ML040910267

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Cycle 15 Startup Report
ML040910267
Person / Time
Site: Catawba Duke Energy icon.png
Issue date: 03/22/2004
From: Jamil D
Duke Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML040910267 (27)
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_ Duke D.M. JAMIL UWPower. Vice President A Duke Energy Company Duke Power Catawba Nuclear Station 4800 Concord Rd. / CNO1 VP York, SC 29745-9635 803 831 4251 March 22,2004 803 831 3221 fax U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Catawba Nuclear Station, Unit 1 Docket No. 50-413 Unit 1 Cycle 15 Startup Report Catawba Unit 1 Cycle 15 (C1C15) completed its transition to Westinghouse Robust Fuel Assembly (RFA) with the introduction of the third batch of RFAs into the core design.

Additionally, CiC15 incorporates eight lead test assemblies which are Westinghouse Next Generation Fuel (NGF) assemblies. Power escalation testing including first flux map at full power was completed on January 6, 2004.

Section 14.3.4, item (3) of the Catawba Updated Final Safety Analysis Report requires a summary report to be submitted within 90 days following resumption of commercial power operation if the fuel has a different design. Accordingly, the Unit 1 Cycle 15 Startup Report dated January 2004 is attached.

There are no regulatory commitments contained in this document. Any questions concerning this report may be directed to Kay Nicholson at 803.831.3237.

Attachments xc: L. A. Reyes Regional Administrator S. E. Peters NRR Project Manager E. F. Guthrie I Senior Resident Inspector 5---'2 I-www. duke-energy. corn

Duke Power Company Catawba Nuclear Station Unit 1 Cycle 15 STARTUP REPORT January 2004

TABLE OF CONTENTS Page 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 2.3 Reactor Coolant System Dilution .2 2.4 Control Rod Drop Timing Test .3 3.0 Zero Power Physics Testing .8 3.1 1/M Approach to Criticality .8 3.2 Reactivity Computer Checkout ........................ 11 3.3 Point of Nuclear Heat Addition .11 3.4 Dynamic Rod Worth Measurement .12 3.5 ARO Boron Endpoint Measurement .12 3.6 ARO Isothermal Temperature Coefficient Measurement .12 4.0 Power Escalation Testing .14 4.1 Core Power Distribution .14 4.2 One-Point Incore/Excore Calibration .18 4.3 Reactor Coolant Loop Delta Temperature Measurement .19 4.4 Hot Full Power Critical Boron Concentration Measurement .20 4.5 Incore/Excore Calibration .20 4.6 Calorimetric Reactor Coolant Flow Measurement .22 i

LIST OF TABLES Page

1. C1 C15 Core Design Data ................................................. 1
2. Cycle 14 and Cycle 15 Rod Drop Timing Results ..................................................4
3. Preliminary NIS Calibration Data .................................................. 6
4. Summary of ZPPT Results ................................................. 9
5. Reactivity Computer Checkout ................................................. 11
6. Nuclear Heat Determination ................................................. 11
7. ITC Measurement Results ................................................. 13
8. Core Power Distribution Results, 18% Power ................................................. 15
9. Core Power Distribution Results, 50% Power ................................................. 16
10. Core Power Distribution Results, 100% Power ................................................. 17
11. Intermediate Power Level One-Point Incore/Excore Calibration Results ............................................ 18
12. Reactor Coolant Delta Temperature Data ................................................. 19
13. Incore/Excore Calibration Results ................................................. 21
14. Calorimetric Reactor Coolant Flow Measurement .................................  ; 22 ii

LIST OF FIGURES PaEe

1. Core Loading Pattern, Catawba 1 Cycle 15.................................................. 5
2. ICRR vs. Demin Water Added During Reactor Coolant System Dilution ............................................. 7
3. ICRR vs. Control Rod Worth During Approach to Criticality .................................................. 10 iii

Page 1 of 22

1.0 INTRODUCTION

C1C15 completes Catawba Unit One's transition to Westinghouse Robust Fuel Assemblies (RFA) with the introduction of its third batch of RFAs. The C1C15 core consists of a feed batch of 69 of these fuel assemblies. The feed batch enrichments are 36 RFAs at 4.32% (w/o U-235) with 6 inch 2.6% (w/o U-235) annular blankets, and 33 F/A's at 4.70% (w/o U-235) with 6 inch 2.6% (w/o U-235) annular blankets.

Additionally, C1C15 incorporates 8 Lead Test Assemblies (LTA). These LTAs are Westinghouse Next Generation Fuel (NGF) assemblies enriched to 4.32% (w/o U-235) with 6 inch 2.6% (w/o U-235) annular blankets.

Burnable absorbers accompanying the feed batch are of two designs: Integral Fuel Burnable Absorber (IFBA) and Wet Annular Burnable Absorber (WABA), both manufactured by Westinghouse.

A total of 16 previously discharged Mark-BW fuel assemblies (8 from 1EOl 0 and 8 from 1EOC1 1) have been reinserted for C1C15.

C1C15 core loading commenced at 1219 on December 2, 2003 and concluded at 1009 on December 4, 2003. Initial criticality for Cycle 15 occurred at 2315 on December 18, 2003. Zero Power Physics Testing was completed at 0600 on December 19, 2003. The unit reached full power at 1900 on January 3, 2004.

