ML20210K928

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Cycle 12 Startup Rept
ML20210K928
Person / Time
Site: McGuire Duke Energy icon.png
Issue date: 08/31/1997
From:
DUKE POWER CO.
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ML20210K927 List:
References
NUDOCS 9708200169
Download: ML20210K928 (49)
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Text

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i Duke Power Company McGuire Nuclear Station Unit 1 Cycle 12 ,

STARTUP REPORT August 1997 hDR A 050 369 P PDR

1. .

.) .

TABLE OF CONTENTS EAge List of Tables .......................................... .................. ...........................................11 Listof Figures ........................................................................... ............................................1i1 1.0 Introduction ...................... ........................................................ .... . .. 1 2.0 : Procritical 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 l 3.4 Reactivity Computer Checkout........ ...................................... =, 8 --

1 l - 3.5 ARO Boron Endpoint Measurement....................., . ..................8 3.6 ARO isothermal Temperature Coefficient Measurement, ...................................9 3.7 Reference Bank Worth Measurement by Dilution...... ......... .9 3.8 - Differential Boron Worth Determination........................ ....................................11 3.9 Control Rod Worth Measurement by Rod Swap.................................................. 11  ;

i 4.0 Power Escalation Testing ................. ...................... .... . . . ... .. 13 4.1 Core Power Dist ribution .............................. ......... .... .................. ...... ...... ................13 4.2

' One-Point incore/Excore Calibration...... ............. ................................17 4.3 Reactor Coolant Delta Temperature Measurement....... ............ ... 17 4.4 Hot Full Power Critical Boron Concentration Measurement. .................................17

- 4.5 - Incore/Excore Calibration ................................... . . .

..................18 4.6 Unit Load Steady State Tests ................... ...............................19

' 4.7 U nit Load Trans ient Test........... ........................... ............... ........... ........ . . ...... . . . . . . . . ., ..... ... 23

..4.8 , Replacement S/G Tuning and Testing of DFCS.......... ............................. ........ ............. .. 44 4,9 Reactor Coolant System Flow Measurement........... ... .. .......... ... ...... ............ ........... 45 I

i J/

  • O 4

LIST OF TABLES Page

1. Core Design Data..... ..... ............................................. ..................1
2. Prelimina ry NI S 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. ReactMty Compute r Checkout.............. .. ........... .. . ..... ...... ........ .... ............. ..... . ... . ............. .8
7. ITC Measurement Results . .. ........... .............................................................................9
8. Control Rod Worth Measurement Data.......... .............. .............................................12 a

Core Power Distribution Results,30% Power ................. .. ............. ...............15

' ore Power Distribution Results, 71 % Power .............................. ......... ...... ...., . ..........16

' ore Pcwer Distribution Resuits.100% Power . .. ...... ...... .... ........................ ............. ......... ... ... 17 1 .*., rieactor Coolant Delta Temperature Data .......... ..........................................................,.18 l 13. Incore/Excore Calibration Results.... .......... ......

l

....... ........ . .. 20

14. NC Loop Cold Leg Temperatures............ ......................... . .. . . . . ....

l ......................20

15. N C Loop Hot Leg Temperature s.... . .. ...... .. .... . . .................... . . . . . . . . . .......... . . ...........  ;;20
16. NC Loop Average Temperatures....................... .. .........................................................21
17. NC Loop Delta Temperatures... .................................................................... 21 18.- Pressurizer Level Data..... . ........ . .. . .. ........e.. .... ............ ............... ................21
19. Tu rbine Gove mor Valve Positions ....... ...... . . . . .. . .. .. ................. . ......... . ....................................., ... 21
20. Tu rbine impulse Pressure......... ... . ....... .......... ... ....... , . ................22
21. S/G Na rrow Range Level....... ...... ..... .................. .... ....... . . . . . . . ....... ..............................22 22, S/G " in Steam Pres su re. . . . . . ...... ............ ........ ....... .. ....... ... . . . . . . . ......... . ..... 22 23.

S/G M a in Ste am Flow . . . ...... ... . . . . . . . . . .. .. . .. .. . .. .. . .. . . . . . . . . . . .. . . .. ... .... . . . . . . . . . .. . . . . . . . .. . . . . . . .. . . . ..

24.

S/G Feedwate r Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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< 1 LIST OF FIGURES Eage

