ML19210E076
| ML19210E076 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 11/21/1979 |
| From: | Cooper T GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19210E072 | List: |
| References | |
| NUDOCS 7911290267 | |
| Download: ML19210E076 (19) | |
Text
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GEORGIA POWER COMPANY EDWIN I. HATCH NUCLEAR PLANT UNIT NO. 1 DOCKET NO. 50-321 BAXLEY, GEORGIA STARTUP Ts'5 REPORT CYCLE 4 PREPARED BY:
T. A. COOPER REACTOR ENGINEERING
_ 1428 337 7 911290 P_( 7
STARTUP REPORT EDWIN I. HATCH NUCLEAR PLANT UNIT 1 CYCLE 4 ABSTRACT The cycle 4 startup test report for the Edwin I. Hatch Nuclear Plant consists of three parts: (1) an introduction, (2) a summary of the startup tests performed, and (3) results of each of the startup tests. This report presents the results of testing performed at various power levels during the power ascension program at the beginning of Cycle 4.
INTRODUCTON -
Edwin I. Hatch Nuclear Plant resumed commercial operation on 8-29 79 after completing a 129 day refueling outage. The following tests were performed as a part of the startup test program.
(a) Full Core Shutdown Margin Demonstration (b) Cold Critical Eigen Valve Comparison (c) Control Rod Friction Testing and Scram Timing (d) Core Average Axial Power Distribution Comparison (e) TIP System Reproducibility and Core Power Symmetry (f) LPRM Calibration (g) 3PRM Calibration (h) Core Verification The purpose for, a brief description of, and acceptance cri;erion for each of the tests listed above is given in the following pages. W1 ere applicable, a definition of the relevant acceptance criteria for the test is given and is designated either, " Level 1" or " Level 2". A Level 1 criterion normally
- relates to the value of a process variable which is useo as a basis for the safety analysis in the reload 3 safety analysis with supplements previously submitted to the commission and/or which are affected by the limiting conditions for operation in the Station Technical Specifications.
A Level 2 criterion is associated with expectations relating to the design performance of systems or components. If a Level 2 criterion is not satisfied, operating and testing plans would not necessarily be altered.
Investigations of the measurements and of the analytical technique used for the predictions would be initiated.
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FULL CORE SHUTDOWN MARGIN DEMONSTRATION 1-URPOSE To demonstrate that the reactor will be subcritical at the most reactive condition during Cycle 4 operation with the highest worth control rod fully withdrawn.
DESCRIPTION This test is performed during the initial start-up of Cycle 4. The shutdown margin will be demonstrated by analytically determining the reactivity of the core at the time of the first Xenon free criticality. Since the beginning of Cycle 4 la not the most reactive point during the cycle, a correction factor is used in determining that the shutdown margin exceeds the Plant Technical Specification limit. This required margin includes allowances for geometric and material asymmetrics in the core resulting from manufacturing tolerances plus an additional allowance for calculational and analytical uncertainties.
CRITERIA Level 1 The fully loaded core must be suberitical, with the analytical highest worth control rod fully withdrawn, by at least 0. 38% dK/K at the most reactive condition during the cycle.
Level 2 Not applicable.
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COLD CRITICAL EIGEN VALUE f,0MDARISON PURPOSE To compare the actual critical control rod pattern in the cold, Xenon-free condition with the predicted critical rod pattern.
DESCRIPTION The initial critical control rod pattern in the cold, Xenon-free condition will be predicted using data supplied by the General Electric Company. The actual and predicted critical control rod pattern will be compared and any difference explained.
CRITERIA Not applicable.
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CONTROL ROD DRIVE FRICTION AND SCRAM TIMING PURPOSE To demonstrate that the Control Rod Drive (CRD) system operates properly following the completion of core alterations, particularly to demonstrate the absence of control rod core component binding and verify compliance of CRD scram insertion times to respective technical specification limiting conditions for operation.
DESCRIPTION A CRD friction testing will be performed on all drives that have been replaced in accort5nce with existing plant operating procedure s . Running and stall flows for all drives will be recorded for both the insertion and withdrawl strokes for comparison to similar, previously recorded data. The comparision will then be used to ascertain CRD performance trends to identify that component showing signs of degradation. For those CRD's replaced during the current outage, these data provide a base line data for future comparisons.
CRD components with degrading performance trends will be scheduled for closer observation during routine surveillance, surveillance tests, and if necessary, preventative maintenance during a subsequent scheduled outage.
CRD scram time testing will be in accordance with existing plant procedures and with the steam dome pressure above 950 psig. Also, as required by plant Technical Specifications, all in-sequence, fully withdrawn CRD's will be scram time tested prior to turbine-generator synchronization; with all other CRD's to be scram time tested prior to achieving 40% of rated core thermal power.
