ML20069F635

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

ML20069F635
Person / Time
Site:	Farley
Issue date:	02/28/1983
From:	ALABAMA POWER CO.
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ML20069F617	List: ML20069F614 ML20069F635
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NUDOCS 8303230252
Download: ML20069F635 (19)

Text

ALABAMA POWER COMPANY i JOSEPH M. FARLEY NUCLEAR PLANT i

i UNIT NUMBER 2, CYCLE 2 i STARTUP TEST REPORT F

l l

4 PREPARED BY PLANT REACTOR ENGINEERING GROUP 4

APPROVED: .

d.O Technical Superintendent (d.. - [ I Plant Manager z 9]

DISK: CYCLE 2 8303230252 830315 '

PDR ADOCK 05000364 '

1 P ppg ._, _ _ , _ , , , _ _ _ , _ _ __ _ _ _ _ _ , _ , , _ _

. . . _ _ _ _ _ _ . _ _ . _ _ _ _ _e . - . ._ _. -_ _ --

TABLE OF CONTENTS i

PAGE 1.0 Introduction 1 2.0 Core Refueling 2 i

i 3.0 Control Rod Drop Time Measurement 5 I

4.0 Initial Criticality 7 5.0 Control Rod and Boron Worth Measurements 8 6.0 ARO HZP Flux Distribution, Moderator 10 Temperature Coefficient and Boron Endpoints 7.0 Power Ascension Procedure 13 8.0 Incore-Excore Detector Calibration 15

! 9.0 Reactor Coolant System Flow Measurement 17

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1.0 INTRODUCTION

l The Joseph M. Farley Unit 2 Cycle 2 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 I

comparison of measured parameters with design predictions, Technical Specifications, or values assumed in the d

FSAR safety analysis.

Unit 2 of the Joseph M. Farley Nuclear Plant is a

, Three Loop Westinghouse pressurized water reactor rated at 2652 MWth. The Cycle 2 core loading consists of l 157 17 x 17 fuel assemblies.

The Unit began commercial operations on July 30,

j 1981, and completed Cycle 1 on October 22, 1962 with an average core burnup of 15350.5 MWD /MTU.

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I 2.0 UNIT 2 CYCLE 2 CORE REFUELING REFERENCES

1. Westinghouse Refueling Procedure FP-APR-R1

2. Westinghouse WCAP 10187 (The Nuclear Design i and Core Management of the Joseph M. Farley I Unit 2 Power Plant Cycle 2) i i

The refueling commenced on 11/6/82 and was completed i in 10 days on 11/16/82. The as-loaded Cycle 2 core is

<t depicted in Figures 2.1 and 2.2, which give the loca-tion of each fuel assembly and insert, and the assembly i enrichments. The Cycle 2 core consists of 1 Region-1 I fuel assembly, 52 Region-2 assemblies, 52 Region-3 i assemblies, and 52 Region-4 assemblies. Fuel assembly

! inserts consist of 48 full length rod cluster control l assemblies, 107 thimble plug inserts, and 2 secondary

{ sources.

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Figure 2.1 -

APR Unit 2, Cycle 2 Reference Loading Pattern R P N M L K J H G F E D C B A ,

, l P 03 P 21 P 24 -

F. F .. F-!

P 38 P-40 P 02 M-43 P-49 P 10 P 33 2

F- F ' F M H7 F - F *- F P 34 P 29 M 51 M 16 N-30 g.1g M.33 P 31 P-20 3

F J: F _, J-4 K5 h F5 G-4 F F -

P 23 N 12 M-09 N-25 N 15 M 24 N-47 N-02 M 31 N 44 P 01 '

F B-11 J-6 G 15 L3' H-3 E-3 J 15 G-6 L 14 F~ <

l P 37 P-18 M 06 N 13 M 17 N 06 M 38 N 04 M 50 N 51 M 47 P 12. P-45  ::

