ML20235X737
| ML20235X737 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/06/1989 |
| From: | Hairston W, Neeley G GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| HL-340, NUDOCS 8903130602 | |
| Download: ML20235X737 (45) | |
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ENCLOSURE POST REPUELING OUTAGE STARTUP TEST REPORT UNIT 1 CYCLE 12 m
i EDWIN I. HATCH NUCLEAR POWER PLANT GEORGIA POWER COMPANY BAXLEY, GEORGIA Prepared by: )j g p }
G.W. Nee /ey Reactor Engineering Supervisor hb /[ . .
4 ABSTRACT 1
THIS REPORT CONSISTS OF A
SUMMARY
OF i SELECTED STATIC AND DYNAMIC REACTOR !
CORE FUNCTIONAL TESTS PERFORMED AS PART OF THE POST-REFUELING OUTAGE STARTUP TEST PROGRAM FOR HATCH UNIT 1 CYCLE 12.
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9 1 I. INTRODUCTION 1.1 PURPOSE This report. consists of a summary of selected static and dynamic reactor core performance tests conducted prior to and during the beginning-of-cycle startup of Hatch Unit 1 Cycle 12.
1.2 PLANT DESCRIPTION The Edwin I. Hatch Nuclear Power Plant Unit 1 is a General Electric design single-cycle boiling water reactor (BHR/4).
Hatch Unit 1 is rated at 2436 MW(th) with a generator rating at this power of 810 MH(e). The plant is located on the South side of the Altamaha River, Southeast of the intersection of the river with-U. S. Highway No. 1 in the Northwestern sector of Appling County, Georgia.
1.3 POST-REFUELING OUTAGE STARTUP TEST DESCRIPTION The Edwin I. Hatch Nuclear Power Plant Unit i resumed commercial operation on 12/09/88 after completing a 72 day refueling outage. The following core performance tests were performed as part of the post-refueling outage startup test program:
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't. t 3.1 Core Verification 3.2 Control Rod Drive Friction Testing 3.3 Control Rod Drive Timing 3.4 Full Core Shutdown Margin Demonstration 3.5 Cold Critical Eigenvalue Comparison 3.6 Whole Core LFRM Calibration 3.7 APRM Calibration.
3.8 Control Rod Scram Time Testing l 3.9' Tip Reproducibility and- A:ymmetry Calculation 3.10 Reactivity Anomaly Calculation The purpose for, a brief description of, and acceptance criterion for each of the tests listed above is enumerated in section III of this report.
1.4 POST-REFUELING OUTAGE STARTUP TEST ACCEPTANCE CRITERIA Where 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 used as the basis for the reload safety analysis with supplements previously submitted to the Nuclear Regulatory Commission and/or which are affected by the limiting condition for operation in the Unit's Technical Specification.
A Level 2 Criterion is associated with expectations related 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 prediction would be initiated.
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, i II. CYCLE DESIGN SlM4ARY 2.1 CORE DESIGN SLM4ARY.
Cycle 12 was designed to operate approximately 379 effective full powsr days (EFPDs) at rated conditions, with an additional 10 EFPDs available from increased core flow and 20 EFPDs from feedwater temperature reduction. One hundred ninety six fresh fuel bundles were loaded in a conventional core configuration.
With the. exception of the four ANF 9x9 Lead Fuel Assemblies, all fuel assemblies loaded in the interior of the core in Cycle 12 have barrier cladding which permit' sthe' relaxation of PCIOMRs.
Because of-the PCIOMR relaxation, control rod sequence exchanges are to be performed at core exposure increments of 2000 MHd/st, instead of every 1000 MHd/st, to minimize fuel duty and improve the capacity factor.
2.2 REACTIVITY / THERMAL LIAIT MARGINS The two parameters which describe the global behavior of the core reactivity throughout the cycle are hot excess reactivity (HER) and cold shutdown margin (CSDM). The beginning-of-cycle (BOC) hot excess reactivity is 0.97% while the peak hot excess reactivity of 2.26% occurs at 5000 MHd/st. The minimum predicted cold shutdown margin of 1.73% occurs at 6000 MHd/st.
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1-The thermal margin design goals for target rod patterns of 10%,
l 10%, and 7% for MFLPD, MAPRAT, and MFLCPR, respectively, could not be met at some points in the cycle without sacrificing the axial power shape. Based on previous experience, however, the difficulty in meeting thermal margin goals is not expected to lead to a derate of the unit.
