ML19351A253

Startup Test Rept (Test Activities from 890404-0719). W/ 891010 Ltr

ML19351A253
Person / Time
Site:	Farley
Issue date:	10/10/1989
From:	Hairston W ALABAMA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8910180203
Download: ML19351A253 (15)

Text

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• A>bama Ptwar Company [

g .o 40 inverness Center Parkway  ;

Post Office Box 1?95 i Birmingharn. Alabama 3L201 1elephone 206 00B 5561 ,

W. G. Heireton, til r senior vice Pre, ment  :

Nuclear Operations /\lahama %%tf f October 10. 1989 a . tre sw:wn we sn.rw . l Docket No. 50-364  :

i i

U. S. Nuclear Regulatory Commission l Attentions Document Control Desk ,.

Washington, D. C. 20555 Gentlemen l Joseph M. Farley Nuclear Flant i Cycle 7 - Startup Report lt

" .Endosed is the Startup Report for Unit 2 Cycle 7 as referenced in i our Cycle 7 Reload letter dated April 13, 1989. j f1 i

'/ l If you have any questions, please advise.  ;

i Respectfully submitted. l A / t

. :E  ? l V. G. Hairston, III  ;

.1  :

VGH,III/MDRimV.1049 ,

Enclosure  !

i cci Mr. S. D. Ebneter ,

Mr. E. A. Reeves  !

Mr. G. F. Maxwell 1

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i 8910180203 891010 PDR ADOCK 0"i000364 P FDC

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. JOSEI41 M. FAR11T NUCIE.AR PLANT r

$, UNIT JulBER 2 CYCLE 7 STAR 7UP TEST REKRT (Test Activities Froni4-4-89 to 7-19-89) y .

4 I

IREPARED BY THE ILANT REACTOR ENGINEERING GRCL'P I

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TABIE OF 00NTENTS PAGE 1.0 Introdtx; tion 1 2.0 Unit 2 Cycle 7 Core Refueling 1 [

3.0 . Control Rod Drop Tine Measurenent 6  !

4.0 Initial Criticality 8 '

5.0 All-Rods-Out Isotherval Tem;erature S Coefficient ard Boron Erdpoint }

f 6.0 Control and Shutdown Bank Worth 9 i Measurements 7.0 Startup and Power Ascension Procedure 10 8.0 Incore-Excore betector Calibration 11 9.0 Reactor Coolant System Flow 13  !

Measurement.

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APPROVED:

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Technical Manager i

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[ _ b Md General Manager - Nuclear Plant l

b/ST-UP_RIT

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1.0 INITODUCTION i

The Joseph M. Farley Unit 2 Cycle 7 Startup Test Report addresses the tests performed as required by plant procedures following core refueling. The re; ort provides a brief synopsis of each test and gives a ocmparison of neasured 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 pressurir.ed water reactor rated at 2652 MWth. The Unit began conmerrial operations on July 30, 1981. The Cycle 7 core loading consists of 157 17 x 17 fuel assemblies.

Cycle Ocnpletion Date ard Average Burnup pycle Numler Completion Date Avg. Durnun (FTnT)#TQtJ 1 October 22, 1982 15.350.5 2 September 17, 1983 10.371.2 3 Jar..ary 4. 1985 14,639.0 4 April 4, 198S 13,183.8 5 October 2, 1987 16,674.0 6 thrrh 24, 1989 16.137.83 2.0 UNIT 2 CYCLE 7 CORE REFUELING REFERD4CES

1. Westinghouse Refueling Prccedure FP-APR-R6

2. Westinghouse WCAP 12193 (The Nuclear Design and Fhnagement of the Joseph M. Farley Unit 2 Power Plant Cycle 7)

Unloading of the Cycle 6 core into the spent fuel yool commenced on 04-04-89 but was not complete until 04-12-89 due to problems with the fuel transfer cart. During the unload each fuel assembly was visually inspected with binoculars; no significant fuel danage or defonnation was noted. Therefore, no changes to the design Cycle 7 core loading pattern were required. Cycle 7 core load tegan on 04-18-8f und was completed on 04-19-89.

The as-loaded Cycle 7 core is shown in Figures 2.1 through 2.4.

which give the location of each fuel assembly and insert, including wet annular turnable absorber insert locations ard configurations.

The Cycle 7 core has a nominal design burnup capability of 16,990.0 MWD #TIV.

