ML20237J814
| ML20237J814 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/21/1987 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 5211-87-2159, NUDOCS 8708260328 | |
| Download: ML20237J814 (38) | |
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e TMI-1 CYCLE 6 STARTUP RCPORT l
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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 Nuclear Instrumentation Overlap . . . . . . . . . . . . . . . . . . . . . . . . 11 3.3 Reacti mete r Checkout . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.4 ARO Critical Boron Concentration . . . . . . . . . . . . . . . . . . . . . . . 13 3.5 Temperature Coef ficient Measurements . . . . . . . . . . . . . . . . . . . 14 3.6 Control Rod Group Worth Measurements . . . . . . . . . . . . . . . . . . . 16 3.7 Di f f e renti al Bo ron Wor th . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 i i
4.0 CORE PERFORMANCE - MEASUREMENTS AT POWER ..................... 24 4.1 Nuclear Instrumentation Calibration at Power . . . . . . . . . . . 25 4.2 ' Incore Detector Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.3 Power Imbalance Detector Correlation Test . . . . . . . . . . . . . . 28 4.4 Core Power Distribution Veri fication . . . . . . . . . . . . . . . . . . . 33 l 4.5 Reactivity Coefficients at Power ....................... 35
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. , 1,0 CORE PERFORMANCE - MEASUREMENTS AT 7ERO POWER -
SUMMARY
Core performance measurements were conducted during the Zero Power Test Program which began on March 23, 1987 and ended on March 24, 1987. This section presents a summary of the Isro power measurements. In all cases, the applicable test and Technical Specifications limits were met. A summary of zero power physics test results is included on Table 1-1.
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- a. Initial Criticality Initial criticality was achieved at 1731 on March 23, 1987.
Reactor conditions were 532*F and 2155 psig and control rod groups 1 through 6 were withdrawn to 100% while group 7 was positioned at 85% withdrawn. Control' Rod group 8 was positioned at 25%
withd rawn. Criticality was achieved by deborating the Reactor Coolant f rom 1745 ppm to 1453 ppm. Initial criticality was achieved in an orderly manner and the acceptance criteria of 1384 100 PPM was met.
- b. Nuclear Instrumentation Overlap At least one decade overlap was measured between the source and intermediate range neutron detectors as required by Technical Specifications.
- c. Reactimeter Checkout An on-line f unctional check of the Mod-Comp reactimeter (using NI-3) was performed af ter initial criticality. Reactivity calculated by the reactimeter was within 5% of the core reactivity determined from doubling time measurements,
- d. All Rods Out Critical Baron Concentration The measured all rods out critical boron concentration of 1449 ppmB was within the acceptance criteria of 1394 1 100 ppmB.
- e. . Temperature Coef ficient Measurements The measured temperature coefficients of reactivity at 532*F, zero power were within the acceptance criteria limits over the range of boron concentrations and rod positions that the measurements were made.
- 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 12.0% which was for CRG-6 worth.
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H. Differential Boron Worth
.The measured differential boron worth at 532*F was 6.15% greater than the predicted value. This is within the bounds of the FSAR and'BW supplied limits of i 15%.
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[ TABLE 1-1 Summary of Zero Power Physics Test Results l Cycle _6 Acceptance Measured '
Parameter Criteria Value Critical Boron 1384 100 ppm 1453 ppm
- NI Overlap >1 decade 1.82 decade 4 Sensible Heat N/A 1.0 x 10-7 amps
- All Rods Out Boron Concentration 13941100 ppm 1449 ppm Temperature Coefficient 0.698 pcm/*F 0.515 pcm/*F (1178 ppm) 1 4 pcm/'F Moderator Coefficient 2.348 pcm/*F 2.165 pcm/'F' t 4 pcm/*F Integral Rod Worths (532*F) GPS-7 3318 PCM i 10% 3250.5 PCM Group 7 994 PCM i 15% 963.5 PCM Group 6 847 PCM i 15% 75f>.5 PCM Group 5 1477 PCM i 15% 1530.5 PCM i
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4
' 2.0 CORE PERFORMANCE - MEASUREMENTS AT POWER -
SUMMARY
-l This section sunenarizes the physics tests conducted with the reactor at power. Testing was performed at power plateaus of 28, 75, and 100% core thermal power. Operation in the power range began on March 24, 1987. i Periodic measurements and calibrations were performed on the plant nuclear instrumentation during the escalation to 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 outputs 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 consist of determining the moderator and power doppler coefficients of reactivity and obtaining detailed radial and axial imbalances. Values of minimum DNBR and maximum linear heat rate were monitored throughout the test program to ensure that core thermal limits were not exceeded,
- a. Nuclear Instrumentation Calibration at Power l l
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 required due to power level, boron and/or control rod configuration changes during testing. '
- b. Incore Detector Testina Tests conducted on the incore detector system demonstrated that all detectors except level 2 of string 20 were functioning as expected, that symetrical detector readings agreed within acceptable limits and that the plant computer applied the correct background, length and depletion correction factors. The backup incore recorders were operational above 80% FP as required by Technical Specifications,
- c. Power Imbalance Detector Correlation Test The results of the Axial Power Shaping Rod (APSR) scans perfonned at 75% FP show that an acceptable incore versus out-of-core offset slope of 0.96 is obtained by using a gain factor of 3.684 in 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.