CRITERIA The mean control rod insertion times after the de-energization of the scram solenoid valves based on the average of all control rods shall not exceed the following:
INSERTION TIME (SECONDS)
PERCENT VESSEL DOME PRESSURE INSERTED GREATER THAN 050 PSIG S 0.375 20 0.40 50 2.0 90 3.50 The mean control rod insertion times after the de-energization of the scram solenoid valves, based on the average of the 3 fastest out of 4, of any 2x?
group of control rods shall not exceed the following:
INSERTION TIME (SECONDS)
PERCENT VESSEL DOME PRESSURE INSERTED GREATER THAN 050 PSIG 5 0.398 20 0.054 50 2.120
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K 90 3.800
The maximum scram insertion time for 90% insertion of any operable control rod shall not exceed 7.00 seconds.
LEVEL 2 With respect to the control rod drive friction tests, if the differential pressure variation exceeds 15 psid for a continuous insertion of the rod, a settling test must be performed, in which case the differential settling pressure should not be less than 30 psid nor should it vary by more than 10 paid over a full stroke. Lower differential pressures in the settling tests are indicative of excessive friction.
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CORE AVERAGE AXIAL POWER DISTRIBUTION COMPARISON PURPOSE To compare the measured average axial power distribution against the predicted distributions.
DESCRIPTION At a core thermal power greater than 50% of rated and at or near rated core flow with equilibrium Xenon, the axial average core power distribution will be evaluated using the neutron monitoring system. This measured distribution will be compared to a predicted distribution supplied by the General Electric Company. Measured and predicted valves, and the percent difference will be reported.
CRITERIA Not Applicable 1428 543
CORE POWER SYMMETRY AND TIP SYSTEM REPRODUCIBILITY PURPOSE
- 1. To confirm the reproducibility of the TIP system readings.
- 2. To determine core power symmetry.
DESCRIPTION Core power distribution data will be obtained during the power ascension program by using complete sets of axial power traces taken with the traversing in-core probe (TIP) system. At power levels greater than 75%, several sets of TIP data will be obtaf ned to determine the overall TIP uncertainty.
TIP data will be obtained with the reactor operating with a symmetric rod pattern and at steady-sta te conditions. The total TIP u%ertainty for the test will be calculated by averaging the total TIP uncertainty for the test will be calculated by averaging the total TIP uncertainty determined from each set of TIP data taken. The total TIP uncertainty is made up of random noise e and geometric components.
Core power symmetry will also be calculated by using the TIP data. As determined from this analysis, uny asymmetry will be verified to be within the uncertainty used in the determination of the MCPR safety limit, using the GETAB method.
ACCEPTANCE CRITERIA Level 1 The total TIP uncertainty (including random noise and geometrical uncertainties) shall be less than 11.4%. This total TIP uncertainty will be obtained by averaging the total uncertainty for all data sets obtained. A minimum of two data sets is sufficient for the determination of total TIP uncertainty. However, if the first two data sets do not meet the criteria, testing may be continued and up to 6 data sets obtained and compared with the criteria. If the 11.4% total TIP uncertainty has not been met by the 6 sets of data, testing may continue and additional data sets obtained provided the effect of the increased uncertainty has been analyzed and, if needed according to the analysis an appropriate adjustment to the MCPR limit is made.
Additional data sets may be obtained to improve the TIP uncertainty by increasing the TIP data base. If the 11.4% total TIP uncertainty berames satisfied, then the MCPR limit, if previously adjusted, can be returned to its original valve.
Level 2 N.A.
142B 344
LOCAL POWER RANGE MONITOR CALIBRATION PURPOSE To calibrate the LPRM system.
DESCRIPTION The LPRM channels will be calibrated to make the LPRM readings proportional to the neutron flux in the narrow-narrow water gap at the chamber elevation.
Calibration factors will be obtained through the use rf either an off-line or process computer calculation that relates the LPRM reading to average fuel assembly power at the chamber height.
ACCEPTANCE CRITERI A Level 1 The meter readings for each LPRM chanber shall accurately provide the necessary in-core local power information used in the determination of the local neutron flux at the respective LPRM's location.
Level 2 N.A.
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AVERAGE POWER RANGE MONITOR CALIBRATION PURPOSE To calibrate the APRM system.
DESCRIPTION A heat balance is made after each major level change and as required. Each APRM channel reading will be adjusted to be consistent with the core thermal power as determined from the heat '-lance. During heat-up, a preliminary calibration will be made by adjusting the APRM amplifier gains so that the APRM readings agree with an initial heat balance based on coolant temperature rise data.
ACCEPTANCE CRITERIA Level 1 The APRM channels must be calibrated to read greater than or equal to the actual core thermal power.
In the startup mode, all APRM channels must produce a scram at less than or equal to 15% of rated thermal power.
