ML20045H981
| ML20045H981 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/02/1993 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20045H980 | List: |
| References | |
| NUDOCS 9307220216 | |
| Download: ML20045H981 (16) | |
Text
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E GEORGIA POWER COMPANY VOGTLE ELECTRIC GENERATING PLANT UNIT 1, CYCLE 5 i
i STARTUP TEST REPORT P
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.i PREPARED BY:
R'lfACY GIN 8ER i
~ / 6 2 f3 REVIEWED B.
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(AlEACTUk ENDf i
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/ 6!l D APPR.OVED BY:
IfEACTOR ENGINEERING SUPERVI'SOR j
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9307220216 930716 Mi E
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PDR ADDCK 05000424 P
a TABLE OF CONTENTS j
i SECTION DESCRIPTION PAGE.
i 1.0 Introduction 3
2.0 Unit 1 Cycle 5 Core Refueling
-3 3.0 Control Rod Drop Time Measurement 4
i 4.0 Initial Criticality 4
5.0 All Rods Out Isothermal Temperature Coeflicient and Boron Endpoint Measurement 5
.i 6.0 Control and Shutdown Bank Wonh Measurements 6
7.0 Startup and Power Ascension 6
8.0 Reactor Coolant System Flow Measurement 7
r Table 1 Control Rod Dashpot Entry Times 8
Table 2 Summary of Control and Shutdown Rod Bank Worth Measurements 9
Table 3 Summary of Power Ascension Flux Map Data 10 Figure 1 Unit 1. Cycle 5 Reference Loading Pattern 11 Figure 2 Control Rod Locations 12 Figure 3 Burnable Absorber Configurations 13 Figure 4 Secondary Source Rod Configuration 14 Figure 5 Burnable Absorber and Source Rod Locations 15 1
Figure 6 Inverse Count Rate Ration Versus Boron Concentration 16 t
e 2
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1.0 INTRODUCTION
The Vogtle Electric Generating Plant 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.
1 Unit 1 is a four loop Westinghouse PWR which was uprated this cycle from 3411 MWth to 3565 AfWth. The cycle 5 core loading consists of 19317 x 17 fuel assemblies.
Unit 1 began commercial operations on May 31,1987 and has completed the first four cycles with the following average burnups:
Cycle 1 Completed 10/08/88 15,852 MWD /MTU Cycle 2 Completed 02/23/90 15,789 MWD /MTU Cycle 3 Completed 09/15/91 18,504 MWD /MTU Cycle 4 Completed 03/13/93 17,856 MWD /MTU e
One IIundred Fifty Six of the 193 assemblies comprising Cycle 5 are based upon the r
VANTAGE 5 design. Two of the VANTAGE 5 assemblies contain a total of 24 demonstration rods clad in ZIRLO (reference Technical Specification 5.3.1) and are located in core locations D14 and M2 for this cycle. One fuel assembly, at location F7, has one stainless steel pin in the place of a leaking fuel pin which was removed this outage.
2.0 UNIT 1 CYCLE 5 REFERENCES Westinghouse WCAP-13607, The Nuclear Design Report for the Voetle Electric Generating Planj
SUMMARY
i i
The ofiload of the Cycle 4 core commenced on March 25,1993 and was completed on March 28,1993. Sipping of all ofIloaded assemblies was performed to detect assemblies which were suspected to be leaking. Assembly 5F22 and assembly SE34 were determined to have leaking fuel rods. Assembly 5F22 was reconstituted on March 30,1993, using a stainless steel rod to replace the leaking fuel rod. Assembly 5E34 was discharged to the speht fuel pool.
3
Core reload commenced on April 5,1993 and was completed on April 11,1993. The as-loaded Cycle 5 core is shown in Figures 1 through 5 which give the location of each fuel assembly and insert. The Cycle 5 core has a nominal design lifetime of 20,900 MWD /MTU and consists of 8 Region 2 (Unit 2 EOC 1),9 Region 3 (Unit 1 EOC 2),12 Region 5A,8 Region 5B,56 Region 6A,16 Region 6B,52 Region 7A, and 32 Region 7B -
assemblies. Fuel assembly inserts consist of 53 fulllength control rods, two secondary sources, and 138 thimble plugs. Assemblies in Regions 7A and 7B have a total of 8288 Integral Fuel Burnable Absorber rods, the pellets ofwhich are coated for a total length of 129 inches and offset upwards 1.5 inches from the center of the fuel stack.'
