ML20105B954

Startup Test Rept,Unit 1,Cycle 5

ML20105B954
Person / Time
Site:	Farley
Issue date:	06/23/1983
From:	Clayton F, Nesbitt C ALABAMA POWER CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 8307070080
Download: ML20105B954 (23)

Text

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.. g ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER 1, CYCLE 5 STARTUP TEST REPORT PREPARED BY PLANT REACTOR ENGINEERING GROUP d

' l APPROVED:

C O. I ~

Technical Superintendent l

[h- , [rc Q in Plant Manager m m ePaavEo 643h3 y

.! / y i DISK: CYCLE 2 EGO 7070000 830623 PDR ADOCK 05000348

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P PDR i k

TABLE OF CONTENTS PAGE 1.0 Introduction 1 2.0 Fuel Inspection and Core Refueling 2 3.0 Control Rod Drop Time Measurement 7 4.0 Initial Criticality 9 5.0 ARO HZP Flux Distribution, Moderator Temperature Coefficient and Boron Endpoints 10 6.0 Control Rod and Boron Worth Measurements 13 s

, 7.0 Power Ascension Procedure 15 8.0 Incore-Excore Detector Calibration 17 9.0 Reactor Coolant System Flow Measurement 19

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1.0 INTRODUCTION

The Joseph M. Farley Unit 1 Cycle 5 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 assumed in the FSAR safety analysis.

Unit 1 of the Joseph M. Farley Nuclear Plant is a Three Loop Westinghouse pressurized water reactor rated at 2652 MWth. The Cycle 5 core loading consists of -

157 17 x 17 fuel assemblies.

Unit 1 began commercial operations on December 1, 1977, and completed Cycle 4 on January 14, 1983 with an average core burnup of 10621.7 MWD /MTU.

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2.0 -FUEL INSPECTION AND CORE REFUELING REFERENCES

1. Westinghouse Refueling Procedure FP-ALA-R4 4

2. Westinghouse WCAP 10308 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant Cycle 5) 2.1 Cycle 4 Fuel Inspection During cycle-3 operation reactor coolant

. activity levels indicated a small number of fuel rod failures in low burnup fuel assemblies. Since several other plants had recently found damaged fuel rods adjacent to baffle center injection joints, it was decided to perform TV Visual inspection of these assemblies and to peen these joints at the end-of-cycle-3 outage. This procedure had been i

used at other plants. Although TV examination showed no damage on the assemblies located at the center injection joints, peening was performed.

At the beginning of cycle-4 RCS activity levels increased significantly but remained below Tech.

Spec. limits. Based upon experience at other plants which had peened at the center injection core baffle joints, it was assumed that fuel rod damage had occurred due to jetting through gaps at the corner injection baffle joints opened by the center

. injection joint peening process. Preparations were therefore made to conduct an extensive fuel 4

inspection campaign to' preclude' the use of damaged assemblies in cycle-5. The inspection programs consisted of a binocular visual examination fol-lowed by vacuum sipping of all assemblies which

'showed no evidence of physical damage and which

were to be returned to the cycle-5 core. TV i ' visual inspection (high/ low magnification) was used in cases where binocular inspection was inconclusive. General. Electric gaseous vacuum sipping equipment was utilized. Additionally all assemblies that were located at-baffle joints (center or corner injection) during cycle-4 that passed both' sipping and TV visual inspection were examined for wear damage beneath the grid-l- straps using Westinghouse SULO probe _ equipment (a differential strain gage technique).

Eleven one-cycle region F assemblies showed gross damage due to baffle. jetting. All damage was.

L .at corner injection joints and was limited almost en-tirely to the top fuel rod span.between grids 7 and 8.

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Damage was limited generally to one or more of the first five fuel rods in from the corner.

The assemblies which suffered gross damage from baffle jetting were F02, FOS, F06, F15,3F17, F18, F19, F20, F30, F31, and F32. Assembly E50 was found to have a hydride blister at the upper end cap weld on one fuel

' rod and was rejected for use in cycle-5. E50 was later found to fail the sipping test. Three assemblies were determined to be leaking from the sipping and were eliminated from use in cycle-5. These were F27 located at a center injection joint, E-14 which was located at a center injection joint in cycle-3, and D34 which was located at a center injection joint in cycle-2.