Power Escalation testing, including first flux map at full power, was completed by 1100 on January 6, 2004.

Table 1 summarizes important characteristics of the Catawba 1 Cycle 15 core design.

TABLE 1 C1C15 CORE DESIGN DATA

1. C1C14 end of cycle burnup: 522.7 EFPD
2. C1 C15 design length: 509 -10/ +15 EFPD Region l FuelType l Number of l Enrichment, l Loading, MTU* l Cycles Burned Assemblies w/o U235 11 Mk-BW 16 3.86 7.2803 3 15A W RFA 12 3.92/2.6* 2 15B W RFA 4 4.47/2.6* 2 16A W RFA 56 4.33/2.6* 1 16B W RFA 28 4.63/2.6* 1 17A W RFA 36 4.32/2.6* 0 17B W RFA 33 4.70/2.6* 35.1665 0 17C W NGF 8 4.32/2.6* ° Totals 193 88.1356
  • 2.60% (w/o U-235) enriched annular Uranium blanket, 6 inches top and bottom

Page 2 of 22 2.0 PRECRITICAL TESTING Precritical testing includes:

  • Core Loading
  • Preliminary Calibration of Nuclear Instrumentation
  • Rod Drop Timing Test Sections 2.1 through 2.5 describe results of precritical testing for Catawba 1 Cycle 15.

2.1 Total Core Reloading The Cycle 15 core was loaded under the direction of PT/0/AN4150/22, Total Core Reloading. Plots of Inverse Count Rate Ratio (ICRR) versus number of fuel assemblies loaded were maintained for each applicable Source Range NIS and Boron Dilution Mitigation System (BDMS) channel.

Core loading commenced at 1219 on December 2, 2003 and concluded at 1009 on December 4, 2003.

Core loading was verified per PT/O/AN4550/03C, Core Verification, which was completed at 1300 on December 4, 2003.

Figure 1 shows the core loading pattern for Catawba 1 Cycle 15.

2.2 Preliminary NIS Calibration Periodic test procedure PT/Q/A/4600/05E, 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 NIS full power currents are similarly adjusted. Intermediate Range (11R) NIS Rod Stop and Reactor Trip setpoints are checked and revised as necessary for initial power ascension. An added conservatism of 20% is applied procedurally to l/R setpoints.

Table 3 shows the calibration data calculated by PT/0/A/4600/05E. Calculations were performed on November 26, 2003. Calibrations were completed on December 14, 2003.

2.3 Reactor Coolant System Dilution The reactor coolant system boron concentration was diluted from the refueling boron concentration to the estimated critical boron concentration per PT/0/A/4150/19B, NC System Dilution Following Refueling. ICRR was plotted versus gallons of demineralized water added.

Initial reactor coolant boron concentration was 2,506 ppmB. The Target Boron Concentration for the dilution was calculated to be 1772 ppmB (Minimum Boron Concentration to Maintain Keff < 0.99 with Shutdown Banks Withdrawn + 50 ppmB conservatism). The calculated volume of demineralized water required was 27,673 gallons. This change in boron concentration was expected to decrease ICRR from 1.0 to 0.51.

Reactor coolant system dilution at -87 GPM was performed from 0339 to 0856 on December 14, 2003. The final reactor coolant system boron concentration, after allowing system to mix, was 1776 ppmB. Figure 2 shows ICRR versus volume of water used.

Page 3 of 22 2.4 Control Rod Drop Timing Test This testing is performed prior to each post-refueling startup to verify that, when dropped from the fully withdrawn position at Hot, No-load conditions, each Rod Cluster Control Assembly (RCCA) completely inserts and that its drop time is < 2.2 seconds (pursuant to Technical Specification Surveillance Requirement 3.1.4.3). The 2.2 second criterion applies to the time measured from beginning of decay of Stationary Gripper coil voltage to Dash Pot entry.

All BOC15 RCCA drop times satisfied the acceptance criterion. Table 2 summarizes not only the BOC15 data, but, for comparison purposes, the BOC14 drop times as well. It should be noted that "Time to DP" is the data to be compared to the 2.2 second criterion. "Time to DP" is a parameter that is measured for the purposes of assessing resistance to the RCCA in the Dash Pot region, which was at one time postulated to be the culprit in increasing drop times industry wide.

i Page 4 of 22 TABLE 2 CYCLE 14 AND CYCLE 15 ROD DROP TIMING RESULTS 1BOC-502 1BOC15-121031 Bank IRod ID Time to DP Time to DP Bottom Bank Rod ID Time to DP Time to DP Bottom CBA H06 1.553 2.014 CBA H06 1.533 2.033 H10 1.513 1.994 H10 1.493 1.993 F08 1.553 2.074 F08 1.553 2.073 K08 1.554 2.075 K08 1.534 2.114 CBB F02 1.534 2.054 CBB F02 1.554 2.115 B10 1.534 2.034 B10 1.554 2.095 K14 1.594 2.134 K14 1.614 2.195 P06 1.555 2.035 P06 1.575 2.076 B06 1.535 1.995 B06 1.555 2.056 F14 1.555 2.095 F14 1.595 2.116 P10 1.536 2.016 P10 1.556 2.097

