ML20093K735

Jm Farley Unit 1,Cycle 6 Startup Test Rept

ML20093K735
Person / Time
Site:	Farley
Issue date:	07/19/1984
From:	ALABAMA POWER CO.
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References
NUDOCS 8407310180
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-

ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER.1, CYCLE 6 STARTUP TEST REPORT PREPARED BY PLANT REACTOR ENGINEERING GROUP Approved:

d.0. U hL Technical Superintendent b

!A P1 Ma'@et' Date Approved:

7-/9- [Y 8407310180 840725 PDR ADOCK 05000348 P

PDR A

. =...

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4 TABLE OF CONTENTS Page 1.0 Introduction

-1 2.0 Fuel Inspection and Core Refueling 1

3.0 Control Rod Drop Time Measurement 6

4.0 Initial Criticality 8

5.0 ARO HZP Flux Distribution, Moderator Temperature Coefficient and Boron Endpoints 9

6.0 Control Rod and Boron Worth Measurements 12 7.0 Power Ascension Procedure 14 8.0 Incore - Excore Detector Calibration 17 9.0 Reactor Coolant System Flow Measurement 19 4

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1.0 INTRODUCTION

The Joseph M. Farley Unit 1 Cycle 6 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 6 core loading consists of 77 new and 80 reused 17 x 17 fuel assemblies as tabulated in 1 2.2.

Unit 1 began commercial operations in December 1, 1977 and completed cycle 5 on February 10, 1984 with an average core burnup of 11096.8 MWD /MTU.

2.0 FUEL INSPECTION AND CORE REFUELING References 1.

Westinghouse Refueling Procedure FP-ALA-RS 2.

Westinghouse WCAP-10525 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant Cycle 6) 2.1 Cycle 5 Fuel Inspection All fuel assemblies were unloaded from the reactor core to permit Control Rod Drive Mechanism Split Pin modification.

Each fuel assembly was visually inspected with binoculars during the core unload.

No significant defects or damage were noted during the visual inspection.

The visual inspection was followed by vacuum sipping of each fuel assembly.

General Electric gaseous vacuum sipping equipment was utilized for this testing.

Final sipping results showed evidence that two assemblies (F-36 and E-09) leaked and another assembly (F-44) was suspected of leaking.

Assembly E-09 was not scheduled for use in Cycle 6.

Assemblies F-36 and F-44 had originally been scheduled for reuse in Cycle 6; therefore, two D assem-blies which had not been used in Cycle 5 were substituted in baffle positions in Cycle 6.

Additionally, F-36, E-09, F-44 and ten other assemblies were tested for leakage using an ultrasonic method developed by Brown Boveri Reaktor of Germany, Failed rods were confirmed during ultrasonic testing of i

F-36 and E-09.

l F-44 and the other ten assemblies tested showed no evidence of leakage by the ultrasonic method.

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m TV visual inspections were also performed on baffle assemblies, F-36, E-09 and F-44.

The only significant defect noted during the TV inspection was a clad crack on a peripheral rod in assembly E-09.

This failure was a

"T" shaped crack where the top end plug connects to the fuel rod.

2.2 Cycle 6 Core Refueling The Cycle 6 core loading commenced on March 27, 1984 following the completion of Control Rod Drive Mechanism Split Pin modifications, and was completed on March 29, 1984.

The as-loaded Cycle 6 core is depicted in Figures 2.1 - 2.3.

The number of assemblies in the various regions of the Cycle 6 core is tabulated below:

No. of Fuel Region Assemblies 4

2 6

38 7

40 8A 44 8B 33 Fuel assembly inserts consist of 48 full length control rods, 2 secondary sources, 49 burnable poison rod inserts, and 58 thimble plug inserts.

