ML20070L743

Corrected Rev 1 to Startup Test Rept

ML20070L743
Person / Time
Site:	Farley
Issue date:	03/06/1991
From:	ALABAMA POWER CO.
To:
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ML20070L746	List: ML20070L743 ML20070U190
References
NUDOCS 9103200213
Download: ML20070L743 (14)

Text

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J. D. Woodard /\lLil)lilllil POWCT ,

Vwe Pic & n-N nes raw Progo " " *" " * """"'

March 14, 1991 U.S. Nuclear Regulatory Connission -

ATIN: Document control Desk

Washington, D.C. 20555 i

Gentlemen

Joseph M. rarley Nuclear flent 1 Unit 1 Cycle 10 StartUp Re y rt Enclosed is Revision i to the Unit 1 Cycle 10 startup Report which was originally sutnitted by letter dated rebruary 1, 1990. Subsequent to this subalttal an error was discovered in the rod worth data provided.

, Additionally, incorrect values for the most limiting P z and r shown in Table 7.1 were included with the retort. Thilr(ev)ision,as

, corrects these errors.

1 I If you have any questions, please advise.

p ,D Woodard m k NQ. v JN/BWa stj3346

,. Enclosure )

cc S. T. Hoffman S. D. Ebneter ,

G. F. Maxwell 1 1

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pkRO3200213 ADOCK 050oo34g 910306 > .; I l'

4 PDR 1

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f I ALABAMA POWER COMPANY l JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER I CYCLE 10 REVISION 1 4

f' STARTUP TEST REPORT PREPARED BY THE PLANT REAC10R ENGINEERING GROUP i l

l t

TABLE OF CONTENTS PAGE l 1.0 Introduction 1 2.0 Unit 1 Cycle 10 Core Refueling 1 3.0 Control Rod Drop Time Measurement 6 4.0 Initial Criticality 8 5.0 All-Roda-Out Isothermal Temperature 8 Coefficient and Boron Endpoint 6.0 Control and Shutdown Bank Worth 9 Measurements 7.0 Startup and Power Ascension Procedure 10 8.0 Incore-Excore Detector Calibration 11 9.0 Reactor Coolant-Systen Flow 12 l Measurement i ,

APPROVED:

Technical Manager.

1 C J.6-f /

8b 2ht/ General Manager - Nuclear Plant 1

/ l PN/STARTRPT.

. . *s '

1.0 INTRODUCTION

The Joseph M. Farley Unit 1 Cycle 10 e.tartup 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 1 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 December 1, 1977. The Cycle 10 core loading consists of 15717 x 17 fuel assemblies.

Previous Cycle Completion Dates and Averare Burnups Date Start of EOL EOL Burnup EOL Burnup Total Cycle Critical CycJe pats (HWD/HTU) (ETPD) EFPY 1 08/09/77 08/18/77 03/08/79 15,450 420.60 1.152 2 10/31/79 11/04/79 11/07/80 10,177 276.70 1.910 3 03/25/81 04/03/81 09/10/81 S.180 140.70 2.296 4 03/03/82 03/07/82 01/14/83 10,622 288.10 3.085 5 03/28/83 03/30/83 02/10/84 11,096 301.30 3.911 6 04/22/84 04/24/84 04/06/85 12,238 333.58 4.825 7 05/26/85 06/27/85 10/03/86 17.231 470.04 6.112 8 11/30/86 12/02/86 ^3/25/88 16,190 443.26 7.327 9 05/20/88 05/21/88 09/23/89 17,456 479.29 8.640 l 2.0 UNIT 1 CYCLE 10 CORE REFUELIN6 l

REFERENCES

1. Westinghouse Refueling Procedure FP-ALA-R9

2. Westinghouse WCAP 12371 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant Cycle 10)

Unloading of the Cycle 9 core into the spent fuel pool commenced on 10-04-89 and was completed on 10-06-89. During the unload each

! fuel assembly was visually inspected with binoculars: there was a l light, small mottled crud pattern on several assemblies, but no l sigr.ificant fuel damage or deformation was noted. Therefore, no changes to the design Cycle 10 core loading pattern were required.

