ML20198T472

TMI-1 Cycle 12 Startup Rept

ML20198T472
Person / Time
Site:	Three Mile Island
Issue date:	12/31/1997
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
6L20-98-20019, NUDOCS 9801270065
Download: ML20198T472 (31)

Text

. _ .

e GPU Nucisat,Inc.

l A Route 441 south Post Othee Box 42 NUCLEAR M * *toa. FAIN 5M4M 6L20 98 20019 te:7i7 s44 7pi January 19, 1998 U.S. Nuclear Regulatory Commission Attn: . Document Control Desk Washington, DC 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI 1)

Operating License No. DPR-50 Docket No. 50 289 Cycle 12 Startup Report Enclosed is the Startup Report for TMI l Cycle 12 operation. Initial criticality for Cycle 12 was achieved at 6:25 p.m. on October 18,1997. Testing addressed by this report was completed on i November 4,1997. In all cases the applicable test and Technical Specifications (TS) limits were met. This repon is being submitted in accordance with TMI I TS 6.9.1.A. No NRC response to this letter is nec, saary or requested.

Sincerely, Y ,

J. Langenbach Vice President and Director, TM1 MRK -,,,

g g,uv~s Enclosure

- cc: Region 1 Adminhtrator fh TMI l Senior Project Manager f.

TMI l Senior Resident inspector l

$OJijilJOJWIRl[Il I

'9901270065 971231 9 _ _ _

DR ADOCK O

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\GUNUCLEAR Th11-1 CYCLE 12 STARTUP REPORT TS11 SIIIFT ENGINEERING DECEh1HER 1997 i {

1 *

\ .

TAllLE OF CONTENTS 1%GE 1.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER -

SUMMARY

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

SUMMARY

.... .. . .. . 3 3.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER , . .. .... .. . 5 3.1 Initial Criticality . . .. .. . . . ..... .. .... . .. .. .. .. ..... .... .. . .... 5 3.2 Nuclear Instrumentation Overlap . .. ... . .. ... .. . . . . . . 9 3.3 Reactimet er Checkout . . . .. .. .. .. . . . . . . ...... . ... .. . ... ... ...... 9 3.4 ARO Critical Boron Concentration . .. .. ..... . . . .. . .. . .. .. . ....... 10 3.5 Temperature Coefficient Measurements . . . .. ... . . ... ... .. . .. . .. . . .... .. ... I1 3.6 Control Rod Group Wonh Measurements . .. .. . ... . . .. .. . . . . . . . . 12 3.7 Differentiellloron Wonh .. .. ... . . . . .. . . . ... ... 17 4.0 CORE PERFORMANCE - MEASUREMENTS AT POWER .. . . . . .... .. .. 19 4.1 Nuclear Instrumentation Calibration at Power . . . . . . . .. ... , 19 4.2 Incore Detector Testing ...... ... . .. .. .. . ... ...... . . ... . 21 43 Power 1mbalance Detector Correlation Test . . . , . ... . . . .. ...... . .. 22 4'.4 Core Power Distribution Verification . .. . . . .. . . . . . . . . . . . . . . . . . 24.

-1

1.0 CORE PERFORM ANCE - MEASliREMENTS AT ZERO POWER - S11MM ARY Core performance measurements were conducted during the Zero Power Test Program which began on October 18,1997 id ended on October 19,1997. 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 ze:o power physics test results appears as Table 1 1.

a. IniliaLCriticality initial criticality was achieved at 1825 on October 18,1997. Reactor conditions were 532*F ar.d 2155 psig. Control rod groups I through 6 were withdrawn to 100%; group 7 was pes sitioned at 86.4% withdrawn; group 8 was positioned at 30.5% withdrawn.

Criticality was achieved by deborating the Reactor Coolant from 2532 ppm to 2152 ppm and repositioning group 7 to 76% withdrawn. Initial criticality was achieved in an orderly manner and within the acceptance criteria of 2146 i 50 l pM.

b. NucleatimtrumentationDycrJan At least one decade overlap was measured between the source and intermediate range detectors as required by Technical Specifications.

c. llencliluctcLGhikelit An on line functional check of the reactimeter using the average of N13 and 4 was performed afler initial criticality. Reactivity calculated by the reactimeter was within 5% of the core reactivity determined from doubling and halving time measurements.

d. A1LRods Out CriticaW9 ton Concentration The measured all rods out critical boron concentration of 2165.9 ppmB was within the acceptance criteria of 2195

• 50 ppmB.

c. Isnipnatuitfacflifienthica!!vtrJnents The measured temperature coellicient of reactivity at 532 F, zero power was within the acceptance criteria limit.

f. Control Rodlitpup Worth McMurements 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 3.3% which was for CRG-7 woith. '

g. Di0'stsntiamoron Worth The measured differential boron worth at 532*F was 3.75% more than the predicted value.

