ML17352A206

Cycle Xiv Startup Rept. W/930820 Ltr

ML17352A206
Person / Time
Site:	Turkey Point
Issue date:	08/20/1993
From:	Plunkett T FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-93-196, NUDOCS 9308300191
Download: ML17352A206 (40)

Text

ACCEI Etui.T~D DOCUMENT DIST I UTION SYSTEM REGULA ~..Y INFORMATION DISTRIBUTIO ,SYSTEM (RIDS)

ACCESSION NBR:930830019l DOC.DATE: 93/08/20 NOTARiZED: NO DOCKET ¹ FACIL:50-251 Turkey Point Plant, Unit 4, Florida Power and Light C 05000251 AUTH. NAME AUTHOR AFFILIATION PLUNKETT,T.F. Florida Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

"Turkey Point Nucl'ear Plant Unit 4,Cycle XIV Startup Rept."

ltr.

W/930820 DISTRIBUTION CODE: IE26D COPIES RECEIVED LTR Startup Report/Refueling Report (per Tech Specs)

J ENCL I SIZE:

I'ITLE:

NOTES:

RECIPIENT COPIES RECIPIENT COPIES lD CODE/NAME LTTR ENCL, ID CODE/NAME LTTR ENCL PD2-2 PD, 1 1 RAGHAVAN,L 2 2 ZNTERNAL: AEOD/DSP/TPAB 1 1 NRR/SR%3 1 1 NUDOCS-ABSTRACT 1 1 EG IL 02 1 1 RGN2 FILE 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS" FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 10 ENCL 10

ik gF

L-93-196 10 CFR 50.36 U. S ~ ..Nuclear Regulatory Commission Attn: Document Control Desk

.Washington, D.C. 20555, Gentlemen:

Re: Turkey Point Unit 4 Docket No. 50-251-Startu 'Re ort In accordance with Technical Specification 6.9.1.1, the enclosed Startup Report is provided for Flori'da Power and Light Company Turkey Point Unit 4. The Unit 4 Cycle XIV Startup Report documents the first use of axial (natural uranium) blankets and snag-resistant spacer grids at the top and'ottom of the fuel assemblies.

If you have any questions, please contact us.

Very, truly yours, T. F. Plunkett Vice President Turkey Point Nuclear TFP/RJT/rt

."Attachment cc: S. D. Ebneter, Regional Administrator, Region II USNRC Senior Resident Inspector, USNRC, Turkey Point Nuclear 9308300191 930820 PDR ADO'500025i PDR an FPL Group company

ATTACHMENT FLORIDA POWER 6 LIGHT COMPANY TUEQCEY POINT NUCLEAR PLANT UNIT O'YCLE XIV STARTUP REPORT

L-93-196 Attachment Page 2 of 18 INTRODUCTION This report contains the official summary of the Startup of Physics Tests performed on Turkey Point Unit 4 at the beginningwith Cycle XIV. The testing program was conducted in accordance of Turkey Point Plant Procedures, and meets the requirements for ANSI/ANS 19.6.1, Revision 0 .(12/13/85), "Startup Physics Tests Pressurized Water Reactors".

Withdrawal of Shutdown banks commenced May 23, '1992 at 0242 and initial criticality was achieved 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 24 minutes later.

WCAP-13682, "The Nuclear Design and Core Management of the Turkey Point Unit 4 Nuclear Power Plant, Cycle 14", was the design source for verifying that acceptance criteria as specified in ANSI/ANS 19.6.1 were met. All tests performed for nuclear design verification meet their acceptance criteria.

The contents of this report provide the documentation required by Technical Specification 6.9.1.1.