Cycle 10 core load began on 10-15-89 and was completed on 10-16-89.

l The as-loaded Cycle 10 core is shown in Figures 2.1 through 2.4, I which give the location of each fuel assembly and insert, including l

wet annular burnable absorber insert locations and configurations.

The Cycle 10 core has a nominal design burnup capability of 16,800 FMD/MTU.

1

1.

i . . ... .

I. '

i l . FIGURE 2.1: UNIT 1 CYCLE 10 REFERENCE LOADING PATTERN l i i '

1 L

l l R FkN =

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OS all 8 03 4A21 aste IA13 3s19 anst gae 496 Bal 1MOs Ill 16W750 stol I M is 439 1 M ts $19 33 ~ ~~~' 4 Kol aASI Isla 3Aa6 asM 241T aset IMO Is68 aAll ros 199 wIIs 1 M al til 1 M it M 18 M IMit Me 1 M 39 6mm 3 - -5 764 eines ages 3Att as27 2A31 2AM IA09 BIS 3Aas 3 33 3347 Pel i .-

M sg 85 1MM e44 IMit All 1M17 H3 1MW RM M3

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tis tie IMF 262s 2Aes acts nos tete s24 Isot u2F 2 31 3A37 3 14 2A48 25e Ama nas 14v7FD 3r t M a6 elf to sit 1Mit s iMas em aman is 7 2Ase Iset 2AM asie 3Ato as39 3A14 2A10 2A19 R37 aAN 230 2Att test 2A45 6es em 1 M ot ric rm ens us tow 6e tas ass se as t M ie att as g_

sts aus asas aAss taas est taol ana9 aats est aAsf aArt asas aut use N suis aos t w 68 618 iMN tos 74 a14 1Maa 4 two ter wse 4e ,

, tot asse 3 Ate au tatt asse anos aAas aAer as41 aAle asef aAct Ist? 2AM 3

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la wste i m te a37 iM27 11 em so imis not i m ot ama e gy l ' P39 tsM tsN IA05 as43 2Act 3Alo RA27 2005 2A13 tatt an95 g34 l

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O y The original w/o U-235 enrichments were No of Fuel Assemblies Region 6 (F) assemblies ... 2.995% Region 6 ........... 5 Region 7 (G) assemblies ... 3.002% Region 7 ........... 4- .

Region 6A (H) assemblies ... 2.999% Region 8A .......... 6 Region-10 (K) assemblies ... 3.597% Region 10 ......-.. 13 Region llA (2A) assemblies .. 3.805% Region llA ........ 40 Region llB (2A) assemblies .. 4.207% Region 118 ........ 28 Region 12A (2B) assemblies .. 3.803% Region 12A ........ 45 Region 12B (2B) assemblies .. 4.200% Region 12B ....... . 16 157-2

TIGURI 8 8 CDNTROL ROD LOCATIONS R P N M L K J H G F E D C 8 A 1

2 A D A 3 SA SA SP s C S W S C

~

3 5 SP 88 SP 58 4 A B D C D B A 7 SA SS SS SP SA s so' D SP C SP C SP D s SA SP SS SS SA 10 A B D C D 8 A 11 SS SP S$ SP i

12 C B SP S C 13 SP SA SA id A D A 1.

l a,...

o' l

FUNCTIDN NUMBER OF CLUSTERS Control Bank D 8 -

Control Bank C 8 Connel Bank B 8 Conuol Sank A 8 Shutdown Bank S8 8 Shutdown Bank SA 8 SF (Spara Rod Locations) 18 l

l 3 1

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

, TICURE 2.3 BURNABLE A550RSER AND SOURCE ASSEMBLY LOCATIONS 1

R P N M L K J H O P E D C 8 A 1

, 2  ! 4 4 3 4 8 12 8 4 4 12 it se it 12 5 4 12 12 12 12 4

. . a l u it u .

