ML20094H223
| ML20094H223 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/24/1992 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20094H217 | List: |
| References | |
| NUDOCS 9203090174 | |
| Download: ML20094H223 (18) | |
Text
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4 GEORGIA POWER COMPANY VOGTLE NUCLEAR PLANT UNIT NUMBER 1, CYCLE 4 STARTUP TEST REPORT PREPARED B ': /-( <f/ . ,.
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REVIEWED BY: i 4R , / /M3!9l_
@EACTOR ENGINEER APPROVED BY: -
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' REACTOR ENGINEER (NG SU>ERVISOR h300174920302 p DOCK 03000424 PDR
TABLE OF CONTENTS PAGE 1.0 Introduction 1 2.0 Unit 1 Cycle 4 Core Refueling 2 3.0 Control Rod Drop Time Measurement 7 4.0 Initial Criticality 9 5.0 All-Rods Out Isothermal Temperature Coefficient and Boron Endpoint Measurement 10 6.0 Control and Shutdown Bank Worth Measurements 12 7.0 Startup and Power Ascension 14 8.0 Reactor Coolant System Flow Measurement 16
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1.0 INTRODUCTION
The Vogtle NuMear Plant Unit 1 Cycle 4 Startup Test Report summarizes results for tests performed as required by plant procedures following a core refueling. The report provides a brief synopsis of each test and gives a comparison of measured values with design parameters, Technical Specifications, or values assumed in the FSAR safety analysis, Unit 1 of the Vogtle Nuclear Plant is a four loop Westinghouse pressurized water reactor rated at 3411 MWth. The Cycle 4 core loading consists of 19317 x 17 fuel assemblies.
Unit 1 began commercial operations on May 31,1987 and has completed the first three cycles with the following average burnups:
Cycle 1 Complete 10/8/88 15,852 MWD /MTU Cycle 2 Complete 2/23/90 15,789 MWD /MTU Cycle 3 Complet's 9/15/91 18,504 MWD /MTU Seventy two of the 193 assemblies comprising Cycle 4 are based upon the VANTAGE 5 design.Two of the VANTAGE 5 assemblies contain a total of 24 demonstration rods clad in ZlRLO (Reference Tech Spec 5.3,1) and are located in core locations K2 and F14, l
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4 2.0 UNIT 1 CYCLE 4 CORE REFUELING REFERENCES-Westinghouse WCAP -13023 (The Nuclear Design Report for the Vogtle Electric Generating Plant, Unit 1, Cycle 4)
SUMMARY
Unloading of the Cycle 3 core into the spent fuel pool enmmenced on 09/26/91 and was completed on 09/28/91. -
Core reload commenced on 10/18/91 and was completed on 10/23/91. The as-loaded Cycle 4 core is shown in Figures 2.1 through 2.4, which give the location of each fuel assembly and insert. The Cycle 4 core has a nominal design lifetime of 17,500 MWD /MTU and consists of 37 Region 4 assemblies,48 Region SA assemblies,36 Region SB assemblies,56 Region 6A asserrblies and 16 Region 68 assemblies. Fuel assemblyinserts consist of 53 fulllenp* introl rod clusters, two secondary sources and 138 thimble plug inserts. The as' ...ivlies in Regions 6A and 6B contain 3200 fuel rods with Integral Fuel Burnable Absorbers (IFBA).
During the reload, assembly 5F48 (core location L5) was discovered to have a bolt and nut attached to the second grid strap from the top of the assembly. This discovery occurred in the spent fuel pool as the assembly was reised from storage for transport to the core. The bolt and nut were removed and the assembly was inspected and cleaned of discoloration in the spent fuel pool area by a representative from Westinghouse's Fuel Plant in Columbia, South Carolina. Af ter cleaning and inspection, the assembly was approved for core loading.