1. Inverse Count Rate Ratio vs. Control Rod Worth During Approach to Criticality.................... 6
2. Integral and Differential Worth of Reference Ban...........................................................10
3. Unit Load 10% Transient Test,78% Power U1 Reactor Thermal Power, Best................. 25 4, Unit Load 10% Transient Test,78% Power U1 Power Range Average Level................-. 26
5. Unit Load 10% Transient Test,78% Power U1 Generator MW..................................... 27
6. Unit Load 10% Transient Test,78% Power - NC Loop Average Temp............................. 28
7. Unit Load 10% Transient Test,78% Power NC Loop " Actual
  • D4................................ 29
8. Unit Load 10% Transient Test,78% Power NC Loop Highest Ave Temp and T Ref.......... 30
9. Unit Load 10% Transient Test,78% Power U1 Pressurizer Press................................ 31
10. Unit Load 10% Transient Test,78% Power - U1 Pressurizer Level................................. 32
11. Unit Load 10% Transient Test,78% Power 1 A S/G Narrow Range Level.... .................. 33 12, Unit Load 10% Transient Test,78% Power 1B S/G Narrow Range Level....................... 34 l 13. Unit Load 10% Transient Test,78% Power 1C S/G Narrow Range Level....................... 35 -
14. Unit Load 10% Transient Test,78% Power 1D S/G Narrow Range Level....................... 36 15.' Unit Load 10% Transient Test,78% Power - S/G Average Steam Pressure...................... 37
16. Unit Load 10% Transient Test,78% Power - CF Pump Speed....................................... 38 17, Unit Load 10% Transient Test,78% Power U1 Turbine impulse Chamber Pressure......... 39
18. Unit Load 10% Transient Test,78% Power - 1 A S/G Feedwater Flow............................. 40
19. Unit Load 10% Transient Test,78% Power - 1B S/G Feedwater Flow. .... ..................... 41
20. Unit Load 10% Transient Test,78% Power 10 S/G Feedwater Flow............................. 42
21. Unit Load 10% Transient Test,78% Power 1D S/G Feedwater Flow............................. 43 iii

.c Page 1 of 45

1.0 INTRODUCTION

McGuire Unit One Cycle 12 includes a feed batch of 68 MkBW fuel assemblies manufactured by Framatome Cogema Fuels (FCF). The feed batch enrichments a,e 3.67% (w/o). Burnable poison rod assemblies used in the feed batch were also manufactured by FCF.

McGuire Unit One Cycle 12 core loading began at 1208 on April 24,1997 and ended at 1758 on April 26, I 1997, initial criticality for Cycle 12 occurred at 0244 on May 18,1997. Zero Power Physics Testing was completed at 0155 on May 19,1997. The unit reached full power at 2306 on May 31,1997. Power 3 Escalation testing, including testing at full power, was completed by June 9,1997.

Table 1 contains some important characteristics of the McGuire 1 Cycle 12 core design.

TABLE 1 M1C12 CORE DESIGN DATA

1. M1C11 end of cycle bumup: 361 EFPD
2. MIC12 design length: 370
  • 10 EFPD Region Fuel Type Number of Enrichment, Loading, Cycles Bumed Assemblies w/o U " MTU" 11A MkBW 1 3.45 0.4562 3 12A MkBW 52 3.60 23.7224 2 13A MkBW 48 3.40 21.8976 1 138 MkBE 24 3.55 10.9488 1 14A MkBW 68 3.67/2.00* 31.0216 0 Totals 193 88,0466 2.00 w/o enriched U blanketed fuel assemblies (6 inches top and bottom)

Design MTU loadings which were used in all design calculations.

i Page 2 of 45 2.0 PRECRITICAL TESTING

Procritical testing includes

. Core Loading

  • Preliminary Calibration of Nuclear Instrumentation Sections 2.1 through 2.2 describe results of precritical testing for McGuire 1 Cycle 12.

2.1 Total Core Reloading The Cycle 12 core was loaded under the direction of PT/0/A/4150/33, Total Core Reloading. Plots of Inverse Count Rate Ratio (ICRR) versus number of fuel assemblies loaded were maintained for each source rango channel.

Core loading commenced at 1208 on April 24,1997 and concluded at 1758 on April 26,1997. Core loading was verified by PT/0/A/4550/03C, Core Verification, which was completed by 2200 9n April 26,1997, 2.2 Preliminary NIS Calibration 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 adjusted 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 Raactor Trip setpoints are checked and revised as necessary for initial power ascension.

Added conservatism was applied to setpoints to account for any uncertainties that may have baen introduced by the T-AVG reduction resulting from replacement of the Steam Generators during the outa; e.

Tabl' 4 shows the calibration data calculated by PT/0/Ai4600/78. Calculations were performed r n February 19,1997. Calibrations were complete by April 30,1997.

Evaluation of the setpoints at approximately 30% power in accordance with PT/0/A/4150f21, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing, indicated these calculated setpoints were conservative.

i t

w _ _ _ _ _ _ _ . _ _ .

Page 3 of 45 TABLE 2 PRELIMINARY NIS CAllBRATION DATA Intermediate Re ege Ratio BOC 12 BOC 12 Channel (BOC 12 +- Reactor Trip Rod Stop Cycle 11) Setpoint, Setpoint, pAmps pAmps N35 0.698 41.88 33.51 N36 0.694 43.29 34.63 Power Range Ratio Axial Offset, Cycle 11 Full Power BOC 12 Full Power Channel (BOC 12 +  %

Cycle 11)- Current, pAmps Current, pAmps Upper Lower Upper - Lower

+30 262.7 183.8 187.9 131.4 N41 0.893 0 -217.5- 233.2 155.3 166.8

-30 172.4- 282.6 123.3 202.1

+30 279.2 190.8 199.4 136.3 N42 0.893 0 231.2 240.7 165.1 171.9

-30 183.3 290.6 130.9 207.5 430 264.8 191.2 189.3 - 136.7 N43 0.893 0 221.9 245.1 158.6 175.2 30 179.0 298.9 I 127.9 213.7

+30 - 259.3 185.2 183.7 131.2 N44 0.885 0 213.0 238.1 150.9 168.7

-30 166.7 291.0 118.1 206.2-

.?

