ML20100H051

Duke Power Co Catawba Nuclear Station,Unit 2 Cycle 8 Startup Rept

ML20100H051
Person / Time
Site:	Catawba
Issue date:	02/29/1996
From:	Mccollum W DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9602270015
Download: ML20100H051 (39)

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l DukelbwerCompany Wlu)A4f R. McCount, JR

. CatawbaNuclear Generation Department licehesident 600ConcordRoad (8W)m-200 Omce Ywk SC29MS (BW)R31-306Fhr DUKEPOWER

-February 21,-1996

' U~.'S. Nuclear Regulatory Commission-

/TTN:

Document Control Desk Washington, DC 20555

Subject:

Catawba Nuclear Station, Unit 2 Docket No. 50-414 Startup. Report,. Unit 2 Cycle 8' In'accordance with Section 6.9.1 of the Catawba Nuclear Station

. Technical Specifications, find attached the Unit 2 Startup Report for Cycle.8 core design.

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

-Very truly yours, j

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R~. McCollum-l i

KEN /U2C8.SR Attachment-xc:

S.-D. Ebneter.

Regional Administrator, Region II R. E. Martin, ONRR R. J. Freudenberger Senior Resident. Inspector I

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Duke Power Company Catawba Nuclear Station Unit 2 Cycle 8 i

STARTUP REPORT

.i February 1996 I

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j TABLE OF CONTENTS EaQB Ust of Tables

...........................11 Ust of Figures iii 1.0 Introduction...................................................

......~.....1 2.0 Precritical Testing...........

....................................2 2.1 TotalCore ReloacAng

........2 2.2 Preliminary NIS Calibration......

2 2.3 Reactor Coolant System Dilution............

2 3.0 Zero Power Physics Testing......

.........6 3.1 1/M Approach to Criticality..............

..........................6 3.2 Source Range / intermediate Range Overlap Data :

.................9 3.3 Point of Nuclear Heat Addition...

..-........9 3.4 Reactivity Computer Checkout................................

.10 3.5 ARO Boron Endpoint Measurement.....

.......................10 3.6 ARO Isothermal Temperature Coefficient Measurement......

-11 3.7 Reference Bank Worth Measurement by Dilution -

.11 3.8 Reference Bank in Boron Endpoint Measurement...

...13 3.9 Differential Boron Worth Determination

13 3.10 Control Rod Worth Measurement by Rod Swap--

13 1

4.0 Power Escalation Testing.=

.15 4.1 Core Power Distribution.....................

15 4.2 One-Point incore/Excore Calibration.....

...19 4.3 Reactor Coolant Delta Temperature Measurement--

.. 20 4.4 Hot Full Power Critical Boron Concentration Measurement.....

21 4.5 incore/Excore Calibration.

21 4.6 Calorimetric Reactor Coolant Flow Measurement.

22 4.7 Unit t.oad Steady State Test.

-23 4.8 Unit Load Transient Test..

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l LIST OF TABLES i

Eage 1.

Core Design Data.....................................................................

.........1 j

i 2.

Preliminary NIS Calibration Data....

.......................................................-...........................4 j

3.

Summary of Zero Power Physics Testing Results..................

.........7 4.-

Source Range / intermediate Range Overlap Data...............

...................9 5.

Nuclear Heat Determination...................................

..10 t

6.

Reactivity Computer Checkout.

-10 7.

ITC Measurement Results................................................................~.....

....... 1 1 8.

Control Rod Worth Measurement Data..........

.................14 9.

Core Power Distribution Results,30% Power -

-- 16 10.

Core Power Distribution Results,60% Power....

............................17 11.

Core Power Distribution Results,100% Power......

.........................................18 12.

One-Point incore/Excore Calibration Results.................

-19 13.

Reactor Coolant Delta Temperature Data...........

.......... 20 14 incore/Excore Calibration Results..........................

.......................22 15.

Calorimetric Reactor Coolant Flow Measurement....

23 16.

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

... 23 17.

NC Loop Hot Leg Temperatures......

=24 18.

NC Loop Average Temperatures.......

.24 19.

NC Loop Delta Temperatures....

24 i

20.

Pressurizer Level Data......

25

^

....... ~...............

21.

Turtine Control Valve Positions.....................

..... 25 22.

NIS Power Range Data =

25 23.

NIS Intermediate Range Data.....

..26 1

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4 LIST OF FIGURES PJIGO 1.

Core Loadog Pattern.......................................................................................

.3 l

2.

ICRR vs. Domin Water Added During Reactor Coolant System Dilution................................... 5 3.

Inverse Count Rate Ratio vs. Control Rod Worth During Approach to Criticality........

.............8 4.

Integral and Difforential Worth of Reference Bank..

..............................................................12 5.

Unit Load Transient Test Power Range Level......................,...............

................27 6.

Unit Load Transient Test NC Loop Highest T-avg T-ref........................................................... 28 7.

Unit Load Transient Test Pressurizer Pressure and Level -

......... 29 8.

Unit Load Transient Test S/G Steam Pressure..................................................

