ML20067C899

Startup Test Rept Unit 2 Cycle 10

ML20067C899
Person / Time
Site:	Farley
Issue date:	02/25/1994
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9403040330
Download: ML20067C899 (17)

Text

e Southern Nuclear Operating Company Post Office Box 1295 Birmingham. Alabama 35201 Telephone (205) 668 5131 L

Southern Nudear Operating Company o.v. uor.y Vice President 1

Farley Project the southern eleCfflC System February 25, 1994 Docket No.

50-364 U.S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington, D.C. 20555 Joseph M. Farley Nuclear Plant - Unit 2 Unit 2 Cycle - Stariuo Report Gentlemen:

Enclosed is the Startup Report for Unit 2 Cycle 10. If you have any questions, please advise.

Respectfully submitted,

'?h ))t;,

i Dave Morey REM / cit.U2 CYCL 10. DOC Enclosure cc:

Mr. S. D. Ebneter Mr. B. L. Siegel Mr. T. M. Ross i

l

$0D 0;).Q 9403040330 940225 l

i PDR ADOCK 05000364

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SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT STARTUP TEST REPORT UNIT 2 CYCLE 10 TABLE OF CONSENTS PAGE 1.0 Introduction 1

2.0 Unit 2 Cycle 10 Core Refueling 1

3.0 Control Rod Drop Time Measurement 8

4.0 Initial Criticality 10 5.0 All-Rods-Out Isothermal Temperature 10 Coefficient and Boron Endpoint 6.0 Control and Shutdown Bank Worth 11 Measurements 7.0 Power Ascension Activities 12 0.0 Incore-Excore Detector. Calibration 13 9,0 Reactor Coolant System Flow 14 Measurement APPROVED:

Technical Manager n

Nuclear Plant General Manager l

l C:\\WP51\\ FILES \\SUFORM

1.O INTRODUCTION The Joseph M.

Farley Unit 2 Cycle 10 Startup Test Report addresses the tests performed as required by plant procedures following core refueling.

The report provides a brief synopsis of each test and gives a comparison of measured parameters with design predictions, Technical Specifications, or values in the FSAR safety analysis.

Unit 2 of the Joseph M. Farley Nuclear Plant is a three loop Westinghouse pressurized water reactor rated at 2652 MWth. The unit began commercial operations on July 30, 1981.

The Cycle 10 core loading consists of 157 17 x 17 fuel assemblies, of which 42 are Westinghouse Low Parasitic (LOPAR) assemblies and the remaining 115 are Westinghouse Vantage 5 fuel assemblies.

Each of the 60 new Vantage 5 fuel assemblies loaded in the Cycle 10 core contains fresh. Westinghouse Integral Fuel Burnable Absorbers (IFBAs). In addition, the core contains two double encapsulated secondary source inserts and 12 previously burned wet annular burnable poison (WABA) in-serts. No thimble plug inserts were used. The design depletion of reac-tivity of the Cycle 10 core is 16500 MWD /MTU.

Previous Cycle Completion Dates and Average Burnuns Date Start of EOL EOL Burnup EOL Burnup Total Cycle Critical Cvele Date (MWD /MTU)

(EFPD)

EFPY 1

05-08-81 05-27-81 10-22-82 15350 416.50 1.141 2

11-30-82 12-03-82 09-16-83 10371 281.68 1.913 3

10-22-83 10-24-83 01-05-85 14639 397.73 3.002 4

03-08-85 03-20-85 04-04-86 13183 359.48 3.987 5

05-11-06 05-13-86 10-03-87 16674 457.67 5.241 6

12-02-87 12-05-87 03-24-89 16138 444.09 6.458 7

05-18-89 05-21-89 10-13-90 17051 468.76 7.742 8

01-03-91 01-06-91 03-06-92 14757 405.69 8.853 9

05-08-92 05-12-92 09-24-93 17352 462.00 10.120 2.0 UNIT 2 CYCLE 10 CORE REFUELING REFERENCES 1.

Westinghouse Refueling Procedure FP-APR-R9.

2.

Westinghouse WCAP 13842, Rev.

1 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 2 Power Plant Cycle 10)

Unloading of the Cycle 9 core into the spent fuel pool commenced on 10/1/93 and was completed on 10/3/93.

