ML20058H061

Core Xvi Startup Physics Test Rept

ML20058H061
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/31/1990
From:	Claffey S, Stanford J, Stetz J CONNECTICUT YANKEE ATOMIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9011140220
Download: ML20058H061 (31)

Text

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4 ONNECTICUT VANMER ATOMIC POWER CuMPANY HADDAM NEr,s 'LANT

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• DOX 127E

• EAST HAMPTON.C1.2 e4 9341 m_

October 19, 1990 Re:

Technical Specification 6.9.1 Docket No. 50 213 U. S. Nuclear. Regulato.

Commission Document Control DetJ Washington, D. C.

20 " 3 L

Dear Sir:

Iladdam Neck Plant Cygle 16 Startup Physics T2st ReppIl in accordt.nce with Section 6.9.1 of the lladdam Neck Plant Technical Specifications, r

Com;wucut Yankee Atomic Power Company (CYAPCO) hereby submits the Startup Physics Test Report for Cycle 16 op: ration for the lladdam Neck Plant.

This report is being submitted within 90 days following completion of the startu,) test program.

Should you have any questions related to this submittal, please contact me.

Very truly yours, fk John P. Stetz '

Station Director J PS/Jhb ec:

(1)

Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 (2)

John T, Shediosky E

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Resident inspector Connecticut Yankee 9011140220 901031 POR ADOCK 05000213 F

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CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT i

CORE XVI STARTUP PHYSICS TEST REPORT t

OCTOBER 1990 l

Prepared by:

uf 7 I'dt //n/

S. F. Claff6f, R6 actor Engineer Approved by:

AS4McA -

'T.' Stanford, Engineering Supervisor Reviewed by:

>W-A - M M6-C. J. Gladding, Engineering Manager i

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t TABLE OF CONTENTS Pagt 1

I n t rod uc ti on.................................................................................. 1 2n trol R od Drop Time M easuremen ts.................................................... 2 New Core Initial Approach to Criticallty................................................... 5 s

4.

A11 Rods Out Criticai Boron Concentration................................................ 6 1

5.

Control Rod Coupling Verification......................................................... 8 6.

Isothermal Te mperatu re Coe fficie n t M casuremen ts...................................... 9 7.

Control Rod Bank Reactivity Worth Measurements..................................... I 1 8.

Rodded Critbl Boron Concen tra tion..................................................... 14 9.

Differen tial Boron Worta................................................................... 15

10. Thir ty Percen t Power flu x M a p........................................................... 17 l

1

11. Elghty Pcrcen P o w c r FI u x M a p s.......................................................... 19

12. One iIu ndted Perce n t Power F1u x Map................................................... 23

13. R eaet or Coola n t S y s t e m Flow Te st.............................................,....... 25 l

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LIST OF TAHLES Inble EA&c 1.

Control Rod Drop Time Measuremen ts.................................................... 4 2.

Delayed Neutron Fractions at 0 EFPD, ARO, I IZP...................................... 7 3.

Co n t rol R od B a n k Wo rt h s.................................................................. 13 4.

D 1 f fe re n t i a 1 B o ro n Wo rt h.................................................................... I 6 5.

Summary of Results from 30% Power Flux Map CY-XVI 1 459...................18 6.

Summary of Results from 80% Power Flux Map CY XVI 2 460.................... 20 7.

Summar', of Results from 80% Power Flux Map CY XVI 3 461.................... 21 8.

Summary of Results from 80% Power Flux Map CY XVI 4-462..................... 22 9.

Summary of Results from 100% Power Flux Map CY XVI 14 472.................. 24

10. Reactor Coolant System Flow Test Results..............

..........................27

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1 1.

INTRODUCTION

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This report documents the Connecticut Yankee Core XVI Startup Physics Test program. The

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testing sequence was completed as follows:

Initialcriticality August 12,1990 Zero power testing completed August 14,1990 Turbine phased to grid August 15,1990 30 % power flux map completed August 31,1990 80% power flux maps completed September 13,1990 100% power aux map completed September 30,1990 RCS flow test completed October 4,1990 The Cycle 16 core loadingis as follows: a Batch 18A fresh feed of 52 stainless steel clad fuel assemblics (4.0 w/o) loaded on the periphery of the core, a Batch 18B fresh feed of four stainless steel clad fuel assemblics (3.0 w/o) loaded in the core interior, one twice burned Batch 15C stainless steel clad fuel assembly loaded in the center of the core, a mixture of 48 Batch 17 and 48 Batch 16 stainless steel clad fuel assemblics loaded in the core interior, and four thrice burned Batch 14B stainless steel clad fuel assemblies loaded on the core periphery.