30 i

9.

Unit Load Transient Test S/G Narrow Range Level.........

.......... 31

,10.

Unit Load Transient Test Generator Megawatts and Turtune 1st Stage Pressure.............

-32 11.

Unit Load Transient Test CF Pumps Turtune Speed.....-..............................

....... 33 12.

Unit Load Transient Test S/G Feedwater Flow...............

.......... 34 c

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1.0 INTRODUCTION

Catawba Unit Two Cycle 8 includes a feed batch of 80 MkBW fuel assemblies manufactured by B&W Fuel i

Company (BWFC). The feed batch was enriched to 3.98% (w/o). Burnable poison rod assemblies used in the feed batch were also manufactured by BWFC.

Catawba Unit Two Cycle 8 core loading be0an at 1046 on November 1,1995 and ended at 1534 on i

November 4,1995. Initial criticality for cycle 8 occurred at 1810 on November 28,1995. Zero Power Physics Testing ended at 0200 on November 30,1995. The unit reached full power at 0041 on December 7,1995.

Power escalation testing, including testing at full power, was completed by December 9,1995.

Table 1 contains some important characteristics of the Catawba 2 Cycle 8 core design 4,

TABLE 1 i

C2C8 CORE DESIGN DATA 1.

C2C7 end of cycle burnup: 428 EFPD 2.

C2C8 design length: 445 i 10 EFPD i

Region FuelType Number of Enrichment,

Loading, Cycles Burned Assemblies w/o U" MTU" 8B MkBW 25 3.75 11.405 2

9A MkBW 40 4.00/0.71*

18.248 1

90 MkBW 8

3.60/0.71*

3.650 1

90 MkBW 40 3.50 18.248 1

10A MkBW 80 3.98 36.496 0

Totals 193 88.0466 i

Natural U blanketed fuel assemblies (0.71 w/o enrichment - 6 inches top and bottom)

Design MTU loadings which were used in all design calculations.

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j Page 2 of 34 2.0 PRECRITICAL TESTING i

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Precriticaltesting includes:

= core loading 4

• preliminary calibration of nuclear instrumentation

• dilution of reactor coolant system to estimated critical boron concentration Sections 2.1 through 2.3 describe results of precritical testing for Catawba 2 Cycle 8.

2.1 Total Core Reloading

' The cycle 8 core was loaded under the direction of PT/0/N4150/22, Total Core Reloading. Plots of Inverse Count Rate Ratio (iCRR) versus number of fuel assemblies loaded were kept for each applicable source range and boron dilution mitigation system (BDMS) channel.

Core loading began at 1046 on November 1,1995 and concluded at 1534 on November 4,1995. Core r

loading was verified by PT/0/N4550/03C, Core Verification, which was completed by 2100 on November 4, 1995.

Figure 1 shows the core loading pattem for Catawba 2 Cycle 8.

2.2 Preliminary NIS Calibration Periodic test procedure PT/0/N4600/05E, Preliminary NIS Calibration, 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 full power currents are similarly adjusted. Intermediate range calibration data is checked and revised as necessary during power escalation testing. Due to the T-hot reduction an added conservatism was used to account for any uncertainties that may have been introduced by the T-hot reduction.

Table 2 shows the calibration data calculated by PT/0/N4600/05E. Calculations were performed on November 13,1995. Calibrations were complete by November 26,1995.

2.3 Reactor Coofant System Dilution The reactor coolant system boron concentration was diluted from the refueling boron concentration to the estimated critical boren concentration per PT/0/N415W19,1/M Approach to Criticality, inverse Count Rate 4

Ratio (ICRR) was plotted versus gallons of demineralized water added.

l Initial reactor coolant boron concentration was 2577 ppmB. The estimated critical boron concentration was calculated to be 1870 ppmB. The calculated volume of demineralized water required was 19489 gallons.

This change in boron concentration was expected to decrease ICRR from 1.0 to 0.4.

Reactor coolant system dilution began at 2215 on November 27,1995 and concluded at 0400 on November

+

28,1995. The final reactor coolant system boron concentration, after allowing system to mix, was 1863 ppmB Figure 2 shows ICRR versus volume of water used.

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Page 3 of 34 FIGURE 1 North CORE LOADING PATTERN, CATAW8A UNIT 2 CYCLE 8 A

i 180 FUEL TRANSFER CANAL ->

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1-------

T27 V06 U87 T31 U81 V32 T02 PD PD PD PD PD PD PD 2------

T03 US22 V38 US43 V57 US08 V59 US48 V03 USO9 T36 j

PD RCCA BP RCCA BP RCCA BP RCCA BP RCCA PD 3----

T20 US02 V75 U63 V13 U72 V50 USO V52 U77 V43 US14 T89 PD PD BP RCCA BP RCCA BP RCCA BP RCCA BP PD PD 4____

US21 V48 US32 V31 US10 V34 US23 V01 US17 V56 US27 V29 US20 RCCA BP RCCA BP PD BP RCCA BP PD BP RCCA BP RCCA 5-T19 V15 U62 V67 US29 V65 058 000 U61 V54 US28 V80 U65 V53 T29 PD BP RCCA BP PD BP PD PD PD BP PD BP RCCA BP SS 4