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- d. Core Power Distribution Verification Core power distribution measurements were conducted at 75% and 100%
L full ~ power under steady state equilibrium xenon conditions for I
specified control rod configurations. The maximum measured and maximum predicted radial and total peaking factors are all in good agreement.. The maximum difference between a measured and predicted value was 1.6% for radial peaking at 75% FP. This met acceptance criteria of <5.0%.
Th'e 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 nornal operational limits.
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- e. Reactivity Coefficients at Power The isothermal temperature coefficient measured at 100% FP was
-6.67 pcm/*F. The measured pover doppler coefficient at 100% FP
' was -8.413 pcm/5 FP. All Technical Specification and Safety Analysis requirements were met.
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' 3.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER This secticn pres;ntri the detailed r;sults 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 6 was achieved at 1731 on March 23,1987. Reactor conditions were 532*F and 2155 psig and control rod groups 1 through 4 were previously withdrawn during the heatup to 532*F. The initial reactor coolant system (RCS) boron concentration was 1745 ppm.
The approach to criticality began by withdrawing control rod group 8 to 25% withdrawn, control rod groups 5 and 6 to 100% withdrawn, and positioning group 7 at 855 withdrawn. Criticality was subsequently achieved by deborating the reactor coolant system to a boron concentration of 1453 ppm. The procedure used in the approach to critical is outlined below in three basic steps:
Step 1 Control Rod Withdrawal Group 8 25% withdrawn Group 5 100% withdrawn Group 6 100% withdrawn Group 7 85% withdrawn Step 2 Deborate using a feed and bleed flow rate of 70 gpm until the inverse count rate is between 0.2 and 0.3. At this point, deborate at a feed and bleed flow rate of 40 gpm.
Step 3 At an inverse count rate of 0.1, stop deboration and increase letdown flow to maximum (120gpm). This enhances mixing between the makeup tank and the reactor coolant system. Achieve initial criticality and position control rod group 7 to control neutron flux as the reactor coolant system boron concentration reaches equilibrium.
Throughoist the approach to criticality, plots of inverse multiplication were maintained by two independent persons. Two plots of inverse count rate (ICR) versus control rod position were maintained during control rod withdrawal. Two plots of ICR versus RCS boron concentration and two plots of ICR versus gallons of demineralized water added were maintained during the dilution sequence. During each reactivity addition (boron dilution and control rad withdrawal), count rates were obtained from each source '
range neutron detector channel.
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- During control rod withdrawal (step 1) ICR plots versus control rod grcup p sition were maintained from the outputs of two source range detectors. The withdrawal interval for each control rod group was limited to no more than half the remaining predicted distance to criticality as detern;ined from the ICR plots.
Deboration of the reactor coolant system was accomplished in two steps as indicated above. First, deboration f rom 1745 ppm was commence 3 using a feed and bleed flow rate of 70 gpm (step 2). RCS boron samples were taken every 30 minutes and samples from the makeup tank and the pressurizer were taken hourly. Two ICR plots were maintained vs. gallons of demineralized water added, and two plots were maintained vs. RCS letdown concentration every 30 minutes. Deboration at a letdown rate of 70 gpm was continued until one of the ICR plots was less than 0.3. At this time, demineralized water addition was reduced to 40 gpm. When one of the ICR plots indicated 0.10, the letdown flow rate was increased to 120 gpm in a recirculation mode to expedite mixing in the RCS (step 3). The reactor went critical during the mixing process.
