ML20082U285

Cycle 6 Startup Rept. W/

ML20082U285
Person / Time
Site:	Catawba
Issue date:	09/30/1991
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9109200098
Download: ML20082U285 (52)

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( harivHe :% C :&l lik'!i l?Ulll?0Sil DUKEPOWER September- 12, 1991-U. S. Nuclear Regulatory Commission ATTN: : Document Control Desk Washington, DC .20555

Subject:

Catawba Nuclear Station, Unit 1 Docket No. 50-413..

,: Cycle 6 Startup Report In accordance with Catawba Technical Specification 6.9.1.1 (3), find attached the Catawba-Unit 1 Cycle 6 Startup Report. This report is necessary because B&W fuel is now being used in approximately one third of the Cycle 6 core. Catawba Unit I resumed commercial operation on June 15, 1991.

Very truly youc,

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. = M. S. Tuckman CRIJU1C6 STAR. TUP .

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S. D. Ebneter

. Regional Administrator, Region II W. T. Orders

- Senior Resident Inspector

. R. E.- Martin, ONRR E .

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DUKE POWER COMPANY CATAWBA NUCLEAR STATION UNIT 1 CYCLE 6 STARTUP REPORT SEPTEMBER, 1991

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TABLE OF CONTENTS Ea&S List of Tables ............., ........ .............. ........ 111 List of Figures ............. ..... ..... .... ..... . ........ iv 1.0 Introduction ........ ..... . . . ... . . ............... 1 2.0 Precritical Testing ........ . ...... ... ... . .... . .. . 2 2.1 Total Core Raloading - PT/0/A/4150/22 ,, .... ...... 3 2.2 Refueling ENB Calibration - PT/1/A/4600/05E ......... 5 2,3 1/M Approach To Criticality - PT/0/A/4150/19 . ...... 8 3.0 Zero Power Physics Testing ...... ....... ................, 12 3.1 So'urce Range / Intermediate Range Overlap Data -

PT/1/A/4153/21 ,, ....... ........ ...... ........ 14 3.2 Point of Nuclear Heat Addition - PT/1/A/4150/21 ..... 16 3.3- Reactivity Computer Checkouc - PT/1/A/4150/21 ....... 18 3.4 ARO Boron Endpoint Measurement - PT/0/A/4150/10 ..., 20 3.5 ARO Isothermal Temperature Coefficient Measurement -

PT/0/A/4150/12A ...................... .............. 21 3,6 Reference Bank Worth Measurement by Dilution PT/0/A/4150/11A ....... ................ ....... .... 23 3.7 Reference Bank In Boron Endpoint Measurement --

-PT/0/Ai4150/10 .................. ............. .... 24 3.8 Differential Boron Worth Dessrmination -

PT/1/A/4150/2; ..................... ........ ..., 25 3.9 Control Rod Worth Measurement by Rod Swap -

PT/0/A/4150/11B ..... ... ............... ....... .., 26

! 3.10 Temporary Rod Withdrawal Limits Determination -

-PT/0/A/4150/20 .......... ..., ..... .... . ....... . 28 4.0 Power Escalation Testing . . ...... .............. ....... 30 4.1 Core Power Distribution - PT/1/A/4150/05 . . . . . . . . . . . . 31 l

L 4,2 Interim Incore/Excore Calibration - PT/0/A/4600/05D . 35.

4.3 Delta Temperature Extrapolation - PT/0/A/4150/16 .. 37 i

. . .. ,._._ _ .__ - - - _ ._ . _ . _ _ . _ . _ _ _ . . _ . . - .....__...___.-.____._m_.___ -

?

. TABLE OF CONTENTS (Continued)  !,

Pane 4.4 Incore/Excore Calibration Check - PT/1/A/4600/05B ... 39 4.5 Incore/Excore Calibration - PT/0/A/4600/05A . . . . . . . . 41 t

4.6- Hot Full Power Critical Boron Conenntration -

PT/0/A/4150/04 .......... ............... ............ 44 I

4.7 Calorimetric Reactor Coolant Flow Hensurement - l PT/1/A/4150/13B ............ ........... . . . . .- . . . . . . . 45

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LIST OF TABLES EaLD

1. Preliminary Calibration Data for Excore Instrumentation 6

2. Summary of ZPPT Results 13

3. Source Range / Intermediate Range OverInp Data 15

-4. -Nuclear llent Determination 17

5. Reactivity Computer Checkout 19

6. ITJ Measurement Results 22

7. Control Rod Worth Measurement Data 27

8. Temporary Rod Withdrawal _ Limits Results 28

9. Core Power Distribution Results, 30% Power 32

10. Core Power Distribution Results, 80% Power 33

11. Core Power Distribution Results, 100% Power 34

12. Interim Incore/Excore Calibration Results 36

13. AT Extrapolation Data 38

14. .Incore/Excore Calibration Check Results- 40

15. Incore/Excore Calibration Results 42 I .

16. Incore/Excare Calibration Input Data 43 i 17. Calorimetric Reactor Coolant Flow Hensurement Data 46 l

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s LIST OF FIGURES faan

1. Core Loading Pattern, Catawba Unit 1 Cycle 6 4 2.: Assemblies Selected For Preliminary Excore Calibration 7 3, 1/M-Approach to Criticality - ICRR vs. Water Addition 10 4 1/H Approach to Criticality - ICRR vs Rod Worth 11 l

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Page 1 140 INTRODUCTION Catawba Unit 1 Cycle 6 core loading began at 0540 on May 5, 1991 and

-concluded at 1959 on May 8, 1991. ]

Criticality was achieved at 2021 on June 11, 1991. Zero Power Physics l

. Testing (ZPPT) was completed at 1800 on June 12. 1991. Power Escalation Testing up to full power was completed on June 24, 1991. All power escalation testing was completed on July 1. 1991.

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Pagn 2 2.0 [RECRITICAL TESTING Procritical testing include = thn procedures used for:

• loading the Cycle 6 core,

• determining initial calibration data for excore power range and

. Intermediate range detectors, and othe approach to criticality Sections 2.1 through 2.3 describe these procedures and results for Catawba 1 Cycle 6,

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

Page_3-2.1 Total Core Rel'oading - PT/0/A/4150/21-The Cycle 6 core was loaded'under the direction of PT/0/A/4150/22, Total Core Reload. Plots of Inverse' Count Rate Ratio (ICRR) versus  ?

number of fuel assemblics loaded were kept for each Iloron Dilution Mitigation System (BDMS) Channel.

Core loading began at 0540 hours0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br /> on Hay 5,1991 and concluded at 1959 on May 8, 1991. The core loading was verified by PT/0/A/4550/03C, Core Verification, which was completed at 0600 on May 9, 1991.

Figure I shows the core loading for Cycle 6.

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Page 4 MGURE 1 CORE LOADING PATTERN CATAWBA UNIT 1 CYCLE 6

-R P N N L K J H 0 F E D C 8 A r IG23 IN19 IG40 IN72 IG24 IN20 IG44 1

-1 I I I I I I i 1 1 16KT 1209ET 1186KT 130KT 136KT 125KT 119KT I 1750 IF48 IN09 IG62 IN27 IG57 lH24 IG64 IN10 1703 IF51 1 2 I I I l- 1 I I I I I I I 2 Il6KT lR22 122KT IR42 18698 IR07 18697 1R09 11KT IR14 17ET 1

.....................................................................c......................

IF22- 1G15 IH58 IF57 lN41 IF41 IG18 IF39 IN61 IF02 IH29 IG55 IF40 1

-3 1 1 I I I I I I I I I I I I 3 1221KT 113KT IB68T 1R30 :869L 1R$1 126KT IR53 IB6A1 IR31 1868T 1260KT 19tf I IF13 IN31 IG25- lN25 lF42 IN42 IG53 IN13 IF64 IH65 IG20 IN64 IF49 i 4 -l I I I I I I I i i 4 1 .I I 1.