The calculations. indicate that it may not be possible to pull all rods out at nominal end-of-cycle (EOC) without violating thermal limits. During the cycle, rod pattern recommendations will take advantage of available thermal inargins in order to burn the bottom'of the core as much as possible, to minimize or eliminate this problem.
2.3 FUEL SlM4ARY/ CORE LOADING PATTERN DESCRIPTION Hatch 1 Cycle 12 is a conventional core loading which was designed to achieve 8099.3 MHd/st. The energy requirements could not be' met using a control cell core design. The Reload 11 batch of fresh fuel contains a total of 196 bundles loaded in the interior of the core. Of these bundles, 108 are GE8B 2.96 weight percent bundles, 84 are GE78 2.84 weight percent, and 4 are Advanced Nuclear Fuels (ANF) Lead Fuel Assemblies-(LFAs) designed to mimic the standard GE7B 2.84 weight percent fresh bundles in nuclear performance. Among the 84 GE7B bundles, there are six Process Variable Lead Test Assemblies (PVLTAs).
The loading pattern in this cycle is quadrant mirror symmetric, except at two Tocations where prematurely discharged bundles were replaced by older bundles of different type and burnup.
The core design constrairds used in developing this loading pattern are described below.
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- 1. ) . Design Flexibility - As in several past cycles, the loading pattern was designed to accommodate fuel bundle changes-based on-inspection results from the refueling outage
.without rerunning licensing calculations. Licensing constraints require that any changes from the licensed pattern must meet certain requirements specified in GESTAR II. These requirements were met.. including the two replacement bundles.
< 2) Inventory' Bundles.- Th.* rcre design was developed to include 84 fresh GE78 bura!!es which were part of.the inventory. maintained in the pool.
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- - ;y r III. -SUMARY OF POST-REFUELING OUTAGE STARTUP TEST RESULTS 1
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- 3.1 CORE VERIFICATION 3.1.1 PURPOSE To verify that all fuel assemblies have been properly loaded into the reactor core as per the licensed final loading pattern including fuel bundle location, orientation and seating.
3.1.2 ACCEPTANCE CRITERIA Level 1 Criteria: Each fuel assembly must be verified to be in its proper location as specified by the' General Electric final loading pattern (Licensed Core) and be correctlyorientedaddseatedinits respective cell.
. Level 2 Criteria: N/A 3.1.3 TEST DESCRIPTION The Hatch Unit 1 Cycle 12 core verification was performed by use of an underwater TV camera to visually inspect the location (by bundle serial number identification),
orientation, and seating of each of the 560 fuel assemblies that comprise the as loaded core.
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3.1.4 TEST RESULTS 1
1% The. core verification was ' performed on 11/14/88 in accordance with engineering procedure 42FH-ERP-014-0S,
'" Fuel Movement Operation". Video tapes of the. core loading pattern indicated that all fuel assemblies were in the proper. location with proper orientation. Fuel Bundle seating. verification indicated that'all but one fuel assembly were properly. seated. .The unseated fuel assembly, was subsequently properly seated and verified as such.
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3.2 CONTROL ROD DRIVE FRICTION TESTING 3.2.1' PURPOSE To demonstrate that the control rod drive system operates properly following the completion of a core alteration.
In particular, this functional test demonstrates the absence of excessive friction in the control rod drive from internal drive obstructions following extensive control rod drive maintenance / replacement.
3.2.2 ACCEPTANCE CRITERIA LEVEL 1 Criteria: The differential pressure variation of all control rod drives to be tested must be less than or equal to 15 PSID for continuous insertion. If this criterion car.not be satisfied, then a settling test must be performed in which case the differential settling pressure should not be less than 30 PSID over the full stroke. Lower differential pressures in the settling test are indicative of excessive friction.
Level 2 Criteria: N/A l
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'3.2.3 TEST DESCRIPTION'
. Control: rod drive friction- testing is normally performed
! on.all control drives that.have been replaced or drives that have undergone. extensive maintenance repair during the refueling outage. In~ essence, the functional test l
measures the differential pressure across the drive piston during_a normal insertion stroke. If necessary, a settle test, which measures the differentia 1' settling pressure of each notch, is performed on a control rod drive during a withdrawal or insertion stroke.
