ML20133A402

DPC CNS Unit 1 Cycle 10 Startup Rept Dec 1996

ML20133A402
Person / Time
Site:	Catawba
Issue date:	12/18/1996
From:	Mccollum W DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9612310081
Download: ML20133A402 (45)

Text

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1) uke (buer Company H1uturR.Ah O rttM.Je Ca >oNuclear Generationikpartment 17ce1%:sident 43 jncordRoad (803)M31-330 Office York,3C29745 (803)R3/41;t6 Fax DUKEPOWER December 18, 1996 U.

S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, DC 20555

Subject:

Catawba Nuclear Station, Unit 1 Docket No. 50-413 Startup Report, Unit 1 Cycle 10 In accordance with Section 6.9.1 of the Catawba Nuclear Station Technical Specifications, find attached the Unit 1 Startup Report for Cycle 10 core design.

Any questions concerning this report may be directed to Kay Nicholson at (803) 831-3237.

Sincerely, W. R. McCollum, J KEN /U1C10.SR Attachment

@d xc:

S. D. Ebneter j/

Regional Administrator, Region II 300041 P. S. Tam, ONRR

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R. J. Freudenberger, SRI 9612310081 961218 PDR ADOCK 05000413 1

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1 Duke Power Company Catawba Nuclear Station Unit 1 Cycle 10 l

STARTUP REPORT December 1996 1

1 4

1

TABLE OF CONTENTS l

Paae List of Tables

. ii 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 l

2.3 Reactor Coolant System Dilution..

.2 2.4 Control Rod Drag Test..

.3 2.5 Control Rod Drop Timing Test.

.4 3.0 Zero Power Physics Testing.

.9 3.1 1/M Approach to Criticality..

.9 3.2 Source Range / Intermediate Range Overlap Data.

.12 3.3 Point of Nuclear Heat Addition..

.12 3.4 Reactivity Computer Checkout.

.13 3.5 ARO Boron Endpoint Measurement..

.13 3.6 ARO Isothermal Temperature Coefficient Measurement..

.14 3.7 Reference Bank Worth Measurement by Dilution.

.14 3.8 Reference Bank in Boron Endpoint Measurement..

.16 3.9 Differential Boron Worth Determination..

.16 3.10 Control Rod Worth Measurement by Rod Swap.,.

.. 16 4.0 Power Escalation Testing..

.18 4.1 Core Power DistribuSon..

.18 4.2 One-Point incore/Excore Calibration.

.22 4.3 Reactor Coolant Delta Temperature Measurement.

.23 4.4 Hot Full Power Critical Boron Concentration Measurement..

.24 4.5 Incore/Excore Calibration.

.24 4.6 Calorimetric Reactor Coolant Flow Measurement..

.25 4.7 Unit Load Steady State Test..

.26 4.8 Unit Load Transient Test..

.29 4.9 Replacement S/G Tuning and Testing of DFCS..

.38 4.10 intermediate Range NIS Setpoint Evaluation..

.40 l

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LIST OF TABLES L

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1.

Core Design Data..

....1 I

3 2i RCCA Drag Test Results..............

.3 l

3.

Cycle 9 and Cycle 10 Rod Drop Timing Results..

.5 l

3 l

4.

- Preliminary NIS Calibration Data............

.7 6

5.

Summary of Zero Power Physics Testing Results.

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Page 6 of 40 FIGURE 1 North CORE LOADING PATTERN, CATAWBA UNIT 1 CYCLE 10 A

I 180 FUEL TRANSFER CANAL ->

1___________

H22 AA21 AA70 J63 AA34 AA42 H18 PD PD PD PD PD PD PD 2--____

K70 AA35 AC03 AA66 AC15 AA58 AC1A AA12 AC2A AA43 K34 PD RCCA BP RCCA BP RCCA BP RCCA BP RCCA PD 3--__

K69 AA39 AC2C AA72 ACO2 AA50 K58 AA13 AC39 AA49 ACOA AA25 K20 PD PD BP RCCA BP RCCA PD RCCA BP RCCA BP PD PD 4____

AA28 AC23 K53 AC24 K46 AC43 AA32 AC50 K01 AC20 K05 AC21 AA46 RCCA BP RCCA BP PD BP RCCA BP PD BP RCCA BP RCCA 5_

H04 Acid AA76 AC19 AA14 AC53 AA65 AC63 AA74 AC64 AA52 AC01 AA64 AC29 H16 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 6_

AA08 AA75 AC13 K15 AC42 J58 AC91 KD6 AC45 JO9 ACSA K52 AC10 AA07 AA63 PD RCCA BP PD BP RCCA BP RCCA BP RCCA BP PD BP RCCA PD 7_

AA73 AC34 AA40 ACSC AA33 AC55 AA22 AC56 AA48 AC4C AA01 ACS2 AA19 AC32 AA68 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 8_

J46 AA53 K68 AA62 AC41 K40 AC62 J20 AC4A K59 AC49 AA02 K51 AA47 J47 go PD RCCA SS RCCA BP RCCA BP RCCA BP r.MA BP RCCA SS RCCA PD 270 g_

AA51 AC26 AA60 AC6C AA26 AC69 AA05 AC61 AA11 ACh 2 AA24 AC60 AA59 AC40 AA15 PD BP RCCA BP PD BP PO BP PD BP PD BP RCCA BP PD 10 _

AA44 AA37 AC25 K04 AC59 J72 AC66 K25 AC46 J53 AC90 K73 AC22 AA16 AA23 PD RCCA BP PD BP RCCA BP RCCA BP RCCA BP PD BP RCCA PD 11 _

