ML020800631

Submittal of Three Mile Island Unit 1 Cycle 14 Startup Report

ML020800631
Person / Time
Site:	Three Mile Island
Issue date:	03/07/2002
From:	George Gellrich AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-02-20054
Download: ML020800631 (29)

Text

AmerGe ....

AmerGen Energy Company, LLC Telephone: 717 944-7621 An Exelon/Britisli Energy Company Three Mile Island Unit 1 Route 441 South, P.O. Box 480 Middletown, PA 17057 March 7, 2002 5928-02-20054 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

THREE MILE ISLAND, UNIT 1 (TMI UNIT 1)

OPERATING LICENSE NO. DPR-50 NRC DOCKET NO. 50-289 Cycle 14 Startup Report Enclosed is the Startup Report for TMI Unit 1 Cycle 14 Operation. Initial criticality for Cycle 14 was achieved at 1:45 PM on December 4, 2001. In all cases the applicable test and Technical Specifications (TS) limits were met.

This report is being submitted in accordance with TMI Unit 1 TS 6.9.1.A. No NRC response to this letter is necessary or requested.

Respectfully, George H. Gellrich Plant Manager Enclosure cc: H. J. Miller, USNRC, Regional Administrator, Region I T. G. Colburn, USNRC, Senior Project Manager, TMI Unit 1 J. D. Orr, USNRC, Senior Resident Inspector, TMI Unit 1 File No. 02035

AmerGen M An Exelon/British Energy Company TMI-1 CYCLE 14 STARTUP REPORT TMI SYSTEM ENGINEERING February 2002

TABLE OF CONTENTS PAGE 1.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER -

SUMMARY

........ 1 2.0 CORE PERFORMANCE - MEASUREMENTS AT POWER -

SUMMARY

................... 4 3.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER ................................ 6 3.1 Initial Criticality .................................................................................................... . 6 3.2 N uclear Instrum entation Overlap ............................................................................ 8 3.3 R eactim eter C heckout ........................................................................................... 8 3.4 ARO Critical B oron Concentration ....................................................................... 9 3.5 Temperature Coefficient Measurements ................................................................. 10 3.6 Control Rod Group Worth Measurements .............................................................. 11 3.7 D ifferential B oron W orth ....................................................................................... 16 4.0 CORE PERFORMANCE - MEASUREMENTS AT POWER ........................................... 18 4.1 Nuclear Instrumentation Calibration at Power ........................................................ 18 4 .2 Incore D etector Testing ......................................................................................... 20 4.3 Power Imbalance Detector Correlation Test ........................................................... 21 4.4 Core Pow er Distribution Verification ..................................................................... 24 1

1.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER -

SUMMARY

Core performance measurements were conducted during the Zero Power Test Program which began on December 4, 2001 and ended on December 5, 2001. This section presents a summary of the zero power measurements. In all cases, the applicable test and Technical Specifications limits were met. A summary of zero power physics test results appears as Table 1-1.

Throughout this report, deviations are calculated as follows:

Deviation = (Measured - Predicted)/Predicted

a. Initial Criticality Initial criticality was achieved at 1345 on December 4, 2001. Reactor conditions were 530.80 F and 2155 psig. Critical conditions were achieved with rod groups 1 through 6 withdrawn to 100%; group 7 at 89% WD; group 8 at 30.4% WD, and boron concentration at 2162 ppmB. Initial criticality was achieved in an orderly manner and within the acceptance criteria of 2186 + 50 ppmB.

b. Nuclear Instrumentation Overlap The overlap between the source and intermediate range detectors was greater than 1.55 decades, exceeding the 1 decade minimum required by Technical Specifications.

c. Reactimeter Checkout An on-line functional check of the reactimeter using the average of NI-3 and 4 was performed after initial criticality. Reactivity calculated by the reactimeter was within 5% of the core reactivity determined from doubling and halving time measurements.

d. All Rods Out Critical Boron Concentration The measured all rods out critical boron concentration of 2183 ppmB was within the acceptance criteria of 2177 +/- 50 ppmB.

e. Temperature Coefficient Measurements The measured temperature coefficient of reactivity at 53 20 F, zero power was +0.84 pcm/F, within the acceptance criteria limit of <+9.0 pcm/F.

f. Control Rod Group Worth Measurements The measured results for control rod worths of groups 5, 6 and 7 conducted at zero power (532°F) using the boron/rod swap method were in good agreement with predicted values.