Level 2 N.A.
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CORE VERIFICATION PURPOSE To verify that all fuel bundles have been properly loaded into the reactor core.
DESCRIPTION The verification may be performed using an underwater TV camera to visually inspect the location, orientation and seating of each of the 560 fuel bundles that make up the core of Cycle 4.
CRITERIA Level 1 Each bundle must be verified to be in its proper location as specified by the General Electric loading plan and be correctly oriented and seated in its respective cell.
Level 2 N.A.
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STARTUP TEST RESULTS FULL CORE SHUTDOWN MARGIN DEMONSTRATION Full-core shutdown margin demonstration was performed on August 24, 1970, at the time of the first critical of Cycle 4. Necessary data was taken to do the shutdown margin calculation by performing an insequence critical. At the beginning of Cycle 4 using a correction factor to compensate for the B'C4 not
'being the most reactive point during the cycle, the required shutdown margin was 0.38% Delta K/K. Corrected for temperature and period, a shutdown margin of greater than 1.46% Delta K/K was demonstrated, well withi- limits.
COLD CRITICAL EIGEN VALUE COMPARISON Control rods were withdrawn in the B1 rod withdrawal sequence on Augast 24, 1979. Criticality was achieved on notch position 10 of the thirty-fifth in-sequence control rod at a moderator temeprature of 1540 F. The resultant period was 61 seconds. A copy of the control rod pattern at criticality is attached here as Figure 1. Based upon accumulated data for GD 023 BWR4 initial core, plus the three reloads , the cold critical Eigen Value was expected to be 1.009. The observed value for Hatch 1 BOC4 was calculated to be 1.0063, an excellent agreement with expectation.
CONTROL ROD DRIVE FRICTION AND SCRAM TIMING Control rod drive friction testing was performed on the seventeen drives replaced during the refueling outage. The results of the friction testing are presented in Table 2A. All drives met the acceptance criteria and were determined to be operating properly.
All drives were scram time tested in accordance with existing plant procedures at a steam dome pressure greater than 950 psig. A summary of the results is presented in Table 2B. All criteria were easily met.
CORE VERIFICATION On July 13, 1479, with the aid of an underwater TV system, the core verification was performed. Each ban ale was verified to be in its proper location and to be correctly oriented and scated in the cell.
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POWER DISTRIBUTION TEST During the reload design process, the BWR Simulator is used to establish the basic control rod patterns to be used during operation and to demonstrate that operation at or near rated power is feasible. However, during power operation process computer data is used to demonstrate compliance with thermal limits and establish the detailed control rod patterns and operating power distributions. Therefore, the accuracy of the BWR Simulator Calculation of the power distribution has no direct impact on compliance with ther.nal limits during operation.
NEDO-20946, BWR Simulator Methods Verification Licensing Topical Report, May 1976, presents the results and conclusions from extensive comparisons of BWR Simulator Calculations and operating reactivity and power distribution data.
The BWR Simulator Power Distribution and Reactivity Results presented were generally in good agreement with operating data.
On Sept. 28, 1979, at a core thermal power of about 2425 MWth and a core flow of about 76 MLB/HR with Equilibrium Xenon, the axial power shape was calculated by a Periodic Core Evaluation. The average axial power distribution as determined by the process computer and that distribution as predicted by the BWR Simulator are presented and compared below in Table 1.
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TIP REPRODUCIBILITY AND CORE POWER SYMMETRY The total TIP uncertainty and core power symmetry was calculated from the average of two data sets. Each data set consists of data necessary to calculate a random noise uncertainty and a total TIP uncertainty. The random noise term is a measure of TIP reproducibility while the geometrical uncertainty term, which is a measure of cere power symmetry, can be obtained by statistically subtracting the TIP reproducibility term from the total TIP uncertainty.
The random noise data consists of a collection of four (4) traces from the common channel from each of the four (4) TIP machines for a total of sixteen (16) traces. After each trace was run using OD 2, then the BASE distribution calculated by OD 2 for that LPRM string was collected. The standard deviation due to noise was calculated from the individual deviations of nodal power at each nodal level five (5) through twenty two (22).
The total TIP uncertainty data censists of a complete OD 1 TIP set and BASE distribution calculated by OD 1 fc" all LPRM strings. For each symmetric TIP pair, the nodal BASE value for tha string in the upper left half of the core in divided by its counterpart in the lower right half of the core and then the average and standard deviation of their ratios are calculated. The total TIP uncertainty is then calculated from the above ratios. The geometrical uncertainty is then available by statistically subtracting the random noise from the total TIP uncertainty.
Data set 1 was collected on 10/3/70 and data set 2 was collected on 10/5/79.