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FIGURE 2.1 ALA CYCLE 6 LOADING PATTERN RPNMLKJHG F

EDCBA 1

G-23 H-03 G-35 l

J-4 F

G-4 F-04 H-16 H-47 G-24 H-54 H-19 F-24 2

D-8 F'

F F-7 F

F H-4 Gj3 G-09' H-68 H-46 D-16 F-03 H-57 H-67 G-36

}

L-6 F

F N-6 J-6 C-6 F

F E-9+

F-01 H-31 H-76 G-13 H-33 G-06 H-02 G-25 H-49 H-41 F-22 K-5 F

F L-4 F

B-7 F

E-4 F

F F-5 l

F-26 H-62 H-63 F-13 H-08 F-48 H-26 F-49 H-27 F-16 H-59 H-60 F-09 5

G-8 F

F H-2 F

L-3 F

E-3 F-B-8 F

,F H-7 H-23 H-53 G-31 H-24 F-40 H-15 G-05 H-'18 F-42 H-01 G-04 H-56 H-30 g

i F

F M-5 F

K-2 F

J-2 F

F-2 F

0-5 F

F G-28 H-45 G- 07 H-14 F-34 H-10 G-26 F-46 F-45 H-38 F-52 H-36 G-19 H-66 F-10 7

M-7 F

K-3 F

N-5 F

D-7 R-8 B-6 F

C-5 F

F-3 F

E-6

/

H-29 G-29 G-10 G-32 H-44 G-02 F-35 H-52 F-50 G-ll H-28 G-21 G-39 G-01 H-25 8

F K-7 J-10 P-7 F

J-14 H-1 F

H-15 G-2 F

8-9 G-6 F-9 F

F-14 H-51 G-30 H-35 F-41 H-09 F-39 F-43 G-18 H-40 F-33 H-07 G-27 H-73 G-34 L-10 F-K-13 F

N-ll F

B-10 A-8 M-9 F

C-11 F

F-13 F

D-9 9

H-17 H-71 G-15 H-34 F-37 H-22 G-22 H-37 F-51 H-39 G-16 H-65 H-43 in F

F M-11 F

K-14 F

G-14 F

F-14 F

D-11 F

F IU F-12 H-77 H-64 F-21 H-04 F-47 H-42 F-38 H-05 F-25 H-61 H-70 F-29 J-8 F

F P-8 F

L-13 F

E-13 F

H-14 F

F H-9 F-08 H-12 H-50 G-08 H-06 G-17 H-13 G-12 H-75 H-21 F-07 K-ll F

F L-12 F

P-9 F

E-12 F

F F-11 G-j8 G-37 H-69 H-58 F-28 D-01 H-48 H-55 G-14 3

l}

L-7 +. F F

N-10 G-T0 C-10 F

F E-10 F-23 H-20 H-72 G-20 H-74 H-11 F-11 p

M-8 F

F K-9 F

F H-12, i

G-33 H-32 G-4 I5 l

J-12 F

G-12 l

xx Assembly ID y

Previous Cycle Location ss Secondary Source Location

+ Cycle 4 Location 3

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FIGURE 2.2 CONTROL R0D LOCATIONS R

P N

M L

K J

H G

F E

D C

B A

c I

A D

A 2

6

'6 A

A SP 3

C B

B C

4 3p SP B

SP 6 8 5

A B

D C

D B

A 6

6 6

b b

A B

B Sp A

7 i' (

D Sp C

Sp C

Sp D

8 b

b b

b A

Sp B

B A

9 A

B D

C D

B A

10 1

6 6

8 SP 8

SP II C

B B

C 12 gg SP b

b A

A I3 i

A D

A 14 Absorber Material, Ag-In-Cd Function Mer of Clusters Control Bank D Control Bank C 8

Control Bank 8 8

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' Control Bank A 8

Shutdown Bank S 8

B 8

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Shutdown Bank S SP(SpareRodL$ cations) 8 13 4