6-V49 US41 V62 US11 V14 T59 V70 U83 V12 T25 V33 US06 V61 US47 V41 PD RCCA BP PD BP RCCA BP RCCA BP RCCA BP PD BP RCCA PD y_

UST V66 U64 V68 U53 V79 US40 V07 US39 V78 U59 V08 U78 V22 U79 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 8-T51 US24 V11 US03 U88 U74 V23 T23 V40 000 U82 US12 V39 US19 T68 go PD RCCA HP RCCA PD RCCA BP RCCA BP RCCA PD RCCA BP RCCA PD 270 g_

U75 V16 U68 V58 U66 V46 US37 V71 US34 V42 U76 V17 U69 V30 U49 PD BP RCCA BP PD BP PD BP PD BP PD BP RCCA BP PD 10 -

VIO US46 V09 US15 V19 T06 V55 056 V72 T40 V20 US18 V04 US45 V02 PD RCCA BP PD BP RCCA BP RCCA BP RCCA BP PD BP RCCA PD 11 _

T12 V28 U64 V77 USOS V73 U70 U51 U73 V76 US38 V27 USS V24 T60 SS BP RCCA BP PD BP PD PD PD BP PD BP RCCA BP PD 12 - - - -

US30 V51 US31 V47 US36 V37 US33 V18 US25 V05 US26 V63 US16 I

I RCCA BP RCCA BP PD BP RCCA BP PD BP RCCA BP RCCA i

13 - - - -

T28 US07 V45 U86 V36 052 V21 U67 V44 U54 V74 USO4 T76 I

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14______

T70 US01 V60 US44 V26 US35 V69 US42 V64 US13 T34 I

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T62 V25 U71 T58 U85 V35 T64 I

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TABLE 2 PREURAINARY NIS CAUBRATION DATA i

Ratio Cycle 7 BOC8 BOC8 c

Channel (BOC 8 +

ReactorTrip ReactorTrip Rod Stop Cycle 7)

Setpoint, Setpoint,

Setpoint, pAmps pAmps pAmps l-N35 0.8 84.8 64.4 51.5 N36 0.8 68.5 52.1 41.7 I

Power Range l

Ratio AxialOffset, Cycle 7 Full Power Current, BOC 8 Full Power Current, Channel (BOC 8 +

pAmps pAmps 4

Cycle 7) 4 t

Upper Lower Upper Lower

+20 282.0 223.9 I

330.2 262.2 N41

~ 0.854 0

290.4 308.8 248.0 263.7 20 250.5 355.2 213.9 303.3 1

-+20 226.3 184.3 l

264.1 215.0 N42 0.857 0

233.0 250.9 199.7 215.0

-20 202.0 286.7 173.1 245.7

+20 259.4 206.1 301.6 239.7 N43 0.860 0

261.4 279.2 224.8 F.40.1

-20' 221.2 318.7 190.2 274.1 t

+20 240.5 194.0 f

280.0 225.9 j

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244.5 262.6 210.0 225.6 20 208.9 299.3 179.4 257.1 i-1 :.

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i 3.0 ZERO POWER PHYSIC 8 TESTING Zero Power Physics Testing (ZPPT) is performed at the beginning of each cycle and is controlled by PT/0/N415001, Controlling Procedure for Startup Physics Testing Test measurements are made below the point of nuclear heat using the output of one power range detector connected to a reactivity computer.

Measurements are compared to predicted data to verify core design. The following tests / measurements are includedin the ZPPT program:

1/M Approach to Criticality Measurement of point of adding heat ReactMty computer checkout All Rods Out critical boron concentration measurement All Rods Ou' isothermal temperature coefficient measurement 1

Measurement of reference bank worth by dilution Reference bank in critical boron concentration measurement Differential boron worth determination Control rod worths measurement by Rod Swap i

Zero power physics testing for Catawba 2 Cycle 8 began at 1447 on November 28,1995 with the beginning j

of rod withdrawal for approach to criticality. ZPPT ended at 0200 on November 30,1995 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 measuremente and results.

3.1 1/M Approach to Criticality Initial crMicality for Catawba 2 Cycle 8 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 1451 on November 28,1995. Criticality was achieved at 1811 on November 28,1995 with Control Bank D at 208 steps withdrawn.