7 4 to '12 12 16 4 8 so' 12 16 12 i 9 4 18 12 12 .

18 4 10 8 12 12 12 12 3 11 4 12 12 12 12 4 12 12 16 as 18 12 1

13 4 3 12 3 4 14 4 4 15 0'

1. 1. Nomber of WASAs Sommary of Inserts 808 WASAs in as Secondary Source m Cluem WABA Clusters 61 Control Rods 48 Thimble Plugs 46 Sec. Sources 2 Total 157 .

4

, T8GURE 3.4 BURNASLE ABSORBER AND SECONDARY SOURCE RCD CONFIGURATIONS

,0 0 O, ,0 0 O, O O O O O O O E O O O O O O O E E O O O O O O O O E O O E

O O O O O O 4 BA Configuration 4 SA Configuremen

_MS~.-

O O E E E O E O E B O E O E O E E O O E E D E O E O E E O E O E O O E E E O E . 5 O E 12 SA Configurttlen 16 BA Configuraten E B C O O O O O O O O O O O O O O O O Sotensory Seurte Aoes 5

3l0 CONTHOL ROD DROP TIME MASUREMENT (FNP-1-STP-ll2) l PURPOSE The purpose of this procedure was to measure the drop time of all full length contt ol rods under hot-full flow conditions in the reactor coolant t.ystem to insuru compliance with Technical Specification Requiremeuts.

$LMY OF RESULTS For the hot full-flow condition (Tavg 1 541 dog,F and all reactor coolant pumps operating) Technical Specification 3.1.3.4 requires that the drop time from the fully withdrawn position shall be f_ 2.2 seconds from the beginning of stationary gripper coil voltsge decay until dashpot entry. All full length rod drop times were measured to be less than 2.2 seconds. The longest drop time recorded was 1.838 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 '!O CONDITIONS DASHPOT ENTRY DASHPOT BOT'!OM Hot full-flow 1.620 see 2.176 see To confirn normal rod mechanism operation prior to conducting the rod drop test, the Verification of Hod control System Operability (FNP-0-EIN3643) was performed. In this test, the stepping waveforms of the at ationary, lift and movable gripper coils were examined, rod speed was measured, and the functioning of the Digital Rod Position Indicator (DRpI) and bank overlap unit was checked. All results were satisfactory.

4 i

6

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_ NORTH ITNIT 1 CYCI.E 10 .

900 k

1.65 1,61 1.61 m p

2.20 2,15 2.18

% 1.63 1.60 s 1 2.16 2.15 N f

1.58 -1.58 1.63 1.64 2.14 2.11 2.19 2,19 M-  !

l.59 1.60 (

2.12 2.17 1.71 1.56 1.60 1.55 1.62 1.63 1.66 2.29 2.07 2.22 2.13 2.26 E-2.16 2.19 ,

1.61 1.54 1.53 1.60 2.15 -

2.09 2.09 2.17- ~J ,

1 S 1.64 1.60 1.58 1.60 3 0o 180o 2.17 2.19 ,2.15 2.14 -N 1.66 1.53 1.57 1.63 2.22 2.10 2.13 2.18 -0' l.70 1.62 1.70 1.55 1.57 1.62 1.60 2.27 2.21- 2.22 2.12 2.20 2.16 2.16 F-1.58 1,64 "

2.12 2.17 [

a 1.62 1,63 1.57 1.62 2.21 2.17 2.13 2.18 0 1.62 1.64 I 2.14' 2.18 C 1.83 1.70 1.84 2.28 2.27 2.42 8 3

g 8

170 f

h 15 14 13 12 11 10 9 8 7 6 5 4 3- 2 i ORIVE LINE "0 ROP TlWE" T ABULATION .

