ML20006E601

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Alabama Power Co Joseph M Farley Nuclear Power Plant Unit 1,Cycle 10 Startup Test Rept. W/900208 Ltr
ML20006E601
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 02/08/1990
From: Hairston W
ALABAMA POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9002260111
Download: ML20006E601 (16)
v • d • e


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  • Alabima Power Company' p ' 4

-~, . 40 inverness Cont:t Parkycy  ;

tb 'l l ~ -l.. . Post Office Box 1295 -

i J' . Birmingham, Alabama 35201

.,,. ' 7blephone 205 868 5581 m

t W. G. Hairston lll - 1 Senior Vice President Nuclear Operations : AlabamaPower the southern electrc system )

February 8,1990

- - . 1 Docket No. 50-348--

u-U.' S. Nuclear Regulatory Commission

~ Attention: Document Control Desk -

Washington, D.'C. 20555 ,

4 Gentlemen Joseph M. Farley Nuclear Plant

' Unit 1 Cycle 10'- Startup Report [

Enclosed is-the.Startup Report for Unit 1-Cycle 10 as referenced in our Cycle 10 Reload letter dated September 27, 1989. s If you have'any questions, please advise.

Respectfully submitted, n

5).h $ %- i V. G. Hairston, III

.i VGH,IIIsemb/NEL-26.28 Enclosure cci Mr. S.'D. Ebneter Mr. E. A. Reeves #

Mr. G.-F. Maxwell

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. n 9002260111 900208

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ALABAMA R)WER OCNPANY.

I JOSEPH M. FARLEY NUCLEAR PLANT s UNIT NUMBER I CYCLE 10  !

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STARTUP TEST REFORT '5 PREPARED BY THE PI M r REACIOR ENGINEERING GROUP I i

I TABLE OF CONTENTS PAGE. j o

1.0. Introduction 1 2.0 Unit.2 Cycle.7 Core Refueling 1 3.0 Control Rod Drop Time Measurement 6 4.0 Initial Criticality 8 5.0 All-Rods-Out Isothennal Temperature 8

-Coefficient and Boron Endpoint 6.0 Control and Shutdown Bank Worth 9 Measurements y 7.0 Startup.and Power Ascension Procedure 10-8.0 Incore-Excore Detector Calibratirn 11 9.0 Reactor Coolant System Flow l 12 Measurement I

l l APPROVED: . 1 ls l'

i E ;> 0O Technical Manager /2 -/F / 7*)

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u General Manager - Nuclear Plant l

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j , [1.0 INIBODUCr10N-q he Joseph M; Farley Unit 1 Cycle 10 Startup Test Report addmsses 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 %ree Loop I Westinghouse pressurized water reactor rated at 2652 MWth. n e Unit began conmdreial operations on December 1,.1977. The Cycle 10 core loading consists of 157 17 x 17 fuel assemblies.

Previous Cycle Completion Dates and Average Burnups Date. Start of EOL EOL Burnup EOL Burnup ' Total Cycle . Critical Cycle Date (MWD /MIU) (EFPD) EFPY 4

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 5,180 140.70 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 12,258 333.58 4.825 7 05/26/25' 05/27/85 10/03/86 17,231 470.04 6.112 8 -11/30/86 12/02/86 03/25/88 16,190 443.26 7.327 9 05/20/88 05/21/88 09/23/89 17,456 479.29 8.640

~

2.0 UNIT'l CYCLE 10 00RE REFbT11NG' 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 light,'small mottled crud pattern on several assemblies, but no  :

1- significant fuel damage or deformation was noted. W erefore, no fd changes to the design Cycle 10 core loading pattern were required.

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

H e as-loaded Cycle 10 core is shown in Figures 2.1 through 2.4, which give the location of each fuel assembly and insert, including wet annular burnable absorber insert locations and configurations.

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

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, FIGURE 2.1: UNIT 1 CYCLE 10' REFERENCE LOADING PATTERN' R P [dN -M L K J H G F E D C 3 A i.E .l Mi.t 48' 47D - 241 ~ ~1  !