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

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Page 7 of 34 TABLE 3

SUMMARY

OF ZPPT RESULT 8 PREDICTED VAL-PARAMETER MEASURED VALUE UE/ ACCEPTANCE CRITERIA Nuclear Heat 3.4 x 16' amps (N44)

N/A ZPPT Test Band 10* to 1.7x10' amps (N44)

N/A ARO Critical Boron 1869 ppmB 1876i 50 ppmB ARO ITC

-3.14 pcmrF

-3.47 i 2 pcm/*F ARO MTC

-1.37 pcmrF

-1.70 pcmrF Reference Bank (Shutdown Bank 999.5 pcm 900 i 135 pcm B) Worth Ref. Bank in Critical Boron 1731 ppmB 1746 ppmB Differential Boron Worth

-7.24 pcm/ppmB

-6.92 i 1.04 pcm/ppmB Control Bank D Worth 593.5 pcm 584 i200 pcm Control Bank C Worth 809.0 pcm 754 i 226 pcm Control Bank B Worth 803.4 pcm 713 i 214 pcm Controf Bank A Worth 373.6 pcm 360i 200 pcm Shutdown Bank E Worth 570.1 pcm 516 i 200 pcm Shutdown Bank D Worth 469.8 pcm 4361200 pcm Shutdown Bank C Worth 469.9 pcm 437 i 200 pcm Shutdown Bank A Worth 283.7 pcm 269 i 200 pcm TotalRod Worth 5372.4 pcm 4969 i497 pcm

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Page 9 of 34 3.2 Source Rangeantermediate Range Overlap Data During the initial approach to criticality, source range and intermediate range data was obtained to verify that at least one decade of overlap existed. If one decade of overlap did not exist, interrnediate range compensation voltage would have been adjusted to provide the overlap.

Overlap data for Cycle 8 was obtained per PT/0/A/4150/01, Controlling procedure for Startup Physics Testing, on November 28,1995. Table 4 contains the overlap data. The acceptance criterion was met.

TABLE 4 SOURCE RANGElINTERMEDIATE RANGE OVERLAP DATA SOURCE RANGE INTERMEDIATE RANGE N31, cps N32, cps N35, amps N36, amps I

i INITIAL DATA:

NIS Meters 550 500 1 x 10 '"

1 x 10 ~"

OAC 478.4 609.0 1.044 x 10" 1.082 x 10" FINAL DATA:

NIS Meters 15,000 15,000 1.5 x 10 '"

1.5 x 10 -"

OAC 12,340 12,130 1.350 x 10 '"

1.197 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 8, the point of nuclear heat addition was determined per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing, on November 28,1995. Table 5 summarizes the data obtained.

The zero power physics test band was set at 10' to 1.7x10' amps on power range channel N44 (connocted 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.

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Page 10 of 34 TABLE 5 NUCLEAR HEAT DETERMINATION Reactivity Computer intermediate Range Intermediate Range (N44), amps Channel N35, amps Channel N36, amps RUN #1 3.4 x 10

4.515 x 10~'

3.893 x 10

RUN #2 4.5 x 10' 3.6 x 10' 3.1 x 10' 3.4 Reactivity Computer Checkout The reactivity computer checkout was performed per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing, to verify that the power range channel connected to the reactivity computer can provide reliable reactivity data. A reactivity insertion of approximately +20, was rnade. The period is measured and used to determine the theoretical reactivity. The measured reactivity is compared to the theoretical reactivity and verified to be within 4.0%.

The checkout was performed for Cycle 8 on November 28,1995. Table 6 lists the results of the reactMty insortion. 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 390.24 17.69 17.77 0.08 0.45 3.5 ARO Boron Endpoint Measurement l

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 8 beginning of cycle, hot zero power, all rods out, critical boron concentration was measured on November 29,1995 per PT/0/A/4150/10, Boron Endpoint measurement. The ARO, HZP boron concentration was measured to be 1869 ppmB. Predicted ARO critical boron concentration was 1876 ppmB. The acceptance criterion, measured boron wittso 50 ppmB of predicted, was met.

1

Page 11 of 34 3.6 ARO loothermol Temperature Coefficient Measurement The all rods out (ARO) 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 cooldown is initiated, within administrative cooldown limits. When sufficient data (at least 5 'F) is obtained, statepoint 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.

4 The beginning of cycle 8 ITC was measured per PT/0/A/4150/12A, Isothermal Coefficient of Reactivity Measurement, on November 29,1995. No additional cooldown/heatup cycles were required because of good agreement between the heatup and cooldown results. Table 7 summarizes the data obtained during the measurement.

Average ITC was -3.14 pcmrF. Predicted ITC was -3.47 pcmrF. Measured ITC was within acceptance criterion of predicted ITC i 2 pcmrF.

The MTC was determined to be -1.37 pcmPF. This value was used with procedure PT0/A/4150/21, Temporary Rod Withdrawal Umits Determination, 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, pcmrF

$L*F Ag1c

,,,Jg1 Cooldown

-7

+22.0 554.8

-3.14 Heatup

+7

-22.0 554.8

-3.14 Average:-3.14 3.7 Reference Bonk Worth Measurement by Dilution The control rod bank predicted to have the highest worth is designated the reference bank and is measured by inserting the bank (with all other rod banks fully withdrawn) in discrete steps while slowly diluting the reactor coolant boron concentration. 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 8 reference bank (Control Bank C) worth was measured on November 29,1995 per PT/0/A/4150/11 A, Control Rod Worth Measurement by Boration/ Dilution. Figure 4 shows integral worth of reference bank versus bank position. The reference bank was measured to be worth 999.5 pcm; predicted worth was 900 pcm. The acceptance criterion, measured worth within i 15% of predicted, was met.