IR15 18690 IR06 It6A6 129ET 18643 lR05 IB6AC 1226KT 1869M IR43 IB68W IR161 I IG36 I N08 . 1701 IN56 IG42 IG32 -IF45 IN68 IF53 IG05 IG54 N66 IF37 IN12 IG59 I 51 1 I I l- 1 I I I I I I I I I i5 110KT 1198KT IR44 lB6A8 1204KT 1199tT 1248KT fB6AD 1196KT .207KT 1250KT 1869T IR37 1269CT 1236KT I IN16 -IG01 IN62 IF34 IG14 IF55 IN51 IG22 IN37 IF26 IG46 IF16 IN01 IG45 IN04 1

-61 1 I I I I I i 1 I I I I I I i6-1224KT IR03 'It69K 111ET 132KT1IR28- 18690 1R18 IB69W IR35 - 1273ET 145KT IA6A2 IR45 146KT I IG48 IN45 IF32 IN53 IF46 IN38 lG41 lG17 lC28 th52 'IF59 IN57' lF47 IN23 IC37 I 7I I l- 1 I I I I I I I I i i i I7 '

1210KT 18692. IR108 1969G 1203ET 18699 121KT. 11BKT 12KT 15690 1227KT lt6AA IR107 18691 115KT I IN69 IG58 IGot IG12 IN36 IG07 IG19 IC04 IG26 .IG67 lN49 IG35 lG56 lG66 (N70 1 .

8l l l 1 1 I I i i I I i i i i I8 1249KT.IR34 10ss2K 1R47 18692 1R10 138KT IR50 120KT .IR12 is69x IR21 10ss1K IR46 1272KT I 91 IG68':IN46- IF43 IN47 IF23 IN67 IG04 IG47 IG02 lH54 IF60 IN02 IF38 IN03 IG49 I 1 'l i I i 1 1: 1 1 I I l l t i9 1262KT IB693 lR41 IB69W-.1232KT lt69A 142KT 144KT 1233ET lt69C 1256KT 1969V IR26 18696 1185KT I lH05- IG65 lN60- IF24 1G60 IF62 IN50 IG30 lH34 IF19 IG16 1F14 IHl4 IG61 INIT I L . 10 1 I i= 1 -I I I I I I i 1 1 1 1 110

[ -' 1235KT IR01 It49J 112KT 1270KT 1R39 1869E 1R64 1869F IR25 123KT I43KT IB69N IR49 1195KT I i

, 'IG34 IN06 .IF44 IN39 -IG51 IG06 - IF56 IN35 IF25 IG33 IG39 IH55 IF09 IN07 IG21 1 L 11 I . ~1 1 1 1 I I I I I- 1 I I I I lit 18KT =1220KT IR16 Ig6AS -1216KT 1190KT 1214KT 18697 1243KT 1276KT 1194KT 186A4 IR$2 140KT 13KT I IF06 lN30 IG31 IN33 IF15 IN32 IG11 IH59 IF31 I N40 . IG38 IN63 IF10 1 12 I l- I. I- 1 I l- 1 I I I I I i 12

. i t08 - 15682. IR20 It69R 1251KT IB6A9 IR11- IB6A7 133KT IS69P 1823 IB68U IR109 I IF05 IG27 IN48 '1712 IN28 IF04 IG29 I F18 . IN44 IF35 lH26 IG09 IF30 1 13 l- l- ..I. -1: 1 1. .I i i 1 1 1 I .I 13 1252KT 841ET 1868V-IR19 IB6AE-IR29 1222KT 1R33 IB6AD IR32 IB68x 1246KT 1275KT I IF36 -IF61 IH11 lG13 IN21 1G10 lN43 lG43 lN15 IF58 IF20 1 14 l 1 1 I . I I I I I I I i 14 124KT. IR158 15KT -IR44 18695 IR36 18694 1:140 14KT IR38 1261KT I

IG52 lN22 IG63 lH71- lG50 lN18 IG03 I lt 15 -l l l 1 1 I I i 15 e 1267KT 1255KT.1193ET 1263KT 1259KT 1245KT 1257KT I j

, R P- N M- L K J. H G .F E D C 8 A

• • • THE CONTROL COMPONENT ID FOR ALL BURNABLE Po! SON RG) AS$ENRLlES (SPRA85) SHOULD END WITH THE "*

"* LETTER 'K'. DUE TO SPACE LINITATIONS ON THIS MAP, ONLY 6 CHARACTERS OF THE CONTROL CCMPONENT "*

• • • ID ARE ASLE TO BE PRINTED ** NOTE THAT ALL BP ID'S $NOULO END WITH A 8K8 *"

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Page 5 2.2 Refuellne ENB Calibration - PT/1/A/4600/03E This procedure is performed for each cycle to ensure that Nuclear instrumentation System (ENB) setpoints and power indications are valid for cycle startup. Final adjustmorits are made using measured data as part of power escalation testing (See Section 4.0).

Power Range NIS calibration data was calculated by obtaining the last calibration data from Cycle 3 and the fuel assembly powers for selected core locations from the corresponding incore flux taeasurements. The beginning of cycle (BOC) 6 predicted powers for these locations were obtained from Duke Design Engineering's "Startup and Operational Report". Thn ratio of predicted BOC 6 power to measured Cycle 5 power was used to adjust the Cycle 5 calibration data.

Intermediate Range NIS 257. F.P. Reactor Trip setpoints were adjusted using selected assembly powers from the first full power flux map of ,

Cycle 5. The ratio of predicted BOC power to the measured BOC 5 power was used to perform this adjustment.

Table 1 lists the ratios and t.alibratton data calculated by this procedure. Figure 2 shows the core locations used to calculate the ratios. Calculations were performed on April 16, 1991. Calibrations were complete before entering Mcxic 3 (Hot Standby) on June 8, 1991, e

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Page 6 TABLE 1

• PREl.IMINARY CALIBRATION DATA >

TOR EXCORE INSTRUMENTATION Intermediate Range a

Ratio Cycle 5 Trip BOC 6 Trip

'- Channel (BOC 6/BOC 5)- Setpoint, g6 Setpoint. uA N35 1.079 98.9 106.7 N36 .1.085 86.6 94.0 Power Range i Ratio Axial Cycle 5 Full Power BOC6 Full Power

_ Channel- (BOC 6/EOC 5) Offset. % Currents, pA Currents, pA.

Upper Lower Upper Lower

+20 338.3 256.2 244.3 185.0

. N41. 0.722 0 294.7 295.8 212.8 213.6

-?0 251.1 355.4 181.3 242.2

+20 339.9 252.3 245.4 182.2

-N42 0.722 0 295.7 292.2 213.5 211.0

-20 251.6 332.2 181.7 239.8

+20 305.5 '233.7 217.8 166.6 N43 0.713 0 266.9 272.5 190.3 194.3 v -20 228.3 311.3 162.8 222.0

+20 346.8. 278.4 254.2 204.1

_ N44 0.733 0- 302.7 321.9 221.9 236.0

-20 258.6 365.4 189,6 267.8 t.

