ML20214L129

Jm Farley Nuclear Plant Unit 2,Cycle 5,Startup Test Rept

ML20214L129
Person / Time
Site:	Farley
Issue date:	08/28/1986
From:	Mcdonald R ALABAMA POWER CO.
To:	Rubenstein L Office of Nuclear Reactor Regulation
References
NT-86-0392, NT-86-392, NUDOCS 8609090405
Download: ML20214L129 (24)

Text

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AIABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER 2, CYCLE 5 STAR'IUP TEST REPORT PREPARED BY PLANT REAC'IOR ENGINEERING GROUP APPROVED:

d,0 kdM Technical Manager General Manager - Nuclear Plant

/ - (, ' *-T./

o DISK CYCLE 2/7 f[)6 j 8609090405 860828 )

PDR ADOCK 05000364 l P pop }

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l TABLE Of C0tm NTS I

PA3E 1.0 Introduction 1 2.0 Unit 2 Cycle 5 Core Refueling 2 l

3.0 Control Rod Drop Time Measurement 7 l

4.0 Initial Criticality 9 5.0 All-Rods-Out Isothermal Temperature Coefficient and Boron Endpoint 10 6.0 Control and Shutdown Bank Worth Measurements 12 7.0 startup and Power Ascension Procedure 14 8.0 Incore-Excore Detector Calibration 17 9.0 Reactor Coolant System Flow Measurement 20 4

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1.0 INTRODUCTION

he Joseph M. Farley Unit 2 Cycle 5 startup Test Report addresses the tests performed as required by plant procedures following core refueling. he 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.

Unit 2 of the Joseph M. Farley Nuclear Plant is a h ree Loop Westinghouse pressurized water reactor rated at 2652 lefth. We Cycle 5 core loading consists of 157 17 x 17 fuel assemblies.

%e Unit began commercial operations on July 30, 1981, completed Cycle 1 on October 22, 1982 with an average core burnup of 15350.5 19fD/MIU, completed Cycle 2 on September 17, 1983 with an average core burnup of 10371.2 MND/MTU, completed Cycle 3 on January 4,1985 with an average core burnup of 14,639.0 P9fD/MTU, and completed Cycle 4 on April 4, 1986 with average core burnup of 13,183.8 75fD/MTU.

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2.0 UNIT 2 CYCLE 5 CORE REFUELING REFERENCES

1. Westinghouse Refueling Procedure FP-APR-R4

2. Westinghouse WCAP 11150 (he Nuclear Design and Core Management of the Joseph M. Farley Unit 2 Power Plant Cycle 5) h refueling comenced on 4/14/86 and was completed in 6 days on 4/20/86. h as-loaded Cycle 5 core is shown in rigures 2.1 through 2.4, which give the location of each fuel assembly and insert, including the burinable poison insert locations and configurations.

he Cycle 5 core has a nominal design lifetime of 16600 MND/MIU and consists of 1 region 4 assembly, 16 region 5 assemblies, 72 region 6 assemblies and 68 region 7 assemblies. Fuel assembly inserts include 48 full length control rod clusters, 68 burnable poison inserts, two secondary sources, and 39 thimble plug inserts.

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l l 3 "08IN 270 -- *- 90 I I 10.. o1 I lH Fuci assembly serial numbers are given in the lower (or lower left) portion of each location. The remaining numbers denote the assembly insert. The original w/o U-235 enrichments were:

Region 4 (P) assemblies ..... 3.096%

Region 5 (R) assemblies ..... 3.402%

Region 6 (S) assemblics...... 3.443%

Region 7 (T) assemblies ..... 3.603%

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I FIGURE 2.2 CONTROL 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 4 C 8 SP B C 5 SB SP S8 8 A 8 D C D 8 A 7 SA S8 S8 SP SA