Page 12 of 34 FIGURE 4 INTEGRAL AND DIFFERENTIAL WORTH OF REFERENCE BANK 1000.0 "

10.0 900.0 9.0 800.0 -

8.0 700.0 7.0 o E

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I 3.8 Reference Bank in Boron Endpoint Measurement This test is performed at the beginning 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 borcn 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 j

reactivity difference from criticality to the " Reference Bank in" configuration is measured and converted to i

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 8 beginning of cycle, hot zero power, reference bank in, critical boron concentration was measured on November 29,1995 per PT/0/A/4150/10, Boron Endpoint measurement. The Reference Bank In Boron Endpoint boron concentration was measure to be 1731 ppmB. Predicted " Reference Bank in" critical boron concentration was 1746 ppmB. There is no quantitative acceptance criteria directly 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 critical boron concentration, and total reactivity worth of reference bank. The calculated value is compared to predicted value to verify consistency. This calculation also provides an indirect check of i

measured reference bank worth and of the boron endpoint measurements.

l The beginning of Cycle 8, hot zero power differential boron worth was calculated to be -7.24 pcm/ppmB per PT/0/A/4150/01, Controlling Procedure for Startup Physics Testing. The predicted value was -6.92 pcm/ppmB. The acceptance criterion (measured within i 15% of predicted), was met.

3.10 Control Rod Worth Measurement by Rod Swap 1

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.

1 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 j

predicted total.

j The beginning of cycle 8 rod worth measurement by rod swap was performed on November 29,1995 per PT/0/A/4150/11B, Control Rod Worth Measurement by Rod Swap. Table 8 summarizes the results. All acceptance criteria were met.

l

e Page 14 of 34 TABLE 8 CONTROL ROD WORTH MEASUREMENT DATA Adjusted Cnbcal Remarrung Difference

% Diff. (Pred -

Reference Position of Worth of Ref.

Measured F%dided (Predicted -

Mees)/Pred x Bank Bank Worth Ref. Bank Bank Alpha Worth. pcm Worth, pcm Measured) 100 Shutdown B (ref. tw*)

N/A N/A N/A N/A 999.5 900

-99.5

-11.1 Shutdown A 1003.5 88 697 1.033 283.7 269

-14.7

-5.5 N^

10032 102.5 611 1.030 373.6 360

-13.6

-3.8 Shutdown D 1002.9 117.5 526 1.013 469.8 436

-33.8

-7.8 Shutdown C 1002.6 118.5 521 1.023 469.9 437

-32.9

-7.5 l

Shutdown E 1002.4 119.5 515 0.839 570.1 516

-54.1

-10.5

• D 1002.1 147 369 1.108 593.5 584

-9.5

-1.6 hB 1001.8 174 225 0.883 803.4 713

-90.4

-12.7 i

wC 1001.5 174.5 222 0.867 809.0 754

-55.0

-7.3 h

5372.4 4969

-403.4

-8.1

e k

Page 15 of 34 4.0 POWER ESCALATION TESTING -

i Power escalation testing is performed during the initial power increase to full power for each cycle and is controlled by PT/0/N4150/01, Controlling Procedure for Startup Physics Testing. Tests are performed 1

from 0% through 100% power with major testing plateaus at ~30%,-65%, and 100% power.

h Significant tests performed during power escalation are:

Core Power Distribution (at ~30%, ~65%, and 100% power)

One-Point incore/Exoore Calibration (at 30% power) '

Reactor Coolant Delta Temperature Measurement (at 65% 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 controlof PT/0/N4150/01) l Unit Load Steady State (at 10%,20%,30%,50%,75%,90% and 100%, This test is not under the controlof PT/0/N4150/01)

Unit Load Transient Test (This test is not unde the control of PT/0/N4150/01)

(

l In addition to the tests listed above, PT/0/N4150/01 performs checks on the incore detector system, on-line thermal power program, intermediate range setpoints, etc. The results of these checks are not included in this report.

s Power escalation testing for Catawba 2 Cycle 8 began on November 30,1995. Full power was reached on December 7,1995. Full power testing was completed on December 9,1995.

Sections 4.1 through 4.6 describe the significant tests performed during power escalation and their results.

4.1 Core Power Distribution i

Core power distribution measurements are performed during power escalation at low power (approximately 30%), intermediate power (approximately 65%), and full power. Measurements are made to verify flux symmetry and to verify enre peaking factors are within limits. Data obtained during this test are also used to check calibration of power range 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 Associates' CORE package and BWFC's MONITOR code, respectively).

The Catawba 2 Cycle 8 core power distribution measurements were performed on December 1,1995 (30% power), December 2,1995 (60% power), and December 7,1995 (100% power). Table 9 through 11 summarize the results. All acceptance criteria were met.