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Page 7 ASSEMBLIES SELECTED FOR PRELIMINARY CALIBRATION OF EXCORE INSTRUMENTATION

!/R 36 I P W R L I J N 5 F t i C e a 8

P/R P/ , t

. N4 ..... .... N43 I I I i 1  ! e i I I ,I i i I I I I I i i t i I I I g I i 1 1 I t l l l 1 1 I I i 1  ! ' I t i I t i I t i 3 9 l 1 1 8 9 8 I I I I i i  ! 3 e i t i i t t I i e i e i t i i i I t i i l I i l i 1 1 4 I t t I I i 1 I I I i 1 I t 4 I t i I t i I I I i l l t t

' I I I I I t i I I I I I I t I 31 1 I I I I I f I I t I t i I l3 1 1 I I I I I I I I I i 1 1 I e i 1 I I I I I I I I i l l 1 I I eI I I I i I I I I I I l l 1 1 16 8 I I t I I I I I I I I I 1 I I I t t I I i t i I I I I I I I I fI- t t I I I I I i t I t i I I IT i t i I I I I i 1 I I I I I I i i t I i  ! i l I I I I I I I i i ei t i i I l i I t i I I I I t se

! I I t  ! I I  ! 1 I I I I t i I t i I I I I I I I I I I I I , t 96 I I I I I I i 1 I I I I I I i9 8 9 t I i 1 8 1 I t t  ! I I e i

, f I I I 1 I I I I t I t I , t li t i 8 I I I t i I I , f I 1 I #13 t i 1 I I I I t I t i I I I I i 9 I t  ! I I I I I i 1 l I i i i 11 1 1 I i i i I t i I I t i I l- q:

1 1 I I 1 I i  ! I i  ! I I I t i I I I I I I I I I I I I I I 12 I I I I I I I I I I I I I I 12 1 1 1 I I I i 1 I I I I I l-i I i i I I I i t i i I i I i 11 I I I I I I I I I t i I I I 13 I t 1 l i I I I t I i t i t i l  ! I l  !  ! I I I I I 14 ' ' ' ' ' ' ' ' ' ' ' ' 14 P/R i I t t t i I t t I I t P/R N44 ...... ...... .... ....... _ .. _ .._. ....... ...... y p' i i iOiO!Oi i i >>

a p a n L E J u s r t e C e a Core Locations Used I/R Core Locations Used for for Pvr Range Cals. Intermed Range Cals.

Page 8 2.3 1/M Anoroach To Criticality - PT/0/A/4150/19 The initial critical!ty for each cycle is achieved under the direction of this procedure. The procedure consists of two major parts: 1) dilution to estinated critical boron concentration (ECB) and 2) rod withdrawal to criticality.

Dilution is performed in Mode 3 (Hot Standby) at essentially 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 enaute 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 (ECP) is calculated based on latest boron concentration. ICRR is plot ted for each source range channel. Rods are withdrawn until source range count rate doubles, the. 1CRR data is obtained and used to project critical rod position. If projected position is acceptable, rod withdrawal continues. This process is repaated until criticality is achieved.

Initial dilution was commenced at 1320 on June 10, 1991 and was concluded at 1733 following addition of 19,750 gallons of demineralized water. Initial NCS boron concentration was 2065 ppmB, the ECB was 1467 ppmB, and the final NCS boron concentration ,

o attained was 1485 ppmB. At 1900 on June 10, 1991, it was  !

discovered that an unacceptable ECP had been used to calculate the ECB (the ECP was established with control bants inserted too deeply, and therefore too close to the allowable Rod Insertion Limit). Via addition of 1052 gallons of boric acid the NCS boren concentration was adjusted to 1591 ppmB (the new ECB was calculated to-be 1570 ppmB) by 0045 on June 11, 1991. Figure 3 shows ICRR vs volume of domin water initially added.

-Rod withdrawal was commenced at 0520-on June 11, 1991. The ECP was calculated to be Control Bank C at 184 steps wd, however, extrapolations using points plotted on the graph of ICRR vs. % rod worth withdrawn indicated that criticality would be achieved below the allowable Rod Insertion Limit (RIL) of Control Bank C at 47 steps wd. Rod withdrawal was continued until actual ICRR was <0.2 at which time it was aborted due to persistent projection of criticality _below RIL.

Following insertion of-all Control Banks at 0830 un June 11, 1991,

, the unexpected.1CRR behavior was evaluated. It was decided that the noted phenomenon must be attributable to the placement of the source bearing fuel assemblies adjacent to the Boron Dilution Mitigation System (BDMS) channels instead of the Source Range NIS Channels-(as had been done in all previous cycles). This core design change resulted in an order of magnitude fewer counts being registered by the Source Range NIS prior to any control rod withdrawal (count rate being further decreased by the presence of l fresh fuel assemblies immediately adjacent to the source range l detectors, which acted as neutron " shields"). As a result, an t-

Page 9 artificially low baseline count rate was used to calculate ICRR as control rods were withdrawn. This lead to a prematurely rapid decrear-e in ICRR as the core's neutron population increased and the Source Range NIS began to register more realistic count rates.

To mitigate this phenomenon (an permit the reactor to achieve criticality) a new ECB of 1682 ppmB (100 ppmB above previous ECB) was established por the adiltion of 984 gallons of boric acid at 1530 on June 11, 1991. This assured an ECP far enough above Rlt to preclude projection of criticality below RIL during control bank w it hd rawa l . At 1620 approach to criticality was again commenced by withdrawal of control banks (shutdown banks still being fully withdrawn).

Withdrawal of control banks to all rods out was completed at 1730 without achieving criticality. Control Bank D was inserted to 100 steps wd. per procedure. New baseline count rates were established on the Sourca Range c; nnnels and a dilution of the NCS at < 30 gpm (monitored by ICRR plots) was commenced at 1813. Upon achievement of an ICRR < 0.6 dilution was halted (1300 gallons of domin water having been added). At 1930, following suf ficient time for mixing of the NCS, Control Bank D withdrawal at 10 step increments was commenced.

Criticality was finally achieved at 2021 on June 11, 1991 at 170 steps wd on Control Bank D. The ultimate c.ritical boron concentration was 1641 ppmB. Figure 4 shows ICRR plots maintained from Source Range Channel N31 during initial rod withdrawal (aborted attempt at criticality) and subsequent withdrawal of Control Banks to ARO (without achieving criticality).

3

Page 10 FIGURE 3 l l

I ICRR vs WATER ADDITION l

1 g2.R.- -31 ICRR (Co/C) l i

0 0 0.9 9 2

0.8 T O OO 0.7 F 00 0.6 i. 0 0 0

0.5 h 0 0 0nn-U 0.4 L

LOWER 0.3 ? LIMIT 0.2 E- EXPECTED 0.1 1. ICRR O= - - ' '- - '- ',, , , 0 .S.R. N-31 0 5000 10000 15000 20000 DEMIN URTER RDDITION g:.R.N-32ICRRCCo/C)

U 0.9 -O O.8 I OO 0.7 F 0

0. 6 i- 0 0.5 [ 0 0nn 0.4 _ 0 LOWER-0.3 f LIMIT 0.2 ? -- EXPECTED 0.1 i. ICRR 0 - - - - ' - - ', , ,,,, , , O S.R. N-32 0 5000 10000 15000 20000 1

DEMIN WRTER RDDITION

Page 11

. FIGURE 4 i ICRR vs ROD WORTH S N-31 ICRR (Co/C) 1_.R. _

0.9 :- _

0.8 E- .

0. 7 i-0.6 h SB e 222 e

0.5 :- SC e 222 e

0.4 .

0.3 ~- SE e 222 0.2 i: . SCB e 26 O

0.1 [-

O' - ' ' '- ' ' ' '- '

O 10 20 30 40 50 60 70 80 90 100 RIL ROD WORTH WITHDRAWN '4 S N-31 ICRR (Co/C) 1 _.R.

0.9 F _

0.8 _

0.7 ~-

ca e n 0.6 5- g 0.5 0.4 : -

CC 1 22 0.3 :- N 0.2 h cc e n u

CD @ 30 0.1 _--

E c- 1 25

' Ilco c 222 0 >

0 10 20 30 40 50 60 70 80 90 100 RIL ROD WORTH WITHDRRUN */.