~

) F F K-7 L2 K3 B7 H5 P7 F3 B-5 F7 F F P 43 M 05 N 49 M-45 N 20 M 15 N 36 M-37 N 23 M 02 N 14 M-12 P 36 g F . M7 A-9 N6 N4 L4 K2 E4 D-3 C-6 R9 D7 F P 52 P 35 M-22 N-03 N 46 M-07 N 45 N 22 N-07 M-39 N 33 N 18 M 23 P 32 P-42 7 F F L6 N3 J-14 M-5 P6 H1 F2 D5 G 14 C-5 E6 F F P 11 M 35 N 37 M-27 M-41 N 21 N 52 L 11 N-41 N 17 M 03 M-08 N 38 M-14 P-47 g F J8 N 12 N-8 L8 P 10 R-8 D 12 A8 B-6 E8 C8 C4 G-8 F P 09 P-22 M 52 N 16 N-40 M-48 N 08 N 31 N-03 M 42 N-39 N 34 M 04 P-41 P 27 g F F L 10 N 11 J-2 M 11 K 14 H-15 B-10 D-11 G2 C-11 E 10 F F P-44 M 36 N 28 M-20 N 29 M-40 N 19 M-29 N 24 M 28 N 43 M-32 P 46 10 F M9 A7 N 10 M 13 L-12 F 14 E 12 C 12 C 10 R7 D-9 F'.

P-51 P 04 M 21 N 42 M-46 N 50 M 25l N-26 M 19 N 10 M 49 P 50 P 05 11 F F K-9 P 11 K 13 B9 H 11 P9 F 13 E 14 F9 F F P 16 N-48 M 11 N 05 N-35 M-44 N 01 N 32 M-34 N 27 P 08 F -

E2 J-10 G1 H 13 E 13 J1 G 10 P5 F 12 L _13_

P 17 P 26 M 13 M 10 ! N 11 g.oj , M 26 P 25 P 13 -

F F J-12 K 11 $13 F 11 G 12 F F P 39 P 30 P 48 M 30 P 28 P 14 P 19 F F F H9 F F F I#

P-07 P 06 P 15 15 F F F COLOR CODE lD LEGEND:

fX X ID: ASSEMBLY REGION 1 2 3 4 IDENTIFICATION FROM CYCLE

• 1 1 FEED XX: PREVIOUS CYCLE LOCATION W/O U 235 2 120 2.601 3.100 3.10 l 3

(. _._.-. _ _ . . . _ _ _ . . . _ . _ . _ _ . - . _ . . . . _ . _ . _ . . . _ _ _ .

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l A D A.

2

SA SS SA SP C 8 SP 8 C SP S8 SP SS 5 A B D C D 8 A SA S8 S8 SP SA 7

. A 0 SP C SP C SP O O SA SP $8 SB SA 9 A B D C D 9 A 10 SB SP SB SP C 8 SP B C SP SA ss SA Absorber Material: A D A Ag-In-Cd 15 Contr ank 0 8

't Control Bank C 8

Control Bank B g

'j Control Bank A 8

Shutdown Sank 5 8 3

Shutdown Bank S 8

SP (Spare Rod Locations) 13 SS (Secondary Source) "

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1 3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-2-STP-112)

PURPOSE The purpose of this test was to measure the drop time of all full length control rods under hot-full flow conditions in the reactor coolant system to

, insure compliance with Technical Specification requirements.

SUMMARY

OF RESULTS For the Hot-full flow condition (T > 541*F and all reactor coolant pumps operating) Te8Kflical Speci-fication 3.1.3.4 requires that the rod drop time from

the fully withdrawn position shall be < 2.2 seconds from the beginning of stationary gripper coil voltage decay until dashpot entry. All full length rod drop 3 times were measured to be less than 2.2 seconds. The longest drop time recorded was 1.45 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 l CONDITIONS DASHPOT ENTRY DASHPOT BOTTOM Hot-full Flow 1.362 sec 1.863 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, lift and moveable gripper coils were examined and rod stepping speed measurements were conducted. All results were satis-factory.

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UNIT 2 CYCII 2 A

MORTH 900 R

1 1.33 1.86' 1.37 1.86 1.40 1.94

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, 1.36 1.35 l i 1.86 -1.86 M 1.35 1.36 '

1.35 1.39 l 1.85 1.86 '

,1.85 1.90 M 1.36 1.35 1.85 1.84 [Lx l

! 1.43 1.36 1.32 1.33 1.35 'l.35 .

1.39 E

l 1.95 1.88 1.80 1.82 1.88 1.86 1.90

- N 1.37 1.33 '1.32 1.36

{ N 1.86 1.83 1.79 1.83 J ,

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) 0' l.38 1.35 1.37 !

