ML20246D971
| ML20246D971 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 08/09/1989 |
| From: | Bryant K, Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ULNRC-2064, NUDOCS 8908280355 | |
| Download: ML20246D971 (28) | |
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. Donald F.Schnell JlEt
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.U.S. Nuclear = Regulatory: Commission Document Control Desk m.
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- V DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 FACILITY OPERATING LICENSE NPF-30
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STARTUP REPORT The.cnclosed Startup Report is submitted fj..
pursuant'to. Sections 6.9.1.1, G.9.1.2,'and 6.9.1.3 of-the s
Callaway Unit'l Technical Specifications.
Very truly yours, o
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Donald F. Schnell DFS/JDB/WRC/bjp Hu,
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U.S. Nuclear Regulatory Commission Page.2 August 18, 1989-
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Mr. A. Bert Davis, Regional Administrator U.S. Nuclear. Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Mr. R. W. DeFayette Chief, Reactor Projects Section 3A U.S. Nuclear Regulatory Commission Region III 799-Roosevelt Road Glen Ellyn, IL 60137 Mr. T. W. Alexion Licensing Project Manager, Callaway Plant Office'of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13-E-21 Washington, DC 20555 Mr. B. H. Little Resident NRC Inspector, Callaway Plant Callaway Resident Office U.S. Nuclear Regulatory Commission RR #1' Steedman, MO 65077 Manager, Electric Department Missouri Public Service Commission P. O. Box 360 L
Jefferson City, MO 65102 4
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Distribution for ULNRC-2064 N. Date (S. L. Auston) (470)
E210.01-A160.761 DFS/ Chrono G. L. Randolph Licensing and Filel (A. C. Passwater/D. E. Shafer/
D. J. Walker) (470)
T. P. Sharkey NSRB (S. L. Auston) (470)
H. Wuertenbaecher (100)
O. Maynard (WCNOCl K. R. Bryant (CA-460)
J. D. Blosser (CA"460) 1 1
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I UNION ELECTRIC COMPANY CALLAWAY PLANT CYCLE 4 A
STARTUP REPORT AUGUST 9, 1989 1'
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l-FACILITY OPERATING LICENSE NPF-30 l.
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1 UNION ELECTRIC COMPANY CALIAWAY PLANT STARTUP REPORT Prepared by K. R. Bryant Submitted:
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Supervising Engineer, Reactor / Secondary Systems Ybh Superintend Systems Engineering
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l TABLE OF CONTENTS l
Pare 1.0 Introduction' 1
'2. 0 Core Refueling 2
3.0 Control Rod Drop Time Measurements 7
4.0-Initial Criticality 9
5.0 Determination of Low Power Physics Testing Range 10 5.1 Dynamic Checkout of R.eactimeter 11
' 5. 2 Boron End Point Measurements 12 5.3 Moderator Temperature Coefficient Measurements 13
.5.4
' Bank Reactivity Worth Measurements 14 6.0 Flux Mapping 15 7.0
- Alignment of Nuclear Instrumentation System 20 8.0 Reactor Coolant Flow Measurement 21 9.0 Core Reactivity Balance 22 mm.
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1.0 INTRODUCTION
This report presents a. summary of plant startup and power escalation testing for Refuel 3 at Union Electric's Callaway Plant.
It satisfies the requirement of. Technical Specifications that a Startup Report be submitted following installation of fuel having a different design.
Callaway Plant is a Westinghouse four-loop pressurized water reactor rated at 3565 MWt.
It is located in Callaway County, Missouri, approximately 80 miles west of the St. Louis metropolitan area, and is owned and operated by Union Electric Company.
Callaway Plant operated in Cycle 3.with a transition core consisting of 13 Westinghouse 17 x 17 low-parasitic (LOPAR) fuel assemblies, 84 Westinghouse 17 x 17 Reconstitutable Optimized fuel assemblies, and 96 Westinghouse Vantage 5 fuel assemblies.
The unit was refueled with Westinghouse Vantage 5 fuel.asseu-blies and the Cycle 4 core now consists of 9 Optimized fuel assemblies and 184 Vantage 5 fuel assemblies.
