ML20196F014
| ML20196F014 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 02/09/1988 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8803010471 | |
| Download: ML20196F014 (34) | |
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DUKE POWER COMPANY McGUIRE NUCLEAR STATION.
UNIT.1 CYCLE 5 STARTUP REPORT February 9, 1988 N
8803010471 890209 PDR P
ADOCK 05000369 PDR g h\
TABLE OF CONTENTS P_agg List of Tables 11 List of Figures iii 1.0 ' Introduction 1 1.1 Prestartup NIS Realignment Following Refueling - 3 PT/0/A/4600/78 2.0 Criticality 5 3.0 Zero Power Physics Testing 6 3.1 All Rods Out Boron Endpoint Measurement - 10 PT/0/A/4150/10 3.2 All Rods Out Isothermal Temperature Coefficient 11 Measurement - PT/0/A/4150/12 3.3 Control Rod Worth Measurement - PT/0/A/4150/11 14 3.4 Control Rod Worth Measurement: Rod Exchange - 17 PT/0/A/4150/11A 4.0 Power Escalation Testing 19 4.1 Incore and NIS Recalibration: Post Outage - 21 PT/0/A/4600/02E 4.2 Reactivity Anomalies Calculation - PT/0/A/4150/04 24 4.3 Incore and Nuclear Instrumentation System 25 Correlation Check - PT/0/A/4600/02A 4.4 Core Power Distribution - PT/0/A/4150/02A 26 4.5 Thermal Power Outpit. Measurement - PT/0/A/4150/03 28 i
LIST OF TABLES.
< Page-
- 1. Overlap Data _7
- 2. . Nuclear Heat 8 3 .- Reactivity Corsputer Checkout 9
- 4. Control Rod Worth Measurement: Rod Exchange 18
- 5. . Core Power Distribution Results - 31% Full Power "20
- 6. Quarter Core Flux Map Data for PT/0/A/4600/02E, Incore 22 and NIS Reca'libration: Post Outage
- 7. Core Power Distribution Results - 97% Full Power 27
- 8. Thermal Powe r Output Measurement Results 29 11
LIST OF FIGURES ~
.z; P_ age
- 1. Unit 1 Cycle 5 Core Loading Pattern 2
- 2. Assemblies to Use for Calculating IR and PR 4 Calibration-
~ 3. ARO ITC Heatup and Cooldown Data: First Run 12
- 4. ARO ITC Heatup and Cooldown Data: Second Run 13
- 5. Shutdown Bank B Rod Worth 15
- 6. Control Bank D Rod Worth 16
- 7. Incore and NIS Recalibration Results 23
$41
1.0 Introduction Core loading for McGuire Unit 1 Cycle 5 was started on October 1, 1987 and was completed October 5. The core for McGuire 1 Cycle 5-consists of 189 Westinghouse optimized fuel assemblies and 4 Babcock and Wilcox Hark-BW demonstration fuel assemblies. To control power peaking and maximize cycle length, 60 Westinghouse Wet Annular Burnable Absorber (WABA) inserts are utilized.
Due to baffle jetting damage discovered af ter McGuire 1 Cycle 3, fuel assemblies residing in potential baffle jetting locations had fuel clips installed to reduce the potential for damage.
Criticality, zero power physics testing and power escalation testing began November 11, 1987. The unit reached 100% power on November 17, 1987. Due to mechanical difficulty with #4 Turbine Governor Valve, the unit was required to reduce power to 97% on November 25. 1987, and remained there until January 19, 1988 at which time full power was achieved by bypassing flow to A and B feedwater heaters. Tests originally scheduled for hot full power equilibrium conditions were performed at 97% full power on December 1, 1987 with the exception of PT/0/A/4150/03, Thermal Power Output Measurement, which was performed January 6,1988 following completion of the NC Flow Test.
Figure 1 gives the Unit 1 Cycle 5 core loading pattern.
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McGuire Jnit :.