2A48 09 2A47 l 32 a29 av6es am amte a21 73 - -2 F14 2A55 2848 2A43 20$1 2A63 K22 1 M 03 l 15 4W63D OW0do a42 a32 ' M SO 4W68 99 ~ ~ '

OS 2353 2005 2A21 2816 2A23 mit IB59 E26 age a45 - 1 M os att t u lo ssos i e to a39 12wt20 att sm ~ ~4 sol IAlt 2332 2AM 2334 2A17 2002 2MO 2942 2A52 E03 19 4WIm IMS a13 12w121 m 19 S6 13v119 nee 12w129 4W59 3 - -5 F44 - 2300 2008 2A22 Is27 2A31 2AM 2Act 2B20 2A38 asas It47 F45 als sW0sp als ' 12W13e ses 1 M 12 att 1Mtf e43 1 M 06 as Ma a16' -

-6 2A67 2825 2400 ISIS NOS 2018 G24 2801 mar 2831 2A37 3 14 2A42 l Iso au6e ass teuf7D 37 12w124 a17 60 a15 1 M 11 a 1wm a06 4u400 18 7 f 2AS6 2ndt 2A04 3 10 2A20 2839 2A14 2A10 2Atl 3s37 2A24 3830 2A11 2849 2A45 em as 12v102 r1D FID - azs se seuseo as a35 as In 12w116 a22 28 8

ris 2Aa0 taas IA3s tus s29 2A03 2:29 2A12 s01 2AS7 2A29 ISM 2A49 KS6 R

96 Austo a09 tem 610 12v104 nos 74 a14 1 M 22 4 two a07 4W3so 40 9

i 2AS9 20$6 2AiG 3B36 2A19 2009 2A06 2A25 2A07 2841 2A30 2007 2A02 2s57 2A64  ;

a64 ewis a26 12w115 a04 1 M 23 a25 IIv105 ene 1 M 13 ast swito e41 10- 1 2 Ass 3 17 Ist less uC4 asta c20 Issa nos a40 2A3s Is44 Ims 3 it 4W550 12w110 m37 1 M 27 11 eae . 50 1 M 14 a01 1 M 09 au6a e 11 F39 3 54 2004 2A05 2843 2A01 2A50 2A27 2s05 2A13 2012 20$5 K34 75 a47 12wtof a27 t w as ss06 towfoo at0 12v126 ass 5'm 12 K04 2A46 2006 2A34 2013 2A36 2821 2A32 2B20 2A41 G7 480 4W540 GWOW a05 1 M 14 a24 SWD70 4W57D M 13 at asso mit 2A33 2533 2A16 5 24 3 52 E09 31 act su6so ato ame ase is 14 L K43 tut Is46 2M4 Issa Im2 #10 o _ .;

1 L 36 100 620 15 2A53 u24 2A34 1

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l o 0 90 - 279 NETu l 0 \ /

e v 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 8A (H) assemblies ... 2.999% Region 8A .......... 6 Region 10 (K) assemblies ... 3.597% Region 10 ......... 13 Region 11A (2A) assemblies .. 3.805% Region 11A ........ 40 Region 11B (2A) assemblies .. 4.207% Region 11B ........ 28 ,

Region 12A (2B) aesemblies .. 3.803% Region 12A ........ 45 Region 12B (2B) assemblies . . 4.200% Region 12B ......... 16 157 2

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

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FIGURE 2.8 c,n, .

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CONTROL ROD LOCATIONS

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R: P N M L. K J H G F S D. C B. A t

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'8 SA SA SP-i 4 C B W S C

5 SP SB SP 88 I,

3 A 5 D C D 8 A r

SA

' 7.- SS SS SP SA S so' D

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SP C SP C SP D 8 S. SP . S. S. ,

6-10 - A B D C D B A

'i 11 - SB SP' SS SP E 12 - C B SP B C i

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13 SP SA SA 7 14 A D' A l

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Abe ggial:

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L PUNCTION NUMBER OF CLUSTERS Control Bank D 8 Control Bank C 8 Control Bank 3 8 Control Bank A 8 l- Shutdown Bank SB 8 Shutdown Bank SA 8 SP (Spare Rod Locations) 18 3

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  • '4' SURNASLE- ABSORBER AND SOURCE' ASSEMBLY LOCATIONS.