During the offload, each fuel assembly was inspected with binoculars for indications of damage or other problems. Two fuel assemblies scheduled for reload (Y39 and 2L35) were noted to have grid damage.

Following the cycle 9 core unload, an EPRI funded fuel inspection to gather baseline data for the Zine Addition program was conducted by Westinghouse on 20 fuel assemblies.

The fuel inspections consisted of oxide thickness measurements, high-magnification TV examinations and crud sampling. During the inspection program, the fuel inspectors performed 1

i l

TV visual examinations of the two grid damaged assemblies identified during the core offload. In addition, the assemblies which were adjacent i

to the damaged assemblies during the previous core cycles were inspected.

l During these examinations, it was confirmed that fuel assemblies Y39 and 2L35 had severe grid damage, and a third fuel assembly, 2L47, also sched-l uled for reload, was found to have a gouge-like defect on a corner rod.

l The cycle 10 core was redesigned to exclude these fuel assemblies from reload.

2L35 and 2L47 are vantage 5 assemblies.

The Cycle 10 Core reload commenced on 10/29/93 and was completed on i

11/2/93.

Due to an error in the core reload procedure, fresh fuel j

assembly 2M25 was mistakenly loaded into core location DS (vice assembly j

2M26). Since these fresh assemblies have the same enrichment, Management i

opted to leave 2M25 in location D5 in order to avoid the risk of damaging j

the assembly by moving it, and ascembly 2M26 was loaded into location M11 J

(vice 2M25).

t The as-loaded, redesigned Cycle-10 core, control rod locations, locations j

of burnable absorbers and source inserts, and the burnable absorber con-figurations are shown in Figures 2.1 through 2.5.

i l

2

Figuro 2.1 Unit 2 Cycle 10 R fsretnco Loading Pattern l

R P

N M

L K

J 11 G

F E

D C

B A

W62 2L32 W57 1

R102 R142 R140 Y34 2Ll6 2M43 2M48 2M50 2L44 Y13 2

R129 R107 Y43 2M56 2 Mil 2LO6 Yll 2LOS 2M18 2M57 Y51 3

R145 4W24D RI19 Ril4 4W20D R103 l

Y37 2L33 2M21 2L15 2M31 2L18 2M32 2L13 2M10 2L39 Y55 4

4W26D R128 12w62D Rll8 4W23D Y23 2M42 2M39 2L48 2L52 2L24 2M13 2L25 2L57 2L43 2M25 2M60 YO7 5

R110 R124 R135 R108 Ril5 R130 R137 2L30 2M27 2Lil 2L50 Y46 2M02 Y12 2M04 Y56 2L56 2Ll4 2M23 2L38 6

Rll7 R144 Ril6 R106 W58 2M44 2L08 2M24 2L21 2M05 Y52 2M17 Y47 2M06 2L20 2M20 2LO3 2M53 W60 7

R127 SS07 12w67D Ril3 R148 12w63D SS08

.R131 2L41 2M54 Y25 2L26 2M09 Y30 2M30 Y48 2M16 YO8 2Mi$

2L22 Y29 2M51 2L42 8

Ril2 R146 R134 R123 W53 2M52 2LO2 2M37 2L28 2M07 Y50 2M35 Y42 2M01 2L23 2M29 21.04 2M47 W51 9

l l

R139 R121 R133 R104 R138 R125 R141 2L49 2M40 2L10 2L55 Y45 2M08 YO6 2M03 Y41 2L54 2L12 2M28 2L31 10 4W25D R122 12w69D R136 4W28D Y16 2M49 2M26 2L40 2L53 2L27 2M34 2L19 2L51 2L37 2M14 2M41 Y22 11 R109 4W21D R120 R126 4W29D R147 Y44 2L34 2M38 2Ll7 2M12 2L29 2M19 2LO9 2M33 2L36 Y38 12 R105 R143 Y53 2M55 2M22 2L01 YO4 2LO7 2M36 2M46 Y40 13 4 North Rlli R132 R101 Y21 2L46

?M45 2M59 2M58 2L45 YO3 14 W56 Y49 W52 15 l

XXX

+ Insert Serial Number XXX

+- Fuel Assembly Serial Number ORIGINAL w/o No. of FUEL REGION U-235 ENRICHMENT ASSEMBLIES Region 9B (W) assemblies

. 4.202%

8 Region 10A (Y) assemblies 3.806%

16 Region 10B (Y) assemblies 4.185%

18 Region 11A (2L) assemblies 3.606%

29 Region 11B (2L) assemblies 4.005%

26 Region 12A (2M) assemblies....

4.201%

40 Region 12B (2M) assemblles 4.415%

20 Total.... 157 3

?

