ML20113J193
| ML20113J193 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/24/1992 |
| From: | Wendt T GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20113J192 | List: |
| References | |
| NUDOCS 9208070143 | |
| Download: ML20113J193 (19) | |
Text
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b GEORGIA POWER COtAPANY <
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VOGTLE NUCLEAR PLANT 4
UNIT 2, CYCLE 3 STARTUP TEST REPORT I
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.I PREPARED BY: F1/u; et - / d[t /9L-t/ REACTOR ENGINEER REVIEWED BY: vf d[7N1 -
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'ACTOf ' ER / f' i APPROVED BY / / /1# /Ir u.v- /4' 95 -
R'EACTO R M EERING SUF4RVISOR 54 MO:
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- 9200070143-920003 a
PDR .ADOCK 05000425 '
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LASLE OF CONTENTS f1GE 1.0 Introduction 1 2.0 Unit 2 Cycle 3 Core Refueling 2 3.0 Control Rod Drop Time Measurement 8 4.0 Initial Criticality 10 5.0 All Rods-Out Isothermal Temperature Coef ficient and Boron Endpoint Measurement 11 6.0 Control and Shutdown Bank Worth Measurements 13 7.0 Startup and Power Ascension
- 15 8.0 Reacter Coolant System Flow Measurement 17 i
P
1.0 INTRODUCTION
The Vogtle Nuclear Plant Unit 2 Cycle 3 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 2 of the Vogtle Nuclear Plant is a four loop Westinghouse pressurized water reactor rated at 3411 MWth. The Cycle 3 core loading consists of 19317 x 17 fuel assemblies.
Unit 2 began commercial operations on May 19,1989 and has completed the first two cycles with the following average burnups:
Cycle 1 Complete 09/14/90 17,161 MWD /MTU Cycle 2 Complete 03/09/92 17,008 MWD /MTU Seventy six of the 193 assemblies comprising Cycle 3 are based upon the VANTAGE 5 design, f
1
2.0 UNIT 2 CYCLE 3 CORE REFUELING
- F.EFERENCES Westinghouse WCAP 13119 (The Nuclear Design Report for the Vogtle Electric Generating Plant, Unit 2, Cycle 3)
SUMMARY
Unfoadin0 of the Cycle 2 core into the spent fuel pool commenced on 03/20/92 and was completed on 03/23/92.
Core reload commenced on 04/04/92 and was completed on 04/07/92. The as loaded Cycle 3 core is shown in Figures 2.1 throu0h 2.5, which give the location of each fuel artsembly and insert. The Cycle 3 cole has a nominal design lifetime of 18,500 MWD /MTU and consists of 17 Region 2 assemblies from Unit- 1,16 Region 3 assemblies, 60 Region 4A assemblies,16 Region 48 assemblies, 48 Re0 l on SA assemblies and 28 Region SB assemblies. Fuoi assembly inserts consist of 53 full length control rod clusters, two secondary sources and 138 thimble plug inserts. The assemblies in Regions BA and 58 contain 4672 fuel rods with Integral Fuel Burnable Absorbers (IFBA).
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FIGURE 2.1 UNIT 2 CYCLE 3 REFERENCE LOADING PATTERN 8 9 C D I f 6 0 4 8 6 8 8 P $
18 5836 pel $ sci $M3 Smot pel M11 9 yds if3pt F798 128 116D0 leet let l 14 litt D44 StOf pel MM S076 Soft le64 SOM Sett la29 i iles tiel Stes tuo D8649 Alfs Siet tel 3aes elfe llel
)
13 ls40 6809 Met $464 SPI 3 H79 $013 SPS3 WG1 late left 6000 ht?J 16006 30165 Mel R3M itsps alt 1 17tos R$M tips out 3 Met Sites 1444 !