Table 3 summarizes the results of the two data sets. The acceptance criterion for the average total TIP uncertainty for all data sets 15 11.4%. The calculated value of 2.325% is well within the acceptance criterion, therefore no adjustments need be made to the MCPR limits as currently implemented.
There is no acceptance criterion on random noise.
LOCAL POWER RANGE MONITOR CALIBRATION I PRM calibrations were performed at approximately 25%, 50%, 75%, and 100% of rated thermal power and as required during the startup test program.
Calibration constants relating LPRM readings to average fuel assembly power at the chamber height were determined by process computer program OD-1 (whole-core LPRM calibration and BASE distributions) and were subsequently employed in the determination of local neutron flux levels at the respective LPRM elevations.
AVERAGE POWER RANGE MONITOR CALIBRATION After each major power level change and APRM calibrations were performed as required throughout the power ascension program based upon thermal power as determined by process computer program OD-3 (core thermal power and APRM calibration). Each instrument was calibrated to read greater than or equal to the actual core thermal power.
In addition to those calibrations performed in the power range, a preliminary APRM calibration was performed during heat-up by adjusting the APRM gains such that the instrument reading agreed with an initial heat balance . based on moderator temperature rise. Normal startup and routine surveillance demonstrated the APRM's ability to produce a scram at less than or equal to 15% of rated thermal power in the startup mode.
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s TABLE 1 RESULTS OF POWER DISTRIBUTION COMPARISON LOCATION 1 2 3 4 5 6 7 8 0 10 11 12 AXIAL REL. PWR (PC) .54 . 04 1.20 1.18 1.11 1.05 1.02 1.09 1.14 1.13 .93 .57 AXIAL REL. PWR (CE) .56 1.15 1.26 1.20 1.11 1.04 1.03 1.10 1.18 1.12 .86 .34
% DIFFERENCE 3.6 9.6 4.8 1.7 0 .96 47 .90 3.4 . 84 -8.1 46.1 1428 351 4
TABLE 2A B0C-4 CONTINU0US INSERT FRICTION TEST DRIVE iREACT07 CRD INSERTION PRESS. DIFF.
LOCATION PRESS. PSIG PRESS. PSIG MAXIMUM MINIMUM 30-35 0 280 6 ._
26-23 0 280 10 5 18-27 0 280 8 5 14-27 0 280 14 9 06-11 0 280 14 8 14-11 0 280 10 4 18- 19 0 280 6 2 22-23 0 280 6 4 26-31 0 280 8 2 14-35 0 280 8 4 34-31 0 280 9 6 46-31 0 280 8 3 26-27 0 280 7 2 42-23 0 280 5 2 50- 19 0 280 5 3 26-19 0 280 5 2 34-51 0 280 6 3 34-35 0 280 6 I 2 1428 352 9 --se'e
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N TABLE 2B -
SCRAM TIME TESTING RESULTS - BOC4
% Inserted Slowest Scram Insertion Average Scram Inserti6n
~
Average Scram Insertion from 48 Time Time Time Criteria Fr'om HNP-1 Tech Specs Sec. 3.3.C.2 Single Rod 2x2 Array Single Rod 2x2 Array Single Rod 2x2 Array 3 Fastest 3 Fastest 3 Fastest 55 .489 .369 .315 .911 .375 .198 20% .919 .737 .674 .664 000 .o54 50% 1.934 1.519 1.412 1.389 2.000 2.120 90% 3.147 2.666 2.464 2.419 3.500 3.800
TABLE 3 TIP REPRODUCIBILITY AND CORE POWER SYMMETRY TEST RESULTS - BOC4 DATA SET 1 DATA SET 2 AVERAGE DATE 10-3-79 10-5-79 CMWT 2436 2436 WT 77 77 RANDOM NOISE UNCERTAINTY 0.638% 0.638% 0.638%
TOTAL TIP UNCERTAINTY 2.550% 2.100 % 2.325%
GEOMETRICAL UNCERTAINTY 2.468% 2.001% 2.2345%
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FIGURE 1 R0D PATTERN AT INITIAL CRITICALITY BOC-4 51 48 00 48 00 48 47 00 00 00 00 00 00 00 00 00 43 00 48 00 43 00 48 00 48 00 48 00 39 00 00 00 00 00 00 00 00 00 00 00 35 48 00 48 00 48 00 48 00 48 00 48 00 48 31 00 00 00 00 00 00 00 00 00 00 00 00 00 27 48 00 48 00 00 00 48 00 10 00 48 00 48 23 00 00 00 00 00 00 00 00 00 00 00 00 00 19 48 00 48 00 48 00 00 00 48 00 48 00 48 15 00 00 00 00 00 00 00 00 00 00 00 11 00 48 00 48 00 48 00 48 00 48 00 07 00 00 00 00 00 00 00 00 00 03 48 00 48 00 48 02 06 10 14 18 22 26 30 34 38 42 46 50
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