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Page 12 3.0 ERO POWER PIIL9]CS TESTiliQ Zero Tower Physics Testing (7. PPT) is performed at the beginning of each cycle and le etentrailed by PT/1/A/4153/21 Post-Rotun!fng Controlling Procedure for Startup Testing. Tcat measurements are mado below the point of nue. lear hont using the output of one power range detector contected to a reactivity cowputer. Hensurements are compared to predicted data to verify coro design. The following measurementa arn Analuded in the ZPPr program:

>

• Source Range /intermediato Rango Overlep eloint of nuclear heat audition eRaar.tivity Computer Chockout

• All Rods Out Critical floron Concentration

• All Rods Out isothermal Temperature Coef ficient

• Reference Bank Worth by dilution

• Reference Bank in critical boron concentration

• Differentist 'ron Wort h determination

• Control F . Worths by rod swap Zero %r physite testing for Cycle 6 began at 0520 on June 10, 1991 with Ji- beginnlug of rod withdrawal for initial approach to criticality.

ZPPT '

complete by 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on June 12, '4H following analysis of rod worth measurements. Tablo 2 summarizes Z..f results. All acceptance critoria were met.

Sections 1.1 through 3.10 describn ZPPT measurements and resnits.

Page 13 TAHl.1: 2

SUMMARY

OF ZPPT RF.SULTS l'RIAtttilt lltaf UIMI Ya LRQ PLtd1cted Valun/Aq.cip10Act G.I.llfIlt!

Nuclest liest 1.19 x 10 arnpu - - -

(on N42)

ZPPT Range 10 to 10 nmps --

(on N42)

ARD Boron Conc. 1662 ppel? 1664 1 50 ppmR ARO ITC + 1.08 pcm/"r + !.49 1 2.0 pcm/*P ARO HTfi + 2.51 pcm/'F + 2.92 pcm/'F Rnference Bank 849.5 pcm 911 pcm (Control Bank C)

Worch Rcf, Bank In 1552.6 ppmB 1561 ppmB Boron Conc.

Differential -7.71 pcm/ppmB -8.84 pcm/ppmD Boron Worth Control Bank D 550.1 pcm 566 pcm Worth Control Bark B 772.5 pcm 705 pcm Worth Control Bank A 333.4 pcm 369 pcm Worth Shutdown Bank E 475.3 pcm $17 pcm Vortti Shutdown Bank D 372.9 pcm. 375 pcm Worth Shutdown Bank C 375.0 pcm 375 pcm Worth Shutdown Bank B 805.1 pcm- 785 pcm Worth Shutdown Bank A 240.3 pcm 232 pcm Worth Total Rod Worth 4774.1 pcm 4835 pcm

.. ._m...m.-__.___._..._____.____. ~,__ _ _ _ _ - - - - - _ _ _ _ _ . _

I Page 14 3.1 figyrce Rance/ Intermediate Range Ovt.depAla PT/1/yt t yyM During the initial approach to criticality, source range and -

intermediate range NIS tlata was obtained to verity than at least e one decade of overlap existed. I f <>ne decade of overlap did not l exist, intermodlate range compensation voltage would be adjusted to '

provirle the overlap. i Overlap data for 0ycle 6 was obtained per PT/1/A/4150/21 Post Refueling controlling Procedure for Startup Testing, on June 11, 1991. Table 3 contains the overlap data. The acceptance >

criterion was met.

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TABLE 3 I SOURCE RANGE / INTERMEDIATE RANGE OVERI.AP DATA -

i SOURCE RANGE INTERMEDIATE RANGE (CPS) (AMrs)

M M M UM INITIAL DATA NIS Cabinet 200 200 1.0 x 10'8' 1.0 x 10'88 OAC 222 156 1.2 x 10'88 1.3 x 10*88 t

SOURCE RANGE B1,0CK NIS Cabinet 9000 6000 1.2 x 10'8' l.3 x 10*8' OAC 7253 5257 1.1 x 10 ~8' 1.2 x 10'8' r

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l.

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

Page 16 3.2 l'q1nLgL.Mughjir HeaLMdit ion - PT/11NM101U The point of nuclear heat addition is measured by trending reactor coolant system tempornture, pressuriznr level, flux Invol and reactivity. A nlow, constant startup rate is initinted by tod withdrawal. An inctenne in reactor coolant nyntem temperature and/or prennurizer level and/or a change in reactivity and/or rate of flux level inct,nso indicate f.hn addition of nuclent hent. The measurement is repeated to ensurn confidence in results.

Por Cycin 6, the point of nucient hont addition was dntermined per PT/1/A/4150/21, Post-Refunling Controlling Proceduto for Startup Testing, on . lune ll, 1991. Tahin 4 nummarizon the datn obtalnnd.

~

The ZPPT tent hand was set at 10 to 10 ' nmps on N 42 (power range channel attached to reactivity ccmputnr). This provided more than a factor of two margin to nuclear heat for ZPPT. The acceptance criterien was satisfied.

L Page 17 i'

IABLE 4 NUCI. EAR HEAT DETERMINATION i

REACTIVITY COMPUTER INTERMEDIATE RANGE i (AMPS) (AMPS)

. N42 EjL. N36.

RUN #1- 1.49 x 10 1.44 x 10 1,51 x 10

RUN #2 1.19 x 10 1.15 x 10 1.21 x 10

• l

- Upper limit of _ test band is minimum reading at point of adding heat divided by 2 ort 1.19 x'10-amps + 2 a 5.95 x 10 ampn l

I I

ZPPT TEST-BAND: 10 to 10 nmps on N42 l l

l h

t I

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,-en&+yem -,pr-r gyy,-,-gne e mw -9tvv- t py r -r+wpw1w'v'ev'ewwvD ww ww

Page 18 b

3.3 Reactivity Computer Cbeckoul - PT/1/A/4150/21 The reactivity computer checkout is perforrned per PT/1/A/4150/21, Post Refuelitig Controllitig Procedure for Startup Testing, to verify that the power range channel connected to the reactivity cornputer can provide reliable reactivity data. Reactivity insertions of approximately 425 -25, 440, and -40 pcm are mado and the period calculated by the reactivity computer in recorded. The sneasurc<l renctivity for each case in cornpared to the theoretical reactivity obtained from the doubling time and verified to be within 4%.

The checkout was performed for Cycle 6 on June 11, 1991. Table 5 lists the results of the 4 trials. The acceptance critorion was met in all 4 cases.

Pago 19 TABI.E 5 REACTIVITY COMPUTER CllECE0T WESTINGilotISE DIGITA!. REACTIVITY C0!iPUTER REACTIVITY PERIOD COMPUTER TilEORET1 cal PERCENT ABS 01.01 E (sec) AP (pcm) REACTIVITY ERROR

• DIFFERENCE A Q qm) _ ,Dd __ ( rtcm )

231.09 +28.61 +28.68 0.24 0.07

-386.13 -21.65 -21.69 0.17 0.04 137.09 +43.98 +44.81 1.86 0.83

-238.97 -37.14 -37.78 1,69 0.64 l

• PERCENT ERROR = densured - Thr.qrp_Ltc3.1 x 100 Theoretical Acceptanco Criteria: Percent error < 4.0% QB absoluto difference < 1 pcm, whichever .is greater.

W l "

l

Page 20 j 3.4 BRO Doron FdLdooint Measurement - PT/0/2/_4D 4 om This test is perforrned at the beginning of each cycle to verify that measured and predicted total core reactivity are consistent. The test in performed near the All Rods Out (ARO) configuration.

Reactor Coolant Systern boron samples are obtained while Control Bank D is pulled to the fully wittdrawn positlon. The reactivity dif ference f rom criticality in measured and converted to an equivalent ht.ron using the predicted differential boron worth. This equivalent boron is added to the avtraged measured boron concentration to obtain the ARO critical baron concentration.

