0 5000 10000 15000 20000 DEMIN URTER RDDITION yg.R.N-32ICRR(Co/C) 0 0.9 F 00 0 0.8 2

00 0.7 p l

0.6 0 0 D 0 0 000 0 00

0. 5 -

0.4 h Louer 0.3 h Limit 0.2 h Expected ICRR O.1 -

0: , , , , , ,, ,, ,, , , ,,,, , , O S.R. N-32 0 5000 10000 15000 20000 DEMIN URTER RDDITION

_ _ _ ___-_-__________.__-_________j

Pass 11 4

FIGURE 4

'ICRR VS. ROD WORTH g.R.N-31ICRR(Co/C) t 0.9 -

0.8 F 0.7 -

2 0.6 - Shutdown E @

112 steps w/d 0.5  ? O 0.4 5 0.3 h cce42swa 0.2  :- O ccelooswa 0.1 - O coe76: :wa 0

''''''''''''''' '',,,,,,, n, O S.R. N-31 0 10 20 30 40 50 60 70 80 90 100 CD@l20swd ROD UORTH WITHDRRUN, */.

I g.R.N-32ICRR(Co/C) _

0.9 F 0.8 -

0.7 h 0.6 h Shutdown E @

_ 112 steps w/d 0.5 9 0 0.4 h 0.3 h cce42swa 0.2 h 0 ccelooswa 2

0.1 O coe76 4wa 0 -''''''' ''''''''''''',,,Pn , O S.R. N-32 0 10 20 30 40 50 60 70 80 90 100 CD@l20swd ROD WORTH WITHDRAWN, */.

l < '. Page 12 l

l 3.0 ZERO POWER PHYSICS TESTING l

Zero Power Physics Testing (ZPPT) is performed at the beginning of each cycle and is controlled by PT/2/A/4150/21, Post-Refueling Controlling Procedure for Startup Testing. Test measurements are made below the point of n'uclear 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 measurements are included in the ZPPT program:

l

• Source Range / Intermediate Range Overlap

' Point of nuclear heat addition

• Reactivity Computer Checkout

• All Rods Out Critical Boron Concentration

• All Rods Out Isothermal Temperature Coefficient

• Reference Bank Worth by dilution

• Reference Bank in critical boron concentration

• Differential Boron Worth determination

• Control Rod Worths by rod swap Zero power physics testing for Cycle 3 began at 1458 on June 6, 1989 with the beginning of rod withdrawal for approach to criticality. ZPPT was complete by 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> on June 7, 1989 following analysis of rod worth j measurements. Table 3 summarizes ZPPT results. All acceptance criteria j were met.

Sections 3.1 through 3.9 describe ZPPT measurements and results.

7

• Page 13.

TABLE 3

SUMMARY

OF ZPPT RESULTS Parameter Measured Value Predicted Value/ Acceptance Criteria Nuclear Heat 2.0 x 10 amps --

(on N41)

ZPPT Range 10 to 10 amps --

(on N41)

ARO Boron Conc. 1400 ppmB 1407 1 50 ppmB ARO ITC . + 0.25 pcm/*F + 0.44 2.0 pcm/*F ARO MTC + 2.08 pcm/ F + 2.27 pcm/'F Reference Bank 1106.5 pcm 1019 i 153 pcm (Control Bank C)

Worth

'Ref. Bank In 1283 ppmB 1300 ppmB Boron Conc.

Differential -9.46 pcm/ppmB -9.52 1.43 pcm/ppmB Boron Worth Control Bank D 564 pcm 566 i 200 pcm

, Worth Control Bank B 624 pcm 628 200 pcm Worth Control Bank A 301 pcm 280 200 pcm Worth Shutdown Bank E 444 pcm 440 200 pcm

i. Worth l .

Shutdown Bank D 350 pcm 376 200 pcm Worth Shutdown Bank C 353 pcm 376 i 200 pcm Worth Shutdown Bank B 868 pcm 832 1 250 pcm Worth Shutdown Bank A 293 pcm 300 200 pcm Worth Total Rod Worth 4903.5 pcm 4817 - 482 pcm l e - - - _ - - _ . - - _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ . - - - - . _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _

4 Page 14-3.1 Source Range / Intermediate Range Overlap Data - PT/2/A/4150/21-During the initial approach to criticality, source range and

-intermediate range data was obtained to verify than at least one decade of overlap existed. If one decade of overlap did not exist, intermediate range compensation voltage would be adjusted to provide the overlap.

Overlap data for Cycle 3 was obtained~per PT/2/A/4150/21, Post-Refueling Controlling Procedure For Startup Testing, on June 6, 1989. Table 4 contains the overlap data. The acceptance criterion was met.