All startup physics test accepumee criteria were met. All requirements of the Technical i

Specifications were fulfilled.

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2.

CONTROL ROD DROP TIME MEASUP.EMENTS Objective The purpose of tha. contml rod drop time measurements is to measure at operating

-7 temperature and pressure the time for each of the 45 contml rods to travel fmm a fully l

withdrawn position to a fully inserted position.

Ikense Reauirements Technical Specification 4.1.3.4 requires that the dmp time for each control rod be determined each refueling.

Pmcedurr The rod position detector primary coil inputs for an entire bank are connected to a Hewlett Packard Multipmgrammer computer system. The bank is dropped using the main control board manual scram button. Pushing the scram button also triggers the computer system to initiate data collection. The coil output voltage signals, which are a function of control rod velocity, are sampled by the computer system at one millisecond intervals for 3.5 seconds. These data are then analyzed by computer code which detennines the time elapsed until the contml rod strikes bottom. Two graphs depicting the drop are also produced from the voltage signal for each rod.

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A backup method is available which uses a high spM recording oscillograph. The backup method is necessary for two reasons: 1) in the event that the computer system is unavailable for any reason, and 2) to retest any questionable rods. The backup method was not required.

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Results 1

l All 45 control rods traveled fmm a fully withdrawn position to a fully inserted position in 2.187 seconds or less with 4 RCPs operating, satisfying the Technical Specification requirement of 2.5 seconds. The measurements were performed on July 30 and 31, 1990. The nominal RCS pressure was 2020 psig and the RCS temperature was at l

Page 2 of 27

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d least 525'F. The maximum drop time was 2.187 seconds (rod #26 of bank A) and the

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minimum drop time was 2.035 seconds (rod #9 of bank D).

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The average drop time was 2.118 seconds; the standard deviation was 0.034 seconds.

Data are presented in Table 1.

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TAHLE1 Control Rod Drop Time Measurements l

Rod No.

Ilank Core location Dron Time (sec) 1 A

H8 2.079 2

A K8 2.120 3

A 116 2.073 4

A F8 2.104 5

A 111 0 2.080 6

D K6 2.109 7

D F6 2.092 8

D F10 2.113 9

D K10 2.035 10 B

M8 2.171 11 B

114 2.162 12 B

D8 2.130 13 B

111 2 2.16!

14 C

M6 2.138 15 C

K4 2.082 16 C

F4 2.112 17 C

D6 2.129 18 C

D10 2.064 19 C

F12 2.071 20 C

K12 2.105 21 C

M10 2.142 22 A

N7 2.161 23 A

J3 2,162 24 A

G3 2.142 25 A

C7 2.150 26 A

C9 2.187 27 A

G13 2.134 28 A

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't.129 29 A

N9 2.163

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l 30 B

M4 2.124 1

31 B

D4-2.118 32 Il D12 2.114 33 B

M12 2.093 34 D

NS 2.138 l

35 D

L3 2.081 36 D

E3 2.077 1

37 D

C5 2.136 l

38 D

Cll 2.070 39 D

E13 2.075 l

40 D

L13 2.101 4i D

Nll 2.134 42 A

P8 2.156 43 A

H2 2.144 44 A

B8 2.163 45 A

1114 2.099 Page 4 of 27

j 3.

NEW CORE INITIAL APPROACII TO CRITICALITY Objective The objecthc of the new core initial critical approach is to provide a safe and ef0cient meitns for achieving the initial criticality.

License Recuirements None Procedure l

At hot zero power conditions, the reactor coolant system (RCS) boron concentration is first reduced from the refueling boron concentration to approximately 450 ppm above the predicted C/D/A @320 and B @200 critical boron concentration. 1/M plots are maintained throughout the approach to criticality. After the initial dilution, control rod banks C, D and A are fully withdrawn and bank B is withdrawn to 200 steps. The final approach to criticality is ' hen made by additional RCS dilution and shimming of Bank B after the dilution has been tenninated.