This is within the bounds of the FSAR and GPUN supplied limits of *15%.

1

i

. i TABLEl1

SUMMARY

OF ZERO POWFR PIIYSICS TEST RESULTS  !

CYCL.E 12 i

Acc:ptance Measured .

f Paramcist _Crilsria_ Value Dryialina Critical Boron 2146 * $0 ppm 2152 ppm 6 ppm N1 Overlap >ldecade >t,4 decade ---

l 4

Sensible lleat N/A 1 x 10 amps ---

All Rods Out Boron l Concentration 2195

• 30 ppm 2165.9 ppm 29.1 ppm l Temperature Cocmcient -0.14 pcm/*F -1.43 pcm/cF l.29 pcm/*F (2154 ppm)

• 2 pcm/*F i

Moderator Coemeient <9.0 pcm/*F 0.19 pcm/*F ---

Integral Rod Worths (532*F) GPS 7 2862 pcm

• 10% 2914.64 pcm 1.8%

Group 7 881 pcm

• 15% 909.84 pcm 3.3%

Group 6 754 pcm

• 15% 741.85 pcm 1,6%

Group 5 1227 pcm

• 15% 1262.95 pcm 2.9%

. Diff Boron Worth 6.51 pcm/ ppm

• 15% 6.754 pcm/ ppm 3.75 %

(2070 ppm) '

i Y

6 2- ._

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. 1 2,0 CORE PERFORMANCE - MEASIIREMENTS AT POWER - SIIMM ARY ,

l Thi,s section summarizes the physics tests conducted with the reactor at power. Testing was perfctmed at power plateaus of approximately 11,40,66, 79, and 100% core thermal power.

Operation in the power range began on October 19,1997. j l

Gadolinia is again present in the TMI-l core as an integral burnable poison. Twenty eight (28) j assemblies containing gadolinia were reloaded from Cycle 11. Sixty eight (68) assemblies containing gadolinia were loaded fresh for Cycle 12. These assemblies require no special ,

monitoring. i Two types ofle d test assemblics were reloaded from Cycle 11 for their second burn.

Four of the LTAs were manufactured by Westinghouse, thus rg resent a departure from the traditional TMI.1 fuel vendor. Two LTAs were manufactured by the B&W Fuct Company, now Framatone Cogema Fuels (FCF), but include a total of sixteen pins manufactured with cladding that uses zirconium alloys M4 and MS. Both types of LTA were monitored during power escalation testing to ensure that they were not the limiting (hottest) assemblies in the core with respect to radi i power distribution power peaking.

a. NuckatJmimmcEtation Calibration at Poxer The power range channels were calibrated as required during the stanup 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 con 6guration changes during testing.

b. Incorgjkitq1011citirw

! Tests conducted on the incore detector system demonstrated that all detectors were fimetioning acceptably. Symmetrical detector readings agreed within acceptable limits except that String 35 at core location F.i was consistently low. Its behavior can be explained by the negative power tilt in its quadrant, which other detectors in that quadrant confirmed. The plant computer applied background, length and depletion correction factors. The backup incore recorders were operational above 80% FP as required by Technical Speci6 cations.

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c. L'anctJmhalance.Drtector Correlation Tes 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 of

>0.96 is obtained by using a gain factor of 3.567 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.

d. Care PowerJ1111tihutioLYeridcatica Core power distribution measurements were conducted at approximately 66% full power under non equilibrium xenon conditions and at 100% full power at equilibrium xenor, conditions. The maximum measured and maximum predicted radial and total peaking factors are allin good agreement. The largest positive percent diflerence between measured and predicted values was $.53% for radial peaking at approximately 66% FP. This met its unique acceptance criteria 006.6%. All other assemblies were 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 litrits.

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, reactinaeter checkout, all rods out critical boron concentration, temperature coefficient measurement, control rod worths, and differential boron wonh.

3.I initinLQitiality initial criticality for Cycle 12 was achieved at 1825 on October 18,1997. Reactor conditions were $32*F and 2155 psig. Control rod groups 1 through 4 were withdrawn during the heatup to 532*F. The initial reactor coolant system (RCS) baron concentration was 2532 ppm.

