ML20199G827
| ML20199G827 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 12/31/1998 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20199G790 | List: |
| References | |
| NUDOCS 9901250037 | |
| Download: ML20199G827 (10) | |
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Comed i Byron Nuclear Power Station Unit 1 Cycle 9 Startup Report l
December,1998 Revision 1 l
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l 9901250037 990106 U g
PDR ADOCK 05000454 6, P
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1.0 Introduction Commonwealth Edison conducted a comprehensive test program following replacement of the Unit 1 Steam Generators (S/Gs) that demonstrated that modified structures, systems, and components perform satisfactorily in service. The test program outlined in this report summarizes events and testing performed during the first heatup and increase to 100% power with Byron 1 Replacement Steam Generators (RSGs). The testing scope included sequencing of special tests (SPPs) and station surveillances to satisfy requirements of the modification.
The Byron Unit 1 Cycle 9 core includes a feed batch of 77 fuel assemblies manufactured by l Westinghouse. The new fuel region incorporates Integral Fuel Burnable Absorber (IFBA) rods with a B-10 loading of 1.5X with a 100 psig backfill pressure. The 1.5X IFBA rods have been used in previous cycles, but unique to Cycle 9 is the reduction of backfill pressure from 200 psig to 100 psig. In addition, enriched annular blankets are used on all feed assemblies (6" top and bottom). Table 1.1 contains characteristics of the Byron Unit 1 Cycle 9 core design.
The Cycle 9 reactor core achieved initial criticality 3/8/98, at 0150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.
The Unit 1 Main Generator was synchronized to the grid 3/9/98 at 0609 hours0.00705 days <br />0.169 hours <br />0.00101 weeks <br />2.317245e-4 months <br />.
Power escalation testing, including testing at full power, was completed 4/4/98.
Table 1.1 Byron Unit 1 Cycle 9 Core Design Data
- 1. Unit 1 Cycle 8 burnup: 433 EFPD
- 2. Unit 1 Cycle 9 design length: 410.2 EFPD Region Fuel Type Number of Enrichment Cycles Burned Assemblies w/o U-235 9A VANTAGE + 24 4.0 2 9B VANTAGE + 16 3.6 2 10A VANTAGE + 36 4.4 1 10B VANTAGE + 40 4.2 1 11A VANTAGE + 44 4.0 0 11B VANTAGE + 32 3.8 0 llc VANTAGE + 1 1.6 0 1
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2.0 Core Testine 2.1 Control Rod Drop Time Measurement This test is performed prior to each fuel cycle with T. , greater than 550 degrees F and with all reactor coolant pumps in operation (per Technical Specification 3/4.1.3.4). Due to the potential for marginally increased reactor coolant system flow impacting control rod drop times, a i verification of acceptable rod drop times was required as part of replacement steam generator testing. The individual full-length shutdown and control rod (RCCA) drop time from the fully withdrawn position is required to be less than 2.7 seconds from the beginning of decay of the ,
stationary gripper coil voltage to dashpot enty. '
All BIR08 RCCA drop times satisfied this acceptance criteria. Table 2.1 summarizes the results ofdrop time measurements from BIR08. In addition, a comparison ofdrop times to previous cycles is provided. Based on this information, there where no changes in the RCCA drop times i due to Steam Generator Replacement. '
i 2.2 Zero Power Physics Testing 1
Zero Power Physics Testing (ZPPT) is performed at the beginning of each cycle as specified by ANS/ ANSI-19.6.1, " Reload Stanup Physics Test for Pressurized Water Reactors." A summary j of the Startup Physics Test results is contained in Table 2.2. All test results were determined to be acceptable.
2.3 Power Escalation Testing Power Escalation Testing is performed during the initial power ascension to full power for each cycle and is controlled by 1/2BVS XPT-3. Tests are performed from 0% through 100% with major testing plateaus at approximately 30%,75%, and 100% power. Significant tests included: ;
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- Core Power Distribution at 25%,46%,60%, and 98% power. l e Reactor Coolant Delta-T Measurement at 60% and 100% power. l Hot Full Power Critical Boron Concentration Measurement (100%).