The inverse count rate plots maintained during the approach are presented in Figures 3.1-1 through 3.1-3. As can be seen from the plots, the response of the source range channels during reactivity additions was very good. Figure 3.1-1 is the plot of ICI versus control rod group withdrawal. Figure 3.1-2 is the ICR plots versus RCS boron concentration and Figure 3.1-3 is the ICR plots versus gallons of demineralized water added to the RCS.
In summary, initial criticality was achieved in an orderly manner.
The measured critical boron concentration was within the acceptance criteria of 1384 100 PPM.
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. 3.2 Nuclear Instrumentation Overlap
- a. PurD0se Technical Specification 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 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 source range high voltage cutoff value was reached.
- c. Test Results The results of the initial NI overlap data at 532'F and 2155 psig have shown a 1.82 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.
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3.3- Reactimeter Cherkout
- a. Purpose
. Reactivity calculations during the Cycle 6 test program were performed using the Mod-Comp Reactimeter. Af ter initial.
criticality and prior to the first physics measurement, an online functional check of the reactimeter was performed to .I verify its accuracy for use in the test program, b'~. Test Method Af ter initial criticality and nuclear instrumentation overlap was established, intermediate range channel NI-3 was connected to the reactimeter and the reactivity calculations were started. After steady state conditions with a constant neutron flux were established, a small amount of positive reactivity was inserted in the core by withdrawing control rod
-group 7. Stop watches were used to measure the doubling time of the' neutron flux and the reactivity was determined from the doubling time reactivity curves. The measurement was for + 80 pcm and - 38 pcm. The reactivities determined from doubling time measurements were compared with the reactivity calculated by the reactimeter.
- c. Test Results The results of the reactimeter verification measurements are I provided below. The measured values were determined to be .
satisfactory and showed that the reactimeter was ready for !
startup' testing. i TEST REACTIMETER D.T. REACTIVTY DEVIATION QAl[ (PCM) (PCM) (5) 1 80 82.2 2.68 l
2 - 38 - 37.9 0.26
- d. Conclusions An on-line functional check of the reactimeter was performed af ter initial criticality. The measured data shows that the I
l core reactivity measured by the reactimeter was in good agreement with the values obtained from neutron flux doubling t.ime s .
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~3.4 All Rods Out Critica'l Boron Concentration
- a. Purpose j.
' -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 532*F l .. and 2155 psig.
- b. Tist Method l
' The Reactor Coolant System was borated to an all rods out condition and steady st&te' conditions were established.
! c.. Test Results The measured boron concentration with group 7 positioned at 1009dD was 1449 ppm.
- d. _ Conclusions The above results show that the measured boron concentration of 1449 ppm is within the acceptance criteria of 1394 i 100 ppm.
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. '3.5 Temperature Coef ficient Measurements
- a. Puroose The moderator temperature coef ficient 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 moderator temperature coefficient cannot be measured directly, but it can be derived f rom the core temperature coefficient and a known fuel temperature (isothermal 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, make-up tank and pressurizer to eliminate reactivity effects due to boron changes during the subsequent temperature swings. The reactimeter and recorders were connected to monitor selected core parameters with the reactivity value calculated by the reactimeter and the core average temperature displayed on a two channel strip chart recorder.
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 10'F. After core temperature and flux were stabilized, core temperature was returned to its initial value. This procedure was completed with control rod group 7 at 885 wd. Calculation of the temperature coefficient from the measured data was performed by dividing the change in core reactivity by the corresponding change in core temperature over a specific time period.
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- c. ' Test Results The results of the isothermal temperature coef ficient measurements are provided below. The calculated values are included for comparison.
In all cases the measured results compare favorably with the calculated values.
.. RCS MEASURED PREDICTED MEASURED PREDICTED BORON ITC ITC MTC MTC (PPM) (PCM/DEG F) (PCM/DEG F) (PCM/DEG F) (PCM/DEG F) 1430 0.51 5 0.698 2.165 2.348
- d. Conclusions The measured values of the temperature coefficient of reactivity at 532'F, zero reactor power are within the <
acceptance criteria of i 4.0 pcm/*F of the predicted value.
An extrapolation of the moderator coefficient to 100%FP indicates that it is well within the limits of Technical Specifications 3.1.7.2.
