ML20082G933

Jm Farley Nuclear Power Plant Unit 1,Cycle 11,Startup Test Rept

ML20082G933
Person / Time
Site:	Farley
Issue date:	08/12/1991
From:	Woodard J ALABAMA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9108220119
Download: ML20082G933 (16)

Text

_. _ _ _ . _ _ . . . _ _ . . - . . . _ _ _ . _ _ _ . _ _ _ . _ . _ . _ . . - . . _ _ _ _ . _ _ _ _ . . . . . _ _ . _ . _ _ _ . _ . _ _ _ .

tu w n.,ro w c m.m oy m imum ccnw em,y O .ist O% e Do. !7)f; O munghxn AWtuu 3%'01 beth.'ne , Utn f#- dt $O *

.m J. D. Woodard vu,onmw-%3w A!O!)illlhllk)WCf

1. arne n,m August 12. 1991 v w e'" " e * " " * "

4

, Docket No. 50-348 U. S. Nuclear Regulatory Commission' Attention: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant-Unit 1 Cycle 11 - Startuo Report Gentlemen:

Enclosed is the.Startup Report for Unit 1 Cycle ll as referenced in our'-

Cycle 11 Reload letter dated April 8,1991.

If you have any questions, please advise.

Respectfully submitted,-

hk W Q. Woodard JDW/MDR:maf5010-Enclosure; cc: Mr.1S.-D.-Ebneter L Mr. S. T. Hoffman l?

Mr. G. F. Maxwell -

l-l-

h 9108220119 910812 PDR- ADOCK:05000348 P -- .pnq_

_ . -. , .._m m . .= .m _ . __ - _ _ _ . //#

l ALABAMA POWER: COMPANY .- i JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER-1 CYCLE 11 STARTUP TEST REPORT l

PREPARED BY THE PLANT REACTOR ENGINEERING GROUP TABLE OF CONTENTS PM[

1.0 Introduction .

1 2.0 Unit 1 Cycle 11-Core Refueling _ l' 30 Control Rod Drop Time Measurement

-7 4.0 Initial Criticality .

. 9 5.0 , All-Rods-Out Isothermal Temperature. 9 Coefficient and Boron Endpoint 6.0 Control-and Shutdown Bank Worth 10 Measurements

,7. 0 Startup and Power Ascension. Procedure 11

-8.0 Incore-Excore Detector. Calibration 13:

9.0, Reactor Coolant System Flow 14 Measurement L

APPROVED:

l b ,, 4 h i ., -

y

Technics 1

Manager

~

-General Manager -: Nuclear P1 ant =

MM/STARTRPT!

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

1.0 INTRODUCTION

The Joseph M. Farley Unit 1 Cycle-ll 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 Final Safety Analysis Report (FSAR).

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 11 core loading consists of 15717x17 fuel assemblies.

Previous Cycle Completion Dates and Averaae Burnuos Date Start of E0L E0L Burnup E0L Burnup Total

,(ycle c Critical Cycle DALq (MWD /MTV) (EFPD) EEP_Y 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 12,238 333.58 4.825 7 05/26/85 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 10 11/08/89 11/10/89 03/08/91 16,910 464.17 9.911 2.0 UNIT 1 CYCLE 10 CORE REFUELING REFERENCES

1. Westinghouse Refueling-Procedure FP-ALA-RIO

2. Westinghouse WCAP 12869- (The Nuclear Design and Core Management of the Joseph M. Farley Unit 1- Power Plant Cycle 11)

Unloading of the Cycle 10 core into the spent fuel pool commenced on 03-22-91 and was completed cn 03-24-91. _ During core unload, each fuel assembly was visually inspected.with binoculars. No unusual fuel damage or deformation was noted.

During the period from 3-24 to '3-26-91,: a series of eddy- current

. control- rod inspections was carried out, as part of ~ an FNP program to

• evaluate-and reduce RCCA wear. -For all but six RCCAs,-the wear was 4

minor:;1ess than' 20% clad ' cross-sectional area removed. Of--the six RCCAs which had noticable wear, the' wear scar ~ for~ RCCA' R26, based on -

1 4

1. g- ,, v-y -ye p+y-* w y -

<a,-yW e- ,w

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

encircling coil (raw) data, was 51% cross-sectional area removed, which exceeded the (50%) wear acceptance criterion for RCCA reuse.