,1 w

,ne

Page 16 of 34 TABLE 9 CORE POWER DISTRIBUTION RESULTS 30% POWER Plant Data Map ID:

FCM/2/08/01 Date of Map:

December 1,1995 Cycle Burnup:

0.210 EFPD Power Level:

30.41 %

4 Control Rod Position:

Control Bank D at 216 steps withdrawn Reactor Coolant System Boron Concentration:

1672 ppmB CORE Results Core Average AxialOffset:

14.589 %

Tilt Ratios for Entire Core Height: Quadrant 1:

0.99223 Quadrant 2:

0.99981 Quadrant 3:

1.00661 Ouadrant 4:

1.00135 Maximum F (nuclear):

1.869 o

Maximum Fu (nuclear):

1.433 Maximum Error between Pred, and Meas Fu:

5.70 %

Average Error between Pred, and Meas. Fu:

1.53 %

Maximum Error between Expected and Measured 5.80 %

Detector Response:

RMS of Errors between Expected and Measured 2.1%

Detector Response:

MONITOR Results Minimum Fo Operational Margin:

23.67 %

Minimum F RPS Margin:

15.21 %

o Minimum F LCO Margin:

54.84 %

Minimum Fu Surveillance Margin:

40.08 %

Minimum Fu LCO Margin:

23.87 %

.1 4.

s t

a Page 17 of 34 TABLE 10 CORE POWER DISTRIBUTION RESULT 8 f

00% POWER Plant Dets Map ID:

FCM/2/08K)2 Date of Map:

December 2,1995 Cycle Bumup:

0.097 EFPD i

Power Level:

59.71 %

Control Rod Position:

Control Bank D at 216 steps withdrawn Reactor Coolant System Boron Concentration:

1510 ppmB 4

CORE Results Core Average AxlalOffset:

7.648 %

Tilt Ratios for Entire Core Height: Quadrant 1:

0.99820 Quadrant 2:

0.99588 Quadrant 3:

1.00347 Quadrant 4:

1.00246 l

Maximum F (nuclear):

1.706 o

Maximum Fu (nuclear):

1.409 Maximum Error between Pred. and Meas F :

6.4%

S Average Error between Pred. and Meas. F -

1.44 %

S Maximum Error between Expected and Measured 6.20 %

Detector Response:

i

^

RMS of Errors between Expected and Measured 2.30 %

Detector Response:

t 1

MONITOR Results Minimum F Operational Margin:

24.99 %

o Minimum F RPS Margin:

20.38 %

o j

Minimum F LCO Margin:

52.19 %

o l

Minimum Fu Surveillance Margin:

34.03 %

l Minimum F LCO Margin:

21.44 %

S i

t

Page 18 0f 34 TABLE 11 CORE POWER DISTRIBUTION RESULTS 100% POWER Plant Data Map ID:

FCM/2/08/03 Date of Map:

December 7,1995 Cycle Burnup:

4.843 EFPD Power Level:

99.87 %

Control Rod Position:

Control Bank D at 211.5 steps withdrawn Reactor Coolant System Boron Concentration:

1262 ppmB CORE hosults Core Average AxialOffset:

-1.733 %

Tilt Ratios for Entire Core Height: Quadrant 1:

0.99563 Quadrant 2:

1.00168 Ouadrant 3:

0.99840 Quadrant 4:

1.00429 Maximum F (nuclear):

1.669 o

Maximum Fu(nuclear):

1.387 Maximum Error between Pred. and Meas Fu:

3.90 %

Average Error between Pred. and Meas. Fu:

1.02 %

Maximum Error between Expected and Measured 3.8%

Detector Response:

RMS of Errors between Expected and Measured 1.4%

Detector Response:

MONITOR Results Minimum F Operational Margin:

2.70 %

o Minimuin F RPS Margin:

22.21 %

Minimum F LCO Margin:

22.27 %

Minimum Fu Surveillance Margin:

4.20%

Minimum Fu LCO Margin:

13.33 %

i

=

l i.

Page 19 of 34

)

4.2 One Point incore/Exoore Calibration j

PT/0/N4600/05D, One-Point incore/Excore Calibration, is performed using results of power range data j

taken at 30% power and the incore axial offset measured at 30%. Power ranges are calibrated before exceeding 50% in order to have valid indications of axial flux difference and quadrant power till ratio for subsequent power increase. The calibration is checked at 60% power. If necessary, power ranges are l

calibrated again per PT/0/A/4600/05D or PT/0/N4600/05A, incore/Excore Calibration.

j Data for Catawba 2 Cycle 8 was obtained on December 1,1995 and all power range calibrations were 1

completed on December 1,1995. Results are listed in Table 12. All acceptance criteria were met.

l 1

TABLE 12 j

ONE-POINT INCORE/EXCORE CALIBRATION RESULT 8 l

4 l

Reactor Power - 30.41%

Axial Offset - 14.589%

Measured Power Rance Currents, pAmps N41 N42 N43 N44 Upper 75.0 56.2 67.2 58.0 Lower 65.7 51.2 61.3 52.5 i