Cyc:.e 5 QUADRANT ASM i 4 1 Figure 1 INS #
Core Loading Pattern 3 2 E17 G04 F45 G58 F34 G23 E32 260KT 306KT 29ET 279(T 23tKT 33ET 31tKT E40 F48 G60 F27 G26 F08 G41 fib G61 F80 E01 325KT R08 4P80K R02 BP7E R28 8PiO4K A06 @75K R15 27ET 2
E02 F53 G43 F5B Fi7 E30 F31 E34 F30 F55 G02 F59 E55 FSE Roi G73K 32ET 236KT 3
24ET 29ET R24 281KT A09 SS6 R12 310KT F54 G53 E21 Gii E58 G35 E35 G35 E10 G31 E25 G33 F33 Rig 8Pii6K 282KT SP7E R22 P95 R20 4
4P9E R14 8P82K 267KT SP77K R13 E04 G6S F81 G22 E15 G25 F02 F49 F04 G01 E46 G40 F64 G67 E41 29ET 26*T SPitik 27ET 8P100N R30 #77K 2 SET 5
65 #76K R03 SP109K 307KT 8P94K 247KT G20 F32 F25 ESS G13 E39 G48 Eii G46 E36 G38 E03 F16 F14 G56 g 32iKT rib 262KT 252KT 8P98K R31 8P87K A04 BP93K R33 8P85 240KT 33ET R32 338KT F63 G37 E24 G34 F05 G45 EOS G69 E49 G12 F11 G50 E38 Gio F37 8Pil2K 28ET #96K 251KT BP10ik R21 BP110K 33ET 7
327XT BP96K R17 BP103N 23ET 253KT 8P91K G57 F12 F23 E33 F51 ES9 G70 dis G71 E52 F52 E09 F28 Foi GS8 g 129 R34 337KT R44 32ET A07 #100K R23 098K Rio 25ET R46 27ET R35 304KT FSB G24 E08 G44 F03 G29 E26 G72 E13 G47 F09 G49 EiS G55 F38 g 23ET BPSE R27 F80K 274T BPiOSK 284T 4P94K 317KT 8PB6K 33ET BP81K R38 BPii4K 272KT GO5 F13 F19 EiB G54 E07 GOS E51 GiB E19 G09 E54 F24 F21 Gi9 8P89K R39 BP92K 255KT 232KT R48 265KT 10 250KT R40 330KT 256KT 8PtiSK R37 8P99K R16 E12 G59 F47 G28 E31 G27 FOS F50 F07 Gi4 E14 G42 F44 G64 E45 gg 261KT F79< R42 F97K, 3tET 8P108k 29ET 257KT 23*T 8PiO60 273KT BP8E R05 #89< 245KT F39 G32 E57 G51 E60 G30 E23 G15 E47 G21 E28 G08 F40 g R28 #93K R25 BP84K 291KT BP95 R47 8P113K 237KT 8P107K R50 @97K R29 EOS F41 G07 F43 F15 E42 F20 E4B F22 F36 G16 F48 E43 g 287KT 241XT #8E R11 29ET R41 SS6 R51 23ET R45 G83K 26ET 290KT E53 F35 G65 F20 G39 F10 G52 F29 G62 F42 E20 g 271KT R36 #74K R49 8P102E R43 8P90K R53 992K R52 258KT E27 G17 F82 G63 F57 G03 E37 15 297KT 302KT 27ET 24ET 32*T 314KT 249(T R P N M L K J H G F E D C B A
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1.1 Prestartup NIS Realignment Following Refueling - PT/0/A/4600/78 This procedure was performed on October '23,1987.
This test was used to calculate preliminary calibration data for the intermediate range (IR) and power range (PR) detectors following refueling.
The Cycle 5 preliminary calibration data.was determined by taking the End of Cycle 4 (EOC4) calibration data and adjusting it by a-weighted average of the ratio of the sum of the predicted assembly powers for the cycle 5 loading to the sum of the measured assembly powers from the last Cycle 4 Incore/Excore calibration. The core locations used to calculate the ratio of the predicted Beginning of Cycle 5 (BOC5) assembly powers to the measured EOC4 values are shown in Figure 2.
The predicted BOC5-to-EOC4 IR ratio was NO.97; the predicted BOC5-to-E004 PR ratio was 40.78. Based on these results, the IR and PR currents were adjusted prior to Cycle 5 Initial Criticality.
3
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-f Figure 2 Assemblies to Use for Calculating IR and PR Calibration Setpoints n
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$5 R P N N L K 3 H G i: E D C OW %s Core locations used for PR calibration 4
Core locations used for IR calibration
. i 2.0 Criticality - PT/0/A/4150/28 On November 11, 1987 boron samples were taken in preparation for the approach to criticality. These samples indicated reactor coolant boron to be 1793 ppm. Since it was desired to go critical with a significant amount of Control Bank D inserted ($750 pcm), a target value of 1424 ppm was chosen for reactor coolant boron concentration. This represented the predicted BOL, AR0, HZP, No Xenon, equilibrium Samarium critical boron concentration of 1499 ppm lass 75 ppm. Calculations using the unit Data Book (OP/1/A/6100/22) indicated a volume of 14016 gallons of demineralized water should be added to the system to dilute from 1793 ppm to 1424 ppm.