H

L-93-196 Attachment Page, 3 of 1'8 TABLE OF CONTENTS INTRODUCTION 1'.0, UNIT 4, CYCLE XIV CORE 1.1 Fuel Design Changes 1' Loading Pattern 1.3 Rod Pattern and Rod Drop Times 2.0 INITIAL CRITICALITY 2.1 Inverse Count Rate Ratio (ICRR) vs. Dilution

.2.'2 Critical Data 3.,0 SUMMPGtY OF 'TESTS 3.1 Nuclear Heating 3.'2 Reactivity vs. Period 3' Boron Endpoints 3.,4 Rod Worth (ppm), Most Reactive Bank 3.'5 Rod Worth (pcm) 3.6 Temperature Coefficient 3.7 Hot Zero Power (HZP) Differential Boron Worth 4.0 SHUTDOWN. MARGIN 5.0 POWER DISTRIBUTION MAPS 6.0 'CRITICAL BORON CONCENTRATION

0 L-93-196 Attachment Page 4 of 18 1.0 UNIT 4 CYCLE XIV CORE'.1 Fuel Desi n Chan es Unit 4 Cycle, 14 fuel is essentially the same as Cycle 13 fuel with the exception that Cycle 14 fuel includes axial blankets and additional snag-resistant grids.

Axial blankets, previously used in Turkey Point Unit 3 Cycle 13 core design are new to Unit 4. Axial blankets consist of a nominal 6 inches of natural VO~ pellets at the top and bottom of the fuel pellet stack. Axial blankets are designed to reduce neutron leakage and therefor improve uranium utilization.

Anti-snag mid-grids were included in the Unit 4 Cycle 13 design. The Unit 4 Cycle 14 design adds top and bottom anti-snag grids to the fuel assembly design. This addition wi;11 reduce the possibility of assembly damage during fuel handling.

1.2 Loadin Pattern This section presents the as-loaded core configuration (Figure 1, page 5).

1'.3 Rod Pattern and Rod Dro Times This section presents the Control and Shutdown Rod pattern and the Rod Drop Times for all rods as measured per Procedure 4-PMI-028..3, "RPI, Hot Calibration, CRDM Stepping Test, and Rod Drop Test" (Figure 2, page 6).

All rods. meet the drop time limit of 2.4 seconds as per Technical Specification 3.1.3.4.

0 4l L-93-196 Attachment Page 5 af 18 FZGURE 1 TURKEY POZNT UNZT 4 CYCLE 14 CORE LOADZNG A

I I

NORtH I

I RR23 RR30 RR15 HF23 HF16 HF06 PP26 SS35 TT38 RR49 T740 SS33 PP55 R52 R54 RR19 TT46 TT19 SS20 TT22 SS18 TT24 TT48 RR07 4M R53 4'M R51 4M RR04 SS48 SS41 RR46 TT03 RR27 7706 RR47 SS29 SS47 RRO6 R57 16M R56 16IJ R55 PP40 TT49 SS40 SS14 TTOS RR39 TT30 RR33 TT14 SS11 SS39 TT50 PP33 R61 16M R59 8M R60 16M R58 SS38 TT31 RR41 T715 RR11 SS01 SS17 SS07 RR10 T702 RR51 TT18 SS37 R66 4M R65 16M R64 16M R63 4M R62 RROS TT41 SS24 TT16 RR36 SS10 TT33 SS03 T734 SS09 RR34 TT04 SS25 TT42 RR20 HF07 R70 16IJ R65 SM SM R69 16M R67 HF13 RR29 RR50 TT32 TT16 . TT20 SS27 SS16 RR25 SS12 SS19 TT27 RR24 TT28 RR48 RR32 HF15 4M 16M 8M R74 R73 R72 SM R71 4M HF05 RR09 TT43 SS21 T711 RR35 TT35 SS13 TT36 SS15 RR40 TT12 SS26 TT37 RR14 HF20 R77 16M R78 SM 8M RSO 16M R76 HF02 SS30 T725 RR43 TT09 RR26 SSOS SS28 SS06 RR01 TT10 RR52 TT26 SS43 R84 4M R53 16M R82 16M R81 4M R79 PP51 TT51 SS32 SS04 T705 RR37 TT21 RR38 TT07 SS05 SS44 TT45 PP45 R89 16IJ R86 SM R87 1QJ R85 RR12 8 $ 46 SS34 RR42 TT01 RR28 T713 RR44 SS31 SS45 RR03 R90 16M R91 16M RSS RR13 TT52 T723 S$ 23 TT17 SS22 TT29 TT47 RR05 4M R92 4M R95 4M SS42 TT44 RR45 TT39 SS36 PP54 R101 R93 RR22 RR31 RR21 HF11 HF10 HF01 key: ASSY INS.

PPxx Rxx ASSY RRxx zzM SSxx HFxx TTxx ....

