ML20217B016
| ML20217B016 | |
| Person / Time | |
|---|---|
| Site: | Mcguire |
| Issue date: | 03/31/1998 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20217B004 | List: |
| References | |
| NUDOCS 9803250328 | |
| Download: ML20217B016 (43) | |
Text
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Duke Power Company McGuire Nuclear Station Unit 2 Cycle 12 STARTUP REPORT March 1998 0R ADO OOOb70 P PDR
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TABLE OF CONTENTS EA98 List of Tables . . . . . . . . . . . . . . . . . . . . ............................... ...........~....................-- ... ...li List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................... .......................iii 1.0 Introduction .... .... . .. . . . . . . . . . . . ......................... ......,.........1 2.0 - Precritical Testing ............. ............. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -2:
-2.1' Total Core Reloading ....... .... ... .........-.............................. . .. .. 2 2.2 Preliminary NIS Calibration..... .... ........ . .....................................2 3.0 ' Zero Power Physics Testing..... ..... .. .. . .... .. .... ... .... . . . . . 4 3.1 1/M Approach to Criticality = . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...............4 3.2 Source Range / Intermediate Range Overlap Data ... ... .............7 3.3 Point of Nuclear Heat Addition ..... . .... .. . ... .... .. .. ..... ... ..... .... . ..................7 3.4 Reactivity Computer Checkout...... .. .. . . . . . .
.,..........8 3.5 Control Rod Worth Measurement by Dynamic Rod Worth.. .. .8 3.6 ARO Boron Endpoint Measurement- .. ...............9 3.7 IsothermalTemperature Coefficient Measurement..... . ... . . . .10 4.0 Power Escalation Testing .. . .... .. ..... . . . . . . . . . . . ...................11 4.1 Core Power Distribution.. ....... . ....................11 4.2 One-Point incore/Excore Calibration...... ...... . . . . . . . . . . .. ..........15 4.3 Reactor Coolant Delta Temperature Measurement. . . . . .....................15 4.4 . Hot Full Power Critical Boron Concentration Measurement . .. .... . .. -16 4.5 incore/Excore Calibration.... . ... . .. . .. . . . . . . . . . . . . . .16 4.6 Unit Load Steady State Test. .. .... ....... . .............................17 4.7 Unit Load Transient Test... ... . .. .. . . . . . . . . . . .... ...............17 4.8 Replacement S/G Tuning and Testing of 0FCS.. ....... . . . ........................38 4.9 Reactor Coolant System Flow Measurement......... ......... ............. ........ .. ........ ... 39 5.0 . Natural Circulation Test.... ....... ....... ...... .........................................................39 m
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. , LIST OF TABLES Eage
- 1. Core Design Data .... . . . . . . . . . . . . . . . .. ..... ........................1
- 2. Preliminary NIS Calibration Data.......... ......................................... .3
- 3. ' Summary of Zero Power Physics Testing Results....... . .. .. .... ... .. ..... .... . . ... ..... ...... . . . . ... 5
- 4. Source Range / intermediate Range Overlap Data... ........... . ... .. 7
- 5. Nuclear Heat Determination ............... ......... ...... ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..........8
- 6. Reactivity Computer Checkout.. ... ..... ............... .............................-.......~.8
- 7. ITC Measurement Results............... . ................... . . . . . . . . . . . . ....... . 10
- 8. Core Power Distribution Results,30% Power. . . . . . . . . ......... ... 12
- 9. Core Power Distribution Results,71% Power.. . .... .. ........ .... ..... ... ... . . . . . . . . . . . . . .. 13
- 10. Core Power Distribution Results,100% Power..... .... .... . . . . .. . ... . . . . . . . . . . . . . . . . 14
- 11. Reactor Coolant Delta Temperature Data.. .... .............. . 15
- 12. Incore/Excore Calibration Results .. .......... .. .... .. .. . .... ......... ... ...... .. ... ..... .. . . . .. .... .... . .... 17
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LIST OF FIGURES EA92
- 1. Inverse Count Rate Ratio vs. Control Rod Worth During Approach to Criticality.... ... ....... .. 6
- 2. Unit Load 10% Transient Test,78% Power - U2 Reactor Thermal Power, Best.................19
- 3. Unit Load 10% Transient Test,78% Power - U2 Power Range Average Lwel.......... .... . 20 4.' Unit Load 10% Transient Test,78% Power U2 Generator MW... ................................. 21