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928 Met'v43 ilms teet 1Flet feel teet loses tiet 3'4s Het slet stas tlfst te SP01 lad lest $319 Sete SP35 W30 St39 $ Pit $P67 9811 SAM 9*65 6069 9757 ilas 8570 238 lites $48 ente Daps tit? 9Mnt MG3 thet Blet 1908 8596 1998 4 latl Self le?3 $* ft k33 p60 6444 WF9 604 W37 lose leep Pet tell pH toes is7DS 2994 17608 79508 taopt 293D9 17tDS 30*pt 1998 79909 1793 Mot 3%De 9eg 8 5645 leM WH SPl4 telt le42 PM SM2 pl8 9120 $f.Il 9 08 ke3 68 74 left test tul ilms 841 1928 3358 1628 elet 428 t174 feet tlM Feet alf? 6468
? Stil 8060 SPf1 SPM Sati Dee Stil pit Stil DM stb WF3 ps2 left lett litel lutDe slet 19es $124 iles 3Z309 les 30500 tot #ttes 17908 R$ft Stet sect 6 Seit lat.3 p3f 1812 tell l>07 5734 6829 W!! Wil ple left W39 lael SP61 test 2406 it?DS 3128 16FDs the 19109 4143 telet ett3 4 500 3829 let ti9e Hept S 9424 Stol R32 M60 H31 W38 Mit leM Stel pel M14 SM4 5437 lett 6401 F198 feet it?! 4108 2834 1400 30008 138D8 Stes sent 5Fes 429 tift Mist 6el e pet till $ Pit lete MM l*69 SP05 UFl 5424 WM pe6 St?1 M47 tu3 fetet 1601 laps teios tems all? 2006 32108 97DS R379 31188 titt 3 lett lett less St23 WZ3 left leM 5877 W32 Stil leet teel loos 1604 3U48 3038 EMF $4s alm 1728 tlft 11168 M40 #FFDS !? tdt 1424 i tale SP'.5 StM le67 last lar) St% lelt la02 P69 Sm64 16?pt alF3 29fDs e604 08130 elf 1 3128 teof Maps elec 14108 i SM pe woi ur3 ui 96: nii BPpl 17109 4109 6?DS 3105 Sees 1534 5 Region 2 (2.601 w/o) SP Region 4B (4.190 w/o)
E Region 3 (3.118 w/o) 5R Region SA (3.805 w/o)
SP Region 4A (3.807 w/o) SR Region 58 (4.198 w/o) 3
4 FIGURE 2.2 CONTROL ROD LOCATIONS R P N W D L. K. J. H. G. F. E. C B A, I 4 t
l SA B C '
i B SA 2 l
SD SB SB SC 3 SA D SE D SA , 4 SC SD -5 B C A C B -6 SB SB -7 900 0 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 ASSORDER WATERIAL: AG-ilHD 00 .
BANK NUWBER OF BANK NUWBER OF IDENilFIER LOCATIONS IDENTIFIER LOCAilDNS A 4 SA 8 8 8 SB 8 C 8 SC 4 D 5 SD 4 SE 4 4
4 FIGURE 2.3 Burnable Absorber Configur6tions i
e e __
i e G 9 9 e 9 9 9 i 4 e i 9 9 9 9 tel 9 9 9 9 91 l 9 9 l 9 8 9 9 9 9 91 0 9 '
9 9 I O 9 9 9 9 9 9 9 9 19 9 9 9 9 9 _
9 9 -
4 9 9 9-9 9 1 9 9 9 9 i9 9 9 9 9 Gi 9 9 i 9 9 9 9 9 9 9 9 9 9 I i 32 IFBA ASSEWBLY 48 IFBA ASSEWBLY i
e e e ce e ce e e 9 9 9 19 0 0 4 ~~-
G 9 19 41 0 9 9 9 9 G # 4 le 9 e i 9 9 10- 9 9 9 9 9 9 91 9 I 4 1 9 9 I 4 19 9 9 9 9 9 9 9 9 9 O!