1.35 180 8

-g 1.88 1.84 1.85 1.83 1.35 1.32 1.34 '

1.39 1.86 1.80- 1.87 1.90 -G 1.37 1.37 1.35 1.33 1.37 1.37 1.35 1.88 1.86 1.83 1.82 1.87 1.85 1.85 F 1.36 1.37 1.87 1.88 E 1.38 1.35 1.33 '

1.36 8 1.87 1.83 1.85 l.86 0

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1.38 1.37 1.87 1.87 0 1.40 '1.40 -1.45

.1.92 1.92 1.98 I

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! 15 14 13 12 ll 10 9 8 7 6 5 4 3 2 1 ORIVE LINE "0 ROP TIME" TA8ULATION TEMPERATURE . 547*F PRESSURE . 2235 psig _ g ptow . 100%

j X.XX BREAKER "0PENING" TO DASHPOT ENTRY - IN SECOMOS DATE -

11-28-82 X.XX BREAXER "0PENING* TO DASHPOT 50TTOM - IN SECOMOS l FIGURE 3.1

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m 4.0 INITIAL CRITICALITY (FNP-2-ETP-3601)

PURPOSE The purpese of this procedure was to achieve initial reacter criticality under carefully controlled conditions, establish the upper flux limit for the i

conduct of zero power physics test, and operationally verify the calibration of the reactivity computer.

SUMMARY

OF RESULTS T

Initial Reactor Criticality for Cycle 2 was achieved during dilution mixing at 1136 hours0.0131 days <br />0.316 hours <br />0.00188 weeks <br />4.32248e-4 months <br /> on November 30, 1982.

The reactor was allowed to stabilize at the following critical conditions: RCS pressure- 2240 psig, RCS temperature 547*F, intermediate range power 1.1x10 -8 amp, RCS boron concentration 1361 ppm, and Control

. Bank D position- 188.5 steps. Following stabilization, the point of adding 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 and immediately following initial criticality to demonstrate that adequate overlap existed.

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5.0 CONTROL ROD AND BORON WORTH MEASUREMENTS (FNP-2-ETP-3601)

PURPOSE Tha objectives of Control Rod and Boron Worth Measure-ments were: (1) to measure the differential and integral reactivity worth of each control rod bank, both individually and when moving in overlap, (2) to determine the differen-l tial boron worth over the range of control bank movement,

and (3) compare results with the design calculations.

SUMMARY

OF RESULTS

! The results of the control bank worth measurements both for banks moving individually and in overlap mode, i together with boron worth determinations are summarized i in Table 5.1. All measurements satisfied their respective review criteria.

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SUMMARY

OF CONTROL ROD AND BORON WORTH MEeIUREMENTS Predicted Bank Measured Design Measured '

Rod Avg. Boron Worth & Review Bank Percent Boron Worth Boron Worth Configuration Conc. (ppm) Criteria (pce) Worth (pcm) Difference (pcm/ ppm) (pcm/ ppm)  !

D 1330 1085 1 163 1048 -3.41 -9.35 -9.29 -

l D+C 1218 1107 1 166 1075.5 -2.85 -9.46 -9.86  ;

D+C+B 1072 1614 1 242 1505.8 -6.70 -----

-8.27 D+C+B+A 898 1675 1 251 1552.5 -7.31 -----

-9.30 Cumulative Data from Control 5481 1 548 5181.8 -5.46 -----

-9.07

, Banks moving individually during dilution Cumulative Data from control 5481 1 548 5250.5 -4.21 -----

-9.17 Banks moving in overlap during boration i Conditions of Measurement:

s Hot Zero Power (547 F; 2235 psig) i i

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l 6.0 ARO HZP FLUX DISTRIBUTION, MODERATOR TEMPERATURE COEFFICIENT, AND BORON ENDPOINTS (FNP-2-ETP-3601) l t

PURPOSE The objectives of these measurements were to:

(1) determine the core flux distribution for the HZP l all-rods-out configuration; (2) determine the hot zero power isothermal and moderator temperature coefficients for the all-rods-out configuration; and (3) measure the i boron end point concentrations for the ARO, D-in, '

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

SUMMARY

OF RESULTS i

Table 6.1 gives a tabulation of the measured i boron end point concentrations compared with the

! design values for each rod configuration considered.

The design acceptance criterion for the all-rods-out

', critical boron concentration was satisfactorily met.

Table 6.2 is a tabulation of measured isothermal and moderator temperature coefficients for the all-rods-out configuration. The design acceptance criterion for

, the ARO isothermal temperature coefficient was met.