1ho Cycle 6 BOC, llot Zero Power (llZP), ARO critical baron concentration was measured on June 12, 1991. The measured boren concentration (average of 3 namplen) was 1659 ppmB. This value was corrected by 3.0 ppmB (equivalent boron worth of withdrewal of Control Bank D from 204 steps wd to 222 steps wd) to yield an ARO concentration of 1662 ppmB. The acceptance criterion, measured boron concentration within 50 ppmB of predicted, was met.

Page 21 3.5 ARO _ ! sot benal_'{tergIALqutJtellichttLJirMHimni - PT/3/ A /4150/12A The ARO Isothermal Temperatuto Coef ficient (110) is measured at the beginning of each cycle to verify consistency with the predicted value. In addit lon, the Moderator Temperature Coci ficient (H1C) is obtained by subtracting the doppler temperature coefficient fron the ITC. The HTC is used to ensure cornpliance with Technical Specification limits. The ITC is measured by slowly (< 10'F/hou.)

changing reactor coolant temperature while measuring reactivity versus coolant temperature. 1ho slope of the reactivity versus temperature line is used to determined the ITC. At least one cooldown and one bestup is per f ortned.

The DOC 6 ARO ITC was measured on June 12, 1991. The initini hentup (and subsequent cooldown) was repeated due to unnnticipated dilution of the NC System during the hent up. This dilution occurred as pntt of a chemical addition to the NCS. The results of t he second heat up and the two cooldownn were averaged to obtain the mensured ITC.

This data is summariznd on Tahin 6 Averaged measured (corrected to 557'F) ITC was + 1.08 pcm/*P.

Predicted ITC was + 1.49 pcm/*P. Hensured ITC wan within the acceptanen criterion of predict ed i ?. pcm/*F.

The HTC was determined to be 4 2.51 pem/'F. This value was used with procedurn PT/0/A/4150/21 Temporary Rod Withdrawal himits Determination, to ensure that HTC would remain within Technical specification limits at all times, PT/0/A/4150/21, Temocrary Rod Withdrawal Limits Determination, is described in Section 3.10.

l'ans 22 TABLE 6 ITC HEASUREMENT liESUI.TS i

AI. C'r) As_lisal T-Avr i"F) JTC.(Pras/*El First Cooldown -4.2 6.0 $$5.7 41.43 (Uncorrect ed)

Second llentup 44.1 45.4 555,4 +1.32 (Uncorrected) i Second Cooldown *4,85 -4.3 S55.3 40,89 (Uncorrected)

Averago 41.08*  ;

• Average-!TC carrected to nominal ll7,P temperature of 557*r BOC llZP doppler coef ficient = -1.43 pcm/*F HTC = Averagn ITC - Doppler Coeff. = 42.51 pt.m/*F i

4 h

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Page 23-4 1

3.6 Ritference Esnk Worth ticanttsment bv Dilution - PT/0/A/4150/11A l

The Control Rod bank predicted to have thn highest worth is ,

designated as the reference bank and is measured by inserting '

)

the bank (with all other rods fully withdrawn) In discreta t.teps, while slowly diluting the NC baron concentration. Tho ,

reactivity worths of the discrete steps of tal insertion are measured by the reactivity computer and summed to obtain tho  ;

inserted worth of thn reference bank as a function of bank height.

5 The 1100 6 Refernnce Bank (Control llank C) Worth was measured on June 12. 1991. The rnference bank was measured to be worth 849.5 pcml predicted worth was 911 pcm; and the acceptanco criterion una that mnasured worth was more than 774.4 pcm and less than 1047.7 pcm (911 pcm i 157.) .

I r

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a w - p yw,, y -

wr-w-,-,,-g- ey e.-g--3 -w-.+i-w---ye- ,-*+>wy--+-wq.wew,-+,=+= e-,wwm<-ew- ww '-= ww-w---,

. . i Page 24 3.7 Reference Bank in Boron Endpoint Menstrement - PT/otA/41so/lo 4

This test is performed to measure the critical boron concentration with the referenco control rod bank fully inserted and all other control rods fully withdrawn. The measured boron ncentration is used with the measured ARO beron concentration and the measured worth of the reference bank to calenlate the dif ferential boron worth. The test is performed by measuring the reactor coolant

., system boson concentration with control rods near the desired configuration. Control rods are inserted or withdrawn to desired configuration concurrently with boron samples. The reactivit) difference from criticality is nonsured and converted to an equivalent boron concentration using the predicted differential boron worth. This equivalent baron is added to the averaga mnasured baron cos. centration to obtain the Reference Bank-in boron concentration.

The BOC 6 Reference !!ank-in boron concentration was measured on Juno 12. 1991. The mnasured boron concentration (average of 3 samples) was-1556 ppmil. This value was corrected by 3.4 ppmB (equivalent boron worth of insertion of Control Bank C from 32 steps wd to O steps wd) to yield a Reference Bank-in baron conenntration of 1552.6 ppmB. The predicted value was 1561 ppmB. There is no acceptanca critnria directly associnted with this test.

l l

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page 25 3.8 Dif f erential Itoren Votiltlti_eLaitLn11En - pT/1/ffl.tl1Q/ll The dif fnrential boron worth in cniculated f rom the moanured AKO critical baron concentration, Reference annk-in critical boron concentration, and integrnl worth of the rnierence bank. The calculated valun is compared to predicted value to verify consistency. This calculation alno providow an indirect check of the mensured reference bank worth and of tbn boron endpoints.

The DOC 6, ilZp ditferentin) boron wotth was calculated to bo

-7.777 pcm/ppmB (849.5 pcm + (1662 ppmB -1552.6 ppm!O). The predicted value wnn -8,84 pcm/ppmB and the accept anco criterion wan -8.84 1 1.33 pcm/ppmH (predicted i 15%).

Page 26  !

3.9 Qqntrol Rod Worth 1:Manutent by Hod Swan - PT/0/A/411QL112 The worth of all cont ol rod banks except the reference bank in measurnd by inserting each bank while withdrawing the reference bank l and/or previously menn tred bank to maintain finar critical conditions. When the t arik boltig measured in fully innert ed, t he ref.s ence bank position is adjusted to achiavo critical conditions E wita all other rod banka fully withdrawn. The wort h of the innerted rod bank is determinnd fr w the critical height of the referenco bank. Thn measured worth is compared to predicted worth to verify conslatency. Also, thn sum of worths of all banks (fricluding reference bank) in compared to predir.ted.

The D006 rod worths were measured by rod bwnp on June 12, 1991.

Table 7 summarir.cn t ha result s of the mensurements. -All

• ncenptatico criterin wnro mot.

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Bc - 2 1 1 1 1 1 e a B i R d

.t o e fi f t e t er o r c RC ) h r e i e A B C D E t h o r d C c A B D r t t e e n n n n n n o r c f r l e w w w w w l l l W o a e P or o o o o o o o o W F R (

re d d d d d r r r l k tf t t t t t t t t a n ne u u u u u n n n t a oR h h h h h o o e o B C( S S S S S C C C T 1 2 3 4 j 1 l 4 1 = 4 f' lI  ! <j  !

, Page 28 j i .'  ;

3,10 '[tE22rary Rod _W111dulgttL,jtim111.llut1Rintilpn - PT/Q/2/3} fig /2q PT/0/A/4150/21 Temporary Rod Withdrawal Limits Determination in .  !

performed at the beginning nf each cycle to determino if any control ,

rod withdrawal limita are needed. The measured ARO HTC from PT/0/A/4150/12A, Isothermal Tempornture Confficient of Ronctivity t

!!onsuromont Ja used with the predicted HTC. vs. Iloron Concentration to '

calculato if the MTC will equal the Tech Spec Limit (minun 20 pcm/*r  ;

for conservatism) at the measurn HZP ARD liron Endpoint. If thn Tech  :

Spec Limit will be excended, rod withdrawnl limit s are imposed bnned on the offeet of.pownr invol on HTC and rod position on boron 4 concentration. Oparation with rods positioned below thlh limit ensures that the !!TC remains below Toch Spec limits.