O i 9 0 91 9 9 9 9 I O 9 9 9 9 9 9 9 -
9 9 9 1 19 9 9 9 9 9 G # # I O 9 9 9 9 ._ 9 9 9 9 9 9 91 4 9 9 9 I e 9 9 e le 9 9 9 9 9 9 et 4 9 9 9 ~
9 9 9 9 9 14 9 9 9 9 9 9 O 9 9 9 9 19 9 9 9 99 9 S ei 6
64 lFBA ASSEWBLY 80 IFBA ASSEWBLY kEGEND t C FUEL ROD D "NSTRUWENTATION I TUBE
@ IFBA ROD 5
FIGURE 2.4 Secondary Source Rod Configuration E E E 5 s E E E E E E E E E E E E E E E 5 s E E E x
Secondary Sources s Secondary Source Rod e
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a FIGURE 2,5 Surnabla Absorber and Source Mod Locations i
R P N W L K J H 0 F E D C B A I
1 48I 321 481 481 841 4ggg 48I 321 , 2 321 48I 801 801 48I 321 3 481 80I 80! 481 , 4 32I 80I 801 801 801 801 801 321 -5 I 481 801 801 80! 481 48I 80I 801 80! 80! 481 -7 908 841 801 801 64I -8 481 80!' 801 80! 80! 481 -9 481 801 801 80! 481 - 10 l 32I 801 80! 801 80! 801 80! 321 - 11 481 801 801 48I 12 m
32I 48I 80I 801 48I 32I 13 32I 481 48I 841 481 321 14 33
. 15 08 TYPE TOTAL fifI..(NUW8ER OF IF8A R0DS)................. 4872 (SSA..(NUW8ER OF SECONDARY SOURCE RODLETS)... 8 7
1 1
4 1 .
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, l full flow conditions in the reactor coolant system to ensure compliance with Technical l Specification requirements.
SUMMARY
OF RESULTS.
For the hot, full flow condition (Tavg .>_ 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 f. 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 1.562 seconds.
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b TABLE 3.1 CONTROL ROD DASHPOT ENTRY TIMES CONTROL ROD DASHPOT ENTRY CONTROL ROD DASHPOT ENTRY LOCATION TIME (MSEC) _ LOCATION TIME (MSEC)
D02 1588 M08 1588 B12 1556 H06 1602 M14 1560 H10 1542 PO4 1582 F08 1592 B04 1600 K08 1584 j D14 1616 F02 1578 ,
P12 1576 B10 1624 !
M02 1564 K14 1528 G03 1542 P06 1570 C09 1540 B00 1544 J13 1544 F14 1600 N07 1532 P10 1554 C07 1558 K02 1528 G13 1558 H02 1570 N09 1556 808 1554 J03 1592 H14 1518 E03 1524 P08 1570 C11 1538 F06 1580 L13 1606 F10 1538 N05 1540 K10 1542 <
C05 1570 K06 1548 E13 1566 004 1554 N11 1556 M12 1500 LO3 1578 D12 1526 H04 1540 M04 1572 D08 1594 H08 1562 H12 1552 c SAMPLE Sl2E = 53 MEAN = 1.562 SIGMA = 0.026 2 SIGMA = 0.052 MEAN 2 SIGMA = 1.610 MEAN + 2 SIGMA = 1.614 Control rods in locations D14, B10 and M12 fell outside of the 2 SIGMA limit and were therefore dropped an additional six times. The drop times for the additional drops were all measured within the 2.7 second Tech. Spec. limit. A summary of the additional rod drop data is provided below.
MAXIMUM MINIMUM AVERAGE CONTROL ROD DASHPOT ENTRY DASHPOT ENTRY DASHPOT ENTRY LOCATION TIME (MSEC) TIME (MSEC) TIME (MSEC)
D14 1,600 1.552 1.585 B10 1.592 1.564 1.577 M12 1.564 1.494 1.527 9
4.0 INITIAL CRITICALITY PURPOSE I
The purpose of this test was to achieve initial rcactor criticality under carefully controlled conditions, establish the upper flux limit for the performance of zero power physics tests, and operationally verify the calibration of the reactivity computer.
SUMMARY
OF RESULTS I
initial reactor criticality for Cy0 e 3 was achieved by dilution at 0315 on May 7,1992. ;
I The reactor was stabilized at the following critical conditions: RCS temperature 556.6*F, intermediate ranDe power approximately 1 x 10 8 amps, RCS boron concentration 2089 ppm, and Control Bank D position at 187.5 (188/187) steps. Followin0 stabilizatiori, 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 I addition, source and intermediate range neutron channel overlap data were taken during !
the flur increase preceding initial criticality to demonstrate that adequate overlap existed, l
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i 5.0 ALL RODS OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT MEASUREMENT PURPOSE The objectives of these measurements were to determine the hot, rero power isothermal and mtiderator temperature coef ficients f t the all rods-out (ARD) configuration and to measure the ARO boron ondpoint concentration.