3.0 CONTROL ROD DROP TLME 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 3.1.3.4 requirements.
SUMMARY
OF RESULTS 5510F and all reactor coolant pumps operating),
For the hot, full flow conditions (T vg > hat the rod drop time from the fully wi a
Technical Specification 3.1.3.4 requires t position shall be < 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 mean drop time was determined to be 1.563 seconds. The rod drop time results for dashpot entry are presented in Table 1.
4.0 INITIA.L CRITICALITY PURPOSE The purpose of this test was to achieve initial reactor criticality under carefully controlled conditions, validate the basis of the shutdown margin curves for Modes 3,4, and 5, establish the power range for performance oflow power physics testing, and operationally verify the calibration of the reactivity computer.
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8 4
SUMMARY
OF RESULTS-Initial reactor criticality for Unit 1 Cycle 5 was achieved, by dilution, at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> on April 24,1993. The reactor was stabilized at the following critical conditions: RCS temperature 557 0F, Intermediate Range power of approximately 1 x 10-8 Amps, RCS boron concentration of 1881 ppm, and Control Bank D at 185/184 steps. Following stabilization, the point of adding nuclear heat was determined and a checkout of the reactivity computer was successfully accomplished. In addition, Source and Intermediate l
Range Nuclear Instrumentation overlap data were taken during the flux increase preceding initial criticality to demonstrate that adequate overlap existed.
ICRR data were obtained during the dilution and plotted to determine if the unit met the Boron Dilution Accident Analysis assumptions. The graph oflCRR versus boron (ppm)is shown in Figure 6. The results indicated that the ICRR values obtained during the dilution were not bounded by the values assumed in the analysis. Westinghouse has evaluated the impact on the design and safety analyses and provided a safety evaluation. In addition, Westinghouse is providing revised Required Shutdown Margin curves which will be incorporated into the Core Operating Limits Report.
5.0 ALL RODS OUT ISOTIIERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT MEASUREMENT PURPOSE The objectives of these measurements were to determine the hot, zero power isothermal temperature and moderator coefficients for the all rods out (ARO) configuration and to measure the boron endpoint concentration.
SUMMARY
OF RESULTS The isothermal temperature coefficient was measured to be + 3.065 pcm/0F which met the design acceptance criteria of being within 2 pcm/0F of+ 3.21 pcm/0F. This gave a calculated moderator temperature coefficient of+ 5.015 pcm/0F which included a Doppler coeflicient of-1.55 pcm/0F and a burnup dependent correction factor of 0.4 pcm/0F and was within the Technical Specification limit of + 7.0 pcm/ OF. Therefore no rod withdrawal limits were required to be established.
The measured ARO critical boron concentration of 1901 ppm was within the design review criterion of150 ppm of the design value of 1891 ppm.
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6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS I
PURPOSE The objective of the bank worth raeasurements was to determine the integral worth of each control and shutdown bank for comparison with design values.
SUMMARY
OF RESULTS The rod worth measurements were performed using the bank interchange (rod swap) method in which the worth of the bank having the highest design worth (the Reference Bank) is measured using the standard boron dilution method and the remaining control and shutdown banks worths are derived from the reference bank reactivity needed to offset full insenion of the bank under test.
The control and shutdown bank worth measurement results are given in Table 2. The reference bank measured value of 780 pcm was not within the design review criteria ofi 10% of the predicted value of 892 pcm (- 12.6%). However, the measured value was within the safety review criteria ofi 15% of the predicted value. Westinghouse has eva, m.ed the impact of the failure to meet the reference bank worth to meet the design review criterion and determined that the current safety analysis contained in the Reload Safety Analysis and the data contained within the Nuclear Design Report are still applicable.