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3.6 Control Rod Group Worth Measurements "
- a. Purpose Thetotalamountofexcessreactivithpresentat beginning-of-cycle (BOC), hot (532*F), clean conditions is 13.76% AK/K. During reactor operations, nearly all of the excess reactivity is controlled by the soluble poison systems. Additional control is provided by moveable control rods. This section provides comparison betwfen the calculated and measured results for the control rod group worths. j, The location and function of each control rod group is shown -
in Figure 3.6-1. The grouping of the control rods shown in '
, 'l Figure 3.6-1 will be used throughout Cycle 6. . Calculated and d measured control rod group reactivity worths for the normal ,'
withdrawal sequence were determined at reactor conditions of zero power, 532'F and 2155 psi. The measured results were obtained using results of reactivity and group position from ;
the strip chart recorders. '
- b. Test Method Control rod group reactivity worth measurements were performed at zero power, 532*F using the boron /rr>d swap method. Both the differential and integral reactivity worths of control rod groups 5, 6, and 7 were determined.
The boron swap method consisted of establishing a deboration rate in the reactor coolant system and compensating for the reactivity changes by inserting the control rod groups in incremented steps. The reactivity changes that occurred during the measuremer.ts were calculated by the reactimeter and dif ferential rod worths were obtained from the measured .
reactivity worth versus the change'in rod group position. The ,
I differential rod worths of each 9toup were then summed to .i obtain the integral rod group s rths. }
- 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 f rom 100% to 0% withdrawn.
The integral reactivity worths for control rod groups 5 through 7 are presented in Figures 3.6-2 through 3.6-4.
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These curves' were obtained by integrating the measured differential worth curves. The integral worth of group.8 was not measured. The calculated worth of group 8 for 532'F, zero power is 'O.19% AK/K.
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 :was 12.0%.
- d. Conclusions Dif ferential. 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.
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FIGURE 3.6 ,
CYCLE 6 CDtGOL RCD GOUP IO7 CONS Fuel Transfer Canal )
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B 1 6 1 C 3 5 5 3 D 7 8 7 8 7 E 3 5 4 4 5 3 F 1 8 6 2 6 8 1 G 5 4 2 2 4 5 H W- 6 7 2 4 2 7 6 M
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1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 X Group Number
.. Grouc No. of Rods Function 1 8 Safety 2 8 Safety 3 8 Safety 4 9 Safety 5 12 Control 6 8 Control 7 8 Control 8 8 APSRs 18 -
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, 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. 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 i during the zero power test.
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- b. Test Method Measurements of the differential boron worth at 532*F was 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 dif ferential 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 9.918 pcm/ ppm 8 at an average boron concentration 1277.5 ppm 8. The predicted value was 9.308 pcm/ ppm 8 t 155. 1
- d. Conclusions The measured results for the soluble poison differential worth at 532*F was within 15% of the predicted differential worth.
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. ' 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 three major power plateaus, 28%, 75%, and.100% of 2535 megawatts core thermal power, as determined f rom primary and secondary heat balance measurements. Operation in the power range began on March 24, 1987.
Power escalations occurred as the required testing at each plateau was successfully completed with the exception of a 1 month hold at 81-85%
power due to OTSG 1evel limits.
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 I 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. 'l l
The major physics measurements performed during power escalation i consisted of determining the moderator and power Doppler coefficients of reactivity and obtaining detailed radial and axial core power distribution measurements for several core axial imbalances. Values of 1 minimum DNBR and maximum linear heat rate were monitored through'out the test program to ensure that core thermal limits would not be exceeded, l
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4.1 Nuclear Instrumentation Calibrotica at Power
- a. Purpose i
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 i 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 j 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:
Test Plateau Bistable Setpoint 5 FP , 5 FP 30 40 75 85 100 105.5
- 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 as expected since the power range nuclear instrumentation measures reactor neutron leakage *.thich is directly related to the above changes in system condition $. 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 12.5% axial offset criteria as determined by the incore monitoring system was also met.
The high flux trip bistable was adjusted to 40, 85 and 105.5%
FP prior to escalation of power to 30, 75 and 100% FP, respectively.
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.' 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, boren, and/or control rod configuration changes during the program..
Acceptance criteria for nuclear instrumentation calibration at power were met in all instances.
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, 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 at 30%FP.
- 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 power of 30, 75, and 100%FP.
Using the corrected SPND maps, calculations were performed to detenmine the detector current to average detector current values per assembly for each incore detector versus axial positions. Any detector levels which were determined to have failed were deleted from scan or substituted.
At 75% FP, SP-1301-5.3, Incore Neutron Detectors-Monthly Check, was performed to calibrate the back-up recorders to its incore depletion value.