On November 11, 1987, this dilution of the reactor coolant system was started. The dilution was secured after 14016 gallons of demineralized water had been added to the system. Af ter appropriate system mixing, Chemistry samples indicated Reactor Coolant System boron was 1446 ppm.
November 12, 1987, rod withdrawal commenced starting with Shutdown Bank A.
As rods were withdrawn, both source range detectors were observed and rod motion was stopped each time flux level doubled. At these points a set of counts were taken on each source range detector and Inverse Count Rate Ratio (ICRR) was plotted to monitor the approach to criticality. The unit achieved criticality at 0415 hours0.0048 days <br />0.115 hours <br />6.861772e-4 weeks <br />1.579075e-4 months <br /> on November 12, 1987 with Control Bank D at 12/13 steps withdrawn. The predicted critical position per OP/0/A/6100/06, Reactivity Balance Calculation was 85 steps withdrawn on Control Bank D.
5
3.0 Zero Power Physics Testing- (ZPPT)
Zero Power Physics Testing for McGuire 1 Cycle 5 started M:::mber 12, 1987 and was completed November 14, 1987. The output si Power Range Detector N43 was used as input to the reactivit.y computer for Zero Power Physics Testing. All acceptance criteria for ZPPT were met.
A minimum of one decade of overlap between the source range and the intermediate range detectors was verified on NovemL-r 12, 1987. The results are shown on Table 1.
The point of adding nuclear heat was determined Novenber 12, 1987. This was done by establishing a slow positive startup rat e and observing a change in plant parameters such as an increase in R: actor Coolant System average temperatures (Tave) with a change in the reactivity trace and an increase in pressurizer level. The test was performed a second time to establish repeatability of the data. Table 2 givea the results of the two tries.
Nuclear be9t was determined to be at an average flux level of 6.75 x 10,7 amps on the reactivity computer picoammeter (N43) and,7 7.03 x 10 amps on Intermediate Range Detector N35 and 8.30 x 10 amps on Intermediate Range Detectog N36. ,Jrom these results the test band for ZPPT was determined to be 10 to 10 amps on the reactivity computer.
On November 12, 1987, an on line checkout of the reactivity computer was perfo rmed. This was done by withdrawing Control Bank D until a positive reactivity insertion of N+25 pcm was indicated on the reactivity computer.
The time for the flux level to double was measured and from this doubling time (DT), the reactor period was calculated (period = DT/.693). Using the reactor period, the amount of reactivity was determined using the predicted data. This reactivity was compared to the reactivity computer indication. The test was repeated for a reactivity insertion of $+50 pcm.
The results met all acceptance criteria and are given in Table 3.
An elestronics only negative reactivity insertion test was also completed satisfactorily as part of PT/0/B/4600/55, Reactivity Computer Periodic Test.
6 .
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4 TABLE 1 4
Overlap Data Source Range Intermediate Ranae'
. cps- amps-
.N31 N32 N35 N36 1.7 x 10 3 1.9 x 10 3
2.8 x 10'II 3.0 x 10'II 1.8 x 10 2.5 x 10 0 -10 -10 2.0 x 10 2.2 x 10 When SR blocked' '4.0 x 10 4 -10 4.0 x 10 0 2.9 x 10 3.1 x 10 10' l
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TABLE 2:
Nuclear Neat Reactivity Computer -Intermediate Range N43 N35 N36 6.37 x 10 ~7
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6.35 x 10' 7.52 x 10
~7 7.15 x 10 7.69 x 10 ~7 9.08 x 10
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AVERAGE '6.75 x 10 amps 7.03 x 10 8.30 x 10 amps Test Band:
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10 to 10 amps on N43.
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3.1 All Rods Out Boron Endpoint Measurement - PT/0/A/4150/10 This test was performed November 12, 1987. Control Bank D was initially at 209/210 steps withdrawn, the Reactor Coolant System boron concentration was 1531 ppm and the Pressurizer boron concentration was 1512 ppm.