_ K02 1.536 2.076 K02 1.616 2.177 CBC H02 1.576 2.096 CBC H02 1.556 2.097 B08 1.517 2.037 B08 1.517 2.058 H14 1.517 2.017 H14 1.537 2.058 P08 1.577 2.057 P08 1.557 2.058 F06 1.558 2.038 F06 1.538 2.019 F10 1.558 2.078 F10 1.518 2.019 K10 1.538 2.058 K10 1.538 2.059

_ K06 1.538 2.038 K06 1.538 2.039 CBD D04 1.539 2.039 CBD D04 1.539 2.060 M12 1.519 2.019 M12 1.519 2.060 D12 1.539 2.039 D12 1.519 2.060 M04 1.540 2.060 M04 1.520 2.061 H08 1.540 2.060 H08 1.520 2.041 SBA D02 1.604 2.124 SBA D02 1.584 2.165 B12 1.544 2.064 B12 1.544 2.085 M14 1.645 2.165 M14 1.605 2.146 P04 1.545 2.065 P04 1.545 2.066 804 1.585 2.085 B04 1.585 2.086 D14 1.626 2.146 D14 1.626 2.167 P12 1.586 2.086 P12 1.586 2.107 M02 1.606 2.146 M02 1.586 2.147 SBB G03 1.606 2.106 SBB G03 1.586 2.107 C09 1.567 2.067 C09 1.547 2.068 J13 1.547 2.067 J13 1.547 2.048 N07 1.567 2.067 N07 1.527 2.048 C07 1.548 2.068 C07 1.548 2.069 G13 1.588 2.128 G13 1.568 2.109 N09 1.568 2.068 N09 1.568 2.109

_J03 1.549 2.049 J03 1.529 2.030 SBC E03 1.569 2.109 SBC E03 1.569 2.070 C11 1.529 2.009

  • C11
  • 15294%0 1O*Z030X5W L13 1.570 2.030 L13 1.570 2.071 N05 1.530 2.030
  • N053
  • 1.t530f IIII2.091 6W.

SBD C05 1.550 2.070 SBD C05PV 0y W1550 1111112.071MAX E13 1.550 2.070 E13 1.550 2.071 N11 1.531 2.051 * 'N11 - i: 1t[531' IL03 1.571 2.051 L03 1.551 2.112 SBE H04 1.531 2.011 SBE H04 1.511 2.032 D08 1.552 2.092 08 1.552 2.073 H12 1.532 2.052 H12 1.512 2.093 M08 1.552 2.072 M08 1.552 2.093

  • NGF Lead Test Assembly locations

Page 5 of 22 FIGURE 1 CORE LOADING PATTERN, CATAWBA 1 CYCLE 15 180° ZAIO ZC4A AA ZA30 AMI ZCs& ZA06 I ______ -- _ _ __ PD PD rD rD PD r'D . D O - 0 PD PD PD PD PD ZDSA AAl9 ZM2i ZD3A ZDSS ZD4C ZDO I ZD4A ZD40I ZC23 AA40 2 ______ - PD R320 I PD i P346 PD R32s PD R327 rD R326 PD 01 I I 0 AA61 I ZA44 I ZDGO I ZD20 J ZC63 I ZCS3 I ZD06 ZC54 IZCA ZD94 ZD66 I ZA43 AAIS rD R310 PD R333 V8W09C R309l PD r337 PD I PD PD 3 _ PD rD 01 1 I L_ 10 ZC31 ZD69 ZC60 ZC96 ZCA4 I ZD23 ZC12 ZD26 zC9C ZC91 zcsc ZDS4

-R350 ZCIA- PD RP330 PD PD SWI03 R336 ZD2A 8WI06 ZC34 PD ZD34 PD ZC03 R315 ZC93 PD ZDIA R345 ZD42 ZA36 ZA22 ZD4I ZD99 ZC94 ZCO6 ZD93 ZC49 8W120 PD 8WIOC PD MW1M4 PD PD R342 PD PD s ____ PD PD R340 PD PD ZD3S ZC2A ZD1O ZC13 ZD12 ZCI 9 Zl92 ZCAI ZC69 ZD56 ZC3A ZC3s ZDGA ZC6I ZCA2 PD R338 SW 19 PD PD N.06 PD 6__ PD R301 SS , PD 8MVS R319 PIrD R323 ZC20 ZC04 ZC2C ZD14 ZC46 ZD36 ZCS2 ZDSO AA2S AA46 ZD4C ZC50 ZD22 ZC40 ZD13 PID PD PD PD PD PD PD PD SWVI16 R343 PD 7__ PD PD R339 8WZ02 ZD24 ZCII ZCos ZD6C ZCO I ZCIo ZD31 ZCIS ZD09 ZD03 ZAI2 90 ZA34 ZD02 ZD0S ZC14 SW104 R303 PD P314 PD R302 SWIII R352 8W099 N300 PD 270° 8 _ __ PD rJ11 8W9A R308 ZDIS ZCIC ZCOA ZC29 ZDOC ZC3C ZD2C ZCs5 ZD45 AA45 AA69 ZDSI ZCS6 ZD32 ZC39 PD PD PD PD SIVog9 - R312 rD PD 9 _ __ PD PD R344 8WI112 PD PD PD ZD9I ZC22 ZD 16 ZCOC ZDII1 ZC32 ZD30 ZCA3 ZC65 ZD64 ZC45 ZC33 ZD63 ZC66 ZC99 R324 I'D R329 PD P341 SIM10 PD SS R349 PD 10 ___ IPD R351 PD PD SwIIC ZD2s ZC4I1 ZD29 ZC42 ZD90 ZCo9 ZC95 ZD96 ZD39 ZA14 ZA24 ZD43 ZD19 ZCGC ZC02 PD PD SW10S PD SWIOA PD SWIIA PD PD R307 PD I'D 11 __ __ PD PD R304

,PD ZCAO I ZC92 I ZC59 ZD6l

~ #sSrfes--A-A ZC25 t-A ZC26 ZD)61 I'I) ZC5A ZC90... ZC9A ZIO I ZCIG;.