Cycle 4 fuel shuffle commenced on April 9, 1989 and was completed on May 4, 1989.
Initial criticality for the cycle occurred on May 21, 1989. The plant was synchronized to the grid on May 23, 1989.
The startup testing program was completed with the flux map at full power on June 1,1989, i
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4 2.01 CORE REFUELING (ETP-ZZ-00035)
OBJECTIVES To specify the final core configuration and to control _the L
unloading and loading sequences.
SUMHARY OF RESULTS Core loading was completed.
DISCUSSION Refueling was performed by completely offloading the Cycle 3 core to the Spent Fuel Fool, changing out fuel inserts, and then loading the_ Cycle 4 core. Ultrasonic inspection of the fuel was performed in parallel with.the core offload.
One fuel-rod was found to be leaking.
In addition, another fuel assembly was visually observed to have a torn grid strap. Attempts to repair these 2 fuel assemblies failed, so they and 2 symmetric assemblies were withheld from the Cycle 4 core.
Four fresh assemblies were obtained snd the core was redesigned to include the replacement assemblies.
The off-load began at 21:05 on 4-9-89, and the reload was completed at 08:49 on 5-4-89, for a total elapsed time of 24.5 days.
The as-built Cycle 4 core configuration is shown in Figures 2.1 through 2.4.
During the refueling, all 53 hafnium rod cluster control assemblies (RCCA) were replaced with Ag-In-Cd RCCA's.
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2 E13 E89lF67 F72 F15 F74 F27 F83 F84 E75 E25 j
5A 6B 6A SC 6A SC 6A SC 6A SC 6A 6B SA E16 F66 F30 E84 F46 E78 F48 E87 F3 E93 F59 F87 E28 SC 6A SC 6A 5B 6A SA 6A 5B 6A SC 6A SC-l 4
l E81 F4 E63 F39 E38 F41 E6 F35 E31 F37 E69 F26 E96 5
4B 6B SC 6A SA 6A 5B 6A 5B 6A 5A 6A SC 6B 4B D63 F68 E66 F29 E5 F1 E30 F45 E41 F24 E3 F6 E76 F78 D48 s
5B 6B 6A 5B 6A SA 6A SA 6A 5A 6A 5B 6A 6B SB E50 F79 F53 E49' F51 E22 F43 E14 F64 E17 F42 E45 F38 F75 E37 7
5B 6A SC 6A 5B 6A SC 6C SC 6A 5B 6A SC 6A 5B E57 F50 E80 F13 E54 F55 E83 F89 E74 F9 E34 F28 E79 F49 E35 g
SC 6B 6A SA 6A 5A 6C 4A 6C 5A 6A 5A 6A 6B SC E90 F76 F33 E10 FS2 E20 F90 D7 F91 E19 F54 E8 F63 F71 E63 F20:lE40 g
5B 6A SC 6A 5B 6A SC 6C SC 6A 5B 6A SC 6A SB E29 F10 E65 F62 E59 F7 E95 F92
,E82 F19 E33 Fil E92 5B 6B I 6A 5B l6A 5A 6A l SA 6A
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E55 F73 lF32 E46 lF31 E26 i F5 i E21. F47 !E27 [F36 E44 F56 F65 E36 i SA 6A 5B 6A SB 6A
!A 6A SC 6B 4B I
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6A 6A 5B ! 6A I 5A I 6A 5B l6A SC 6A SC 12 SC F17lSC E88 F44 E56 l F57 l E4 IF18 E42 'F14 E94 F2 i E73 E61 5A 6B 6A SC 6A SC 6A SC
}6A SC 6A 6B 5A 13 E24 F69 F23 E86 F61 E77 F58 E70 lF60 E91 F22 F86 E23 5A SC 6B 6B 6A 6B 6A 6B 6B SC 5A E15 E62 F85 F81 F21 F88 F40 F80 F77 E85 E18 l
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R e REGION NuuggR ID ID FUEL ASSEMBLY IDENTIFICATION
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4A - 3.4 W/0 SA - 3.6 W/0 6A - 3.6 W/0 4B - 3.8 W/0 5B - 3.8 W/0 6B - 4.0 W/o l
SC - 4.2 W/0 6C - 4.0 W/0 i
ITEL/ REGION LOCATIONS FIGURE 2.1 (
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SA B
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SA 3
SA D
SE D
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B C
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D A
SE C
9 SB SB 10 B
C A
C B
SE D
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SD 4
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Control and Shutdown Rod Locations FIGURE 2.2 _ _ - _
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MD MD MD TC TC TC TC TC TC 13 MD MD MD MD
- MD MD MD MD lTC TC TC TC TC TC 15 MD MD TC D DEG.