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l FIGURE 2.1 A. w. V0GTLE UNIT 1, CYCLE 4 REFERENCE LOADING PATTERN R P N W L K J H G F E D C B A 5D54 5004 SE44 SE14 SE11 5D21 SD14 ~~ " '
F-15 N-14 E-12 H-11 L-12 C-14 K-15 5045 5036 SF02 5F59 SF14 SF16 SF17 SF61 SF04 5031 5D78 J-15 L-15 FEED FEED FEED FEED FEED FEED FEED E-15 G-15
'5019 5F05 5F64 SF27 SE63 SE79 SE15 5E82 SE60 5F34 SF70 SFIO SD22 A-7 FEED FEED FEED F-13 M-14 M-7 D-14 K-13 FEED FEED FEED R-7 5056 SF71 SEB1 SE19 5E62 SEOS 5F37- SE47 SE59 SE16 SE78 SF58 SD64 ~
A-5 FEED H-3 G-6 E-2 E-10 FEED L-10 L-2 J-6 C-8 FEED R-5 5D57 SF11 SF47 SE18 5F48 SE03 5F26 5028 5F55 SE17 SF29 SE23 5F42 SF01 SD30 A-10 FEED FEED K-9 FEED J-2 FEED H-1 FEED G-2 FEED F-9 FEED FEED R-10 SD11 5F72 5E70 SE84 SE33 5E07 SE73 5F43 SE67 SE34 SE48 SE69 SE76 5F60 5018 B-3 FEED C-10 P-11 P-7 M-9 M-3 FEED D-3 J-4 B-7 B-11 N-10 FEED P-3 SE21 SF22 SE58 SE22 5F39 5E64 5F40 SE41 SF33 5E65 SF36 SE06 SE68 5F23 5E13 D-11 FEED B-4 F-11 FEED N-4 FEED H-7 FEED C-4 FEED K-11 P-4 FEED M-11 SE09 SF13 5E35 ST25 504u 5F46 SE28 5046 SE3B 5F32 SD24 Sf31 SE36 5F24 SE31 g
E FEED J-12 FEED R-8 FEED J-8 F-14 G-8 FEED A-8 FEED G-4 FEED L-8 SE25 5F15 5E80 SE30 SF53 5E75 SF51 5E37 SF45 5E55 SF56 5E08 SE66 SF20 SE45 D-5 FEED B-12 F-5 FEED N-12 FEED H-9 FEED C-12 FEED K-5 P-12 FEED M-5 5D09 SF67 SE57 SE61 5E43 SE26 SE53 SF44 SE77 SE32 SE42 SE49 5E71 SF68 5007
.8-13 FEED C-6 P-5 .P-9 G-12 M-13 FEED D-13 D-7 B-9 B-5 N-6 FEED P-13 5073 5F12 SF41 SE10 6F50 SE24 5F30 5052 SF54 SE12 SF28 SE39 SF35 5F03 5D69 A-6 FEED FEED K-7 FEED J-14 FEED H-15 FEED C-14 FEED F-7 FEED FEED R-6 5076 SF65 SE56 SE46 SE52 SE01 SF49 5E27 5E83 SE04 SE74 SF66 5053 ,,
A-11 FEED N-8 G-10 E-14 E-6 FEED L-6 L-14 J-10 H-13 FEED R-11 5033 5F08 5F61 SF38 5E54 SE51 $E40 SE72 SE50 5F52 5F63 5F09 5D16 A-9 FEED FEED FEED F-3 M-2 0-9 D-2 K-3 FEED FEED FEED R-9 5066 5005 SF06 SF57 SF21 SF18 5F19 SF69 5F07 503FSD77 ,
J-1 L-1 FEED FEED FEED FEED FEED FEED FEED E-1 G-1 5072 5D63 SE02 SE29 5E20 5D35 5062 F-1 N-2 E-4 H-5 L-4 C-2 K-1 00
- E0!0N 4 (3.405*/o) REGION 6A (3.800"/o) l5E 5F j 2E 101 5A (3.997*/o) REGION 66 (4.100*/o)
!5E i
OE ". 59 (4.305*/o)
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FIGURE 2.2 CONTROL ROD LOCATIONS t R P N W E D C ',
L. K. J. H. G,. F, B A i 1 ,
SA B C B SA 2 SD SB SB SC 3 SA D SE D SA I 4
SC SD -5 B C A C B -6 i SB SB -7 900 C SE A D A SE C -8 SB SB -9 B C A C B - 10 SD SC - 11
'SA D SE D SA 12 SC SB SB SD 13 SA B C B SA 14
-15 ABSOR9ER WATDti AL: AG-leH:0 00 -
BANK NUW8ER OF BANK NUWBER OF--
IDENTIFIER LOCATIONS IDENilFIER LOCAll0NS A .4 SA 8 8 8 SB 8-C 8 SC 4
FIGURE 2.3 Burnable Absorber and Secondary Source Rod Configurations i
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64IFBA Secondary Sovrees Non-IFBA Rod IFBA Rod Guide Tube Secondary Source Rod 5
4 FIGURE 2.4 Burnable Absorber and Source Rod Locations R P N W L K J H G F E D C B A
. 1 321 481 321 321 481 2 433, 4 48! 641 641 481 3 481 641 4SI , 4 641 641 641 64! 64! 641 -5 48I 641 481 -6 321 641 641 641 641 321 -7 900 32.1 641 641 641 641 321 -8 321 641 641 64f 641 321 -9 48I 641 481 - 10 641 641 641 641 641 641 - 11 48I 641 481 12 481 641 641 481 13 48'I I 32I 481 14 433, 15 00 TYPE TOTAL f f fl . , (NUWBER OF IFBA R0DS) . . . . . . . . . . . . . . . . . 3200 lSSA..(NUWBER OF SECONDARY SOURCE RODLETS)... 8 6
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.3.0 CONTROL ROD DROP TIME MEASUREMENT PURPOSE
-The purpose of this test was to measure the drop time of all control rods under hot, full flow conditions in the reactor coolant" system to ensure compliance with Technical Specification requirements.