547 TEMPERATURE - PRESSURE -

2232 gp(gy. 100 X. XX BREAKER "0PEMIMG" TO DASHP0T ENTRY - IN SECOMOS 11-6-89 -j DATE -

X.XX BREAKER "0PENING" TO DASMP0T 8OTTOM - IN SECOMOS

-7

4.0 INITIAL CR'TICALITY (PNP-1-ETP-3501)

PURPOSE The purpose of this procedure was to achieve initial criticality under carefully controlled conditions, establish the upper flux limit for the conduct of zero power physics tests, and operationally verify the calibration of the reactivity computer.

SUMMARY

OF RESULTS Initial Reactor Criticality for Cycle 10 was achieved during dilution mixing at 0441 hours0.0051 days <br />0.123 hours <br />7.291667e-4 weeks <br />1.678005e-4 months <br /> on November 8, 1989. During the approach to criticality, NIS source and intermediate range overlap data were taken to demonstrate that adequate cheNnel overlap existed.

The reactor was allowed to stabilize at the following ccr.ditions:

RCS Pressure 2235 psig RCS Temperature 5460F Intermediate Range Power 1 x 10-s Amp RCS Boron Concentration 1961 ppe Bank D Position 212 steps Following stabilization, testing was delayed for five hours by a ground loop problem in the reactivity computer AC power connection.

When the problem was resolved, the point of adding nuclear heat was determined and a checkout of the reactivity computer using positive l and negative flux periods was performed.

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

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

StP44ARY OF RESULTS l

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

ARO.HZPISOTHERMALANDMODhRATORTEMPERATURECOEFFICIENT Boron Measured ITC Design Acc. Calculated Rod Conc. ITC Criterion MTC Confiruration p.Is ,, pen /0 F pcm/0 F p.pe/o F .

All Rods Out 1985 -0.62 -0.5012 +1.75 8

t

.._.-._.y . , . - . . _ . m . .,..v.,m. - .r., -

l

. i

, ' where.  ;

I IN = Isothensal Temperature coefficient, includes -2.26 pas /or l Doppler coefficient MM = Moderator Temperature coefficient, normalised to the ARO j condition i No rod withdrawal limits were needed to maintain the moderator i temperature coefficient within the Technical Specification limit of l

+5.0 pcm/or. f ARO. HZP BORON ENDPOINT CONCENTRATION >

Rod configuration Measured Cs (ppa) P.ggirn-predicted Ca (ops) i All Rods Out 1988 2050 1 50 l The measured value is approximately 62 ppe below the predicted RZP boron endpoint, which exceeds the 50 ppe design review criterion given  ;

above. Westinghouse evaluated this discrepancy and concluded that the [

measured value falls within acceptable limits (see attached letter). ,

e 6.0 CONTROL AND SHUTDOWN BANK WORTR MEASUREMENTS (TNP-1-ETP-3601)

PURPOSE ,

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

1 SUPHARY OF RESULTS  !

The rod worth measurements were performed using the bank inter-change eethod in which: (1) the worth of the bank having the highest design w wth (des- rtd as the " Reference Bank") is carefully  ;

measured using th  ; idard dilution method; then (2) the worths of and shutdown banks are derived from the change-the remaining con,-

in the reference beu reactivity needed to offset full insertion of i

the bank being measured. .