The maximum deviation between measured and predicted worths was 5.2%, which was for CRG-7 worth. This was within the acceptance criterion for group worth of +/-15%. The deviation for the combined Group 5 - 7 worth was approximately zero, well within the

+10% acceptance criterion.

g. Differential Boron Worth The measured differential boron worth at 532'F was 1.4% more than the predicted value.

This is within the bounds of the FSAR and Framatome ANP (FRA-ANP) supplied limits of

+15%.

TABLE 1-1

SUMMARY

OF ZERO POWER PHYSICS TEST RESULTS CYCLE 14 Parameter Acceptance Criteria Measured Value Deviation Critical Boron 2186 + 50 ppmB 2162 ppmB -24 ppmB NI Overlap >1 decade >1.55 decade N/A Sensible Heat N/A 1.3 x 10-7 amps N/A All Rods Out Boron 2177 + 50 ppmB 2183 ppmB -6 ppmB Concentration Temperature Coefficiient -0.55 pcm/°F -0.70 pcm/°F -0.15 pcm/0 F (2154 ppmB) + 2 pcm/ 0 F Moderator Coefficieent <9.0 pcm/°F +0.84 pcm/0 F N/A Integral Rod Worths 2987 pcm + 10% 2986.4 pcm 0%

(532-F) GP 5-7 951 pcm + 15% 1001 pcm +5.2%

Group 7 Group 6 870 pcm +/- 15% 825 pcm -5.1%

Group 5 1166 pcm + 15% 1160 pcm -0.5%

6.342 pcm/ppmB + 15% 6.431 pcm/ppmB +1.4%

Differential Borox Worth (1923 ppm]B) 2.0 CORE PERFORMANCE - MEASUREMENTS AT POWER -

SUMMARY

This section summarizes the physics tests conducted with the reactor at power. Testing was performed at power plateaus of approximately 10, 30, 45, 79, and 100% core thermal power.

Operation in the power range began on December 5, 2001.

Gadolina is again present in the TMI- 1 core as an integral burnable poison. Ninety six (96) assemblies containing gadolina were reloaded from Cycle 13 and earlier. Seventy two (72) assemblies containing gadolina were loaded fresh for Cycle 14. These assemblies require no special monitoring.

A Lead Test Assembly (LTA) was created for Cycle 14. Four fuel pins were removed from a Mark B 10 fuel assembly and replaced with four M5 fuel pins. While the assembly as a whole is being burned for its third cycle, the M5 pins are being burned for their fourth cycle to determine the characteristics of M5 cladding at high burnups. The M5 pins were manufactured by the B&W Fuel Company, now Fra-ANP. The LTA was monitored during power escalation testing to ensure that it was not the limiting (hottest) assembly in the core with respect to radial power distribution power peaking.

a. Nuclear Instrumentation Calibration at Power The power range channels were calibrated as required during the startup program based on power as determined by primary and secondary plant heat balance. These calibrations were performed due to power level, boron and/or control rod configuration changes during testing.

b. Incore Detector Testing Tests conducted on the incore detector system demonstrated that all detectors were functioning acceptably. Symmetrical detector readings agreed within acceptance limits.

The plant computer applied background, length and depletion correction factors. The backup incore recorders were operational above 79% full power (FP).

c. Power Imbalance Detector Correlation Test The results of the Axial Power Shaping Rod (APSR) movements performed at approximately 79 %FP show that an acceptable incore versus out-of-core offset slope could be obtained by using gain factors ranging from 3.330 to 3.755 for the power range scaled difference amplifiers. The measured values of minimum DNBR and maximum linear heat rate for various axial core imbalances indicate that the Reactor Protection Trip Setpoints provide adequate protection to the core. Imbalance calculations using the backup recorder provide a reliable alternative to computer calculated values.

d. Core Power Distribution Verification Core power distribution measurements were conducted at approximately 45 %FP under non-equilibrium xenon conditions and at 100 %FP at equilibrium xenon conditions. The maximum measured and maximum predicted radial and total peaking factors are all in good agreement. The largest positive percent difference between measured and predicted values was 4.42% for total peaking at approximately 45 %FP. This met its acceptance criterion of 4.8%. All other assemblies were also within their limits for radial and total peak.