SUMMARY
OF RESULTS For the hot, full flow condition (Tavg 2, 551*F and all reactor coolant pumps operating), Technical Specification 3.1.3.4 requires that the rod drop time from the fully withdrawn position shall be S. 2.7 seconds from the beginning of stationary gripper coil voltage decay until dashpot entry. All rod drop times were measured to be less than 2.7 seconds. The rod drop time results for dashpot entry are presented in Table 3.1. The mean drop time was determined to be 1.554 seconds.
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TABLE 3.1 CONTROL ROD DASHPOT ENTRY TIMES cot < TROL ROD DASHPOT ENTRY CONTROL ROD DASHPOT ENTRY LOCATION TIME (MSEC) LOCATION TIME (MSEC)
D02 1588 M08 1536 812 1540 H06 1548 M14 1582 H10 1556 PO4 1538 F08 1576 804' 1564 K08 1592 D14 1618 F02 1550 P12 1580 B10 1546 M02 1648 K14 1582 G03 1516 P06 1518 C09 1550 B06 1554 J13 1558 F14 1546 N07 1530 P10 1534 C07 1528 K02 1558 G13 1544 H02 1574 N09 1570 808 1512 J03. 1554 H14 1570 E03 1528 P08 1528 C11 1556 -F06 1544 L13 1556 F10 1552 N05 1534 K10 1546 C05 1544 K06 1522 E13 1546 D04 1570 N11 1536 M12 1544 LO3 1604 D12 1558 H04 1540 M04 1538 D08 1546 H08 1586 H12 1552 SAMPLE SIZE = 53 MEAN = 1.554 SIGMA = 0.025 2 SIGMA = 0.050 MEAN-2 SIGMA = 1.504_ _ MEAN t 2 SIGMA = 1.604 Since the control rods in locations D14 and M02 fell outside of the 2 SIGMA limit, they were dropped an additional six times. The drop times for these extra six drops were all measured at less than 1.610 seconds which was within the 2.7 second Tech. Spec. limit. ,
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i 4.0 INITIAL CRITICALITY PURPOSf; The purpose of this test was to achieve initial rea: tor criticality under carefully controlled
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conditions, establish the upper flux limit for the conduct of zero power physics tests, and operationally verify the calibration of the reactivity computer.
SUMMARY
OF RESULTS Initial reactor criticality for Cycle 4 was achicved by dilution at 1417 on November 19, 1991. The reactor was stabilized at the folkwing critical conditions: RCS temperature 556.6*F, intermediate range power approtimately 1 x 10 8 amps, RCS boron concentration 2094 ppm, and Control Bank D position at 166 steps. Following stabilization, the point of adding nuclear heat was determined and a checkout of the reactivity computer using both positive and negative flux periods was successfully accomplished, in addition, source and intermediate range neutron channel overlap data ware taken during the flux increase preceding initial criticality to demonstrate that adequate overlap existed.
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5.0 ALL RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT MEASUREMENT PURPOSE The objectives of those measurements were to determine the hot, zoro power isothermal and moderator temperaturo coef ficients for the all rods out (ARO) configuration and to measure the ARO boron endpoint concentration.
SUMMARY
OF RESQLIS The measured ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration for HZP are shown in Table 5.1. The isothermal temperature coefficient was measured to be + 2.77 pcm/*F which meets the design acceptance criteria. This gives a calculated moderator temperature coefficient of +4.55 pcm/'F which is within the Technical Specification limit of + 7.0 pcm/'F at HZP. Thus, no rod withdrawal limits were nooded to ensure the +7.0 pcm!'F limit was met. The design acceptance criter:an f or the ARO cridcal boron concentration was satisf actorily met using the revised critoria of +50/-70 ppm as recommoded by Westinghouse. This change reflects the high boron concentration at BOL required for extended life cores: however these limits remain within the Technical Specification requirernents of 3.1.1.3.