- Reactor Coolant System Flow Measurement at 60% and 100% power. l 2.4 Core Power Distribution i Core power distribution measurements were performed during power escalation at low power
(<30%), intermediate power (40-75%), and full power. Measurements are made to verify flux ;
I symmety and to verify core peaking factors are within limits. Data obtained during these tests are used to check calibration of Power Range NIS channels and to calibrate them if required.
Measurements are made using the Moveable Incore Detector System and analyzed using the INCORE 3-D computer code. j 2
. . 1 Following successful completion of the low power flux map, power was increased to the 50%
S/G testing plateau (calorimetric power of 46%). At this power level, the maximum corrected F,y h
was greater than Fxy""', but less than Fxy ""' This was the result of the core load pattern and INCORE constants, not the result of S/G replacement. This resulted in the allowable power increase limited to 66% power. Reactor power was subsequently stabilized at 60.4% &nd a core power distribution flux map obtained. The corrected Fxy at this power level was less than Fxy""',
and the unit was released for increase to full power.
Results of the core power distribution measurements at 25%,46%,60%, and full power are l l shown in Tables 2.3, 2.4, 2.5, and 2.6, respectively.
2.5 Full Power Loop Delta-T Determination i
The purpose of this test is to determine the full power Delta-T for each Reactor Coolant loop in !
order to recalibrate any loop with significant change. This procedure is applicable in MODE 1 !
and is performed above 95% Rated Thermal Power (RTP) after each refueling outage. Results ,
are contained in Table 2.7.
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Table 2.1 Unit 1 RCCA Rod drop Time Comparison B1R06 B1R07 #1 B1R07 #2 B1R08 Startup D.02 1.53 D-02 1.555 D-02 1.520 D-02 1.54 B-12 1.5 B 12 1.530 B-12 1.490 B-12 1.51 M-14 1.535 M-14 1.545 M-14 1.520 M-14 1.55 P-04 1.495 P-04 1.505 P-04 1.495 P-04 1.52 B-04 1.49 B-04 1.510 B-04 1.475 B44 1.49 D-14 1.53 0-14 1.575 D-14 1.510 D-14 1.55 P-12 1.465 P-12 1.490 P-12 1.475 P-12 1.49 M-02 1.555 M-02 1 ?s80 M-02 1.535 M-02 1.59 G-03 1.51 G-03 1.535 G-03 1.515 G-03 1.52 C-09 1.475 C-09 1.52C C-09 1.470 C-09 1.5 J-13 1.53 J-13 1.565 J-13 1.520 J-13 1.57 N-07 1.5 N-07 1.510 N 07 1.485 N47 1.52
. C-07 1.475 C.07 1.515 C-07 1.465 C-07 1.51 4 G-13 1.485 G-13 1.520 G-13 1.485 G-13 1.52 N49 1.5 N-09 1.525 N-09 1.470 N-09 1.5 J-03 1.5 J-03 1.525 J-03 1.490 J-03 1.54 E-03 1.55 E43 1.585 E-03 1.535 E-03 1.56 C-11 1.515 C-11 1.560 C-11 1.500 C-11 1.52 L-13 1.53 L-13 1.550 L-13 1.520 L 13 1.53 N-05 1.525 N-05 1.520 N-05 1.520 N-05 1.52 C-05 1.485 C-05 1.520 C-05 1.495 C-05 1.49 E-13 1.51 E-13 15/5 E-13 1.510 E-13 1.54 N-11 1.505 N-11 1.525 N-11 1.490 N 11 1.5 L-03 1.54 L-03 1.545 L-03 1.535 L-03 1.55 I H-04 1.525 H-04 1.530 H-04 1.480 H-04 1.5 D-08 1.495 D-08 1.535 D-08 1.480 D-08 1.51 H-12 1.5 H-12 1.515 H-12 1.490 H-12 1.51 M-08 1.505 M-08 1.525 M-08 1.475 M-06 1.49 s H-06 1.5 H-06 1.525 H-06 1.495 H-06 1.5 H-10 1.49 H-10 1.560 H-10 1.480 H-10 1.52 F 08 1.525 F-08 1.555 F-08 1.505 F-08 1.5 K-08 1.52 K-08 1.535 K-08 1.490 K-08 1.51 F-02 1.55 F-02 1.560 F-02 1.505 F-02 1.52 B-10 1.515 B-10 1.525 B-10 1.490 B.10 1.51 K-14 1.53 K-14 1.540 K-14 1.500 K-14 1.54 I P-06 1.49 P-06 1.500 P-06 1.480 P-06 1.52 ;