Page 4 of 45 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 credicted data to verify core design. The following tests / measurements are included in the ZPPT program:

+1/M Approach to Criticality

  • Measurement of Point of Adding Heat
  • Reactivity Computer checkout
  • All Rods Out Critical Boron Concentration measurement (Boron Endpoint)
  • All Rods Out isothermal Temperature Coefficient measurement

+ Measurement of Reference Bank worth by dilution

  • Control Rod Worth Measurement by Rod Swap Zero power physics testing for McGuire 1 Cycle 12 began at 0114 on May 18,1997 commencing with rod withdrawal for approach to criticality. ZPPT ended at 0155 on May 19,1997 following analysis of Rod Swap data. Table 3 summarizes results from ZPPT. All acceptance criteria were met.

Sections 3.1 through 3.10 describe ZPPT measurements and results.

3.1 - 1/M Approach to Criticality Initial criticairty for McGuire 1 Cycle 12 was achieved per PT/0/A/4150/28, Criticality Following a Change in 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 0114 on May 18,1997. Criticality was achieved at 0244 on May 18,1997 with Control Bank D at 111 steps withdrawn.

- Figure 3 shows the ICRR behavior during the approach to criticality. All acceptance criteria of PT/0/A/4150/28 were met.

'T.

Page 5 of 45 TABLE 3

SUMMARY

OF ZPPT RESULTS l

PREDICTEDVALUE OR PARAMETER MEASURED VALUE ACCEPTANCE CRITERIA Nuclear Heat 4.75 x 10' amps (N41) N/A ZPPT Test Band 10* to 10' amps (N41) N/A ARO Critical Boron 1803 ppmB 1778 50 ppmB ARO ITC 1.24 pcnfF 1.37 2 pcm/ F ARO MTC +0.44 penfF +0.31 penfF Reference Bank (Shutdown Bank 922.0 pcm 881 132 pcm B) Worth Ref. Bank in Critical Boron 1683 ppmB 1662 ppmB Differential Boron Worth 7,75 pcm/ppmB 7.55 pcm/ppmB Control Bank D Worth 007.5 pcm 614 200 pcm Control Bank C Worth 727.8 pcm 730 219 pcm Control Bank B Worth 655.0 pcm 622 200 pcm Control Bank A Worth 385.1 pcm 390 2 200 pcm Shutdown Bank E Worth 524.7 pcm 508

  • 200 pcm Shutdown Bank D Worth 425.7 pcm 424 200 pcm Shutdown Bank C Worth 418.7 pcm 423
  • 200 pcm Shutdown Bank A Worth 266.2 pcm 268 2 200 pcm Total Rod Worth 4932.6 pcm 4860 486 pcm

Page 6 of 45 FIGURE 1 ICRR vs. CONTROL ROD POSITION DURING APPROACH TO CRITICALITY l,--- -- - - -- - - - -

y V '

O.9 - - - -- - - - - - - - -

0.8 - -- -- - - - - - - - . .

0.7 - - ---- - - -

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SDD @ 0 SDE @ 0 CBA@0 CBB @ 0 CBC @ 0 CBD @ 0 Control Bank Position l

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[ Page 7 of 45 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 overlap 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/4150/28, Criticality Following a Change in Core Nuclear Characteristics, on May 8,1997. Table 4 contains the overlap data. The acceptance criterion was met.

'l TABLE 4 SOURCE RANGE / INTERMEDIATE RANGE OVERLAP DATA SOURCE RANGE INTERMEDIATE RANGE N31, cps N32, cps N35, amps N38, amps g .

INITIAL DATA:

NIS Meters 1100 850 2 x 10 '" 2 x 10 '"

OAC 1268 835 2 x 10~" 2 x 10~"

FINAL DATA:

NIS Meters 14,000 8,000 - 2.0 x 10 "* 2.0 x 10 "'

OAC 10586 7692 1.3 x 10 "' 1.4 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 change in reactivity 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/4150/21, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing, on May 18,1997, Table 5 summarizes the data obtained.

The Zero Power Physics Test Band was set at 10* to 10" amps on Power Ran0e channel N41 (connected to reactivity computer). This test band provided more than a factor of two margin to nuclear heat for zero power physics testing. Acceptance criterion was satisfied.