The results of the core power distribution measurements are given in Table 4.4-1. All quadrant power tilts and axial core imbalances measured during the power distribution tests were within the Technical Specification and normal operational limits.

3.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER This section presents the detailed results and evaluations of zero power physics testing. The zero power testing program included initial criticality, nuclear instrumentation overlap, reactimeter checkout, all rods out critical boron concentration, temperature coefficient measurement, control rod worths, and differential boron worth.

3.1 Initial Criticality Initial criticality for Cycle 14 was achieved at 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br /> on December 4, 2001. Reactor conditions were 532. I°F and 2147 psig. Control rod groups 1 through 4 were withdrawn prior to the approach to criticality. Deboration from the refueling concentration to the concentration for criticality of 2162 ppmB occurred prior to the approach to criticality.

Criticality was achieved by withdrawing control rod groups 5 and 6 to 100% and control rod group 7 to 89%.

Throughout the approach to criticality, four plots of inverse subcritical multiplication were maintained by three independent persons. Count rates were obtained from the source range neutron detector channels. Plots of inverse count rate (ICR) versus control rod position were maintained during control rod withdrawal.

The inverse count rate plots maintained during the approach to criticality are presented in Figure 3.1-1. As can be seen from the plot, the response of the source range channels during reactivity additions was very good.

In summary, initial criticality was achieved in an orderly manner. The measured critical boron concentration was 2162 ppmB, within the acceptance criteria of 2186 + 50 ppmB.

Figure 3.1-1 1/M vs. Rod Position 1.2 1

0.8

• NI-11 MNI-11A S0.6 - NI-12 SNI-12A 0.4 0.2 0

0 50 100 150 200 250 300 Rod Index, %

3.2 Nuclear Instrumentation Overlap

a. Purpose Technical Specification (TS) 3.5.1.5 states that prior to operation in the intermediate nuclear instrumentation (NI) range, at least one decade of overlap between the source range NIs and the intermediate range Ms must be observed.

b. Test Method To satisfy the above overlap requirements, core power was increased until the intermediate range channels came on scale. Detector signal response was then recorded for both the source range and intermediate range channels. This was repeated until the maximum source range value was reached.

c. Test Results The results of the initial NI overlap data at 532°F and 2155 psig have shown a >1.55 decade overlap between the source and intermediate ranges.

d. Conclusions The linearity, overlap and absolute output of the intermediate and source range detectors are within specifications and performing satisfactorily. There is at least a one decade overlap between the source and intermediate ranges, thus satisfying T. S.

3.5.1.5.

3.3 Reactimeter Checkout

a. Purpose Reactivity calculations during the Cycle 14 test program were performed using the reactimeter. After initial criticality and prior to the first physics measurement, an online functional check of the reactimeter was performed to verify its accuracy for use in the test program.

b. Test Method After initial criticality was established, the reactimeter and the reactivity calculations were started. Steady state conditions were established and a small amount of positive reactivity was inserted in the core by withdrawing control rod group 7.

Reactivity Measurement and Analysis System (RMAS) software compared the reactivity calculated from the doubling times to the values calculated by the reactimeter. Measurements were taken at approximately +81, and -48 pcm.

c. Test Results The measured values were determined to be satisfactory and showed that the reactimeter was ready for startup testing.

d. Conclusions An on-line functional check of the reactimeter was performed after initial criticality.

The measured data shows that the core reactivity measured by the reactimeter was in good agreement with the values obtained from neutron flux doubling times.

3.4 All Rods Out Critical Boron Concentration

a. Purpose The all rods out critical boron concentration measurement was performed to obtain an accurate value for the excess reactivity loaded in the TMI Unit 1 core and to provide a basis for the verification of calculated reactivity worths. This measurement was performed at system conditions of approximately 532°F and 2155 psig.

b. Test Method Starting from the critical condition, the Group 7 control rods were withdrawn to the full-out position. The resulting reactivity was measured with the reactimeter. The boron equivalent of this reactivity was calculated and added to the measured RCS boron concentration.

c. Test Results The measured boron concentration, corrected for B-10 depletion, with group 7 positioned at 100% withdrawn (WD) was 2183 ppmB.

d. Conclusions The above results show that the measured boron concentration of 2183 ppmB is within the acceptance criteria of 2177 + 50 ppmB.