R31 S PD PD M10'o R321 8%Vl 13 PD I'D I r305 PI) R336 12 .__

_ _ R332 I'D ZA42 ZD)59 ZD95 ZC64 ZC555 OA ZC4C ZC62 ZDIC I ZDSC ZA41 AA13 AASO 13 __-_--- PD PD PD R.322 PD R334 j 51V R3P13 PD R331 PD PD PD 0_

I __ __ _ - _I P)-I 3 I I P

.0 ZC24 ZD44 ZDS3 Z71)52 ZD04 ZDJ39 ZDGS ZD3C ZC30 AAGO AAS9 *I PD R33s PD R348 rD R328 rD R317 PD I 14 .__ _ _ _. PD PI347 0- I I I I 0 l _ g U

_4A16 l . _ ag A a ZC43 v i B AA35 l

ZA39 AA43I ZC44 I I LAtC 1 ^

I I A PD I'D PD I'D PD PD PD I 15 01 _ 0 I I I I I I I I I I

I I I I I

I I I I 1.1 L K J II G F E D C B R P 14

_ _ FUEL ID: AA =BATCI/ 11ZA-=BATCH 15.ZC-=B3ATCI1 1G.ZD-=BATCH 17

_ COMPONENT ID. R---=CONTROL ROD. SS=SECONDARY SOURCE: -W---=BURNABLE POISON.

PD = PLUGGING DEVICE (TIIIMBLE PLUG).

OCycic II Rcinwcit Ocyck 1II I i'zsc.is

Page 6 of 22 TABLE 3 PRELIMINARY NIS CALIBRATION DATA Intermediate Range Ratio Cycle 14 BOC 15 BOC 13 Channel (BOC 15 + Reactor Trip Reactor Trip Rod Stop Setpoint, Cycle 14) Setpoint, Amps Setpoint, Amps iAmps N35 0.7972 6.851 E-05 5.462 E-05 4.370 E-05 N36 0.7410 l 7.899 E-05 5.853 E-05 4.682 E-05 Power Range Ratio Axial Cycle 14 Full Power BOC 15 Full Power Channel (BOC 15 + Offset, % Current, pAmps Current, LAmps Cycle 14)Curn mpuretIms C Upper Lower Upper Lower

+20 293.9 237.1 219.5 177.1 N41 0.7468 0 254.6 277.3 190.1 207.1

-20 215.2 317.5 160.7 237.1

+20 286.0 212.8 208.1 154.9 N42 0.7277 0 247.0 248.6 179.7 180.9

-20 207.9 284.4 151.3 207.0

+20 283.0 219.3 210.6 163.2 N43 0.7442 0 246.2 257.9 183.2 191.9

-20 209.4 296.5 155.8 220.7

+20 226.1 181.2 169.1 135-5 N44 0.7477 0 195.6 213.1 146.3 159.3

-20 165.0 245.1 123.4 183.3

Page 7 of 22 FIGURE 2 ICRR vs. DEMIN WATER ADDED DURING REACTOR COOLANT SYSTEM DILUTION 1.2 1.1 I1 0.9 Ca

=c0.8 0.7 0.6 0.5 0.4 0 5000 10000 15000 20000 25000 Demineralized Water Added, gallons I- expected ICRR - lower limit

  • BDMS A N BDMS B
  • N31 A N32

Page 8 of 22 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, and PT/0/A/4150/01A, Zero Power Physics Testing. Test measurements are made below the Point of Nuclear Heat Addition using the output of one Power Range NIS detector connected to a Westinghouse Advanced Digital Reactivity Computer (ADRC). Measurements are compared to predicted data to verify core design. The following tests/measurements are included in the ZPPT program:

  • 1/M Approach to Criticality
  • Reactivity Computer checkout
  • Measurement of Point of Nuclear Heat Addition
  • Control Rod Worth Measurements via Dynamic Rod Worth Measurement
  • All Rods Out Critical Boron Concentration measurement
  • All Rods Out Isothermal Temperature Coefficient measurement Zero power physics testing for Catawba 1 Cycle 15 began at 2200 on December 18, 2003 commencing with implementation of bucking (gamma compensation) current on the ADRC. ZPPT concluded at 0545 on December 19, 2003 with completion of the ITC Measurement. Table 4 summarizes results from ZPPT. All acceptance criteria were met.

Sections 3.1 through 3.10 describe ZPPT measurements and results.

3.1 1M Approach to Criticality Initial criticality for Catawba 1 Cycle 15 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 BDMS or Source Range (S/R) NIS count rates double.