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3.2.4. TEST RESULTS Control. Rod Friction Testing was performed on 11/14/88 for
.the fifteen control rod drive units replaced during the outage in accordance with engineering procedure 42IT-C11-001-0S, " Control Rod Friction Testing". The test results indicated that thirteen of the fifteen control rod drives were satisfactory either by the normal insertion-differential pressure test.or the settle test. .Two of the drives encountered excessive friction as the control rod was stroked. The tw) defective drives were replaced with new drives and then the control rod drive friction test was rerun for these two drives with satisfactory results.
Subsequent analysis indicated that the defectiver'ebuilt control rod drives had bent index tubes thereby resulting in .the observed. friction. A summary of the results of the control rod friction testing is given in Attachment 1 of this report.
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'3.3- CONTROL ROD DRIVE TIMING l
3.3.1 PURPOSE To demonstrate that the control rod drive system operates properly following the completion of a core alteration.
In particular, this functional test verifies that the insert and withdraw capability of the control rod drive system is within acceptable limits.
3.3.2 ACCEPTANCE CRITERIA Level 1 Criteria: The insertion and withdraw drive time for each control rod drive must be between 38.4 and 57.6 seconds. In the-event that a control rod drive fails to meet this criteria, then the applicable drive must be adjusted and a new criteria of 43.2 to 52.8 seconds is applied to the adjusted drive.
Level 2 Criteria: N/A i
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3.3.3 TEST DESCRIPTION Control Rod drive timing is performed once per operating I
cycle on all the control rod drives in the reactor core.
Normal withdrawal and insertion times are recorded for each of the drives, under normal drive water pressure. If acceptable withdrawal and/or insertion cannot be obtained for normal drive water pressure, then the respective neddle valve for the applicable withdrawl and/or insertion stroke must be adjusted until an acceptable drive time is achieved in accordance with the above criteria.
3.3.4 TEST RESULTS Control rod timing test was performed on 12/04/88 for all 137 control rod drives in accordance with plant procedure 34SV-C11-004-1S, "CRD Timing". Each control rod drive was determined to have, or adjusted (HHERE NECESSARY) to have a normal insertion and withdraw speed as required. A summary of the results of this functional test is given in Attachment 2 of this report.
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3.4 FULL CORE SHUTDOHN MARGIN DEMONSTRATION 3.4.1 FURPOSE i
To demonstrate that the reactor can be made subcritical.
for any reactivity condition.during Cycle 12 operation with the analytically determined, highest worth control rod capable of withdrawal, fully withdrawn, and all other rods fully inserted.
3.4.2 ACCEPTANCE CRITERIA:
Level 1 Criteria: The fully loaded core must be subtritical by at'least 0.38%' delta K with the analytically determined, highest worth control rod capable of withdrawal, fully withdrawn, and all other rods fully inserted at the most reactive condition during the cycle.
LEVEL 2 Criteria: N/A 3.4.3 TEST DESCRIPTION The full core shutdown margin demonstration was performed analytically during the Hatch Unit 1 Cycle 12 BOC
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insequence critical with the reactor core in a xenon-free state. To account for reactivity effects such as moderator temperature, reactor period, and one rod out criterion, correction factors are used to adjust the startup condition to cold conditions with the highest worth control rod fully withdrawn.
. c Furthermore, since the beginning-of-cycle 12 was not the most reactive time in the operating Cycle, a correction factor (R Factor) of 0.36 delta K was applied to the results of this calculation to ascertain the cycle minimum cold shutdown margin.
3.4.4 TEST RESULTS The full core shutdown margin demonstration was performed on December 7, 1988, in accordance with Core Calculation l Procedure 42CC-ERP-010-05, " Shutdown Margin, Demonstration". Results of this calculation yielded a cold shutdown margin of 1.92% delta K for the beginning of cycle cold shutdown margin and a cycle minimum cold shutdown margin of 1.56% delta K. A summary of the shutdown margin calculation is given in Attachment 3 of this report.
3.5. COLD' CRITICAL EIGENVALUE COMPARISON 1
1 3.5.1 PURPOSE- 'f To' compare the actual xenon-free critical control rod configuration with the analytically predicted xenon-free control rod configuration that has been corrected for moderator' temperature and reactor period ' reactivity effects.