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-Page 15 TABLE 4 i

SOURCE RANGE / INTERMEDIATE RANGE OVERLAP DATA SOURCE RANGE _ INTERMEDIATE RANGE (CPS) (AMPS)

N31 N32 N35 N36 l' INITIAL DATA NIS Cabinet 300 350 1.0 x 10 ~11 1.3 x 10 ~28 OAC 280.9 354.9 9.964 x-102 1.320 x 10~11 SOURCE RANGE BLOCK NIS Cabinet 40,000 37,000 4.0 x 10~8' 3.0 x 10 ~28 OAC 34,270 36.240 3.292 x 10 ~2' 2.647 x 10~2'

'*: '4-- Page 16 3.2 Point of Nuclear' Heat Addition - PT'2/A/4150/21 /

The point of nuclear heat addition is measured by trending reactor coolant system temperature, pressurizer level, flux level and reactivity. A slow, constant startup rate is initiated by rod withdrawal.. An increase in reactor coolant system temperature and/or pressurizer level and/or a change in reactivity and/or rate of flux icvel increase indicate the addition of nuclear heat. The measuremer.t is repeated to ensure confidence in results.

For Cycle'3, the point of nuclear heat addition was determined per PT/2/A/4150/21, Post-Refueling Controlling Procedure for Startup Testing,'on June 6, 1989. Table 5 summarizes the data obtained.

The ZPPT test band was set at 10 to 10 amps on N-41 (power range channel attached to reactivity computer). This provided more

.than a factor of two margin to nuclear heat for ZPPT. The acceptance criterion was satisfied.

. + Page 17 TABLE 5 NUCLEAR HEAT DETERMINATION REACTIVITY COMPUTER. INTERMEDIATE RANGE (AMPS) (AMPS)

N41 N35 N36

~

RUN #1 2.11 x 10 ~8 1.464 x 10 ~5 1.142 x 10 '

~

RUN #2 2.04 x 10 ' 1.381 x 10 1.060 x 10

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

~ ~

2.04 x 10 ' amps + 2 = 1.02 x 10 ' amps

~ ~

ZPPT TEST BAND: 10 ' to 10 ' amps on N41 e

_ _ _ _ _ _ _ .____.__m____..__ _ _ _ _ _ _ _ - _ . _ . _ _ . . _

4 Page 18 3.3 Reactivity' Computer Checkout - PT/2/A/4150/21 The reactivity computer checkout is performed per PT/2/A/4150/21, Post-Refueling Controlling Procedure for Startup Testing, to verify that the power range channel connected to the reactivity computer can provide reliable reactivity data. Reactivity insertions of approximately +25, -25, +40, and -40 pcm are made and the doubling.

(or halving) time is recorded. The measured reactivity for each case is compared to the theoretical reactivity obtained from the doubling time and verified to be .ithin 4%.

The checkout was performed for Cycle 3 on June 6, 1989. Table 6 lists the results of the 4 trials. The acceptance criterion was met in all 4 cases.

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Page 19 TABLE 6 REACTIVITY COMPUTER CHECKOUT t

WESTINGHOUSE REACTIVITY COMPUTER DOUBLING KEACTIVITY

-PERIOD (HALVING) EJMPUTER THEORETICA7. PERCENT ABSOLUTE (sec) TIME (sec) AP (pcm) REACTIVITY ERROR

• DIFFERENCE AP (pcm) (%) (pcm) 264.0' 183 +25.0 25.4 1.6 0.4

-318.8 221 -27.0 -26.8 0.7 0.2 168.8 117 +37.0 37.5 1.3 0.5 1 .

1

-219.3 152 -41.0 -41.8 '1. 9 0.8 l

l 1

• PERCENT ERROR = l Measured - Theoretical l x 100 l Theoretical l Acceptance Criteria: Percent error < 4.0% O_R absolute difference < 1 pcm, whichever is greater.

i

-Page 20 3.4 ARD Boron Endpoint Measurement - PT/2/A/4150/10 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 is measured and converted to an equivalent boron using the predicted differential boron worth. This equivalent boron is added to the averaged measured boron l: concentration to obtain the ARO critical boron concentration.

l l

The Cycle 3 BOC, Hot Zero I'ower (HZP), ARO critical boron l concentration was measurer' on June 7, 1989. The measured boron i concentration (average of 3 samples) was 1401 ppmB. This value was corrected by -I ppmB (reactor was slightly subcritical with ARO) to yield an ARO concentration of 1400 ppmB. The acceptance criterion, measured boron concentration within 50 ppmB of predicted, was met.

._m.-...___m_-m____m__ ____ _ _ _ . . . = _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ . . _ _ . = _=_.