FIGURE 2.3:

Control and Shutdown Rod Locations 9

R P

N M

L K

J H

Q F

I D

C B

A i

1 A

D A

2 SA SA SP 3

l l

C B

SP 3

C 4

i SP SB SP SB 5

A B

D C

D 3

A 6

.9A SB SB SP SA 7

900 D

SP C

SP C

SP D

- 8 SA SP SB SB SA

- 9 A

B D

C D

5 A

10 SB SP SB SP 11 C

3 SP 3

C 12 SP SA SA 13 A

D A

14 Assonsa uAtum As-tv.cD 15 00 BANK NUMBBR OF BANK NUMBER OF IDENTIFIER LOCATIONS ID$NTIFIER LOCATIONS A

e SA e

B 0

85 8

C 8

SP

~ 13 D

8 4

1 i

I i

FIGURE 2.3:

Burnable Absorber and Source Assembly Locations s

i i

1j R

P N

M L

K J

E G

F E

D C

3 A

q e

e i

i i

1 1.

k 64I 64I 64%

2 l

i 64I 104I 104I 64I 3

i 4W 4W 104I 1041 4

j 104I 1041 I

4W 12W 4W 64I 44I 5

104I 104I 104%

i i

104I UCI 801 104I 6

1 1

642 104I 80I 104I 80I 104I 64I

-7 i

12W 12W 900 64Z 4SSA 104I 104I 4SSA 44I 8

104I 104I 64I 104I SOI 104I 80I 104I 64I

- 9 i

i 104I SOI SOI 104I 10 4W 12W

' 4W 7

64I 64I 11 104I 104I 104I i

4W

~4W i

104I 104I 12 1

104I 104I J

f.

64I 104I 104I 64%

13 i

64I 64I 64I 14 i

15 00 1

i TYPE TOTAL 1

99W...(NUMBER OF WABA RODLETS)...............

80

1. 1. 1. I..(NUMBER OF IFRA R0DS)..................

5248-

1. SSA..(NUMBER OF SECONDARY SOURCE RODLETS)...

8

)

l 1 -

5 i

n...

,--,w,r a

7 w..

3 FIGURE 2.4:

Secondary Source Rod Configurations

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4 i

1 l

i i

I

)

4 i

1 a

SECONDARY SOURCE ASSE!43LY a

4 6

l FIGURE 2.5:

Durnable Absorber Configurations 1

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1 i

30 CONTROL ROD DROP TIME MEASUREMENT (FNP-2-STP-112)

PURPOSE The purpose of this procedure was to measure the drop time of all full length control rods under hot full-flow conditions in the reactor coolant system to ensure compliance with Technical Specification Requirements.

SUMMARY

OF RESULTS For the hot full-flow condition (Tavg 2 541 *F and all reactor coolant pumps operating) Tw hnical Specification 3.1.3.4 requires that the drop time from the fully withdrawn position shall be s 2.7 seconds from the beginning of stationary gripper coil voltage decay until dashpot entry.

All full length rod drop times were measured to be less than 2.7 seconds.

The longest drop time recorded was 1.55 seconds for rod B-6.

The rod drop time results for both dashpot entry and dashpot bottom are presented in Figure 3.1.

Mean drop times are summarized below:

TEST MEAN TIME TO MEAN TIME TO' CONDITIONS PASHPOT ENTRY PASHPOT BOTTOM Hot full-flow 1.363 sec.

1.829 sec.

To confirm normal rod mechanism operation prior to conducting the rod drop test, the Verification of Rod Control System Operability (FNP-0-ETP-3643) was performed.