Ratio, Extrapolated (from measured l Currents to " Expected" (from last calibration) Currents j

N41 N42 N43 N44 Upper 0.7689 0.7113 0.7320 0.7003 Lower 0.7953 0.7409 0.7680 0.7258 i

1 New Calibration Currents, pAmps i

Axial N41 N42 N43 N44

Offset, l

Upper Lower Upper Lower Upper Lower Upper Lower i

+20 254.9 206.1 190.9' 161.1 229.3 193.0 197.5 165.3 0

224.1 242.7 168.4 188.0 198.7 224.8 172.4 192.2 i

-20 193.4 279.2 146.0 214.9 168.1 256.6 147.3 219.0 1

t j

L

Page 20 of 34 4

4.3 Reactor Coolant Loop Delta Temperature Measurement Reactor coolant system (NC) hot leg and cold leg temperature data is obtained between 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 (AT,) are valid. AT, is used in the overpower and overtemperature delta temperature reactor protection functions.

PT/0/A/4600/26 was performed at 75% power on December 4,1995 and at 100% power on December 5, 1995. No adjustments were required based on results obtained at 75% power. Loops A, B, C, and D were calibrated using full power results. Table 13 summarizes the test results.

TABLE 13 REACTOR COOLANT DELTA TEMPERATURE DATA Reactor Power = 75.36%

Loop A Loop B Loop C Loop D Meas. T,, 'F 599.9 602.4 599.9 601.3 Meas. Too, 'F 553.8 554.7 554.2 555.1 Calc. Ah, BTU /lb 61.22 63.67 60.60 61.61 Calc. Ah, BTU /lb 81.24 84.49 80.42 81.75 4

o Calc. AT., 'F 59.0 61.1 58.5 59.3 Current AT.,'F 58.1 60.7 58.6 58.9 Difforence,'F 0.9 0.4

-0.1 0.4 Reactor Power = 98.08%

Loop A Loop B LoopC Loop D Meas. T,, 'F 613.9 616.9 614.1 615.8 Meas. Tmo, 'F 555.3 556.3 555.8 557.0 Calc. Ah, BTU /lb 79.92 83.09 79.49 80.64 Calc. Ah, BTU /lb 81.49 84.72 81.05 82.22 o

Calc. AT, 'F 59.7 61.6 59.3 59.9 Current AT,'F 58.1 60.7 58.6 58.9 Difference, *F 1.6 0.9 0.7 1.0

I 4

~

Page 21 of 34 4

4.4 Hot Full Power Critical Boron Concentration Mesourement 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 t

any off-reference condition (e.g. Inserted rod worth, temperature error, difference from equilibrium xenon) and compared to predicted value.

I 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 (150 ppmB).

For Catawba 2 Cycle 8, the hot full power critical boron concentration was measured on December 6, 1995. The measured critical boron concentration was 1259 ppmB. Predicted critical boron concentration

)

was 1270 ppmB; when adjusted for difference at zero power, the adjusted predicted critical boron concentration was 1263 ppmB. The difference between measured and adjusted predicted critical boron concentration was -4 ppmB, which met the acceptance criterion.

4.5 incore/Excore Calibration Excore power range channels are calibrated at full power per PT/0/A/4600/05A, incore/Excore Calibration.

Incore data (flux maps) and power range currents are otAained 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 Catawba 2 Cycle 8 on December 8 and December 9,1995. All power range calibrations were completed on December 9. Eight flux maps, with axial offset ranging from -8.570% to

+3.561% were used. Table 14 summarizes the results. All acceptance criteria were met.

J J

i i

I 4

L

Page 22 of 34 TABLE 14 INCORE/EXCORE Call 8 RATION RESULTS 4

Full Power Currents, Microamps Axial N41 N42 N43 N44

Offset, Upper Lower Upper Lower Upper Lower Upper Lower

+20%

291.8 229.6 220.0 180.7 263.1 216.0 227.0 184.7 0%

253.8 265.9 193.6 210.8 228.8 250.9 197.6 213.3

-20%

215.8 302.3 167.3 240.9 194.5 285.9 168.2 241.8 Correction (M) Factors N41 N42 N43 N44 i

1.396 1.434 1.383 1.415 4.6 Calorimetric Reactor Coolant Flow Measurement With clean venturls, 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 pe r Tech Specifications.

This test was performed on December 5,1995. The Primary Loop D/T Correction was calculated to be 0.98011. Table 15 summarizes the results. All acceptance criteria were met.

Page 23 of 34 TABLE 15 1

CALORIMETRIC REACTOR COOLANT FLOW MEASUREMENT Run Number Secondary Thermal Output %

Primary Thermal Output %

1 98.1885 100.1737 2

98.1309 100.1279 3

98.0303 100.0166 Average 98.1166 100.1077 Primary Loop D/T Correction 0.98011 (Secondary Power / Primary) 4.7 Unit Load Steady State Test in order to verify satisfactory steady state plant operation with reduced T-hot implemented (T-hot reduced

- 3' to extend Steam Generator Ufe) per NSM CN-21367, TT/2/A/9200/092, Unit Load Steady State for NSM CN-21367 was performed at approximately 10%,20%,30%,50%,75%,90% and 100%. With the plant at steady power level data on the following parameters was obtained.

