ELV-00318, Cycle 2 Startup Test Rept
| ML20246H429 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/13/1989 |
| From: | Bockhold J, Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 0191E, 191E, ELV-00318, ELV-318, NUDOCS 8903200172 | |
| Download: ML20246H429 (22) | |
Text
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l GEORGIA POWER COMPANY l
VOGTLE ELECTRIC GENERATING PLANT UNIT NUMBER 1, CYCLE 2 STARTUP TEST REPORT l
PREPARED BY SITE REACTOR ENGINEERING GROUPt EDITED BY D.R. MARNON-APPROVED BY A.G. RICKMAN j i;
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TABLE OF CONTENTS 1.0 Introduction 2.0 Unit 1 Cycle 2 Core Refueling 3.0 Control Rod Drop Time Measurement ,
l 4.0 Initial Criticality 5.0 All-Rods-Out Isothermal Temperature Coefficient and Boron Endpoint 6.0 Control Bank Worth Measurement 7.0 Startup and Power Ascension Procedure
- 8.0 Reactor Coolant System Flow Measurement 9.0 Incore-Excore Detector Calibration
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1.0 Introduction The Vogtle Electric Generating Plant Unit 1 Cycle 2 Startup Test Report summarizes results for some of the tests performed following core refueling. The report provides a brief synopsis of each test and gives a comparison of measured parameters with design predictions, Technical Specifications, or values assumed in the FSAR safety analysis.
Unit 1 of the Vogtle Electric Generating Plant is a four loop Westinghouse pressurized water reactor rated at 3411 MWth. The Cycle 2 core loading consists of 193 17 x 17 fuel assemblies.
Unit 1 began commercial operations on May 31, 1987 and completed Cycle 1 on October 8, 1988 with an average core burnup of 15852 MWD /MTU.
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2.0 Unit 1 Cycle 2 Core Refueling REFERENCES
- 1. Westinghouse WCAP 11980 (The Nuclear Design and Core Physics Characteristics of the Alvin W. Vogtle Unit 1 Nuclear Power Plant Cycle 2)
SUMMARY
Unloading of the Cycle 1 core into the spent fuel pool commenced on 10/18/88 and was completed on 10/23/88. Visual examination of the Cycle-1 fuel assemblies unloaded into the spent fuel pool showed no damaged assemblies.
Core reload commenced on 10/27/88 and was completed on 11/03/88. The as-loaded Cycle-2 core is shown in Figures 2.1 through 2.4, which give the location of each fuel assembly and insert, including burnable poison insert location and configurations. The Cycle-2 core has a nominal design lifetime of 15,600 MWD /MTU and consists of 45 Region 2 assemblies, 64 Region 3 assemblies and 84 Region 4 assemblies.
Fuel assembly inserts consist of 53 full length control rod clusters, 60 wet annular burnable absorber inserts (WABAs),
two secondary sources and 78 thimble plug inserts.
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' Figure 2.1 Unit 1 Cycle 2 Reference Loading Pattern l
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R P N M L K J H G F E D C B A 1
2 3
4 5
6 7
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9 10 11 12 13 14 15 o'
COLOR CODE LEGEND.
REGION '
2 3 4 FROM CYCLE 1 1 FEED w/o U-235 2.601 3096 3405 L -- -- _ ---- - - - - - _ - - - - - - - - - - - -
l-Figure 2.2 CONTROL ROD LOCATIONS l R P N M L K .J H G F E D C B A l 1
l 5 SC SD 8 B C A C B 7 SB S8 ,
l 8 so* C SE A D A SE C 3 9 SB S8 y 1
10 B C A C B l.
11 SD SC 12 SA D SE D SA 13 SC S8 S8 SD i
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15 l Absogrgterial: !
o-l CONTROL BANKS LWER SHUTDOWN BANKS YbER[
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j A 4 S8 8 J l
TOTAL 25 SA 8 TOTAL 28
Figure 2.3 BURNABLE ABSORBER AND SOURCE ASSEMBLY LOCATIONS R P N M L K J H G F E D C B A k
1 2 SS 3 4 8 4 8 4 4 4 12 8 8 12 4 5 12 8 8 8 12 6 8 8 8 8 8 8 7 8 8 4 8 8 t
8 so' 4 8 4 4 8 -4 'i I
9 8 8 4 8 8 10 8 8 8 8 8 8 11 12 8 8 8 12 12 4 12 8 8 12 4 13 4 8 4 8 4 4
14 SS 15 o' l 1
- Number of WABAs 448 WABAs in g
g SS Secondary Source 80 Clusters
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1 Figure 2.4 )
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. BURNABLE ABSORBER AND SECONDARY SOURCE ROD CONFIGURATIONS I
,O O 0, , O O O, O O O O O O O E O O O O O O O E E O O O O O O O O E O O E E E E O O O O O O .