3.5.2 ACCEPTANCE CRITERIA Level 1 Criteria: N/A Level 2 Criteria: N/A 3.5.3 TEST DESCRIPTION The cold critical eigenvalue is the value of the core Keff at which criticality is achieved with the reactor in a xenon-free state and the coolant at 68'F. This value is determined _ analytically by the Core Analysis Group in Birmingham, Alabama by use of the BWR simulator model PANACEA. Once the actual critical state is achieved during the beginning of cycle startup, the applicable data is sent to the Core Analysis Group and a comparison between the actual (corrected for moderator temperature reactivity effects and reactor period) and predicted critical control rod configuration is performed. The results of this computation is used for predictive model benchmarking.
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3.5.4 TEST RESULTS.
r The beginning-of-cycle startup for Hatch Unit 1 Cycle 12
=was performed on December 7, 1988. The following reactor core conditionc were observed when a critical state was achieved:
SEQUENCE A2 RSCS Group 1 Fully Hithdrawn RSCS Group 2 12 Control Rods fully withdrawn and the 13th control rod (30-11) rod withdrawn to notch 20.
Moderator Temperature 167.5F Reactor Period 154.6 sec Control Rod Density 0.7853 Calculated MCSDM 1.56% delta K A cold critical eigenvalue of 1.0015 was calculated from the actual critical data given above. .This compares to an assumed value of '1.0030. A surnmary of the cold critical eigenvalue comparison input and result is given'in Attachment 4 of this report.
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3.6 MHOLE CORE LOCAL POWER RANGE MONITOR (LPRM) CALIBRATION 3.6.1 PURPOSE To determine 1). The LPRM calibration constants such that when multiplied by the actual LPRM readings will produce L calibrated LPRM readings proportional to the traversing incore probe (TIP) signal readings at the LPRM locations and 2). The BASE and BASLP arrays which contain the machine normalized full power adjusted TIP signals at every node.and LPRM detector location 'respectively.
3.6.2 ACCEPTANCE CRITERIA Level'1 Criteria: N/A ,
Level 2 Criteria: N/A 3.6.3 TESTING DESCRIPTION The whole core LPRM calibration and BASE distribution .;
calculation determines the LPRM calibration constants and the BASE and BASLP distributions used in the axial power distribution calculations. In essence, the TIP system is used in conjunction with the process computer to generate the axial distribution of machine-independent / core average power-independent TIP signals. The axial distribution of ,
machine normalized full power adjusted TIP data is used to d generate LPRM calibration constants required for TIP normalized LPRM readings.
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. o In addition, machine normalized full power adjusted TIP readings are generated at every axial node (BASE distribution) and at every LPRM detector location (BASLP i distribution). These arrays are used as input data in the core calculation / monitoring programs to accurately calculate the power distribution at every node in the core.
t 3.6.4 TEST RESULTS Nhole core LPRM Calibration and BASE distribution was performed in accordance with Engineering Procedure 42CC-ERP-012-0S, "001 And OD2 TIP Operation", at approximately 25%, 50%, 75%, and at 100% power. LFRM calibration constants, BASE and BASLP arrays were computed by the process computer and subsequently used successfully by the process computer to calculate the nodal power distribution and the core thermal limits.
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3.7 APRM CALIBRATION-L 3.7.1 PURPOSE ~
To calibrate the APRM system to actual core thermal power, as determined by a heat balance.
'3.7.2 ACCEPTANCE CRITiRIA Level 1 Criteria: The APRM readings must be within a
-tolerance of 21. of core thermal power as determined from a heat balance.
Level 2 Criteria: N/A 3.7.3 TEST DESCRIPTION The APRM gains are adjusted after major-power level changes, if required, to read the actual core thermal power as determined by a heat balance in accordance with procedure 34SV-SUV-021-0S "APRM-Adjustment to Core Thermal Power". The heat balance required for the calibration process may be obtained from process computer program P1 (Periodic Core Evaluation) and 003 (Core Thermal Power And APRM Calibration) or from a manual heat balance in accordance with core calculation procedure 42CC-ERP-001-1S, " Core Heat Balance-Power Range".
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3.7.'4 TEST RESULTS After major power level changes, and anytime deemed necessary'during the Hatch Unit 1 Cycle 12 B0C Startup, an APRM calibration was performed in accordance with plant procedure 34SV-SUV-21-0S, "APRM Adjustment To Core Thermal Power". Each APRM was calibrated within a tolerance of 21.;
to read core thermal power as calculated by the process computer.-
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3.8 CONTROL R00 SCRAM TIME TESTING 3.8.1 PURPOSE To demonstrate that the control rod drive system functions as designed with respect to scram insertion times following:the completion of core alteration.