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R P- N' 'M L. K J H G F E D- C 8 A L ,

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4 12 18 ss 18 12 l

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-6 8 12 12 12 12 8 7 4 18 - '12 .12 : '18 4-US so' 12 18 12 1 J9 4- 18 12 12 -14 4 10 ' 8 12 12 12 .12 8

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J 12 ' 12 ' 18 as 18- '12 l

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    1. - Number of WABAs Summary of Inserts 808 WABAs in

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WABA Clusters 61 81 Cluem l Control Rods 48 l

'7 nimble Plugs 46 Sec. Sources 2  !

Total 157 4 l

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- naar 2.4 1

suRNAaLa Ass 0ResR ANo sec0NDARY s0URCE RCD CONPl0URADONS I

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(. l B B B B C O D D D D D E O O O O O O O E 5 O O O O O O O O E O O  !

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,0 0 O, O O O O O O O O O O O O O O O O O Seconeary Sowse Rees 5

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3.0 . 00NmOL IOD Diop TIME MEASUREMENT (FNP-1-STp-112) l ISRFOSE ,

i ne purpose of this procedure was to neasure the drop time of  ;

all full length control rods under hot-full flow oorditions in the  !

L mactor coolant system to insum compliance with Technical {

Specifloation Requirenents.  ;

Sttt%RY OF RESULTS ,

j For the hot full-flow cordition (Tava jt 541 deg.F ard all I reactor coolant peps operatirW) Technical Specification 3.1.3.4 requires that the drop time from the fully withdrawn position shall te 12.2 seconds from the beginning of stationary gripper ooil voltage decay until dashpot entry. All full length rud drop times were measured to be less than 2.2 secords. %e longest drop time  !

recortled was 1.838 seconds for rod B-6. %e rod drop time results  !

for both dashpot entry and dashpot bottom are presented in Figum '

3.1. Mean drop times are sunziarized below:  ;

TEST MEAN TIME 'IO HEAN TIME TO ,

00NDI'lONS PASW Ur E m V IW95Ur EMU 1 Hot full-flow 1.620 sec 2.170 sec To confinn noaml red na:heism o;wint.ior, price to conductir* the tsd dmp test, the Verificatiaa of Rod 03nt ol Systen Oparsbility ,

ifKP-0-f?!P-3043) was perforwd. In this tect, the steppird we.veforms of the otacionary, lift a r1 novable gripper calls were cumined, tyd

  • speed was ressured, and the functioning of the D!gital pod 1%2tien  :

Irdicator (DRPI) ard tuA overlap unit was olmekeA. A1.1 results were satisfactory. ,

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NotT,N L' NIT 1 CYCLE 10  ;

900 i t i 1.65 2.20 1.61 2,15 1.61 2.18

\Y ' __p 1.63 1.60 s  !

2.16 2.15 N

[ g 1.58 'l.58 1.63 1.64 2.14 2.11 2.19 2.19- M i

1.59 1.60 2.12 2.17

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1.71 1.56 1.60 1.55 1.62 1.63 1.66 K E

2.29 2.07 2.22 2.13 2.26 2.16 2.19 [

1,61 1.54 1.53 1.60 2.15 2.09 2.09 2.17 ~J 1.64 1.60 1.58 l.60 o

, 0o 2.17 100 -- H 2.19 l 2.15 2.14

, 1.66 l 1.53 1.57 1.6, i j 2.22 2.1C i 2.13 2.18 ~G

! 1. 7'O t.62 1.70 .

l.55 ~" 1.$7 1.b? 1760 l i <2.27' 2.21 2.22 2.12 2.00 2.16 L 16 F L i L ,;,,

=

?.12 2.17 --

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I 1.62 1.63' 1.57 1.62 l

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2.21 2.J 7! 2.13  :'2.13 , l 1.62 1.64 I  !