In this test, the stepping waveforms of the sta-tionary, lift and movable gripper coils were examined for anomalies, rod speed was measured, and the functioning of the Digital Rod Position Indicator (DRPI) and bank overlap unit were checked. In addition, the bank overlap unit switch settings and functions were verified to be correct. No abnormal indications were found during this test.

l I

l l

l

}

8 I

!\\,

Figure 3.1:

Cycle 10 Drive Line " Drop Timo" Tabulation R

P N

M L

K J

11 G

F E

D C

B A

1 1.417 1.367 1.400 l

1.867 1.833 1.850 2

l l

1.350 1.367 1.833 1.850 3

1.367 1.333 1.367 1.350 1.850 1.833 1.833 1.817 4

1.333 1.367 1.800 1.850 5

1.417 1.333 1.433 1.317 1.317 1.333 1.550 1.900 1.783 1.883 1.833 1.800 1.783 2.017 6

1.350 1.283 1.317 1.367 l

!.817 1.750 1.783 1.850 7

l.350 1.300 1.350 1.417 1.833 1.767 1.800 1.833 8

1.350 1.317 1.383 1.350 1.800 1.817 1.783 1.800 9

1.383 1.350 1.317 1.350 1.317 1.333 1.417 1.867 1.817 1.767 1.817 1.800 1.800 1.850 10 l.350 1350 1.817 1.817 11 1.350 1.350 1.350 1.350 l

1.800 1.833 1.833 1.800 12 1.350 1.350 1.817 1.800 13 4 North 1.500 1.383 1.417 1.950 1.867 1.883 14 15 X.XX

• Dreaker " opening" to dashpot entry (seconds)

X.XX

- 11reaker " opening" to dashpot bottom (seconds)

TEMPERATURE 546.90 PRESSURE 2272.54

% FLOW 100 l

DATE 11-27-93 9

i,*

1 f

4.0 INITIAL CRITICALITY (FNP-0-ETP-3601)

I PURPOSE i

]

The purpose of this procedure was to achieve initial criticality under carefully controlled conditions, establish the upper flux limit for the i

conduct of zero power physics tests, and operationally verify the cali-bration of the reactivity computer, s

SUMMARY

OF RESULTS i.

Initial reactor criticality for Cycle 10 was achieved during dilution J

mixing at 2251 hours0.0261 days <br />0.625 hours <br />0.00372 weeks <br />8.565055e-4 months <br /> on November 29, 1993.

The reactor was allowed to

}

stabilize at the following conditions:

RCS Pressure 2236.8 psig i

RCS Temperature 547.0 *F Intermediate Range Power

1. 2 x 10 Amp RCS Boron Concentration 1628.5 ppm Bank D Position 206 steps q

once criticality was achieved, the point of adding nuclear heat was determined in order to define the flux range for physics testing, and the reactivity computer calibration was verified by making positive and i

negative reactivity changes and comparing the reactivity indicated by the reactivity computer with values determined from the Inhour Equation.

4 5.0 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT (FNP-0-ETP-3601) 1 PURPOSE l

The objectives of these measurements were to determine the hot, zero power isothermal and moderator temperature coefficients for the all-rods-out (ARO) configuration and to measure the ARO boron endpoint concentration.

SUMMARY

OF RESULTS The ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration are tabulated below:

ARO. HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT Boron Measured ITC Design Acc. Calculated Rod Conc.

ITC Criterion MTC Confiouration nom ocm/*F ocm/*F ocm/*F All Rods Out 1643.5

+0.345

+0.38 i 2

+2.30*

• MTC result was normalized to all rods out (ARO) and to the ARO critical boron concentration (1643 ppm),

where:

ITC = Isothermal Temperature Coefficient, includes -1.92 pcm/*F Doppler coefficient.

MTC = Moderator Temperature Coef ficient, corrected to the ARO condition.

10

=-

EQLE:-The objective of the MTC determination is to verify that the most i

positive MTC that occurs during the cycle at power levels below 70% does not exceed the Technical Specification limit of +7 pcm/*F.

In the cycle 10 design, it was determined that the MTC would reach its maximum value followino BOL, at which time it would exceed the measured BOL value by 0.4 pcm/*F.

Thus, at powers less than 70%:

Cycle 10 maximum MTC = (2.3 pcm/*F +

0. 4 pcm/

• F) = +2.7 ocm/*F ARO. HZP BORON ENDPOINT CONCENTRATION Rod Confiouration Measured C.

(com)

Desion-oredicted C. (nom)'

All Rods Out 1646.8 1643 f, 5 0 Since the maximum Cycle 10 MTC

(+2. 7 pcm/* F) was less positive than the Technical-Specification limit of +7.0 pcm/*F, no rod withdrawal limits were required. The design review criterion for the ARO boron concentra-tion was also satisfied.