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Results l

Core XVI initial boron concentation at hot conditions was approximately 2190 ppm. This j

concentration was reduced to 2001 ppm by adding demineralized water to the reactor coolant l_

system. Control rod banks C, D, and A were then withdrawn to 320 steps and bank B l

withdrawn to 200 steps. Criticality was achieved at 08:53 en August 12,1990 by adding l

approximately 11,210 gallons of demineralized water. The critical conditions were 535.6*F, bank B at 213 steps, and 1564 ppm boron. The conected critical boron concentration with bank B at 200 steps and 535'F is 1559 ppm boron. The predicted boron concentration with L

bank B at 200 steps is 1546 ppm. The difference of 13 ppm between the corrected test data and the predicted critical boron concentration is well within the acceptance criteria of 100 ppm.

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4 4.

ALL RODS OUT CRITICAL HORON CONCENTRATION Ob'metive The objective is to measure the all rods out critical bomn concentration at hot zero power conditions.

License Requirements Technical Specifications 4.1.1.1 and 4.1.1.4 require verification of adequate shutdown margin prior to exceeding five percent power following a refueling. This test provides information used to verify core design and thus, adequate shutdown margin.

Procedure Bank B controls rods are borated to 290 steps. The remaining control rod banks are all fully withdrawn. Critical bomn concentration is measured. De predicted worth of bank B from 290 to 320 steps then is used to correct the critical bomn concentration with bank B at 290 steps to the all rods out condition.

Results The all rods out critical baron concentration was 1599 ppm at 535'F. The predicted hot zero power boron concentration was 1584 ppm. The difference of S ppm between the predicted and measured critical bomn concentrations is well within the 100 ppm acceptance criteria.

Reactivity Comouter l

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Nuclear Instrumentation System power range channel 1 upper and lower fission chambers were used as inputs to the reactivity computer together with the hot zero power delayed neutron fractions for all zero power tests. Reactor coolant temperature and pressurizer level were also input to the reactivity computer. He reactivity computer was calibrated against several stable reactor periods varying from 50 seconds to 400 seconds. The all rods out, hot zem power beta fractions are listed in Table 2.

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i TABLE 2 Delayed Neutron Fractions at 0 EFFD, ARO,llZP Gmg Beta. eff i mmbda. Sec 1

2.060E-4 1.280E 2 2

1.287E-3 3.150E 2 3

1.165E-3 1.208E-1 4

2.501 E-3 3.217E 1 5

9.150E-4 1.402E+0 6

2.230E-4 3.875E40 Beta (Total) = 6.2978 3 Relative importance (I) = 0.970 lleta Effective = 6.108E 3 e

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CONTROL ROD COUPLING VERIFICATION

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i Obiective The objective of this test is to verify that each control rod assembly is connected to its respective drive shaft.

License Requirenents None Procedures After the All Rods Out Just Critical Boron Concentration procedure is completed, each control rod is inserted into the core until at least a five pcm change is observed on the reactivity computer. His is accomplished by disconnec'!ng the lift coils for all rods except the test rod in the test bank and driving the rod into the core. The test rod is then returned to its initial position. All control rods are tested in this manner, Results All control rods exhibited at least a five pcm reactivity change upon insertion into the core.

The acceptance criteria was met. This is not a Technical Specification surveillance, l

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6.

ISOTilERMAL TEMPERATURE COEFFICIENT MEASUREMENTS Objective The objective of these measuremer's is (1) to determine the isothemial temperature coefficient (ITC) for the new core at hot zero power conditions at two contml rod con 0gurations and (2) correct the ITC data at AROSIZP to ARO41ZP/BOL, ARO/lIFP/BOL and AROMIFP/EOL moderator temperature coefficients (MTC) for verificat.)n of the Technical Specifications i

limits.

License Reauirements Technical Specifications, Section 4.1.1.5," Moderator Temperature Coefficient," requires that I

the temperature coefficient be detemiined for a new core.

i Procedure Hot zem powerjust critical conditions are established. The reactor coolant temperature is reduced and then increased by approximately 5'F using the atmospheric steam dump. The reactivity computer calculates the reactivity change due to the change in temperature and displays reactivity as a function of temperature during the cooldown and heatup on an X Y plotter. Temperature coefficients are obtained at the ARO and a rodded control rod bank configuration.

l Ecst Six isothermal temperatum coefficient measurements were obtained at the ARO configuration.

l The average measured isothennal temperatum coefficient was -3.37 pcWF. The predicted L

. value is -3.71 pcWF. The difference of 0.34 pcWF between the measured and pmdicted 1

temperature coefficients is well within the acceptance criteria of 4 pcWF. This result was l ~

then extrapolated to ilZP BOL, liFP BOL and HFP EOL conditions.