H17 AC31 AA27 AC0C AA04 AC44 AA38 AC51 AA56 AC65 AA09 AC36 AA20 AC09 H05 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 12 _ _ _ _

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i Page 7 of 40 TABLE 4 PRELIMINARY NIS CALIBRATION DATA Intermediate Range Ratio Cycle 9 BOC 10 BOC 10 Channel (BOC 10 +

Reactor Trip Reactor Trip Rod Stop Cycle 9)

Setpoint, Setpoint,

Setpoint, pAmps Amps Amps N35 1.010 78.1 74.9 59.9 N36 1.025 72.0 70.1 56.1 Power Range Ratio Axial Cycle 9 Full Power BOC 10 Full Power Channel (BOC 10 +

Offset, %

Cycle 9)

Current, Amps Current, pAmps Upper Lower Upper Lower

+20 303.7 238.4 323.1 253.7 N41 1.064 0

261.9 276.7 278.7 294.4

-20 220.1 315.0 234.2 335.2

+20 291.1 214.0 310.0 227.9 N42 1.065 0

250.7 248.8 267.0 265.0

-20 210.3 283.4 224.0 301.8

+20 254.8 196.6 269.6 208.0 N43 1.058 0

221.3 228.7 234.1 242.0

-20 187.7 261.0 198.6 276.1

+20 247.4 197.9 263.7 211.0 N44 1.066 0

214.4 231.1 228.6 246.4

-20 181.6 264.4 193.6 281.9

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Page 9 of 40 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/4150/01, Controlling Procedure for Startup Physics 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 j

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 All Rods Out Isothermal Temperature Coefficient measurement

• Measurement of Reference Bank worth by dilution

• Reference Bank in Critical Boron Concentration measurement

• Differential Boron Worth determination l

Control Rod Worth Measurement by Rod Swap i

)

l Zero power physics testing for Catawba 1 Cycle 10 began at 0232 on October 2,1996 commencing with

^

l rod withdrawal for approach to criticality. ZPPT ended at 1422 on October 3,1996 following analysis of l

Rod Swap data. Table 5 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 criticality for Catawba 1 Cycle 10 was achieved per PT/0/A/4150/19,1/M Approach to Criticality. In this procedure, Estimated Critical Rod Position (ECP) is calculated based on latest available Reactor Coolant boron concentration. Control rods are withdrawn until Boron Dilution Mitigation System (BDMS) or Source Range count rate doubles. Inverse Count Rate Ratio (ICRR) is plotted for each source range and BDMS 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 0232 on October 2,1996. Criticality was achieved at 0707 on October 2,1996 with Control Bank D at 207 steps withdrawn.

f Figure 3 shows the ICRR behavior during the approach to criticality. All acceptance criteria of PT/0/A/4150/19 were met.

d 4

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I Page 10 of 40 I

TABLE 5 1

SUMMARY

OF ZPPT RESULTS I

l i

l PREDICTED VALUE OR j

PARAMETER MEASURED VALUE ACCEPTANCE CRITERIA Nuclear Heat 7.7 x 10'7 amps (N43)

N/A ZPPT Test Band 10* to 2.0x10-7 amps (N43)

N/A ARO Critical Boron 1840 ppmB 1833 i 50 ppmB ARO ITC 3.26 pcm/ F

-3.05 i 2 pcm/ F 1

i l

ARO MTC

-1.59 pcm/ F

-1.58 pcm/ F Reference Bank (Shutdown 893.5 pcm 857 i 128.6 pcm I

l Bank B) Worth Ref. Bank in Critical Boron 1725 ppmB 1712 ppmB Differential Boron Worth

-7.77 pcm/ppmB

-7.09 i 1.06 pcm/ppmB Control Bank D Worth 599.0 pcm 562 i 200 pcm Control Bank C Worth 778.8 pcm 807 i 242 pcm Control Bank B Worth 692.2 pcm 634 i 200 pcm 1

Control Bank A Worth 310.0 pcm 346 i 200 pcm Shutdown Bank E Worth 455.2 pcm 466

• 200 pcm Shutdown Bank D Worth -

496.7 pcm 443

• 200 pcm Shutdown Bank C Worth 486.2 pcm 438 i 200 pcm Shutdown Bank A Worth 346.0 pcm 296 i 200 pcm Total Rod Worth 5057.6 pcm 4849 i 485 pcm 4

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I Page 12 of 40 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 existance 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 10 was obtained oer PT/0/A/4150/01, Controlling procedure for Startup Physics Testing, on October 2,1996. Table 6 s *tains the overlap data. The acceptance criterion was met.

TABLE 6 SOURCE RANGE / INTERMEDIATE RANGE OVERLAP DATA

)

I SOURCE RANGE INTERMEDIATE PANGE N31, cps N32, cps N35, amps N36, amps INITIAL DATA:

NIS Meters 500 400 1 x 10 '"

1 x 10 '"

OAC 570 390 1.10 x 10~"

1.00 x 10'"

FINAL DATA:

NIS Meters 15,000 15,000 1.50 x 10 "

2.00 x 10 "

OAC 23,950 12,200 1.442 x 10 "

1.788 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 stariup rate is initiated by rod withdrawal. An increase in Reactor Coolant System temperature and/or j

Pressurizer level accompeied by a change in reactivity and/or rate of flux increase indicates the addition of Nuclear Heat, j

For Cycle 10, the Point of Nuclear Heat Addition was determined per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. on October 2,1996. Table 7 summarizes tha data obtained.