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TC = THERMDCDUPLE 50 i
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3 80 128 128 l{8 128 128 80
- 4 158 128 128 128 128
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80 128 128 128 128 128 80 3
80 128 128 128 128 128 128 80
'7 128 128 128 20 128 128-128 8
80 128 128 20 20 128 128 80 5
128 128 128 20 128 128 128 10 80 128-128 128 128 128 128 80 I
80 128
,128 128 128 l128 80 12 128-128 128 128 128 128 80 128 128 128 12S 128 80 13 SS
-14 80 80 128 80 128 80 80 15
.J Number of Burnable Absorber Rods SS Secondary Source Rods 10112 Fresh lFBA Rods 80 Fresh WABA Rods 3urnable Absorber and Source Assembly Locations FIGURE 2.4 6-
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3.0 CONTROL ROD DROP TIME MEASUREMENTS '
OBJECTIVES To measure the drop time of each Rod Control Cluster Assembly (RCCA) under. hot, full flow conditions in accordance with Technical Specification 3/4.1.3.4
SUMMARY
OF RESULTS Technical Specification 3/4.1.3.4 requires that the drop time for each RCCA from the fully withdrawn position be less than or equal to 2700. msec from the beginning of decay of stationary grip-per coil voltage to darhpot entry with Tavg greater than or equal to 551*F and all reactor coolant pumps operating. Under these con-ditions, the longest drop time was 1568 msec for rod D02.
A summary of all rod drop times is presented as Figure 3.1.
The wean drop time to dashpot entry was 1527 msec.
DISCUSSION The Westinghouse Automatic Rod Drop Test Set was used to measure rod drop times.
The test set consists of two Remote Units (one connected to each Digital Rod Position Indication (DRPI) cabinet inside containment) and a Central Unit (connected to the Rod Control Logic cabinet). A rod bank is withdrawn and the control circuit for one group of rods is interrupted from the Central Unit, causing the selected group of rods to drop. Analog signals created by the rods passing through the DRPI coils are fed into the Remote Units, where they are con-verted to digital signals and stored. Upon command, the digital data is transmitted to the Central Unit for display (or printout) of a graphic representation of the rod drop signals and associated drop times. Following completion of data transmission the process is repeate,d for the remaining withdrawn group and, subsequently, for each' bank.
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FIGURE 3.1
-ROD DROP TIME TALULATION
-i HOT, FULL FLOW; CYCLE 3 A
B C
D-E F
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O' 13 1556 1504 1514 1526 1548
'g 2196 2081 2134 2146 2198 1544 1536 1522 1524 2184 2166 2142 2154 12
'1539 1544 1512 1496 1538 2209 2194 2162 2126 2148 11 1538 1508 2168 2148 10 1516 1520 1524 1542 1496 2136 2160 2154 2192 2126 g_
1532 1522 2172 2142 270*
8 1488 1544 1542 1560 1522 1562 1500 l
90' 2108 2194 2182 2230
'2192 2212 2120 1524 1522 7
2144 21 ;2 6
1514 1530 1539 1562' 1498 I
2144 218n
???9 2202 2118 3
1536 1530 2186 _
2150 6
1544 1524 1502 1552 1534 2174 2164 2142 2202 2174 3
1544 1530 1524 1520 2174 2160 2144 2150 1
1568 1516 1494 1514 1530 210R 2]46 2114 2144 2]70 1
5 180*
xxxx TIME TO DASHPOT ENTRY (MSEC) xxxx TIME TO DASHPOT BOTTOM (MSEC) 8-
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4.0 INITIAL CRITICALITY-(ETP-ZZ-ST002)
OBJECTIVES To achieve initial criticality following refueling in a cautious and controlled manner.