ICRR is plotted for each S/R NIS and BDMS channel. ICRR data is used to project critical rod position. If projected critical rod position is acceptable, rod withdrawal may continue.

The ECP for C1C15 initial criticality was determined to be Control Bank D at 226 SWD. Rod withdrawal for the approach to criticality began at 2220 with Criticality subsequently achieved at 2315 on December 18, 2003 at a control rod position of 188 SWD on Control Bank D.

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

Page 9 of 22 TABLE 4

SUMMARY

OF ZPPT RESULTS PREDICTED VALUE OR PARAMETER MEASURED VALUE ACCEPTANCE CRITERIA Nuclear Heat 6.174 x 10 7 amps (N41) N/A ZPPT Test Limit 5.557 x 10'7 amps (N41) N/A ARO Critical Boron 1879 ppmB 1845 +/- 50 ppmB ARO ITC -4.26 pcmPF -4.04 +/-2 pcm/PF ARO MTC -2.60 pcmPOF -2.38 pcmPF Control Bank D Worth 750.3 pcm 689+/- 103.4 pcm Control Bank C Worth 822.8 pcm 816+/- 122.4 pcm Control Bank B Worth 635.4 pcm 659 +/- 100 pcm Control Bank A Worth 367.7 pcm 349 +/- 100 pcm Shutdown Bank E Worth 397.9 pcm 418 +/- 100 pcm Shutdown Bank D Worth 456.3 pcm 445 t 100 pcm Shutdown Bank C Worth 437.7 pcm 442t 100 pcm Shutdown Bank B Worth 813.2 pcm 855 +/- 128.3 pcm Shutdown Bank A Worth 220.0 pcm 224 +/- 100 pcm Total Rod Worth 4901.3 pcm 4897 +/- 391.8 pcm

Page 10 of 22 FIGURE 3 ICRR vs. CONTROL ROD WORTH DURING APPROACH TO CRITICALITY 12 1

0.8

= 0.6 C>

0.4 0.2 0

0 10 20 30 40 50 60 70 80 90 100 Rod Worth, %Wfthdrawn 1-- N31 -+-N32

  • ECP 13 lowerIimR 13upperlimit x rod insertionmit

Page 11 of 22 3.2 Reactivity Computer Checkout The reactivity computer checkout was performed per PT/0/A/4150/01A, Zero Power Physics Testing, to verify that the Power Range channel connected to the reactivity computer can provide reliable reactivity data. A Reactivity Insertion of between +25 and +40 pcm via control rod withdrawal is used to establish a slow, stable startup rate over which determination of Reactor Period is performed by the ADRC. The resulting Period is then used by the ADRC to determine the corresponding Theoretical Reactivity.

Measured Reactivity is compared to the Theoretical Reactivity and verified to be within 4.0% or 1.0 pcm (whichever is greater). This evolution is repeated as necessary to ensure compliance with acceptance criterion.

The checkout was performed for Cycle 15 on December 19, 2003. Results are summarized in Table 5.

TABLE 5 REACTIVITY COMPUTER CHECKOUT Period Theoretical Measured Absolute Error Percent Error (seconds) Reactivity (pcm) Reactivity (pcm) (pcm) (%)

199.7 1 31.0 l 30.3 1 0.7 1 -2.27 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 15, the Point of Nuclear Heat Addition was determined per PT/OA/4150/01A, Zero Power Physics Testing, on December 19, 2003. Table 6 summarizes the data obtained.

The Zero Power Physics Test Limit was set at 7.11 x 10-7 amps on Power Range channel N41 (connected to reactivity computer). This test limit provides 10% margin to Nuclear Heat for performance of Dynamic Rod Worth Measurement.

TABLE 6 NUCLEAR HEAT DETERMINATION

Page 12 of 22 3.4 Dynamic Rod Worth Measurement Using the Westinghouse Advanced Digital Reactivity Computer (ADRC), the reactivity worth of each RCCA Bank is measured using Dynamic Rod Worth Measurement (DRWM) technique as follows:

  • Control Bank D is withdrawn (in MANUAL) to fully withdrawn position
  • Flux level is allowed to increase to just below ZPPT Test Limit
  • First RCCA Bank to be measured is inserted in Bank Select Mode in one continuous motion to a Step Demand Counter indication of - 2 Steps Wd
  • Once the ADRC has signaled that it has acquired sufficient data for measurement, the RCCA Bank is returned to fully withdrawn position
  • The next Bank to be tested is then selected and, once flux level has recovered to just below ZPPT Test Limit, the measurement process is repeated
  • This test sequence is repeated until all Control and Shutdown Banks have been measured The measured worth of each RCCA Bank is verified to be within 15% or 100 pcm (whichever criteria is greater) of predicted worth. The sum of the worths of all banks is verified to be within 8% of the sum of predicted worths. This sum is also verified to be > 90% of the predicted total.

The Beginning of Cycle 15 rod worth measurements via DRWM were performed on December 19, 2003 per PT/O/A/4150/01 A, Zero Power Physics Testing. Results are summarized in Table 4. All acceptance criteria were met.