- c. Conclusions Incore detector testing during power escalation demonstrated that all detectors except level 2 of string 20 were functioning as expected, that symmetrical detector readings agreed within acceptable limits, and that the computer applied correction f actors are accurate. The backup incore recorders were calibrated at 75% FP and operational above 80% FP as required by the Technical Specifications.
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, 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. indicated out-of-core power distribution and the actual incore
.. power distribution.
- 2. To demonstrate axial Xenon control using the Axial Power Shaping Rods (APSR's).
- 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 Inbalance."
- 4. To determine the core maximum linear heat rate and minimum DNBR at various power imbalances.
- b. Test Method This test was conducted at 75% 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 limits would not be exceeded by operating within the flux / delta flux / flow envelope set in the Reactor Protection System.
Specified plant data was recorded at the following target imbalances:
l
- 1. 2 to 3%
- 2. O to 1%
- 3. -14 to -16%
- 4. -10 to -14%
- 5. -5 to -10%
- 6. -1 to +1%
As CRG-8 was moved to establish the above imbalances, the integrated control system automatically compensated for reactivity changes by repositioning CRG-7 to maintain a constant power level.
- c. iestResults The relationship between the ICD and 0C0 offset was determined at 75% FP by performing an imbalance scan with the APSR's.
The average slope measured on the four out-of-core detectors was 1.385. The lowest slope was 1.312 for NI-7. The scaled '
difference amplifier gain was 5.035.
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' A comparison of the Incore detector (ICD) of fset' versus the cut-of-core (OCD) detector of fsct obtained for each N1 channel is shown in Table 4.3-1.
Core power distribution measurements were taken in conjunction with the most positive and most negative imbalances at 75% FP and the values of minimum DNBR and worst case MLHR and compared to the acceptance criteria.
.. The worst case values of minimum DH8R and maximum linear heat rate determined at 75% FP are listed in Table 4.3-2.
Th'e worst case DNBR' ratio was greater than the minimum limit of 1.3 and the maximum value of linear heat rate was less than the fuel melt limit of 20.5 kw/f t after extrapolation to 105.5 FP. These results show that Technical Specification limits have been met.
Backup offset calculations using AP 1203-7 agree with the computer calculated of fset. Table 4.3-3 lists the computer calculated offset as well as offsets obtained using the incore detector 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. A revised difference amplifier K factor of 3.684 will provide a slope greater than or equal to 0.96 when OCD offset is plotted versus ICD offset.
Minimum DNBR and Maximum Linear Heat Rate parameters were well within Technical Specifications limitations.
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______-a
TABLE 4.3-1 --
'INCORE'0FFSET-VS OUT-OF-CORE OFFSET INCORE-- OUT-OF-CORE OFFSET (%)
0FFSET (5) NI-5 NI-6 _
NI-7 NI-8 3.77' . 5.51 5.55 4.68 5.20
.1.07 2.25 2.28 1.86 2.10'
-1.76 -1.26 -1. 30 - -1.30 -1.23
-5.12 -5.80 -5.86 -5.40 -5. 4 5 .
-6.40 -7.46 -7.64 -6.97 -7.05
-10.65 -13.74 -14.09 -12.84 -12.92
-14.72 -20.47 -20.98 -19.08 -19.13
-19.90 -27.86 -28.47- .-25.80' -25.98
-21.39 -29.81 -30.45 -27.52 -27.79 l
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m.
c.- ,
TABLE 4.3-2 WORST CASE DN8R AND LHR IMBALANCE MINIMUM EXTRAPOLATED WORST CASE LHR EXTRAP. MAX. LHR (1) ONBR MDNBR (KW/FT) (KW/FT)
-15.93 4.70 3.14 10.06 13.67 2.67 4.84 3.15. 9.28 12.37 4
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l TABLE 4.3-3 FULL INCORE OFFSET VS BACKUP RECORDER OFFSET FULL INCORE BACKUP RECORDER OFFSET OFFSET
(%) (%)
-5.12 - -3.30 i
+3.77 +3.86 l
-21.39 -17.16 l l
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. 4.4 Core Power Distribution Verification
- a. Purpose To measure the core power distributions at 75 and 100 percent full power 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 perfomed at 75 and 100% full power, under steady state conditions, for specified control rod configurations. To provide the best comparison between measured and predicted results, three-dimensional equilibrium xenon conditions were established. Data collected for the measurements consisted of detailed power distribution information at 364 core locations from the incore detector system and the worst case core thermal conditions were calculated using this data. The measured data was compared with calculated results.