_.._________._-._._m____.____________.___-________.-__m.___m_- - _ _ _ _ _

e

• Page 21 a

- 3.5 ARO Isothermal Temperature Coefficient Measurement - PT/2/A/4150/12A s.

The ARD Isothermal Temperature Coefficient (ITC) is measured at the beginning of each cycle to verify consistency with the predicted value. In addition, the Moderator Temperature Coefficient (MTC) is

, obtained by subtracting the doppler temperature coefficient from the L ITC. The !!TC is used to ensure compliance with Technical Specification limits. The ITC is measured by slowly (< 10*F/ hour) i changing reactor coolant temperature while measuring reactivity versus coolant temperature. The slope of the reactivity versus temperature line is used to determined the ITC. At least one cooldown and one heatup is performed.

The BOC 3 ARD ITC was measured on June 7, 1969. Only one cooldown and one heatup were required because of the good agreement between the results. Table 7 summarizes the data obtained and the test results.

Averaged measured ITC was + 0.25 pcm/*F. Predicted ITC was + 0.44 pcm/*F. Measured ITC was within the acceptance criterion of predicted 2 pcm/*F.

The MTC was determined to be + 2.08 pcm/"F. This value was used with procedure PT/2/A/4150/20, Temporary Rod Withdrawal Limits Determination, to ensure that MTC would remain within Technical specification limits at all times. No rod withdrawal limits were l required.

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.Page 22 l, . .

TABLE'7 ITC MEASUREMENT RESULTS l' ,

AT, *F ap, pcm Avg. Temp., *F ITC, pcm/*F Cooldown -4 .-3.0 553.5 +0.75 (Uncorrected) lleatup +4 -3.0 554.0 +0.75 (Uncorrected)

Average +0.253*

• Average ITC corrected to nominal I!ZP temperature of 557'F BOC EZP doppler coefficient = -1.83 pcm/*F MTC =. Average ITC - Doppler Coeff. = +2.08 pcm/*F l

I 9-

_ __-u_ -_-___m.___-_._.mm _ _ _ _ _ __. _____.m

• ** Page 23 3.6 Reference Bank Worth Measurement by Dilution - PT/2/A/4150/11A The Control Rod bank predicted to have the highest worth is designated as the reference bank and is measured by inserting the bank (with all other rods fully withdrawn) in discrete steps, while slowly diluting the NC boron concentration. The reactivity worths of the discrete steps of rod insertion are measured by the reactivity computer and summed to obtain the inserted worth of the reference bank as a function of bank height.

The BOC 3 Reference Bank (Control Bank C) Worth was measured on June 7, 1989. Figure 5 gives integral rod worth as a function of reference bank height. The reference bank was measured to be worth 1106.5 pcm; predicted worth was 1019 pcm; and the acceptance criterion was that measured worth was more than 866 pcm and less than 1172 pcm (1019 pcm i 15*.) .

9

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[* - Page 25 3.7 Reference Bank in Boron Endpoint Measurement - PT/2/A/4150/10 This test is performed to meas.ure the critical boron concentration with the reference control rod bank fully inserted and all other control rods fully withdrawn. The measured boron concentration is used with the measured ARO boron concentration and the measured worth of the' reference bank to calculate the differential boron worth. The test is performed by measuring the reactor coolant system boron concentration with control rods near the desired configuration. Control rods are inserted or withdrawn to desired configuration concurrently with boron samples. The reactivity.

difference from criticality is measured and converted to an equivalent boron concentration using the predicted differential boron worth. This equivalent boron is added to the average measured boron concentration to obtain the Reference Bank-in boron concentration.

The BOC 3 Rererence Bank-in boron concentration was measured on June 7, 1989. The measured boron concentration (average of 3 sampics) <

was 1287 ppmB. This value was votre&ted by -4 ppmB to yield a Referance Bank-in boron concer!r.rattoa of 1283 ppmB. The predicted value was 1300 ppmB. There 1 ne acceptance criteria directly associated with this test.

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Page 26 3.8 Differential Boron Worth Determination - PT/2/A/4150/21 The differential boron worth is calculated from the measured ARO critical boron concentration, Reference Bank-in critical boron concentration, and integral worth of the reference bank. The calculated value is compared to predicted value to verify consistency. This calculation also provides an indirect check of the measured reference bank worth and of the boron endpoints.

The BOC3, HZP differential boron worth was calculated to be -9.46 pcm/ppmB (-1106.5 pcm + (1400 ppmB-- 1283 ppmB)). The predicted value was -9.52 pcm/ppmB and the acceptance criterion was -9.52 i 1.43 pcm/ppmB (predicted i 15%).