Page 8 of 45

! TABLE 5 NUCLEAR HEAT DETERMINATION Reactivity Computer Intermediate Range Intermediate Range (N41), amps Channel N35, amps Channel N36, amps RUN #1 3.5 x 10' 3 x 10* 2 x 10 3

-- )

RUN #2 '6x10 6 x 10' 6 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 can 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%

The checkout was performed for Cycle 12 on May 18,1997.' 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 186,11 33.37 33.04' O.33 -0.97 223.86 - 28.42 28.24 0.18 -0.61 3.5 - 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 near the all rods out (ARO) configuration. Reactor Coolant System boron samples are obtained while Control Bank D is pulled to the fully withdrawn position. _The reactivity difference from criticality to the ARO configuration is measured and 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 May 18,1997 per PT/0/A/4150/10, Boron Endpoint Measurement. The ARO, HZP boron concentration -

-was measured to be 1803 ppmB. Predicted ARO critical boron concentration was 1778 ppmB. The acceptance criterion, measured boron within 50 ppmB of predicted, was met.

.o Page 9 of 45 3.6 ARO Isothermal Temperature Coefficient Measurement The all rod. out (ARO) Isothermal Temperature Coefficient (lTC) is measured at the beginning of each j

cycle to verify consistency with predicted value. In adoition, the Moderator Temperature Coefficient (MTC) is obtained by subtracting the Doppler Temperature Coefficient from the ITC. The MTC is used to ensure compliance with Technical Specification limits.

To measure the ITC, state point data is obtained prior to cooldown. A Reactor Coolant System cooldown is initiated, within administrative cooldown limits. When sufficient data (at least 4 *F) is obtained, state point data is again obtained. A heatup is performed while again maintaining administrative limits. The Delta Reactivity divided by the Delta Temperature (for each cooldown/heatup) are used to determine the ITC. The cooldown/heatup cycle is repeated if additional data is required.

The Beginning of Cycle 12 ITC was measured per PT/0/A/4150/12, Isothermal Temperature Coefficient Measurement, on May 18,1997. No additional cooldown/heatup cycles were required because of good agreement between initial heatup and cooldown results, Table 7 summarizes the data obtained during the

.. easurement.

Average ITC was determined to be -1.24 pcmFF. Predicted ITC was -1.37 pern?F. Me:.sured ITC was therefore within acceptance criterion of predicted ITC

  • 2 pcmFF.

The MTC was determined to be +0.44 pcmFF. This value was used with procedure PT/0/A/4150/31, 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 withdrawal limits were required,

TABLE 7 ITC MEASUREMENT RESULTS ITC, pcmFF AT, 'F Ap, pcm Tm ,'F Cooldown 4.8 +5.2 554.6 -1.08 Heatup +5.0 -7.0 554.6 -1.40 h

l Average: -1.24  ;

3.7 Reference Bank Worth Measurement by Dilution The control rod bank predicted to have the highest worth is designated the Reference Bank. This RCCA bank is measured by irmerting the bank (with all other mi banks fully withdrawn) in discrete steps while slowly diluting the Reactor Coolant System (at rate < 500 pcm/hr). The reactivity worths of the discrete steps of rod insertion are measured using the Reactivity Computer and summed to obtain the integral worth of the Reference Bank.

The Beginning of Cycle 12 Reference Bank (Shutdown Bank B) worth was measured on May 18,1997 per PT/0/A/4150/11, Control Rod Worth Measurement. Figure 3 shows integral worth of Reference Bank versus bank position. The Reference Bank was measured to be worth 922 pcm; predicted worth was 881 pcm. The acceptance criterion, measured worth within 15% of predicted, was met.

.c Page 10 of 45 FIGURE 2 INTEGRAL AND DIFFERENTIAL WORTH OF REFERENCE BANK 1000,0 7.0

! 900.0 '--

- 6.0 800.0 -

l

-- 5.0 700.0 -_

h 600.0 --

j --4.0 500.0 -

E - 3.0 p 400.0 -

3 300.0 - _ 2.0 200.0 -

, - 1.0 100.0 I-0.0

, O.0 0 50 100 150 ,. 200 250 Bank Position

-at-Integral Rod Worth

-+-Differential Rod Worth

________a

.o Page 11 of 45 3.8 Ditterential Boron Worth Determination The differential boron worth is calculated from the measured ARO critical boron concentration, " Reference Bank in' critical boron concentration, and total measured reactivity worth of Reference Bank. The calculated value is compared to predicted value to verify consistency. This calculation also provides an l

Indirect check of measured Reference Bank worth and of the Boron Endpoint measurements.

The Beginning of Cycle 12 Hot Zero Power differential boron worth was calculated to be 7.75 pcm/ ppm 8 per PT/0/N4150/f1. Control Rod Worth Measurement. The predicted value was .7.55 pcm/ppmB. There is no acceptance criterion on this measurement.

3.9 Control Rod Worth Measurement by Rod kwap The worths of all control rod banks except the Reference Bank are measured by inserting each bank while withdrawing the Reference Bank and/or previously measured bank to mahtain near critical conditions.