The reason for the error in reference bank worth is beleived to be misprediction of radial power distribution. Potential causes of power distribution differences are being investigated.
7.0 STARTUP AND POWER ASCENSION
_P_URPOSE The purpose of the power ascension program was to provide direction for Intermediate and Power Range detector calibration prior to startup and as needed during power ascension and to perform measurements and provide calibration data for incore-excore detector axial flux difference, core hot channel factor, Reactor Coolant System AT, and Reactor Coolant System flow at specified power level plateaus. In addition, Unit I was uprated this cycle from 3411 MWth to 3565 MW -
th 6
W
+
SUMMARY
OF RESULTS Full core flux maps were obtained at 28.7%, 50.8%, and 80.6%. Hot channel factors were evaluated at each power plateau and the results are shown on Table 3. Reactor Coolant Flow was determined from precision calorimetric measurements at 94.6% power.
Calibration constants for AT were determined at 95.5% power.
8.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT PURPOSE The purpose of this test was to determine the flow rate in each reactor coolant loop to confirm that the total flow met the minimum flow requirement given in Technical Specification 3.2.5.
S_UMMARY OF RESULTS The total flow was determined to be 396,864.5 gpm which is > the Technical Specification minimum limit of 384,509 gpm.
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TABLE 1 i
CONTROL ROD DASHPOT ENTRY TIMES CONTROL DASHPOT CONTROL DASHPOT ROD ENTRY ROD ENTRY LOCATION TIME (msec)
LOCATION TIME (msec)
D02 1592 M08 1532 B12 1596 H06 -
1522 M14 1598 H10 1546 PO4 1552 F08 1594 B04 1594 K08
'1546 D14 1610 F02 1560 P12 1578 B10
-1570 M02 1604 K14 1576 G03 1522 P06 1534 C09 1556 B06 1582 J13 1572 F14 1556 N07 1530 P10 1570 C07 1556 K02 1580 G13 1562 H02 1558 N09 1535 B08 1578' J03 1534 H14 1540 E03 1566 P08 1548 Cll 1590 F06 1546 L13 1544 F10 1548-NOS 1546 K10 1562 C05 1540 K06 1580 E13 1558 D04 1544 N11 1568 M12 1538 LO3 1602 D12 1556 H04 1570 M04 1554 DOS 1582 H08 1580 H12 1594 Sample Size = 53 Mean = 1.563 seconds 2a Limits = 1.519 and 1.607 seconds Control rod D14 fell outside of the 2a limits with a drop time of 1.610 seconds and was dropped an additional six times. The drop times for the extra drops were measured to be between 1.572 and 1.606 seconds which fell within the 20 limits and were consistent with the drop times of the other rods.
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4 TABLE 2
SUMMARY
OF CONTROL AND SHUTDOWN ROD BANK WORTH MEASUREMENTS PREDICTED INTEGRAL MEASURED BANK WORTH AND
' BANK REVIEW CRITERIA WORTH PERCENT BANK (ocm)
(ocm)
DIFFERENCE Control A 357 100 360.6
+ 1.0 Control B 696 104 607.3
- 12.7 Control C 775ii16 744.7
- 3.9 Control D 4841100 507.7
+ 4.9 Shutdown A 287i100 278.6
- 2.9 Shutdown B 892189 780
- 12.6 (Reference)
Shutdown C 409 100 384.9
- 5.9 Shutdown D 4061100 380.9
- 6.2 Shutdowm E 436 i 100 391.4
- 10.2 All Banks 4742 + 474 4436.1