- 5. Unit Load 10% Transient Test,78% Power- NC Loop Average Temp......... ...... . ......... 22
- 6. Unit Load 10% Transient Test,78% Power - NC Loop " Actual" D/T.. .. .... .................. .. 23
- 7. Unit Load 10% Transient Test,78% Power - NC Loop Highest Ave Temp and T-Ref.. ....... 24
- 8. Unit Load 10% Transient Test,78% Power - U2 Pressurizer Press............ ... .... ........ . 25
- 9. Unit Load 10% Transient Test,78% Power - U2 Pressurizer Level....... . ... .................. 26
- 10. Unit Load 10% Transient Test,78% Power - 2A S/G Narrow Range Level......... ....... ... . 27
- 11. Unit Load 10% Transient Test,78% Power- 2B S/G Narrow Range Level........... ........ .. 28
- 12. Unit Load 10% Transient Test,78% Power 2C S/G Narrow Range Level....................... 29
- 13. Unit Load 10% Transient Test,78% Power - 2D S/G Narrow Range Level....... ............. . 30
- 14. Unit Load 10% Transient Test,78% Power - S/G Average Steam Pressure.. .... ... ....... . 31
- 16. Unit Load 10% Transient Test,78% Power - U2 Turbine impulse Chamber Pressure......... 33
- 19. ' Unit Load 10% Transient Test,78% Power - 2C S/G Feedwater Flow............... ........ .... 36
- 20. Unit Load 10% Transient Test,78% Power - 2D S/G Feedwater Flow.... .. . ........ ......... 37 iii O
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~ 1.0 ' INTRODUCTION McGuire Unit Two Cycle 12 includes a feed batch of 68 MkBW fuel assemblies manufactured by Framatome Cogema Fueir. (FCF). The feed batch enrichments are 3.78% (w/o) and axially blanketed with 2.0%(w/o) upper and lower blankets. Burnable poison rod assemblies used in the feed batch were also manufactured by FCF.
McGuire Unit Two Cycle 12 core loading began at 1037 on December 1,-1997 and ended at 1941 on December 3,1997. Initial criticality for Cycle 12 occurred at 0127 on December 17,1997. Zero Power Physics Testing was completed at 1701 on December 17, 1997. The unit reached full power at on December 23,1997. Power Escalation testing, including testing at full power, was completed by December 31,1597.
Table 1 contains some important characteristics of the McGuire 2 Cycle 12 core design.
TABLE 1 M2C12 CORE DESIGN DATA
- 1. M2C11 end of cycle burnup: 409.5 EFPD
- 2. M2C12 design length: 410 10 EFPD Region Fuel Type Number of Enrichment, Loading, Cycles Bumed Assemblies w/o U" MTU" 12A MkBW 24 3.85 10.9488 2 128 MkBW 25 3.95 11.4050 13A MkBW 48 3.90 21.8976 1 138- MkBW 28 4.15 12.7736 1 14A- MkBW 68 3.78/2.00* 31.0216 0 Totals 193 $8.0466
' 2.00 w/o enriched U blanketed fuel assemblies (6 inches top and bottom)
Design MTU loadings which were used in all design calculations.
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Page 2 of 39 2.0 . PRECRITICAL TESTING Precritical testing includes:
. Core Loading
. Preliminary Calib.ation of Nuclear instrumentation Sections 2.1 and 2.2 describe results of precritical testing for McGuire 2 Cycle 12.
2.1 Total Core Reloading I
The Cycle 12 core was loaded under the direction of PT/0/A/4150/033, Total Core Reloading. Plots of j l
Inverse Count Rate Ratio (ICRR) versus number of fuel assemblies loaded were maintained for each source range channel.
Core loading commenced at 0052 on Deember 1,1997 and concluded at 1941 on December 3,1997. Core loading was verified by PT/0/A/4550/003C, Core Verification.
2.2 Preliminary NIS Ca;ibration Periodic test procedure PT/0/A/4600/78, Prestartup NIS Realignment Following Refueling, is performed before initial criticality for each new fuel cycle. Intermediate range reactor trip and rod stop setpoints are adjt.sted using measured power distribution from the previous fuel cycle and predicted power distribution for the upcoming fuel cycle. Power Range NIS full power currents are similarly adjusted. Intermediate Range NIS Rod Stop and Reactor Trip setpoints are checked and revised as necessary for initial power ascension.
Westinghouse standard importance factors and conservatisms were applied in determining the setpoints.
This aided in accounting uncertainties that may have been introduced by the T-AVG reduction resulting from replacement of the Steam Generators during the outage.
Table 4 shows the calibration data calculated by PT/0/A/4600/78. Calculations were performed on August 27,1997. Calibrations were complete by December 15,1997.
Evaluation of the setpoints at 25% and 30% power in accordance with PT/0/A/4150/21, Post Refueling Cantrolling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing, indicated these calculated setpoints were conservative.