The control and shutdown bank worth measurement results are given below. . The measured worths satisfied the review criteria both for the banks measured individually and for the combined worth of all the banks. t t'

Control or Predicted Bank -

Shutdown Worth & Review Measured Bank Percent Bank criteria (ocal Worth (pca) Difference i A 378 1 100- 392.4 3.82 1247.5*i 0.93 a B (Ref.) 1236 1 124 C 714 1 107 642.5 -10.02-L 9

1

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-m--..- --. --- . , . . . . r . , , , , ,- ...c y. .w.., w-.,., , ,_,--y,-.w...n *..v.,,-m.m.w,-y--r-.,,. --

St W ARY OF CONTROL AND SHUTDOWN BANK WORTHS (CONTINUED 1 Control or Predicted Bank Shutdown Worth & Review Measured Bank Percent 13nh criteris (oca) Worth (pes) Difference D 1005 1 151 951.2 -5.35 SD - A 1064 1 160 1085.6 2.03 SD - B 1020 1 153 961.4 -5.74 All Banks $417 1 541.7 5280.6 -2.52 l

• Measured by the dilution method 7.0 STARTUP AND POWER ASCENSION PROCEDURE (TNP-1-ETP-3605)

PURPOSE f

The purpose of this procedure was to provide controlling instructions .

fort <

1. NIS intennediate and power rante setpoint changes, as required prior to startup and during power ascension.

2. Ramp rate limitation and control rod movement recommendations.

3. Conduct of startup and power ascension testing, to includes

a. HZP physics tests (TNP-1-ETP-3601).

b. Incore movable detector system alignment (TNP-1-ETP-3606).

c. Incore-excore AFD channel recalibration (FNP-1-STP-121).

d. Core hot channel factor surveillance (TNP-1-STP-110). ,

e. Reactor coolant system flow measurement (FNP-1-STP-ll5.1).

StW ARY OF RESULTS In order to satisfy Technical Specification requirements for .

invoking special HZP physics test exceptions, preliminary trip set-  !

points of less than 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 804 to allow power escala-tion above 25% for calorimetric recalibration of the power range '

channels. (The 80% setpoint was used instead of 109% in case the . >

i uncalibrated power range channels were indicating non-conserva- ,

j tively.) At approximately 31% power, the power range channels were j recalibrated, the high-range trip setpoint was restored to 109%, and.

setpoint currents were determined for the intermediate range channels. 3 10 5

- - - . - , , - . - . , . . . , -. , -, - , .,,w r ~

i I i

The Westinghouse fuel warranty limits the power ramp rate to 34  ;

of full power per hour between 20% and 100% power until full power has been sustained for 72 cumulative hours out of any seven-day operating  :

period. This ramp rate was observed during the ascension to 100%

power.

Determination of incorn movable detector system core limit  ;

settings (FNP-1-ETP-3606) was accomplished during the ascension to 31%

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

(FNP-1-STP-121) which was perforwed in two phases (at 35% power and 100% power), and the reactor coolant flow test (FNP-1-STP-115.1) '

j performed at 100% power, which are described in Sections n.0 and 9.0 of this report. As summarized in Table 7.1 core hot chaeel factors were evaluated from the incore-excore full-core base case flux map taken under non-equilibrium xenon conditions at 354 power, and from the full-core flux maps performed at 33.5% and-100% power under equilibrium xenon conditions.

TABLE 7.1 StM1ARY OF POWER ASCENSION PLUX MAP DATA i

Parameter Map 238 Man 244 Man 245 Avg.

• power 34.5 33.5 100.1 Max FDH 1.5437 1.5357 1.4703 Max power tilt

• 1.0030 1.0030 1.0020 Avg. core % A.O. +9.977 +13.378 +1.715 Most limiting FQ(Z)" 2.1301 2.1859 1.8060 l;

, FQ Limit 4.5124 4.5124 2.2620 l Flux map Non-equilibrium Equilibrium Equilibrium l

conditions xenon xenon xenon

'Cale: 'ated power tilts based on assembly TDHN from all assemblies. -

Based on percent margin to FQ limit. ,

8.0 INCORE-EXCORE DETECTOR CALIBRATION (TNP-1-STP-121) i BIR_P_0,3 l The objective of this procedure was to determine 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 calibrating the delta flux penalty to the overtemperature delta-T protection system.