• For:Il006, PT/0/A/4150/21 wan performnd on June 12 14, 1991. No withdrawal limits were nocennary bnlow 70% full power. Ilowever,  ;

control-rod withdrawal limits woro nnn!.ed and imposed for > 95% full  ;

power until cycle burnup was grontor than 12 EITD DE baron concentration wan lonn than 1635 ppmH. Tahic 8 lists the results of .

the test.-nnd shown the temporary rod withdrawal limits that wern imponed via the Unit 1 Ronctor Operating Data Manual.

i.

_ _ _ _ _ . _ _ . __ _ __ . _ . _ _ _ _. ~..~..__ _ . _ _ ... _ ..,_.. ..._._ . _ . . _ _ _.

Page 29 1

TABLE 8 TEMPORARY ROD WITilDRAWAL LIMITS RESULTS ARO.117.P, Boron Endpoint ( f rom PT/0/A/4150/10): 1662 ppmi) ting Boron Concentrations (ppmB);

Rod Position  % Pull Power (Control Bank D) 100 222 1635 203 1642 183 1652 163 1663 Rod Witharawal Limits imposed:

CONTROL BRNK D POS1110N, STEPS 220_ _

210: :-

4, 200 -

1 190 : /

180  ! ' -- 1001 170 :_

~

160 ,

150 :

140  :,

HZP,RRO BORON 130  :

-*-- 100*/ POWER 120 1500 1550 1600 1650 1700 NC BORON CONCENTRRTION, PPMB

1 ,

Page 30 4.0 l'QEER ESCALATION TESIlliQ Power escalation testing in performed during the initial power escalation of each cycle and in controlled by PT/1/A/4150/21, Pont Refueling Controlling frocedure for Startup Testing. Tests are per formed f rom 0%

through 100% power with major platenun at ~ 30%, - 801, and ~ 100% power.

The following significnnt tents arn performed in the power escalation tent programs eCorn power distribut.fon (~ 30%, ~ 801, and ~ 100%)

einterim incore/excorn calibration (~ 30%)

• Full power delta temperaturo extrapolation (~ 30%, ~ 801, and ~ 1001) eincorc/Excore Calibration Check (.~ 801)

• Incore/Excorn Calibration t~ 1001,)

• Precision colorimet ric reactor coolant flow measurement (- 100%)

ellot Full Power (llFp) critical boron concentrat ion mennurement

(~ 100%)

l Also during power escalation, checka are performed for the incorn dntactor system, thermal power program and inputs, various other procons computer programs and functions related to corn monitoring, atens flow feed flow mismatch, turbino impulan pressure, intermediate ranga trip satpoints, nte. The renults are not included in this report.

Power escalation testing for Cycle 6 beenn on June 14, 1991. Full power was reached on June 24, 1991. Full power testing was completed on July 1, 1991.

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

9 s

l

, e .

• l'a r, o 31 4.1 [ m g _l'c C ] h i d q u ejt _1 1 [ 1] M 4j j pf_0j Core power dist ribut ion monstren:ent s are pet formed during power escalation testing at low power (~ 30',), Int ermodint e power (~ 807.),

and full power (~ 100',) . Mensurements nie made to ver ify that thn llent Flux hot Channel Enctorn, F are witt in llenits and to veriiy ilux symmetry. 11ennuremerit a n re m1dh, wit h the inco e moveable detector syNtem and analyzed with Tihangattram Nucient Annor.inten CORF, packnge, linbcock and Wilcox Monitor pncknge is used to cynlunte penkirig f actors with tempect to predicted limits not forth in ihn COI.R . This ptogram determinen Fg Operational, r gkPS, niid F 3g margina. Additional mensurements may be requireil if these margins are exceeded during power onenintlon.

The 1100 6 core power dist ribut ton niennurements were per for med on June 15, 1991 (30% power), June 12, 1991 (75% power), and Jutie 2, 1991 (1001, power). Tablen 9 through 11, respectively, nummarizn the rnsults. Acceptance Crit erin for core power distributtoti were met in all comen.

4 4

_ . _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ w

o ..

Pago 32 l

TAllLE 9  !

CORE POWER DISTRIBUTION RESULTS l 30?. POWl;R eMap ID: FCM/1/06/001

• Date: June 15, 1991

• Power Loyol 29.658% F.P.

eCycle Bernup: 0.274 EFPD, 11.07 MWD /MTV

• Boron Concentration: 1497 ppmil

• Control Bank D Position: 214.5 ntops withdrawn

• Core Average Axial Offset: 16.709%

eHaximum F"A(X, Y, Z): 2.1171 at Core Location 11-11

• Hinimum Margin to Surynllianco F : 52.96% at Corn Location B-11 9

• RHS Error of Detector Responso: 4.01%

• Reactor Coolant Flow: 395,501 gpm; 102.047 of Tech. Spec. Flow

'Haximum F AH Error (From Predicted): 8.59% at Corn Location J 08

• Hean of Absoluto F AH Errors: 3.08%

Entire Top italf, Bottom llalf,

• Ti1t Ratios: Core 11einh1 Ngigilirr,i Egiptq),ited Quadrant 1 0.98756 0.98767 0.98742 Quadrant 2 1.00703 1.00775 1.00601 Quadrant 3 1.00947 1.00542 1.01515 Quadrant 4 0.99593 0.99915 0.99142

'Hir.imum F q Operational Margin: +5.43% at Corn Loention P-09

+Hinimum Fq RPS Margins -3.69% at Core Location D-12*

• Minimum F 3g Margin: +2.05% at Corn Location 11-13

• Margin violation required no extra surveillance (l.c. additional flux mapping).

Page 33 TABLE 11 CORE POWER DISTRIDUTION RLSULTS 80% POWER eMap ID: FCM/1/06/002 eDate: June 23, 1991

• Power Level: 75.51% P.P

• Cycle Burnup: 2.520 EFPD, 101.84 MWD /MTU eBoron Concentration: 1335 ppmD eCont rol Bank D Ponit ion: 211 steps withdrawn score Average Axial Offset: 3.754%

eMaximum F ^(X, Y, Z): 1.8591 at Corn Location B-11 eMinimum Margin to Surveillance P q: 38.717. at Core Location B ll

• RMS Error of Detector Response: 2.55%

• Reactor Coolant Flow: 393,029 gpm; 102.09'. of Tech. Spec. Flow

• Maximum F AH Err r (From Predicted): 6.10% at Corn Location C-08 eMean of Absolute F AH E " I" ' I'89%

Entire Top Half, Bottom llalf,

• Tilt Ratios: C. ore lleigts. Normaligni Normalized Quadrant 1 0.99294 0,99623 0.98939 Quadrant 2 1.00770 1.00585 1.00969 Quadrant 3 1.00318 1.00178 1.00469 Quadrant 4 0.99619 0.99615 0.99623 eMinimum qF Operational Margin: 44.92* at Corn Location J-14

• Minimum F q RPS Margin: 44.93% at Core !.ocation D-12

• Minimum F Margin: +2.931 at Core Location 11-13 AH

_ _ _ . . . . . . _ __ ._ . - _ _ _. . = . . . _ _ _ _ . . . _ __

Page 34 TABIE 11 CORE POWER DISTRIBUTION RESULTS 100% POWER eMap ID: FCM/1/06/003 eDate: June 26, 1991

• Power Level: 99.73%

• Cycle Burnup: 5.95 EFPD, 240.47 MWD /MTU eBoron Concentrution: 1187 ppmB

• Control Bank D Position: 214 steps withdrawn eCoro Average Axial Offset: -0.679%