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- - TABLE 2 PRELIMINARY NIS CALIBRATION DATA Intermediate Range Ratio BOC 12 BOC 12 Channel (BOC 12 High Flux High Flux
+ Cycle Reactor Trip Rod Stop
pAmps pAmps N35 0.623 44.85 35.89 N36 0.726 54.03 43.22 Power Range Ratio Axial Offset, Cycle 11 Full Power BOC 12 Full Power Channel (BOC 12 + %
Current, pAmps Current, pAmps Cycle 11)
Upper Lower Upper Lower
+30 224.1 151.5 166.7 118.0 N41 0.732 0 187.0 202.1 136.1 148.9 30 149.8 245.7 105.6 179.8 i
+30 315.4 227.4 237.9 170.3 N42 0.740 0 263.6 285.1 193.9 212.4
-30 211.6 342.9 150.0 254.6
- +30 275.0 206.8 208.2 159.4 N43 0.743 0 232.0 260.2 172.0 194.0
-30 189.0 313.6 135.8 233.7
+30 291.2 213.7 217.7 159.0 N44 0.738 - 0 244.4 267.1 179.4 198.4 30 197.6 320.6 141.2 237.7
Page 4 of 39 3.0 ZERO POWER PHYSICS TESTING Zero Power Physics Testing (ZPPT) is performed at the beginning of each cycle and is controlled by PT/0/A/4150/21, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing. Test measurements are made below the Point of Nuclear Heat using the output of one Power Range NIS detector connected to a reactivity computer. Measurements are compared to predicted data to verify core design. The following tests / measurements are included in the 2. PPT program:
- 1/M Approach to Criticality
- Measurement of Point of Adding Heat
- Reactivity Computer checkout
- Control Rod Worth Measurement by Dynamic Rod Worth
- All Rods Out Cntical Boron Concentration measurement (Boron Endpoint) alsothermal Temperature Coefficient measurement Zero power physics testing for McGuire 2 Cycle 12 began at 0611 on December 17,1997. ZPPT ended at 1701 on December 17,1997. Table 3 summarizes results from ZPPT. All acceptance criteria were met.
Sections 3.1 through 3.7 describe ZPPT measurements and results.
3.1 1/M Approach to Criticality l f
Initial criticality for McGuire 2 Cycle 12 was achieved per PT/0/A/4150/28, Criticality Following a Change in j Core Nuclear Characteristics. In this procedure, an Estimated Critical Rod Position (ECP) is calculated based on latest available Reactor Coolant boron concentration. Control rods are withdrawn 50 to 60 steps at a time while monitoring source range channel response. Inverse Count Rate Ratio (ICRR) is plotted for each source range channel. ICRR data is used to project critical rod position. If projected critical rod position is acceptable, rod withdrawal may continue.
Rod withdrawal for the approach to criticality began at 2347 on December 16,1997. Criticality was achieved at 0127 on December 17,1997 with Control Bank D at 182 steps withdrawn.
Figure 3 shows the ICRR behavior during the approach to criticality. All acceptarce criteria of f PT/0/A/4150/28 were met.
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Page 5 of 39
- TABLE 3
SUMMARY
OF ZPPT RESULTS PREDICTEDVALUE OR PARAMETER MEASURED VALUE ACCEPTANCE CRITERIA Nuclear Heat 3.87 x 10' amps N/A ZPPT Test Band below 3.87 x 10' amps (N41) N/A ARC Critical Boron 2037 ppmB 2036 50 ppmB ITC -1.205 pcnfF -0.99 2 pcmrF MTC +0.465 penfF +0.68 pcmrF Control Bank D Worth 646.5 pcm 614
- 100 pcm Control Bank C Worth 790.2 pcm 812 100 pcm Control Bank B Wortu 687.8 pcm 644 100 pcm Control Bank A Worth 300.9 pcm 337 100 pcm Shutdown Bank E Worth 501.4 pcm 5061100 pcm Shutdown Bank D Worth 529.9 pcm 491 100 pcm Shutdown Bank C Worth 530.2 pcm 490 2100 pcm Shutdown Bank B Worth 1105.4 pcm 'iO40 2156 pcm Shutdown Bank A Worth 314.2 pcm 288 100 pcm Total Rod Worth 5406.5 pcm 5221.5 522 pcm
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Page 7 of 39
( 3.2 . Source Range / intermediate Range Overlap Data During the initial approach to criticality, Source Range and Intermediate Range NIS data was obtained to verify the existence of at least one decade of overlap. If one decade of overtap did not exist, intermediate range compensation voltage would have been adjusted to provide the overlap.
Overlap data for Cycle 12 was obtained per PT/0/A/4150/028, Criticality Following a Change in Core Nuclear Characteristics, on December 17,1997. Table 4 contains the overlap data. The acceptance criterion was met.