The approach to criticality began by withdrawing control rod group 8 to 30.5% withdrawr; control rod groups 5 and 6 to 100% withdrawn, and positioning group 7 at 86.4%

withdrawn. Criticality was subsequently achieved by deborating the reactor coolant system to a boron concentration of 2152 ppm and repositioning group 7 to 76% withdrawn Throughout the approach to criticality, plots ofinverse multiplication were maintained by two independent persons. Count rates were obtained from each source range neutron detector channel. One person used NI-ll A and the other used NI-12A. Plots ofinverse count rate (ICR) versus control rod position were maintained during control rod withdrawal. Plots ofICR versus RCS boron concentration and plots oflCR versus gallons of demineralized water added were maintained du-ing the dilution sequence.

The inverse count rate plots maintained during the approach to criticality are presentei 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 ofICR 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 2146 50 PPM.

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nuclear instrumentation (NI) range, at least one decade of overlap between the source range N1's and the intermediate range N1's must be observed.

b. Is3t Method To satisfy the above overlap requirements, cos e power was increased until the intermediate range channels came on scale. Detector signal response was then recorded for both the source range and intermediate twge channels. This was repeated until the maximum source range value was < ched. ,

c. 1c11E18:115 The results of the initial N1 overlap data at 532*F and 2155 psig have shown a >l.4 decade overlap between the source and intermediate ranges.

d. . Conclusions The linearity, overlap and absolute ( atput of the intermediate and source range detectors are within specifiutions 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 Bractimeter Chesinut

a. PREDJs Reactivity calculations during the Cycle 12 test program were performed using the reactimeter. After initial criticality and prior to the first physics measurement, an online ft.nctional check of the reactimeter was performed to verify its accuracy for use in the test program.

b. IcstAicthed Afler 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.

RMAS soflware compared the reactivity calculated from the doubling times to the values calculated by the reactimeter. Measurements were taken at approximately

+64, 80, and 483 pcm, 9

c. Issdesults The measured values were determined to be satisfactory and showed that the  !

reactimeter was ready for startup testing.

d. Conclusicas 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 AlLRods out criticaLiloton concentralien

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

b. Inthic1had Starting from the critical condition, the Group 7 control rods w.:re 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 Resul13 The measured boron concentration with group 7 positioned at 100% WD was 2165.9 ppm.

d. Conclusions The above results show that the measuied boron concentration of 2165.9 ppm is -

within the acceptance criteria of 2195 * $0 ppm.

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' 3.5 lenipetAlutc_CQC.IflCifIllMCAEutf1HC1115

a. Eurpq$c The moderator temperature coemeient 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 coemeient shall not be positive while greater than 95% FP. The moderator temperature coemeient cannot be measured directly, but it can be derived from the isothermal temperature coeflicient and a known fuel temperature (Doppler) coeflicient.

b. IcsLMelhnd 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* amps on the intermediate range. Equilibrium boron concentration wa: 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 reactivity value and the RCS average temperature was disphyed on the RhiAS monitor.

Once steady state conditions were established, a heatup rate was started by closing the turbine bypass valve Aller the core average temperature increased by about 5*F core temperature au Jux were stabilized and the process was reversed by decreasing the core average temperature by about 10oF. Afler core temperature and flux were stabilized, core temperature was returned to its initial value. Caiculation of the temperature coeflicient 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 'coeflicient measurements are provided below. The predicted values are included for comparison.

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

RCS hiEASURED PREDICTED hiEASURED REQUIRED BORON ITC ITC hiTC h1TC (EEM) (PChi/DEG F) (PCht/DEG F) (PChi/DEG Fj ( PChi/DlQf',

2154 -1.43 -0.14 +0.19 <+9.0

. d. CptcJm!Ons The measured values of the temperature coemcient of reactivity at $32'F, zero '

, reactor power are within the acceptance criteria of 2.0 pcm/*F of the predicted value. An extrapolation of the moderator coemeient to 100%FP indicated that it was well within the limits of Technical Specifr.ations 3.1.7.2.

3.6 Conitodedfttopp Wonh Mcasurements

a. hl[PDIC 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 3.61 will be used throughout Cycle 12. Calcultted and measured control rod group reactivity worths for the normal withdrawal sequence were determined at reactor conditions of zero power,532*F and 21$5 psi. The measured results were obtained using results of reactivity and group position from the RMAS system.

b. LitMelbad Control rod group reactivity wonh measurements were performed at zero power, 532*F using the boron / rod swap method. Both the differential and integrni reactivity wonhs 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 insening 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 wonh 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. hsdtsults 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% withdrawn.

The integral reactivity wonhs for control rod groups 5 through 7 sre presemed in Figures 3.6-2 through 3.6-4.