The Zero Power Physics Test Band was set at 10* to 2.0x10-7 amps on Power Range channel N43 (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 13 of 40 TABLE 7 NUCLEAR HEAT DETERMINATION Reactivity Computer Intermediate Range intarmediate Range (N43), amps Channel N35, amps Channel N36, amps RUN #1 7.70 x 10

4.50 x 10~7 5.70 x 10~7 RUN #2 8.10 x 10

5.30 x 10'7 6.58 x 10 7

3.4 Reactivity Computer Checkout The reactivity computer checkout was performed per PT/0/A/4150/01, Controlling Procedure for Startup Physics TeMing, to verify that the Power Range channel connected to the reactivity computer can provide reliable reactivity data. Reactivity Insertions of approximately +25 and -25pcm (but no greater than +45 and -45 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%

j The checkout was performed for Cycle 10 on October 2,1996. Table 8 hsts the results cf the reactivity _

insertion. The acceptance criterion was met.

TABLE 8 REACTIVITY COMPUTER CHECKOUT Period, seconds Theoretical Reac-Measured Reac.

Absolute Error, Percent Error,%

tivity, pcm tivity, pcm pcm 220.99 29.12 28.87 0.25

-0.84

-259.56

-33.43

-32.74 0.69

-2.07 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 i

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 10 beginning of cycle, hot zero power, all rods out, critical boron concentration was measured on October 2,1996 per PT/0/A/4150/10, Boron Endpoint Measurement. The ARO, HZP boron concentration was measured to be 1840 ppmB. Predicted ARO critical boron concentration was 1833 ppmB. The acceptance criterion, measured boron within 50 ppmB of predicted, was met.

Page 14 of 'O 3.6 A.*tO Isothermal Temperature Coefficient Measurement i

The all rods at (/sRO) 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 Temperature Coefficient from the ITC. The MTC is used to ensure compliance with Technical Specification limits.

To measure the ITC, statepoint data is obtained prior to cooldown. A Reactor Coolant System we! deem

!s initiated, within administrative cooldown limits. When sufficient data (at least 5 F) is obtained, statepoid data is again obtained. A heatup is performed while again maintaining administrative limits.

The Delta Reactivity divided by the Delta Temperaure (for each cooldown/heatup) are used to determine the ITC. The cooldown/heatup cycle is repeated if additional data is required.

4 The Beginning of Cycle 10 ITC was measured per PT/0/A/4150/12A, Isothermal Coefficient of Reactivity Measurement, on October 2,1996. No additional cooldown/heatup cycles were required because of good ag,eement between initial heatup and cooldown results. Table 9 summarizes the data obtained during the n easurement.

Average ITC was determined to be -3.26 pcm/ F. Predicted ITC was -3.05 pcm/ F. Measured ITC was th(refore within acceptance criterion of predicted ITC i 2 pcm/ F.

Tne MTC was determined to be -1.59 pcm/ F. This value was used with procedure PT/0/A/4150/21, Temporary Rod Withdrawal Limits Determination, to ensure that MTC would remain within Technical Specification limits at all power levels. No rod withdrawal limits were required.

TABLE 9 ITC MEASUREMENT RESULTS h

l l

ITC, pcm/oF f

AT, 'F. ',,ggcr,p,,f,,b,,*F, Cooldown

_,;, ;,gg,,,,,,~313;,0 554.58 5

-3.19 j

Heatup

+6.00

-20.0 554 60

-3.33 mm mm,mm mJm.. m.m.

mm m. m

,m mmm..

.m.,

l Average: -3.26 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 inserting the bank (with all other rod 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 10 Reference Bank (Shutdown Bank B) worth was measured on October 3,1996 per PT/0/A/4150/11 A, Control Rod Worth Measurement by Boration/ Dilution. Figure 4 shows integral worth of Reference Bank versus bank positiro. The Reference Bank was measured to be worth 893.5 pcm; predicted worth was 857 pcm. The acceptance criterion, measured worth within i 15% of predicted, was met.

Page 15 of 40 FIGURE 4 INTEGRAL AND DIFFERENTIAL WORTH OF REFERENCE BANK 900.0.-

10.0 9.0 800.0 q

700.0 7.0 o E 600.0 g

o.

a i

I 6.0 3 E 500.0 E

5:

=

l 5.0 g E 400.0 y

_g 4.0 g 5 300.0 3.0 3 200.0

- 2.0 100.0 1.0 I

0.0 0.0 0

50 100 150 200 250 Reference Bank Position, steps withdrawn l-e-integral Worth -*-Predicted Worth -+-Differential Worth -*-Predicted Diff. Worth l

. ~ - -

l.

e Page 16 of 40 3.8 Reference Bank in Beton Endpoint Measurement This test is performed at the be: ginning of each cycle to measure the critical boron concentration with the Reference Bank fully inserted and all other control rod banks fully withdrawn. The measured boron i

concentration is used with the measured ARO critical boron concentration and the measured worth of the reference bank to calculate the differential boron worth. Reactor Coolant System boron samples are obtained while control rods are inserted or withdrawn to the " Reference Bank In" configuration. The

[

~

reactivity difference from criticality to the " Reference Bank In" 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 " Reference Bank In" critical boron concentration.

The Cycle 10 Beginning of Cycle, Hot Zero Power, Reference Bank In, critical boron concentration was measured on October 3,1996 per PT/0/A/4150/10, Boron Endpoint Measurement.

This boron concentration was measured to be 1725 ppmB.

Predicted " Reference Bank in" critical boron i

concentration was 1712 ppmB. There is no quantitstive acceptance creerion associated with this test.

3.9 Differential Boron Worth Determination The differential boron worth is calculated from the measured ARO critical boron concentration,

" Reference Bank in" critica! t,0mn 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 indirect check of measured Reference Bank worth and of the Boron Endpoint rneasurements.