TABLE 4 SOURCE RANGE / INTERMEDIATE RANGE OVERLAP DATA SOURCE RANGE INTERMEDIATE RANGE N31, cps N32, cps N35, amps N36, amps INITIAL DATA:
NIS Meters 1100 1000 1.2 x 10 '" 2 x 10 '"
OAC 1050 963 1.148 x 10'" 1.958 x 10~"
FINAL DATA:
NIS Meters 15,000 15,000 1.5 x 10 1.5 x 10 "'
OAC 11688 8690 1.210 x 10 "' 1.479 x 10 "'
3.3 Point of Nuclear Heat Addition The Point of Nuclear Heat Addition is measured by trending Reactor Coolant System temperature, Pressurizer level, flux level, and reactivity while slowly increasing reactor power. A slow, constant startup rate is initiated by rod withdrawal. An increase in Reactor Coolant System temperature and/or Pressurizer level accompanied by a decrease in reactivity as calculated by reactivity computer and/or rate of flux increase indicates the addition of Nuclear Heat.
For Cycle 12, the Point of Nuclear Heat Addition was determined per PT/0 A/4150/021, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Esca! mon Testing, on December 17, 1997 Table 5 summarizes the data obtained.
The Zero Power Physics Test Band was set below 3.87x10' amps on Power Range channel N41
- (connected to reactivity computer). Acceptance criterion was satisfied.
Page 8 of 39 TABLE 5 NUCLEAR HEAT DETERMINATION Intermediate Range Intermediate Range Channel N35, amps Channel N36, amps RUN #1 3.73 x 10" 3.73 x 10
RUN #2 4 x 10~' 4 x 10' 3.4 Reactivity Computer Checkout The reactivity computer checkout was performed per PT/0/A/4150/21, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing, to verify that the Power Range channel connected to the reactivity computer could provide reliable reactivity data. Reactivity insertions of approximately +25 and +40 pcm are made. The resulting Periods are measured and used to determine the corresponding theoretical reactivities. The measured reactivity is compared to the theoretical reactivity and verified to be within 4.0% or 1pcm.
The checkout was performed for Cycle 12 on December 17,1997. Table 6 lists the results of the reactivity insertion. The acceptance criterion was met.
TABLE 6 REACTIVITY COMPUTER CHECKOUT Period, seconds Theoretical Reac. Measured Reac- Absolute Error, Percent Error,%
tivity, pcm tivity, pcm pcm 147.0 39.9 39.8 0.1 -0.18 181.9 33.3 32.7 0.6 -1.91 3.5 Control Rod Worth Measurement by Dynamic Rod Worth The worth of all control Banks are measured using the Dynamic Rod Worth Methodology. With Control Bank D inserted less than 75 pcm, the bank is fully withdrawn and flux is allowed to increase. Once flux reaches a predetermined level, the bank to be measured is fully inserted. Upon completion of data gathering, the bank is fully withdrawn. The measurement is repeated for all temaining banks.
The measured worth is compared to the predicted worth to verify design and measurement accuracy. The sum of the worths of all banks is compared to be within 90% of the predicted total bank worth.
> The beginning of Cycle 12 rod worths were measured on December 17,1997 using PT/0/A/4150/011B, Control Rod Worth Measurement: Dynamic Rod Worth. Table 3 summarizes the results, all acceptance criteria were rnet.-
. .. .. .. .. . . .)
Page 9 of 39 3.6 ARO Boron Endpoint Measurement This test is performed at the beginning of each cycle to verify that measured and predicted total core reactivity are consistent. The test is performed at the all rods out (ARO) configuration. Reactor Coolant System boron samples are obtained prior to withdrawing Control Bank D. Subsequently Control Bank D is fully withdrawn -while measuring the reactivity worth of this rod movement. The reactivity difference from criticality to the ARO configuration is then converted to an equivalent boron worth using the predicted differential boron worth. The average measured boron concentration is adjusted accordingly to obtain the ARO critical boron concentration.
The Cycle 12 beginning of cycle, hot zero power, all rods out, critical boron concentration was measured on December 17,1997 per PT/0/A/4150/10, Boron Endpoint Measurement. The ARO, HZP boron concentration was measured to be 2037 ppmB. Predicted ARO critical boron concentration was 2036 ppmB. The acceptance criterion, measured boron within 50 ppmB of predicted, was met.
Page 10 of 39 3.7 isothermal Temperature Coefficient Measurement
- The isothermal Temperature Coefficient (lTC) is measured at the beginning of each cycle'to verify consistency with predicted value, In addition, the Moderator Temperature Coefficient (MTC) is obtained by subtracting the Doppler Temperatura Coefficient from the ITC. The MTC is used to ensure compliance with Technical Specification limits.