Control Bank D was pulled to the All Rods Out (ARO) Configuration and the resulting reactivity change was converted to equivalent boron using the predicted Differential Boron Worth. Control Bank D was then reinserted to the just critical condition and the test was repeated.
The results of these reactivity changes were each added to the initial Reactor Coolant System boron concentration to give two values for the ARO Boron Endpoint. These values were averaged to give the j fincl result of 1537 ppm. This value met the acceptance criterion for the Hot Zero Power (HZP) ARO Critical Boron concentration of 1499 150 ppm.
10
3.2 All Rods Out Isothermal Temperature Coefficient Measurement -
PT/0/A/4150/12 This test was performed on November 12, 1987. The test measures Isothermal Temperature Coefficient (ITC) by plotting Reactivity versus Average Reactor Coolant System Temperature. The Moderator Temperature. Coefficient (MTC) is found using the relationship as follows:
MTC (pcm/*F) = ITC - Doppler Temperature Coefficient The acceptance criterion on the ARO ITC was 1.64 13.0 pcm/ F. The predicted Doppler Temperature Coefficient was -1.31 pem/*F.
Control Bank D was at 194 steps withdrawn and the Reactor Coolant System boron concentration was 1547 ppm at the start of the test. A heatup/cooldown was performed while keeping rod position and boron ,
constant to determine reactivity change versus temperature. The heatup/cooldown was performed a second time to establish repeatability of the data. The results are shown in Figures 3 and 4.
The average ARO ITC was found to be +1.90 pcm/ F. This fell within the acceptance criterion band. This gave an ARO MTC of +3.21 pcm/ F which was witbin acceptable Technical Specification limits.
- Following the completion of this test, PT/0/A/4150/31, Determination of Rod Withdrawa2 Limits to Ensure Moderator Temperature Coefficient within Limits of Technical Specifications was performed. The results of this test indicated there were no rod withdrawal limits needed for Cycle 5. i i
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4 3.3 Control Rod Worth Measurement - PT/0/A/4150/11 On November 13, 1987, Shutdown Bank B rod worth was measured using the established boration/ dilution method. There were no other rods in the core at the time. Shutdown Bank B was predicted to be the highest worth bank and was measured using this method so as to serve as the reference bank for Control Rod Worth Measurements by Rod Exchange. The measured worth of Shutdown Bank B was 717 pcm. The predicted worth was 789 pcm i 118 pea. This represented an error of 9.1% and was within the acceptance criterion of 115%. Figure 5 shows the measured integral and differential rod worths for Shutdown Bank B. Following the performance of PT/0/A/4150/11A Control Rod Worth Measurement: Rod Exchange (discussed in Section 3.4), the core was left in a configuration suited for performing rod worth measurement for Control Bank D by boration/ dilution. This was done on November 13, 1987. Again, there were no other rods in the core. The measured worth of Control Bank D was 492 pcm. The predicted worth was 515 pcm 1 52 pcm. The error was 4.5% well within the acceptance criterion, also. The resulting integral and differential rod worths
; for Control Bank D are shown in Figure 6.
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SHUTDOWN BANK 8 WOATH Oifferential and Integral RCC Bank (RCCA) Worth
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3.4 Control Rod Worth Measurement: Rod Exchange - PT/0/A/4150/11A On November 13, 1987 the rod exchange method of control rod worth measurement was performed. Shutdown Bank B was used as the reference bank and its worth was measured by the boration/ dilution method (see Section 3.3). With the reference bank essentially all the way in and the reactor just critical, each conti o1 and shutdown bank was exchanged with the
' reference bank. The integral worth of the bank being measured (i.e.,