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-; t Page 27 3.9 Control Rod Worth Measurement by Rod Swap - PT/2/A/4150/11B The worth of all control rod banks except the reference bank is 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 position is adjusted to achieve critical conditions with all other rod banks fully withdrawn. The worth of the inserted  ;

rod bank is determined from the critical height of the reference '

bank. The measured worth is compared to predicted worth to verify consistency. Also, the sum of worths of all banks (including reference bank) is compared to predicted.

l The BOC3 rod worths were measured by rod swap on June 7, 1989.

Table 8 summarizes the results of the measurements. All acceptance l criteria were met.

4

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Page 29 4.0 POWER ESCALATION TESTING Power escalation testing is performed during the initial power escalation of each cycle and is controlled by PT/2/A/4150/21, Post-Refueling Controlling Procedure for Startup Testing. Tests are performed from 0%

through 100% power with major plateaus at ~ 30%, ~ 80%, and ~ 100% power.

The following significant tests are performed in the power escalation test program:

' Core power distribution (~ 30%, ~ 80%, and ~ 100%)

• Interim incore/excore calibration (~ 30%)

• Full power delta temperature extrapolation (~ 30*4, ~ 80%, and ~ 100%)

• Incore/Excore calibration (~ 80*4)

• Precision calorimetric reactor coolant flow measurement (~ 100%)

• Hot Full Power (HFP) critical boron concentration measurement (~ 100%)

• Incore/Excore Calibration Check (~ 100%)

In addition to the above tests, a transient test was conducted from ~ 40%

und ~ 100% to verify performance of Steam Generator level control following a modification to the IcVel instrumentation, and core power distribution was measured at ~ 98% power to ensure peaking factors would not be violated at full power.

Also during power escalation, checks are performed for the incore detector system, thermal power program and inputs, various other process computer programs and functions related to core monitoring, steam flow-feed flow mismatch, turbine impulse pressure, intermediate range trip setpoints, etc. The results are not included in this report.

Power escalation testing for Cycle 3 began on June 8, 1989. Full power was reached on June 15, 1989. Full power testing was completed on June 22, 1989.

Sections 4.1 through 4.8 describe the significant tests performed during power escalation and their results.

1 1

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l. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _.______________.____._____________________9

1?

"., Page 30 4.1 Core Power Distribution - PT/2/A/4150/05 Core power distribution measurements are performed during power escalation testing at low power (~ 30%), intermediate power (~ 80%),

and full power (~ 100%). Measurements are made to verify that core peaking factors are within limits and to verify flux symmetry.

Measurements are made with the incore moveable detector system and analyzed with Shangstram Nuclear Associates CORE package. Additional measurements are required if there is potential for exceeding limits during power escalation.

l-l The BOC 3 core power distribution measurements were performed en June 11, 1989 (30% power), June 13, 1989 (79% power),' June 14, 1989 (98% power), and June 16, 1989 (100% power). Tables 9 through 12, respectively, summarize the results. A measurement was required at 98% power because the margin to peaking factor limits at 80% was insufficient to permit unrestricted full power operation. Acceptance Criteria for core power distribution were met in all cases.

[..

7,* Page 31' TABLE 9 CORE POWER DISTRIBUTION RESULTS 30% POWER Map ID: FCM/2/03/001 Date: June 11, 1989 Power Level: 30.31%

Cycle Burnup: 0.357 EFPD, 14.85 MWD /MTU Boron Concentration: 1241 ppmB Control Bank D Position: 217 steps withdrawn Core Average Axial 0ffset: 24.162%

T Maximum F : 2.2125 at Core Location C-13 9

T Maximum Fq (Z)/K(Z): 2.2867 at Core Location C-13, Axial Location 44 Minimum Margin to Surveillance F :

q 41.66%

1.4310 at Core Location C-13 Maximum.FAH:

i Measured R: 0.7953 l

Reactor Coolant Flow: 397,949 gpm; 102.66% of Tech. Spec. Flow Maximum F AH Error (From Predicted): 8.9C% at Core Location B-13 I Mean of Absolute F Errors: 2.65%

AH Tilt Ratios: Top Half Bottom Half 1

I Quadrant 1 0.99698 0.99668 i Quadrant 2 1.02975 1.03397 I

Quadrant 3 1.00378 0.99560 Quadrant 4 0.96950 0.97375 L -- - - - - - - - _ _ _ _ _ _ _

-?