3.8.2 ACCEPTANCE CRITERIA Level 1 Criteria: A) The average scram insertion time for all operable control rods from the fully withdrawn position, based on de-energization of the scram pilot solenoids, with reactor steam dome pressure above 950 psig shall not exceed the following:
Notch Position Average from Fully Insertion Withdrawn Time (Secs) 46 0.358 36 1.096 26 1.860 06 3.419 l 1
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B) The average' scram insertion time,
-from the. fully withdrawn position, for the 3 fastest control rods in each~
group of four control rods arranged in a 2x2 array, basedion the de-energization of the scram pilot solenoids, shall not exceed the following:
Notch Position Average from Fully Insertion Hithdrawn Time (Secs) 46 0.379 36 1.162 26 1.972 06 3.624 C) The maximum scram insertion time of each control rod, from.the fully-withdrawn position to position 06, based on the de-energization of the scram piolot solenoid, shall not exceed 7.0 seconds.
Level 2 Criteria: N/A 4 i
3.8.3 TEST DESCRIPTION The control rod drive scram time testing was performed in accordance with engineering procedure 42SV-Cll-001-1S,
" Control Rod Scram Time Testing", with the steam dome pressure above 950 PSIG. The test consists of scramming each control rod, collecting the resulting scram time data, and analyzing the data in accordance with the procedure to ensure compliance with acceptance criteria noted above.
3.8.4 TEST RESULTS All control rod drives were tested in accordance with engineering procedure 42SV-Cll-001-1S, " Control Rod Scram Testing", with the steam dome pressure greater than 950 PSIG. A summary of the results is given in Attachment 5 of this report.
3.9 TIP REPROD'JCIBILITY AND ASYMETRY CALCULATIONS 3.9.1 PURPOSE To confirm the reproducibility of the TIP (Traversing Incore Probe) system by calculating a random noise uncertainty and to determine.TIP asymmetry by calculating a geometric uncertainty.
3.9.2 ACCEPTANCE CRITERIA Level 1 Criteria: The total TIP uncertainty utilizing a minimum of 2 data sets shall be less than 11.4%. If the average of the 2 data sets do not meet the criteria, then up to 6 data sets may be obtained and averaged into the total TIP uncertainty. If the 11.4% total uncertainty has not been met with the 6 data sets, additional data sets may be obtained, provided the effect of the increased uncertainty has been analyzed by General Electric, and if needed, an appropriate adjustment to the MCPR limit is made.
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There is no acceptance criteria for-random noise uncertainty. It is expected.that.the random noise {
uncertainty will be a major contributor to the total uncertainty. Total uncertainty contains both random noise-and geometric uncertainty components.
Level 2 Criteria: N/A 3.9.3 TEST DESCRIPTION _.
Data collection for TIP reproducibility and uncertainty calculations will be obtained by running OD1 (Whole Core LPRM Calibration And BASE Distribution) and OD2 (LPRM Substitute Value and BASE Distribution) to obtain axial power traces with the TIP system at power levels greater than 50% and the reactor operating with a symmetrical rod pattern at steady state conditions. Two sets of data will be obtained and the total TIP uncertainty will be determined by averaging the total TIP uncertainty for each data set. -Each data set consists of data'necessary to calculate a random noise term and a total TIP uncertainty. The geometric uncertainty term can be obtained by subtracting the random noise factor from the total TIP uncertainty.
The random noise term is obtained by running OD2 in the common channel (string 28-29) of each TIP machine once.
The LPRMs in the string 28-29 are then deleted from scan and 3 more OD2s are run in string 28-29. After each individual scan, BASE data is collected by use of the "T1PSYM" program on the VAX minicomputer.
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The standard deviation due to random noise is calculated from the individual deviations of nodal power at each nodal level.5 through 22. The total random noise deviation is the average of the standard deviation for each nodal level.
The total uncertaining is.obtained by running OD1. After the 001 run, BASE data is obtained again with the "TIPSYM" program on the VAX minicomputer.