PP Reload Cvcle 11 RR Reload Cycle 12 SS Reload Cycle 13 TT Feed CycLe',14 R Control Rod M MASA insert HF Hafniisa inserts xx Sequence amber zz Hwher of MASA fingers

4I 0 L-93-196 Attachment Page 6 of 18 FIGURE 2 TURKEY POINT UNIT 4 CYCLE 14 RCCA BANK PATTERN AND DROP TZMES A

I I

NORTH I

I CB-8 CS-S 1.37 1.35 SB-A SB A 1.36 1.35 CB-C CB 0 CB C 1.35 1.34 1.32 SB-B CB A CB A SB-S

'.33 1.37 1.31 1.32 CB.B CB-C SB 8 CB.C CS ~ 8 1.39 1.32 1.37 1.33 1.34 SB-A CB A CB A SS.A 1.34 1.32 1.33 1.33 CB-D SS.B 'CBD SB 8 CB-D 1.35 1.37 1.35 1.35 1.34 SB A CB-A CB-A SB A 1.34 1.35 1.35 1.35 CS 8 CB.C SB-B CB-C CB 8 1.37 1.34 1.35 1.32 1.35 SS.B CB.A CS-A SS-S 1.35 1.34 1.32 1.30 CS C CB.D CB C 1.34 1.35 1.34 SS A SB A 1.34 1.36 CB 8 CB-S 1.37 1.40 RCCA TINE keys SB-x sec.

RCCA CS-x SB Shutdown Bank CB Control Bank x Bank Identifier sec. Drop Time toDashpot

li L-93-196 Attachment Page 7 of 18

2. 0 XNITIAL CRITICALITY

2. 1 INVERSE COUNT RATE RATXO (XCCR) vs DILUTXON The approach to criticality began May 23, 1993 at approximately 0242 when the stepping of shutdown banks began in accordance with Procedure 0-OSP-040. 6, "Initial Crit'icality After Refueling.'" Criticality .was achieved approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 24 minutes later on May .23, 1993 at 0906 by diluting 14,460 gallons of water with control bank D at 180 steps.

Figure 3 (page 8) is a plot of the ICRR during the approach to criticality.

2.2 CRITICAL DATA Upon attaining criticality, the flux level was increased'o 1 x 10 amps on the reactivity computer to obtain critical data, as follows:

Tavg = 547.1'F Control Bank D = 178 Steps Reactor Coolant System (RCS) Boron = 1689 ppm Picoammeter Flux = 1 x 10 '

N35 Flux 1.1 x 10 A N36 Flux 1.8 x 10 '

4k Page 8 FIGURE 3 Turkey Point Unit 4 (1/M) on Approach to Criticality (1/M) DURING ROD WITHDRAWAL 1.3 N31 1.2 1'. 1 0 0 0 0 0.9 N32 0 0 0 0 0.8 0.7 T 'I 50 100 150 228 SHUTDOWN BANK A 50 100 150 SHUTDOWN BANK B 228 I

50 CBAH 100 150 228 CBB I

22 100 150 CB-C I I 228 H

CBD~

(1/M) DURING DILUTION 1 0 0 0.9 p 0 0

0.8 p 0 0 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0

5000 10000 15000 20000 Gallons Water Added

L-93-196 Attachment Page 9 of 18

3. 0

SUMMARY

OP TESTS This section provides a summary of the results of the low power physics tests for Unit 4, Cycle XIV along with the Nestinghouse design data. For each test, the acceptance criteria is listed at the bottom of the table. This report compares design and measured data using Difference and Percent Difference.

Difference = Predicted Measured For calculating Percent Difference, the equation is:

~

PredfctedValue 11&00 2tfeaauredVal ue 3.1 Nuclear He tin The point of adding Nuclear Heat was determined in accordance with Procedure 0-OSP-040. "Initial Criticality After Refueling". This is 6,performed by establishing a small positive startup rate and measuring the flux level at which T, departs from its established steady state value. Nuclear Heating was measured to first occur at values presented on Table 3.1.1.

T?LB' ~ 1 1C FLUX LEVEL (AMPH)

Pico Oter N-35 N-36 1 S x 10 2.1 x 10 3.4 x 10 8 All physics tests were conducted at or below 1 x 10 " amps on the picoammeter connected to N-44 to assure Nuclear Heating did not occur.