(

8 D SP C SP C SP D 9 SA SP S8 S8 SA 10 A 8 D C D 8 A 11 S8 SP S8 12 C 8 SP 8 C 13 SP SA SA 14 A D A Absorber Material Ag-In-Cd is FUNCTION NUMBER OF CLUSTERS Control Bank D 8 Control Bank C 8 Control Bank B 8 Control Bank A 8 Shutdown Bank SB 8 Shutdown Bank SA 8 SP (Spare Rod Locations) 13 LOCATIONS NS & C11 = CORE WATER LEVEL '.tlERM0 COUPLE PROBES 4

FIGURE 2.3 i

BURNABLE ABSORBER AND SOURCE ASSEMBLY LOCATIONS R P N M L K J H G F E D C B A 1

2 4 4 3 .4 12 20 12 4 4 16 24 SS 24 16 ,

5 4 16 24 24 24 16 4 6 12 24 24 24 24 12 7 4 24 24 20 24 24 4

\

S 20 24 20 20 24 20 9 4 24 24 20 24 24 4 10 12 24 24 24 24 12 11 4 16 24 24 24 16 4 12 16 24 SS 24 16 13 4 12 20 12 4

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14 4 4 15 SS Secondary Source 4 1120 Fresh Standard BA's 5

FIGURE 2,4 BURNABLE ABSORBER CONFIGURATIONS O D E O M

,0 E O O O O O O O E O E O E O O O O O E E O O O O O O E O E O 5 D

O O O E O E l 4 Fresh BA 12 Fresh BA Configuration Configuration 5 O E, 5 O E O E l O E E O 5 O 5 O E l

E O E I 16 Fresh BA i

Configuration a e, ,a a e,

,a .

! e O a O a a e a ms a e a a e a e a E O E O E E 5 E E E g 3 3 E E E 20 Fresh BA 24 Fresh BA Configuration Configuration 6

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

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1 4 3.0 CONTROL ROD DROP TIME MEASUIWGMT (INP-2-STP-112)

)

i' PURPOSE i The purpose of this test was to measure the drop time of all full t length control rods under hot-full f1w conditions in the reactor

coolant system to insure compliance with Technical Specification i requirements.

SUI 9tARY OF RESULTS i '

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For the Hot-full flow condition (T > 541'r and all reactor i

coolant pumps operating) Tech'ical Spec 1M cation 3.1.3.4 requires that the rod drop time from the fully withdrawn position shall be < 2.2 l seconds from the beginning of stationary gripper coil voltage decay j until dashpot entry. All full length rod drop times were measured to i

be less than 2.2 seconds. The longest drop time recorded was 1.49

! seconds for rod B-6. 'Ihe rod drop time results for both dashpot entry i and dashpot bottom are presented in rigure 3.1. Mean drop times are sununarized below: (

4 TEST MEAN TIME 'IO MEAN TIME 'IO l CONDITIONS DASHPOP ENTRY DASHPOT BOI"IOM Hot-full Flow 1.37 sec 1.86 sec l To confirm normal rod mechanism operation prior to conducting the

! rod drops, a Control Rod Drive Test (PNP-0-ETP-3643) was performed.

In the test, the stepping waveforms of the stationary, lift and  ;

4 moveable gripper coils were examined, and the functioning of the l Digital Rod position indicator and the bank overlap unit was checked. [

l Rod stepping speed measurements were also conducted. All results were ,

j satisfactory.

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A UNIT 2 CYCLE 5 900 1.36 1.38 1.41 \p P 1.85 1.88 1.90 1.37 1.36 1.37 1.37.

1.87 1.82 1.36 1.38

[ N 1.85 1.85' 1.82 1.85 N 1.38

'1.88 1.34 1.83 [l N 1.48 1.35 1.35 1.35 1.35 1.36 1.40 I

1.99 1.83 1.84 1.84 1.83 1.85 1.90 NN 1.36 1.33 1.30 1.36 l

l 1.86 1.81 1.80 1.82 ~J

% o 1.41 1.38 . 1.37 1.38 g 1.87 180 -H 1.91 1.85 1.85 1.34 1.32 1.34 1.39 1.84 1.82 1.78 1.88 ~0 1.40 1.38 1.36 1.43 1.36 1.36 1.37 1.82 1.95 1.83 F 1.88 1.87 1.87 1.88 1.36 1.40 E

1.85 1,91 1.36 1.33 1.33 1.36 l 0

, 1.85 1,82 1,82 1.83 1.37 1.38 1.79 1.86 C 1.39 1.43 1.49 .