2.2 Cycle-5 Core Refueling The C following'ycle-5 core loading the completion of commenced fuel inspection on 3/2/83 and debris cleanup activities, and was completed on 3/4/83. The as-loaded Cycle-5 core is depicted in Figures 2.1 - 2.3, which shows the location of each fuel assembly and insert, and gives the assembly enrichments. The number of assemblies in the various regions of the Cycle-5 core is tabulated below:

4 Region No. of Fuel Assemblies 1 4 3 5 4 16 4A- 2 5 50

6 40

7 40 Fuel assembly inserts consist of 48 full length

>ontrol rods, 2 secondary sources, 51 burnable ~

poison inserts, and 56 thimble plug inserts.

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Figure 2.1 AL A Umt 1. Cyclu 5 Rulcrence Lowlmy Pattern R P N M L K J H G F E D C B A l

180u A 34 F 35 A 18 I

nR F' H2 0 19 F 40 G 05 F 13-' G 11 F42 0 39

.; .,; 2 L 10 F F l. 2 F F E 10 0 15 F48 G 07 E 09 E 21 E 27 G 19 F 49 D-07

- .- SS 7 g: 3 FM 4

F 13 F F L-3 H1 E3 F K 13 D 27 E-04 G 13 E 37 G 23 G 35 E 45 F 21. G 25 C 32 D 09

~. 5, n :.; 4 J4 C 10 F M5 F Fl2 ' F D-5 F G 11 N 10 D-05 F 34 G 31 E 31 E 16 E40 F-Q1 E-08 F-22 E-29 G 04 F 52 D47

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• r^ i- 5 F5 F F M-7 R7 J2 .

J6' G-8 A7~ D7 F F 7' K-5 F 38 G 3G E 38 F 03 E 02 G D3 E49 G 39 E-48 F 10 E 33 G 09 f,

i F L4 .. .. t 6 F J1 H 14 F H7. F H2 '%.1

.O i:E4 F F A :'.

! C 49 G 32 E 24 G 28 E 20 G 29 E 43 F Op l

,, E 36 G 24 E 34 G 26 E 03 G 06 C 14 l 59 F N5 K7 r-a

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o 7 F F P7 P 6 B7 F F7 F C5 F 45 F 21 F L9 E 11 F 23 20 3 E 22 ( 12 E 10 f 26, E 41 ZD4

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F-04 E-47 f 16 [33 F P 11 R-8 K-2. N8 J8 86' G2 G8 C8

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4 .) 2700- 8 C 28 G 17 M 13 K 14 A8 .::5.{dle 85' FSN E 15 G 18 E 13 G 20 E 44 F *29

++ E-4 G G 01 E 17 G 34 E 07 G 21 C-43 E7 F

.:. H.E 9 N 11 F K9 F P9 F10 B9 F F9 F C 11 F L7 F 44 G 14 E 39 F 14 E 32 G 10 E 28 G-38 E 35 .F 28, E 19 G 37 F 39 F' F L 12 315 P8 F H9- F B8

. w G 15 E 12 F w1 F h* S 10 D-45 F-41 G 15 E-25 F 0,3 E 23 E 30 E-01 F 07 E 26 G 16 F 33 D-37

,,,  : c *. --

i.. .,l F 11 F F M9 R 9' 11 J 10 J 14 G 10 A 97 D-9 F F. >a K 11 D 42 E 18 G 08 E 51 G 33 F 11 G-40 l

• E 05 G 12 E 12 D 11 C-6 J 12 F M 11 F F 14 ' 12 F D 11 F G 12 N6 l

0 24 F-47 G 30 E 52 E-06 E 42 G 27 F 38 D-03 F3 F F SS f ,; 13 L 13 H 15 E 13 _F F-K3 D 33 F 37 G 02 F 25 G 22 5 51 D 38 t:G F F 14 1 E 14 F F E6 A 43 F 60 A 35 XX Represents Cycle 1 Location * +

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++ XX Represents Cycle 3 Location i

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,.,of ID ID ID ID ID

" r Code ID ID Legend:

XX XX XX XX XX XX XX ID: Assembly identification Aegion 3 4 i 1 4A S G 7 From Cycle ZD: Demonstration Assembo 1 3 1A 4A 4A 4A/ FEED FEED XX: Previous Cycle Location.