SUMMARY
OF RESULTS Initial criticality was achieved on 5-21-89 at 0957.
DISCUSSION From an initial condition of all rods in and a boron concentration of 2166 ppm, the Sh'atdown and Control Banks were withdrawn to a final position of Control Bank D at.160 steps with all other banks fully withdrawn.
Inverse Count Rate Ratio (ICRR) plots were maintained during bank withdrawal.
Reactor Coolant System dilution was initiated at 60 gpm.
During dilution, the ICRR was plotted.
k' hen the ICRR reached 0.1, the dilution rate was reduced to 30 gpm and the ICRR was normalized to a value of 1.0.
k' hen the ICRR decreased' below 0.2, dilution was stopped; subsequent mixing brought the reactor.
critical.
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5.0 DETERMINATION OF LOW POWER PHYSICS TESTING RANGE (ETP-ZZ-0T003)
OBJECTIVES
-To determine the power level at which detectable reactivity feedback from fuel heating occurs and to establish the power
. range for low power physics testing.
SUMMARY
OF RESULTS Detectable reactivity feedback was observed at 3.0 E-7 amps as indicated by the Reactimeter.
The low power physics testing range was set at 1.0 E-8 to 1.0 E-7 amps.
DISCUSSION-With the reactor critical at a power level of approximately 1 E-8 amps (as indicated by the Intermediate Range channels),
approximately 30 pcm of positive reactivity was added by with-drawal of Control Bank D.
Power was allowed to increase until reactivity feedback effects were observed as indicated on strip chart recorders by an increase in Tavg and a decrease in reac-tivity.
The upper physics testing range limit was set at one-third of the current at which fuel heating occurred. The lower physics testing range limit was set at 100- times the background level.
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5.1 DYNAMIC CHECK 0NT OF REACT 1 METER (ETP-SE-ST002)
OBJECTIVES To verify accuracy of reactivity measurements obtained from the Reactimeter.
SUMMARY
OF RESULTS
'The absolute deviation between indicated and theoretical reactivity was determined to be 2.34%.
This inet the acceptance criteria of 4.0%.
DISCUSSION The Reactimeter is a cigital computer-based system for the on-line computation of dynamic reactivity during low power physics. testing.
For dynamic checkout of the Reactimeter, the controlling bank was withdrawn approximately-25 pcm while monitoring reac-tivity using the Reactimeter. The resulting reactor period was determined from the increasing flux level. The measured reactor period was used to calculate a theoretical reactivity which was compared with the reactivity measured by the Reactimeter.
This process was repeated for positive reactivity additions of approximately 50 and 75 pcm.
The average absolute deviation between indicated and theoretical reactivity was then calculated for comparison with the acceptance criteria. - _ _ _ _ _ - _ _ _ _ _.
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5.2 BORON ENDPOINT MEASUREMENTS
.(ETP-ZZ-ST004)
OBJECTIVES To measure the critical boron concentration at various-Control Bank configurations.
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SUMMARY
OF RESULTS All boron endpoint measurements met the design criteria, as summarized below.
Bank Measured Boron Acceptance Configuration Concentration (ppm)
Criteria (ppm)
All Rods Out 1701 1708 1 50 All Control Banks In 1315 1327 1 199 DISCUSSION Conditions were established with the controlling rod bank within 50 pcm of its endpoint configuration with the reactor critical in the Icw power physics testing range.
The control-ling bank was inserted or withdrawn as applicable to its endpoint configuration while monitoring reactivity. The changes in reactivity due to bank movement and Tavg deviation from Tref were converted to equivalent boron concentration units and used to correct the initial boron concentration, yielding the endpoint boron concentration.
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. 4 5.3' MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS p
(ESP-ZZ-00009) l
OBJECTIVES To measure the Isothermal Temperature coefficients (ITC) and determine the Moderator Temperature Coefficients.(MTC) under J.c various Control Bank configurations.
