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 Table of Contents Page Acknovledgements Introduction 1.0 Unit 4 Cycle III Core 1.1 Loading Pattern 2 1.2 Rod Pattern 3 1.3 Rod Drop Times 4 2.0 Initial Criticality 2.1 ICRR Vs. Rod Mithdrav 7 2.2 ICRR Vs. Dilution 8 3.0 .Summary of Tests 3.1 Nuclear Heating. 10 3.2 Reactivity Vs. Period 11 3.3 Boron Endpoints 12 3.4 Rod North (PQ1) 13 3.5 Rod llorth (PPH) 13 3.6 Temperature Coefficient 14

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

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

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

90'4 13 12 11 10 5 4 S26 R51 S13 83 ~R51 14 S36 S31 S40 L39 S10 Sol S14 25 R59 12P12 RS13 12P126 R71 15 S16 P40 R18 N28 P46 N15 R34 P33 S20 N 12 21 50 R60 12P77 R72 64 93 88 S19 P42 R14 P05 R42 P38 R27 R43 P20 S23 89 24 47 R65 71 R70 57 '13 RSO 72 18 87 S35 P35 R02 P22 R21 P02 R26 P16 R45 P36 R32 P47 S18 L 95 32 38 R89 10 R78 56 R54 90 R76 62 04 23 S39 R47 P09 ROS P23 R30 N50 R29 P15 R36 P08 R01 S28 R94 75 R86 79 RS18 60 R57 59. RS11 80 R69 37 R83 S06 94 S04 12P10 N05 R44 P21 R24 P32 R04 P18 R38 '31 R13 N13 S29 S03 R81 73 R77 54 07 39 19 67 R56 46 R85 12P12 100 R31 L4l) P37 P52 R46 N17 R33. 'LOS R22 N12 RI'0 P50 P03 L20 R52 Ei 61 RS12 P54 RS8 74 R84 63 R59 52 R51 69 R61 PS3 RS14 >R52 1 \p S12 101 S38 12P85 N23 R73 R10 43 P ll R64 R37 66 P27 91 R06 41 08'2051 P26 P17 R79 R48 76 N19 R68 S02 12P13 S15 05 S05 R15 P14 R19 P25 R23 N24 Rll P06 R07 P07 R35 S37 R87 48 R75 98 RS17, 53 R67 44 RS16 81 R62 65 R82 Sll P30 R25 P43 R03 P12 R50 P01. R17 P28 R05 P45 S08 09 01 55 R55 35 R74 78 R52 26 R91 40 85 17 S33 P19 R49 P10 R12 P39 R39 P04 R28 P48 S24 96 30 77 R66 45 R53 68 R92 58 S6 20 S30 P44 R09 418 P29 NOS R16 P34 S21 27 29 42 R93 12P92 R63 49 02 11 S07 S09 S25 L14 S34 S27 S17 )3 33 R90 12P99 RS15 12P90 R88 13 S22 R41 31 70 16

Thimble Plugs 270'32 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 Si C DP Sg C Df PL S)

Sg PL IO Sp II 12 I

I IJ Absorber ala ter i a 1:

Ag-In-Cd B

l5 Control Rod Designation Function Huoibe.r of Clusters

. ~

Control Bank Control Bank 0

C 8 Control Bank B 8 Control Bank A 8 Shutdown Bank SB 8 Shutdown Bank SA 8 Part Length PL 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 UNIT NO.~

M L K J H G D C B I I I 1.26 1.25 1.74 1.70 1.23 1.24 1.70 1.72

1. 22 1,22 1.24 1.68 1.68 1.70 1.25 1.24 1.26 1.25 1.71 1.71 1.72 1.71 1.22 1.27 1.23 l. 29 1.67 1.73 1.70 1.82 1.24 1.70 1.25 1.25 l. 24 1.70 1 71

~ 1.72 1.22 1.27 1.25 1.27 1.25 1.68 1.73 1.71 1.74 1.71 1 23

~ 1. 27 1.26 1.25 1.70 1.73 1.71 1.70 1.26 1.71 1.28 1.75 1.26 1.71 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 1.22 1.78 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

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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 N-35 ~

-7

7. 1 x 10 amps N-36 ~

-7 7.3 x 10 amps All physics tests -8 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) Reactivity (pcm) Reactivity (design)~ Diff (%)

140. 8 41 40.5 1~2 162 36,4 36. 5 +0. 3 199.4 30.5 30 -1.7 233.8 26.5 +1.9 248.2 24.5 24.8 +1.2 259 24 24 0 285.1 22.5 22 2~3

PRX delayed neutron data used.

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

+Not measured 12

ll 3.4 Rod Worths (pcm)

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

SBA 2490 2474 2710 7128 7164 7610 Diff(%) +0.5% +6. 0%

3.5 Rod Worth (ppm)

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

308 264 SBB ARI Less Most 556 505 551 Reactive Rod

~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 Heasured Design (PRN) Westinghouse

-1.2 pcm/'F Not Gale -2.3 pcm/'F DIN~* -8.1 pcm/'F Not Calc -5.0 pcm/'F

>Hoderator Temperature Coefficient Rods Heasured Design (PRN) Mes tinghouse

+0.6 pcm/'F -4.5 pcm/'F -0.5 pcm/'F DIN:>* -6.3 pcm/'F Not Gale -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.

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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 EOL (pcm) (pcm)

Rod !forth Stuck Rod 4324 3765 Doppler Defect Hoderator Defect 10% Rod Uncertainty 538 510 Rod Insertion 500 500 Flux Redistribution 500 850 Shutdown Hargin 2786 1905 Required by Technical Specifications 1000 1770

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