the test bank) was deternined from the difference in the critical rod position of the reference bank with and without the test bank in-the core. The measured worths were compared with predicted worths and all banks were within the acceptance critcria of 30% or +200 pcm whichever was greater. The total rod worth sas <10% from predicted which met the acceptance cirteria. The results of the rod exchange test are given on Table 4. d b I i i r 17
; TABLE 4 Control Rod Worth Measurement: Rod Exchange
+
Bank Predicted Worth Measured Worth Percent Identification pcm pcm++ Difference Shutdown Bank B 789 717* - 9.1 (reference) Control Bank D 515 491 - 4.7 Control Bank C 786 721 - 8.3 Control Bank B 732 693 - 5.3 Control Bank A 279 267 - 4.3 Shutdown Bank E 348 336 - 3.4 Shutdown Bank D 452 432 - 4.4 Shutdown Bank C 451 441 - 2.2 Shutdown Bank A 274 309 +12.8 TOTAL ROD WORTH. 4626 4407 - 4.7
- Measured by boration/ dilution method
, Measured - Predicted x 100 Predicted i
I
++ Rounded to nearest pcm 18
4.0 Power Escalation Testing McGuire Unit 1 Cycle 5 Power Escalation testing started November 14, 1987 at the conclusion of ZPPT and was completed January 6, 1988. The unit went on line November 14 at 0705 hours. The unit experienced several holds during power escalation which were scheduled to allow testing of the Main Turbine (due to work conducted on the High Pressure Turbine and replacement of High Pressure Turbine blades). During a hold for turbine testing at 31% power on November 15, Core Power Distribution, PT/0/A/4150/02A, was performed. An incore tilt of 3.1% was measured in Quadrant 2, and was reported to appropriate Nuclear Design Engineering staff for evaluation. Otherwise, the results from the test indicated that Core Power Distribution Technical Specification Limits for operation at 100% power would not be violated, and all test acceptance criteria were met. Table 5 shows the test results. Following completion of turbine testing at 31%, power was increased to 49% at 2.5%/hr. Power was held at 49% for approximately 10 hours for turbine testing. On November 16, power escalation resumed. From 50% to 80%, PT/0/A/4600/02E, Incore and NIS Recalibration: Post Outage, was performed (see Section 4.1). The excore detectors were calibrated at 76% power on November 17, power escalation then resumed, and the unit achieved 100% Full Power late that dav. The remaining tests designated for Hot Full Power Equilibrium Conditions were scheduled for 100% Full Power. However, on November 18, the unit was shutdown due to High Reactor Coolant System leakage. Upon return to service on November 20, mechanical difficulty with #4 Turbine Governor Valve held power to 97%. Therefore, high power testing as described in Sections 4.2-4.5 was completed at this power level. 19
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TABLE 5 3 Core! Power Distribution Results. 131% Full Power Unit 1 Cycle 5 Map'FCM/1/05/001 Date/ Time Map Taken 11/15/87 0557 hours .; Power Level 31.0% ! Cycle Burnup 0.21.EFPD 10.52 MWD /NTU l Boron Concentration 1490 ppa ! Control Rod Position Control Bank D at 141 steps withdrawn Maximum F : O 2.1048 at Axial Loc. 32, Horiz. Loc B-11 Maximum Fg : 1.2663 at Axial Loc. 32 Maximum pin F 1.4926 at Horiz. Loc. E-14 r Maximum error AH (fr il Predicted) 8.47% at Horiz. Loc. D-04 Maximum F /K(Z) 2.1091 at Axial Loc. 32 Maximum % Reduction in Axial Flux None [ i a Difference (AFD) Wings ! Minimum % Margin to AFD Wings -48.0112% at Axial' Loc.-33 ! R,,, (Tech Spec 3/4.2.3) 0.8287 . Total Reactor Coolant Flowrate 407,933 gallons / minute t (Process Computer) ' i 4 Total Incore-Axial Offset +3.221% ! P
. Incore Tilts %:
!r Upper Core Lower Core ; r Quadrant 1: -0.965% Quadrant 1: -1.257% a Quadrant 2: 3.111 Quadrant 2: 3.098% i Quadrant 3: -0.777 Quadrant 3: -0.367% Quadrant 4: -1.369% Quadrant 4: -1.473% NOTE: Axial location 1 is the bottom of the core. [
! Axial location 61 is the top of the core.
i l s 4 i 20 .