3.5 Temperature Coefficient Measurements

a. Purpose The moderator temperature coefficient of reactivity can be positive, depending upon the soluble boron concentration in the reactor coolant. Because of this possibility, the Technical Specifications state that the moderator temperature coefficient shall not be positive while greater than 95 %FP. The Core Operating Limits Report (COLR) is more restrictive, requiring a negative coefficient above 80% power. The moderator temperature coefficient cannot be measured directly, but it can be derived from the isothermal temperature coefficient and a known fuel temperature (Doppler) coefficient.

b. Test Method Steady state conditions were established by maintaining reactor flux, reactor coolant pressure, turbine header pressure and core average temperature constant, with the reactor critical at approximately 10-9 amps on the intermediate range. Equilibrium boron concentration was established in the Reactor Coolant System (RCS), make-up tank and pressurizer to eliminate reactivity effects due to boron changes during the subsequent temperature swings. The reactivity value and the RCS average temperature was displayed on the RMAS monitor.

Once steady state conditions were established, a heatup rate was started by closing the turbine bypass valves. After the core average temperature increased by about 5°F core temperature and flux were stabilized and the process was reversed by decreasing the core average temperature by about 7°F. After core temperature and flux were stabilized, core temperature was returned to nearly its initial value.

Calculation of the temperature coefficient from the measured data was performed by dividing the change in core reactivity by the corresponding change in RCS temperature.

c. Test Results The results of the isothermal temperature coefficient measurements are provided below. The predicted values are included for comparison.

In all cases the measured results compare favorably with the predicted values.

RCS Boron Measured ITC Predicted ITC Measured MTC Required MTC p_1mB pcrm/F pcm/F p.cm/F pgm/F 2160 -0.70 -0.55 +0.84 <+9.0

d. Conclusions The measured values of the temperature coefficient of reactivity at 532°F, zero reactor power are within the acceptance criteria of -2.0 pcm/IF of the predicted value. An extrapolation of the moderator coefficient to 80%FP indicated that it was well within the limits of TS 3.1.7.2 and the more restrictive limits of the COLR.

3.6 Control Rod Group Worth Measurements

a. Purpose This section provides comparison between the calculated and measured results for the control rod group worths. The location and function of each control rod group is shown in Figure 3.6-1. The grouping of the control rods shown in Figure 3.6-1 will be used throughout Cycle 14. Calculated and measured control rod group reactivity worths for the normal withdrawal sequence were determined at reactor conditions of zero power, 532°F and 2155 psig. The measured results were obtained using results of reactivity and group position from the RMAS system.

b. Test Method Control rod group reactivity worth measurements were performed at zero power, 532 0F using the boron/rod swap method. Both the differential and integral reactivity worths of control rod groups 5, 6, and 7 were determined.

The boron/rod swap method consists of establishing a deboration rate in the reactor coolant system, then compensating for the reactivity changes by manually inserting the control rod groups in incremental steps.

The reactivity changes that occurred during the measurements were calculated by the reactimeter. Differential rod worths were obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each group were then summed to obtain the integral rod group worths.

c. Test Results Control rod group reactivity worths were measured at zero power, 532'F conditions. The boron/rod swap method was used to determine differential and integral rod worths for control rod group 5 - 7 from 100% to 0% WD The integral reactivity worths for control rod groups 5 through 7 are presented in Figures 3.6-2 through 3.6-4.

These curves were obtained by integrating the measured differential worth curves.

Figure 3.6-1 Control Rod Locations and Group Descriptions for TMI-1 Cycle 14 A

B C

D E

F G

H K

L M

N 0

P R

wx- Group Number Group Number of Control Rods Control Rod Number in the Group Function 1 8 Safety 2 8 Safety 3 8 Safety 4 8 Safety 5 12 Control 6 8 Control 7 9 Control 8 8 APSRA Figure 3.6-2 Integral Worth for CRG-5 1200 1000 800 E

_ 600-400 200 0

-~An n~f 60 70l 80 90 100 URI UG--UPou"tiun,  % Witdraw CRG-5 Position, % Withdrawn Figure 3.6-3 Integral Worth for CRG-6 900 800 700 600 E

0. 500 Cu 400 300 200 100 0

0 10 20 30 40 50 60 70 80 90 100 CRG-6 Position, % Withdrawn Figure 3.6-4 Integral Worth for CRG-7 1200 1000 800 E

' 600 400 200 0

10 20 30 40 50 60 70 80 90 100 0

CRG-7 Position, % Withdrawn Table 3.6-1 provides a comparison between the predicted and measured results for the rod worth measurements. The results show good agreement between the measured and predicted rod group worths. The maximum deviation between measured and predicted worths for a group was +5.2%.

d. Conclusions Differential and integral control rod group reactivity worths were measured using the boron/rod swap method. The measured results at zero power, 532°F indicate good agreement with the predicted group worths. All individual group worths and the combined worth met their acceptance criteria.