The Beginning of Cycle 10, Hot Zero Power differential boron worth was calculated to be -7.77 ocm/ppmB per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. The predicted value was -7.09 pcm/ppmB. The acceptance criterion (measured within i 15% of predicted), was met.

3.10 Control Rod Worth Measurement by Rod Swap i

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 maintain 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 total.

The Beginning of Cycle 10 rod worth measurement by Rod Swap was performed on October 3,1996 per PT/0/A/4150/11B, Control Rod Worth Measurement by Rod Swap. Table 10 summarizes the results. All acceptance criteria were met.

I n

---e w

f 1

i i

Page 17 of 40 1

TABLE 10 CONTROL ROD WORTH MEASUREMENT DATA i

Ad}usted Cntcal Remaseng Difference

% Diff. (Pred -

Reference Posde of Ref.

Worth of Ref.

Measured Predcted (Predcted -

Meas)/Pred =

Bank Bank Worth Bank Bank Alpha Worth, pcm Worth. pcm Measured) 100 Shutdown B (Ref Bank)

N/A N/A N/A N/A 893.5 857

-36.5

-4.3 SMdown A 912.5 100 558 1.015 346.0 296

-50.0

-16.9 i

Control A 912.8 101 553 1.091 310.0 346

+36.0

+ 10.4 s

sixddownc 913.1 121 438 0.974 486.2 438

-48.2

-11.0 Swdown D 913.4 123 427 0.976 496.7 443

-53.7

-12.1 Shutdown E 913.6 105 530 0.865 455.2 466

+10.8

+2.3 conts D 913.9 149 286 1.101 599.0 562

-37.0

-6.6 i

contd B 914.2 148 291 0.762 692.2 634

-58.2

-9.2 controi c 914.5 177 148 0.915 778.0 807

+28.2

+3.5 i

mal 5057.6 4849

-208.6

-4.3 i

}

Page 18 of 40 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/4150/01, Controlling Procedure for Startup Physics Testing. Tests are performed from 0% through 100% power with major testing plateaus at ~30%,~75%, and 100% power.

Significant tests performed during C1C10 Power Escalation were:

. Core Power Distribution (at ~30%, ~71%, and 100% power)

. One-Point incore/Excore Calibration (at 30% power)

. Reactor Coolant Delta Temperature Measurement (at 71% and 100% power)

. Hot Full Power Critical Boron Concentration Measurement (at 100% power)

. Incore/Excore Calibration (at 100% power)

. Calorimetric Reactor Coolant Flow Measurement (at 100% power, This test is not under the control of PT/0/A/4150/01)

. Unit Load Steady State - at 30%,50%,75%, and 100% (Steam Generator Replacement Post-Mod testing) e Unit Load Transient Test - at 50% and 71% (Steam Generator Replacement Post-Mod testing)

. Replacement S/G Functional Tuning and Testing of DFCS - at 10%,30%,50%, and 71%

(Steam Generator Replacement Post-Mod testing)

. Evaluation of Intermediate Range NIS Rod Stop and Rx Trip Setpoints in addition to the tests listed above, PT/0/A/4150/01 performs evaluations of the Movable !ncore Detector System, and on-line (OAC) Thermal Power program. The results of these are not included in this report.

Power Escalation Testing for Catawba 1 Cycle 10 began on October 4,1996. Full power was reached on October 10,1996. Full power testing was completed on October 18,1996. Sections 4,1 through 4.10 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 limits Data obtained during this test i

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 CORE and MONITOR codes (adapted from Shangstrom Nuclear i

Associates

• CORE package and FCF's MONITOR code, respectively).

The Catawba 1 Cycle 10 Core Power Distribution measurements were performed on October 5,1996 (30% power), October 8,1996 (71% power), and October 14,1996 (100% power). Tables 11 through 13 summarize the results. All acceptance criteria were met.

Page 19 of 40 TABLE 11 CORE POWER DISTRIBUTION RESULTS 30% POWER Plant Data Map ID:

FCM/1/10/001 Date of Map:

October 5,1996 Cycle Burnup:

0.30 EFPD Power Level:

29.95% F.P.

Control Rod Position:

Control Bank D at 215 Steps Wd Reactor Coolant System Boron Concentration:

1667 ppmB CORE Results Core Average Axial Offset:

12.421 %

Tilt Ratios for Entire Core Height: Quadrant 1:

0.97997 Quadrant 2:

1.03602 Quadrant 3:

1.01995 Quadrant 4:

0.96407 Maximum Fo (nuclear):

2.083 Maximum F s (nuclear):

1.555 3

Maximum Error between Pred and Meas F s:

9.0%

3 Average Error between Pred, and Meas. Fas:

3.62%

Maximum Error between Expected and Measured 9.50 %

Detector Response:

RMS of Errors between Expected and Measured 4.5%

Detector Response:

MONITOR Results Minimum Fa Operational Margin:

29.38 %

Minimum Fo RPS Margin:

9.79 %

Minimum Fo LCO Margin:

54.95 %

Minimum F s Surveillance Margin:

35.43 %

3 Minimum F s LCO Margin:

19.25%

3

t Page 20 of 40 TABLE 12 CORE POWER DISTRIBUTION RESULTS 71% POWER Plant Data Map ID:

FCM/1/10/003 Date of Map:

October 8,1996 Cycle Burnup:

1.77 EFPD Power Level:

70.897% F.P.