SUMMARY
OF RESULTS
. As required by Technical Specification 3/4.1.1.3, the MTC for the all rods out (ARO) configuration must be less positive than +5 pcm/*F. The subject MI6 was determined to be 3.82 pcm/"F.
The ARO ITC was determined to be 1.79 pcm/*F.
This met the acceptance criteria of 1.38 i 3 pcm/*F.
DISCUSSION The ITC measurement was performed.by first decreasing, then increasing Tavg using steam dumps and measuring the resulting i-reactivity changes. The ITC is the change in reactivity divided by the associated change in temperature.
The MTC was determined by subtracting the design Doppler Temperature Coefficient from the ITC.
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5.4' BANK REACTIVITY WORTH MEASUREMENTS (ETP-ZZ-STOC5)
OBJECTIVES l-To measure integral reactivity worth of.each Control Bank.
SUMMARY
OF RESULTS A11' design criteria were met, as summarized below.
Control
, Integral Acceptance Bank Worth (pcm)
Criteria (pcm)
D 678 719 i 72 C
928 943 i 94 B
867 866 i 87 A
637 656 i 66 DISCUSSION Integral bank worths were measured using the boron exchange method. Reactor Coolant System dilution was initiated at a constant rate.
The subject bank was periodically inserted in response to the change in reactivity caused by dilution.
The reactivity resulting from each incremental bank movement was measured using the Reactimeter and summed to yield integral bank worth. w__-._____-_______
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g 6.0 FLUX MAPPING (ETP-SR-ST001)
OBJECTIVES To. verify adequate flux cymmetry and power distribution during initial startup following refueling.
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SUMMARY
OF TEST RESULTS Low power' flux map results are presented as Table 6.1.
The Safety. Review Criteria that incore tilt is $1.04 was met.
The Design Review Criteria that incore tilt is $1.02 and the error between measured and predicted normalized enthalpy rise for each instrumented assembly is g 10% was also. net.
Intermediate power flux map results are presented as Table l
6.2.
The Design Review Criteria that the error between measured and predicted normalized enthalpy rise for each instrumented assembly is 5 10% was met.
The Design l',aview Criteria that incore tilt is 5 1.02 was also met.
Full power flux map results are presented as Table 6.3.
The Design Review Criteria that the error between measured and predicted normalized onthalpy rise for each instrumented assem-bly is $ 10% was met.
The Design Review Criteria that incore tilt is 5 1.02 was also met.
DISCUSSION A full core flux map was taken prior to criticality to check the operability of the Flux Mapping System.
At approximately 30 percent power a full core flux was taken. Safety review criteria for incore tilt $1.04 was met, as well as the design review criteria for incore tilt $1.02 and
$10% error between measured and predicted normalized enthalpy rise for each instrumented assembly.
At approximately 73% power the nuclear enthalpy rise hot channel factor and the heat flux hot channel factor were veri-fied.to be within the limits prescribed by Technical Specifica-tions.
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At approximately.100% power the following surveillance were performed:
1.
The absolute incore versus indicated axial flux difference-was verified to be'less than 3%.
.2.
-The nuclear entaalpy rise hot channel factor and the ha.at flux hot channel factor were verified to be.within the limits prescribed by Technical Specifications.