3.7 Differential Boron Worth

a. Purpose Soluble poison in the form of dissolved boric acid is added to the moderator to provide additional reactivity control beyond that available from the control rods, burnable poison rod assemblies, and integral burnable poisons. The primary function of the soluble poison control system is to control the excess reactivity of the fuel throughout each core life cycle. The differential reactivity worth of the boric acid was measured during the zero power test.

b. Test Method Measurements of the differential boron worth at 532°F were performed in conjunction with the control rod worth measurements. The control rods worths were measured by the boron swap technique in which a deboration rate was established and the control rods were inserted to compensate for the changing core reactivity. The reactimeter was used to provide a continuous reactivity calculation throughout the measurement. The differential boron worth was then determined by summing the incremental reactivity values measured during the rod worth measurements over a known boron concentration range. The average differential boron worth is the measured change in reactivity divided by the change in boron concentration.

c. Test Results Measurements of the soluble boron differential worth were completed at the zero power condition of 532°F. The measured boron worth was 6.431 pcm/ppmB at an average boron concentration of 1932.3 ppmB. This corresponds to a 1.4%

deviation, which is well within the predicted value of 6.342 pcm/ppmB +/- 15%.

d. Conclusions The measured results for the soluble poison differential worth at 532°F was within 15% of the predicted differential worth.

TABLE 3.6-1 COMPARISON OF PREDICTED VS MEASURED ROD WORTHS CnntrAl TAd flrniin Predicted Worth_ nero Measured WorthW icm Percent Difference CrNntrnl -pnrl CTrou L Predicted Worth I pern Measured Wo h- i)cm 5 1166 -4 15% 1160.3 -0.5%

6 870+/- 15% 825.3 -5.1%

7 951 +/- 15% 1000.8 +5.2%

5-7 2987 +/- 10% 2986.4 0%

4.0 CORE PERFORMANCE - MEASUREMENTS AT POWER This section presents the results of the physics measurements that were conducted with the reactor at power. Testing was conducted at power plateaus of approximately 10%, 30%, 45%, 79%, and 100% of 2568 megawatts core thermal power, as determined from primary and secondary heat balance measurements. Operation in the power range began on December 5, 2001.

Periodic measurements and calibrations were performed on the plant nuclear instrumentation during the escalation to full power. The four power range detector channels were calibrated based upon primary and secondary plant heat balance measurements. Testing of the incore nuclear instrumentation was performed to ensure that all detectors were functioning properly and that the detector inputs were processed correctly by the plant computer. Core axial imbalance determined from the incore instrumentation system was used to calibrate the out of core detector imbalance indication.

The major physics measurements performed during power escalation and at full power consisted of obtaining detailed radial and axial core power distribution measurements. Also, during power escalation, nuclear instrument response was determined for several core axial imbalances. Values of minimum DNBR and maximum linear heat rate were monitored throughout the test program to ensure that core thermal limits would not be exceeded.

4.1 Nuclear Instrumentation Calibration at Power

a. Purpose The purpose of the Nuclear Instrumentation Calibration at Power was to calibrate the power range nuclear instrumentation indication to be no less than 2 %FP of the reactor thermal power as determined by a heat balance and to within +/- 2.5% incore axial offset as determined by the incore monitoring system.

b. Test Method As required during power escalation, the top and bottom linear amplifier gains were adjusted to maintain power range nuclear instrumentation indication to be not less than 2% of the power calculated by a heat balance.

When directed by the controlling procedure for physics testing, the high flux trip bistable setpoint was adjusted. The major settings during power escalation are given below:

Nominal Test Plateau, %FP Nominal Bistable Setpoint, %FP 40 50 80 90 100 105. 1

• Normal full power setpoint

c. Test Results An analysis of test results indicated that changes in Reactor Coolant System boron and xenon buildup or burnout affected the power as observed by the nuclear instrumentation. This was expected since the power range nuclear instrumentation measures reactor neutron leakage which is directly related to the above changes in system conditions. Each time that it was necessary to calibrate the power range nuclear instrumentation, the acceptance criteria of calibration to be no less than 2.0

%FP of the heat balance power was met without any difficulty. Also, each time it was necessary to calibrate the power range nuclear instrumentation, the + 2.5% axial offset criteria as determined by the incore monitoring system was also met when required.