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4 BA Configuration 8 BA Configuration I
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B u O 0, 0 O E 5 O E O E O O O O O O E E O O O O O E O E O E O O O O O E O E O O O j 12 BA Configuration Secondary Source Rods I
3.0 CONTROL ROD DROP TIME MEASUREMENT PURPOSE The purpose of this test was to measure the' drop time of all full length control rods under hot-full flow conditions- a in the reactor coolant system to insure ~ compliance with !
Technical Sp' edification requirements.
SUMMARY
'OF RESULTS For the hot-full flow condition ( T > 551 F'and all reactor coolant pumps operating) TechniE!al'" Specifications 3.1.3.4 requires that the rod drop t'ime from the fully-withdrawn position.shall be $ 2.2 seconds'from the beginning of' stationary gripper coil voltage decay until ' dashpot entry. - ,
All full length rod drop times were measured to be less than l 2.2 seconds. The rod drop time results for both dashpot entry i and dashpot bottom are presented in Figure 3.1. 'Mean drop i times are summarized below- l l
TEST MEAN TIME TO MEAN TIME TO l CONDITIONS DASHPOT ENTRY DASHPOT BOTTOM I l
Hot-full Flow 1.452 sec 1.941 sec l
To confirm normal rod mechan' ism operation prior to conducting the rod drops, the Rod Control System Functional Test (53005-C) was performed. In the test, the Control Rod Drive Mechanism' full withdrawal and operability'were checked.
l The functioning of the Digital Rod position indicator and the bank overlap unit were checked. All results were satisfactory.
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Vogtle Unit 1 Cycle 1 And 2 Hot Rod Drop Times 1 Table 3.1: Comparison Of Vogtle Unit 1 Cycle 1 And 2 Drop Times DASHPOT ENTRY (MBEC) TURNAROUND TIME (MBEC)
RCCA CYCLE 1 CYCLE 2 DIFFERENCE
- CYCLE 1 CYCLE 2 DIFFERENCE
- D02 1462 1444 -18 1962 1914 -48 B12 1464 1450 -14 1974 1910 -64 !
M14 1478 1450 -28 2008 1930 -78 P04 1444 1450 +6 1924 1900 -24 ,
BO4 1454 1422 -32 1964 1882 -82 D14 1462 1444 -18 1942 1904 -38 l P12 1474 1438 -36 2004 1928 -76 i M02 1480 1456 -24 1920 1926 +6 G03 1444 1428 -16 1924 1898 -26 C09 1438 1472 +34 1938 1932 -6 J13 1438 1446 +8 1988 1906 -82 N07 1466 1456 -10 1926 1916 -10 C07 1446 1444 -2 1946 1964 +18 l l
G13 1452 1434 -18 1942 1924 -18 N09 1414 1456 +42 1964 1956 -8 J03 1424 1440 +16 1894 1940 +46 E03 1458 1444 -14 1918 1914 -4 C11 1464 1450 -14 1944 1960 +16 L13 1456 1492 +36 1996 1952 -44 ;
N05 1448 1472 +24 1938 1902 -36 C05 1444 1435** -9 1934 Multiple *** -
i E13 144d 1438 -10 1958 1968 +10 N11 1432 1446 +14 1952 1916 -36 LO3 1476 1478 +2 1936 1988 +52 l i
H04 1434 1440 +6 1914 1930 +16 '
l D08 1438 1454 +16 1918 1964 +46 i 9
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H12 1428 1432 +4 1948 1932 -16 1 M08 1446 1444 -2 1916 1904 -12 i H06 1410 1420 +10 1910 1920 +10 H10 1444 1450 +6 1924 1990 +66 F08 1430 1444 +14 1930 1914 -16 )
i K08 1446 1456 +10 1926 1926 0 l
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Vogtle Unit 1 Cycle 1 And 2 Hot Rod Drop Times Table 3.1: Comparison Of Vogtle Unit 1 Cycle 1 And 2 Drop Times DASHPOT ENTRY (MBEC) TURNAROUND TIME (M8EC)
RCCA CYCLE 1 CYCLE 2 DIFFERENCE
- CYCLE 1 CYCLE 2 DIFFERENCE
- F02 1428 1470 +42 1868 1970 +102 B10 1444 1458 +14 1934 1948 +14 K14 1482 1418 -64 1992 1938 -54 P06 1420 1472 +52 1870 1912 +42 B06 1476 1414 -62 1936 1924 -12 F14 1478 1450 -28 1968 1960 -8 P10 1414 1464 +50 1904 1964 +60 K02 1428 1420 -8 1848 1920 +72 H02 1402 1424 +22 1902 1914 +12 B08 1454 1490 +36 1914 1980 +66 H14 1442 1490 +48 1972 2040 +68 l P08 1424 1456 +32 1914 1976 +62 F06 1432 1488 +56 1902 1958 +56 l l