When the_ bank being measured is fully inserted, the Reference Bank is positioned to achieve critical conditions with all other rod banks fully withdrawn. The worth of the fully inserted bank is determined from the critical position of the Reference Bank. The measured worth is compared to predicted worth to verify consistency. The sum of the worths of all banks, including the reference bank, is also compared to predicted tothl.

The Beginning of Cycle 12 rod worth measurement by Rod Swap was performed on May 18 19,1997 per PT/0/A/4150/11 A, Control Rod Worth Measurement Rod Swap. Table 8 summarizes the results. All acceptance criteria were met.

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Page 13 of 45 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/4150/21, Post Refueling Controlling Procedure br Criticality, Zero Power Physics, and Power Escalation Testing. Tests are performed from 0% through 100% power with major testing i plateaus at -30%,-75%, and 100% power.

Significant tests performed during McGuire 1 Cycle 12 Power Escalation were:

I e Core Power Distribution (at -30%, -78%, and 100% power) l

  • One-Point incore/Excore Calibration (at -30% power) e neactor Coolant Delta Temperature Measurement (at 90% and 100% power)

. Hot Full Power Critical Boron Concentration Measurement (at 100% power) e 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 Steady State at 30%,78%,90%,98%, and 100% (Steam Generator Replacement Post Mod testing) 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%,30%,78%, and 100%

(Steam Generator Replacement Post Mod testing) e Evaluation of intermediate Range NIS Rod Stop and Rx Trip Setpoints Power Escalation Testing for McGuire 1 Cycle 12 began on May 19,1997. Full power was reached on May 29,1997. Full power testing was completed on June 11,1997. Sections 4.1 through 4.9 describe the l

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 limits. Data obtained during this test are 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 Shangstrom Nuclear Associates' CORE package and FCF's MONITOR code, respectively).

The McGuire 1 Cycle 12 Core Power Distribution measurements were performed on May 21,1997 (30%

power), May 25,1997 (78% power), and June 9,1997 (100% power). Tables 9 through 11 summarize the results. All acceptance criteria were met.

Page 14 of 45 TABLE 9 CORE POWER DISTRIBUTION RESULTS 30% POWER Plant Data Map ID: mic12f001 Date of Map: May 21,1997 Cycle Burnup: 0.21 EFPD Power Level: 29.14% F.P.

Control Rod Position: Control Bank D at 203 Steps Wd Reactor Coolant System Boron Concentration: 1773 ppmB COMET Results -

Core Average Axial Offset: 12.328 %

Tilting Factors for Entire Core Height: Quadrant 1: 0.90450 Quadrant 2: 1.00481 Quadrant 3: 0.99569 Quadrant 4: 1.00499 Maximum F. (nuclear): 1.983 Maximum F (nuclear): 1.459 Maximum Error between Pred, and Meas F;%

4.22 Average Error between Pred, and Meas. F; 1.20 %

Maximum Error between Expected and Measured 4.39%

Detector Response:

RMS of Errors between Expected and Measured 1.62 %

Detector Response:

MONITOR Results Minimum F OperationalMargin: 22.04 %

Minimum F. RPS Margin: 14.80 %

Minimum F. LCO Margin: 57.11 %

Minimum F, Surveillance Margin: 37.40%

Minimum F LCO Margin: 25.15%

Page 15 of 45 TABLE 10 CORD POWER DISTRIBUTION RESULTS l 78% POWER Plant Data Map ID: mic12f002 Date of Map: May 25,1997 Cycle Burnup: 1.78 EFPD Power level: 78.39% F.P.

Control Rod Position: Control Bank D at 200 Steps Wd Reactor Coolant System Boron Concentration: 1488 ppmB COMET Results Core Average AxlalOffset: 0.383 %

Tilting Factors for Entire Core Height: Quadrant 1: 0.99691 Ouadrant 2: 1.00009 Quadrant 3: 1.00235 Quadrant 4: 1.00064 Maximum F. (nuclear): 1.782 Maximum F,(nuclear): 1.417 Maximum Error between Pred, and Meas F,: 3.57 %

Average Error between Pred, and Meas. F,: 1.02 %

Maximum Error between Expected and Measured 3.74 %

Detector Response:

RMS of Errors between Expected and Measured 1.40%

Detector Response:

MONITOR Results Minimum F. Operational Margin: 9.18%

Minimum F RPS Margin: 14.83 %

Minimum F LCO Margin: 39.57 %

Minimum F, Surveillance Margin: 16.65 %

Minimum F, LCO Margin: 20.31 %

Page 16 of 45 TABLE 11 CORE POWER DISTRIBUTION RESLLTS 100% POWER Plant Dats l Map lD: mic12f005-l Date of Map: June 8,1997 Cycle Burnup: 12.46 EFPD Power Level: 99.82% F.P.