Control Rod Position:

Control Bank D at 215 Steps Wd Reactor Coolant System Boron Concentration:

1425 ppmB CORE Results Core Average Axial Offset:

3.319 %

Tilt Ratios for Entire Core Height: Quadrant 1:

0.99037 Quadrant 2:

1.01704 Quadrant 3:

1.01166 Quadrant 4:

0.98092 Maximum Fo (nuclear):

1.822 Maximum F s (nuclear):

1.466 3

Maximum Error between Pred. and Meas Fas:

6.77 %

Average Error between Pred. and Meas. F s:

2.31 %

3 Maximum Error between Expected and Measured 6.80%

Detector Response:

RMS of Errors between Expected and Measured 3.10%

Detector Response:

MONITOR Results Minimum Fo Operational Margin:

23.16%

Minimum Fo RPS MarDin:

15.58 %

Minimum Fo LCO Margin:

43.79 %

Minimum Fw Surveillance Margin:

21.95 %

Minimum F s LCO Margin:

18.27%

3

. - - ~

s 8

4 Page 21 of 40 TABLE 13 CORE POWER DISTRIBUTION RESULTS 100% POWER Plant Data Map ID:

FCM/1/10/004 Date of Map:

October 14,1996 Cycle Burnup:

6.70 EFPD 1

Power Level:

99.834% F.P.

Control Rod Position:

Control Bank D at 215 Steps Wd Reactor Coolant System Boron Concentration:

1234 ppmB CORE Results Core Average Axial Offset:

-1.596 %

Tilt Ratios for Entire Core Height: Quadrant 1:

0.99172 Quadrant 2:

1.01714 4

Quadrant 3:

1.00269

]

Quadrant 4:

0.98845 Maximum Fo (naclear):

1.784 Maximum Fas (nuvear):

1.449 Maximum Error between Pred. and Meas F s:

7.06 %

3 Average Error between Pred. and Meas. F s:

1.58 %

3

[

Maximum Error between Expected and Measured 7.0%

Detector Response:

l RMS of Errors between Expected and Measured 2.2%

Detector Response:

MONITOR Results j

Minimum Fo Operational Margin:

6.74 %

Minimum Fo RPS Margin:

13.84 %

r Minimum Fo LCO Margin:

23.01 %

Minimum F s Surveillance Margin:

1.51 %

3 Minimum F s LCO Margin:

12.64 %

3 1

1 J

s Page 22 of 40 4.2 One-Point incore/Excore Calibration PT/0/A/4600/05D, 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 order 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 (71% F.P. for C1C10). If necessary, Power Range NIS is recalibrated per i

PT/0/A/4600/05D or PT/0/A/4600/05A, Incore/Excore Calibration.

Data for Catawba 1 Cycle 10 was obtained on October 5,1996 and all Power Range NIS calibrations were completed on October 6,1996. Results are presented in Table 14. All acceptance criteria were

{

met.

J i

i i

TABLE 14 ONE-POINT INCORE/EXCORE CAllBRATION RESULTS Reactor Power = 30.00%

Axial Offset = 12.421%

)

Measured Power Range Currents, Amps l

N41 N42 N43 N44 Upper 75.0 77.0 72.0 62.0 Lower 69.0 67.0 64.0 58.0 Ratio, Extrapolated (from measured) Currents to " Expected"(from last calibration) Currents N41 N42 N43 N44 Upper 0.9604 1.0363 1.1236 1.0243 Lower 0.9991 1.0903 1.1468 1.0534 New Calibration Currents, Amps l

Axial N41 N42 N43 N44

Offset, Upper Lower Upper Lower Upper Lower Upper Lower

+20 263.7 216.8 271.0 210.3 252.6 200.9 217.7 181.7 0

227.4 251.6 233.4 244.4 219.3 233.7 188.7 212.2

-20 191.1 286.4 195.8 278.5 186.1 266.6 159.8 242.7

l s

Page 23 of 40 4.3 Reactor Coolant Loop Delta Temperature Measurement Reactor Coolant System (NC) Hot Leg and Cold Leg temperature data is normally obtained between j

50% and 80% power and at 100% power per PT/0/A/4600/26, NC Temperature Calibration, to ensure that full power delta temperature constants (ATo) are valid. ATo is used in the Overpower and i

Overtemperature Delta Temperature reactor protection functions.

I In the case of C1C10, the four loop ATo's were each preliminarily established at 57.9 F per Steam Generator Replacement Project. Due to the fact that NC Loop 1B has always been significantly cooler than the other loops, PT/0/A/4600/26 was required to be performed (and Loop B ATo adjusted accordingly) at 30.0% F.P. on October 6,1996, because of the resulting discrepency this created between Loop 1B's and the other Loops' %F.P.AT's. PT/0/A/4600/26 was subsequently repeated at 71%

F.P. on October 9,1996 and at 100% F.P. on October 11,1996. All foJr NC Loop ATo's were adjusted per results obtained at 71% power. Loops A, B, and C were ultimately recalibrated using full power results. Table 15 summarizes the test results.