B-06 1.49 B-06 1.505 B-06 1.475 B46 1.5 I F-14 1.55 F-14 1.605 F-14 1.535 F-14 1.59 {
P-10 1.515 P-10 1.525 P-10 1.495 P-10 1.52 l K-02 1.54 K-02 1.580 K-02 1.530 K-02 1.57 H-02 1.51 H-02 1.510 H-02 1.510 H-02 1.52 B-08 1.5025 B-08 1.515 B48 1,485 B-08 1.49 H-14 1.5 H-14 1.520 H-14 1.495 H-14 1.52 P-08 1.515 P-08 1.545 P-06 1.520 P-08 1.56 F-06 1.475 F 1.530 F-06 1.505 F-06 1.51 F-10 1.515 F-10 1.580 F-10 1.525 F-10 1.53
! K-10 1.475 K-10 1.545 K-10 1.490 K-10 1.52 K-06 1.465 K-06 1.51 0 K-06 1.470 K-06 1.47 D-04 1.505 D-04 1.525 D-04 1.485 D-04 1.5 M-12 1.495 M-12 1.615 M-12 1.500 M-12 1.52 D-12 1.485 D-12 1.625 D-12 1.480 D-12 1.5 l
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Table 2.1 Unit 1 RCCA Rod drop Time Comparison l i
B1R06 B1R07 #1 B1R07 #2 B1R08 Startup M-04 1.47 M-04 1.495 M-04 1.490 M-04 1.5 H-08 1.525 H-08 1.575 H-08 1.505 H-08 1.52.
Average 1.508 Average 1.540 Average 1.498 Average 1.521 Std. Dev. 0.02306 Std. Dev. 0.03046 Std. Dev. 0.01935 Std. Dev. 0.02563
+2 Sigma 1.554 +2 Sigma 1.600 +2 Sigma 1.537 +2 Sigma 1.572
-2 Sigma 1.462 -2 Sigma 1.479 -2 Sigma 1.459 -2 Sigma 1.469 Total 0.03 Sigma Total Average 1.516 Total +2 Sigma 1.58 Total -2 Sigma 1.46 ,
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i Table 2.2 BIR08 Startup Physics Test Results Review Acceptance Parameter Predicted Measured Difference Criteria Criteria i ARO Critical Boron 1437 ppm 1397 ppm 40 ppm 250 ppm 1000 pcm i Critical Ca with Reference 1304 ppm 1265 ppm 39 ppm 500 pcm N/A l Bank Fully Inserted ,
Differential Boron Worth -8.32 -8.37 0.6% 210% of design N/A pcm/ ppm pcm/ ppm >
ARO ITC -2.814 -2.82 0.0 2 pcmm of N/A -
pcm/F pcm/F design value j
/ 29 MTC -l.134 -1.14 0.0 N/A Within Tech pcm/F pcm/F Spec 3.1.1.3 '
Reference Bank I109 pcm I102.7 pcm -0.57% $10% between 515% between (Shutdown Bank B) measured & measured &
Worth design design Control Bank A Worth 309 pcm 254.6 pcm -54.4 pcm s15% or $100 $30% or 5200 ,
pcm of design pcm of design l Control Bank B Worth 789 pcm 818.3 pcm 3.71 % 515% or $100 530% or s200 l l pcm of design pcm of design Control Bank C Worth 686 pcm 648.1 pcm 5.52 % sis % or $100 <30% or <200 i pcm of design pcm of design Control Bank D Worth 573 pcm 551.5 pcm -21.5 pcm <15% or < loo 530% or s200 l pcm of design pcm of design j Shutdown Bank A 233 pcm 227.8 pcm -5.2 pcm 515% orsloo 530% or s200 l ,
pcm of design pcm of design Shutdown Bank C 448 pcm 444.4 pcm -3.6 pcm 515% or 5100 530% or $200 i pcm of design pcm of design Shutdown Bank D 450 pcm 445 pcm -5.0 pcm 515% or $100 530% or 5200 pcm of design pcm of design Shutdown Bank E 534 pcm 503.5 pcm -30.5 pcm 515% or $100 530% or $200 pcm of design pcm of design Total Rod Worth 5131 pcm 4995.9 pcm -2.63% 510% between >90% of the measured & predicted sum design of bank worths i
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t Table 2.3 Core Power Distribution Results 25% Power .