Control Rod Position: Control Bank D at 214 Steps Wd Reactor Coolant System Boron Concentration: 1211 ppmB COMET Results Core Average AxialOffset: 0.685 %

Tilting Factors for Entire Core Height: Quadrant 1: 0.99515 Quadrant 2: 1.00413 Quadrant 3: 0.99874 Quadrant 4: 1.00198 Maximum F. (nuclear): 1.737 Maximum F (nuclear): 1.405 Maximum Error between Pred, and Meas F,: 2.82 %

Average Error between Pred. a'nd Meas. F,: 0.78 %

Maximum Error between Expected and Measured 3.07 %

Detector Response:

-3 RMS of Errors between Expected and Measured 1.04 %

Detector Response:

MONITOR Results Minimum F. Operational Margin: 10.04 %

Minimum F. RPS Margin: 17.61 %

Minimum F. LCO Margin: 24.99 %

Minimum F, Surveillance Margin: 6.30 %

Minimum F, LCO Margin: 15.50 %

Page 17 of 45 4.2 One Point incore/Excore Calibration PT/0/A/4600/02FD, One Point incore/Excore Calibration, is performed using results of Power Range NIS I

data taken at 30% power and the incore axial offset measured at 30%. Power Rany channels are calibrated before exceeding 50% in order to have valid Indications of Axial Flux Differena and Quadrant Power Tilt Ratio for subsequent power ascension. The calibration is checked using the intermediate power level flux map (78% F.P. for M1012). Il necessary, Power Range NIS is recalibrated per PT/0/N4600/02F or PT/0/N4600/02G, incore and NIS Rocalibration.

Data for McGuire 1 Cycle 12 was obtained on May 21,1997 and all Power Range NIS calibrations were completed on May 23,1997. 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/N150/40, NC Loop Delta T, RTAS, and OPDT &OTDT Channel Check Criteria Evaluation, to ensure that full power detta temperature constants (AT.) are valid.

AT,is used in the Over power and Over temperature Delta Temperature reactor protection functions.

In the case of M1012, the four loop AT/s were each preliminarily established at 59.97 F, 57.82*F.

58.97*F, and 57.49'F for Loops A D, per Steam Generator Replacement Project and previously observed biases. Portions of PT/0/N4150/40 were completed at 90% power on May 26, 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 ATls were adjusted using full power results. Table 12 summarizes the test results.

TABLE 12 REACTOR COOLANT DELTA TEMPERATURE DATA Reactor Power a 99.67%

Loop A Loop B Loop C Loop D Meas. T,,, 'F 614.5 613.5 614.6 612.9 Meas. Tm, 'F 554.6 555.4 555.5 554.8 Calc. AT,, 'F 59.9 58.1 59.1 58.1 4.4 Hot Full Power Critical Boron Concentration Measurement The Hot Full Power critica! boron correntration is measured using PT/0/N4150/04, 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.

Page 18 of 45 For the purposes of Startup Physics testing, the predicted entical boron concentration is adjusted for the diff.rence between predicted and measured critical boron concentration measured at Zero Power. Tiie difference between measured boron concentration and adjusted predicted value is used to compare to acceptance enterion (150 ppmB).

For McGuire 1 Cycle 12, the Hot Full Power critical boron concentration was measured on June 11,1997, The measured critical boron concentration was 1195.9 ppmB. Predicted critical boron concentration was 1179.2 ppmB; when adjusted for difference at zero power, the adjusted predicted critical boron concentration was 1170.9 ppmB. The difference between measured and adjusted predicted critical boron concentration was 8.3 ppmB, which met the acceptance criterion.

4.5 incore/Excore Calibration 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 N.S 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. Tha 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 1 Cycis 12 on June 7 and 8,199/. A!I Power Range NIS calibrations were completed on June 9. Eight flux maps, with axial offset ranging from -0.568% to +3.989% were used. Table 13 summarizes the results. All acceptance criteria were met.

Page 19 of 45 TABLE 13 INCOREKXCORE CALIBRATION RESULTS Full Power Currents, Microamps Axial N41 N42 N43 N44 Offset, Upper Lower Upper Lower Upper Lower Upper Lower l

+30% 220.3 153.9 236.9 162.3 218.1 158.7 211.9 150.1 l

0% 178.3 195.2 193.4 205.6 179.3 202.9 171.9 193.6 30% 136.2 230.5 150.0 248.9 140.5 247.1 132.0 237.2 Correction (M s) Factors N41 N42 N43 N44 1.340 1.379 1.382 1.312 4.6 BOC12 Unit Load Steady State Test in order to verify satisfactory sts ',dv state plant operation with Replacement Steam Generators (NSM MG-19815 TT/1/A/9815/00/02E, BOL 12 Unit Load Steady State Test for NSM MG 19815 was performed at approximately 30%,78%,89%,98% and 100%. With the plant at steady power level data on the following parameters was obtained, e Reactor Power

. NC Loop Cold Leg Temperature e NC Loop Hot Leg Temperature

. NC Loop Average Temperature e NC Loop Delta Temperature e Pressurizer Level Turbine Control Valve Positon

Page 20 of 45

  • Turbine impulse Pressure e S/G Narrow Range Level

)

e S/G Main Steam Flow e S/G Steam Pressure e S/G Feedwater Flow The test was performed at -30% power on May 20,1997, -78% on May 25,1997, -89% on May 26, 1997, -98% on May 26,1996, and -100% on August 13,1997. All acceptance criteria were rnet. Tables 14 through 24 document the results.