~l

L-93-196 Attachment Page 10 of 18 3.2 Reactivit vs. Period Reactivity Computer checkout was done in accordance with Procedure . 0-0SP-040.6, "Initial Criticality After Refueling." This checkout is performed by inserting small positive and negative reactivities using rod motion. The period of the flux change is used to calculate the design reactivity. The measured reactivity is taken directly from the reactivity computer. The results of this test are given in Table 3.2.1.

TABLE 3.2.1: MEASURED REACTIVITY VS. DESIGN MEASURED DESIGN PERIOD REACTIVITY REACTIVITY (SEC) (PCM) (P CM) DIFF*

+151.2 +39.0 +38.8 +.5

-249.6 -33.'0 33 7

~ 2 1

~

+78.3 +65. 0 +65.4 .6

+129.1 +44.0 +44.2 5

• Acceptance Criteria is 4% for positive period.

3.3 Boron End ints The Boron Endpoint measurement is a way, of measuring the steady state boron concentration of an under-rodded core (positive period in effect) or an over-rodded core (negat'ive period'n effect). In FPL's testing program the first case is an unrodded core and the second case is a core with the reference bank at the bottom. The Boron Endpoint is measured using Procedure 0-OSP-040.5, "Nuclear Design Verification." In this methodology a just-critical condition is established as near as practical to the required rod configuration. The rods are then moved into the desired configuration and back to equilibrium. The RCS boron concentration which was measured at equilibrium is then adjusted for the ppm worth of the rods. The results of the two boron endpoint measurements are given in Table 3.3.1.

TABLE 3.3. 1: BORON ENDPOINTS (ppm)

MEASURED MESTINGHOUSE DIFFERENCE*

(ppm) (ppm) (ppm) 1698 1693 SB-B 1552 1547

• Acceptance Criteria is +/- 50 ppm.

41 '

L-93-196 Attachment Page 11 of 18 3.4 Rod'orth Most Reactive Bank Rod Worth was measured per Procedure O-OSP-040.5, "Nuclear Design Verification." The reference bank (highest predicted worth) was first inserted dilution as the controlling bank was .withdrawn. Then a was used to adjust, the reference bank to approximately 30 steps from the bottom. Finally a Boron Endpoint (see section 3.3) was performed. By graphing the rod worth, measured by the reactivity computer, versus rod insertion, a differential rod worth curve is generated.. Bycurve summing is the differential worth an integral rod worth generated and the bank worth determined. The total bank worth is presented in Table 3.4.1. The Integral and differential bank worth of the reference bank is displayed in Figure 4 (page 12).

TABLE 3.4.1: ROD,WORTH (pcm)

MEASURED WESTINGHOUSE DIFF*

(pcm) (pcm) (+)

SB-B 1173 1189 1.4

• Acceptance Criteria is less than 10%

3.5 Rod Worth cm Remainin Banks R

The remaining RCCA bank worth was measured per Procedure 0-OSP-040.5, "Nuclear Design Verification," using, the rod swap technicgxe. This technique involves swapping the negative reactivity of the bank being inserted with the positive reactivity from the bank being withdrawn. Each bank is sequentially swapped for the reference bank. The worth of each bank can then be determined from the integral rod worth curve. The results of this measurement are given in Table 3.5.1.

TABLE 3.5.1: ROD WORTH (pcm)

MEASURED WESTINGHOUSE 0 DIFF*

(pcm) (pcm) ()

SB-A 1052.3 1098 4.3 CB-A 1085.6 1104 1.7 CB-B 435.3 484'163

.11.2 CB-C 1093.0 6.4 CB-D 635.9 664 4.4 TOTAL 5475.1 5702 4.1 NOTE: The total rod worth includes the reference bank.

• The acceptance criteria for rod worth measurements are:

Individual banks within +/- 15% or +/- 100 pcm of design, whichever is greater and Total of all measured banks within

+/- 10% of design.