2.14 2.18 C 1.83 1.70 1.84 '

2.28 2.27 2.42 I ,

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> g i 270' [

l 15 14 13 12 il 10 9 8 7 6 5 16 3 2 1 DRIVE LINE "0 ROP TlWE" TABULATION 547 2232 100 TEMPERATURE - PRES $URE -  % FLOW -

X.XX BREAKER "0PENING" TO DASHPoi ENTRY - IN SECONOS 11-6-89 DATE -

X.XX BREAKER "0PENING" TO DASHPOT BOTTON - IN SECONOS 7

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, 4.0 INITIAL CRITICALITY (FNp-1-ETp-3601)

RRKEE We purpose of this procedum was to achieve initini criticality j_ under carefully controlled corditions, establish the upper flux limit for the corduct of zero yower physica tests, and operationally verify i the calibration of the reactivity computer. '

SLRt%RY 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 intemediate range overlap data were taken to demonstrate that adequate channel overlap existed.

he reactor was allowed to stabilize at the following corditions:

RCS Pressure 2235 psig  ;

RCS Temperatune 546*F  !

Internediate Range power 1 x 10-

  • Amp  !

RCS Boron Concentration 1961 ppm Bank D Position 212 steps ,

Following stabilization, testing was delsfed for iive how a by a i ground loop problem in the reactivity ocanputer AC pcwer e cnnection.

When the problem was renolved, the point of adding nuclear heat, uns detennined ard a chechout of t.he sv:actitity corapoter usire pn8Ativo .

and negative flux }erirais was ptfermed.  !

5.0 All-RODS-OLTT ISCTDIERMAL TDIPFJMSEE OC/5FICIEMT AND FiWG IETOINT (RTP-1-ETp-3601)

  • 11LR3g f he objectives of these sensurerents were to deternine the hot, ,

zero power isothermal and moderator temperature coefficients for the all-rods-out ( ARO) configuratiori and to noasu'e the ARO boron endpoint concentration.

St#tRRY OF RESULTS i

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

l ARO. HZP IS0!HEFPML AND IODERA'KR 7DIPERARRE 00 EFFICIENT

,. Boron Measured 170 Design Acc. Calculated l Rod Conc. I'IC Criterion MIC j Confiruration _ pra _ pom/8 F pam/0 F pom/* F All Rods Out 1985 -0.62 -0.50 1 2 +1.75 l

8 l

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, .oere:

I'IC = Isothernal Temperatum Coefficient, incitdes -2.26 pom/* F Doppler coefficient 1110 : Moderator Temperature Coefficient, norunlized to the AHO oordition No rod widdrawal limits were needed to maintain the moderator temperature coefficient within the Technical Specification limit of

+5.0 pom/0F.

ARO. HZP BORON ENDFOItif CONCENTRATION Rod Configuration Measured Ca (mm) Design-predicted Ca (un)

All Rods Out 1988 2050 1 50 The measured value is approximately 62 pyn below the predicted HZP toron endpoint, which exceeds the 50 pyn design review criterion given above. Westinghouse evaluated this discrepancy and concluded that the measured value falls within acceptable limits (see attached letter) .

6.0 00NTf0L AND SWfDOWN BANK WOR'IH MEASURD4ENTS (FNP-1-EIT.-3601)

BM553E The objective of the inci worth neasurenents was to deture.iine the integrul reactivity war th of esch control md shutdoan bank f x comparj aon with the vales pred.cted by design.

NJ.cE1. 9f RFME

'It c rw unr t.n acts.ren.nt.s ec perfenned uu ng the bank int.er-change method in which: '4) the vorth of the bank having the highest design worth (uesignated as the "Refemnoe bank") ic carefu'ly measured usir.g the stands.rd dilution mJie d theri (2) the worths of the remaining control ard shutdown banks are derived from the change in the reference bank reactivity needed to offset full insertion of the bank being measured.