The high flux trip bistable was adjusted to a nominal setpoint of 50, 90 and 105.1

%FP prior to escalation of power to nominal plateaus of 40, 80 and 100 %FP, respectively.

d. Conclusions The power range channels were calibrated based on heat balance power several times during the startup program. These calibrations were required due to power level, boron, and/or control rod configuration changes during the program.

Acceptance criteria for nuclear instrumentation calibration at power were met in all instances.

4.2 Incore Detector Testing

a. Purpose Self-powered neutron detectors (incore detector system) monitor the core power density within the core and their outputs are monitored and processed by the plant computer to provide accurate readings of relative neutron flux.

Tests conducted on the incore detector system were performed to:

(1) Verify that the output from each detector and its response to increasing reactor power was as expected.

(2) Verify that the background, length and depletion corrections applied by the plant computer are correct.

(3) To measure the degree of azimuthal symmetry of the neutron flux.

b. Test Method The response of the incore detectors versus power level was determined and a comparison of the symmetrical detector outputs made at steady state reactor powers of approximately 12.5, 30, 45, 79, and 100%FP.

At approximately 79 %FP, 1301-5.3, Incore Neutron Detectors-Monthly Check, was performed to calibrate the backup recorder detectors to their incore depletion value.

c. Conclusions Incore detector testing during power escalation demonstrated that all detectors were functioning as expected. Symmetrical detector readings agreed within acceptable limits and the computer applied correction factors are accurate. The backup incore recorders were calibrated and were operational above 80 %FP.

4.3 Power Imbalance Detector Correlation Test

a. Purpose The Power Imbalance Detector Correlation Test has four objectives:

(1) To determine the relationship between the core power distribution as measured by the out-of-core detector system (OCD) and the incore detector system (ICD) instruments.

(2) To demonstrate axial power shaping control using the Axial Power Shaping Rods (APSRs).

(3) To verify the adequacy and accuracy of backup imbalance calculations as done in AP 1203-7, "Hand Calculation for Quadrant Power Tilt and Core Power Imbalance."

(4) To determine the core maximum linear heat rate and minimum DNBR at various power imbalances.

b. Test Method This test was conducted at about 79 %FP to determine the relationship between the core axial imbalance as indicated by the incore detectors and the out-of-core detectors. Based upon this correlation, it could be verified that the minimum DNBR and maximum linear heat rate (MLHR) limits would not be exceeded by operating within the flux/delta flux/flow envelope set in the Reactor Protection System.

CRG-8 was moved to establish the various imbalances. The integrated control system (ICS) automatically compensated for reactivity changes by repositioning CRG-7 to maintain a constant power level. The RCS boron concentration was adjusted to obtain additional imbalance data. Again, the ICS compensated for the boron change by inserting CRG-7 to maintain constant power.

c. Test Results The relationship between the ICD and OCD offset was determined at about 79 %FP by changing axial imbalance through adjustment of the APSRs, boron concentration, and resulting Group 7 control rod position. The average slope measured on the four OCDs was 0.966. The lowest slope was 0.949 for NI-6. The scaled difference amplifier gain was changed so that each detector would respond with an acceptable slope.

A comparison of the ICD offset versus the OCD detector offset obtained for each NI channel is shown in Table 4.3-1.

Core power distribution measurements were taken at the most positive and negative imbalances at 79 %FP. The values of minimum DNBR and worst case MLHR were within their respective acceptance criteria.

Backup offset calculations using 1203-7, Hand Calculations for Quadrant Power Tilt and Core Power Imbalance, agree with the computer calculated offset. Table 4.3-2 lists the computer calculated offset as well as offsets obtained using the ICD backup recorders.

d. Conclusions Backup imbalance calculations performed in accordance with AP 1203-7 provide an acceptable alternate method to computer calculated values of imbalance.

Minimum DNBR and MLHR parameters were well within Technical Specifications limitations.

The final slopes of the ICD to OCD correlations were within the acceptance criteria.