WiO U 235 2.115 3.102 3.113 3.108 2.801 2.995 3.00 SS: Secondary Source 4

Figure 2.2

( CONTROL ROD LOCATIONS RPNM L KJ HG F EDCB A I

A ID A SA SA SP C B SP B C SP SB SP SB 4 5 A B D C D B A SA SB SB SP .SA D SP C SP C SP D b SA l SP SB SB SA 9 A B D C D B A 10 SB SP SB SP  !!

C B SP B C SP SA SA Absorber Material: A D A id

• Ag-In-Cd ;r Con ro nk D 8 C ntrol Sank C 3 Centrol Bank B e Control Bank A 8 Shutdown Bank S o g ShutdwnBankS{ 8

~S? (Spare Rod Locations) 13 Figure c:NTROL ROD LOCATIONS 5

Figure 2.3 Source Assembly Locations RPNM LKJ HG F EDCB A I

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,4 2 12 SS 12 3 8 12 4 12 8 8 4 4 8 $

12 4 12 12 4 12 b .

( 4 12 12 12 12 4 7

+

s 8 4 12 12 12 12 4 9 12 4 12 12 4 12 10 8 4 4 8 Il 8 12 4 12 8 !2 12 SS 12 [3 4 4 15 l

SP Spent BP Assembly 424 Fresh Standard bps l ('

SS Secondary Source 12 Spent Standard bps 6

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, 3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-1-STP-ll2)

PURPOSE The purpose of this test was to measure khe drop time of all full length control rods under hot-full flow conditions in the reactor coolant system to insure com-pliance with Technical Specification requirements.

SUMMARY

OF RESULTS For the Hot-full condition (T > 541*F and all reactor coolant pumps operating) T83Hnical Specification 3.1.3.4 requires that the rod drop time from the fully withdrawn position shall be < 2.2 seconds from the begin-ning of stationary gripper coil voltage decay until dash-pot entry. All full length rod drop times were measured to be less than 2.2 seconds. The longest drop time recorded was 1.8 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 CONDITIONS DASHPOT ENTRY DASHPOT BOTTOM Hot-full flow 1.628 sec. 2.182 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 step-ping 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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. Figure 3.1 l

UNIT 1 CYCLE 5 4

NORTH I

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1.64 2.20 1.63 2.17 1.60 2.15

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P h 1.63 2.20 1.65 2.20 [

g 1.59 1.60 1.63 1.65 2.15 2.17 2.22 2.20 M 1.59 '1.60

'2.13 2.17 N (

1.70 1.58 1.67 1.60 1.66 1.65' 1.60 x 2.25 2.15 2.22 2.15 2.22 2.22 E 2.15 h j

g l.60 1.55 1.57 . 1.60 -g g 2.17 2.10 2.12 2.17

% 1.65 1.64 1.60 '

1.63 o Oo 2.20 180 1

2.20 2.15 2.18 -M 1.66 1.57 1.58 1.64 _,g 2.22 2.12 2.12 2.19

' 1.68 1.63 1.66 1.58 1.59 1.63 1.59 2.22 2.20 2.21 2.10 2.14 2.20 2.15 I 1.61 1.64 2.17 2.21 E l

1.66 1.63 1.60 1.60 D 2.20 2.20 2.16 2.16 1.61 1.65 2.17 C 2.21 1.70 1.72 '1.80 2.22 8 2.19 2.38 I >

8 .A 270

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l 15 14 13 12 11 10 9 8 7 6 5 4 3 2 i DRIVE LINE "0 ROP TIME" TABULATION TEMPERATURE . 5470F PRESSURE - 2235 psig g Flow .100 t

X.XX BREAKER *0PENING" TO DASHPOT ENTRY - IN SECOMOS 3-27-83 DATE -

X.XX BREAKER "0PENING" TO DASHPOT BOTTOM - IN SECOMOS 8

l 4.0 INITIAL CRITICALITY (FNP-1-ETP-3601)

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PURPOSE The purpose of this procedure was to achieve initial reactor criticality under carefully controlled conditions, establish the upper flux limit for the conduct of zero power physics test, and operationally verify the calibra-tion of the reactivity computer. -

SUMMARY

OF RESULTS Initial Reactor Criticality for Cycle 5 was achieved during dilution mixing at 0614 hours0.00711 days <br />0.171 hours <br />0.00102 weeks <br />2.33627e-4 months <br /> on March 28, 1983.