6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-0-ETP-3601)

PURPOSE The objective of the bank worth measurements was to determine the integral reactivity worth of each control and shutdown bank for comparison with the values predicted by design.

SUMMARY

OF RESULTS The rod worth measurements were performed using the bank interchange method in whicht (1) the worth of the bank having the ' highest design worth (designated as the " Reference Bank") is carefully measured using the standard dilution method; then (2) the worths of the remaining con-trol and shutdown banks are derived from the change in the reference bank reactivity needed to offset full-insertion of the bank being measured.

For Cycle 10, control bank B was the reference bank. The measured bank worths satisfied the review criteria both for the banks measured indi-vidually and for the total worth of all banks combined.

I

SUMMARY

OF CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS Control or Predicted Bank Shutdown Worth & Review Measured Bank Percent Bank Criteria (ocm)

Worth (oem)

Difference A

384 i 100 383.99

-0.003 B (Ref.)*

1191 119 1162.50

-2,39 C

055 i 128 827.78

-1 18 D

999 1 150 995.21

-0.38 SD - A 974 i 146 949.09

-2.56 SD - B 1008 t 151 976.38

-3.14 All Banks 5411 541.1 5294.95

-2.14

• The reference bank worth was measured by the dilution method.

11

-- - _ = - -

- _ - _. _. _- - - -. -l

~

7.0 POWER ASCENSION ACTIVITIES Upon completion of HZP physics tests, the following activities were performed during power ascension, or at full power:

1.

Incore movable detector system alignment.

2.

Measurement of NIS intermediate range (IR) channel currents in I

order to determine IR high flux trip and rod stop setpoints.

3.

Incore-excore AFD channel recalibration.

l 4.

Core hot channel factor surveillance.

5.

Reactor coolant system flow measurement.

6.

Generation of rescaling data for the OPAT and OTAT protection loops based on the 100% loop ATs measured during the RCS flow test.

At approximately 10% - 30% power, the determinat Of the incore system core limit settings (FNP-2-ETP-3606) was performed. The purpose of this procedure is to align the system so that the movable detectors stop at the correct core heights during flux mapping.

In order to invoke Technical Specification 3.10.3 test exceptions for HZP physics testo, preliminary intermediate and power range trip setpoints of less than or equal to 25% power were used for initial reactor startup and physics testing.

Since NIS intermediate range detector N36 was re-placed, the preliminary N36 channel trip setpoint and rod stop currents were set to 80% of the previous, Cycle 9 currents for this channel.

Following the completion of physics tests, the NIS power range high range high flux trip setpoint was increased to 80% to allow power escalation above 25%.

(The 80% setpoint, vice 109%, was administratively imposed to address the possibility that the power range channels initially could be indicating nonconservatively.) A power ascension limitation of 30.31%

(derived from the projected Cycle 9 - 10 change in core neutron leakage) was recommended prior to the first thermal power measurement to prevent inadvertently exceeding 35% power prior to channel calibration. Interme-diate Range detector currents measured at 29.5% power were used to gen-erate rod stop and high flux trip setpoints for Intermediate Range NIS Channels N35 and N36.

Af ter ramping the reactor to 48% power, the Incore-excore test (described in par. 8.0) was performed and the power range N41-N44 delta flux chan-nels were recalibrated, Following delta flux channel calibration, the power range NIS high flux trip setpoint was increased from 80% to 109%,

and a full-core flux map was performed at equilibrium xenon conditions at 48% power for core hot channel factor surveillance (FNP-2-STP-110).

At approximately 99% power, the RCS flow test (described in par. 9.0) was performed and the 100% power loop ATs were determined.

Since greater 1

than a 1% difference existed between the'new values and the ATs to which AT protection loops -1 and -3 were scaled, new OPAT and OTAT scaling data was given to I&C for recalibration of these channels.

As described in Table 7.1, core hot channel surveillance was initially performed under non equilibrium conditions using the incore-excore base case full core flux map taken at 48% power, and then under equilibrium conditions using full-core flux maps performed at 48% and 99% power. As shown in Table 7.1, all results were satisfactory.

12

-. - ~.

TABLE 7.1

SUMMARY

OF POWER ASCENSION FULL CORE FLUX MAP DATA Parameter Fuel Tvoe Mao 246 SID 252 Man 253 Avg. % power N/A 47.4%

48.7%

99.6%

Max power tilt

• N/A 1.0143 1.0148 1.0141 Avg. core % A,0.