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TABLE 6.1 HZP BORON ENDPOINT CONCENTRATIONS Rod Configuration Measured CB Design-Predicted C (ppm) (ppm) B ARO 1387.0 1381 i 50 ppm

• D in 1273.0 1265 D+C in 1163.0 1148 D+C+B in 981.0 979 D+C+B+A in 815.5 803

• Design Acceptance Criterion.

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TABLE 6.2 IIZP ISOTiiERMAL AND MODERATOR TEMPERATURE COEFFICIENT Rod Configuration Boron Measured Calculated a,f Design Acceptance Concentration a a Criterion i T mod ppm pcm/*F pcm/*F pcm/*F All Rods Out 1387.5 -2.03 -0.13 -2.3 1 3 a,7 - Isothermal temperature coefficient, includes -1.9 pcm/*F doppler coefficient

'u - M derator only temperature coefficient mod C

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i 7.0 PQWER ASCENSION PROCEDURE (FNP-2-ETP-3605)

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

l 1. Ramp rate and control rod movement limitations l 2. Incore movable detector system final alignment

.j 3. Flux map at less than 50% power i

4. Adhering to the delta flux band during ascension to 75% power

5. Incore/Excore calibration at 75% power.

'l

SUMMARY

OF RESULTS i

, In compliance with Westinghouse recommendations

! and fuel warranty provisions, the power ramp rate was L]i limited to 3% of full power per hour between 20% and 100% power until full power was achieved for 72 cumula-l' tive hours out of any seven-day operation period. Control rod motion during the initial return to power was minimized, and the startup was conducted with the rods i withdrawn as far as possible. The rod withdrawal rate l was limited to 3 steps per hour above 50% power.

! Final alignment of the incore movable detector system was completed during power ascension (at power

, levels above 5%) prior to performing the flux max at

, 49% power.

Full core flux maps were taken at 49% and 80% power.

The results were within Technical Specification Limits and are summarized in Table 7.1.

An incore/excore calibration check was performed at 49% power. A full recalibration of the excore AFD channels was performed at approximately 75% power to comply with Technical Specification requirements. The incore/excore recalibration is described in section 8.0.

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TABLE 7.1 3

SUMMARY

OF POWER ASCENSION FLUX MAP DATA Parameter Map 34 Map 35 Date 12/06/82 12/08/82 i Time 18:34 15:52 i

i Avg. % Power 49.54 80.375

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Max. Fg (Z) 1.9654 1.8646 l Max. FAH 1.5025 1.4629

.1 j Max. Power Tilt * +1.0111 +1.0080 t

Avg. Core % A.O. +4.168 +5.009 f;

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

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. 8.0 INCORE-EXCORE DETECTOR CALIBRATION (FNP-2-STP-121) i PURPOSE _

The objective of this procedure was to determine the relationship 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 1 calibrating the control board and plant computer axial

j flu;; difference ( AFD) channels.

1

SUMMARY

OF RESULTS

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A preliminary verification of excore AFD channel calibration was performed at 49% power to insure AFD

could be kept within the target band during the ascension to 80% power. Flux maps for incore-excore recalibration were run at 75% - 80% power at average p arcent core axial offsets of + 17.909, +5.009, -9.806,

j and -17.873, as determined from the incore printouts.

i The measured detector currents were normalized to 100% power, and a least squares fit was performed to

obtain the linear equation for each top and bottom

+

detector current versus core axial offset.

l Using these equations, detector current data was generated and utilized to recalibrate the AFD channels and the delta flux penalty to the overtemperature AT setpoint. (See Figure 8.1) i i'

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' DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:

.l I-Top = 1.5357*AO + 292.2089 pa

1 I-Bottom = -1.7647*AO + 302.1177 pa CHANNEL 42:

, I-Top = 1.5517*AO + 287.7846 pa I-Bottom = -1.8000*AO + 298.4756 pa l

,l CHANNEL N43:

l I-Top = 1.5793*AO + 299.2300 pa I-Bottom = -1.8412*AO + 301.0934 pa CHANNEL N44:

I-Top = 1.5736*AO + 279.0687 pa I-Bottom = -1.8654*AO + 300.3199 pa 1

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9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-2-STP-115.1)

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 2 Technical Specifications.

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SUMMARY

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

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the average of six calorimetric heat balance measure-ments, the total core flow was determined to be 286,607.7 gpm, which meets the above criterion.

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