F10 1434 1498 +64 1904 2018 +114 I K10 1414 1516 +102 1934 2006 +72 K06 1448 1464 +16 1868 1994 +134 D04 1428 1440 +12 1918 1910 -8 l M12 1432 1420 -12 1892 1950 +58 D12 1428 1426 -2 1898 1936 +38 M04 1428 1446 +18 1918 1966 +49 H08 1488 1470 -18 2028 1940 -88 AVERAGE 1445 1452 1933 1941 ****
STD. DEV. 40 22.4 36.7 33.5
- Positive means Cycle 2 time was greater than Cycle 1 time.
- Average of six rod drops.
- Multiple drops; no single turnaround time used.
- The almost equal increase in drop times and turnaround times was accompanied by a decrease in standard deviation.
NOTE: Cycle 1 times were taken from Initial Startup Procedure 1-5SF-04.
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4.0 INITIAL CRITICALITY i PURPOSE The purpose of this test was to achieve initial reactor i criticality under carefully controlled conditions, establish .
the upper flux limit for the conduct of zero power physics tests, and operationally verify the calibration of the f reactivity computer.
SUMMARY
OF RESULTS Initial Reactor Criticality for Cycle 2 was achieved during dilution mixing at 0320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> on November 26, 1988.
The reactor was allowed to be stabilized at the following i critical conditions: RCS temperature 558 *F, intermediate range power 1 x 10-8 amps, RCS boron concentration 1880 ppm, and Control Bank D position 150 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 l addition, source and intermediate range neutron channel l l overlap data were taken during the flux increase preceding l i
initial criticality to demonstrate that adequate overlap I existed. i l
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m 5.O ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT 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 bo +2.55 pcm/ *F which meets the design acceptance criteria. This gives a' calculated moderator temperature coefficient of +4.75 pcm/
- F which' is ' within the Technical-Specification limit of'+7.0 pcm/*F at HZP. Thus, no rod withdrawal limits were needed to ensure the +7.0 pcm/
- F limit.
was met. The design acceptance criterion for the ARO critical boron concentration was satisfactorily met.
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. i 6.0 CONTROL BANK WORTH MEASUREMENTS l
PURPOSE l
The objective of the bank worth measurements was to '
determine the integral reactivity worth of-each control bank for comparison with the values predicted by design.
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SUMMARY
OF RESULTS 1
The rod worth measurements were performed using the !
procedure 53002-C which specifies use of the boron dilution method.
l The control bank worth measurement results_are given in j 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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7.0 POWER ASCENSION AFTER REFUELING PROCEDURE (55019)
PURPOSE-The purpose of this procedure was to provide controlling instructions for:
- 1. NIS intermediate and power range calibration as required prior to start up and during startup.
power ascension to take into account the effect of-a low leakage core..
I Conduct of startup and power ascension testing, to' I 2.
include:
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- a. HZP reactor physics tests (53002-C) _
- b. reactor coolant system flow measurement . (54014-C)
- c. core hot channel factor surveillance (54004-C) !