To measure the ITC, a Reactor Coolant System cooldown is initiated, within administrative cooldown limits. The RCS temperature must change by 1.4 *F. Upon stabilization of spatial reactivity effects, the reactivity computer automatically samples data and perforrns an on-line ITC calculation. The test is terminated when the ITC error is s 0.1 perrfF. A heatup is performed using the same methodoingy while maintaining administrative limits. The cooldown/heatup cycle is repeated to obtain confidence and repeatability of data.
The Beginning of Cycle 12 ITC was measured per PT/0/A/4150/012, Isothermal Temperature Coefficient Measurement, on December 17,1997. Table 7 summarizes the data obtained during the measurement.
Average ITC was determined to be -1.205 pcm/ F. Predicted ITC was -0.99 perrfF. Measured ITC was therefore within acceptance criterion of predicted ITC 2 perrfF.
The MTC was determined to be +0.465 perrfF. This value was used with procedure PT/0/A/4150/031, Determination of Temporary Rod Withdrawal Limits to Ensure Moderator Temperature Coefficient Within Limits of Technical Specifications, to ensure that MTC would remain within Technical Specification limits at all power levels. No rod withdrawallimits were required.
TABLE 7 ITC MEASUREMENT RESULTS ITC, penfF AT, T ,,'F
- F Cooldown 1.6 555.5 -1.36 Heatup 1.6 555.< -1.05 Average: -1.205
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Page 11 of 39 4.0 POWER ESCALATION TESTING Power Escalation Testing is performed during the initial power ascension to full power for each cycle and is controlled by PT/0/A/4150/021, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing. Tests are performed from 0% through 100% power with major testing plateaua at -30%,-75%, and 100% power.
Significant tests performed during McGuire 2 Cycle 12 Power Escalation were:
. Core Power Distribution (at ~30%, -78%, and 100% power) e One-Point incore/Excore Calibration (at ~30% power) e Reactor Coolant Delta Temperature Measurement (at 90% and 100% power) e Hot Full Power Critical Boror, Concentration Measurement (at 100% power)
- Incore/Excore Calibration (at 100% power) e Calorimetric Reactor Coolant Flow Measurement (at 100% power, This test is not under the controlof PT/0/A/4150/21) e Unit Load Transient Test - at 38% and 78% (Steam Generator Replacement Post-Mod testing) e Replacement S/G Functional Tuning and Testing of DFCS - at 10%
(Steam Generator Replacement Post Mod testing)
. Evaluation of intermediate Range NIS Rod Stop and Rx Trip Setpoints Power Escalation Testing for McGuire 2 Cycle 12 began on December 29,1997. The unit reached 98%
RTP on December 31 and increased to 100% RTP by January 2,1998. Full power testing was completed on December 31,1997. Sections 4.1 through 4.9 describe the significant tests performed during power escalation and their results.
4.1 Core Power Distribution Core power distribution measurements are performed during power escalation at low power (approximately 30%), intermediate power (approximately 75%), and full power. Measurements are made to verify flux symmetry and to verify core peaking factors are within allowable limits. Data obtained during this test is also used to check calibration of Power Range NIS channels and to calibrate them if required (see sections 4.2 and 4.6). Measurements are made using the Moveable incore Detector System and analyzed using Duke Power's COMET code (adapted from the Shangstrom Nuclear Associates' CORE package and the FCF MONITOR code).
The McGuire 2 Cycle 12 Core Power Distribution measurements were performed on December 20,1997 (29% power), December 23,1997 (78% power), and December 29,1997 (98% power). Tables 8 through 10 summarize the results. All acceptance criteria were met.
Page 12 of 39 TABLE 8 CORE POWER DISTRIBUTION RESULTS 30% POWER Plant Data Map ID: m2c12f001 Date of Map: December 20,1997 Cycle Burnup: 0.3 EFPD Power Level: 29.02% F.P.
Control Rod Position: Control Bank D at 200 Steps Wd Reactor Coolant System Boron Concentration: 1868 ppmB COMET Results Core Average AxialOffset: 12.339%
Tilting Factors for Entire Core Height: Quadrant 1: 1.00371 Quadrant 2: 0.99312 Quadrant 3: 1.00798 Quadrant 4: 0.99519 Maximum Fo (nuclear): 2.084 Maximum Fu (nuclear): 1.531 Maximum Error between Pred. and Meas Fu: 4.92%
Mean Error between Pred. and Meas. Fu: 1.81 %
Maximum Error between Expected and Measured 4.99 %
Detector Response :
RMS of Errurs between Expected and Measured 2.5%
Detector Response:
MONITOR Results Minimum F Operational Margin: 34.08 %
Minimum Fo RPS Margin: 15.23 %
Minimum Fo LCO Margin: 54.93 %
Minimum Fu Surveillance Margin: 28.15%
Minimum Fu LCO Margin: 24.59 %
Page 13 of 39 TABLE 9 CORE POWER DISTRIBUTION RESULTS 78% POWER Plant Data Map ID: m2c12f002 Date of Map: December 23,1997 Cycle Burnup: 1.4 EFPD Power Level: 77.75% F.P.