4 11

I

~

SLM MRY OF RESULTS 4

At an indicated power of approximately 354, a full-core base case flux map and five quarter-core flux maps were performed at various positive and negative axial offsets to develop equations relating detector current to core axial offset. To reduce error, all flux maps  !

were performed at essentially the same NCS temperature. The power range NIS channels were adjusted to incorporate the revised calibra-tion data.

' At 100% power, a full-core flux map (No. 245) was performed to verify that the core peaking factors were satisfactory and to revise the detector equation currents for 100% power, sero percent axial offset (the "l-zero current /'). This was done primarily to correct <

for the effects of the change in programmed Tavg which occurs as power is increased from 35% to 100% power. The resulting final equations for NIS Channels N41 - N44 are given in Table 8.1. The power range NIS channels were adjusted to incorporate the revised calibration.

TABLE B.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS

, OBTAIN FROM INCORE-EXCORE CALIBRATION TEST CHANNEL, N41:

I-Top 0.7050*A0 + 156.65 uA I-Bottom = -0.9551*A0 + 151.66 uA CilAVNRL N42:

I-Top = 0.7219*AO 4 148.07 uA ,

I-Bottom = -0.9203*AO + 146.45 uA CHANNEL N43 1-Top = 0.7283*A0 + 156.34 uA I-Bottom = -1.01218A0 + 161.83 uA J

CHANNEL N44:

I-Top = 0.7522*A0 + 153.44 uA I-Bottom = -0.9792*A0 + 150.36 OA 9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-1-STP-115.1)

PtrRPOSE The purpose of this procedure was to measure the flow rate in I cach reactor coolant loop in order to confirm that the total core flow.

l met the minimum flow requirement given in the Unit 1 Technical Specifications..

12

. . . _ _ . ~.,_.__...r , . , . . ___m,,., .m._., _ . , - , . . . . . -

\*

St> MARY OF EstrLTS To comply with the Unit 1 Technical Specification, the total reactor coolant system flow rate measured at normal operating temperature and pressure must equal or exceed 265,500 gym for three loop operation. From the average of six calorimetric beat balance measurements, the total core flow was determined to be 279,703 spe, which meets the above criterion.

i 13

' ---m____,,____ _ _ , ,. , _ _ _ _, ._ ___

[h 3q Westingficuse Commercial Nucleat av m Electric Corporation Fuel DMslon " sees *

• v e its n o m November 8, 1989 89AP*.G.0064 Mr. L.K. Mathews, Manager Nuclear Fuel Services Southern Company Services, Inc.

P.O. Box 1295 Birmingham, AL 35201

Dear Mr. Mathews:

JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 EVALUATION OF BORON ENDPOINT MISPREDICTION Based on discussions with Alabama Power personnel at Farley Unit 1, we have been cade aware of the discrepancy between measured and predicted critical boren concentration (at HIP, all rods out) of minus 62 ppm (negative 443 pcm l ThisdiscrepancyexceedstheWestinghousedesignreviewcriteriaofplusor).

minus 50 ppm, but falls within the acceptance criteria of plus or thinus one cercent (1%) delta-k/k (plus or minus 1000 pcm), as specified in Westinghouse report WCAP 9648.

Westinghouse established the review criteria as a point of notification; therefore, the discrepancy in itself has no safety significance, and does not preclude continuation of power ascension. Westinghouse has evaluated this i

situation specifically for the Farley Unit 1 Cycle 10 design and concludes that, contingent upon meeting the review criteria for rod worth measurements (i.e., Red Swap), Farley Unit 1 can proceed with power ascension.

David E. McKinnon 7 ::, h Fuel Project Engineer-

/kph cc: M.M. Fiedler -

W.H. Andrews R.A. Hommerson M.D. Rickels

. Morty

. Nesbitt

.H. Marlow IN MsM9 house Comme @llht@kfyt/0mt&1 - Wnwofthe 1.WMot0en BMtye Nomulk%ty Angt ,

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