The extrapolated IIZP BOL MTC was -1.27 pcWF. The predicted value is 1.61 pcWF.

The Technical Specification requimment that the liZP BOL MTC shall be less positive than l

l 5.0 pcWF is met.

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e The extrapolated liFP BOL MTC was -8.37 pcWF. The predicted value is -8.71 pcWF.

The Technical Specification requirement that the IIFP BOL MTC shall be less positive than 0.0 pcmrF is met.

The extrapolated liFP EOL MTC was -26.97 pcmfF. The predicted value is 27.31 pcWF.

The Technical Specification requirement that the IIFP EOL MTC shall be less negative than 32.0 pcmrF is met.

Six isothermal temperature coefficient measurements were obtained at hot zero power with banks A, B, and D fully inserted and bank C fully withdrawn. The average measured isothermal temperature coefficient was 16.27 pcmrF. The predicted value is -13.90 pcmrF.

The difference of 2.37 pcm/"F between the measured and predicted temperature coefficients is within the acceptance criteria of 4 pcmrF.

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e 7.

CONTROL ROD BANK REACTIVITY WORTH MEASUREMENTS Obiective The objective of this test is to meast a the differential and integral reactivity worths of control roci banks B, A, and D.

Licene Reauirements Technical SpeciGcation 4.1.1.1 and 4.1.1.4 require verification of adequate shutdown margin prior to exceeding five percent power following a refueling. This test provides infomiation used to verify core design and thus, adequate shutdown margin.

Procedure At hot zero power, contml rod banks are inserted into the core in small increments using the normal sequence for operating bank insertion. Contml banks B and A as well as shutdown bank D were measured. The rate of insertion is governed by a reactor coolant dilution established by adding demineralized water to the RCS at 30 gpm. The tractivity of the core is continuously calculated and displayed on a strip chan by the reactimeter. The strip chart is then analyzed to determine the reactivity worth of each control rod bank movement.

t Results The memsured and predicted values for control rod bank worth are presented in Table 3. The total rod wonh measured was 9.58% greater than the piedicted worth. The results are well within the acceptance criteria of115% for an individual contro1 rod bank but close to the i

i10% criteria for total measured wonh.

Possible causes of the discrepancy were evaluated. An independent measurement of the total' control rod wonh was perforined to verify the design prediction. The reactivity difference between the all mds out and rodded bomn concentration was less than 0.7 percent of the predicted total control rod wonh It was determined that negative bias signals in the power -

L range channels (used for physics testing) created higher than expected rodded isothermal temperature coefficient and reactivity worth measurements. The nuclear instrumentation system was replaced during the refueling outage. A signincant difference was the use of I

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fission chamber detectors. The old system used ion chambers. The circuitry in the preamplifiers contain an offset for alpha decay of uranium 238 in the fission detector, a bias from a protection circuit, and a bias from a hardware testing circuit. Proposed corrective action includes the substitution of an attemate upper and lower preamplifier designed for physics testing and adjusting them to correct for the alpha decay signal. The alternate preamplifiers will climinate the testing and protection circuits. The attemate equipment will be l

put in service prior to and during startup physics testing. These amplifiers will be used in one of the four power range nuclear ins:rument channels. The other three channels will remain operable for reactor protection.

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,b TABLE 3 Contml Rod Bank Worths Coritrol Rod Bank Measured Worth. PCM Predicted Wonh. PCM

% Deviation B

-1043 956

-8.34 A

2361 2103

-10.93 D

2296 2095 8.75 Total

-57(X)

-5154 9.58

% Deviation = Predicted - Measured x 100 Measured 1

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RODDED CRITICAL BORON CONCENTRATION Objective The objective is to measure the rodded critical txron concentration at hot rem power conditions.

Licente Reautregggg None Procedure Bank C controls rods arr fully withdrawn. The remaining control rod banks are all fully inserted. Critical boron concentration is measured.

Results The rodded critical boron concentration was 895 ppm at 535'F. The predicted hot rem power boron concentration is 885 ppm. The diff erence of 10 ppm between the predicted and measured critical boron concentrations is well within the 100 ppm acceptance criteria.

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DIFFERENTIAL HORON WORTH Obiective The objective of this test is to measure the reactivity worth of the soluble poison in tenns of penVppm.