TABLE 15 REACTOR COOLANT DELTA TEMPERATURE DATA Reactor Power = 71.0685%

Loop A Loop B Loop C Loop D Meas. Tmt, "F 597.7 593.3 597.0 596.2 Meas. Teow, *F 552.1 551.9 551.2 552.5 Calc. Ah, BTU /lb 60.225 54.265 60.298 57.671 Calc. Aho, BTU /lb 84.742 76.356 84.845 81.148 Calc. ATo, *F 61.6 56.1 61.7 59.3 Current ATo, *F 57.90 51.44 57.90 54.94 1

Difference, *F 3.7 4.7 3.8 4.3 Reactor Power = 99.5611%

Loop A Loop B Loop C Loop D Meas. Tmi, *F 615.6 610.5 615.0 613.7 Meas. Tcao, F 553.7 553.5 553.0 554.3 Calc. Ah, BTU /lb 84.414 76.952 84.408 80.672 Calc. Aho, BTU /lb 84.786 77.291 84.780 81.028 Calc. ATo, *F 62.1 57.2 62.2 59.6 Current ATo, *F 61.6 56.1 61.7 59.3 Difference, 'F 0.5 1.1 0.5 0.3

I s

Page 24 of 40 4.4 Hot Full Power Critical Boron Concentration Measurement The Hot Full Power critical boron concentration is measured using PT/0'T/4150/04, Reactivity Anomaly Calculation. Reactor Coolant boron concentration is measured (average of three samples) with reactor 3

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 the purposes of Startup Physics testing, the predicted critical boron concentration is adjusted for the difference between predicted and measured critical boron concentration measured at Zero Power. The difference between measured boron concentration and adjusted predicted value is used to compare to acceptance criterion (tSO ppmB).

For Catawba 1 Cycle 10, the Hot Full Power critical boron concentration was measured on October 14, 1996. The measured critical boron concentration was 1221 ppmB. Predicted critical boron concentration was 1250 ppmB; when adjusted for difference at zero power, the adjusted predicted critical boron j

concentration was 1257 ppmB. The difference between measured and adjusted predicted critical boron concentration was -36 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/4600/05A, incore/Excore Calibration. 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 perforrned for Catawba 1 Cycle 10 on October 14 and 15,1996. All Power Range NIS calibrations were completed on October 17. Eight flux maps, with axial offset ranging from -8.443% to l

+4,054% were used. Table 16 summarizes the results. All acceptance criteria were met.

._.. _..=_

j I

Page 25 of 40 TABLE 16 INCORE/EXCORE CAllBRATION RESULTS l

Full Power Currents, Microamps Axial N41 N42 N43 N44

Offset, YO Upper Lower Upper Lower Upper Lower Upper Lower

+20%

294.6 230.6 287.8 215.4 275.1 211.0 228.5 184.8 0%

252.0 269.4 247.5 250.9 236.9 247.3 196.8 216.2

-20%

209.5 308.3 207.3 286.3 198.7 283.7 165.1 247.6 Correction (Ma) Factors N41 N42 N43 N44 1.278 1.316 1.297 1.305 4.6 Calorimetric Reactor Coolant Flow Measurement With clean venturis. PT/2/A/4150/138, Calorimetric Reactor Coolant Flow Measurement, is performed to establish a Primary Loop Delta T correction value to correct Primary power to Secondary Power. This is needed due to the fact that the NC Loop Elbow Tap Correction Factors are now fixed constants per Tech Specifications.

This test was performed on October 21,1996. The Primary Loop D/T Correction was calculated to be 0.972. Table 17 summarizes the results. All acceptance criteria were met.

i j

- a Page 26 of 40 TABLE 17 l

CALORIMETRIC REACTOR COOLANT FLOW MEASUREMENT I

Run Number Secondary Thermal Output %

Primary Thermal Output %

1 99.775 100.288 2

99.670 100.252

-3 99.633 100.245 Average 99.693 100.262 Primary Loop D/T Correction 0.972 (Secondary Power / Primary) 4.7 Unit Load Steady State Test in order to verify satisfactory steady state plant operation with Replacement Steam Generators (NSM CN-19815, TT/1/A/9200/085, Unit Load Steady State Test for NSM CN 19815 was performed at.

approximately 15%,30%,50%,75%, and 100%. With the plant at steady power level data on the i

following parameters was obtained.

Reactor Power NC Loop Cold Leg Temperature NC Loop Hot Leg Temperature j

NC Loop Average Temperature NC Loop Delta Temperature Pressurizer Level Turbine Control Valve Positon f

Turbine impulse Pressure NIS Power Range Indication NIS Intermediate Range Indication Main Steam Pressure The test was performed at -15% power on October 4,1996, ~30% on October 5,1996, ~50% on October 6,1996, -75% on October 8,1996, and ~100% on October 11,1996. All acceptance criteria were met. Tables 18 through 25 document the results.

1 l

Table 18 NC Loop Cold Leg Temperatures i

Power Level NC Loop A NC Loop B NC Loop C NC Loop D l

l 15.58 556.9 556.8 556.3 557.1 I

l 30.66 554.0 554.0 553.3 554.4 l

l 50.36 553.5 553.4 552.6 553.8 i

j l

71.03 552.0 551.9 551.1 552.5 l

99.53 553.7 553.5 552.9 554.3 i

1 l

a

6 i

Page 27 of 40 i

Table 19 NC Loop Hot Leg Temperatures l

Power Level NC Loop A NC Loop B NC Loop C NC Loop D 15.58 567.2 564.7 566.9 566.3 30.66 573.9 571.8 573.6 572.8 50.36 585.9 582.9 585.6 584.7 71.03 596.3 592.6 596.1 595.6 99.53 613.8 609.6 613.9 612.9 l

Table 20 NC Loop Average Temperatures l

l Power level NC Loop A NC Loop B NC Loop C NC Loop D 15.58 562.2 561.0 561.5 561.8 30.66 564.2 562.9 563.3 563.6 50.36 570.2 568.3 569.2 569.5 71.03 574.8 572.5 573.7 574.3 l

99.53 584.7 581.9 583.7 583.9 Table 21 NC Loop Delta Temperatures l

i Power Level NC Loop A NC Loop B NC Loop C NC Loop D 15.58 10.3 7.8 10.5 9.1 30.66 19.7 17.7 20.1 18.3 l

50.36 32.5 29.5 33.0 30.8 l

71.03 44.3 40.7 45.0 43.1 l

99.53 60.1 56.1 61.0 58.7 j

Table 22 Pressurizer Level Data Power level PZR Level PZR Level PZR Level PZR Level Setpoint Channel 1 Channel 2 Channel 3 15.58 30.5 30.5 N/A N/A 30.66 32.7 32.8 33.3 33.4 50.36 39.0 39.0 39.6 39.9 71.03 44.0 44.0 44.9 44.9 99.53 54.4 54.4 55.1 55.4 4

i..