I 1.88 1.93 1.95 A

270' ['

l 15 14 13 12 11 10 9 8 7 6 5 4 3 2 i ORIVE LINE "0 ROP TIME" TA80LATl0M TEMPERATURE . 549.7 F pg[33ygg . 2215 onir _  % FLOW 100 X.XX 8REAKER "0PENING" To DASHP0T ENTRY - IN SECOMOS DATE 5-11-86 X.XX BREAKER "0PENING" TO DASHPOT BOTTCH - IN SECONDS FIGURE 3.1 8 _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ , . _ _ _ _ _

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4.0 INITIAL CRITICALITY (INP-2-ETP-3601)

PURPOSE

%e purpose of this procedure was to achieve initial reector 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.

SLAMARY OF RESULTS Initial Reactor Criticality for Cycle 5 was achieved during dilution mixing at 1211 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.607855e-4 months <br /> on May 11, 1986. he reactor was allowed to stabilize at the following critical conditions: RCS 223 pressurp 8 x 10- amp,5RCSpsig, boronRCS temperature concentration 548.0'r, 1730 ppm, intermediate and Control Bank D range power i

position- 190.5 steps, rollowing stabilization, the point of adding i

nuclear heat was determined and a checkout of the reactivity computer

! using both positive and negative flux periods was successfully accomplished. In addition, source and intermediate range neutron channel overlap data were taken during the flux increase preceding initial criticality to demonstrate that adequate overlap existed.

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5.0 ALL-RODS-Our ISOINERMAL TEMPERA'IURE COErrICIENT AND BORON ENDPOIffr (fNP-2-r1P-3601)

PURPOSE

! h e 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 measure the ARO boron endpoint concentration.

SUMMARY

OF RESULTS The measured ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration are shown in Table 5.1. We isothermal temperature coefficient was measured to be -3.10 penV'r which meets the design acceptance criteria. This gives a calculated moderator temperature coefficient of -0.77 penV'r which is within the Technical Specification limit of +5.0 penV'r. %us, no rod withdrawal limits were needed to ensure the +5.0 penv'r limit was met. The design acceptance criterion for the ARO critical boron concentration was satisfactorily met.

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TABLE 5.1 i

ARD, HZP ISO 1HERfRL APO 70DERATOR TEMPERATURE CDEFFICIENT ,

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Rod Configuration Boron Measured Calculated g Design Acceptance Concentration g a, , , Ghedm ppa pcW'T pcm/*F pen /'F All Rods Out 1723.8 -3.10 -0.77 -3.07 3 g - Isothermal temperature coefficient, includes -2.33 pcm/*F doppler coefficient a,,, - M ra w M y wrahe emded ARO, HZP BORON ENDPOINT CONCENTRATI(N Rod Configuration Measured C, (ppm) Design - predicted C, (ppm)

All mods out 1730.2 1751 50

6.0 CONTROL AND SHt7tDOWN BANK WORDI MEASUREMENTS (INP-2-ETP-3601)

PURFOSE he 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 h e rod worth measurements were performed using the bank I 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 reference bank reactivity needed to offset full insertion of the bank being measured.

The control and shutdown bank worth measurement results are given in Table 6.1. The measured worths satisfied the review criteria both for the banks measured individually and for the combined worth of all banks.

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

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TABLE 6.1

! SulTERY OF CQfrROL APO SIRTIDOM BANK WORTH MEASUREMENFS 1

Predicted Bank Measured Worth & Review Bank Percent

. Bank Criteria (pcm) Worth (pcm) Difference Control A 472 100 465.8 -1.3 J

Control B (Ref.) 1357 136 1311.8* -3.3 1'

control C 859 129 795.6 -7.4 i'

control D 1092 164 1023.6 -6.3.