4.1 Incore and NIS Recalibration: Post Outage - PT/0/A/4 iOO/02E This test was started on November 16, 1987 and was run during the power escalation from 50% to 80% Full Power. The data obtairied from this test were used to set the nuclea r instrumentation lystem amplifier gains, the axial flux difference function of ;he overpower AT setpoints and to determine the correlation between incore and excore axial offsets. The data were collected by taking quarter core flux maps and associated excore detector currents at eight different axial offsets i as indicated in Table 6. (The quarter core flux map pattern had ' previously been verified as an accurate representation of axial offset through PT/0/A/4150/23, Quarter Core Flux Map Qualification Test). These data were then input into a benchmarked off-line computer program which generated the output shown in Figure 7. The appropriate factors were then input into the plant instrumentation systems and all acceptance criteria were met. t t l r I E I L 5 t l 2 I J 6 21
IT ; -
;ie:
TABLE 6
' Quarter Core Flux Map Data for H_
PT/0/A/4600/02E, Incore and NIS Recalibration: Post Outage Map Average Thermal Power (%) Incore Axial Offset (%)
- 1) QCM/1/05/004 52.120 20.104
- 2) QCM/1/05/005 55.250 17.682
- 3) QCM/1/05/006 57.940 14.998
- 4) QCM/1/05/007 60.870 12.106
- 5) QCM/1/05/008 63.170 9.286
- 6) QCM/1/05/009 66.370 7.391
- 7) QCM/1/05/010 69.040 5.870
- 8) QCM/1/05/011 71.580 4.207
- 9) QCM/1/05/012 74.300 2.252
(- 10) QCM/1/05/013 75.670 1.148
- 11) QCM/1/05/014 75.630 0.318 22
Figuro 7 Encore Currents ud Voltages Correlated to 1001 Full Poser at Various Antal Offsets Unit ! Cycle 5 FULL POWER MTECTOR C:.* RENTS (MICROAMPSI CORRESPON0 lug TO VARIOUS INCORE A!!AL OFFSETS INCORE CETECTCR N 41 MTECTOR N-42 MTECTOR N 43 MTECTOR N-44 A11AL OFFSET T B i i i 3 i I 20.0 251.7 191.9 233.2 171.6 228.4 256.0 225.4 172.4 20.0 237.2 198.4 219.9 187.0 214.6 285.0 213.6 199.6 10.0 222.7 214.9 204.6 202.4 204.9 314.1 201.1 204.9 0.0 208.2 231.4 193.2 217.0 193.1 343.1 190.0 224.2
-10.0 193.7 247.9 179.9 233.1 181.3 372.1 178.2 241.5 20.0 179.2 264.4 166.6 241.5 169.5 401.2 164.3 258.7
-30.0 164.7 291.0 153.2 263.9 157.7 430.2 154.5 276.0 r8 0.9793 -0.9125 0.9713 -0.9926 0.9723 -0.9851 0.9700 0.9865 NORMAlllED DETECTOR VOLTAGES (VOLTSI AT VARIOUS AIIAL OFFSETS IhCCPE DETECTOR N 41 OETECTCR N 42 DETECTOR i 43 MIECTOR N-44 AI!AL OFFSET i 8 T-3 i i TB i i T1 i i TI 30.0 10.071 6.546 3.525 10.054 6.544 3.491 9.854 6.214 3.641 9.884 6.404 3.479 20.0 9.491 7.140 2.350 9.479 7.152 2.327 9.347 6.920 2.420 9.366 7.046 2.319 10.0 9.910 7.735 1.175 8.905 7.741 1.164 8.639 7.625 1.214 9.640 7.684 1.160 0.0 9.330 8.330 0.000 8.330 8.330 0.000 9.330 9.330 0.000 8.330 8.330 0.000 10.0 7.750 E.925 -1.175 7.755 8.919 -1.164 7.921 9.035 1.214 7.812 9.972 -1.160 20.0 7.169 9.520 -2.350 7.181 9.506 -2.327 7.313 9.740 2.429 1.294 9.614 2.319
-30.0 6.529 10.114 -3.525 6.6% 10.096 -3.491 6.804 10.446 -3.641 6.776 10.256 3.479 AFD thCORE/EICCRE RAi!OS FOR GUADRANTS 1 - 4 G'JAL 4 QUAD 2 GUA0 1 GUAD 3 N 41 N 42 N 43 h-44 M=1.418 R = l.432 R = 1.373 R = 1.437 PPEPARED IV.C%). M.DATE.").l.7/97 n
4.2 Reactivity Anomalies Calculation - PT/0/A/4150/04 This test compared the actual core reactivity to the predicted core J reactivity by taking into account-the actual Reactor Coolant System 3 . boron concentration, Xenon and Samarium wortns, rod positions and power level and adjusting these to the ARO, Hot Full Power (NFP), equilibrium Xenon and Samarium condition. Theoretical and actual Reactor Coolant System boron concentration for this conditions were then compared. The test, performed at 97% on December 1, 1987 indicated that the actual ARO, HFP, equilibrium Xenon and Samarium condition boron concentration was 1049 ppm. This compares to a predicted value of 1022 ppm. The 27 ppa difference translated into a 245 pcm error between actual and predicted. reactivity worths. This was, however, within the acceptance criterion for the test of 11000 pcm. i, M i t 1 0 i i 4 s 1 1 1 e a t f' 4 24 I