- d. incore-excore'AFD channel calibration (55003-C)
SUMMARY
OF RESULTS I Full core flux maps were obtained at about 30%, 50%,
75% and 100% RTP.. Hot Channel factors were evaluated at l each power plateau and are shown in Table 7.1. The incore ;
and excore delta-I were also evaluated at each plateau. !
Reactor coolant flow was determined from a series of-three I calorimetric measurements at just below 75% RTP. An incore- )
excore recalibration test was performed at 100% RTP. I I
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SUMMARY
OF POWER ASCENSION FLUX MAP DATA Param Mao 67 Mao 68 Mao 69 Mao 70 Avg. % 30 49 75 100 Power .l 1
FDHN 1.872 1.788 1.6680 1.5520 l Limit i
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FDHN 1.4778 1.469 1.4804 1.4699 l Measured .q i
Core 1.4 3.3 0.2 4.0 Avg. AFD I I
Avg. 4.639 6.699 0.319 3.999 ,
Core % )
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- Maximum 2.1310 2.0363 1.9179 1.8324 l l FQ(Z) j FQ Limit 4.5652 4.5536 3.0645 2.2459 I l
8.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT PURPOSE The purpose of this procedure was to determine the' flow rate in each reactor coolant loop in order to confirm that the total core flow met the minimum flow requirement given, ,
in the Unit 1 Technical Specifications.
SUMMARY
OF RESULTS To comply with the Unit 1 Technical Specifications, the total reactor coolant system flow rate determined at normal operating temperature and pressure must equal or exceed 396,198 gpm for four loop operation.
From the average of 3 calorimetric heat balance measurements, the total core flow was determined to be 408,562 gpm, which meets the above criterion.
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9.0. INCORE-EXCORE= DETECTOR CALIBRATION PURPOSE The objective of this procedure was to determine the relationship between power range upper and lower excore i detector currents and incore axial offset for the purpose of calibrating the control board and plant computer axial flux difference (AFD) channels, and for calibrating the delta flux penalty to the over. temperature delta-T protection system.
SUMMARY
OF RESULTS At approximately 100% power, a full-core base case' flux map and three quarter core flux maps were run at various positive and negative axial offsets to develop equations' relating detector current to. core axial offset. To reduce-error, all flux maps were performed at the same RCS temperature. The power range NIS channels were adjusted.to incorporate the revised calibration data.
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DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST-2 .!
CHANNEL'N411 I-Top = .1.705 A.O. + 242.0 yA.
I-Bottom = -1.777 A.O. + 255.7 #A CHANNEL N42:
I-Top = 1.688 A.O. + 233.9 #A I-Bottom = -1.610 A.O. + 251.0 pA CHANNEL N43:
I-Top = 1.685 A.O. + 249.4 #A I-Bottom = -1.742 A.O. + 263.4 yA CHANNEL N44:
I-Top = 1.637 A.O. + 234.7 pA l I-Bottom = -1.694 A.O. + 253.2 yA 1
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ELV-00318 March 13, 1989 0191e U. S. Nuclear Regulatory Commission ATTN: Document Control Desk l Washington, D. C. 20555 PLANT V0GTLE - UNIT 1 NRC DOCKET 50-424 OPERATING LICENSE NPF-68 UNIT 1, CYCLE 2 STARTUP TEST REPORT Gentlemen:
By letter VL-102, dated October 25, 1988, Georgia Power Company committed to provide a Startup Test Report summarizing results of some of the tests performed following core refueling for Unit 1 Cycle 2. The enclosed report includes a core loading map and summaries of the following measurements:
- Control rod drop time;
- Critical boron concentration;
- Control rod bank worth;
- Moderator temperature coefficient;
- Startup power distribution using the incore flux mapping system.
Should you have any questions, please contact this office at any time.
Sincerely, W. k V W. G. Hairston, III NJS/gm Enclosure xc: Georgia Power Company Mr. P. D. Rice Mr. C. K. McCoy Mr. G. Bockhold, Jr.
6 Mr. J. P. Kane r l NORMS U. S. Nuclear Regulatory Commission Mr. S. Ebneter, Regional Administrator Mr. J. B. Hopkins, Licensing Project Manager, NRR (2 copies)
Mr. J. F. Rogge, Senior Resident Inspector - Operations, Vogtle