Control Rod Position: Control Bank D at 201 Steps Wd Reactor Coolant System Boron Concentration: 1653 ppmB COMET Results Core Average Axial Offset: 2.020 %
Tilting Factors for Entire Core Height: Quadrant 1: 1.00299 Quadrant 2: 1.00006 Quadrant 3: 0.99972 Quadrant 4: 0.99722 Maximum Fo (nuclear): 1.840 Maximum FS (nuclear): 1.468 Maximum Error between Pred. and Meas FS : 5.81 %
Mean Error between Pred. and Meas. FS : 1.38 %
Maximum Error between Expected and Measured 6.44 %
Detector Response:
RMS of Errors between Expected and Measured 2.3%
Detector Response: _
MONITOR Results Minimum Fo Operational Margin: 18.26%
Minimum Fo RPS Margin: 12.15 %
Minimum Fo LCO Margin: 38.13 %
Minimum FS Surveillance Margin: 18.27 %
Minimum FS LCO Margin: 21.44 %
Page 14 of 39 TABLE 10 CORE POWER DISTRIBUTION RESULTS 100% POWER Plant Data Map ID: m2c12f004A Date of Map: December 29,1997 Cycle Burnup: 7.7 EFPD Power Level: 97.76% F.P.
Control Rod Position: Control Bank D at 210 Steps Wd Reactor Coolant System Boron Concentration: 1420 ppmB COMET Results Core Average Axial Offset: 1.106 %
Tilting Factors for Entire Core Height: Quadrant 1: 1.00170 Quadrant 2: 1.00097 Quadrant 3: 1.00258 Quadrant 4: 0.99474 Maximum Fo (nuclear): 1.792 Maximum Fu (nuclear): 1.461 Maximum Error between Pred. and Meas F S: 3.65%
Mean Error between Pred. and Meas. FS : 1.24 %
Maximum Error between Expected and Measured 3.8%
Detector Response:
RMS of Errors between Expected and Measured 1.8%
Detector Response:
MONITOR Results Minimum FoOperational Margin: 6.56 %
Minimum Fo RPS Margin: 15.35%
Minimum Fo LCO Margin: 24.17 %
Minimum Fu Surveillance Margin: 6.67 %
Minimum FS LCO Margin: 17.27 %
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Page 15 of 39 4.2 One-Point incore/Excore Calibration PT/0/N4600/002F, One-Point incore/Excore Calibration, is performed using results of Power Range NIS data taken at 30% power and the incore axial offset measured at 30%. Power Range channels are calibrated before exceeding 50% in ordar to have valid indications of Axial Flux Difference and Quadrant Power Tilt Ratio for subsequent power ascension. The calibration is checked using the intermediate power level flux map (78% F.P. for M2C12). If necessary, Power Range NIS is recalibrated per PT/0/N4600/002F or PT/0/N4600/002G, incore and NIS Recalibration.
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Data for McGuire 2 Cycle 12 was obtained on December 20,1997 and all Power Range NIS calibrations were completed. All acceptance criteria were met.
4.3 Reactor Coolant Loop Delta Temperature Measurement Reactor Coolant System (NC) Hot Leg and Cold Leg temperature data is normally obtained at approximately 90% and 100% power per PT/0/N4150/40, NC Loop Delta-T, RTAS, and OPDT &OTDT Channel Check Criteria Evaluation, to ensure that full power delta temperature constants (AT ) are valid.
AT,is used in the Over-power and Over-temperature Delta Temperature reactor protection functions.
In the case of M2C12, the four loop AT/s were each preliminarily established at 57.90 F, 56.38 F, 58.87'F, and 56.46 F for Loops A-D, per Steam Generator Replacement Project and previously observed biases. Portions of PT/0/N4150/040 were completed at 90% power on December 23,1997 to verify proper conservatism in the calibrations, and the entire procedure was performed after 100% power equilibrium conditions were achieved. All four NC Loop AT/s were adjusted using full power results. Table 11 summarizes the test results.