1

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FIGURE 2.1 UNIT 2 CYCLE 7 RETERENCE LOADING PATTERN f i

A B C D E F G H J K L M N P R j

216 303 58 - - - 15 I see USS eso  ;

6Wl10 275 ~ g4 )

IN titt 6WiSD R164 tiet ~ i U16 UN mt u68 WO Utl 930

- .. .. .. ... . .. .. .. l 216 6WR3B 13WDee 0143 13W660 titt item 6Was M1 _ ,_,, g 3 fl0 de wie U64 Wit Un6 WIS We f24 221 01M 13WB3B R103 14Wpp 3906 tente R130 1 M 1D R144  !?S - - 12 961 WS WOD WI? U36 W36 U10 W11 u63 116 118 390 4WBOD 1W700 6136 1 M 10 IFS 13W750 336 1W760 81M 1 M iD 4Wl60 263 _ _ g1 ,

URI Wl9 Wt2 Utl WRI til wie 106 WO6 Ulf Wil WD4 Uti  !

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.. . ...... .... .. . - .. ~ . ~..

- .. .~. . .. .. ~ .

A135 1 M 79 8145 13W900 0108 1W730 stia 13WE0 tief i M 10 8124 1NT7D G137 _ _ go  !

U31 Wpl USS W7 Ulf WO6 168 me 969 W3 UO6 W1 UM ,

... . ...... ...... .... ... . . . ... ...... l 16diD IM 13d30 tiff ISS R132 1*W730 254 i m 7D R131 643D 740 g 296 6W17D tit!

664 W1 W60 Wie 108 W38 Uit USS W33 W19 ft1 WDF u62 We SM  ;

269 R1M 1WFID 276 iW8?D R106 Ill 13dB 211 8130 13W100 MS 13dle ti&O 237 g i u67 Ull we U30 WO9 T49 USS W u61 TSO W39 U00 W38 ue0 U37 3 1W101 Ret 16m30 0110 4WI2D 250 7 235 .Wie 8101 tende 254 13W90D R115 264 titt Mt Wil U64 W9 160 W33 Ute Ul4 U0S WO 156 WI U39 WS 566 .. .... ...... ... - . .. - .  !

... . . .. . .. .. .~ . ..

0117 1N060 0119 itW40 R105 itW630 8138 0123 12W9?D tiOT tant 2120 13dcD 6 UO3 W10 Uti Wil 963 W34 167 W0i Ulf WID Wie W1T WI6 itW98 R139 1NP30 373 1W79 F30 13W990 8133 13Wh6 6WtiD 262 5 287 4WB00 >

Wie 734 161 W50 U02 WO4 W64 U23 Wil Wil WIS V17 W31 239 R162 13 Weed till 16mm 8803 tema 8111 13Wpl0 0164 707 4 110 un E6 U18 WDR U01 WF Uit W35 Vll fl4 223 6Wte 13W640 0102 13 Whee R116 1W7DD 4W250 210 3 101 m3 W21 W65 W18 US0 WC3 Wl7 161

.. . .. . ... .- ._ ~ .. .. ...... l 791 810F 6Wt?D R1M 6400 0116 205 2 U32 Utt wl use et Uit vor t Il4 283 270 g S3F Ul1 lie  ;

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("* FU(L A$$(MlLY IN$(Ri $(RIAL NUMl(R , North '

("* FU(L A$$(M6LY $[R] AL NWl[R j The original w/o U-235 enrichernts were: No. of Tuel Assemblies Region 6 (S) assemblies .... 3.443% Region 6 ............ 8 Region 7 (T) assemblies .... 3.603% Recion 7 ........... 20 Region 8A (U) assemblies ... 3.598% Region BA .......... 36 Region 8B (U) assemblies ... 3.994% Region 8B .......... 28 Region 9A (W) assemblics ... 3.800% Region 9A .......... 49 Region 9B (W) assemblies ... 4.200% Region 9B ........... 16 Total 157 2

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• I TIct'RE 3.2 3 1

4 i CONTROL ROD LOCATIONS -

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R P N M L K J H 0 F E D C B A 1

1 2 A D A 3 M M W  !

i 4 C S P B C  !