Table 14 NC Loop Cold Leg Temperatures L 1 Power Level NC Loop A NC Loop B NC Loop C - NC Loop D 29.37 555.35 555.66 555.77 555.16 78.24 553.72 554.46 554.45 553.84 88.92 555.47 556.26 556.27 555.64 98.21 555.18 556.03 556.04 555.43 99.79 554.63 555.32 555.50 554.78 Table 15 NC Loop Hot Leg Temperatures Power Level NC Loop A NC Loop B NC Loop C NC Loop D 29.37 573.53 572.99 573.26- 572.80 78.24 600.22 600.12 600.71 599.42 88.92 607.63 607.67 608.29 606.78 98.21 612.39 012.54 613.19 611.46 99.79 613.11 612.78 613.48 611.83

e Page 21 of 45 Table 16 NC Loop Average Temperatures Power Level NC Loop A NC Loop B NC Loop C NC Loop D 29.37 564.74 564.50 564.88 564.40 78.24 ,_

577.67 577.61 578.20 577.34 88.92 582.29 582.29 582.90 582.00 98.21 584.59 584.62 585.24 584.29 99.79 584.60 584.40 585.15 584.13 Table 17 NC Loop Delta Temperatures Powor level - NC loop A NC Loop B NC Loop C NC Loop D

, 29.37 18.19 17.32 17.49 17 64 l 78.24 46.50 45.66 46.27 45.57 88.92 52.16 51.40 52.01 51.13 98.21 57.20 59.49 57.14 56.03 99.79 58.49 57.45 57.96 57.07 Table 18 Pressurizer Level Data Power Level PZR Level PZR Level PZR Level Channel 1 Channel 2 Channel 3 29.37 34.23 34.50 33.26 78.24 48.71 47.60 47.88 88.92 53.15 52.03 52.25 98.21 55.73 54.60 54.94 99.79 54.94 53.79 54.13 Table 19 Turbine Governor Valve Positions Power Level GV1 GV2 GV3 GV4 29.37 9.2 11.5 8.5 8.0 78.24 6.6 73.0 70.1 0 88.92 26.1 97.8 94.8 0 98.21 95.4 97.8 94.8 8.2 99.79- 95.4 97.8 94.7 15.5

i-1 Page 22 of 45 Table 20 i Turbine impulse Pressure 4 (PSIG)

Power level Chi Ch2 '

29.37 186.21 185.51 78.24 557.44 560.89 88.92 641.83 643.78 98.21 713.89 713.80 99.79 728.49 728.12 Table 21 S/G Narrow Range Level i

(%)

Power Level S/G A S/G B S/G C S/G D 29.37 45.42 44.67 47.20 46.68 78.24 57.72 58.39 58.90 58.98 88.92 60.88 61.42 62.15 62.07 98.21 63.74 63.77 64.27 64.23 99.79 64.29 64.24 64.50 64.20 Table 22 S/G Main Steam Pressure (PSIA)

Power Level S/G A S/G B S/G C S/G D 29.37 1061.8 1058.1 1056.9 1065.5 78.24 1019.1 1016.4 1015.4 1022.9 88.92 1027.6 1024.9 1024.1 1031.5 98.21 1018.6 1016.1 1015.4 1023.3 99.79 1007.0 1004.2 1004.0 1011.1 Table 23 S/G Main Steam Flow (MLB/HR)

Power Level S/G A S/G B S/G C S/G D 29.37 1.001 0.877 0.995 0.979 78.24 2.922 2.831 3.022 2.975 88.92 3.378 3.274 3.491 3.434 98.21 3.784 3.670 3.903 3.851 9979 3.899 3.757 3.987 3.944

4 4

Page 23 of 45 Table 24 S/G Foodwater Flow (MLB/HR)

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l Power Level S/G A S/G B S/G C S/G D 29.37 0.995 1.014 0.967 1.029 78.24 2.860 2.912 2.907 2.910 88.92 3.290 3.352 3.348 3.346 98.21 3.674 3.744 3.734 3.730 99.7s 3.738 3.800 3.794 3.781 l

i l

4.7 UNIT LOAD 10% TRANSIENT TEST TT/1/A/9815/00/03E, Unit Load 10% Transient Test for NSM MG 19815, was performed to verify proper operation of the modifications performed on various control systems per NSM MG 19815, 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 Turbine / Generator Control). The test venfies that the control systems work as designed to prevent the following plant transients (in response to a -10% step load change):

  • Actuation of Safety injection Pressurizer and Steam Safeties or PORVs Lifting This test satisfies the transient retest as required for the Post Modification Testing for Replacement Steam Generator instrumentation and Control.