Page 32 TABLE 10-CORE POWER DISTRIBUTION RESULTS 80% POWER Map ID: FCM/2/03/008 Date: June 13, 1989 Power Level: 79%

Cycle Burnup: 1.980 EFPD, 82.48 MWD /MTU Boron Concentration: 1009 ppmB Control Bank D Position: 181.5. steps withdrawn Core Average Axial Offset: 0.060%

T Maximum Fq : 1.9415 at Core Location B-09 T

plaximum Fq (Z)/K(Z): 1.9460 at Core Location B-09, Axial Location 34 Minimum Margin to Surveillance F :

q 20.16% at Core Location B-09 Maximum F AH:

1.4228 at Core Location B-09 Measured R: 0.8972 Reactor Coolant Flow: 396,282 gpm; 102.24% of Tech. Spec. Flow Maximum F AH Error (From Predicted): 7.26% at Core Location A-09 Mean of Absolute F AH rr rs: 2.11%

Tilt Ratios: Top Half Bottom Half Quadrant 1 0.99977 . 0.99878 Quadrant 2 1.01892 1.03015 Quadrant 3 0.99768 0.99046 Quadrant 4 0.98363 0.98061 l

L_ _ _ _ . _ . _ _ _ __ _ _ . _ _ _ __ .__.___.___.______.___.____.____________._____.__________.________.___.______________________________a

',* Page 33 l TABLE 11 CORE POWER DISTRIBUTION RESULTS 98% POWER Map ID: FCM/2/03/009 Date: June 14, 1989 Power Level: 97.70%

Cycle Burnup: 2.831 EFPD, 117.92 MWD /MTU Boron Concentration: 984 ppmB Control Bank D Position: 201 steps withdrawn Core Average Axial Offset: -1.052%

T Maximum Fq : 1.8793 at Core Location G-08 T

, Maximum Fq (Z)/K(Z): 1.8818 at Core Location G-08, Axial Location 33 Minimum Margin to Surveillance F :

q 3.94%

Maximum FAH:

1.4041 at Core Location E-08 Measured R: 0.9348 Reactor Coolant Flow: 395,534 gpm; 102.04% of Tech. Spec. Flow Maximum F AH rr r (From Predicted): 5.73% at Core Location B-13 Mean of Absolute F AH ## #8 I

Tilt Ratios: Top Half Bottom Half Quadrant 1 1.00037 . 1.00149 Quadrant 2 1.01730 1.03043 Quadrant 3 0.99440 0.98503 Quadrant 4 0.98793 0.98305 I

_ _ _ _ _ _ _ _ _ _ _ - _ _ - - __ - _ _ _ _ -_ _ a

Page 34 TABLE 12 CORE POWER DISTRIBUTION RESULTS 100% POWER.

Map ID: FCM/2/03/010 Date: June 16, 1989 Power Level: 99.59%

Cycle Burnup: 4.514 EFPD, 188.02 MWD /MTU Boron Concentration: 961 ppmB Control Bank D Position: 209 steps withdrawn.

Core Average Axial Offset: 0.193%

T Maximum Fq : 1.8850 at Core Location G-08 T

Maximum Fq (Z)/K(Z): 1.8923 at Core Location G-08, Axial Location 40 Minimum MLrgin to Surveillance F : 2.39%

9 Maximum F 1.4167 at Core Location G-08 AH:

Measured R: 0.9487 Reactor Coolant Flow: 395,424 gpm; 102.02% of Tech. Spec. Flow Maximum F AH rr r (From Predicted): 5.81% at Core Location B-13 Mean of Absolute F AH Errors: 1.98%

Tilt Ratios: Top Half Bottom Half Quadrant 1 1.00208 0.99851 Quadrant 2 1.02184 1.02676

, Quadrant 3 0.98613 0.99166 l' Quadrant 4 0.98994 0.98308 i

l l .' ".' Page 35 i

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l 4.2 Interim Incore/Excore Calibration - PT/2/A/4600/05D I

The test is performed using power range currents measured during the

30% core power distribution measurements with the axial offset from '

L that power distribution to obtain calibration data for the power ranges. Power ranges are calibrated before exceeding 50% power to ensure that an accurate indication of core axial flux difference (AFD) and quadrant power tilt ratio (QPTR) is available for Technical Specification surveillance. This "I point" calibration is limited in l that new correction factors cannot be obtained. The factors are assumed to be the same as those from the end of the previous cycle.

Data for the BOC 3 interim calibration was obtained on June 11, 1989.

Power range calibrations were completed on June 11, 1989. The rest 0 ts are given on Table 13. All acceptance criteria were met.

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.- 4 Page 36 TABLE 13 INTERIM INCORE/EXCORE CALIBRATION RESULTS

1) Flux Map Data: Power Level 30.35%

Axial Offset +24.2%

Power Range Currents, pA N41 N42 N43 .N44 Top: 133.0 107.5 121.4 118.8 Bottom: 102.5 88.2 99.9 102.7

2) Ratio of measured currents to expected (from last calibration data) currents:

N41 N42 N43 N44 Top: 0.7739 0.7873 0.8921 0.7471 Bottom: 0.8042 0.8088 0.9297 0.7872

3) New Calibration Data:

Power Range Currents, pA Axial N41 N42 N43 N44 Offset, % Upper Lower Upper Lower Upper Lower Upper Lower

+20 426.8 349.0 345.3 300.6 389.0 340.5 380.9 349.0 0 372.6 402.9 302.6 347.9 336.9 394.7 331.0 399.4

-20 318.5 456.9 260.0 395.3 284.6 449.0 280.9 449.7 Correction Factors (Mj)

N41 N42 N43 N44 1.432 1.443 1.368 1.443 l

l u_____________________________ _ _ _ _ _ . . _ . . _ _ _ _ _ __ _ _ _ _ .