I FIGURE 4 Page 12 Differential and Integral Reference Bank Worth 12 1400 INTEGRAL ORTH Q DIFFE EMTIALWORTH

6) 1 200

~10 (0

1000 g)

~8 Q

U3 Q

0 800 ~

Ne 600 ~~

C6 Q3 4

U 0$ O S~ 400 2 0) 200 Cl 0

50 100 150 200 228 Rod Position (steps withdrawn)

L-93-196 Attachment, Page 13 of 18 3.6 Tem erature Coefficient The isothermal and moderator temperature coefficients were determined using Procedure 0-OSP-040.5, "Nuclear Design Verification." The isothermal temperature is measured by varying the moderator temperature below the point of adding nuclear heat. The reactivity change is then simply divided by the temperature change to obtain the isothermal temperature coef ficient. The moderator temperature coefficient is calculated from the isothermal temperature coefficient by subtracting the doppler coefficient. The values determined for this testing sequence are presented on Tables 3.6.1 and 3.6.2 below:

TABLE 3.6.1: ISOTHERMAL TEMPERATURE COEFFICIENT (pcm/ F)

MEASURED WESTINGHOUSE DIFF*

(pc /F) (pcm/ F) (pcm/ F)

-1.44 .43 1.01

• Acceptance Criteria is +/- 2 pcm/ F of design.

TABLE 3. 6. 2: MODERATOR TEMPERATURE COEFFICIENT (pcm/ F)

MEASURED* WESTINGHOUSE (pcm/ F) (pcm/ F)

.26 1.27

• Acceptance Criteria is < 5 pcm/'F.

3.7 HZP Differential Boron Worth The Hot Zero Power (HZP) Differential Boron Worth was measured using Procedure O-OSP-040.5, "Nuclear Design Verification." The worth of the reference bank is divided by the boron change from ARO to the reference bank fully inserted. The value obtained for this test is presented on Table 3.7.1.

TABLE 3.7.1: HZP DIFFERENTIAL BORON WORTH (pcm/ppm)

MEASURED WESTINGHOUSE DIFF*

(pcm/pr ) (pcm/pe ) (pcm/pz )

8.56 8.32 2.8

• Acceptance Criteria < +/- 15%.

4i L-93-196 Attachment Page 14 of 18 4.0 SHUTDOWN MARGIN The Shutdown Margin vas calculated prior to power escalation to verify adequate shutdown capability. For this calculation, the total of the design rod worth (minus. the most reactive stuck rod) vere reduced by 7%. The results show adequate shutdown margin at Beginning of Life (BOL) and End of Life (EOL). The following is a summary of the data used:

TABLE 4.1: UNIT 4, CYCLE XIV SHUTDOWN DATA BOL EOL HKP Control Rod North Re irement All Rods Inserted Less Most Reactive Stuck Rod 6.28 6.72 (1) Less 7% 5.84 6.25 Hot Full Power HFP to HKP Reactivit Insertion Reactivity Defects (Doppler, T,~,

Void, Redistribution) 1.72 2.71 Rod Insertion Allowance 0.50 0.50 (2) Total Requirements 2.22 3.21 Shutdown Margin (1) (2) '(%a@) 3.62 3.04 Required Shutdown Margin (%wp) 1.00 1.77

• Source: WCAP 13682 5.0 POWER DISTRIBUTION MAPS The core,was mapped using incore instrumentation for power levels of 30%, 50% and 100%. A summary of the results are presented- on pages 15 through 17.

0 FIGURE 5 Page 15 SVNKARY OF 30X POUER FLUXHAP HEASURED ASSEHBLY POMER AND PERCENT DIFF. TO EXPECTED PCMER INSTR. LOC. ONLY 15 14 '13 12 11 10 9 8 7 6 5 4 3 2 1 0.222. 0.232. 0.223.

-1.8.

0.349. 0.824. 1 ~ 115. 0.774. 1.103. 0.808. 0.345.

0.393., 1.110 ~ 1.328. 1.262. 1.315. 1.229. 1.298. 1.100. 0.366.

3.7. 1.4. -0.7. .2.7. 0.9.

0.386. 0.941. 1.272. '1.076. 1.286. 0.900. 1.284. 1. 103. 1.304.'0.933. 0.348.

-8.0.

~ ~ ~

0.341. 1.083. 1.198. 1.196. 1.250. 1.167. 1'.331. 1.202. 1.339. 1.292. 1 '94. 0.999. 0.313.

-0.3. .6.5. 1.0. 4.0. 3.3. 1.0.