Page 28 of 40 Table 23 Turbine Control Valve Positions Power Level CV1 CV2 CV3 CV4 15.58 11.3 %

11.5 %

0.0%

0.0%

30.66 21.0 %

20.0 %

0.0%

0.0%

50.36 34.7 %

34.4 %

0.0%

0.0%

71.03 91.1 %

91.0 %

7.3%

0.0%

99.53 100.0 %

100.0 %

88.1 %

25.9%

Table 24 Turbine impulse Pressure (PSIG)

Power level Ch1 Ch2 15.72 70.4 71.2 30.67 167.5 168.0 48.49 297.2 297.8 71.01 440.0 440.4 99.83 692.1 691.4 Table 25 NIS Power Range Data

( Amps)

Power Level N-41 N-42 N-43 N-44 15.72 74 77 72 63 30.67 146 146 138 120 48.49 240 238 227 197 71.01 350 340 327 284 99.83 518 495 481 411 Table 26 NIS Intermediate Range Data

( Amps)

Power Level N-35 N-36 15.72 43.1 45.6 30.67 78.9 82.7 48.49 138.0 140.0 71.01 196.0 194.0 99.83 286.0 283.0

s Page 29 of 40 Table 27 Main Steam Pressure (PSIG)

_ Power Level S/G A S/G B S/G C S/G D 15.58 1055.5 1051.8 1065.8 1063.1 30.66 1022.0 1017.9 1032.0 1029.7 50.36 1004.2 999.9 1014.0 1011.7 71.03 980.5 976.1 990.2 988.2 99.53 975.7 971.1 985.6 983.9 4.8 Unit Load Transient Test TT/1/A/9200/86, Unit Load Transient Test for NSM CN-19815, was performed to verify proper operation of the modifications performed on various control systems per NSM CN-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 verifies that the control systems work as designed to prevent the following plant transients (in response to a ~10% step load change):

Reactor Trip Turbine Trip 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 50% F.P. on October 7,1996 and from 70% F.P. on October 8,1996. All acceptance criteria for the test were met as follows:

1)

Reactor and Turbine did not trip 2)

Safety injection was not initiated 3)

No Manual Operator Intervention was required to stabilize the Unit 4)

Pressurizer PORV's did not lift 5)

Pressurizer Code Safety Valves did not lift 6)

Monitored plant parameters did not indicate sustained or diverging oscillations 7)

Nuclear Power undershoot/ overshoot was < 3%

8) initial step load chan0e was 2 80 MWe Figures 5 through 20 illustrate response of plant parameters during the tests.

= _ -.

J s

Page 30 of 40 FIGURE 5 UNIT LOAD TRANSIENT TEST POWER RANGE LEVEL 50% FULL POWER l

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s Page 31 of 40 FIGURE 6 UNIT LOAD TRANSIENT TEST NC LOOP HIGHEST T-AVG / T-REF 50% FULL POWER i

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Page 33 of 40 FIGURE 8 UNIT LOAD TRANSIENT TEST S/G STEAM PRESSURE j

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Page 38 of 40 4.9 Replacement S/G Tuning and Testing of DFCS TT/1/A/9200/96, Replacement Steam Generator Functional Tuning and Testing of the Feedwater Control System, was performed to record the behavior of S/G Level Controls and the course of action taken to optimally tune the system following replacement of the Westinghouse D3 Steam Generators. Testing was performed ptior to Startup (as DFCS was initially placed in AUTO) and at various power level during i

initial power ascenen to allow monitoring of control algorithms of S/G Level Controls and the Feedwater Pump Speed Control System. The Main and Bypass Feedwater Control Valves along with the Feed Pump D/P Program received the dominant scrutiny during this testing. The following discussion summarizes the results of this testing:

Monitoring System Being Placed in Automatic Control The DFCS was placed in automatic control on September 27,1996. While the Unit was in mode 3, 'B' Main Feedpump was placed in service. Due to the plant conditions, the pump speed had to be brought up to minimum speed setting before the pump could be placed in auto. When pump discharge pressure was approximately 1070 PSID and FW/STM Hdr D/P at approximately 160 PSID, the pump was placed in automatic control. The Pump went immediately to minimum speed (approx. 2950 RPM) and remained constant. The CF Bypass valves momentarily opened from 8% to 20% to adjust to new D/P and 4

stabilized immediately with no oscillations observed.

Tuning Tests at 10% Power i

Steam Generators A through D were subjected to five (5) percent level perturbations on October 3,1996.

These tests were conducted on one generator at a time. A decreasing step change was applied to the na. Tow range level setpoint followed by an increasing step change of the same magnitude to retum the setpoint to its programmed position. The S/G levels responded adequately for the decreasing step i

change and a slight overshoot of 2% was noticed for the increasing step change. The overshoot was 1

expected due to the magnitude of the change which required the Main Reg. valves to momentarily open to approximately 12%. This test was to validate the response of the system while utilizing the ' Low-Power' controllers. No problems encountered or tuning adjustments were required for this phase of testing.

Monitoring the Placemot of Main Turbine On-Line Placement of the Main Turbine On-Line occuned on October 4,1996. All S/G levels, valve demands, and Feedwater Pump parameters being monitored for this event. This event caused the S/G levels to decrease approximately 2-3%.