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

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Tab!c 3.61 provides a comparison between the predicted and measured results for the ed werth measurements. The results show good agreement between the meased und predicted rod group worths. The maximum deviation between measud and predicted worths for a group was +3.3%.

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.

3.7 RMerentialDoron Worth

a. Ihttppig 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 wonh of the boric acid was measured during the zero power test.

b. Ic1LMelbstd A

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c. Isfdsstits Measurements of the soluble boron difTerential worth were completed at the zero power condition of 532*F. The measured boron worth was 6.754 pcm/ppmB at an average boron concentration of 1939.5 ppmB. This corresponds to a 3.75%

deviation, which is within the predicted value of 6.51 pcm/ppmB i 15%.

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

• TABLE 3.61 I COMP ARISON OF PREDICTED VS MEASURED ROD WORTilS i

MEASURED PREDICTED PERCENT  !

CRG. WORTil WORTil DIFFERENCE N(L _. (PCM) _1PCM)  %

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40 CORE PERFORMANCE- MEASIIRD1ENTS AT POWER This section presents the results of $c physics rneasurements that were conducted with the reactor at power. Testing was conducted at power plateaus of approximately 11%,40%,66%,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 October 19,1997.  ;

Periodic measurements and calibrations were perfo med on the plant nuclear instmmentation 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 instmmentation 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 instmmentation 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 monito cd throughout the test program to ensure that core thermal limits would not be exceeded.

4.1 Nucleatimtatumulatio1LCalihtatiplaLPowcr

a. Ihtrnosc 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 ofTset as determined by the incore moni toring system.

b. Icf1Mc1hnd 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 phyrics testing, the high flux trip bistable setpoint was adjusted. The major settings during power escalation are given

. below:

Nominal Nominal Test Plateau Bistable Setpoint

% FP  % FP 40 50 80 90 100 105.1

c. Icst Retalts An analysis of test results indicated tha, .hanges in Reactor Coolant System boron and xenon buildup or burnout afTected the power as observed by the nuclear instrumentation. This was expected since the power range nuclear instrumentation measures scactor neutron leakage which is directly related to the above chareges in system conditions. Each time that it was necessar/ to calibrate the power rar,te nuclear instrumentation, the acceptance criteria of calibration to be no less thaa 2.0% FP of the heat balance power was met without any difliculty. Also, each ime 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 wher required. ,

The high flux trip bistable was a6jasted 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 roo configuration changes during the program.

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

• 4.2 Jrt 01g_Qritetor Testirig

a. IMrpg3g 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 syn' metry of the neutron Dux.

b. Test Medted 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 approximately 11. 40,66,79 and 100%FP.

Using the corrected SPND maps, calculations were performed to determine the detector current to average detector current values per assembly for each incore detector versus axial positions.

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

c. Cattclusintts 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 as required by the Technical Speci0 cations.

b

' 4.3 P_QECLimhal401c_Dsitslor Correlation Test

- a. hi.rpose The Power Imbalance Detector Correlation Test has four objectives:

1. To determine the relationship between the core power distribution as measured by the otd of-core detectors and the incore instruments.

2. To demonstrate axial power shaping control using the Axial Power Shaping Rods (APSR's). ,,

3. To verif the f adequacy and accuracy of backup imbalance calculations as done in Al' .203-7, " 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. IntAhthed 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 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 was deborated to obtain more negative imbalance data. Again, the ICS compensated for the boron cha.. f inserting CRG-7 to maintain constant power.

c. Test Resitlts The relationship between the ICD and OCD offset was determined at about 79% FP by changing axial imbalance with tne APSRs. The average slope measured on the four out-of-core detectors was 1.005. The lowest slope was 0.946 for NI-7. The scaled ditTerence amplifier gain was changed to 3.567 A comparison of the incore detector (ICD) offset versus the out-of-core (OCD) detector offset obtained for each N1 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 compared to the acceptance criteria.

TABLE 4.3-1 INCORE OFFSET VS. OUT-OF-CORE OFFSET INCORE OUT-CF-CORE OFFSET (%)

OFFSET

% NI .1 N.l. 6 N1-7 Nl-H 7.56 7 32 6.98 6.33 6.65 4.12 4.05 3.84 3.50 3.66

-3.16 -3.25 -3.35 -3.08 -3.08

-6. I 5 -6.36 -6.48 -5.98 -5.95

-12.03 -12.86 -12.91 -11.83 -12.00

-20.01 -21.61 -21.61 -19.54 -20.23

-24.41 -25.99 -26.04 -23.59 -24.26 i

- _ - - _ . _ - - - - . . - - - . - - - - - - . - . - - - - __-. ---__s

\ .