0.817. 1.337. 1'.084. 1.292. 0.988. 1.223. 1.255. 1.235. 0.996. 1.273. 1'.056. 1.261. 0.786.

0.228. 1.109.

0.5.

1' 4.5..

267. 1.330. 1.207. 1.222 ~ 1.252 2.1 ~

'.178. 1.253. 1.169. 1.066.

1.9. . -7.7.

1 ~ 208. 1.218. 1.110. 0.227.

-3. 1 ~

~ ~ ~

0.244: O.798. 1.274. 0.908. 1.352. 1.275. 1.184. 0.820. 1.181. 1.223. .1.211. 0.843. 1.326. 0.785. 0.238.

~ 3.8. 5.7. 0.2. -5.5. 0.1. 0.7.

0.240. 1.195. 1.294. 1.278. 1.164. 1.209. 1.257. 1.194. 1.285. 1.199. 1.101. 1.199. 1.217. 1. 140. 0.231.

G ~ 6.8. 4.5; 0.865. 1.384. 1.122. 1.238. 1.016.. 1.224. 1.261. 1.208. 0.963. 1.243 '.996. 1.266. 0.835.

5.7. -4.1. 2.3. 2.9. . -3.6. -9.0. 2.2.

0.361. 1. 162. 1.337. 1.368. 1.299.. 1.163. 1.310. 1.151. 1.285. 1.294. 1.246. 1.045. 0.338.

7.0. 1.2.

0.411. 1.035. 1.331. 1.051. 1.213. 0.837. 1.206. 1.109. 1.304. 0.947. 0.361.

8.5. -4.1. 1.9. -4.9.

0.411. 1.087. 1.255. 1.227. 1.258. 1.275. 1.346. 1.097. 0.376.

.5.0.

0.327. 0. 779. 1. 125. 0.809. 1.191. 0.843. D.344.

.4.6. 2.7. 6.4. 0.9.

0.243. 0.249. 0.236.

6.9.

POMER TILT IN POUER TILT IN "CORE AVERAGE UPPER HALF OF CORE LOUR HALF OF CORE AXIAL OFFSET

(-,+) ~ (+,+) (-,+) . (+,+)

0.9974 . 0.9665 1.0087 . 1.0129 0. 208

~'

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.0486 . 0.9875 0.98'l7 . 0.9966

(- -) ~ (+ -) (- -) (+ )

TOP TEN NUCLEAR F DELTA H 245 G14NI FDHN*1.5422 309 B 7IC FOHN>1.5359 260 F13LC FOHN"-1.5285 274 E12LF FOHN<1.5262 273 E13HB FDHK>1.5020 287 D11JD FDHN=1.4936 185 N 5FL FOHN>1.4909 293 D SFD FDHHi1.4907 233 H11HL FOHNa1'.4859 198 L 4DJ FDHHa1.4787 LINITING 3 FO ELEVATIOHS ( TOP 15X, BOTTOK 15X, AND'NIDDLE 7OX OF THE CORE )

AXIAL FO(Z) HEAS. PERCENT SOURCE.

POINT LIHIT FO(Z) TO.LIN. NO. ID 28 4.6052 2.2654 50'81 402 G14XX 1D 4.3964 1.7459 60.29 417 F13XX 52 4.6400 1.6294 64.88 322 P 7XX

FIGURE 6 Page 16 QPQQRY OF 50X POWER FLUXHAP MEASURED ASSEHBLY POWER AND PERCEN'I DIFF, TO EXPECTED POWER INSTR. LOC. ONLY 15 14 13 12 11 10 9 8 7 6 5 4 3

. 0.225. 0.236. 0.226.

-0.2.

0.342. 0.801. 1.111. 0.774. 1.103. 0.773. 0.324.

~ ~

0.386. 086. 1 '89. 1.245. 1.318. 1.195. 1.242 1.031. 0.379.

-0.4.

2.0. -1.6. -5.3.

. 0.382. 0.962. 1.282. 1.080. 1.253. 0.870. 1.253. 1.075. 1.264 '.937. 0.378.

.0.1 ~

0.340. 1.081. 1.280. 1 277. 1.277. 1.136. 1.264. 1.158. 1.308. 1.261. 1.236. 1.065. 0.340.

-0.6.

~

-0.2. .1.7. .1.3. 0.8..1.5. -3.5.

0.811. 1.319. 1.105. 1.307. 0.991 ~ 1.177. 1.214. 1.197. 0.995. 1.272. 1-067. 1.262. 0 810. ~

. 0.226. 1.133. 1.278. 1 '96. 1.182. 1.221. 1.206. 1.157. 1.232. 1.182. 1.136. 1.249. 1.240. 1.112. 0.225.

J 1.8. 0.0. -1.7.

~ 0.239. 0.794. 1.332. 0.902. 1.304. 1.264. 1.206. 0.815. 1.164. 1.210. 1.259. 0.880 ~ 1.305. 0'. 785. 0.236.

H 0.7. 1.8. -1.3. -1.3. 0.0.

. 0.231. 1.143. 1.276. 1.290. 1 178. 1.228 1.283. 1.194. 1.269. 1.211 ~ 1.147. 1.243. 1.234. 1. 113. 0.225.

G ~ 4.2. 3.0.

~ ~

0.832. 1.339. 1.120. 1.323. 1.031. 1.226. 1.255. 1.221. 1.015. 1.283. 1.060. 1.277. 0. 799.

2.3. 2.4. 1.5. .3.3. -2.1.

0.349. 1.120. 1.319. 1.313. 1 '21. 1.179. 1.299. 1.170. 1.296. 1.264. 1.253. 1.060. 0.335.

2.7.

0.393. 0.990. 1.314. 1 ~ 106. 1.288. 0.899. 1 ~ 280. 1.091. 1.266. 0.942. 0.375.

3.8. 1 ~ 0. 0.9. -1.2.

0.394. 1.103. 1.312 ~ 1.269. 1.345. 1.279. 1.312. 1.073. 0.376.

1.7.

0.339. 0.809. 1.150. 0.809. 1.145 ~ 0.821. 0.339.

-0.9. 2.4. -0 '.

0.240. 0.246. 0.231.

5.8.

POWER TILT IN POWER TILT IN CORE AVERAGE UPPER HALF OF CORE LOWER HALF OF CORE AXIAL OFFSET

(-,+) . (+,+) (-,+) . (+,+)

1.0010 . 0.9787 0.9978 . 0.9865 -0.272

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.0284 . 0.9920 1.0128 . 1.0028

(-, ) . (+i ) ( ,-) . (+,-)

TOP TEN NUCLEAR F-DELTA-H 274 E12LF FOHN%1.5074 287 D11JD FDHK$ 1.5016 307 8 9GC FDHK$ 1.4814 273 E13NB FDHN$ 1.4814 260 F13LC FOHN%1.4800 309 8 7IC FDHN*1.4773 245 G14NI FOHN%1.4747 215 J14HG FDHN%1.4671 179 H11JL FDHN$ 1.4653 297 C11NC FOHN%1.4650 LINITING 3 FQ ELEVATIONS ( TOP 15X, BOTTGH 15X, AND HIDDLE 70X OF THE CORE )

AXIAL FQ(Z) KEAS. PERCEHT SOURCE POINT L INIT FQ(Z) TO LIN. NO. ID 27 4.5936 '.0692 54.95 372 J14XX 52 4.6400 1.7941 61.33 384 J 2XX 10 4.3964 1.5974 63.66 431 E12XX

FIGURE 7 Page 17 SUHHARY OF 100X FLUXNAP HEASURED ASSEHBL'Y POWER AND PERCENT DIFF. TO EXPECTED POWER - INSTR. LOC. OWLY 15 - 14 13 12 11 10 9 8 7 6 5 4 3

. 0.269. 0.286. 0.270.

13.9.

0.359. 0.742. 1.119. 0.860. 1.184. 0.919. 0.380.

0.441. 1.081. 1.162. 1.207. 1.351. 1.269. 1.148. 1.146. 0.425.

13.6. -7.4 ~ 2.8.

0.410. 0.970. 1.232. 1.003. 1.225. 0.986. 1.221. 0.987. 1.174. 0.976. 0.424.

9.3.

0.340. 1.026. 1.145. 1.151. 1. 174 ~ 1.095. 1.224. 1.087. 1.162. 1.226. 1.218. 1.090. 0.379.

-2.4. -7.4. -6.8. -8.7. -1.4.

0.864. 1.226. 1.001. 1.243. 1.002. 1-206- 1.240. 1.183. 0.995. 1.216. 1.059. 1.235. 0.871.

. 0.281. 1.203. 1.255 '.190. 1.156. 1.270. 1.324. 1.275. 1.323. 1.218.

3.1. 3.0.

1 ~ 150. 1.259. 1.236. 1. 163. 0.268.

-2.2 J . 18.4 ~ ~

~ ~ ~ 0 ~ ~

. 0.281. 0.857. 1.335. 0.982. 1.226. 1.250. 1.256. 0.955. 1.211. 1 ~ 221. 1.274. 1.012. 1.319. 0.849. 0.286.

H 1.6. -6.7. -3.6. 0.3. 13.6.

. 0.248. 1.138. 1.237. 1.242. 1.141. 1.203. 1.283. 1.226. 1.206. 1.178. 1.155. 1.281. 1.263. 1. 161. 0 '60.

G . . 5.0. -0.2. -6.1.

~ ~

~

0.797. 1. 189. 1.043. 1.279. 1.004. 1.200. 1.325. 1.209. 0.989. 1.267. 1.091. 1 '94. 0.844.

-5.2. 0.7. -2.6. 4.7. -3.3. 1.0. 5.4 ~

~ ~ ~ ~ ~ ~ ~ ~ ~

0.330. 1.063. 1.230. 1.174. 1.251. 1.189. 1.320. 1.178. 1.238. 1.280. 1.265. 1.120; 0.380.

-5.6. 2.9.

. 0.419. 1.012. 1.250. 1.092. 1.239. 0;971. 1.059. 1 ~ 145. 0.966. 0.439.

-0.7. -7.3.

1.1. 1.257'.264. ~

13.0.

~ ~

. 0.419.,1.123. 1.334. 1.225. 'l.194. 1.259. '1.073. 0 '11.

.6.8, 0.388. 0.892. 1.180. 0.776. 1.062. 0.837. 0.388'.

11.4. -2.0. 11.6.

0.282. 0.274. 0.231.

19.1.

POWER TILT IN POWER TILT IN CORE AVERAGE UPPER'ALF OF CORE LOWER HALF OF CORE AXIAL OFFSET

(-,+) . (+,+) (-,+) . (+,+)

0.9843 . 1.0014 0.9939 ~ 1.0010 3.144

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.0128 . 1.0015 '1.0005 .'.0046

( ) (+ -) (-, ) ~ (+,-)

TOP TEN NUCLEAR F-DELTA-H 215 J14HG FOHN%I.5287 165 P 7IN'07 FDHK01 ~ 5057 8 9GC FOHN~1.4991 227 J 2CG FOHN=1.4787 257 G 2CI FDHN*1.4773 245 G14HI FDHK>1.4495 172 N 8NL FOHN-"1.4444 283-E 3CB FOHN*1.4437 282 E 4DF FDHN=1.4391 222 J 78C FDHN=1.4323 LIHITING 3 Fa ELEVATIONS ( TOP 15K, BOTTON 15X, AND NIDDLE 70X OF THE CORE ')

AXIAL FQ(Z) HEAS. PERCENT SOURCE POINT L I HI T FQ(Z) TO LIH. NO. ID 20 2.2585 1.9780 12.42 372 J14XX 6 2.0692 1.7324 16.27 379 J 7XX 52 2.3223 1.9026 18.07 322 P 7XX

0 0 L-93-196 Attachment Page 18 of 18

~

0 CRITICAL BORON CONCENTRATION The critical boron concentration was calculated by adjusting a measured boron concentration to the equilibrium hot full

-power, all rods out condition, as per Operating Procedure 1009.6, "Critical Boron Concentration-Full Power." For Unit 4, Cycle XIV, this calculation was performed at 600 Megawatt days/metric-ton-uranium (MWD/MTU). The following is a summary of the results.

TABLE 6.1:

SUMMARY

OF CRITICAL BORON CONCENTRATION (ppm)

MEASURED WESTINGHOUSE DIFF*

(ppm) (ppm) (ppm) 1170 1187 17

• Acceptance Criteria +/- 50 ppm.

4i

'I S