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LOW POWER FLUX MAP RESULTS Map.ID': 89-004 Date Performed:
5-25-89' Power Level:
26%
Cycle Burnup:
21.7 MWD /MTU
' Boron Concentration:
1421 ppm Rod Positioni CBD at 170~ steps Limiting Measured F :
'2.1461 Axial Location 31 Core Location L3 q
T.S. Limit:
4.2363' Limiting Measured F
- 8 AH T. S... Limit:
1.9430 Total Core Axial Offset:
6.315%
Maximum F Error::
6.5% Core Location C5 g
- Incore' Tilts:
QUADRANT UPPER CORE TILT LOWER CORE TILT 1
0.9990 0.9932 2
1.0057 1.0068 3
0.9968 1.0010 4
0.9985 0.9990 1 _ _ - _ - _ - _ _ _ _ _ _ _
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- INTERMEDIATE POWER F1DX MAP RESULYS
' Map ID: 89-005 Date Performed;
'5-27-89 Power Level:
~72%
Cycle Burnup:
77.8 MWD /MTU Boron'Conenatration:
1278 ppm Rod Position:
CBD at 208 steps Limiting Measured F :
2.0361 Axial Location 31 Core Location'L3 q
T.S. Limit:
2.9419 Lim'iting Measured FAH:
1.5527LCore Location L3 T.S. Limit-1.7231 Total Core Axial.0ffset:
6.743%
Meximum FAH Error:
5.2% Core Location L1 Incore Tilts:
OUADRANT UPPER CORE TILT LOWER CORE TILT 1
1.0044 0.9932 2
1.0012-1.0077 3
0.9999 1.0004 j
4 0.9945 0.9987 1
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FULL POWER FLUX MAP RESULTS Map ID: 89-006
'Date Performed:
6-1-8'9
- Power Level:
99%
H
- Cycle Burnup:
258 MWD /MTU Boron Concentration:
1208 ppm Rod Position:
CBD'at 215 steps
' Limiting Measured F :
2.0109 Axial Location 38 Core Location N11 q-q T.S. Limit:
2.1754 i
Limiting Measured F 1.5281 Core Location L3 l
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- T.S.' Limit:
1.592 Total Core Axial Offset: 4.657%
1 Maximum F rr r:
re ca i n All AH i
Incore Tilts:
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' QUADRANT UPPER CORE TILT LOWER CORE TILT l
1 1.0018 0.9973 2-1.0012 1.0035 l
3 0.9968 0.9987 4
1.0002 1.0005 i
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1 7.0 ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM (ETP-SE-ST001)
OBJECTIVES To generate alignment data for the Intermediate and Power l
Range detectors prior to criticality.
To verify adequate overlap between the Inte;; mediate. and Power Range channels.
To verify that initial calibration of the Intermediate and g
Power Range channels using calculated alignment data did not result in non-conservative trip setpoints.
To normalize Power Range detectors during power ascension.
SUMMARY
OF RESULTS The Nuclear Instrumentation System was set up for operation for power ascension and the rest of Cycle 4.
DISCUSSION A preliminary alignment of the Intermediate and Power Range channels was performed prior to criticality in anticipation of reduced currents and a mismatch between the Intermediate and Power Range channels (based on Cycle 3 alignment).
The preliminary alignment was based on calculations which multiplied Cycle 3 currents by the ratio of the sum of selected assembly powers (predicted) of Cycle 4 to the same (actual) from Cycle 3.
Intermediate and Power Range trip setpoints were monitored during power ascension to ensure that the initial alignment did not result in non-conservative setpoints.
The percent power indications of the Power Range channels were adjusted during power ascension based on secondary calorimetric..
A preliminary incore/excore calibration was performed at approximately 73% RTP. Another incore/excore calibration was performed at approximately 100% RTP.
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8.0 REACTOR COOLANT FIDW MEASUREMENT (ESP-BB-03015)
OBJECTIVES To determine the Reactor Coolant System (RCS) flow rate by precision heat balance measurements.
SUM 1ERY OF RESULTS The.RCS flow rate was determined to be 402,647 gpm.
This met the acceptance criteria that the RCS flow rate be > 382,630 gpm.
DISCUSSION The RCS flow rate measurement was performed at approximate-
~
1.00% RTP. The instrumentation used for determination of st.am pressure, feedwater temperature, and feedwater venturi delta-P was calibrated within seven days of performing the calorimetric.
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- 9 9.0 CORE REACTIVITY BA1ANCE
~(ESP-ZZ-00013)
OBJ ECTIVES To compare the overall core reactivity balance with pre-dicted values at full power.
SUMMARY
OF RESULTS The equivalent reactivity difference between measured and predicted boron concentration was 63?. pcm (80 ppm) which met the
-acceptance criteria of 1000 pcm, as required by Technical Specification 4.1.1.1.2.
DISCUSSION Under equilibrium conditions at approximately 100% RTP, the Reactor Coolant System boron concentration was corrected to yield the Hot Full Power, All Rods Out, Equilibrium Xenon / Samarium boron concentration for comparison with the predicted boron concentration.
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