The S/G levels leveled off to setpoint evenly with no oscillations occurring. The pump responded accordingly with a slight decrease in the Fw/Stm Hdr D/P and returned to program level with no oscillations. No problems encountered or tuning adjustments were required for this phase of testing.

Tuning Tests at 30% Power The identical level pertuibation tests performed at 10% power were repeated for the 30% plateau testing on October 5,1996. A sight amount of undershoot and overshoot was observed for these tests which settled out within one cycle. This test was performed to validate the response of the system while utilizing the 'High-Power' controllers. No problems encountered or1

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Page 39 of 40 Tuning Tests at 50% Power Steam Generator A was subjected to five (5) percent level perturbations and the Feedpump was subjected to a 20 PSID increasing / decreasing step change to the Fw/Sim D/P program on October 6, 1996. The level tests were conducted on S/G 'A' and then followed by the Feedpump test.

l Prior to the test, it was noticed that the Steam Flows for all generators indicated approximately 20%

higher than expected. This error effects the level controls by inducing a steam flow / feed flow mismatch to the S/G flow controllers and offsetting the Fw/Stm D/P program of approximately 20%. The effects of these errors and their implications were discussed to insure valid testing results could be achieved for this plateau. It was decided that the errors would have minimal impact to the test results and testing would continue.

All tests were performed satisfactory with no problems encountered or tuning adjustments required for this power level.

10% Load Rejection Test at 50% Power The unit was run back approximately 10% on October 7,1996. All systems performed as expected, it was noticed that 'D' Steam Generator lagged the other three slightly but after the initial swell in the S/G's, all levels returned to program setpoint with no oscillations being noted. No adjustments were required for this test.

10% Load Rejection Test at 70% Power Prior to the start of testing at this power level, it was determined by the test coordinator that final tuning constants are adequate and will not have to be adjusted. Further level perturbation testing will not be required. The only plant differences for this test was an increase of 20% reactor power and both Main Feed pumps were now in service.

The unit was run back approximately 10% on October 8,1996. All systems performed as expected. It was noticed that 'D' Steam Generator continued to lag the other three S/G's slightly but after the initial swell, all S/G levels retumed to program setpoint with no oscillations being noted. No adjustments were required for this test.

Observations & Conclusions 1)

Steam Flows were corrected / normalized at 50% power.

2)

Nozzle Swap from CA to CF did not operate properly. It was tietermined that the stoke time for the CA valves has increased, thus timing out the DFCS timers (60 sec.). Engineering is to pursue the reasons for incrt.51 sed stroke time and/or adjust DFCS intemal timers to resolve this problem.

3)

The original tuning constants for thc 03 Steam Generators were adequate for the Replacement Steam Generators and did not require mei 6astment. The level controls responded as good as or better than the old S/G's. Due to this fact, further testing beyond 70% power was unnecessary.

4)

S/G 'D' lags the other S/Gs during transients. This is due to piping differences and not considered a problem with the DFCS level control. Adjustments have been made to Unit 2 to resolve these differences and similar changes could be applied to Unit 1 if desired. The differences are small and not recommended at this time.

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i Page 40 of 40 5)

The final position of the Main Reg. valves are approximately 55% open. Adjustment could be made to the Fw/Stm D/P program to allow them to go further open thus reducing speed on the Feedpumps.

4.10 Intermediate Range NIS Setpoint Evaluation PT/0//./4150/01, Controlling Procedure for Startup Physics Testing, performs an evaluation of Intermediate Range NIS response in comparison to 70% (Rod Stop) and 25% F.P. (Rx Trip) setpoints preliminarily instated per PT/0/A/4600/05E prior to Unit Startup. This evaluation acquires N35 and N36 indication data as close to 20% and 25% Thermal Power as possible, and then uses it to perform a linear extrapolation to derive expected I/R NIS Channel responses at 31% F.P.

The extrapolated channel responses are then compared to the instated 25% F.P.1/R NIS Rx Trip setpoints to ensure that each channel's indication exceeds the corresponding setpoint value at 31% F.P. This verifies that the Rx Trip setpoints have been established conservatively enough to ensure compliance with alloivable Tech Spec tolerance on this Reactor Protection function.

In the case of C1C10 Startup, this evaluation indicated that 1/R NIS Channel N35 was extrapolated to indicate 68.0 amps at 31% Thermal Power. This indication was noted to be less than the 78.1 amps preliminarily incorporated as the 25% F.P. Rx Trip setpoint, resulting in the channel's declared inoperabilty and entry into the Tech Spec Action item Log with the required Tech Spec action of recalibration of the setpoint at the next available opportunity (Rx Power < 10% F.P., P10 enabled) noted.

The fact that the setpoint was unacceptably non-conservative despite having been derived with an additional 5% of conservatism applied to account for lower Downcomer temperature, was attributable to unanticated assymetry of the radial core power distribution. The 30% F.P. flux map indicated excessive incore Tilt in Quadrants 2 and 3, meaning that the flux levels in Quadrants 1 and 4 (adjacent to N35) were low enough to cause the channel's response to be deficient with respect to the I/R Nis Rx Trip setpoint.

A Past Operability evaluation concluded that N35's indication at an actual Thermal Power level of 31%

did, in fact, exceed the 25% F.P. Rx Trip setpoint. The channel was therefore actually operable at the time the Protection Function was enable (< 10% F.P., P10 enabled) during initial power ascension.

Inherent inaccuracy associated with the I/R NIS response extrapolation ultimately lead to N35's inoperabilty.

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