5 W W W M S A B D C D B Al i I 7 SA SS St SP SA l  :

8 po- D SP C SP C SP D l i

9 SA SP St SS SA l 10 A B D C D B A i 11 SS p l SS SP i i

12 C S SP B C f 13 SP SA SA i

14 A D A ,

IS Abs glet:

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FUNCTION NUMBER OF CLUSTERS Control Bank D 8 Control Bank C 8 Control Bank B 8 Conwel Bank A 8 Shutdown Bank S8 8 '

Shutdown Bank SA 8 SP (Space Red Locatione) 13 3

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--- --- . _ . . . - . , . . . . _ . ~ . - . . - . _ . - - - . . - , . - - - - - . . . - . . . . - . . . . . . . . . - . . - . . , _ . - . . . . . . .

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. .. j FIGURE 2.3 j

, SURNABLE ASSORBNL APG SOURCE ASSEMBLY LOCATIONS I

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i R P N M L K J H G F E O C S A l l

1 l

2 4 4 3 4 12 12 12 4 l 2

18 SS it 12 i 4 12 I

\ i 5 4 12 12 12 12 12 4 l i

12 12 12  !

8 12 12 12 .

7 4 16 12 12 to 4 ,

i 8 po* 12 12 12 12 12 4 16 12 12 16 4  ;

9  !

12 12 12 12 12 12 10 4 12 12 12 12 12 4 ,

l 11 ,

i 12 14 SS 14 12  ;

12 13 4 12 12 12 4 14 4 4  ;

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1. 1. Number of WASAs Summary of Inserts g

SS Secondary Source l

I k'ABA Clusters 65 684 k'ABAs in Control Rods 48 65 Clusters Thimble Plugs 42 l

Sec. Sources 2 Total 157 1 4 l

FIGURE 2.4 cuRwasta AssonssR Awo stCONDARY s0URCE ROD CONMOUMTIONS 1

i O O O E O E  !

E E O O l O O O O O E D E D E )

O O O O O E E O  !

O O O O O E D E D E ,

E E O O O O O E D E t 4 sa c nviour si.n 12 sa c.ntieursii.n l i

, 0 0 O,

,5 O E, E O E O E O O O O O i O E E G O O O O E O E O E O O O O O E O E O O O 16 sA C.nfleurati.n secondary tourse Rods 5

1 l.--. -- . - . . . . _ . , _ . . . - _ . . _ . _ _ _ _ _ _ _ _ , _ . . _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

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, 3.0 CON 1ROL NOD DROP TIME HEASLRDENT ( a'P-2-STP-112) t,< >

The purpose of this procedure was to neasure the drop tine of all '

full length control rods under hot-full flow corxiitions in the reactor F

coolant system to insure compliance with Techrilcal Specificatiori Requirements.

SW1AIN OF RESLLTS l For the hot full-flow condition (Tava 1541 deg.F ard all reactor l coolant pumps operating) Technical Specification 3.1.3.4 requires that  :

the drop tine from the fully withdrawn position shall be 12.2 seconds  ;

from the beginnira of stationary gripper coil voltage decay until  :

dashpot entry. All full lerdth rui drop times were measured to be less than 2.2 seconds. The longest drop tine recorded was 1.49 l seconds for rod B-6. The rcxi drop tine results for Loth dashpot entry l'

ard dashyot lottom are presented in Figure 3.1. Mean drop tines are

stmstarized below

j TEST MEAN TIME TO MEAN TIME 10 CONDITIONS DASHIVT BTRY DASHTOT Dcrtin]  ;

a f Ilot full-flow 1.372 see 1.87 see I i

To confirm nornal rod nechanism operation prior to conducting the  ;

rod drop test, the Verification of Rcd Control System Operability t

! (FNP-0-ETP-3643) was perfonned. In this test, the stepping waveforms of the stationary, lift and movable gripper coils were examined, rcd speed was measured, and the functioning of the Digital Rod Position  :

Indicator (DRPI) and innh overlap unit was checked. During the test, the follouirs problems were noted:

(1) In the 1AC power cabinet, a 30 amp fuse was blown. This was replaced.  :

L (2) A ground was found on the lift coil cimuit of rod J13, which was traced to Module F of containment penetration B032. bkdule F was i replaced and the three rods (M12. J13 and G9) which had leads passing

• through this module were re-tested with satisfactory results.