MA eMaximum Fq (X, Y, Z): 1.8445 at Core Location B 11

• Minimum Margin to Surveillance Fq : 19.88%

• RMS Error of Detector Responses: 1.98%

• Reactor Coolant Flow: 391,583 gpm; 101.71% of Tech. Spec. Flow

• Maximum F AH Error (From Predicted): 4.86% at Core Location C-08 eMean of Absolute F Errors: 1.31%

AH Entire Top H.Sif, Bottom llalf,

' Tilt Ratios: Core Heicht Normalized Normalized Quadrant 1 0.99288 0.99431 0.99146 Quadrant 2 1.00943 1.00694 1.01189 Quadrant 3 1.99874 0.99638 1.00108 Quadrant 4 0.99895 1.00238 0.99557 eMinimumqF Operational Margin: +5.84% at Core Location B-11 eMinimum F q RPS Margin: +6.57% at Core Location N-12 eMinimum'F Margin: +4.92% at Core Location J-10 AH

'+

lage 35 i 3

. L e

4.2 Interim incore/Excore Calibration - PT/0/A/4600/05D The test is performed using power range currents measured during the 30% core power' distribution measurements with the axial offset from that power distribution to obtain calibration data for the power ranges. Power rangen are calibrated before execeding 50% power to ensure that an accurate indication of corn axial flux difference (APD) '

and quadrant power tilt ratio (QPTR) is available for Technical Specification surveillance. This "1 point" calibration in 11 mind in ,

that now 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 5 interim calibration was obtained on June 15, 1991.

Power range calibrations were completed on June 18, 1991. The results  ;

are given on Table 12. All acceptance criteria were met.

v l' .

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-, ,,n-.w.,-a- 3 . - ,,- .-ovq r -ev---vv e-w=--r-o---www= we e---w ivm-m _._ +_w.-e w e e-- e -e. = -+ - -- + - - * --<v

- - %%---sew - + ' w*--vw.wei ++em-w'--- + -----r

, Page 36 TABLE 12 INTERIM INCORE/EXCORE CALIBRATION RESULTS

1) Flux h., p- Flux Hap I.D. FCM/1/06/001 Power Level 29.658% F.P.

Axial Offset +16.709%

Measured Pownr Rango Currents, pA 11 4_1 ti42 ti4] R44 Top: 79.4 72.2 63.3 68.9 Bottom: 67,8 58.8 52.8 59.7

2) Ratio of measured currnnts to expected (f rom last calibration data shown on Table 1) currents:

14.1 tii2 lihl ILS Top: 1.1195 1.0137 1.0007 0.9334 Botto.': 1.2050 1.0605 1.0401 0.9614

3) New Calibration Dato:

Powar Range Cuirents, pA Axial N41 N42 N43 N44 Offant, % Upper Lower Upper Lower Upper Lower Upner Lower

+20 273.5 222.9 248.5 193.2 217.9 173.3 237.3 196.2 0 238.2 257.4 216.4 223.8 190.4 202.1 207.1 226.9

-20 203.0 291.8 184.2 254.3 162.9 230.9 177.0 257.5 Correction Factors (Hj)

N41 N42 N43 N44 1.418 1. 19 1.393 1.425 I

l I

4 j

e Page 37 l

, l t

l 4.3 Delta Te perature Extrapolition - PT/0/A/4150/16 i

Reactor Coolant System (NC) hot leg and cold leg temperature data in '

obtained per PT/0/A/4150/16. Unit 1.ond Steady Stato, at ~ 80%, and ,

~ 100% power to ensure that delta temperaturn (AT) indicativns are accurate. AT. In units of percent of full power AT (% FPAT), is an important parameter in the reactor protection nystem. The reactor ,

trips associated with AT. overtemperaturn AT (OTAT) trip and overpower  !

AT (OPAT) trip. are important with respect to TSAR accident analysns.  !

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

-setpoints.

The % TPAT indications rely upon a valid number for thn full power AT

('T). or AT,. The AT. valun is the valuo.which in determined and +

checked by PT/0/A/4150/16. An extrapolation is performed from ~ 80%  !

'using measured data at that power level to check AT for nach loop.

The extrapolation fm based on enthalpy, as AT la noE completely linear with power. If extrapolated full power AT is more than 0.6*F (~ 1%)

different from current AT , AT is adjusted by station instrument and electrical (IAF.)technicfEns. At 100% power, the AT are checked again. preferably using data obtained during.the proEision calorimetric flow mt surement (see section 4.?) and adjusted as necessary/ desired.

Table 13 summarizes the results of the AT extrapolation. A,l l acceptance critoria were met.

h 4

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- - , , . , - . . . .a.. -,m.- ,4-.,,. ,,.w,_,.. v,,., c.n.,,,. _ , . . , , . _ . . , . . , . . , , , , . . , , , , , _ , , , _ , , , . , . . , . . , . , , . . . , . ,.,,.-.,....,..-,.,..,..,.n.-, .,-.,,s

i l'a g e 38 TAlil,E 13 AT f.XTRAPol.ATION l'ATA POWER = 75.5156% LegitA Lggiti! bggy,,y hgpyt[}

T COLD, "F 557.3 556.6 557.3 556.9 T il0T, 'F 602.5 600.7 603.0 601.9 Ah liTV/lb 60.510 $8.812 61.230 60.152 Extrapolated Full Power Values:

7 C01.D 'T 561.0 591.0 561.0 561.0 T-il0T, 'F 618.8 617.3 619.4 618.5 Ah, BTU /lb 80.129 77.881 81.083 79.655 FPAT, 'F 57.8 56.3 58.4 57.5 Current AT , 'F 59.23 56.92 59.72 57.51 Error (AT7'FPAT), 9

• r -1.43 0.62 1.32 -0,01 POWER = 98.48% bep1LA Lopn E hggn_X hgqlLJ)

T C01.D . 'F 561.0 560.0 561.0 560.3 T-il0T, *F 618.6 616.I 619.0 617.7 Ah, DTU/lb 79.837 77.254 80.457 79.333 Ex t rapol a t egl Full Power Value:

T C01.D . " F 561.0 561,0 561.0 561.0 T-ll0T, 'F 619.4 616.9 619.8 618.5 Ah, BTU /lb 81.069 78.446 81.699 80.557 FPAT, "F 58.4 56.9 58.8 58.2 Current AT , *F 57.80 56.30 58.40 57.51 Error (AT7FPAT), g

'T 0.6 0.6 0.4 -0.7

[iQIlf t 1. Extrapolation is a linear extrapolatton of Ah, including Ah =0 at 0% power. For purposes of extrapolation, T-C01.0 is assumed to be 561 F at full power. This assumption has small effect on extrapolated AT (~ 0.l*r AT error /1.0*F T COLD error).

2. AT for !. oops A D, and C were adjusted following extrapolat on using ~ 80% dat a.

3. AT for each loop was adjusted following extrapolation us. tag

- T00% data.

Page 39 4.4 Jngne/Excore Calibn11trLQkeik - PT/1/ A/4600/05B An incore/excore calibration check is performed with the ~ 80% power flux map results. The purposn of the check is to verify that the calibration performed at a lower power level in still valid. The chango in reactor coolant temperature and the (.hange in radial power distribution associated with the change in power affects the number of neutrons detected by the power range detectors and therefore can affect the calibration.

The check was pntformed for BOC 6 on Juno 23, 1991. Power level wns.

75.51% and cycle burnup was 2.520 Erl'D. Table 14 lists the results of the check.

A result of > 2%, but i 31 dif f erence between measured Incoro AFD (from flux map) and indicated AFD (f rom Excore NIS) requires that PT/0/A/4600/0$D Interim Incore/Excorn calibration be repeated. A result of > 3% dif ference requires that PT/0/A/4600/05A, Incore/Excore calibration be performed prior to continuing power escalation. In the case of this test, the calculated differences for all four Excore his channels were < 2%, allowing power escalation to continue to 100% F.P.

without further NIS calibration.