However, subsequent, in-depth analysis determined that, although the wear scar probably had penetrated the cladding, the cross-sectional wear was slightly less than 50%. Therefore, it was recommended that

, R26 could be reused provided that it was axially repositioned so that the wear scar was no longer aligned with an RCCA guide card. The five remaining RCCAs showed noticable wear, but were acceptable for reuse.

These were R46 (37% w ar), R24 (35% wear), R30 (23% wear), R18 (22%

wear), and R16 (22% wear). RCCA tip cracking was observed on most '

RCCAs, but, in all cases, was within 12 inches of the lower tips of  ;

the rods and, therefore, was deemed acceptable.

Prior to the end of Cycle 10, ultrasonic fuel leak testing had not been planned because there had been no indications of fuel assembly leakage. However, during the final ramp-down in power for the refueling outage, radiocFemistry analysis detected an iodine activity peak in the reactor coolant, indicating a probable leaking fuel-assembly. Therefore, arrangements were hurriedly made and fuel inspection equipment arrived on site while the RCCA inspections were being conducted. The equipment setup in the spent fuel pool was-expedited by the discovery that the large underwater manipulator device could be disassembled and brought in through the new fuel hatch, rather than having to lift the spent fuel pool roof slab to bring in the fully. assembled unit.

The fuel assembly ultrasonic leak testing (UT) was conducted from 3-27-91 through 3-30-91, following completion of_ the RCCA inspections. Bccause of the possibility that the leaker might have been a fuel assembly designated for-core reload, the reload assemblies were tested first in-order to expedite the core redesign. Two-adjacent fuel rods, E2 and F2, of reload' assembly 2A57 were identified as the leakers. The leak was characteristic of a debris-induced failure.- However, no debris could be found by_ visual TV inspection, probably as a' result of the interior position of- the damaged rods.

Fuel assembly. 2A57 was replaced with an identical assembly, 2A06, in the redesigned core.

Core reload for Cycle 11 commenced on'4-19-91 and was completed' on 4-21-91. The as-loaded, redesigned Cycle Ilicore is shownzin Figures 2.1. through 2.4, which give the location of each fuel assembly and insert, including wet annular burnable absorber insert locations and

'Jurations. The Cycle 11. core has a nominal design burnup-

. U111ty of 17,450 MWD /MTU.

I 1

c -

2 i

u, ,_ _ _. __ . , _ . - -

. Pigurn 2.1 : Unit 1 Cycle 11 Reference ImdirF] Pat.lfrn A B C D [ F G H J l L M N P R

--  :- l ,

I l l 47p g . . . . _

e.

75 }$

2A55 3 35 2A63 450 R17 39 R45 20 R08 47 -- l 4 2A51 3 38 2060 3 32 2046 3 22 2M9 660 W17D 1W156D '

R13 4 IW1500 R46 IW147D W190 360

- - 13 2A50 2057 2043 3 17 2028 3 44 2C36 2062 2A03 74 R41 , 19114D R09 191100 SS06 19109D R06 led R22 720 12 2A25 2014 2025 3 56 2011 3 12 M4 3 57 2C30 2006 2M5 1

100 $180 191260 R20 IWil3D 620 40 15  !$1180 R19 IW12SD W140 12 -

11 2M3 2C38 2026 2A56 2017 3 36 3 59 3 07 2020 2M7 2C32 2058 2A52 R43 131460 R32 IW116D R18 191320 R39  !$1290 R29 191020 R42 I N1490 R48 10 3 41 2049 3 51 2021 2M5 2C52 3 16 2064 2A54 2027 3 40 2053 3 09 690 5 R05 IW122D 66 19133D R40 $60 R37 19123D 250 19127D R12 1 3 9 2A67 2059 3 19 2029 3 02 2C61 3 40 3 54 3 43 2C39 3 34 M7 3 03 2040 2M2 32 R47 IN151D 150 17D R23 190 300 18 R34 73 241 IN1550 R44 65 A 3 05 3 04 2008 3 42 3 60 2B45 3 52 3 29 3 53 3 26 3 55 3 06 2C22 3 23 3 27 26 38 til IW100D 60 IW1050 R01 610 R04 19131D 36 !W121D R10 17 56 4 7

2A58 2054 3 24 2001 3 21 2033 3 20 3 47 3 15 2044 3 13 2012 3 11 2041 2M8' R35 13153P R27 !Wil7D R36 191300 R25 191280 R30 IW1190 R24 . 131540 R28 --6 3I7

' _ .3 58 M9 2M8 2048 2047 3 33 2051 2A59 2016 3 46 2C50 3 39 50 0K 9 IW103D R21 191040 37  % 99 191060 R33 191240 W150 4 -- 5 2M6 2C37 2015 2A53 2019 3 10 3 50 3 30 2018 2E4 2005 2C63 2M4 80 R03 IWil50 R38 19107D SS05 IWillD A07 191 2 R16 60 4 2A06 2031 20W 3 61 2024 3 08 2(23 3 49 2013 2010 2A10 230 W16D IW152D R26 IN1480 R31 IW157D W13D IW ---~ 3 2Al2 2034 2C56 3 25 M3 3 14 2055 2045 2A66 650 R15 77 M2 70 R14 29D 2 2MO 3 01 2042 3 20 2C35 3 10 2M1 11 26 31 1 2M1 3 31 2A62

(~ Fla ASSOSLY INTRI SERIAL 28ER - Itrth

(- F00. AS0BLY SERIAL MSER The original v/o U-235 enrichments were No. of fuel assemblies:

Region 11 A (2A) assemblies... 3.805% Region 11A......... 5 Region 11B (2A) assemblies. . . 4.207X - Region.11D........ 27 Region 12A-(2n) assemblies... 3.803% Region 12A........ 45 Region 12B (2B) assemblies... 4.19'fX . Region 12B........ 16 Region 13A (2C) assemblies... 3.801% Region 13A........ 32 Region 13B (2C) assemblies... 4.195% Reqion 138........ 32 Total............ 157 3-

Figura 2.2 : Control cnd Shutdown Rod Locations R P N M L K J H G F D E C 3 A 1

2 A D A 3 SA SA SP 4 C 8 SP S C 5 SP S3 SP Se e A 2 D C D B- A 7 SA 38 SB SP SA 8 so- D SP C SP C SP O 9 SA SP SS M SA 10 A B D C D B A 11 SS W SS SP 12 C S SP S C 13 SP SA SA l

M A D A

, in FUNCTION NUMSSR OF CLUSTWIS Centrol Sank D 8

! Conwei Senk C S l Centret Sank 8 8 Control Bank A 8 Shutdown Sank SS S Shutdown Bank SA 8 SP (Spare Rod Locations) 13-4

Piguro 2.3 : Burcoblo Abotrber c d Storca Accccbly 1,ocot.icco R P N H L K J H 0 F E D C B A I

1 2

3 8 12 12 12 4 4 18 18 SS 18 to 5 8 18 18 to 14 8 8 12 18 18 16 16 12 7 16 18 to 18 8 9F 12 12 9 16 18 is 14 to 12 16 18 14 18 12 11 8 14 18 18 16 8 12 14 14 SS 14 IS 1

13 8 12 12 12 4 13 15 0'

""FF- Number of WASNs sumary of Inserts 720 WARNS in I" " WABA Clusters 52 Control Rods 48 Thimble Plugs 55 Sec. Sources 2 Total 157 5

l Figuro 2.4 : Hurnobio Ab crier end S:condcry S Orca R:d Configuratic33

] =

I J. 3 7 D

• D Q

I I a I E g _ _ _ _

l.

I _ - _ __

8 tA CONnOURAT10N 12 8A CONnOURAT10N

_ _ _ I e e lE

_ a B'J e e e i

_g 9 e e

e. e 5 5 e e e

it EA C00 H 3URAT)CN MCOpeARY SOURCE LOCAT)ONS

. . _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - "^ ~ ~ - - ~ ^

i 3.0 CONTROL ROD DROP TIME MEASUREMENT '

t (FNP-1-STP-ll2)

PURPOSL The purpose of this procedure was to measure the drop time of all full i

length control rods under hot-full flow conditions in the reactor coolant system to insure compliance with Technical Sp6cification Requirements.

SUMMARY

OF RESMlli for the hot full-flow condition (Tavg 1541 deg.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 s 2.2 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 2.2 seconds. The longest drop time recorded was 1.85 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 CONDlil0NS DASHPOT ENTRY DASHPOT BOTTOM ,

Hot full-flow 1.63 sec 2.18 sec In order to confirm normal rod mechanism operation prior to conducting -

the rod drop test, the Verification of Rod Control System Operability test (FNP-0-ETP-3643) was performed. In this procedure, the stepping waveforms of the stationary, lift and movable gripper coils are .

• examined, rod speed is measured, and the functioning of the Digital Rod Position Indicator (DRPI) and the bank overlap unit are verified.

As a part of the RCCA Wear Reduction Program, the Cycle 11 control rod fully-withdrawn-position was changed from 228 steps to 231 steps by <

increasing the bank overlap from 100 steps to 103 steps .During the Rod Control System Operability test,- the- required changes to the bank overlap unit switch settings were verified, and a measurement of the actual fully-withdrawn position of each RCCA was performed in order to provide data for planning RCCA repositioning for future' fuel cycles.-

. Problems noted during the rod operability test were: (1) A 24 volt power supply failed in the 2AC power cabinet and had-to be replaced.

(2) The rod _ speed for both the control and shutdown banks was found to be out of tolerance and required adjustment of the master cycler in the_ logic cabinet to ccerect. (3) The bank overlap unit would not.

- count down when the "-t" tesr pushbutton was pressed, which required

- replacement _of a circutt-ca'/d.. Once these= problems were corrected, performance of the rod control system was _ satisfactory.

Following the F.od control-System Operability Test, Operat uns' >

exercised each control snd shutdown bank by performing five complete bank withdrawals and insertions in order to. verify there were no further problems with the system _ No problems occurred at this time.

Later, following heatup in. preparation for the rod drop test, it was found that- 24 volt power-supply PS2 in.both.the 1BD and 2BD power cabinets'had failed and had to be replaced.

y i

w r*

er > gs r'-w*t *'-awn w m--s' , -by+>ma vu 9 w's-? = y_i+cetwewr e un m n+ 9wwm+*me.s,-y y -g g, , - y e- 9-- - ps t-m t' e- -e- ww ert um9

_ _ _ _ _ _ . . _ _ _ _ . . . _ . _ _ _ . _ _ _ _ _ . _ . _ _ _ _ . _ . _ _ _ . _ _ _ _ ~ . _ _ _ . . _ . _ . _ _ _ _ . _ .

2 ggp.jg Dai t, 1 Cyclo 11 90 -

g'y 1.62 2.15 1.59 2.14 1.64 2.17

\Y '

P I

l.64 1.62 s "

2.16 2.17 [

1.59 1.58 1.63 1.65 M i

, 2.15 2.09 2.18 2.21

~

1.62 1.63 M ~

2.20 2,29 1.75 1.57 1.65 1.57 1.62 1,63 1,67 . .K 2.35 2.12 2.18 2.17 2.21 2.13 2.24 1.63 1.56 1.55 1.62 ,

~"

R' 2.15 2.09 2.04 2.12 s

N 1.65 1.61 1.59 1.61 N O 2.21 2.11 2.13 180 -H 2.18 1.68 1.56 1.58 1.62 -

2.25 2.08 2.11 2.17 1.69 1.62 1.62 1.57 1.59 1.63 1.62 r 2.26 2.16 2.18 2.12 2.14 2.18 2.13 1.62 1.65 -

2.17 2.20 "

k.65 1.62 1.58 1.63 3 2.22 2.16 2.11 2.15 1,62 1.66 C 2.15 2.18 1.69 1.72 1.85 2.24 2.28 2.50 3 270 / A A

-7 15 14 13 12 il 10 9 8 7- 6 5 4. 3 2 1 DRIVE LINE " DROP TIME" TABULATION _

TEMPERATURE - 547 F PRESSURE- 1944 991  % FLOW - 100 i X.XX BREAKER " OPENING" TO DASHPOT ENTRY ~ IN SECONDS BREAKER "0PENING" TO -DASHPOT BOTTOM - IN SECONt}S DATE. 5/ 17/. 91 X.XX

'8

4.0 INITIAL CRITICALITY (FNP-1-E1P-3601)

NE0E 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 RESUL15 Initial Reactor Criticality for Cycle 11 was achieved during dilution mixing at 0739 hours0.00855 days <br />0.205 hours <br />0.00122 weeks <br />2.811895e-4 months <br /> on 05/18/91. During the approach to criticality, >

NIS source and intermediate range overlap data were taken to demonstrate that adequate channel overlap existed. The reactor was allowed to stabilize at the following conditions:

RCS Pressure RCS Temperature 225gfpsig 548 Intermediate Range Power 8 x 10~9 Amp RCS Boron Concentration 2034 ppm Bank D Position 191 steps following stabilization, the point of adding nuclear heat was determined and a checkout of the reactivity computer using positive and negative flux periods was performed, initial criticality for Cycle 11 was achieved with the flux in the source range as a result of withdrawing the source-range (SR) detectors to the maintenance position to reduce the SR count rate.

During previous cycles, the high SR count rate resulted in criticality being achieved in the intermediate range. Testing performed prior to criticality (fNP-0-E1P-3652) demonstrated that withdrawal of the SR detectors reduced the count rate to approximately one-third of the value with the detectors in their normal, inserted position.

5.0 ALL-R005-0VT IS01tiERMAL TEMPERATURE COEfflCIENT AND BORON ENDPOINT (fNP-1-ETP-3601)

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

SUMMARY

Of RESULTS The AR0, hot zero power temperature coefficients and the ARO boron endpoint concentration are tabulated on the following page.

9-l

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

AE0J1Zf_1101HERMAL @D MOMBATOR 1[MPERATURE COEfflCIENT Boron Measured ITC Design Acc. Calculated Rod Conc. IT~ Critgrio" Hi Confiaur. alls ,_pp_m__ pl.m.d[ pcm/4 Pimigf All Rods Out 2038 0.0 40.45 1 2 +1.71 where:

ITC = lsothermal Temperature Coefficient, includes -1.66 pcm/O r Doppler coefficient HTC = Moderator Temperature Coefficient, normalized to the ARO condition No rod withdrawal limits were needed to maintain the moderator temperatu 45.0pcm/gecoefficientwithintheTechnicalSpecificationlimitof f.

ARO. HZP CORON ENDPOINT CONCENTRATION Rod Confingn ite Measured Cg._{npM Desian-prediriqd_Cg_.{tpal All Rods Out 2051 2082 1 50 Although the measured value falls within the 150 ppm tolerance band of the review criterion, it is noticeably below the design value. How-ever, Westinghouse (in letter 91AP*-G-0040, which transmitted document CDD-91-ll6) notified FNP that they estimated that their design code had overpredicted the HZP boron endpoint by 20 - 45 ppm, so this result was expected.

6.0 CONTROL AND SHVIDOWN BANK WORTH MEASUREMENTS (FNP-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 comparison with the values predicted by design.

SUMMARY

OF RESULTS The rod worth measurements were performed using the bank interchange method in which: (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 control and shutdown banks are derived from the change in the refer-ence bank reactivity needed to offset full insertion of the bank being' measured. The control and shutdown bank worth measurement results are given on the following page. The measured worths satisfied the review criteria both for the banks measured individually and for the combined worth of all the banks.

10' L

-~

j

$QtiMARY Of CON 1ROL AND SHU1DQWN BANK WORTH MEASUREMENTS Control or Predicted Dank Shutdown Worth & Review Hoasured Dank Percent DJink (riteria-(RCal Worth...fpg Differente A 315 1 100 317.5 0.8 B 1328 1 124 1331.0 0.2 C (Ref.)* 963 i 107 966.2 0.3 0 926 1 151 923.1 -0.3 i SD - A 926 i 160 935.1 1.0 SD - B 1008 1 153 989.1 ~1.9

All Danks 5406 1 541.7 5462.0 -0.07 ,

• Measured by the dilution method 7.0 STAR 10P AND POWER ASCENSION PROCEDURE (fNP-1-ETP-3605)

PURPOSI 4

The purpose of this procedure was to provido controlling instructions fort-

1. NIS intermediate and power range setpoint changes, as required prior to startup and'during power ascension.

2. Ramp rate limitatinn and control rod movement recommendations.

3. Conduct of startup and power ascension-testing, to_ include: _

a. HZPphysicstests(fNP-1-ETP-3601). '

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

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

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

e. Reactor coolait system flow measurement (fNP-1-STP-il5.1). I E MMARY OF RESULTS in order to. satisfy Technical Specification requirements for invoking special-HZP physics test exceptions, preliminary trip setpoints of--

E less than or equal to 25% power were used for the NIS intermediateEand l power range channels. When physics tests were completed, the power '

range setpoint was-increased to 80% to allow power escalation-above

- 25% for calorimetric recalibration'of the power range channels'.

- 11

~,

_=_ _ _ _ . _.,

c y m, .( # e- y er~ r*, +, e a s we r ee  % ,rg - W < , , - + - , e --c4 -% - ve r d- ++ e , + , - - * + +vw,--e--e-* w >--=-e--*-.

v e r--r--r e d' 4-.-*- - - --w - e- *-v- t

(The 80% setpoint was used instead of 109% in case the uncalibrated power range channels were indicating non-conservatively.) At a>proxi-mately 31% power, the power range channels were recalibrated, tie high-range trip setpoint was restored to 109%, and setpoint currents were determined for the intermediate range channels.

The Westinghouse fuel warranty limits the power ramp rate to 3% 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.

The determination of incore movable detector system core limit settings (FNP-1-ETP-3606) was accomplished during the ascension to 30%

pnwer. During this arocedure, it was found that incore detector "A" was inoperable, so t1at all startup test flux maps had to be performed using only four detectors. The incore excore recalibration test (FNP-ISTP-l?!), which was performed in two phases (at 30% power and 77% power), and the reactor coolant flow test (FNP-1-STP Il5.1) performed at 77% power, are described in Sections 8.0 and 9.0 of this report. As summarized in Table 7.1, core hot channel factors were evaluated from the incore-excore full-core base case flux map taken under non-equilibrium xenon conditions at 30% power, and from the full-core flux maps performed at 30% power, equilibrium xenon and at 77% power, non-equilibrium xenon.

TABLE 7.1

SUMMARY

Of POWER ASCENSION FLUX MAP DATA Parameter Man 262 Man 268 Mao 269 Avg. % power 30.37 30.0 76.68 Max FDH 1.5596 1.5567- 1.5156 Max power tilt

• 1.0174 1.0164 1.0142 Avg. core % A.O. 47.594 +7.488 42.623 Most limiting FQ(Z)** 2.1365 2.1370 1.9143 f4 Limit 4.5240 4.5240 2.9348 flux inap Non-equilibrium Equilibrium Non-equilibrium i conditions xenon- xenon xenon

• Calculated power tilts _ based on assembly FDHN from all assemblies.

l ** Based on percent margin to fQ limit.

12

8.0 INCORE-EXCORE DE1ECTOR CAllBRATION (fHp-1-Sip-121)

IML l The objective of this procedure was to determine the relationship i between power range upper and lower excore detector currents and axial l 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 delti-T protection system.

.511MMARY Of RESULl$ ,

At an indicated power of approximate'y 30%, a full-core base case flux map and five quarter-corv 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 RCS temperature. The power range NIS channels were adjusted to incorporate the revised calibration data.

Due to increasing QpTR indications as power was increased, a full-core flux map (No. 269) was performed during the hold at 77% power (rather than at 100% power) to verify that the core peaking factors were satisfactory and to perform a calibration correction to the detector current equations, primarily, this corrects for the effects of the '

change in programmed Tavg and for other factors which may affect core neutron leakage as power is increased from 30% to 100%. The resulting final equations for NIS Channels N41 - N44 are given in Table 8.1.

The aower range NIS channels were adjusted to incorporate the revised caliaration.

TABLE 8.1 DETECTOR CURRENT VERSUS AX1AL OFFSET EQUATIONS '

0 BLAIN FROM INCORE-EXCORE CAllBRATION TEST

[HANNEL N41:

1-Top = 0.7724*A0 + 156.82 uA l-Bottom = -l.0191*A0 + 152.71 uA CHANNEL N4h I-Top = 0.7248*A0 + 148.07.uA l Bottom = -1.0221*A0 + 152.65 uA

[lMtl[gl N43i 1-Top = 0.7314*A0 + 153.85 uA l-Bottom = -1.0586*A0 .+ 161.62 uA CHANNEL N44:

1-Top =- 0.7705*A0 4 154.73 uA-1-Bottom = -1.0620*A0 + 156.41 uA 13-

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

9 9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-1-STP-llS.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 Unit 1 Technical Specifica-tions.

SUMMARY

Of RESULTS in order 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 267,600 gpm for three loop operation. From an average of twelve calorimetric heat balance measurements performed at 76.7% power, the total core flow was determined to be 282,723 gpm, which meets the above criterion.

As a result of the removal of the RTD bypass manifold during the Cycle 10 - 11 outage and the replacement of all RCS loop RTDs, it was necessary to rescale the Overpower and Overtemperature delta-T protection systems to measured loop. delta-Ts. Since the RCS flow test involves making precision measurements of RCS loop T-hot and T-cold, the test data were also utilized to determine the 100% delta-Ts for protection loop rescaling. The 100% power loop delta-T valuos were found by dividing the average measured delta-T for each loop by the average (decimal) calorimetric power.

l 14

.. . .