Text

Submit Report Contains Official Summary of Startup Physics Tests, Unit 4 Cycle Iii. the Tests Were Conducted in Accordance with Operating Procedure 0204.5, Startup Sequence After Refueling
	ML18227A290
Person / Time
Site:	Turkey Point
Issue date:	08/15/2018
From:	Kaminskas V; Florida Power & Light Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18227A290 (49)
	v • d • e

FLORIDA POWER 6 LIGHT Ce!fPANY TURKEY POINT PLANT UNIT 4 CYCLE III STARTUP REPORT

Acknowledgement I would like to acknowledge all the members of the Turkey Point Nuclear Operations Group for their support during the Unit 4 Cycle III Startup Tests.

I would also like to thank the members of the Power Resources Nuclear staff for their core data book and technical support given during the Startup Physics Tests.

Finally, I would like to acknowledge Jonell Collier for typing this report.

Introduction.

This report contains an official summary of the Startup Physics

Tests, Unit 4 Cycle III.

The tests were conducted in accordance with Operating Procedure 0204.5, Startup Sequence After Refueling.

The testing program commenced on June 6, 1976, with initial criticality and was completed June ll, 1976.

Completed by

~

4"~r~r4 ~A+

Vxto A.

Kaminskas'pproved by Robert E.

Dawson Reactor Supervisor

0

Acknovledgements Introduction 1.0 Unit 4 Cycle III Core 1.1 Loading Pattern 1.2 Rod Pattern 1.3 Rod Drop Times 2.0 Initial Criticality Table of Contents Page 2

3 4

2.1 ICRR Vs.

Rod Mithdrav 2.2 ICRR Vs. Dilution 7

8 3.0

.Summary of Tests 3.1 Nuclear Heating.

3.2 Reactivity Vs. Period 3.3 Boron Endpoints 3.4 Rod North (PQ1) 3.5 Rod llorth (PPH) 3.6 Temperature Coefficient 10 11 12 13 13 14

4. 0 Shutdo~im hfargin 4.1 Calculation 5.0 Power Distribution 1'taps 5.1 HZP Flux i'lap 5.2 60% Flux Ifap 17 20 21

1.0 Unit 4 Cycle III 1.1 Loading Pat tern 1.2 Rod Pattern 1.3 Rod Drop Times

El

Date 5-1/f-Pg Florida Power I3r Light Company REACTOR FUEL LOCATION TURKEY POINT PLANT UNIT NO.

90'4 13 12 11 10 5

4 S36 25 S26 83 S31 S40 R59 12P12 R51

~R51 L39 RS13 S13 14 S10 12P126 Sol R71 S14 15 S16 12 P40 21 R18 50 N28 R60 P46 N15 12P77 R72 R34 64 P33 93 S20 88 N

S35 95 S19 89 P35 32 P42 24 R02 38 R14 47 P22 R89 P05 R65 R21 10 R42 71 P02 R78 P38 R70 R26 56 R27 57 P16 R54

'13 RSO R45 90 R43 72 P36 R76 P20 18 R32 62 S23 87 P47 04 S18 23 L

S06 94 S39 R94 R47 75 S04 N05 12P10 R81 P09 R86 R44 73 ROS 79 P21 R77 P23 R30 RS18 60 R24 P32 54 07 N50 R57 R04 39 R29 59.

P18 19 P15 R36 RS11 80 R38 '31 67 R56 P08 R69 R13 46 R01 37 N13 R85 S28 R83 S29 S03 12P12 100 R31 61 L4l)

RS12 P37 P54 P52 RS8 R46 74 N17 R84 R33.

63

'LOS R59 R22 52 N12 R51 RI'0 69 P50 R61 P03 PS3 L20 R52 RS14

>R52 Ei 1 \\p S12 101 S38 N23 12P85 R73 R10 43 Pll R64 R37 66 P27 91 R06 41 P2608'20 51 P17 R79 R48 76 N19 R68 S02 12P13 S15 05 S05 R87 R15 48 P14 R75 R19 98 P25 R23

RS17, 53 N24 R67 Rll 44 P06 R07 RS16 81 P07 R62 R35 65 S37 R82 Sll 09 P30 01 S33 96 R25 55 P19 30 P43 R55 R49 77 R03 35 P10 R66 P12 R74 R12 45 R50 78 P39 R53 P01.