This test was performed from 38% Reactor Power. on May 23,1997 and from 78% Reactor Power on May 26,1997. All acceptance criteria for the test were met as follows:

1; Reactor did not trip --

2) Turbine did not trip

. 3) Safety injection was not initiated

4) No Manual Operator Intervention was required to stabilize the Unit
5) Pressurizer PORV's did not lift
5) Pressurizer Code Safety Valves did not lift
6) Steam Generator PORV's did not lift
7) Steam Generator Code Safety Valves did not lift i

Page 24 of 45

8) Monitored plant parameters did not indicato sustained or diverging oscillations Figures 3 through 21 illustrate response of plant parameters during the 78% Reactor Power test.

e

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Page 44 of 45 4.8 Replacement S/G Tuning and Testing of Feedwater Control TT/1/A/9815//00/04E, Functional Tuning and Tasting of the Feedwater Control System, was performed to record the behavior of the S/G Level Controls and the course of action taken M optimally tune the system following replacement of the Westinghouse D3 Steam Generators. Testing was performed prior to Startup and at various power levels during initial power ascension to allow monitoring of S/G Level Controls and Feed Pump Speed Controls. Difficulties with level swings at very low powet levels (-3% power) prevented optimization of level control due to inaccuracies in main steam flows and changes in the CF header to MS header differential pressure program for Feed Pump Speed Control Program. These effects were unexpected since CNS did not remove the Nozzle Swap startup evolution while MNS did. The following discussion summarizes the results of the testing.

Monitoring System Being Placed in Automatic Control i The Feedwater Control System Bypass Valses were placed in AUTO on May 20,1997 with the A Main Feed Pump in MANUAL at approximately 130 pounris on the D/P Program. Reactor Power was l approximately 3% and level swings of 4 to 8 % narrow range were observed. This instability improved with i

a reduction to approximately 76 pounds D/P on the Feed Pump Speed Program and a Bypass Valve

! Controller Gain reduction from 4.7 v/V to 3.0 v/v., however, operations was uncomfortable with testing at

! these level swings. Testing was canceled at this power level.

Monitoring the Placement of the Main Turbine On-Line Placing of tbs Turbine Generator on line at approximately 8% power was uneventful and occurred on May 20,1997. All S/G levels fluctuated slightly (decreased) and the Feed Pump Speed Program D/P dropped slightly. S/G levels retumed to setpoint evenly with no oscillations occurring. The A Feed Pump retumed smoothly to program.

Monitoring the Transition onto the Main Feed Reg Valves As approximately 15% power, the Main Feed Rag Valves were opened in MANUAL with the associated Bypass Valve closing in AUTO.The Feed Pump Program D/P was 130 pounds. Again there were difficulties with stability. Reduction of Feed Pump Speed D/P Program to 76 pounds resulted in improved behavior and the CF System was placed in full automath operation.

Tuning Tests at 30% Power Steam Generators D through A (in that order) were subjected to five (5) percent level perturbations on May 21,1997. These tests were conducted on one generator at a time. An increasing step change was applied to the narrow range level setpoint followed by a decreasing step change of the same magnitude to retum the setpoint to its programmed position. The S/G levels responded adequately for the increasing step change , overshooting approximately 1 to 2%. The setpoint decrease produced a similar undershoot.

No problems or tuning adjustments were required for this phase of testing.

A Feed Pump Speed Program D/P perturbation was simulated by introducing a five (5) percent Feed Pump Speed demand on the Feed Pump Speed Controller in MANUAL The five percent speed demand resulted in a CF Header to SM Header D/P change of nearly 100 pounds. The Main Feedwater Reg Valves compensated (closed) and S/G levels changed less than 2 percent and settled within one cycle.

The Feed Pump Controller was returned to program by placing the controller back to AUTO. A reverse of

h

,4

}

Page 45 of 45 f the above behavior resulted. No problems encountered or tuning adjustments were required for this phase of testing.

Monitoring the Placement of the B Feed Pump into the Feedwater Header Tne second Feed Pump was introduced into sea,'ce on May 25,1997 at approximately 40% power. No problems were encountered or tuning adjustments required for this transition.

Tuning Tests at 78% Power S/G Level control was well behaved through power ascension to 78% power, therefore, no further level perturbation testing was performed. The Feed Pump Speed Perturbation was performed with both Feed Pumps in MANUAL. CF Header to SM Header D/P was raised from approximately 160 pounds to 185 h

pounds. The A and D S/G Main Feed Reg Valves showed signs of oscillation between 25% and 45% open

{ and the test was terminated.

Observations and Conclusions

1) The elimination of Nozzle Swap at MNS had a significant effect on the Feedwater Control System.

The shift in CF/SM Header D/P resulted in excessive Pump sped 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 tuning adjustments as warranted. The issue 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.

4.9 Reactor Coolant Flow Testing PT/1/A/4150/13, NC Flow Calculation, y performed on May 28,1997 at approximately 97.3% power.

The average reactor coolant system flow (as determined by four one hour test runs, measured by the elbow taps, is 401,932 gallons per minute. The NC flow prior to Steam Generator replacement was approximately 385,600 gpm, or an increase of 16,332 gpm, or approximately 4.25%.

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