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

., Page 40

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l

-TABLE 14 .

INCORE/EXCORE CALIBRATION CHECK RESULTS l Flux Hap ID: FCM/1/06/002 Cycle Burnupt 2.520 EFPD Average Power Level: 75.51%

.Incore Axial Offset: +3.734%

.Incore Axial Flux Difference: +2 83%

c Average Excore AFD and difference from incoro AFD

v AEQ Differengg Quadrant 1 (N43) +4.17% +1.34%

Quadrant 2 (N42) +4,40% +1,57%

Quadrant 3 (N44)' +4,22%~ +1,39%

l Qurdrant 4 (N41) +4,07% 31,24%

i L

Acceptance Criteria: Absolute value of diff rence between excore AFD and incorn AFD is 5 3,0%,

F

. , . . . _ . . - - ~

l

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y page 41 i

i 4.5- Incore/Excore Calibration - PT/0/A/4150/05A The excore power range channels are calibrated por PT/1/A/4150/05A,  ;

Incore/Excore Calibration, at 100% F.P., Quarter Core Plux Maps (QCMs) l and full Core Maps (FCMs) are taken at various axist offsets while at a

, constant power level. Power range upper and Inwer chamber indicated powers (%) averaged are during each map. A linear least squares fit of the output of each chamber (normalized to 100% power) is determined.

The slope and intercept of the upper and lower chamber of each power range are used to determiae the <:a ' 'bration data (full power currents

and correction factor, Mj) for t' -

S nnel.

This test was perf.rme d ',r B0ct June 27, 1991. Eight maps with axial of fsets ranging from -9.648% to 44.746% were obtained. An additional QCFM was taken but not used as its data was not consistent with the leaet squares fit.

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

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Page 42 TABLE 15 INCORE/EXCORE CALIBRATION RESULTS 4

-Powec Range Calibration Currents, pA

.i Axial .N41 N42 N43 N44 Offset, 7 Upper Lower Upper Lower Upper Lower Upper Lower

+20 298.7 239.1 272.9 208.0 240.9 188.3 258.7 209.6 0 264.9 276,4 241.9 240.1 213.5 217,9 228.7 241.6

• 20 231.1 313.7 210.9 2:2.2 186.1 247.5 198.8 273.6 e s

Correction Factors (Hj)s N41 N42 N43 N44

-1.522 1.527 1.515 1.519 t

W

+

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TABLE 16 INCORE/EXCORE CALIERATION INPUT DATA Power Range Indicated Power, .*

Power Azial NI 1 h42 N43 N44 Map ID Level,% Offset,% Upper Lower Upper Lower Upper Lower Upper Lower FCM/1/06/003 99.73 -0.679 110.3 106.9 110.8 106.8 111.0 107.3 109.3 106.0 QCM/1/06/005 100.30 -6.202 107.0 11 107.6 111.6 107.9 112.3 106.1 110.7-QCM/1/06/006 .99.98 -8.102 105.8 113.5 106.4 113.3 106.7 114.0 105.0 112.4 FCM/1/06/007 99.89 -9.648 104.1 '114.5 104.7 114.4 105.0 115.0 103.4 113.5 QCM/1/06/008 99.90 0.536 110.9 107.0 111.6 106.9 111.9 107.4 110.2 106.1 QCM/1/06/009 99.65 1.693 111.9 105.9 112.5 105.8 112.9 106.3 111 3 105.0

, QCM/1/06/010 99.66 3.235 113.0 104.7 113.7 104.7 114.1 105.2 112,4 103.9 QCM/1/06/011 99.70 4.746 114.7 103.9 115.4 103.8 115.7 104.3 114.1 103.1 1

Least Squarcs Fir Data:

Fit Correlatior 0.9964 -0.9961 0.9964 -0.9958 0.9962 -0.9959 0.9957 -0.9954 i

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Page 44 1

! 4,6 Ilot Full Poqqt Critical Boron. Concentration - PT/0/A!4150/04 The hot full power critical boron concentration is measured per PT/0/A/4150/04, Reactivity Anomaly Calculation. The measurement is made as close to the reference conditions (full power, TAVG = 590.8 F, equilibrium xenon, ci,stgn samarium, all rcds out) as practicable, then 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 systen baron samples

• Reactor power

• Control rod position and inserted worth

• Xenon worth

• Semarium worth

• T-AVG

• Cycle Burnup For BOC 6, the test was performed on Juua 27, 1991, at 100.0% power with a burnup of 5.861 EFPD, The evaluation of the measured ilFP Critical Boron Concentration was parformed per PT/1/A/4150/21 Post-Refueling Controlling Procedure for Startup Testing.

The results of the evaluation are as follows:

Predicted Boron Concentration -

1190.7 ppmB Predicted adjusted for ilZP error -

1188.7 ppmB Measured Boron Concentration -

1173.1 ppmB Difference between measured and adjusted prediction -

-15.6 ppmB The acceptance criterion (absolute value of dif ference < 50 ppmB) was met, i

,1 g ..=%,

i page 45 s

4.7 - Calorimetric Reactor Coolant Flogleasurement - PT/1/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 plant 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 6, the calorimetric flow measuremnnt was performed on

-June 26, 1991., Three test runs were performed; the average of the results were used to generate the elbow tap correction factors. Table 17 summarizes the results. All seceptance criteria were met.

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TABLE 17 Page 46 CALORIMETRIC REACTOR COCLANT FLOW MEASUREMENT DATA

• RUN #1:

Lpon A Loon B Loon C Lggp_Q Iyid Power, MBTU/Hr 2976.75 2852.17 2938.24 2845.64 11,612.8 T-HOT, 'F 619.3 617.1 619.6 618.3 (99.776%)

T-COLD, 'F 561.3 560.2 561.9 560.7 NC Flow, MLBM/4r 36.885 36.292 36.621 35.568 NC Flow, GPM 99,714 97,955 99,094 96,048 392,811 (102.03%)

Elbow Tap Coef ficients:

Channel 1 0.300304 0.305301 0.289790 0.299121 Channel 2 0.2954f4 0.309843 0.290461 0.294867 Channel 3 0.297363 0.307063 0.291520 0.304643

+]LUjLf/21 Power, MBTU/Hr 2977.36 2850.82 2935.45 2845.89 11,609.52 T-HOT, 'F 619.4 617.0 619.6 618.3 (99.748%)

T-COLD, 'T 561.3 560.2 561.9 560.7 NC Flow, MLBM/.h- 36,877 36.276 36.582 35.549 NC Flow, GPM 99,693 97,909 98,988 95,997 392,587 (101.97%)

Elbow Tap Coefficients:

Channel 1 0.300241 0.305159 0.289480 0.299010 Channel 2 0.295358 0.309647 0.290150 0.294616 Channel 3 0.297300 0.306920 0.291208 0.304429

'RUN #3:

Power, MBTU/Hr 2969.56 2847.16 2937.18 2843.87 11,597.77 T-HOT, 'F 619.2 616.9 619.5 618.2 (99.65%)

T-COLD, 'F 561.2 560.1 561.9 560.6 NC Flow, MLBM/Hr 36.861 36.283 36.638 35.556 NC Flow, GPM 99,645 97,922 99,135 96,006 392,708 (102.0%)

Elbow Tap Coefficients:

Channel 1 0.300012 0.305158 0.289750 0.298957 Channel 2 0.295180 0.309697 0.290420 0.294753 Channel 3 0.297121 0,306868 0.291477 0.304526

• AVERAGE:

NC Flow, GPM 99,684 97,929 99,072 96,017 392,702 (102.0%)

Elbow Tap Confficierts:

Chsnnel 1 0.300186 0.305206 0.289673 0.291029 Channel 2 0.295334 0.309729 0.290344 0.294,'5 Channel 3 0.297261 0.306950 0.291402 0.30i333 l

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