R52 R39 68 R17 26 P04 R92 P28 R91 R28 58 R05 40 P48 S6 P45 85 S24 20 S08 17 S30 27 P44 29 R09 42 418 R93 P29 NOS 12P92 R63 R16 49 P34 02 S21 11 S07 33 S09 S25 R90 12P99 L14 S34 RS15 12P90 S27 R88 S17 13

)3 S22 31 R41 70 270'32 16

Thimble Plugs I.I GEND Assembly No.

Insert No.

Coze IIl FonM sii45 pEY. tG/70

Control, Shutdown and Part Lergth Rod Locations RpNM I

KJ HG F EDOH C

Si DP Sg C

Df S)

Sg PL Sp PL IO II 12 IJ I

I Absorber ala ter i a 1:

Ag-In-Cd B

l5 Function Control Bank 0

~ Control Bank C

Control Bank B

Control Bank A

Shutdown Bank SB Shutdown Bank SA Part Length PL Control Rod Designation Huoibe.r of Clusters 8

8 8

8

0'l

ROD DROPS The following is a table of rod drops times as measured prior to Unit 4 Startup Physics Tests.

The two times, given are time to dashpot, I

a Technical Specification of 1.8 seconds or less, and time to the bottom of the core, which does not have a Technical Specification requirement.

All rods were dropped and met the Technical Specification requirement.

Florida Power 5 Light Company REACTOR FUEL LOCATION TURKEY POINT PLANT UNITNO.~

M L

K J

H G

I I

I D

C B

1.26 1.74 1.25 1.70 1.24 1.70 1 ~ 23 1.70 1.22 1.67 1.22 1.68 1.26 1.71 1.25 1.71 1.25 1.70

1. 27 1.73

1. 22 1.68 1.27 1.73 1.23 1.70 1.24 1.71 1,22 1.68 1.27 1.73 1.25 1.71 1.28 1.75 1.24 1.72 1.26 1.72 1.24 1.70 1.27 1.74 1.25 1.71 1.25 1 ~ 71 1.26 1.71 1.23 1.70 1.25 1.71 1.26 1.71

l. 24 1.72 1.25 1.70

l. 29 1.82 1.26

l. 74

10 1.25 1.70 1.26 1.70 1.25 1.72 1.25 1.71

11

l. 24 1.70 1.23 1.69 1.23 1.70

12 1.23 1.68 1.25 1.72

13 1.28 1.78 1.22 1.75

15 LEGEND Time to dash pot (sec)

Time to bottom or core (sec) r'OR!8 5045 RLV. ]a i70

2.0 Initial Criticality The approach to criticality began June 6,

1976 at 1459 hours0.0169 days 0.405 hours 0.00241 weeks 5.551495e-4 months in accordance with Operating Procedure 0204.3, Initial Criticality After Refueling.

Criticality was achieved June 6, 1976 at 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months by withdrawing control rods to 119 step on Bank D and diluting '25,300 gallons of water.

Upon attaining criticality the flux level was increased to 1 x 10 8 amps on the immediate range to obtain critical data.

Tavg = 547'F Control Bank = 119 steps Flux = 1 x 10 amps Boron

=. 10SO ppm The following two graphs are a plot of the ICPR during the approach to criticality.

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3. 0 Summary of Tes ts 3.1 Nuclear Heating 3.2 Reactivity Vs. Period Check 3.3 Boron Endpoints 3.4 Rod Morths (PCM) 3.5 Rod Uorths (PPM) 3.6 Temperature Coefficient

0 0

o

3.1 Nuclear Heating Nuclear Heating was determined in accordance with Operating Procedure 0204.3, Initial Criticality After Refueling.