(3) During the rod stepping test, small waveform anomalies were noted -

for rods K4 and D6. The results were reviewed with Westinghouse engineers and it was determined that the traces were satisfactory.  ;

(4) Due to small waveform anomalies, rcxis F14 and K8 were re-tested with the different operating coil leads reversed (to see if the leads were connected in reverse polarity). No problems were disclosed.

(5) The lift coil leads for rcxi D12 were found to be connected in reverse polarity. This reversed the nagnetic polarity of the rod D12 drive shaft. causing its rod drop trace to be reversed (with respect to the rennining rod drop traces). However, the drop tine of rod D12 was nornal . The lift coil leads were restored to correct polarity.

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O UNIT 2 CYC1.E 7 l

, NORT!!  !

KY i

1.36 1.37 1.39 p l 1.82 1.91 1.89 f 1.38 1.35 ,  :: !

1.90 1.85 [

1.35 1.38 1.36 1.38 M 1.87 1.87 1.82 1.89 [

\ 1.37 1.35 A

m. ' E 1.84 1.85 K !

1.47 1.36 1.34 1.32 1.35 .l.36 1.39 1.99 1.88 1.85 1.75 1.88 J.86 1.90 $ f 1.36 1.34 1.34 1.36 _g ,

g 1.87 1.87 1.83 1.87 ,

N 1.40 1.47 1.35 1,38 -H I 1.91 1.86 1.80 1.87 i 1.37 1.33 1.34 1,38 _

1.85 1.83 1.86 1.89 1.38 1.40 1.34 1.32 1.38 1.38 1.38 F 1.89 1.80 1.86 1.89 1.87  !

1.87 1.90 1.37 1.37 g 1.87 1.89 ,  ;

i 1.38 1.35 1.34 1.35 D 1,82 1.85 1.84 1.82 1.37 1.38 C ,

1.85 1.89  :

1.38 1.43 1.49 B 1.89 1.94 2.04 j

/ A g- i i

1f '

5 4 3 2 1 15 14 13 12 11 10 9 8 7 6 l

DRIVE LINE " DROP TIME" TABULATION TEMPERATURE - 546.s91 PRESSURE- 2236.67  % FLOW - 100 X.XX BREAKER " OPENING" TO DASHPOT ENTRY - IN SECONDS DATE- 5-11-89 X.XX BREAKER " OPENING" TO DASHPOT BOTTOM - IN SECONDS FIG. 3.1 7

h 4.0 INITIAL CRITICALITY ( FAT-2-ETP-3601 )

ItTUOSE The p trpose of this procedure was to achieve initial criticality under carefully controlled conditions, establish the upper flux limit for the corrf.uct of zero power physics tests, ard operationally verify the calibration of the reactivity computer.

Sltt%RY OF RESlLTS Initial Reactor Criticality for Cycle 7 was achieved during dilution mixir.g at 0205 hours0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br /> on May 18, 1989. The reactor was allowed to stabilize at the following conditions:

RCS Pressure 2235 psig RCS Temperature 547.10F Intermediate Range Power 1.1 x 10 6 A RCS Boron Concentration 1991.5 p n Bank D Position 180 steps Following stabilization, the paint of adding nuclear heat was detennined and a checkout of the reactivity computer using positive and negative flux periods was perfonmxi. In addition. NIS source and intermediate range overlap data were taken during the flux increase prior to criticality t.o denonstrate that adequate overlap existed.

5.0 ALL-RODS-OLT IS711tER'1AL TD!pEROLTtE COEFFICIENT AND BORON ENDTOIhT IFNp-2-ETp-3601) pt'RFOSE 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 measum the ARO boron erd} mint concentration.

SIM1WY OF RES1*LTS The ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration follow below.

ARO. HZp ISOIWERMAL AND FODERA70R TDIPERA7tTtE 00FIFICIENT Boron Measured Olt Design Acc. Calculated Rod Conc. c2 r Criterion of .. o Configuration pra Lg/0 F rem /0 F pcm/0 F All Rods Out 2006 -0.63 -0.86 1 3 +1.71 at t : Isothennal temperature coefficient, includes -2.26 pom/0F doppler coefficient af, .. e : Fkxierator only temperature coefficient, nonnalized to the ARO Corxlition 8

Q ,

=

~,;

ARO. HZP BORON ENDICINT 00NCENTRATION

• r Rod Confirutstion Measured Ca ( pre ) Desiei-predieted Ca ( proJ  !