For each symmetric TIP pair, the nodal BASE value for the string in the upper left half of the core is divided by its counterpart in the lower right half of the core. The average and standard deviation of these ratios are than-calculated. The total TIP uncertainty is then obtained and the geometric uncertainty ~is calculated by statistically subtracting the random noise factor from the total TIP uncertainty.
3.9.4 TEST RESULTS TIP reproducibility and asymmetry calculations were performed in accordance with engineering procedure 42FH-ENG-023-IS, "TIP Reproducibility and Symmetry' Uncertainty Calculations", using 2 sets of data collected on January 27, 1989. A summary of the results are given in Attachment 6 of this report. Uncertainy was less than two percent.
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- 3.10 REACTIVITY ANOMALY CALCULATION 3.10.1 PURPOSE To check for possible reactivity anomalies as the core excess reactivity changes with exposure.
3.10.2 ACCEPTANCE CRITERIA Level 1 Criteria: The corrected control rod density shall not differ from a control rod density equivalent by more than plus or minus 1% delta K.
Level 2 Criteria: t;/A 3.10.3 TEST DESCRIPTION During the BOC startup following a refueling outage and every month thereafter, a reactivity anomaly calculation is performed to monitor the core reactivity during the cycle. Since anticipated operation or unanticipated events may place the reactor in a condition other than that for which the baseline anomaly curve was developed, the actual control rod density is corrected for off-rated conditions. The corrected control rod density is then compared to the reactivity anomaly curve provided in the !
Cycle Management Report to ensure that the corrected control rod density is within the plus or minus 1% delta K limit of the curve. l I
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4 3.10.4 TEST RESULTS.
Reactivity anomaly calculation was performed'in accordance with 42CC-ERP-007-0S. " Reactivity Anomaly Calculation", on December 16, 1988; The corrected
- control rod density was well within the acceptance criteria range as specified above. The results of this calculation is given in Attachment 7 of this report.
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ATTACHMENT 3 SHUTDOWN MARGIN CALCULATION RESULTS K ,,, (STRONGEST ROD OUT) 0.9792 K
cn T (EIEWRW M MITININ) 1.nu VALUE OF R 0.36% AK REACTIVITY CORRECTION FOR -0.20% AK MODERATOR TEMPERATURE REACTIVITY CORRECTION FOR 0.04% AK PERIOD COLD SHUTDOWN MARGIN (CSDN) 1.92% A K NINIMUN COLD SHUTDOWN 1.56% AK NARGIN (MCSDN)
MININUM REQUIRED COLD SHUTDOWN 0.38% A K ,
MARGIN (MRCSDN)
- - a' Southern Company Servic s, Inc.
Post Offc3 Box 2625 ;
Birmingnam - Alabama 35202 '
Telephone 205 870.-6011 ATTACHMENT 4 h
- l. Southern CompanyServiceS the southern erectic system
.)
l January 4,1989
< Mr. H. C. Nix,'Jr.
Edwin I. Haten Nuclear Plant CAH-NSF-979 i
s -P.O. Box 442 PA-1061 Baxley, Georgia 31513 EWO: 3005-!Z
Reference:
Letter, T. R. Powers to K. S. Folk, December 21, 1988,
" Hatch-1 Cycle-12 Cold Critical Data."
Dear Mr. Nix:
The observed cold critical conditions of Haten-1 BOC12 (cf. referenced ,
letter) were modelled with PANAC07 in order to verify the validity of the shutdown margin calculations perfomed and reported in the Cycle Management Report.
The calculated cold critical eigenvalue based on the observed cold critical conditions was 1.0015. The assumed value used in the Cycle Management Report was 1.003. Tne difference between real and assumed eigenvalues is acceptable and will have no impact in meeting the shutdown margin reouired by the Technical Specifications throughout the cycle.
Please let me know if you need any more infomation regarding shutdown margie demonstration.
Si ncerely, g l T. C. Kane11opoulos BWR Core Analysis Engineer Approved by:
K. S. Folk, Manager BWR Core Analysis !
TCK/ gps cc: Georgia Power Con:pany J. T. Beckham R. D. Baker T. R. Powers G. W. Neeley Southern Company Services L. K. Mathews W. R. Sutton W. R. Mertz 2876H
, l lI .
1 1l YT AS E RE 9 2 2 4 MM RT 7 6 7 2 AI AS 3 1 9 6 RT A .