C Shutdown A 1142 171 1106.0 -3.2 I

Shutdown B - 1017 152 930.9 -8.5 j

I All Banks Combined 5939 594 5633.7 -5.1

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• Measured by dilution method i

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7.0 STARTUP AND POWER ASCENSION PROCEDURE (FNP-2-ETP-3605)

PURPOSE ne purpose of this procedure was to provide controlling instructions for:

1. NIS intermediate and power range 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 include:

a. HZP reactor physics tests (FNP-2-ETP-3601).

b. incore movable detector system alignment (FNP-2-ETP-3636).

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

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

e. reactor coolant system flow measurement (FNP-2-STP-ll5.1).

SUMMARY

OF RESULTS In order to satisfy Technical Specification requirements for invoking special core physics test exceptions, preliminary trip setpoints of less than or equal to 25% power were used for the NIS intermediate and power range channels. When physics tests were completed, the power range setpoint was increased to 80% to enable power escalation (above 25%) for calorimetric recalibration of the power range channels. (The 80% setpoint was used instead of 109% in case the uncalibrated power range channels were indicating non-conservatively.) At approximately 35% 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.

h e 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 startup test program is addressed elsewhere in this report except for the following notes:

Determination of the incore movable detector system core limit settings (FNP-2-ETP-3606) was accomplished for all modes of system operation during the ascension to 35% power.

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In previous startups, the incore-excore recalibration was performed at approximately 75% power. During the Cycle 5 startup, however, a preliminary recalibration was performed at 35% power and was refined using additional data collected at higher power levels.

No excessive quadrant power tilt ratios or delta flux channel calibration problems were encountered during the ascension above 35%

power.

h e revised incore-excore recalibration program resulted in seven quarter-core, and four full-core flux maps being taken between 35% and 100% power. h e results from the full-core maps were within Technical Specification Limits, and are sunnarized in Table 7.1.

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TABLE 7.1

SUMMARY

OF POWER ASCENSION FLUX MAP DATA Parameters Map 118 Map 124 Map 125 Map 128 Date 5/13/86 5/24/86 5/25/86 06/02/86 Time 23:11 17:53 22:47 12:45 Avg. % Power 34.19 31.75 48.32 98.42 Max FAH 1.4812 1.4926 1.4951 1.4645 Max Power Tilt

• 1.0059 1.0072 1.0042 1.0052 Avg. Core % A.O. -3.441 -0.164 -0.328 -3.918 Maximum FQ(Z) 2.1608 2.0643 2.0488 1.9171 FQ Limit 4.6084 4.5969 4.5969 2.3400 Xenon Conditions Non- Equilibrium Non- Equilibrium Equilibrium Equilibrium

• Calculated power tilts based on assembly F6HN from all assemblies.

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8.0 INCORE-EXCORE DETECIOR CALIBRATION (FNP-2-STP-121)

PURPOSE

, h objective of this procedure was to determine the relationship i between power range upper and lower excore detector currents and incore axial offset for the purpose of calibrating the delta flux penalty to the overtemperature AT protection system, and for calibrating the control board and plant computer axial flux difference (AFD) channels.

SUMMARY

OF RESULTS During previous startups, incore-excore recalibration was performed at approximately 75% power. However, during the Cycle 5 startup, the following modified sequence was used:

(a) At approximately 35% power, a full-core base case flux map and five quarter-core flux maps were run to perform the basic incore-excore recalibration. he power range NIS channels were adjusted to incorporate the revised calibration data.

(b) At a later time, a full-core flux map was performed at approximately 32% power under more stable xenon conditions than the original base case map. The evaluation of this map included a verification of the effectiveness of the 35% power NIS recalibartion.

(c) During power escalation, a full-core map was taken at 48% power and quarter core maps were taken at 51% and 55% power to develop additional data for comparison with the 35% power results.

Problems with the incore movable detector system prevented five quarter-core maps from being obtained as was planned.