i 4.3 Incore and_ Nuclear Instrumentation System Correlation Check - PT/0/A/4600/02A This test was used to compare the incore axial offset as indicated by a full core flux map to the axial offset indicated on the plant computer by the excore detectors. This test also verifies the incore/excore calibration data that b.ad been implemented during PT/0/A/4600/02E, Incore and NIS Recalibation: Post Outage. The test was performed at 97% on December 1,1987. The indicated incore axial offset from flux map FCM/1/05/016 was 2.039%. The core average axial offset from the excore detectors was 4.033%. These results gave an absolute difference of 1.994% and was within the acceptance criterion of 13% difference. 35
g 4 4.4 Core Power Distribution - PT/0/A/4150/02A On December 1, 1987, PT/0/A/4150/02A, Core-Power Distribution, was performed to verify the core power distribution technical specification limits for operation would not be violated. The reactor was at 97% Full Power and equilibrium conditions. Again a tilt (2.3% in magnitude) was measured in Quadrant 2. Appropriate Nuclear Design personnel were notified of the test results for evaluation. All acceptance criteria for this test were met. Table 7 gives the test results. 4 't i f i I 1 l l i ! 26
i k-TABLE 7 Core Power Distribution Results 97% Full Power Unit 1 Cycle 5 Map FCM/1/05/016 Date/ Time Map Taken 12/1/87 1305 hours Power Level 97% Cycle Burnup 14.1 EFPD 586.1 MWD /MTU Boron Concentration 1050 ppm Control Rod Position Control Bank D at 206 steps withdrawn Maximum F : 1.8879 at Axial Loc. 32, Horiz. Loc. B-11 Maximum F : 1.2011 at Axial Loc. 33 Z Maximum pin F 1.4327 at Horiz. Loc. E-12 Maximum error (from predicted) 6.28% at Horiz. Loc. B-13 Maximum F /K(Z) 1.9222 at Axial Loc. 40 Maximum % Reduction in Axial Flux None Difference (AFD) Wings Minimum % Margin to AFD Wings -5.6903% at Axial Loc. 40 R (Tech Spec 3/4.2.3) 0.9521 Total Reactor Coolant Flowrate 405,338 gallons / minute (Process Computer) Total Incore Axial Offset +2.039% Incore Tilts %: Upper Core Lower Core Quadrant 1: -0.517% Quadrant 1: -0.921% Quadrant 2: 2.346% Quadrant 2: 2.284% Quadrant 3: -0.814% Quadrant 3: -0.053% Quadrant 4: -1.016% Quadrant 4: -1.310% NOTE: Axial location 1 is the bottom of the core. Axial location 61 is the top of the core. 27
4.5 Thermal Power Output Hessurement - PT/0/A/4150/03 This test was .used'to verify that the primary and secondary heat balances on the plant computer were consistent with primary and secondary heat balances on a benchmarked offline computer. The test was run on January 6,1988 at 1500 hours at 96% F.P. The results are shown in Table 8. The acceptance criterion of 12% difference between the offline computer and the plant computer was met. l 28
t
,_ .' s .
n TABLE 8 Thermal Power Output Measurement Results - Plant Computer off-Line Computer !
% Wt %
Primary Heat Balance 96.55 3293.40 96.69 3298.00 Secondary Heat Balance 96.22 3282.30 -96.28 3284.30 29
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e DUKE POWER GOMPANY P.O. Isox 33180 CitARLOTTE. N.O. 98949
!!AL H. M'CKER rsutessons rua enemisert (704) 073 4gu wixasan emoet-February 24, 1988 U.S. Nuclear Regulatory Coussission A cument Control Desk Washington, D.C. 20555
Subject:
McGuire Nuclear Station, Unit 1 Docket No. 50-369 Cycle 5 Startup Report Gentlemen Pursuant to McGuire Technical Specification 6.9.1.1, attached is the Startup Re-port for McGuire Unit 1 Cycle 5. The report documents the initial use of Wet Annular Burnable Assemblies in a McGuire core. Very truly yours.
- /Y > y n _
Hal B. Tucker SAG /103/j ge Attachment xct Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta St., NW, Suite 2900 Atlanta, Georgia 30323 Mr. Darl Hood Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 j6Eb Mr. W.T. Orders [g NRC Resident Inspector l McGuire Nuclear Station}}