.[.

, Page 37 4.3 Delta Temperature Extrapolation - PT/2/A/4150/16 Reactor Coolant System (NC) hot leg and cold leg temperature data is obtained per PT/2/A/4150/16, Unit Load Steady State, at 30%, 80%, and 100% power to ensure that delta temperature (AT) indications are accurate. AT, in units of percent of full power AT (% FPAT), is an important parameter in the reactor protection system. The reactor trips associated with AT, overtemperature AT (OTAT) trip and overpower AT (OPAT) trip, are important with respect to FSAR accident analyses.

AT indications which are too low (non-conservative) could invalidate many FSAR analyses and allow an unanalyzed transient. AT indications which are too high (conservative) could limit power level because of turbine runbacks which are designed to occur 2% below the trip setpoints.

The % FPAT indications rely upon a valid number for the full power AT

(*F), or AT . The AT,value is the value which is determined and checked by PT/2/A/4150/16. An extrapolation is performed from 80%

using data obtained at 30% and 80% power to check AT 9 for each loop.

The extrapolation is based on enthalpy, as AT is not completely linear with power. If extrapolated full power AT is more than 0.6 F (~ 1%)

different from current ATg , AT is adjusted by station instrument and 9

electrical (IAE) technicians. At 100% power, the AT are checked again, preferably using data obtained during the precision calorimetric flow measurement (see Section 4.7).

Table 14 summarizes the results of the AT extrapolation. All acceptance criteria were met.

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Page 38 TABLE 14 AT EXTRAPOLATION DATA 1 Power = 30.28%:

' Loop A Loop B Loop C Loop D T-COLD, 'F 555.2 556.0 555.7 556.0

. T-110T, 'F 573.2 574.6 574.2 574.5 Ah. BTU /lb 23.065- 23.896 23.748 23.764 Power = 78.86%

T-COLD, 'F 558.2 559.3 559.2 559.7 T-IlOT, *F 603.6 606.2 605.8 606.2 Ah BTU /lb 60.960 63.346 62.889 62.839 Extrapolated Full Power Values:

T-COLD, F 560.0 560.0 560.0 560.0 T-110T, *F 616.7 618.7 617.7 617.6 Ah, BTU /lb 77.230 80.238 79.664 79.608 FPAT, 'F 56.7 58.7 57.7 57.6 Current AT , F 9

55.6 57.4 55.5 56.1 Error (AT , FPAT), F -1.1 -1,3 -2.2 -1.5 Power = 99.84%:

T-COLD, 'F 559.5 561.1 560.8 561.8 T-Il0T, F 616.7 619.3 618.6 619.1 Ah, BTU /lb 78.74 80.75 80.05 79.57 Extrapolated Full Power Value:

T-COLD, F 560.0 560.0 560.0 560.0 T-IlOT, *F 617.2 618.5 618.1 617.8 Ah, BTU /lb 78.863 80.876 80.175 79.69 FPAT, F 57.2 58.5 58.1 57.8 Current ATg , F 56.7 58.7 57.7 57.6 Error (ATg , FPAT), F -0.5 0.2 -0.4 -0.2 NOTES: 1. Extrapolation is a linear extrapolation of Ah, including Ah =0 at 0% power. For purposes of extrapolation, T-COLD is assumed to be 560 F at full power. This assumption has small effect on extrapolated AT (~ 0.1 F AT error /1.0 F T-COLD error).

2. AT for each loop was adjusted following extrapolation using 30% and 80% data.

3. ATg check at full power used data from precision calorimetric flow measurement.

u___________________.________________.__

[

Page 39 4.4 Incore/Excore Calibration - PT/2/A/4600/05G The excore power range channels are calibrated per PT/2/A/4600/05G, Post-Refueling Incore/Excore Calibration, before increasing power above the 80% testing plateau. Quarter Core Flux Maps (QCMs) are taken at

.various axial offsets while at a constant power level. Power range upper and lower chamber currents are measured during each QCM. A linear least square fit of the output of each chamber (normalized to 100% power) is determiaed. The slope and intercept of the upper and lower chamber of each power range are used to determine the cali!> ration data (full power currents and correction factor, Mj) for that channel.