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 in conjunction with DRWM. Reactor Coolant System boron samples are obtained at 30 minute intervals during DRWM. The reactivity difference from criticality to the all rods out (ARO) configuration is measured 9 times over the course of DRWM. These reactivities are averaged to determine the amount of control rod insertion at just critical core conditions. This reactivity is converted to equivalent boron (using the predicted differential boron worth) and added to the average of the boron samples obtained during DRWM to obtain the ARO critical boron concentration.

The Cycle 15 beginning of cycle, hot zero power, all rods out, critical boron concentration was measured on December 19, 2003 per PT/0/AN4150/01A, Zero Power Physics Testing. The ARO, HZP boron concentration was measured to be 1879 ppmB. Predicted ARO critical boron concentration was 1845 ppmB. The acceptance criterion (measured boron within 50 ppmB of predicted) was therefore met.

3.6 ARO Isothermal Temperature CoefficIent Measurement The ARO Isothermal Temperature Coefficient (ITC) 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.

Page 13 of 22 The Isothermal Temperature Coefficient of Reactivity is measured as follows:

  • A cooldown at -10 OF/hour is initiated.
  • Once a constant cooldown rate is established, data gathering on the reactivity computer is initiated.
  • After at least 1.1 OF of data is obtained and the error analysis performed by the reactivity computer indicates < 0.1, the cooldown is halted.
  • A heatup at -10 OF/hour back to 557 OF is then initiated. Once a constant heat-up rate is established, data gathering on reactivity computer is initiated and subsequently halted when measurement criteria are satisfied.

Control rod motion is limited to that required to maintain flux below the testing limit. The cooldown/heatup cycle is repeated if additional data is required.

The Beginning of Cycle 15 ITC was measured per PT/0NA/4150/01A, Zero Power Physics Testing, December 19, 2003. No additional cooldown/heatup cycles were required due to good agreement between initial heatup and cooldown results (difference between the measurements S 1.0 pcm/OF). Table 7 summarizes the data obtained during the measurement.

Average ITC was determined to be -4.26 pcm/IF. Predicted ITC was -4.04 pcm/OF. Measured ITC was therefore within acceptance criterion of predicted ITC +/- 2 pcm/OF.

The MTC was determined to be -2.60 pcm/OF. Since the MTC was measured to be negative, compliance with Tech Spec 3.1.3 and SR 3.1.3.1 was ensured without performance of procedure PT/0/AN4150/21, Temporary Rod Withdrawal Limits Determination. Performance of this procedure was waived per PT/0/AN4150/01, Controlling Procedure for Startup Physics Testing.

TABLE 7 ITC MEASUREMENT RESULTS I Average Temp

('F)

ITC (pcmPF) 1 Cooldown 556.3 II -4.22 Heatup 556.6 l -4.29 l Average I 556.5 -4.26

Page 14 of 22 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/4150101, Controlling Procedure for Startup Physics Testing. Tests are performed from 0%through 100% power with major testing plateaus at -18%,-50%, and 100% power.

Significant tests performed during C1 C1 5 Power Escalation were:

  • Core Power Distribution (at -18, -50, and 100% power)
  • One-Point Incore/Excore Calibration (at -50% 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/O/A/4150/01)
  • 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 Incore Detector System, and on-line (OAC) Thermal Power program. The results of these are not included in this report.

Although ZPPT for Catawba 1 Cycle 15 was completed on December 19, 2003, Power Escalation Testing was not commenced until December 31, 2003. During this interval, Reactor Power was limited to 10%

Full Power (F.P.) due to unavailability of the Main Generator (due to Hydrogen Cooler seal leakage). Full power was reached on January 3, 2004. Full power testing was completed on January 8, 2004. Sections 4.1 through 4.7 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 (< 40%

F.P.), Intermediate Power (between 40% F.P. and 80% F.P.), and High Power (> 90% F.P.).

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 (evolved from Shanstrom Nuclear Associates' CORE package and FCF's MONITOR code, respectively).

The Catawba 1 Cycle 15 Core Power Distribution measurements were performed on December 31, 2003 (18% power), January 1, 2004 (50% power), and January 6, 2004 (100% power). Tables 8 through 10 summarize the results. All acceptance criteria were met.

Page 15 of 22 TABLE 8 CORE POWER DISTRIBUTION RESULTS 18% POWER Plant Data Map ID: FCM/1/15/001 Date of Map: December 31, 2003 Cycle Burnup: 1.091 EFPD Power Level: 17.698% F.P.

Control Rod Position: Control Bank D at 210 Steps Wd Reactor Coolant System Boron Concentration: 1707 ppmB COMET Results Core Average Axial Offset: 26.929%

Tilt Ratios for Entire Core Height: Quadrant 1: 1.01743 Quadrant 2: 1.00553 Quadrant 3: 0.99935 Quadrant 4: 0.97769 Maximum F0 (nuclear): 2.358 Maximum Fm (nuclear): 1.517 Maximum Error between Pred. and Meas Fm: 8.99%

Average Error between Pred. and Meas. FH: 2.99%

Maximum Error between Expected and Measured 9.17%

Detector Response:-__

RMS of Errors between Expected and Measured 3.87%

Detector Response: N Minimum F0 Operational Margin: 25.80%

Minimum Fa RPS Margin: 4.89%

Minimum F0 Steady State Margin: 49.00%

Minimum Fm Surveillance Margin: 24.09%

Minimum FAH Steady State Margin: 21.25%

Page 16 of 22 TABLE 9 CORE POWER DISTRIBUTION RESULTS 50% POWER Plant Data Map ID: FCM1/1 3/002 Date of Map: January 1, 2004 Cycle Burnup: 1.369 EFPD Power Level: 49.39% F.P.