License Renuirements None Procedure Reactor coolant and pressurizer bomn samples are taken and analyzed at the equilibrium ARO f

and banks B, A, and D inserted configurations. The critical boron concentrations are corrected for temperature and rod configustion. The differential boron worth is calculated by dividing the measured bank wonh by the cha:

i boron concentration.

n Results Table 4 presents the data of the predicted and corrected just critical bomn concentration for the ARO and rockled configurations, the piedicted and measured total n.Tetivity wonh of banks B, A, and D and differential boron wonh. The measured differential bort,1 wonh was 9.01 %

greater than prulleted.

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d TABLE 4 Differential Boron Worth Menmred Predicted ARO/IIZP critical boron (ppm) 1599 1584 Rodded IIZP critical boron (ppm)

(Banks B, A Dinsened) 895 885 Boron Difference (ppm) 704 699 Hank worth (pcm)

(Sum of B, A.D) 5700

-5154 Differential boron wonh (pem/ ppm) 8.10

-7.37 Page 16of 27

.o 10.

TillRTY PERCENT POWER FLUX MAP Obiective The objective of the nominal 30% power Oux map is to determine if any gross neutron flux abnormalities exist.

1.leense Renuirements None Procedure One flux map is taken using the incore flux mapping system and evaluated using the INCORE computer code.

Results The results of the flux map demonstrated that the core power distribution is as predicted. A summary of the results is shown in Table 5, j.'

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TABLE 5 Summary of Results fmm 30%

Power Flux Map CY-XVI 1-459 Power. 31%, Burnup 43.8 mwd /Mtu (37 EFPH), Boron 1431 ppm, Bank B - 304 steps Core Penks 1

Adiusted kW/ft F. delta Il Menenred Limh Mensured Limh 3.4 12.9 1.447 1.93 1

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i incore Oundrant Power Tilt l

l hiCI3Ed LEDit 1

1.0056 l 1.0063 l

N/A 0.9984 1 0.9897 Core Averace AxialOffset 9.46 %

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11.

EIGHTY PERCENT POWER FLUX MAPS Obiective ne objective of the threc 80% power flux maps is to confirm the predicted core power distribution and to establish the incore/excore axial offset correlation.

License Requirements Technical Specifications 4.2.2 and 4.2.3 require that the linear heat generation rate and enthalpy rise hot channel factor be determined from incore measurements and evaluated before exceeding 80% of rated power. Additionally, the excorchncore axial offset calibration must be performed prior to exceeding 80% power.

Procedure Three flux maps are performed at approximately 80% power. A new incore/excore axial offset correlation is established based on these three maps. The axial offset indication is calibrated and the power distribution parameters are evaluated prior to increasing power from 80% to 100%.

i Results The results of the %% power flux maps produced power distributions that compared well with predicted y, lues and were within the Technical Specification limits. Based on the evWation of f. e 80% power flux maps, power was increased to 1(X)% A summary of the results is shown in Tables 6,7, and 8.

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TAHLE 6 i

Summary of Results From 80%

Power Flux Map CY.XVI 2-460 Power. 79%, Burnup.134.5 mwd /Mtu (114 EFPil), Boron. I170 ppm, Bank B - 312 steps t

o Core Peaks Adinsted kW/ft F. delta H Mencured Limh Measured Limh 8.2 12.9 1.44 1.70 Incore Oundrant Power Tilt Mensured Limk 1.0015 l 1.0052

......... l 1.02 0.9972 1 0.9962 Core Averace Axial Offset 2.76%

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0 TABLE 7 Summary of Results from 80%

Power Flux Map CY.XVI.3 461 Power. 80%, Burnup.150.9 mwd /Mtu (128 EFPH), Boron. I 160 ppm, Bank B. 285 steps care Peaks Adiusted kW/ft F delta Il Measured Limit Measured Limit 8.3 12.9 1.46 1.70 L

1 Incore Onndrant Power Tilt Measured Limil 1.0021 1 1.0069 1.02

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0.9988 1 0.9921 Core Average Axial Offset l

0.03 %

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TABLE 8 Summar/ of Results from 80%

Power Flux Map CY.XVI 4 462 Power. 80% Burnup.173.3 mwd /Mtu (148 EFPil), Boron. I130 ppm, Bank B. ' 65 steps Core Peaks i

Adiusted kW/ft F-delta H Mencured Lhnh Mensured Limh 8.5 12.9 1.45 1.70 Incore Ouadrant Power Tilt Mencured Limh 0.9974 1 1.0055 l

1,02

......... l 0.9993 1 0.9979 Core Avernce Axial Offset

. l.66%

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ONE HUNDRED PERCENT POWER FLUX MAP Objective The objective of the 1(X)% power flux map is to confirm the predicted core power distribution parameters and to verify the excore/incore axial offset correlation detemiined at 80% power.