TABLE 11 REACTOR COOLANT DELTA TEMPERATURE DATA Reactor Power = 99.67%
Loop A Loop B Loop C Loop D Meas. T., *F 613.8 610.6 613.8 610.9 Meas. T , *F 555.3 554.8 554.5 554.8 Calc. AT , 'F 59.81 56.98 60.60 57.37 i
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Page 16 of 39 4.4 Hot Full Power Critical Boron Concentration Measurement The Hot Full Power critical boron concentration is measured using PT/0/A/4150/04, Reactivity Anomaly Calculation. Reactor Coolant boron concentration is measured (average of three samples) with reactor at essentially all rods out, Hot Full Power, equilibrium xenon conditions. The measured boron is corrected for any off-reference condition (e.g. inserted rod worth, temperature error, difference from equilibrium xenon) and compared to predicted value.
For McGuire 2 Cycle 12, the Hot Full Power critical boron concentration was measured on January 5, 1998. The measured critical boron concentration was 1385.2 ppmB. Predicted critical boron concentration was 1386.7 ppmB. The difference between measured and adjusted predicted critical boron concentration was 1.5 ppmB, which met the acceptance criterion.
. 4.5 incore/Excore Ca;ibration
, Excore NIS Power Range channels are calibrated at full power per PT/0/A/4600/02G, incore and NIS Recalibration. Incore data (flux maps) and Power Range NIS currents are obtained at various axial power distributions. A least squares fit of the output of each detector (upper and lower chambers) as a function of measured incore axial offset is determined. The slopes and intercepts of the fit for the upper and lower chamber for each channel are used to determine calibration data for that channel.
This test was performed for McGuire 2 Cycle 12 on December 29-31, 1997. All Power Range NIS calibrations were completed on December 31. Seven flux maps, with axial offset ranging from -6.882% to
+2.192% were used. Table 12 summarizes the results. All acceptance criteria were met.
Page 17 of 39 TABLE 12 INCORE/EXCORE CALIBRATION RESULTS Full Power Currents, Microamps Axial N41 N42 N43 N44 Offset, Upper Lower Upper Lower Upper Lower Upper Lower .
+30% 207.4 137.4 307.7 203.1 260.3 183.1 273.7 188.2 0% 167.2 180.5 249.7 265.9 213.1 239.9 223.8 244.4
-30% 127.1 223.6 191.6 328.8 165.8 296.6 173.9 300.7 Correction (M,) Factors N41 N42 N43 N44 j
1.253 1.280 1.310 1.324 4.6 BOC12 Unit Load Steady State Test The Unit Load Steady State Test was not performed on Unit 2 due to favorable results on Unit 1.
Standard procedure guidance directed monitoring of significant system parameters per PT/0/A/4150/028, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing.
No adverse trends or unacceptable system response was noted.
4.7 UNIT LOAD 10% TRANSIENT TEST TT/2/A/9815/00/03E, Unit Load 10% Transient Test for NSM MG-29815, was performed to verify proper l operation of the modifications performed on various control systems per NSM MG 29815, Replacement Steam Generator instrumentation and Control. The purpose of the test was to demonstrate proper plant response, including automatic control system performance, to a ~10% step load change (initiated via l
1 l
. . . .- 1
4 3 Page 18 of 39 Turbino/ Generator Control). The fest verifies that the control systems work as designed to prevent ihe following plant trancients (in respore to a ~10% stop load change):
- Actuation of Safety injection
- Pressurizer and Steam Safet.ies or PCRVs Lifting This test satisfies the transient remst as required for the Post Modification Testing for Replacement Steam Generator Instrumentation and Co itrol.
This test was performed from 38% Reactor Power. on Decemoer 22,1997 and from 78% Reactor Power on Occt' mber 23,1997. All acceptance criteria for the test were met as follows
- 1) Reactor did not tr,p
- 2) Turbine did not trip
- 3) Safety injection was no+. initiated
- 4) No Manual Operator intervention was requiied to stabilize the Unit
- 5) Pressurizer PORV's did not lift
- 5) Pressurizer Code Safety Valves did not lift I
- 6) Steam Generator PORV's did not lift
- 7) Steam Generator Code Safety Vanes did not lift
- 8) Monitored plant parameters did not indicate sustained or diverghg oscillations Figures 3 - 21 illustrate response of plant parameters during the 78% Reactor Power test.
I
l 1
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Page 36 of 39 I
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, s. e-Page 38 of 39 4.8 Replacement S/G Tuning and Testing of Foodwater Control TT/2/A/9815//00/04E, Functional Tuning and Testing of the Feedwater Control System, was performed to record the behavior of the S/G Level Controls and the course of action taken to optimally tune the system following installation of the BWI Replacement Steam Generators. Testing was performed at 10% power level. S/G level controls and Feed Pump speed controls were monitored during power ascension. No additional tuning was required. Based on Unit 1 performance, Generator level swings were anticipated during low power operation.