The worst case values of minimum DNBR and maximum linear heat rate -

. determined at 79% FP are listed in Table 4.3-2.

The worst case DN'3R ratio was greater than the minimum limit and the maxin.um value oflinear heat rate was less than the fuel melt limit of 20.5 kw/fl afler cxtrapolation to 105.1 FP. These results show that Technical Specification limits have been met.

Backup offset calculations using AP 1203-7 agree with the computer calculated offset. 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 accoroance with AP 1203-7 provide an acceptable alternate method to computer cakulated values ofimbalance. A difference amplifier K factor of 3.567 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.

4.4 Core Power Distdbytion Verifration

a. barposg To measure the core power distributions during the power escalation and at 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 measuremen:s were performed at approximately 66%FP during the power escalation and at 100% full power 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 locat;ons from the incore detector system. The worst case core thermal conditions were calculated using this data. The measured data was compared with calculated predictions. One location (N-13) initially failed its acceptance criterion. A new limit specifically for this location was calculated and justified with a safety evaluation.

1

TABLE 4.3 2 WORST CASE DNBR AND LHR IMBALANCE OFFSET MINIMUM EXTRAPOLATE WORST CASE LHR EXTRAP. MAX. LHR

&  % DNBR MDNBR KW/FT KW/FT 6.00 7.56 3.48 2.40 10.80 13.36

-19.38 -24.41 3.38 2.43 12.04 15.27

(

l 1

TABLE 4.3-3 FULL INCORE OFFSET VS BACKUP RECORDER OFFSET FULL INCORE BACKUP RECORDER OFFSET OFFSET (m (N _

7.56 6.91 4.12 4.557

-3.16 -1.19

-6.15 -4.221

-12.03 -9.54

-20.01 -15.7

-24.41 -20.38

. c. Init.Enuhs Th- acceptance criteria for power distribution require that hil new fuel be within limits for radial and total peaking. Also, the RMS of the differences between meawed and predicted 11FP 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 which gives the core power level, control rod pattern, cycle burnup, boron concentration, axial imbalance, maximum quadrant tilt, minimum DNBR, maximum LliR 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. Data for the three dilTerent radial peaks and the two total peak limits are shown. The highest Worst Case MLHR was 12.65 kw/fl at 100% FP which is well below the limit of 20.5 kw/ft The lowest minimum DNBR value was 2.912 at 100% FP which is well above the 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 distiibution.

d. Conclusions Core power distribution measurements were conducted at approximately 66% and 100% full power. Comparison of measured and predicted results show good agreement. The largest difference between the maximum measured and maximum predicted peak value was 5.53% for radial peaking at approximately 66% FP for location N-13. This met its acceptance criterion of <6.6%. All other fuellocations met their lower 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.

, TABI E 4.41 CORE POWER DISTRIBIRION RESULTS POWER PLATEAU Escalation 100%FP DATE 10-20-97 10-24-97 Actual Power (%FP) 65.42 99.96 CROl6 (%WD) 100 100 CRG7 (%WD) 91.3 91.4 CRG8 (%WD) 30.4 30.1 Cyc!c Burnup (EFPD) 0.41 3.74 Boron Conc. (PPM) 1935 1582 Imbalance (%) 1.94 -2.71 Maximum Tilt (%) 1.60 1.91 MDNBR 4.51 2.91 Worst Case MLHR (KW/FT) 8.57 12.65 Maximum Radial Peak Difference, New Fuel Location N 13 N-13 Measured Peak 1.230 1.211 Predicted Peak 1.165 1.157 DifTerence (%) 5.53 4.667 Acceptance Criterion (%) 36.6 % $6.6%

Location L-14 L-14 Measured Peak 1.134 1.136 Predicted Peak 1.106 1.115 Difference (%) 2.53 1.883 Acceptance Criterion (%) 55.8 55.8 Location K-14 K-14 Measured Peak 1.146 1.154 Predicted Peak 1.122 1.141 Dif':rence (%) 2.09 1.139 Acceptance Criterion (%) 53.8 53.8 Maximum Total Peak DifTerence, New Fuel Location L-14 L-14 Measured Peak 1.323 1.302 Predicted Peak 1.294 1.289 DifTerence (%) 2.28 1.01 Acceptance Criterion (%) 57.5 57.5 Location N- 13 N-13 Wasured Peak 1.416 1.390 Predicted Peak 1.371 1.347 Difference (%) 3.25 3.I9 Acceptance Criterion (%) 55.5 55.5 Eighth-Core RMS of Absolute DitTerences for Radial Peaks, All Fuel Measured 0.036 0.032 Acceptance Criterion 0.05 0.05 i

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