N/A

+4.177

+0.762

+0.552 Max FAH Lopar 1.125 1.1344 1.1453 Vantage 5 1.5904 1.5750 1.5532 FAH Limit Lopar 1.791 1.789 1.551 Vantage 5 1.907 1.904 1.651 Limiting FQ(Z)"

Lopar 1.5323 1.5287 1.4227 Vantage 5 2.0970 2.0859 1.9474 FQ Limit Lopar 4.5571 4.5985 2.3062 Vantage 5 4.8125 4.8453 2.4574 Flux map N/A Non-equilibrium Equilibrium Equilibrium conditions Calculated power tilts based on assembly FDEN from all assemblies.

" Based on percent to FQ limit.

Fuel types referenced above are Lopar (low parasitic fuel, 42 assemblies) and Vantage 5 fuel (115 assemblies).

8.0 INCORE-EXCORE DETECTOR CALIBRATION (FNP-2-STP-121)

PURPOSE The objective of this procedure was to determine the relationship between power range upper and lower excore detector currents and core axial of f-set for the purpose of calibrating the main control board and plant computer axial flux difference (AFD) channels, and for calibrating the delta flux penalty input to the overtemperature delta-T protection system.

SUMMARY

OF RESULTS At an indicated power of approximately 48%, a full core base-case flux map was performed at the AO (+4.177%) obtained immediately following power ascension. Five additional (quarter-core) flux maps were performed at various positive and negative axial offsets ranging from -30.754% to j

+21.429% in order to develop equations relating - detector current to incore axial offset. Prior to ascending above 48% power, the power range NIS channels were adjusted to incorporate the revised calibration data.

During the refueling outage preceding the cycle 10 startup, the original analog power range channel detector current meters were replaced with permanently installed digital meters on all channels (N41 - N44).

The digital meters enhanced the accuracy and precision of detector current readings and reduced the error in the incore-excore test.

As a result, the excore quadrant power tilt ratio (QPTR) remained well within its 13 I

g

,.wi-

,--7

.r%.,

y<g

,y

limits during the ascension to full power and, at 99.6% power, the maximum QPTR was only 1.0029, well within the required limit of 1.02.

The revised detector current vs AO equations resulting from the Incore-Excore recalibration are tabulated below:

TABLE 8.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:

0.8026

• AO

+

159.1434 uA I-Top

=

-1.0551

• AO

+

158.8756 uA I-Bottom

=

CHANNEL N42:

i i

l I-Top 0.8079

• AO

+

163.2901 uA

=

-1.0928

• AO

+

158.4032 uA I-Bottom

=

CHANNEL N43:

I I-Top 0.8270

• AO

+

167.5974 uA

=

-1.1082

• AO

+

165.0391 uA I-Bottom

=

l CHANNEL N44:

I I-Top 0.9013

• AO

+

176.7149 uA

=

l I-Bottom

-1.2405

• AO

+

176.0114 uA

=

l l

9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-2-STP-115.1)

PURPOSE The purpose of this procedure was to measure the flow rate in each reactor coolant loop in order to confirm that the total core flow met the minimum flow requirement given in the Technical Specifications. In addi-tion, the RCS loop 100% delta-T values measured during this test are used to evaluate and, if necessary, to rescale the OPAT and OTAT protection channels.

l

SUMMARY

OF RESULTS In order to comply with the Unit 2 Technical Specifications, the total reactor coolant system flow rate measured at normal operating temperature and pressure must equal or exceed 267,880 gpm for three loop operation.

From the average of 12 sets of measurements, the measured RCS loop flows were:

Loop A = 94,375.9 gpm Loop B = 89,702.0 gpm Loop C = 92,206.6 gpm These combine to give a total measured core flow of 276,284.5 gpm, which satisfies the Technical Specification-requirement.

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l' The measured loop-aTs (normalized to 100.0% power) obtained during the i

RCS flow teat were:

Loop At 63.762 'F Loop B:

67.675 'F Loop C:

64.054 'F since more than a 1% difference existed between these values and the ATs to which Loops -1 and -3 were scaled, scaling calculations were performed and provided to I&C for recalibration of the OPAT and OTAT protection channels.

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