Steam Dump Valves Setup:
Steam Dump Valves were adjusted under hot conditions resulting in a large improvement in level control during a low power operation over Unit 1. Operations rarely had to make an adjustment to the Steam Header Pressure Controls.-
Feedwater Bypass Valve Controllers:
The Feedwater Bypass Valve Controllers were re-tuned to make them more responsive to the level changes in the new Steam Generators. Placing the Feedwater Bypass Valves in Manual and closing, while opening the Main Feodwater Regulating Valves in Manual occurred without any problem. As soon as the Bypass Valve was closed the Main Regulation Valve was placed in Auto. The Feedwater Pump was placed in Auto after the last Main Feedwater Valves were placed in Auto.
Feedwater Pump Speed Demand:
The New Feodwater Pump Speed Control Program (45-175 PSID) had no effect on the controls system at lower power since Operations maintained the Feedwater Pump in manual during this time.175 PSID at 100% power is the correct operating point.
Operations:
Operations used the expanded Wide Range Levels Indication to maintain the proper levet control.
Operations further avoided Steam Generator oscillations by varying only one parameter at a time (Pump Speed control, only manually assisting one valve at a time, adjusting Blow Down Flow, etc.).
Turbine:
Rolling and placing the Turbine on-line between 12% 15% power, generally improved the Primary Temperature decrease that was observed on Unit 1.
Observations and Conclusions
The shift in CF/SM Header D/P resulted in excessive Pump speed with the Bypass or Main regulating Valves compensating by positioning at less than optimal throttling positions.
- 2) - The Feedwater Bypass Valve Controllers had their loop gains reduced from 4.7v/v to 3.0v/v to make them less responsive to level swings.
- 3) The adjusted tuning constants for the Bypass Valve controls were adequate for the Replacement Steam Generators at higher power levels. At startup and lower power levels the level control was poor. Engineering is studying this phenomenon and has plans to thoroughly investigate the associated control circuitry behavior and make further tunirig adjustments as warranted. The issus has been deemed an Operator Workaround and been added to the Major Equipment Problem Resolution Program. Work Orders for additional testing / tuning have been added to the Plant Trip List should the opportunity for further investigation arise.
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4.9 Reactor Coolant Flow Testing PT/2/A/4150/13, NC Flow Calculation, was performed on February 10, 1998 at approximately 99.7%
power. The average reactor coolant system flow (as determined by four one hour test runs, measured by the elbow taps, is 398,479 gallons per minute. The NC flow prior to Steam Generator replacement was approximately 385,147 gpm, or an increase of 13,332 gpm, or approximately 3.5%
5.0 Natural Circulation Verification Test
- On November 29,1997 a Natural Circulation Verification Test was performed on Catawba Nuclear Station Unit 1. This test was performed to verify the ebility of the NSSS to remove heat via natural circu'ation for Catawba Unit 1, McGuire Unit 1, and McGuire Unit 2 after the installation of BWI Replacement Steam Generators.
Method:
While in Mode 3 at End-of-Cycle 10 refueling outage, Catawba Nuclear Station Unit 1 tripped all four NC Pumps to verify Natural Circulation. Natural Circulation was verified by Wide Range loop temperatures as well as core exit Thermocouples. Pressurizer and Steam Generator Pressures and levels response were monitored throughout this test as wait. Stable natural circulation was maintained for approximately 30 minutes while data was gathered to verify the Acceptance Criteria and Review Criteria.
Results:
All Acceptance Criteria for this test were successfully met as described below:
- 1. No Safety injection occurred or was required. Pertinent parameters remained within all safety guidelines specified within the procedure.
- NC Subcooling remained greater than 15 F.
- Steam Generator Pressure remained stablu or decreasing.
- NC T-Hots remained stable or decreasing.
- NC T-Colds remained at the saturation temperature for the Steam Generator pressure.
- 2. Establishment of the natural circulation was verified by the simultaneous achievement of the following conditions over an approximately 30 minute interval.
- The highest calculated loop AT remained s 45 F.
- The highest loop AT changed s 13 F during the test interval.
- The AT between the 5-highest incore Thermocouples readings and lowest Wide Range T-Cold reading remained s 45 F.
- The AT between the 5-highest incore Thermocouple readings and the lowest Wide Range T-Cold reading stabilized such that a s t 3 F change occurred during the test interval.
- The NC Subcooling margin remained 2 25 F.
- Pressurizer pressure remained at 2235125 psig during the test interval.
- Pressurizer level decreased as a result of the RCS shrinkage. The lowest observed level was 23.4 %
- All Steam Generator levels remained above 25% (Narrow Range).
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