Control Rod Position: Control Bank D at 215 Steps Wd Reactor Coolant System Boron Concentration: 1618 ppmB COMET Results Core Average Axial Offset: 9.133%

Tilt Ratios for Entire Core Height: Quadrant 1: 1.01417 Quadrant 2: 1.00171 Quadrant 3: 1.00064 Quadrant 4: 0.98349 Maximum F0 (nuclear): 1.905 Maximum FAH (nuclear): 1.498 Maximum Error between Pred. and Meas FAH: 9.58%

Average Error between Pred. and Meas. FAH: 3.31%

Maximum Error between Expected and Measured 10.50%

  • Detector Response:

RMS of Errors between Expected and Measured 4.20%

Detector Response:

Minimum F0 Operational Margin: 23.48%

Minimum F0 RPS Margin: 9.5%

Minimum F0 Steady State Margin: 58.80%

Minimum Fm Surveillance Margin: 23.78%

Minimum FAH Steady State Margin: 27.28%

  • Nuclear Design and Reactor Support (NDRS) reviewed the 50% flux map, with particular attention to errors which challenged the UFSAR Section 14.3.3 acceptance criteria for Flux Symmetry check. Errors determined to be attributable to unanalyzed Shutdown Cooling (Pu build-in) sustained by Mark-BW reinsert assemblies during extended period in Spent Fuel Pool. Based on NDRS recommendation, power ascension continued to allow evaluation of core power distribution at Full Power. NDRS evaluation is documented in PT/OA/4150/001.

, ; . :I I

- ~::  :

Page 17 of 22 TABLE 10 CORE POWER DISTRIBUTION RESULTS 100% POWER Plant Data Map ID: FCM/1/15/003 Date of Map: January 6, 2004 Cycle Burnup: 5.377 EFPD Power Level: 99.854% F.P.

Control Rod Position: Control Bank D at 215 Steps Wd Reactor Coolant System Boron Concentration: 1230 ppmB COMET Results Core Average Axial Offset: -1.391%

Tilt Ratios for Entire Core Height: Quadrant 1: 1.01316 Quadrant 2: 0.99808 Quadrant 3: 0.99531 Quadrant 4: 0.99345 Maximum Fl (nuclear): 1.728 Maximum Fm (nuclear): 1.512 Maximum Error between Pred. and Meas Fm: 7.76%

Average Error between Pred. and Meas. Fm: 3.01%

Maximum Error between Expected and Measured 8.15%

  • Detector Response: l RMS of Errors between Expected and Measured 3.79%

Detector Response:

Minimum Fa Operational Margin: 0.17%

Minimum Fa RPS Margin: 8.12%

Minimum Fa Steady State Margin: 25.28%

Minimum F&H Surveillance Margin: 11.96%

Minimum F,&H Steady State Margin: 19.43%

  • Reaction Rate Error in excess of 10% noted at 50% F.P. now within UFSAR Section 14.3.3 Flux Symmetry check criteria due to depletion of excess plutonium in Mark-BW reinsert assemblies.

Page 18 of 22 4.2 One-Point Incore/Excore Calibration PT/0/A/4600/05D, One-Point Incore/Excore Calibration, is performed as necessary using results of Power Range (P/R) NIS data taken during power ascension flux maps and measured incore axial offset obtained from them. Calibration of the P/R NIS is necessary if difference between indicated excore and measured incore AFD exceeds 2%. For C1 C1 5 Startup, no calibration was necessary for Low Power Flux Map.

However, calibration was required by the Intermediate Power Level Flux Map (obtained at 50% F.P.).

Power Range channel calibration was required to be completed prior to exceeding 90% in order to have valid indications of Axial Flux Difference and Quadrant Power Tilt Ratio for subsequent power ascension.

Data for the Intermediate Power Level calibration was obtained on January 1, 2004 and all P/R NIS calibrations were completed on January 2, 2004. Results are presented in Table 11. All acceptance criteria were met.

TABLE 11 INTERMEDIATE POWER LEVEL ONE-POINT INCORE/EXCORE CALIBRATION RESULTS Reactor Power = 49.39% Axial Offset = 9.133%

Measured Power Range Currents, uAmos N41 N42 N43 N44 l Upper 78.1 81.4 77.0 61.2 Lower j 75.0 73.8 70.7 60.5 Ratio, Extrapolated (from measured) Currents to "Expected" (from last calibration) Currents l N41 N42 N43 N44 Upper 0.6409 0.6927 0.6671 0.6748 l Lower Jj 0.6497 0.7176 0.6665 0.6996 New Calibration Currents, pAmps Axial N41 N42 N43 I N44 Offset,

+20 Upper 170.5 l Lower 139.1 1 Upper 178.0 (Lower 136.9 Upper 167.8 J Lower 130.6 Upper 133.7 j Lower 111.8 0 147.7 162.7 153.7 160.0 146.0 153.6 115.7 131.5

-20 124.8 186.3 129.4 183.0 124.1 176.6 97.6 151.3

Page 19 of 22 4.3 Reactor Coolant Loop Delta Temperature Measurement Reactor Coolant System Hot Leg and Cold Leg temperature data is obtained at a stable power level between 75% F.P. and 95% F.P.; 'and then subsequently at 100% F.P. per PT/O/A14600126, NC Temperature Calibration, to ensure that full power delta temperature constants (ATo) are valid. AT0 is used in the Overpower and Overtemperature Delta Temperature reactor protection functions.