'Ihe control and shutdown bar.k 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.

Control or Predicted Bank Shutdown Worth & Review Measured Bank Percent Bank Criteria (nom) Worth (nom) Difference A 378 1 100 337.7 -10.7 B (Ref.) 1252 1 125 1226.7* -2.0 C 749 1 112 728.7 -2.7 9

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Stit1ARY OF 00N7HOL AND SH(TITXhW BANK W1RTHS (00NTINUED)

Control or Predicted Bank Shutdown Worth k Review Heasured Bank Percent Bank Criteria (rom) Worth (txw) Difference  !

i 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 All Banks ~5594 1 559.4 5429.8 -2.94 .

  • Measured by the dilution method  ?

7.0 STAR 1UP AND POWER ASCENSION PROCEDURE (FNP-1 +OP-3605)

PURK1SE  ;

The purpose of this procedure was to provide controlling instructions for:

1. NIS intermediate and power range setpoint darvtes, as required ,

prior to otArtup arxi during power ascension.

2. Rup rate limitation and control rod movernent reconnesxiaMocsd. l
3. Conduct of startup and power ascensson testing, to include:  ;

n" .

a. HZP physics tests (FNP-1-ETP-36011 s
b. Incore movable detector cystae n11grunent (FNP-1-ETP-5fo6). l l
c. Incere-execie AFD channel reanlibsstion (FNP-2-:7t?-121).  ;
d. Core hot channel factor surveillance ' IMP-1-STP-110). <
e. Reactor coolant system flow measurement (FWP-1-STP-115.1).

l SUtt1ARY OF RESULTS l

In ortler to antit Te mnical Specification requirements for invoking sper:ial HZP pr,' sics 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 80% to allow power escala-p tion above 25% for calorimetric recalibration of the power range L channels. (1he 80% setpoint was used instead of 109% in case the l uncalibrated power range channels were indicating non-conserva-tively.) At approximately 31% power, the power range channels were recalibrated, the high-range trip setpoint was restored to 109%, arri setpoint currents were determined for the intermediate range channels.

10 i

A

- he Westirabouse fuel warranty limits the Tower ramp rate to 3%

of full power yer hour between 20% ard 100% power until full power has been sustained for 72 cumulative hours out of any seven-day operatirW period. his ramp rate was observed during the ascension to 100%

yower.

Deterinitation of incore novable detector system core limit settirwa (h'P-1-ETP-3006) was accomplished during the ascension to 31%

power. his was followed by the iroore-exoore recalibration test (INP-1-STP-121) which was performed in two phases (at 35% yower and 100% yower), ard the reactor coolant flow test (INP-1-STP-115.1) performed at 100% yower, which are described in Sections 8.0 ard 9.0 cf this re; ort. As stasaarized in Table 7.1, core hot channel factors were evaluated irom the incom-exoore full-core base case flux map taken urder non-equilibritsu xenon oorditions at 35% power, ard frxm the full-core flux maps perforined at 33.5% and 100% power under equilibrium xenon corditions.

TABLE 7.1 St> MARY OF IWFR ASCENSION F1UX MAP DATA e

!! Parameter Man 238 tho 244 thn 245 Avg. % Iower 34.5 33.5 100.1 h u; iT4! 1.5437 1.5357 1.4703 Mw rower tilts 1.0030 1.0030 1i0020 Avg. core % A.O. +9.977 +13.378 +1.715 Mast limitin,t 19tZ)* t 2-1301 2.1859 1.8441 yt Linut 4.5121 4.5124 2.3026 Flux faap Non-equGib. lima Equilibritan Equilibrium conditions xenon xenon xenon

' Calculated power tilts based on assembly FDHN from all assemblies.

Based on percent nargin to 19 limit.