All Rods out 2011 1996 1 50

%us, no red withdrswal limits were needed for the moderator temper-  !

ature coeffielent to be within the Technical Specification limit of j

+5.0 pom/*F. The design acceptance criterion for the ARO  ;

oonoentration was satisfactorily met.  !

P 6.0 00NTHOL AND SHUITJ0WN BANK WCRTH MEASURD11NIS (IHP-2-ETP-3601)  :

RRPOSE he objective of the bank worth measurements was to determire the integral reactivity worth of each control and shutziown bank for ,

comparison with the values predicted by design.  !

Sltt1ARY OF RESULTS l The rod worth measurements were performed using the bank inter- .

change acthod in which: (1) the worth of the bank having the highest design worth (designated as the " Reference Bank") is carefully neasured using the standarti dilution nethod; then (2) the worths of the remaining control and shutdown banks are derived from the change in the reference bank reactivity needed to offset full insertion of the bank being measured.

The control and shutdown bank worth measurement results are given ,

below. %e measured worths satisfied the review criteria both for the -

banks measured individually arxl for the combined worth of all the banks. .

Stlt1ARY OF 00f(IROL AND SHl.TTDOWN BANK WORTH MEASUR&DNTS Control or Predicted Bank Shutdown Worth & Review Measured Bank Percent Bank Criteria (ocal Worth (nce) Differetcq A 378 1 100 337.7 -10.7 B (Ref.) 1252 1 125 1220.7 -2.0 C 749 1 112 728.7 -2.7 D 1134 1 170 1110.5 -2.1 SD - A 1079 1 161 1035.5 -4.03 SD - B 1002 1 150 990.7 -1.1  !

Combined 5594 1 559.4 5429.8 -2.94 '

• Measured by dilution method 9

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7.0 STARTUp AND POWER ASCENSION PROCEDURE (WP-2-ETP-3605) ,

t R EFOSE j i

ne purpose of this procedure was to provide controlling instructions )

I for:- l

1. ' NIS intermediate and power range setpoint changes, as required  !

prior to startup and during power macension.

2. Rasp rate limitation and control rod movement r=-j .istions. I L  :

3. Corduct of startup and power macension testing, to include:

a. H2P physics tests (FNP-2-ETp-3601). .

b. Incore movable detector system alignment (FNP-2-ETP-3606). l

c. Incore-excore AFD channel recalibration (FNP-2-STP-121).

d. Core hot channel factor surveillance (FNP-2-STP-110). j

e. Reactor coolant system flow measurement (FNP-2-STP-115.1). ,

SlM1ARY OF REStLTS -

In order to satisfy Technical Specification requirunents for  !

invoking special HZP physics test exceptions, preliminary trip set-points of less then or equal to 25% power were used for the NIS inter-  !

mediate and power range channels. When physics tests were completed.  !

the power range setpoint was increased to 80% to allow power escala- -

tion above 25% for calorinetric recalibration of the power range  ;

channels. (The 80% setpoint was used instead of 109% in case the uncalibrated power range channels were indicating non-oonserva-tively.) At approximately 40% power, the power range channels were recalibrated, the high-range trip setpoint was restored to 109%, and setpoint currents were determined for the intermediate range channels. j

%e Westinghouse fuel warranty limits the power ramp rate to 3% t of full power per hour between 20% and 100% power until full power has been sustained for 72 etanulative hours out, of any seven-day operating i period. His ramp rate was observed during the ascension to 100%

power.

As exnlained in Unit 2 Incident Report No. 2-89-205, it was found during power ascension that the power range nuclear instruments were l reading low (by approximately seven percent at 55% power). %e cause l of the problem was traced to leaking bypass valves in two of the three i feedwater flow instrunents, which introduced errors into the calcula-ted thermal power used to calibrate the power range nuclear instru-ments. The leaking bypass valves were repaired and the power rarute nuclear instruments were readjusted to indicate to within 2% of actual thermal power. %e affect of this problem on startup testing is out-lined in Section 8.0.

10

, Detensination of incore movable detector system core limit l settings (FNP-2-ETP-3606) was accomplished during the maoension to 40% >

power. 'Ihis was followed by the incore-excore recalibration test  !