License Requirements

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Technical Specifications 4.2.2 requires that the linear heat generation rate be evaluated based on incore measurernents at rated power following each refueling outage.

Procexlure A flux map is taken at rated power. Excore readings are also taken during the flux maps.

1 After the evaluation of the flux map, the excore/incore axial offset correlation is veriGed.

Results An incore flux map was performed at 1(X)% power. Data generated by this flux map were used in the evaluation of the excore/incore correlation. Allincore results were within the Technical Specification limits. A summary of the results is shown in Table 9,

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e, TABLE 9 Summary of Resultr fmm 1(X)%

Power Flux Map CY-XVI 14-472 Power 99.4%. Burnup. 439.6 mwd /Mtu (374 EIPil), Boron - 1031 ppm, Bank B - 310 steps Core Penks Adiusted kWlh F-deltn ll Mensured Limit Mensured Limit 10.11 12.90 1.45 1.60 Incore Oundrant Power Tilt Mensured Limit 1.0019 1 1.0045

......... l 1.02 0.9999 0.9937 Core Average AxlalOffset l-l

.l.169 %

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- 13.

- REACTOR COOLANT SYSTEM FLOW TEST Obiecdve l

The purpose of the teactor coolant system flow test is to measure the total vessel flo. ate (including core bypass flow) at 100% rated power operating cor.ditions. A precision heat balance was used and the results were corrected for all measurement uncertainties. The results of this test provide flow constants used for shiftly RCS ficw surveillance.

License Reauirenwnts Technical Specification 4.2.5.2 requires that the reactor coolant system flow rate be determined by a heat balance within 7 EFPD of achieving 100% rated thermal power r.ner each refueling outage.

e Procedure

'i A precision heat balance was established for each loop using the steam generators as the

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control volumes.; The following parameters were measun:d:

yn reactor coolant system pressure y

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hot leg temperatures 1

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+ ~ cold leg temperatures feedwater temperatures

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feedwater flow rates j

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feedwater pressure 1

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u, p

p

+- - steam generator pressun:s Since steam generator blowdown error was not considered in the flow uncertainty, blowdown i

was isoired during the period of data acquisition. The above data were used 'to calculate the 4 following time averaged paranieters for each of the four loops: steam generator heat transfer rate, primary coolant enthalpy change and cold leg specific volume. The hot leg temperatures were corn:cted for stratification effects by using empirically derived wo st case s*.atification values. The 4 loop test was conducted at 1811 MWth,563 'F average coolant temperature and 2014 psig average %nant pressure.

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q g :i-An uncertainty analysis was perfonned in 1986 for this measurement in accordance with NUREG/CR 3659,"A Mathematical Model for Assessing the Uncertainties of

- Instrumentation Measurements fse Power and Flow of PWR Reactors." The analysis m

considered the effects of all sw 3 of uncertainty in each instrumentation loop which was

~used. The uncertainty calculatie ns were updated in 1989. The flow uncertainty value for the

- four loop configuration was established to be 2.982% of the nominal flow rate,4.281% for the three loop configuration.

Results.

The best estimate reactor coolant system flow rate was determined to be 273,917 gpm.

Corrected for measurement uncertainties, the flow rate was established to be 265,749 gpm.

This flow rate is 19749 gpm (8.03%) greater than the minimum value of 246,000 gpm, as required by Technical Specification 4.2.5.2 Rated 'hermal power was reached at 13.1 EFPD.

The reactor coohnt system flow testing was completed at 19.8 EFPD, thus the flow test was conducted 6.7 EFPD after 100% rated thermal power operation was achieved. The flow test data is summarized in Table 10.

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Reactor Coolant System Flow Test Results Stratification Iggs 4TfB Correction (*F)

Flow Rate (GPM) 1 45.50

+ 0.35 70617

'2' 45.27

+ 0.35 67510 3-46.04

+ 0.45 69040 s-'

4 46.12

+ 0.85 66750 Total RCS Loop Flow Rate:

273,917 gpm 2.982 % Uncertainty Penalty:

8168 gpm f,

Minimum Guaranteed Flow Rate:

265,749 gpm t

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