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TABLE 5.1 ARO HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT l
Measured ITC Design Calculated Rod Configuration Boron ITC Acceptance MTC Concentration
+ 2.77 pcmrF + 2.96 pcmrF +4.55 pcmrF All Rods Out 2107 ppm ITC -Isothermal temperature coefficient, includes -1.78 pcmrF doppler coefficient l
MTC - Moderator only temperature coefficient, normalized to the ARO condition ARO HZP BORON ENDPOINT CONCENTRATION
' Rod Configuration Measured C. (ppm) Design - predicted C. (ppm) 2166 l All Rods Out 2112 l l
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6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS PURPOSE The objective of the bank worth measurements was to determine the integral reactivity worth of each control and shutdown bank for comparison with the valess predicted by design.
SUMMARY
OF RESULTS The rod worth measurements were performed using the bank interchange method in which: (1) the worth of the bank having the highest design worth (the " Reference Bank")is carefully measured using the standard boron dilution method; then (2) the worths of the remaining control and shutdown banks are derived from the change in .
the reference bank reactivity needed to offset fullinsertion of the bank being measured.
The control and shutdown bank worth measurement results are given in Table 6.1.
The measured worths satisfied the review criteria both for the banks measured individually and for the combined worth of all banks.
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TABLE 6.1
SUMMARY
OF CONTROL BANK WORTH MEASUREMENTS i Predicted Inteorel Bank Measured Bank Worth Percent Difference Bank Worth & Review Criteria (ocm) focm) locm) 267.5 -20.6 Control A 337 1 100 717.8 -2.1 Control B 733 i 10 768.8 -9.0 Control C 845 i 127 483 100 459.0 -5.0 Control D 207.2 -1.3 Shutdown A 210 1 100 896.0 -4.7 Shutdown B 940 1 94 (Reference) 423.9 -2.6 Shutdown C 435 i 100 429.5 -1.5 l Shutdown D 436 1 100 434.7 -11.3 Shutdown E 490 A 100 4604.4 -6.2 All Banks Combined 4909 + 491 13 t , , ,
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7.0 STARTUP AND POWER ASCENSION PURPOSE The purpose of the power ascension program was to provide controlling instructions for:
- 1. NIS intermediate and power range calibration as required prior to startup and during power ascension to take into account the effect of a low leakage core.
- 2. Conduct of startup and power ascension testing, to include:
- a. HZP reactor physics tests
- b. Reactor coolant system flow measurement
- c. Core hot channel factor surveillance
- d. Incore-excore AFD channel calibration
- e. Reactor Coolant System Delta T Calibration
SUMMARY
OF RESULTS Full core flux maps were obtained at about 30%, 49%, 76% and 98% RTP. Hot Channel factors were evaluated at each power plateau and are shown in Table 7.1. The incore and excore delta-l were also ovatusted at each plateau. Reactor coolant flow was ,
determined from calorimetric measurements at 96% RTP. Anincore-excore recalibration test was performed at 76% RTP.
Delta T calibration constants were determined at 98% power using the calorimetric .
measurements and measured values of T HOT and T-COLD.
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TABLE 7.1
SUMMARY
OF POWER ASCENSION FLUX MAP DATA Entam Mao 130 . Mao 131 Mao 135 Mao 13.5 Avg. % 30 49 76 98 Power LOPAR 1.899 1.807 1.683 1.580 FDHN Limit VANTAGE 1.995 1.904 1.768 1.659 5 FDHN Lirnit LOPAR 1.5670 1.4560 1.4336 1.4093 F DHN Measured VANTAGE 1.6084 1.5977 1.5447 1.5310 5 FDHN Measured Core Avg. -0.1 8.4 5.1 4.9 AFD Avg. Core -0.290 17.275 6.659 4.973
% A.O.
Most 1.8978 2.3219 1.8393 1.8089 Limiting FO(2) -
+2%
Transient 3.4023 4.2511 2.4276 1.8311 FO Limit 15
4 8.0- REACTOR COOLANT SYSTEM FLOW MEASUREMENT PURPOSE-The purpose of this test was to determine the flow rate in each reactw coolant loop in order to confirm that the total core flow rnet the minimum fiow requiremant given in Technical Specifications.
SUMMARY
OF RESULTS To comply with the Technical Specifications, the total reactor coolant system flow rate determined at normal operating temperature and pressure must equal or exceed 393,136 gpm. The total core flow was determined to be 400,231 gpm.
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