In the case of C1 C1 5, power ascension was halted at 94% F.P.-on January 3, 2004, to allow evaluation of the four pre-existing loop ATo's. All four channel constants were found to be unacceptable (calculated ATo constants exceeding existing constants by more than 0.6 0F). New ATo constants were generated per PT/O/A/4600/026. Upon completion of AT Process Channel calibrations, power ascension was resumed.

At 100% F.P., on January 5, 2004, ATO's were re-evaluated, and all were found to be acceptable.

Table 12 summarizes the test results.

TABLE 12 REACTOR COOLANT DELTA TEMPERATURE DATA Reactor Power = 94.2048%

Loop A Loop B Loop C Loop D Meas. THOT, OF 611.7 607.9 613.2 609.8 Meas. TcOLD, 'F 552.0 551.8 551.7 552.7 Calc. Ah, BTU/lb 80.55820 75.08170 83.16302 76.79067 Calc. AhN, BTU/lb 85.51390 79.70100 88.27895 81.51460 Calc. ATO, 'F 63.1 59.2 64.9 60.3 Current ATO, OF 61.7 58.0 63.7 59.5 Difference, 0F +1.4 +1.2 +1.2 +0.8

Page 20 of 22 4.4 Hot Full Power Critical Boron Concentration Measurement The Hot Full Power critical boron concentration is measured using PT/O/A/4150/04, Reactivity Anomaly Calculation. Reactor Coolant boron concentration is measured (average of three samples) with reactor at essentially all rods out, Hot Full Power, equilibrium xenon conditions. The measured boron is corrected for any off-reference condition (e.g. inserted rod worth, temperature error, difference from equilibrium xenon) and compared to predicted value.

A simple assessment of the accuracy of the predicted excess reactivity of the new core is performed by comparing the difference between predicted Beginning of Life HZP and HFP critical boron concentrations with the difference between measured BOL HZP and HFP critical boron concentrations. Acceptance criteria is met by verifying that Measured ABoron is within +/-50 ppmB of Predicted ABoron.

For Catawba 1 Cycle 15, the Hot Full Power critical boron concentration was measured on January 5, 2004. The measured HFP, ARO critical boron concentration was 1234 ppmB. Predicted HFP critical boron concentration was 1219 ppmB. The ARO Boron Endpoint Measurement during ZPPT yielded a measured HZP Boron Concentration of 1879 ppmB (prediction being 1845 ppmB). The Predicted ABoron was therefore 626 ppmB, while the Measured ABoron was 645 ppmB. The difference of 19 ppmB between these two parameters satisfied 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 P/R NIS currents are obtained at various axial power distributions. -A least squares fit of the output of each detector (upper and lower chambers) as a function of measured incore axial offset is determined. The slopes and intercepts of the fit for the upper and lower chamber for each channel are used to determine calibration data for that channel.

This test was performed for Catawba 1 Cycle 15 on January 6, 7 and 8, 2004. All Power Range NIS calibrations were completed on January 8. Nine flux maps, with axial offsets ranging from -12.007% to

+6.006% were used. Table 13 summarizes the results. All acceptance criteria were met.

I; Page 21 of 22 TABLE 13 INCORE/EXCORE CALIBRATION RESULTS Full Power Currents, Microamps Axial N41 N42 N43 N44

Offset, e Upper J Lower Upper ppr Lower Upper] Lower

+20% 185.6 146.4 190.9 142.7 179.2 136.2 142.8 116.1

.0% 161.4 172.4 166.3 167.9 156.7 160.7 124.8 137.7

-20% 137.2 198.4 141.7 193.1 134.2 185.2 106.8 159.3 Correction (Mj) Factors

Page 22 of 22 4.6 Calorimetric Reactor Coolant Flow Measurement With clean Main Feedwater Flow venturis, PT/1/A/4150/13B, Calorimetric Reactor Coolant Flow Measurement is performed to validate the Operator Aid Computer's calculations of Reactor Thermal Power and Reactor Coolant Flowrate.

The results of this test, performed for C1 C15 on January 29, 2004, are summarized in Table 14. The test was not performed at Full Power due to reduction to 99% F.P. imposed by Digital Feedwater Control System malfunction. However, performance of this test at any power level > 97% F.P. is permissible. All acceptance criteria were met and adequate margin to Technical Specification Minimum Reactor Coolant Flow limit of 388,000 GPM was demonstrated.

TABLE 14 CALORIMETRIC REACTOR COOLANT FLOW MEASUREMENT Run Number Total Calculated Reactor Coolant Flow (GPM)

Percent Tech Spec Flow

(%)

} Calculated Thermal Power Level

(%)

1 391,354 100.864 98.776 l 2 391,413 100.880 98.816 1 3 391,543 100.913 98.753 Average l 391,437 1 100.886 l 98.782

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