The reactor was allowed to stabilize at the following critical conditions: RCS pressure - 2229 psig,_gCS tem-perature 547*F, intermediate range power 8 x 10 amp, RCS boron concentration 1333 ppm, and Control Bank D position 170 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 sucessfully 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 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT, BORON

( ENDPOINT AND FLUX DISTRIBUTION (FNP-1-ETP-3601)

PURPOSE The objectives of these measurements were to:

(1) determine the hot, zero power isothermal and moder-ator temperature coefficients for the all-rods-out (ARO) configuration; (2) measure the ARO boron endpoint con-centration; and (3) determine the hot, zero power ARO flux distribution in the reactor core.

SUMMARY

OF RESULTS The measured ARO, hot zero power temperature co-efficients and the ARO boron endpoint concentration are shown in Table 5.1. The moderator temperature coef-ficient was found to be slightly positive (+0.38 pcm/ F).

The NRC was notified by special report and control rod withdrawal limits were established in accordance with Technical Specifications to maintain a negative MTC during normal plant operation. The design acceptance criterion for the ARO critical boron concentration was satisfactorily met.

' Following the control and shutdown bank worth measurements (Section 6.0) a flux distribution map was obtained at the ARO configuration. As summarized in Table 5.2, the differences between measured and design-predicted relative assembly powers satisfied the design criteria, but incore tilt was in excess of 1.02.

Westinghouse was immediately notified in accordance with their recommended policy and the plant commenced power escalation. The next two flux distribution maps, taken at 48% and 78% power, demonstrated that incore tilt had decreased to below 1.02.

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e TABLE 5.1 ARO, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT

'R:d Configuration Boron Measured Calculated aTDesign Acceptance Concentration a Criterion T " mod pcm/*F ppm pcm/*F pcm/*F All Rods Out 1347.5 -2.32 +0.38 -3.2 1 3 .

e - Is thermal temperature coefficient, includes -2.7 pcm/*F doppler coefficient T

a - M derator only temperature coefficient mod C

ARO, HZP BORON ENDPOINT CONCENTRATION Rod Configuration Measured CB (ppm) Design. predicted CB (ppm)

All Rods Out 1357.5 1325 i 50

TABLE 5.2 l'

RESULTS OF HZP, ARO FLUX DISTRIBUTION MAP A. FAH percent error between measured and design - predicted values versus relative assembly power Pi of assembly i.

Item Value Pi Design Criterion Maximum positive percent error +10.2% 0.783 i 15% for Pi < 0.9 Maximum negative percent error -7.6% 1.062 i 10% for Pi > 0.9 B. Incore Quadrant Tilt:

Maximum Incore Tilt Design Criterion 1.037* < l.02

• Corrected value obtained by subtracting design asymmetry (1.0127) from the measured core tilt (1.0497). The (un-corrected) measured core tilts at 48% and 78% power were 1.0168 and 1.0135, respectively.

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6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS ,

( (FNP-ETP-3601) l 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; and (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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TABLE 6.1

SUMMARY

OF CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS Predicted Bank Measured Worth & Review Bank Percent Bank Criteria (pcm) Worth (pcm) Difference Control A 630 i 94 658.7 +4.6 ,

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Control B (Ref.) 1301 i 130 1292.3* -0.7 Control C 732 i 110 715.3 -2.3 i Control D 1237 1 185 1196.4 -3.3

, g Shutdown A 1147 1 172 1129.1 -1.6

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Shutdown B 973 146 943.6 -3.0

! All Banks Combined (020 1 602 5935.4 -1.4

• Measured by dilution method i

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i 7.0 POWER ASCENSION PROCEDURE (FNP-1-ETP-3605)

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

1. Ramp rate and control rod movement limitations

2. Incore movable detector system final alignment

3. Flux map at less than 50% power

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

5. Incore/Excore calibration at 75% power.

SUMMARY

OF RESULTS In compliance with Westinghouse recommendations and fuel warranty provisions, the power ramp rate was limited to 3% of full power per hour between 20% and 100% power until full power was achieved for 72 cumula-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 withdrawn as far as possible. The rod withdrawal rate was limited to 3 steps per hour above 50% power.

t 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 48% power.