C 2F 1 1 3 SN x OA 23 EI I GTR E ARE L D 8 6 0 9 s RET G 5 9 6 1 .
ESI N O 3 0 8 4 I R VNR AI C S 1 1 3 2
1 C YT O AS B RE 7 1 5 1 RT 9 1 4 7 AS 2 8 3 4
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MM 2F 1 2 AI x S RT 23 T C SN L O D 5 U EI O S GT R AR 1 1 2 5 T E RE E 0 2 6 0 N R ES 3 8 3 5 VN L .
E AI G 1 2 M G N I
H N S C I A T T S YT T E AS RE 6 5 A T E RT 0 4 2 1 MM AS 2 6 4 6 E AI A 3 8 . .
RT 2F . .
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- ATTACHMENT TIP REPRODUCIBILITY AND CORE. POWER
' SYMMETRY TEST RESULTS -
BOC12 DHTA SET 1 DATA SET 2 AVERAGE i DATE 1/27/89 1/27/89 CMWT 2435 2435 MANDOM. NOISE .739% .569% .654%
UNCERTAINTY-l' TOTAL TIP 1.600% 1.795% 1.697%
UNCERTAINTY GEOMETRICAL 1.419% 1.702% 1.561%
UNCERTAINTY
-p: ,,
' ATTACH M ENT 7 REACTIVITY ANOMALY CALCULATION RESU LTS Date Performed: December 16, 1988 Unit 1 Cycle 12 Sequence A2 CMWT: 2430 W T: 77.37 Mlb/hr Dome Pressure: -1001 psia Actual Control Rod Density: 0.086 Corrected CRD = Actual CRD + Correction Correction: -3.1503 E- 1 *(1 -(CMWT/2436))
+ 2.1328 E- 1 *(1 -(WT/ RATED CORE - FLOW))
+2.82369E-3 *(DESIGN INLET SUBCOOLING - DHS)
+4.17224E-5 *(1 O20 PR) - 0.0015056685 i
Corrected CRD = 0.08751 Predicted CRD - 0.068 I
+1 % Value - O.166 -1 % Value - 0.018 l~ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ _ _ - _- - ._.____1
, 4.
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- phore o 6 f/i$id? C nv e6 h. > sr. 1 W. G. Hairston, til Ser or Vice Plei;i@nt NxkA Opetations HL-340 0028V.
X7GJ17-H000 March 6, 1989 U. S. Nuclear Regulatory Commission
' ATTN: . Document Control Desk Washington, D. C. 20555 PLANT HATCH - UNIT 1 NRC DOCKETS 50-321 OPERATING LICENSES DPR-57 1' PLANT STARTUP TEST
SUMMARY
REPORl Gentlemen:
L Per- the requirements of Plant Hatch Unit 1 Technical . Specification !
l 6,9.1.1, Georgia ' Power - Company (GPC) submits . a Plant Startup Test Summary -l b Report. The report presents the results' of the static and dynamic functional- core tests performed during . startup from the Unit 1^
maintenance / refueling (M/R) outage. - The . unit is now-operating in Cycle 12.
The report is being submitted because GPC loaded '108 fuel bundles from General. Electric (GE) that are of the GE-8 design. This fuel bundle design
'has been reviewed and approved by the NRC via' NEDE-24011-P, " General I
Electric Standard Application for Reactor Fuel." Required Technical Specification changes for Plant Hatch were reviewed and approved by the-NRC prior to loading the assemblies. As expected, the testing identified no unexpected behavior of the reactor core and instrumentation.
Please contact this office if you have questions.
Sincerely, y), . /bn,? Y W. G. Hairston, III l
GKM/eb j
Enclosure:
Post Refueling Outage Startup Test Report - Unit 1 Cycle 12 c: (See next page.)
'i 1
- i. 0.'S. Nuclear Regulatory Commission l March 6, 1989 l
Page Two c: Georgia Power Comoany Mr. H. C. Nix, General Manager - Hatch Mr. L. T. Gucwa, Manager Engineering and Licensing - Hatch GO-NORMS Q. S. Nuclear Regulatory Commission. Washinaton. D.C.
Mr. L. P. Crocker, Licensing Project Manager - Hatch U. S. Nuclear Regdatory Commission. Region II Mr. M, L. Ernst, Acting Regional Administrator Mr. J. E. Menning, Senior Resident Inspector - Hatch
_. . . .