Reactor Power NC Loop Cold Leg Temperature NC Loop Hot Leg Temperature NC Loop Average Temperature NC Loop Delta Temperature Pressurizer Level Turbine Control Valve Positon NIS Power Range Indication NIS Intermediate Range Indication The test was performed at ~10% power on November 30,1995, ~20% on November 30,1995, ~30% on December 1,1995, ~50% on December 1,1995, ~75% on December 4,1995, ~90% on December 4, 1995 and ~100% on December 5,1995. All acceptance criteria were met. Tables 16 through 23 document the results.

Table 16 NC Loop Cold Log Temperatures Power Level NC Loop A NC Loop B NC Loop C NC Loop D 13.94 554.7 555.3 555.0 555.2 17.69 554.5 555.1 554.8 555.1 30.01 553.1 553.8 553.4 553.8 46.96 553.5 554.2 553.9 554.5 75.58 554.1 555.0 554.6 555.5 85.71 553.7 554.6 554.2 555.2 98.54 555.8 556.9 556.4 557.5

Page 24 of 34 Table 17 NC Loop Hot Log Temperatures Power Level NC Loop A NC Loop B NC Loop C NC Loop D 13.94 563.4 564.1 563.3 563.7 17.69 565.5 566.3 565.5 565.9 30.01 571.3 572.8 571.4 572.2 i

46.96 581.9 583.9 582.2 583.2 75.58 599.1 602.2 599.7 601.1 85.71 604.5 607.6 605.2 606.8 98.54 613.2 616.6 614.0 615.6 Table 18 NC Loop Average Temperatures Power Level NC Loop A NC Loop B NC Loop C NC Loop D 13.94 559.5 559.8 559.0 559.4 17.69 560.4 560.8 560.1 560.4 30.01 562.8 563.5 562.4 563.0 46.96 569.4 569.3 568.1 568.9 75.58 577.4 579.0 577.3 578.5 85.71 580.0 581.5 579.9 581.2 98.54 585.4 587.3 585.4 586.9 Table 19 NC Loop Delta Temperatures Power Level NC Loop A NC Loop B NC Loop C NC Loop D 13.94 8.7 8.9 8.3 8.5 17.69 10.9 11.2 10.7 10.8 30.01 18.2 19.0 18.0 18.3 46.96 28.3 29.7 28.3 28.8 75.58 44.9 47.1 45.2 45.7 85.71 50.8 53.0 51.0 51.6 98.54 57.4 59.7 57.7 58.1 j

Page 25 of 34 Table 20 Pressurizer Level Data Power Level PZR Level PZR Level PZR Level PZR Level Setpoint Channel 1 Channel 2 Channel 3 13.94 27.7 26.8 27.4 25.5 17.69 28.7 27.5 28.1 26.2 30.01 31.3 30.7 31.2 29.3 46.96 37.1 36.6 37.0 35.2 75.58 46.6 45.9 46.4 44.8 85.71 49.1 48.4 48.9 47.2 98.54 54.7 54.3 54.7 53.1 Table 21 Turbine Control Valve Positions Power Level CV1 CV2 CV3 CV4 13.97 10.40 %

10.70 %

0.0%

0.0%

17.69 12.5%

12.8 %

0.0%

0.0%

30.01 20.5 %

20.7 %

0.0%

0.0%

46.96 32.0 %

32.3 %

0.0%

0.0%

75.58 100.0 %

100.0 %

17.5 %

0.0%

85.71 100.0 %

100.0 %

46.6 %

0.1%

98.66 100.0 %

100.0 %

100.0 %

35.30 %

Table 22 NIS Power Range Data

( Amps)

Power Level N-41 N-42 N-43 N-44 13.97 65 49 58 51 17.69 77 58 70 59 30.01 140 105 127 110 46.96 226 173 206 176 75.53 378 292 347 297 85.71 432 336 400 343 98.70 509 397 471 402

i Page 26 of 34 Table 23 NIS Intermediate Range Data (gAAmps)

Power Level N-35 N-36 13.94 42 37 17.69 54 47 30.01 88 77 46.96 144 128 75.58 239 214 85.71 273 245 98.54 328 293 4.8 Unit Load Transient Test The Unit Load Transient Test TT/2/A/9200/91, was performed to verify proper operation of the modifications performed on various control systems for NSM CN-21367. T-Hot Reduction. The purpose of the test was to demonstrate proper plant response, including automatic control system performance, to a

~10% step load change. The test verified that the control systems worked 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 Ufting This test satisfied the transient retest as required for the Post Modification Testing for T-Hot Reduction.

This test was performed at ~75% power on December 4,1995. All acceptance criteria for the test were met. Figures 5 through 12 show plant parameters during the test.

FIGURE 5 Page 27 of 34 Unit Load Transient Test Power Range Level 76.00 75.50 N

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