Due to the lower neutron leakage of the cycle 5 core, design-predicted NIS detector currents equal to 80% of the Cycle-4 values were used for initial reactor trip and rod stop setpoints. At 30% power, detector current readings and calorimetric data were obtained to verify the adequacy of the initial settings and to provide data for rescaling the NIS intermediate range setpoints.

Full core flux maps were taken at 48% and 76% ower.

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

An incore/excore calibration check was performed at 48% power with satisfactory results. A full recalibration of the excore AET channels was performed at approximately 75% power to comply with Technical Specification require-ments. The incore/excore recalibration is described in l section 8.0.

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TABLE 7.1

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SUMMARY

OF POWER ASCENSION FLUX MAP DATA Parameter Map 119 Map 120 Date 4/1/83 4/2/83 Time 05:00 11:47 Avg. % Power 48.82 75.87 Max. Fg (Z) 2.0322 1.9416 Max. FAH 1.4863 1.4634 Max. Power Tilt * +1.0168 +1.0135 Avg. Core % A.O. +2.922 -0.030

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

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8.0 INCORE-EXCORE DETECTOR CALIBRATION (FNP-1-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 calibrating the control board and plant computer axial flux difference (AFD) channels.

SUMMARY

OF RESULTS A preliminary verification of excore AFD channel calibration was performed at 48% power to insure AFD could be kept within the target band during the ascension to 76% power. Flux maps for incore-excore recalibration were run at approximately 75% power at average percent core axial offsets of + 15.097, -0.030,

-18.181, and -25.916, as determined from the incore printouts.

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.

I 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) t 17

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FIGURE 8.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:

I-Top = 1.0393*AO + 194.3945 pa I-Bottom = -1.0822*AO + 194.2360 pa CHANNEL 42:

I-Top = 1.0609*AO + 189.6089 pa I-Bottom = -1.1447*AO + 183. 9130 p a CHANNEL N43:

I-Top = 0.9731*AO + 180.6990 pa I-Bottom = -1.1589*AO + 193.4700 pa e CHANNEL N44:

I-Top = 1.0932*AO + 184.6970 pa I-Bottom = -1.1680*AO + 190.1275 pa t 18

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9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-1-STP-ll5.1)

PURPOSE The purpcse 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 Specifications.

SUMMARY

OF RESULTS To comply with the Unit 1 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 measure-ments, the total core flow was determined to be 283,178.5 gpm, which meets the above criterion.

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• Alabama P wer Com*),any ~

• 600 North 18th Street l Post Offica Box 2641 Birmingharn. Alabama 35291 Telephone 205 250-1000 F. L CLAYTON, JR.

j Senior Vice President Alabama Pbwer 11e souttwm electre system June 28, 1983 Docket No. 50-348 j

Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. S. A. Varga J. M. Farley Unit 1 Startup Report Gentlemen:

Enclosed is the Startup Report for Unit 1 Cycle 5 as required by the March 17, 1983 letter from F. L. Clayton, Jr. to Mr. S. A. Varga.

If you have any questions, please advise.

Yours very truly, I

A F. L. Clayton Jr. I FLC,Jr/MDR:cl Enclosure xc: Mr. R. A. Thomas Mr. G. F. Trowbridge Mr. J. P. O'Reilly Mr. E. A. Reeves Mr. W. H. Bradford l

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~ Director, Nuclear Reactor Regulation June 28, 1983 bc: Mr. W. O. Whitt Mr. R. P. Mcdonald Mr. H. O. Thrash Mr. O. D. Kingsley, Jr. (w/ Enclosure)

Mr. W. G. Hairston, III (w/ Enclosure)

Mr. K. W. McCracken (w/ Enclosure)

Mr. J. W. McGowan Mr. R. G. Berryhill Mr. D. E. Mansfield Mr. W. G. Ware Mr. J. R. Crane Mr. L. B. Long t