Nuclear heating first occurred at:

-7 Kiethly Pico Ammeter 7.5 x 10 amps

-7 N-35

~ 7. 1 x 10 amps

-7 N-36

~ 7.3 x 10 amps

-8 All physics tests were conducted at a flux level at or below 9.5 x 10 amps to assure nuclear heating did not occur.

10

Cl

3.2 Reactivity Vs. Period i

Reactivity Computer Checkout was done in accordance with Operating Procedure 0204.3, Initial Criticality After Refueling.

The results are as follows:

Period (sec) 140. 8 162 Reactivity (pcm) 41 36,4 Reactivity (design)~

40.5

36. 5 Diff (%)

1 ~ 2

+0. 3 199.4 233.8 248.2 30.5 24.5 30 26.5 24.8

-1.7

+1.9

+1.2 259 24 24 0

285.1 22.5 22 2 ~ 3

PRX delayed neutron data used.

3.3 Boron Endpoints Measured Design Westinghouse ARO D in D,C in D,C,B in D,C,B,A in D,C,B,A,SBB in N-1 (F14) 1095 1032 894 832 524 539 604 589 504 552 492 1057 1043 993 977 862 834 768 756

+Not measured 12

ll

3.4 Rod Worths (pcm)

Rods Measured 673 Design 669 Wes tinghous e 699 1409 1356 1523 823 979 852 1733 1686 1826 SBB)

SBA 2490 7128 2474 7164 2710 7610 Diff(%)

+0.5%

+6. 0%

3.5 Rod Worth (ppm)

Rods D

+Measured 63 64 Design Wes tinghouse 138 131 143 62 94 78 A

SBB 308 264 ARI Less Most 556 Reactive Rod 505 551

~Measurement uncertainty on PPM Boron is an average of + 18 ppm 13

Temperature Coefficient Temperature coefficient was measured in accordance with Operating Procedure 1604.6, Nuclear Design Check Tests.

The measured numbers are:

Isothermal Temperature Coefficient Rods DIN~*

Heasured

-1.2 pcm/'F

-8.1 pcm/'F Design (PRN)

Not Gale Not Calc Westinghouse

-2.3 pcm/'F

-5.0 pcm/'F

>Hoderator Temperature Coefficient Rods DIN:>*

Heasured

+0.6 pcm/'F

-6.3 pcm/'F Design (PRN)

-4.5 pcm/'F Not Gale Mes tinghouse

-0.5 pcm/'F

-3.2 pcm/'F

>Hoderator Temperature Coefficient can be defined as Isothermal Temperature Coefficient minus Doppler Coefficient.

Doppler Coefficient

= -1.8 pcm/'F.

>+Bank C = 214 steps 14

From the data taken the ARO case shows a positive moderator temperature coefficient. of +0.6 pcm/'F.

Technical Specification 3.1.2.6 states "the reactor shall not be made critical unless the moderator temperature coefficient is zero or negative

. except for low power physics tests".

The point at which the moderator temperature coefficient is zero is defined as a boron concentration equal to or less than 1085 ppm or control bank D equal to or less than 194 steps.

By assming an error of -1 pcm/'F the limit for a zero or negative moderator temperature coefficient is defined as a boron concentration equal to or less than 1076 ppm or control bank D equal to or less than 160 steps.

These limits assure a zero or negative moderator temperature coefficient for BOL operations.

15

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6/9/76 Unit 4 C cle III 1033 10 38 104'!4 49

~1055 0:1065:.

107 107 108 I

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10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 Rank D

0

4.0 Shutdown Hargin Shutdown margin was calculated prior to power escallation to verify adequate shutdown capability.

For this calculation design, numbers were used because of the excellent agreement with measured number.

The following is a summary of the results:

4.1 BOL (pcm)

EOL (pcm)

Rod !forth Stuck Rod Doppler Defect Hoderator Defect 4324 3765 10%

Rod Uncertainty 538 510 Rod Insertion Flux Redistribution Shutdown Hargin 500 500 2786 500 850 1905 Required by Technical Specifications 1000 1770

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CBD P/L 169 228

~ax P 1.829

ML18227A290

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