(d) When xenon equilibrium was achieved at approximately 100% power, a full-core flux map was taken to renormalize the calibration data to compensate for temperature decalibration (the change in reactor core neutron leakage caused by the changes in coolant temperature l associated with changes in power). he power range NIS channels were recalibrated to incorporate this correction.

l At 35% power, the six flux maps were performed at axial offsets of approximately + 18%, + 7%, -3%, -15%,.-22%, and -27%. the detector currents measured during the flux maps were normalized to 100% power, and a least squares fit was performed to derive the output current vs.

axial offset equation for each top and bottom detector. Calibration values obtained from these equations were used to recalibrate the NIS i

channels.

Evaluation of the flux map taken at 32% power showed a small error in the recalibration of NIS channel N44, but confirmed that the calibration equations derived at 35% power were satisfactory.

l he calibration error on channel N44 was corrected.

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Between 48% and 55% power, three flux maps were performed at axial offsets of approximately +6%, -0.3% and -5%. By subtracting the current contributions due to temperature decalibration, the currents measured during these maps were, in effect, normalized to 35% power so that they could be merged with the 35% power data. Using the combined 35% and 48% - 55% power data, the incore-excore equations were recalculated to obtain improved equation slopes. (The resulting changes in slope were small.) No recalibration of the NIS channels was performed at this time.

At approximately 100% power, revised zero percent axial offset (I-zero) currents were determined from flux map data and were combined with the merged 35% and 48% - 55% equation slopes to yield the finalized incore - excore equations given in Table 8.1. Using these results, the NIS channels were recalibrated.

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TABLE 8.1 DETEC'IOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:

I-Top = 0.7889*AO + 185.20 pa I-Bottom = -1.1085*AO + 182.84 a CHANNEL 42:

I-Top = 0.8766*AO + 180.55 a I-Bottom - -1.1619*A0 + 179.08 pa CHANNEL N43:

I-Top = 0.8250*A0 + 189.63 pa I-Bottom = -1.1836*A0 + 184.58 a CHANNEL N44:

I-Top = 1.0003*AO + 202.10 a I-Bottom = -1.3355*A0 + 202.23 pa 19

l 9.0 REAC'IOR C00IANT SYSTEM FIM MEASURDiENT (PNP-2-STP-115.1) l PURPOSE

'Ihe purpose of this procedure was to measure the flw rate in each reactor coolant loop in order to confirm that the total core f1w met the minimum flow requirement given in the Unit 2 Technical Specifications.

SiJfflARY OF RESULTS To comply with the Unit 2 Technical Specifications, the total reactor coolant system flow rate measured at normal operating temperature and pressure must equal or exceed 265,500 gpm for three loop operation. From the average of six calorimetric heat balance measurements, the total core flow was determined to be 285,767.6 gpm, which meets the above criterion.

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.. '.' NT 86-0392 I Calling Address Alab ma Power Company 600 North 18th Street Post office Box 2641 i Birmingham. Alabama 35191 Telephone 205 783-6090 R. P. Mcdonald Senior %ce President -

Flintridge Building AlabamaPbwer IN saut en e,cr c s, v . '

August 28, 1986 Docket No. 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. L. S. Rubenstein Joseph M. Farley Nuclear Plant - Unit 2 Cycle 5 - Startup Report Gentlemen:

Enclosed is the Startup Report for Unit 2 Cycle 5 as required by the April 29, 1986 letter from Mr. R. P. Mcdonald to Mr. L. S.

Rubenstein.

If you have any questions, please advise.

Yours ve trul ,

c' R. P. Mcdonald RPtyMDR:emb Enclosure cc: Mr. L. B. Long

! Dr. J. N. Grace l Mr. E. A. Reeves Mr. W. H. Bradford s

e,vP

@ 9"'9 u ap

g. W' s 4s i\

bc: Mr. W. O. Whitt Mr. W. G. Hairston, III Mr. K. W. McCracken Mr. J. D. Woodard i

Mr. J. W. McGowan Mr. C. D. Nesbitt Mr. R. G. Berryhill Mr. D. E. Mansfield Mr. J. A. Ripple Mr. J. K. Osterholtz Mr. J. T. Brantley Mr. D. E. Dutton Mr. J. R. Crane Mr. K. C. Gandhi i

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