The test was performed for BOC3 on 6/13/89. Five QCMs with' axial offset ranging from +3.567% to -5.181% were used. An additional flux map was taken but not used as its data was not consistent with the least squares fit. Axial offset swing was stopped after an ~ 9% span was. achieved because of excessive rod insertion required to achieve the negative axial offset swing.

Table 15 summarizes the data obtained during the test. Table 16 lists the new calibration data. All acceptance criteria for this test were met.

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.,' Page 41

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TABIE 16 INCORE/EXCORE CALIBRATION RESULTS

~ Power Range Calibration Currents, 1 A

. Axial ~N41 N42 N43 N44 Offset- Upper Lower Upper Lower Upper Lower Upper Lower

+20 440.2' 353.7 360.8 310.2 403.0 348.5 401.2 357.9 0- 396.3 422.3 327.1 366.5 359.5' 410.6 359.5 425.0

-20 352.3 490.9 293.4 422.7 315.9 472.6 317.8 492.1' Correction Factors (tij)

N41 N42 N43 N44 1.464 1.559 1.470 1.461 l

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Page 42 4.5 Hot Full Power Critical Boron Concentration - PT/2/A/4150/04 The hot full power critical boron concentration is measured per PT/2/A/4150/04, Reactivity Anomaly Calculation. The measurement is made as close to the reference conditions (full power, T,y = 590.8 F, equilibrium xenon, design samarium, all rods out) as practicable, taen adjusted for any off-reference conditions. For evaluation of the full power boron concentration, the predicted value is adjusted by the error obtained for the hot zero power boron concentration.

The data obtained for this test comprises:

• 3 reactor coolant system boron samples

• Reactor power

• Control rod position and inserted worth

• Xenon worth

• Samarium worth

• T-AVG

• Cycle Burnup For BOC 3, the test was performed on June 16, 1989, at 99.51*4 p;wer with a burnup of 4.577 EFPD. The evaluation of the measured HFP Critical Boron Concentration was performed per PT/2/A/4150/21, Post-Refueling Controlling Procedure for Startup Testing.

The results of the evaluation are as follows:

Predicted Boron Concentration -

959.8 ppmB Predicted adjusted for HZP error -

952.8 ppmB Measured Boron Concentration -

957.5 ppmB Difference between measured and adjusted prediction -

4.7 ppmB The acceptance criterion (absolute value of difference < 50 ppmB) was met.

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\* I ',' Page 43 4.6 Incore/Excere Calibration Check - PT/2/A/4600/05B.

.An'incore/excore calibration check is performed with the full power core power distribution. The purpose of the check is to verify that the calibration performed at a lower power level (~ 80%) is valid at full power. The change in reactor coolant temperature and the change in radial power distribution associated with the change in power s .affects.the number of neutrons detected by the power range detectors and therefore can affect the calibration.

The check was performed for BOC 3 on June 16, 1989. Power level was-99.59% and cycle burnup was 4.514 EFPD. Table 17 lists the results of the check. The acceptance criterion (difference between incore and excore.AFD < 3.0%) was met.

PT/2/A/4600/05D, Interim Incore/Excore Calibration, was performed because the error between.incore and excore AFD, while less than 3%,

was greater than 1%. .If the error had been more than 3%, a full incore/excore calibration would be required. The method for the interim incore/excore calibration is discussed in Section 4.2.

Table 18 gives the results of this calibration. Power ~ranga, calibrations were complete on June 20, 1989. All acceptance criteria were met.

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4 "~t Page 44

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. TABLE 17 INCORE/EXCORE CALIBRATION CHECK RESULTS Flux Map ID: FCM/2/03/010 Cycle Burnup: 4.514 EFPD Average Power Level: 99.59%-

Incore Axial Offset: +0.193*4 Incore Axial Flux Difference: +0.19%

Average Excore AFD and difference from incore AFD AFD Difference N41 +2.45% +2.26%

N42' +1.99% +1.80%

N43 +2.15% +1.96%

N44 +2.15% +1.96%

Acceptance Criteria: Absolute value of difference between excore AFD and incore AFD is less than 3.0%.