Plant Data Map ID: Byl0901 '
Date of Map: 3/11/98 .
Cycle Burnup: 0.9 EFPD l Power Level: 24.9 %
Control Bank D Position: 153 steps INCORE 3-D Results :
Core Average Axial Offset -0.139 l Tilt Rations for Entire Cose Height: Quadrant 1: 1.0008 ,
Quadrant 2: 1.0107 l Quadrant 3: 1.0197 Quadrant 4: 0.9688 Maximum corrected Fxy: 1.8674 Fxy "- 1.930 Table 2.4 Core Power Distribution Results 46% Power Plant Data Map ID: Byl0902 Date of Map: 3/13/98 Cycle Burnup: 1.51 EFPD Power Level: 45.9 %
Control Bank D Position: 172 steps INCORE 3-D Results Core Average Axial Offset .795 Tilt Rations for Entire Core Height: Quadrant 1: 0.9933 Quadrant 2: 0.9922 Quadrant 3: 0.9964 Quadrant 4: 1.0182 Maximum corrected Fxy: 1.9556 Fxy *- 1.930 Max. Nuclear Enthalpy Rise Hot Channel Factor: 1.7248 Nuclear Enthalpy Rise Hot Channel Factor Limit: 1.9178 7
l Table 2.5 I Core Power Distribution Results !
! 607c Power l I l
Plant Data i Map ID: By10903 l
- . Date of Map
- 3/17/98 2 Cycle Burnup: 3.42 EFPD Power 12 vel: 60.4 %
.j Control Bank D Rod Position: 185 steps )
l INCORE 3 D Results Core Average AxialOffset -0.619 Tilt Ratios for Entire Core Height: Quadrant 1: 0.9998 Quadrant 2: 0.9982
, Quadrant 3: 0.9893 '
Quadrant 4: 1.0226 Maximum corrected Fxy: 1.7216
- Exv"*- 1.930 l
- . Max. Nuclear Enthalpy Rise Hot Channel Factor
- NA I Nuclear Enthalpy R.ise Hot Channel Factor Limit: NA 2
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Table 2.6 Core Power Distribution Results Full Power Map
- Plant Data Map ID
- Byl0904 Date of Map: 3/30/98 Cycle Burnup: 14.6 EFPD Power Level: 97.97c Control Rod Position: 219 steps INCORE 3-D Results i
Core Average AxialOffset -2.681-Tilt Ratios for Entire Core Height: Quadrant 1: 1.0017 Quadrant 2: 0.9833 Quadrant 3: 1.0026 Quadrant 4: 1.0124 Maximum corrected Fxy: 1.6760 Fxy""- 1.930 Max. Nuclear Enthalpy Rise Hot Channel Factor: 1.5421 Nuclear Enthalpy Rise Hot Channel Factor Limit: 1.7107 8
Table 2.7 Full Power Loop Delta-T Loop That Tcold Full Power Previous Delta-T Cycle Delta-T A 610.2 553.0 57.2 59.1 B 609.5 551.9 57.6 60.6 C 610.6 553.1 57.5 60.8 D 611.2 552.7 58.6 62.1 i
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