- c. Test Results A sunenary of the cases studied in this report is given in Table 4.4-1 which gives the core power level, control rod pattern, cycle'burnup, boron concentration, axtal imbalance, maximum quadrant tilt, minimum DNBR, msximum LHR and power peaking data for each measurement. The highest Worst Case MLHR was 11.17 at 100% FP which is well below the limit of 20.5 kw/ft. The lowest minimum DNBR value was 3.978 at 1005 FP which is well above the limit of 1.30.
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 75% and 100% full power. Comparison of measured and predicted results show good agreement. The maximum difference between a measured and predicted value was 1.6% for radial peaking at 755 FP This met the acceptable criteria of < 5.05.
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.
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. TABLE 4.4-1 CORE POWER DISTRIBUTION RESULTS POWER PLATEAU 75% 100%
DATE 03-29-87 05-05-87 Actual Power (%FP) 75.0 99.62 CRG 1-6 . (%WD) 100 100 CRG 7 (5WD) 93 90 CRG 8 (%WD) 31.5 27.5 Cycle Burnup (EFPD) 1.79 31.78 Boron Conc. (PPM) 1078 965 Imbalance (%) -2.42 -4.69 Maximum Tilt (%) 0.51 0.32 MDNBR 5.33 3.978 Worst Case MLHR (KW/FT) 8.65 11.17 Maximum Radial Peak Measured 1.281 1.270 Predicted 1.260 1.250 Difference (5) 1.639 1.575 Acceptance Criteria (%) 15% $55 Maximum Total Peak l
Measured 1.514 1.512 Predicted 1.510 1.500 Difference (%) 0.264 0.794 Acceptance Criteria (%) $7.5% $7.5%
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a
,I l 4. 5.- Reactivity Coefficients at Power .
i
-a.- . Purpose The'purpo'se of this test is to measure the temperature'and power doppler coef ficients of. reactivity at power. This information is then used to assure that Tech. Spec. 3.1.7.1, which states that the moderator temperature coefficient shall
'not be positive at power levels above 95% of rated power, is
. satisfied.
- b. Test Method For measuring the temperature coefficient of reactivity, the average RC temperature was decreased and then increased by about 5 Degrees F. The reactivity associated with each temperature change was obtained from the change in controlling 7 rod group. position, and the values for the coefficient were calculated. I For measuring the power doppler coefficient of reactivity, reactor power was decreased and then increased by about 5 percent FP. The reactivity change was obtained from the change in controlling rod group position, and'_the values for the coefficient were calculated.
1 In conjunction with both reactivity coefficient measurements,.
differential controlling rod group worth measurements using
'the fast insert / withdrawal method were perfonned,
- c. Test Results At 1005 FP, temperature and power doppler coefficient !
measurements were performed. The moderator temperature '
coefficient measured at 1005 FP was -5.204 pcm/*F. This verifies that the moderator temperature coefficient is negative above 955 FP.
i The measured power doppler coefficient at 1005 FP was -8.413 pcm/5FP and the measured fuel doppler coefficient was -1.219 pcm/0F. This meets the acceptance criteria of being more negative than -0.9 pcm/0F.
- d. Conclusions !
The measured moderator temperature coefficient (MTC) results indicate that the MTC will be negative above 955 F.P.
The measured fuel doppler coefficient (FDC) results meet the requirement that the FDC be more negative than -0.9 pcm/0F.
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e -.
t GPU Nuclear Corporation Nuclear :=,o=gr8o Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct 0121 Number:
August 21, 1987 52T1 2159 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
]
Gentlemen: J 1
Three Mile Island Nuclear Station, Unit 1 (TMI-1) I Operating License No. DPR-50 f Docket No. 50-289 l Cycle 6 Startup Report l l
Enclosed is the Startup Report for TMI-1 Cycle 6 operation. The testing l addressed by this report was completed and approved as of May 22, 1987. This report satisfies Technical Specification 6.9.1. A. No NRC response to this letter is necessary or requested. q J
~
This is being submitted two days late as discussed with R. Conte, Senior I Resident Inspector.
l Sincerely, H. D. H 11 Vice President & Director, TMI-1 HDH/CWS/spb:0963A
)I cc: R. Conte j W. Russell l G. Edison Enclosure i
j l
l GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation u