ML19268A570

AL Power Co Joseph M Farley Nuclear Plant Unit 1,Cycle 3 Startup Test Rept. Proprietary App a Withheld (Ref 10CFR2.790)

ML19268A570
Person / Time
Site:	Farley
Issue date:	09/10/1981
From:	Foster G, Macdonald W, Mcdaniel S ALABAMA POWER CO.
To:
Shared Package
ML19268A567	List: ML19268A566 ML19268A570
References
NUDOCS 8112170263
Download: ML19268A570 (20)

Text

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ALABAMA POWER COMPANY

. JOSEPII M. FARLEY NUCLEAR PLANT

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UNIT NUMBER 1, CYCLE 3 STARTUP TEST REPORT PREPARED BY PLANT REACTOR ENGINEERING GROUP S. MCDANIEL

_ G. FOSTER W. S. MACDONALD

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App):oved:

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, .adug,g_ Generating Plant Engineer - Supervising

,wa r5,w , Technical Supervisor o

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[r f30, .// ecimical Superintendent 4).I. /l.b2'r-w.c Plant Manager 9112170263 510910' PDR ADOCK 03000349

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TABLE OF CONTENTS PAGE 1.0 Introduction 1 2.0 Core Refueling 2 3.0 Control Rod Drop Time Measurement 7 4.0 Initial Criticality 9 t

5.0 Control Rod and Boron Worth Measurements 10 6.0 ARO CZP Flux Distribution, Moderator 12 Temperature Coefficient and Boron Endpoint.s 7.0 Power Ascar>;. ton Procedure 15 8.0 Incore-Excore Detector Calibration 17 we .

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1.0 INTRODUCTION

The~ Joseph M. Farley Unit I Cycle 3 Startup Test Report addresses the tests identified as requirements 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 assumed in the FSAR safety analysis.

The Joseph M. Farley Nuclear Plant consists of two pressurized water reactors, each rated at 2652 MW L. The 3-l'oop Nuclear Steam Supply Systems and Turbine Generator sets were supplied by Westinghouse, and provide a gross electrical output of 860 MW per unit. The initial core loading for each reactor consists of 157 fuel assemblies composed of 17 x 17 rod arrays.

Unit I began Cycle 1 commercial operation on December 1, 1977, following completion of the initial startup and powcr ascension tast program Cycle 1 operation ended on Marcli 8,1979, with an average core burnup of 15,450 MWD /MTU. Cycle 2 operation ended on November 8, 1980, with an average core burnup of 10,177 MWD /r.TU. Initial criticality for Cycle 3 was achieved on March 25, 1981, and full power operation was resumed on April 2 0 , 1 9.8 1~.

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2.0 Unit 1 CYCLE 3 CORE REFUEL'NG REFERENCES

1. Westinghouse Refueling Procedure FP-ALA-R2

2. Westinghouse WCAP 9761 (The Nuclear Design and Core hcnagement of the Joseph M. Farley .

Unit 1 Power Plant Cycle 3)

3. Westinghouse FP-AP-256 (Joseph M. Farley Nuclear Design Revisions for Unit 1 Cycle 3) 2.1 FUEL SHUFFLE OPERATIONS .

Nuclear ftiel (52 assemblies) was received on site for Cycle 3, Core Region 5, during the period from February 19, 1980, to March 13, 1980. No problems of significance were disclosed by fuel receipt inspections.

The core fuel. shuffle commenced on November 30, 1990. The shuffle resulted in: (1) the removal of all assemblies to the spent fuel pit; (2) shuffling of the Region 1, 3, and 4 assemblies into an approximate checkerboard pattern in the inner section of the core; (3) arranging the fresh (Region 5) assemblies into a ring surrounding the inner checker-board; and (4) relocation of inserts (such as control rods), and replacement of spent burnable poison inserts with thimble plug inserts.

During the fuel shuffle,.two optimized luel assemblies (Region-4A) were inspected by a team from the Westinghouse Nuclear Fuel Division. As _

detailed in the Westinghouse Preliminary Report, 2

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the optimized assemblies were found to be in good condition following one cyle of irradiaticn, and suitable for continued use. (A copy of the '

Westinglovce Report is attached as Appendix A.)

When the refueling operations were completed, the core was scanned and videotaped with an underwater TV camera to verify the correct locption of each fuel assembly.

2.2 CYCLE 2 CORE DESCRIPTION

- The as-loaded Cycle 3 ccre is depicte:1 in Figures 2.1 through 2.3, which give the location of each fuel assembly and component, and provide information on assembly enrichment and burnup.

The core is composed of 1 Region-1 fuel assembly, 52 Region-3 assemblies, 50 Region-4 assemblies, two Region-4A optimized test assemblies, and 52 Region 5 assemblies. Fuel assembly inserts consist of 48 full length rod cluster control assemblies, 107 thimble plug inserts, and 2 secondary sources.

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. C'C D ZD .{ 10: ASSEMBLY i

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ZD: DEMONSTR ATIO

- REGION 1 3 4 4A 5 ASSEMBLY FROM CYCLE 1 2 2 2 FEED R: R EMOV AG t. E W/O U 235 2.115 3.102 3.113 3.108 2.801 FUEL RODS

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3.0 CONTROL ROD DROP TIME MEASUREMENT (Procedure FNP-1-STP-112)

PURPOSE The purpose of this procedure was to measure the drop time of all full length control rods under. hot-full. _

flow conditions in the reactor coolant system- to insure system performance was in' compliance with Technical Specification requirements.

SUMMARY

OF RESULTS For the Hot-full flow condition.(T avg > 541.F an.d all reactor coolant pumps operating) Technical Speci-fication 3.1.3.4 requires that the rod drop time from 1 the fully withdrawn position shall be i 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 1.780 seconds for rod B-6. A tabulation of the drop times for the remaining full length control rods is presented in core diagram form in Figure 3.1 together with drop times to dashpot bottom. -Mean drop times are summarized below.

TEST MCAN TIME TO MEAN TIME TO

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CONDITIONS DASHPOT ENTRY DASEPOT 30TTOM Hot-Full Flow 1.58 sec 2.15 sec To confirm normal rod mechanism operation prior to conducting the rod drops, a Control Rod Drive Test (FNP-0-IMP-230.3) was performed. In the test, the stepping waveforms of the . stationary, lif t and movable gripper coils. vere examined and rod specd measuremente vero conducted. All results were saticfactory.

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- TEMPERATURE - 547*F PRESSURE- 2236 PSIC  % FLOW - 100

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X.XX BREAKER " OPENING" TO DASIIPOT ENTRY IN SECONDS DAT~e- 3/22/81 X.XX DREAKER " OPENING" TO DASl! POT DOTTOM - IN CECONdS FIGURE 3.l'

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4.0 11:1TIAL CR.ITICALITY (Procedure FNP-1-ETP-80)

PURPOSE The purpose of this procedure was to achieve initial reactor criticality under carefully contro1{ed _

conditions, establish the upper flux limit for the conduct of zero power physics tests, and operationally verjfy the calibration of the reactivity computer.

SUMMARY

OF MSULTS Initial Reactor Criticality was achieved during dilution mixing at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> on March 25, 1981.

Conditions were allowed to stabilize at 1X10-8 Amps in the intermediate range at a Control Bank D position of 145.5 steps and an RCS Boron Concentration of 1333 ppm.

The approach to criticality proceeded without incident.

Following stabilization, the point of adding nuclear heat was determined and a checkout of the reactivity computer was successfully accomplished. In addition,

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source and intermediate range neutron channel overlap data ver'e taken during the flux increase preceding and

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immediately following initial criticality to demonntrate adequate overlap existed.

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5.0 CONTROL ROD AND BORON WORTH MEASUREMENTS (Procedures FNP-1-ETP-81, -83, -84, -85, -86, and -88)

PURPOSE The purpoco of these procedures was to meacure the differential and integral reactivity worth of each .

control rod bank, both individuelly and when moving in overlap, and to determine tile differential boron worth over the range tf control bank movement.

_ SIR 4 MARY OF RTEU_L_T_S_

The results of the contro.1 bank worth measurements, both for banks moving individually and in overlap mode, together with boron worth determinations are summarized in Table 5.1. All measurements satisfied

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their respective review criteria.

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6.0- ARO HZP FLUX DISTRIBUTION, MODERATOR TEMPERATURE COEFFICIENT, AND LORON ENDP02iTS

, (Procedures FNP-1-ETP-82, -83, -84, -85 and -86)

PURPOSE .

The objectives of these procedures vere to:

(1) measure the hot zero power moderator temperature coefficient in the all-rods-out configuration; (2) Determine the core flux distribution for the, all-rods-out configuration; and (3) measure the boron end point concentrations for the ARD, D-in, D~ +

C-in, D + C + B-in and the D + C + B + A-in rod ,

configurations . ,

SU11MA),1Y OF RESULTS

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Table 6.1 gives a tabulation of the measured boron end point concentrations compared with the design values for each rod configuration considered.

The vendor's recommended acceptance criterion for the all-rods-out critical boron concentration was

- aatisfactorily met.

Table 6.2 is a tabulation of measured isothermal and moderator temperature coefficientc for the HZP all-rods-out configuration, together with the isothermal temperature coefficient obtained from core design curves. The design acceptance criterion for the isothermal temperature coefficient was satisfactorily met, and the moderator temperature coefficient was found to be negative, as. required by plant Technical Specifications.

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TABLE 6.1 HZP BORON ENDPOINT CONCENTRATIONS Rod Configuration Measure,d C B Design-Predicted C B (ppm) (ppm) ,

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ARO 1357.8 1374 1150 p'm*p D in 1252.4 1260 D+C in 1143.4 1145 D+C+B in 971.4

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( FNP-.1-EIP-100 )

1 P'sRPOSE

'lhe pu;. pose of this procedure was to provide controll.tng instructions fcr the following: ,. _

1. 1: amp rate old control rod movement limitations

2. Incore mov able detector system final alignment

3. Flux nap at less than 50% power

4. Delta flux band for ascersion to 75% power Incore/Excore calibr:r.;ir.n at /5% power.

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[;UFjMARY OF RESULTS To corrply with tha ft:el warranty, the power ramp rato uas limited to 3% of full power per hour bi'. ween

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20% and 100% power unP.il f.ul'l: power was achieved for 72 cumulative hours coat of any seven-day operation period. Control rod notion during the initial return

o power was. minimized, and the startup was conducted with the rods withdrawn as far as ponsible. The. rod withdrawal rate was limi;ced -to 3 steps per hour .above 50% power.

The finL1, detailed alignment of-tha incore

.4-movabic' detector system was performed during power i

ascens' ion (at powc.r levels shove 5%) prior to taking -

the 48% po'wer flux map.

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Flux maps were taken at 48%, and 74," power. The results, including power. tilts, were well within Tech-nical Specification Limits.

A preliminary incere/excore calibration verification was performed at 48%,,and a final calibration was performed

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at 75% Power. Results of the final incore/excore calibratio are given under Section 8.0.

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INCORE-EXCORE DETECTOR CALIBRATION (Procedure FNP-1-STP-121) .

PURPOSE The purpose of this procedure was to determine the relationchip between pouer range upper _and lower. _

execre detector currents, and incore axial offset - -

to provide data for calibrating the delta flux penalty to the overtemperature AT protection system, and for calibrating the control board and plant corputer axial flux difference (AFU) M'annels.

SUMMARY

OF RESULTS ,

?reliminary and final verifications of encore AFD channel calibralion were performed at 48% and 75% power,

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resnectively. The flux maps for the final verification vere run at ave?: age core axial offsets of +3.635, -12.601,

-21.717, #10.17 and +15.29, as determined from the incore resul ts .

The detector currents were normali' zed to 100% power, and a le~ast squares fit was performed to obtain the linear equations for each detector's top and botton current.

Using these equations, the detector current cali-

__ bration data was generated and utilized to recalibrate the delta flux instrumentation.

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