1

' # "* Page 45

,. .~ - e TABLE 18 FULL POWER INTERIM INCORE/EXCORE CALIBRATION RESULTS Ii

1) Flux Map Data: Power Level 99.65%

Axial Offset +0.193*4 Power Range Currents, pA N41 N42 N43 N44 Top: 412.2 340.8 373.0 375.7 Bottom: 428.7 372.8 415.5- 432.5

2) Ratio of measured currents to expected (from last calibration data) currents:

Power Range Currents, pA

^

N41 N42 N43 N44 Top: 1.0429 1.0446 1.0400 1.0475 Bottom: 1.0204 1.0224 1.0171 1.0226

3) New Calibration Data:

Power Range Currents, pA Axial N41 N42 N43 N44 Offset, % Upper Lower Upper Lower Upper Lower Upper Lower

+20 459.1 360.9 376.9 317 2 419.1 354.5 420.3 366.0 0 413.3 430.9 341.7 374.7 373.9 417.6 376.6 434.6

-20 367.4 500.9 306.5 432.2 328.5 480.7 332.9 503.2 Correction Factors (Mj)

N41 N42 N43 N44 1.464 1.559 1.470 1.461 l

_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ . _ . _ . _ _ _ _ __ _ -. _ I

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Page 46 4.7 Calorimetric Reactor Coolant Flow Measurement - PT/2/A/4150/13B A calorimetric reactor coolant flow measurement must be performed at-least once every 18 months per Technical Specifications. It is

' desirable that this is done as soon as possible-in the cycle to ensure that.the'feedwater venturis are not fouled and to ensure that plent changes (such as steam generator tube plugging) have not significantly affected reactor coolant flow. The results of the calorimetric flow measurement are used to correct the reactor coolant elbow tap flow indications, which provide the indication of reactor coolant flow in normal operation.

For Cycle 3 the calorimetric flow measurement was performed on June 22, 1989. Three test runs were performed; the average of tha results were used to generate the elbow tap correction factors. Table 19 summarizes the results. All acceptance criteria were met.

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~ * ] ** Page 47 TABLE 19 CALORIMETRIC REACTOR COOLANT FLOW MEASUREMENT DATA RUN #1:

Loop A Loop B Loop C Loop D Total Power, NBTU/Hr. 2818.46 2940.88 2926.03 2935.24 11,620.61 T-HOT, *F 616.7 619.3 618.6 619.1 (99.84%)

T-COLD, "F .

559.5 561.1 560.8 561.8 NC Flow, MLBM/Hr. 35.760 36.409 36.546 36.878 145.593 NC Flow, GPM 96,326 98,302 98,629 99,673 392,930 (101.38%)

Elbow Tap Coefficients:

Channel 1 0.302121 0.301037 0.311063 0.298507 Channel 2 0.289845 0.283989 0.290469 0.298687 Channel 3 0.298160 '0.302423 0.298821 0.300418 RUN #2:

Power, MBTU/Hr 2810.35 2932.18 2914.91 2925.18 11,582.62 (99.52%)

T-HOT, *F 616.5 619.1 618.5 619.0 T-COLD,.*F 559.5 561.1 560.8 561.8 NC Flow, MLBM/Hr 35.756 36.420 36.509 36.850 145.535 NC Flow, GPM 96,319 98,331 98,532 99,597 392,779 (101.34%)

Elbow Tap Coefficients:

Channel 1 0.302144 0.301078 0.310599 0.298235 Channel 2 0.289776 0.283954 0.290097 0.298371 Channel 3 0.298134 0.302416 0.298479 0.300145 RUN #3:

Power, MBTU/Hr 2810.01 2932.58 2916.15 2925.86 11,584.60 (99.53%)

l T-HOT, *F 616.5 619.0 618.4 618.9 T-COLD, F 559.5 561.0 560.8 561.8 i

NC Flow, MLBM/Hr 35.769 36.437 36.545 36.891 145.642 NC Flow, GPM 96,353 98,380 98,631 99,711 393,075 (101.41%)

Elbow Tap Coefficients:

Channel 1 0.302200 0.301036 0.310907 0.298483 Channel 2 0.289877 0.283974 0.290300 0.298618 Channel 3 0.298239 0.302421 0.298683 0.300346 AVERAGE:

NC Flow, GPM 96,333 98,338 98,597 99,660 392,928 (101.37%)

Elbow Tap Coefficients:

Channel 1 0.302155 0.301050 0.310356 0.298408 Channel 2 0.289833 0.283972 0.290289 0.298559 Channel 3 0.298178 0.302420 0.298661 0.300303 1

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'j }. Page 48 4.8 Transient Test for Steam Generator Level Tap Modification -

'IT/2/A/9200/58 A transient test to verify proper response of Steam Generator narrow range level indication was included..in the BOC 3 power escalation test program. This test was part of the post-modification testing following relocation of the lower tap for the narrow range channels in an effort to improve the transient response of the channels.

The transient test was similar to an initial startup test performed during initial power escalation. A step load decrease of approximately 10% power was affected using turbine controls.

- Transient data was captured by the transient monitor system. Data was plotted and analyzed to verify proper response to the transient.

The test was performed from initial power levels of 40% and 100%.

L Figure 6 shows response of one steam generator level indication to the step decrease from 100%. All acceptance criteria were met.

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