(m'P-2-STP-121) at 40% power and the reactor ooolant flow measurement  !

in'P-2-STP-115.1) at 100% power, which are described in Sections 8.0 l and 9.0 of this report. As summarized in Table 7.1. core hot channel I factor surveillance was perfonned on the incore-exoore full-oore base case flux nap taken under non-equilibrium conditions at 40% power, and >

on the full-core flux maps performed at 55% and 100% power under i equilibrium corxtitions. i TABLE 7.1 i

SWt4ARY OF IWER ASCENSION FLUX MAP DATA Parameter dmo 170 Man 177 Map 178 j

?

Avg. % power 40.190 54.75 99.91 Max FAH 1.5067 1.4994 1.4704 '

Max power tilts 1.0043 1.0030 1.0023 Avg. core % A.O. 1.256 1.450 0.179 Maximtso M(Z) 2.0750 2.0075 1.8617 N Limit 4.6168 4.6052 2.3084 Xenon conditions Non- Equilibritan Equilibritan Equilibritzn

• Calculated power tilts based on assembly FAHN from all assemblies. '

I 8.0 IN00RE-EXOORE DETECTOR CALIBRATION (FNP-2-STP-121) ,

l PURFOSE l The objective of this procedure was to detennine the relationship between power range upper and lower excore detector currents and axial offset for the purpose of calibrating the control board and the plant computer axial flux difference ( AFD) channels, and for calibrsting the delta flux penalty to the overtemperature delta-T protection system.

SUtiARY OF RESUL'lE At an indicated power of approxinately 35%, a full-core base case flux map and five quarter-core flux maps were perfonned at various positive and negative axial offsets to develop equations relating detector current to core axial offset. To reduce error, all flux maps were perfonned at essentially the same RCS temperature. The power range NIS channels were adjusted to incorporate the revised calibration data.

11

p When the indicated power was increased to approxinately 48%.

further power escalation was halted due to the develoyment of exces- '

sive calculated quadrant power tilt ratios (QTTRs). At approxinately the same time. it was discovered that the byymss valves on two of the feedwater flow instrunents had teen leaking, causing the calculated (calorinetric) reactor thermal gewer to te lower than actual jewer.

l (When the instrunent problem was corrected, it was found that the actual power at the previously-irdicated 48% was nearly 55%. ) Thus, it is susrected that the QITR problems noted during power ascension may have resulted largely from the error introduced into the incore- l

! excore equations by use of the wmng percents power in the test I calculations.  ;

i

, To verify that core peakirut factors at 55% power were E

satisfactory, a full core flux nap was perfonned. Then the incore-excore calibration was corrected to eliminate further spurious QFfR indications by perfonning an I-zero normalization. This pennitted

ower to be increased to 100% with no further QPI'4 problems. A full- ,

core flux map performed at 100% power demonstrated that the incore-e>, core calibration was still satisfactory at full lower and that no further re-normalization was required.

TABLE 8.1 I

DETECTOR CURRENT YERSUS AXIAL OFFSET EQUATIONS OBTAIN FTO) INOORE-EXOORE CALIBRATION TEST CHANNEL N41:

1-Top = 0.6324*AO + 162.70 uA I-Bottom = -1.22228AO + 160.02 uA t

CHANNEL N42: .

I-Top = 0.7873*AO + 162.99 uA I-Bottom : -1.3251*AO + 156.97 uA CHANNEL N43:

I-Top = 0.78078AO + 169.79 uA I-Bottom = -1.34 *AO + 163.73 uA C RNNEL N44:

1-Top  : 0.8118*AO + 181.01 uA I-Bottom = -1.6376*AO + 184.51 uA 12

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,4 l g '.

.> 9.0 RFAC'IrJR 000LWT SYSTEM FIN MEAStREMENT (FWP-2-STF-115.1) {

i A? MRPOSE l 1he purpose of this procedure was to measure the flow rate in each reactor coolant loop in oMer to confim that the total core flow i met the minimum flow requirement given in the Unit 2 Technical -

i Specifications. l

, .4 l

Stewn w nEsttTs  ;

To comply with the Unit 2 Technical Specification, the total [

reactor coolant system flow rate measured at nomal operating

• temperature and pressure must equal or exceed 265,500 rpm for th me ,

loop operation. From the average of six calorimetric heat balance  :

! measurements, t.he total com flow was detemined to be 279,148 gpn,  !

L which meets above the criterion.

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