- 6.5 9
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4 TABLE 3
SUMMARY
OF POWER ASCENSION FLUX MAP DATA -
Parameter Map 154 Man 155 Man 156
^
Average % Power 28.7 50.8 80.6 LOPAR FN Limit 1.859 1.759 1.621 AII LOPAR FN Measured 1.1330 1.1273 1.1304 AII VANTAGE 5FN Limit 2.003 1.894 1.746 AH VANTAGE 5FN Measured 1.5386 1.5566 1.5478 i
All Transient FQ Limit 3.5210 3.5814 2.2426 Most Limiting F (Z) + 2%
2.0812 1.9275 1.8473 Q
Core Average AFD (%)
3.2 1.2 0.6 Core Average Axial Offset (%)
11.147 2.392 0.795 f
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s FIGURE 1 UNIT I CYCLE 5 REFERENCE LOADING PATTERN i
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Sg4 5001 5058 5039 SC,12 5028 SF40 5034 Sept 5042 5057 5002 5g1 E-10 FEED FEED FEED P-8 FEED J-7 FEED 8-8 FEED FEED FEED 'L-10 W88 5082 SF31 SFS2 5044 SF09 5050 SF38 SF49 5083 WS7 lSis FEED D-8 E-13 FEED C-13 FEED L-13 H-12 FEED N E80 5019 5051 SF35 5087 SF11 5072 SF32 5874 SF01 5080 SF41 5038 5006 KS6 N-9 FEED FEED C-11 FEED P-5 FEED F-8 FEED 8-5 FEED N-11 FEED FEED C-9 SF54 5020 5(jjt5 5052 SF02 SF18 SF21 5881 SF19 SF13 SF04 5040 Srps 5008 SF30 G-11 FEED K-2 FEED L-2 H-14 J-14 FEED G-14 P-8 E-2 FEED F-2 FEED J-11 SE45 5035 5022,5F58' 5088 SF23 5084 SF48 5088 SF22 5077 5038 5024 K25 A-9 FEED FEED /N FEED B-7 FEED L-5 FEED P-7 FEED FEED FEED R-9
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SF55 5015 5p 5047 SFOS SF24 SF14 5079 SF17 SF18 5F07 5048 Sp 5018 SF28 4-5 FEED K-14 FEED L-14 5-8 J-2 FEED 0-2 H-2 E-14 FEED F-14 FEED J-5 K58 5013 5045 SF42 5070 SF12 5089 SF48 5083 SF03 5082 SF47 5448 5014 KS8 ~"
N-7 FEED FEED C-5 FEED P-11 FEED K-8 FEED 8-11 FEED N-5 FEED FEED C-7
<5g 5055 SF37 SF34 5041 g 5F05 g 5043 SF27 SF25 5000 #72, SW FEED H-4 E-3 FEED N N-3 75e35 FEED L-3 W-8 FEED M g2, 5003 5081 5037 5(pt 5025 SF45 5027 SC,02 5049 5058 5004 5g1
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si 5009 5011 5021 5459 5033 5005 5017 gl fj k/ FEED TEED FEED FEED FEED FEED FEED ' ett K '
SE72 SF38 SE11 SE14 SE44 5F39 K51 ASSEWBLY FROW CYCLE 2 0-13 E-7 G-1 H-1 J-1 L-7 J-13 REGIDN 2 ASSEleLIES AME FROM v007tE trilf 2 CYCLE 1 g
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SA 14 15 ABSORBDt IAATDtI AL: AG-lM-CD 00 BANK NUWBER OF BANK NUWBER OF IDENTIFIER LOCATIONS IDENTIFIER LOCATIONS A
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1 48I 48I 1041 104I 104I 481 48I 2
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- 11 104I 128I 128I 1041 12 1041 1281 104I 4SSA 104I 128I 1041 13 48I 48I 1041 1041 104I 48I 48I 14 15 00 TYPE TOTAL fffl..(NUWBER OF IFBA R0DS).................
8288 (SSA..(NUMBER OF SECONDARY SOURCE RODLETS)...
8 15 j
t FIGURE 6 INVERSE COUNT RATE RATIO VERSUS BORON CONCENTRATION ICRR for Vogtle Unit 1 Cycle 5 12 1.e 0.8- --~--------1.~~~---~~~-t-~~-~---~~t--~~~~-~-t.----
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$0 lb lb 200 250 360 32 Boron Charige from Critical (ppm) 4' N-31 m N-32 BDMS Umit 1
I 16
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