SUMMARY
OF RESULTS 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 +3.02 pcm/*F which meets the design acceptanco criteria. This gives a calculated moderator temperature coefficient of +4.81 pcm/'F which is wittyin the Technical Specification limit of + 7.0 pcm/'F at HZP. Thus, no rod withdrawalliriits were needed to ensure tho + 7.0 pcm/'F Iimit was mot.
Because of the high burnup core design of the Unit 2 Cycle 3 core,it was expected that the design acceptance criteria of .i. 60 ppm for the ARO critical boron concentration would not be met. As shown in Table 5.1, the measured ARO critical boro:1 concentration was 85 ppm below the predicted ARO critical boron concentration. The boron concentration corresponding to the 1000 pcm Technical Specification acceptanco criterion is approximately 134 ppm. Westinghouse conducted a review of the Reload Safety Evaluation (RSE) and determined that the RSE conclusions would remain valid using 134 ppm for the acceptance criterion as long as the remaining HZP physics test results mot the relevant design review criteria. Since all of the remaining HZP physics test results met tne relevant design review critoria, we conclude that the RSE remains valid.
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TABLE 5.1
)
ARO HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT Boron Measured ITC Design Calculated Rod Confiouration Concentration ITC Acceotance MTC All Rods Out 2078 ppm +3.02 pcmfF +3.47 pcmfF +4.81 pcmfF ITC - Isothermal temperature coefficient, includes -1.71 JcmfF doppler coefficient MTC - Moderator only temperature coefficient, normalized to the ARO condition ARO HZP BORON ENDPOINT CONCENTRATION Rod Conficuration Measured C. (com) Desian - credicted C. (com)
A!! Rods Out 2091 2176 12 9
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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 values predicted by design.
SUMMARY
OF RESQLTS .
The rod wt . measurements were performed using the bank interchange method in which: (1) tt 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 WGRTH MEASUREMENTS Bank Predicted Intearal Bank Measured Bank Worth Percent Difference Worth & Review Criteria (ocm) (Dem) focm)
Control A 268 1 100 209.2 -21.9 Control B 794 1 119 820.4 + 3.3 Control C 778 i 117 748.1 - 3.8 Control D 502 1 100 466.8 - 7.0 Shutdown A - 244 1 100 234.9 - 3.7 Shutdown B 957 i 96 996.0 *- 4.1 (Reference)
Shutdown C 456 1 100 447.1 - 2.0 Shutdowr. D 451 1 100 438.8 - 2.7 Shutdown E 452 1100 427.3 - 5.5 All Banks Combined 4902 + 490 4788.6 - 2.3 14
.- . . - . -. - . . _ =... - . - . . - . _ . . _ . - - - - _ . . - . . -- . . -
7.0 STARTUP AND POWER ASCENSION PURPQSE 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 ascemion to account for the effect of a low leakage core.
- 2. Performance 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 l l
SUMMARY
OF RESULTS Full core flux maps were obtained at plateaus of approximately 31 %,49 %,76 %
and 80% 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 evaluated at each plateau. An incore excore recalibration test was performed at 76% RTP. Reactor I coolant flow was determined from calorimetric measurements at 94.8% RTP.
I Delta T calibration constants were determined at 94.8% 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 Param Mao 67 Mao 68 Mao 62 Mao 71 Avg. % 31.2 48.9 75.9 80.3 ,
Power LOPAR 1.902 1.831 1.681 1.659 FDHN Limit VANTAGE 1.991 1.903 1.769 1.748 5 FDHN Limit s
LOPAR 1.4716 1.4418 1.4256 1.4197, FDHN Measured VANTAGE 1.6209 1.5829 1.5537 1.5454 5 FDHN Measured ,
Coro Avg. 4.0 6.0 4.0 3.3 AFD Avg. Coro 12.796 12.287 5.309 4.065
% A.O. ,
Most 2.1594 2.2008 1.7851 1.7830 Limiting FO(Z)
+2%
Transient 4,0702 4.2654- 2.3825 2.2032 FO Limit 16
. __ __ . _ _ . _ . _ _ _. . . . _ , . .. ... - ._ .~ _ _ .. _ .- _ . , ._ .- _,. _ . _ -
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l
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8.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT PURPOSE The purpose of this test was to determine the flow rate in each reactor coolant loop in order to confirm that the total cota flow met the minimum flow requirement 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 415,033 gpm. !
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