TABLE 4.3-1 INCORE OFFSET VS OUT-OF-CORE OFFSET Incore Offset, % Out-of-Core Offset, %

M-5 NI-6 NI-7 NI-8 7.931 6.169 6.230 8.294 6.995 5.925 4.944 5.019 7.036 5.700 1.054 0.410 0.631 2.394 1.150

-1.059 -1.329 -1.021 0.585 -0.644

-18.276 -18.411 -18.169 -17.146 -18.000

-23.734 -23.535 -23.249 -22.478 -23.223 TABLE 4.3-2 FULL INCORE OFFSET VS BACKUP RECORDER OFFSET Full Incore Offset, % Backup Recorder Offset, %

7.931 8.54

-1.059 0.61

-18.276 -15.15

-23.734 -19.58 4.4 Core Power Distribution Verification

a. Purpose To measure the core power distributions during the power escalation and at 100

%FP to verify that the core axial imbalance, quadrant power tilt, maximum linear heat rate and minimum DNBR do not exceed their specified limits. Also, to compare the measured and predicted power distributions.

b. Test Method Core power distribution measurements were performed at approximately 45%FP during the power escalation and at 100 %FP under steady state conditions. To provide the best comparison between measured and predicted results, three-dimensional equilibrium xenon conditions were established for the full power test. Data collected for the measurements consisted of power distribution information at 364 core locations from the incore detector system. The worst case core thermal conditions were calculated using this data. The measured data was compared with calculated predictions.

c. Test Results The acceptance criteria for power distribution require that all new fuel be within limits for radial and total peaking. Also, the RMS of the differences between measured and predicted HFP radial peaks for all fuel (eighth core) should be less than 0.05.

A summary of the cases studied in this report is given in Table 4.4-1. The table lists the core power level, control rod pattern, cycle burnup, boron concentration, axial imbalance, maximum quadrant tilt, minimum DNBR, maximum LHR and power peaking data for each measurement. Note that the radial and total peak data is not necessarily for the maximum peaks in the core, but for the locations with the largest difference between the predicted and measured data for new fuel. The radial peak and total peak limits are shown. The highest Worst Case MLHR was 12.91 kw/ft at 100 %FP which is well below the maximum limit of 20.5 kw/ft. The lowest minimum DNBR value was 3.03 at 100 %FP which is well above the minimum limit.

The quadrant power tilt and axial imbalance values measured were all within the allowable limits. Table 4.4-1 also gives a comparison between the maximum calculated and predicted radial and total peaks for an eighth core power distribution.

d. Conclusions Core power distribution measurements were conducted at approximately 45%FP and 100 %FP. Comparison of measured and predicted results show good agreement. The largest difference between the maximum measured and maximum predicted peak value was 4.42% for total peaking at approximately 45 %FP for location N-13. This met its acceptance criterion of<4.8%. All fuel locations met their acceptance criteria.

The measured values of DNBR and MLHR were all within the allowable limits. All quadrant power tilts and axial core imbalances measured during the power distribution test were within the Technical Specifications and normal operational limits.

TABLE 4.4-1 CORE POWER DISTRIBUTION RESULTS Power Plateau Escalation 45% Steady State 100%

Date 07 December 2001 11 December 2001 Actual Power (%FP) 45.38 100.03 CRG 1-5 (%WD) 100 100 CRG 6 (%WD) 100 100 CRG 7 (%WD) 60.1 91.6 CRG 8 (%WD) 30.2 30.3 Cycle Burnup (EFPD) 0.224 4.15 Boron Concentration (ppmB) 1920 1609 Imbalance (%) -8.34 -2.02 Maximum Tilt (%) 2.74 2.59 MDNBR 6.25 3.033 Worst Case MLHR (kW/ft) 6.83 12.91 Maximum Radial Peak Difference, New Fuel Location N-13 N-13 Measured Peak 1.146 1.173 Predicted Peak 1.123 1.156 Difference (%) 2.05 1.47 Acceptance Criterion (%) *3.8 *3.8 Maximum Total Peak Difference, New Fuel Location N-13 N-13 Measured Peak 1.723 1.423 Predicted Peak 1.650 1.379 Difference (%) 4.42 3.19 Acceptance Criterion (%) *4.8 *4.8 Eighth-Core RMS of Absolute Differences for Radial Peaks, All Fuel Measured 0.022 0.021 Acceptance Criterion _<0.05 _<0.05