8.0 IN00RE-EX00RE DtmwivR CALIBRATION (IHP-1-STP-121)

IURTME no objective of this procedure was to detemine the relationship between power range upper and lower excore detector currents ard axial offset for t.he purpose of calibrating the control boarti and the plant computer axial flux difference (AFD) channels, and for calibrating the delta flux penalty to the overtemperature delta-T protection system.

11

1

~.

Sttt%RY OF RESULTS At an irxiioated power of approximately 35%, a full-core base case flux map and five quarter-core flux maps were Ierformed 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 amme RCS temperature. %e power range NIS channels were adjusted to incorporate the revised calibra-tion data.

At 100% power, a full-core flux amp (No. 245) was performed to verify that the core peaking factors were antisfactory and to revise the detector equation currents for 100% power, zero percent axial offset (the "I-zero currents"). h is 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. he power range NIS channels were adjusted to incorporate the revised calibration.

TABLE 8.1 DlaurmR CURRENT VERSUS AXIAL OFFSET EQUATIONS

w. ODTAIN FROM INOORE-EXOORE CALIBRATION TEST DIANNEL N41:

I-Top = 0.7050 AO + 156.55 uA I-Bottne : -0.9551*AO + bH 6S uA plA%'EL N42:

1 Top = 0.72196AO + 148.07 uA I-Bottom : -0.9293*AO + 146.45 uA DIA%.,_Th.E121 9- I-Top = 0.7283*AO + 156.34 uA I Ik,ttom = -1.0121 TAO + 161.83 uA GIANNEL N44:

I-Top = 0.7522sAO + 153.44 uA I-Bottom = -0.9792*AO + 150.36 uA

9.0 REACIOR 000LANT SYSTIN FIDW MEASURDEC (FNP-1-STP-115.1)

PURPOSE The purpose of this procedure was to measure the flow rate in each reactor coolant loop in oixier to confirm that the total core flow met the minimum flow requirement given in the Unit 1 Technical Specifications.

12

_ _ _ _- __ __ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ - _ - - _ _ _ _ _ - __ _

b-

. Sth MRY OF RESULTS L 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 rpn for three loop operstion. From the average of six calorimetric heat balance measurements, the total core flow was determined to be 279,703 apn, which meets the above criterion.

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November 8, 1989 89AP* G.0064  ;

Mr. L.K. Mathews, Manager  !

Nuclear Fuel Services Southern Company Services, Inc. '

P.O. Box 1296 ,

Sirmingham, AL 35201

Dear Mr. Mathews:

j ,

JO$tPH M. FARLEY NUCLEAR PLANT UNIT 1

! VALUATION OF BORON ENDPOINT MI$ PREDICTION .;

L Based on discussions with Alabama Power personnel at Farley Unit 1, we have been made aware of the discrepancy between measured and predicted c"itical boren concentration (at HIP, all rods out) of minus 62 ppm (negative 443 pcm).

This discrepancy exceeds the Westinghouse design review criterte of plus or

~,

minus 50 ppm, but falls within the acceptance criteria of plus or minus one percent (M) delta k/k (Pus or minus 1000 pcm), as specified in Westinghouse report WCAP 9648; w .

' Westinghouse established the review criteria ss a point tf notification' therefore, the discrepancy in itself hss no safety significance, and dess not i

?

preclude continuation of power ascension. Westinghouse has evaluated this situation specifically for the Farley Unit 1 Cycle 10 detipn and concludes

.that contingent upon meeting the review criteria-for rod vorth measurei. ants t ,

i,

'{i.e., Rod Sway), l'arley Unit 1 can proceed with power ascension .

L '

L Very uk y ,

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David E. McKinno v ,

Fuel Pro. ject Engineer

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1 cc: M.M. Fiedler '

W.M. Andrews i

R.A. Hommerson -

M.D. Rickels

. Morey

. Nesbitt

  1. n.H. Marlow

.=

o neranywar Ammme/Ateev4rthsk - Mrwe'er imMontn Se%pr Newe'AnwAnet 3 .

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