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FIGURE -2.3 BURNABLE POISON AND SOURCE ASSEMBLY LOCATIONS R

P N

M L

K J

H G

F E

D C

B A

l 2

R-SS 8

3 12 8

8 12 4

12 8

12 8

12 5

8 8

12 12 8-8-

6 i

8 12 4

L 12 8

7 12 4

16 4

12 8

(

8 12 4

12 8

9 8

8 12 12 8

8-10 12 8'

12 8

12 ll 12 8

8 12 12 8

SS 8

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14 15 SS Secondary Source 464 Fresh Standard bps 5

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3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-1-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 (Tavg > 541*F and all reactor coolant pumps operating) Technical Specification 3.1.3.4 requires that the rod drop time from the fully with-drawn position shall be < 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 1.81 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 Dachpot Bottom Hot Full-flow 1.626 sec.

2.173 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 I

stationary, lift and movable gripper coils were examined and rod stepping speed measurements were conducted.

All results were satisfactory.

6 i

A FIGURE 3.1 NORT!!

UNIT 1 CYCLE 6 y

90*

R 1.64 1.64 1.65 Y

2.20 2.17 2.20 P

N 1.64 1.61 N

s 2.18 2.' l E

[

1.59 1.60 1.63 1.63 M

2.14 2.14 2.19 2.20 1.60 1.62 L

2.14 2.18 l.73 1.57 1.64 1.54 1.63 1.64 1.65 g

2.31 2.12 2.18 2.09 2.15 2.22 2.22 g

s 1.61 1.55 1.55 1.60 s'

2.16 2.05 2.09 2.19

-J s\\

0*

1.68 1.59 1.61 1.63 s'

i 2.24 2.16 2.15 2.17 180*

H 1.67 1.56 1.58 1.61 2.21 2.09 2.11 2.15 1.71 1.63 1.62 1.59 1.60 1.63 1.61 F

2.25 2.19 2.15 2.10 2.13 2.2'l 2.16 1.58 1.65

_g 2.13 2.19 1,63 1.61 1.57 1.62 D

2.17 2.15 2.09 2.16 1.61 1.64 C

2.16 2.17 l.74 1.72 1.81 2.30 2.25 2.39 B

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y 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

DRIVE LINE " DROP TIME" TABULATION i

TEMPERATURE -

547 F

PRESSURE-2100 psia 4 FLOW -

100 X.XX BREAKER " OPENING" TO DASHPOT ENTRY - IN SECONDS X.XX BREAKER " OPENING" TO DASHPOT BOTTOM - IN SECONDS DATE-4/20/84 i

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

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 tests, and operationally verify the calibration of the reactiv-ity computer.

Summary of Results dilution mixing at 0616 hours0.00713 days <br />0.171 hours <br />0.00102 weeks <br />2.34388e-4 months <br /> on April 22, Initial Reactor Criticali 1984.

The reactor was allowed to stabilize at the following critical conditions:

RCS pressure - 2235 psig, RCS.t mediate range power - 1.3 x 10-gmperature - 545

• F, inter-amp, RCS boron concentration

- 1807 ppm, and Control Bank D position - 185 steps.

Follow-ing stabilization, the point of adding nuclear heat was deter-mined and a cileckout of the reactivity computer using both positive and negative flux periods was successfully accom-plished.

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.

4

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5.0 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT, BORON ENDPOINT AND FLUX DISTRIBUTION Purpose The objective of these measurements was to: (1) determine the hot, zero power isothermal and moderator temperature co-efficients for the all-rods-out (ARO) configuration; (2) measure the ARO boron endpoint concentration, and (3) determine the hot, zero power ARO flux distribution in the reactor core.

Summary of Results The measured ARO, hot zero power temperature coefficients and the ARO boron endpoint concentration are shown in Table 5.1.

The moderator temperature coefficient (MTC) was found to be positive (+ 4.55 pcm/*F) as expected from the core design. ; Technical Specification 3.1.1.3.a was changed before the outage to allow a maximum MTC of

+5.0 pcm/*F.

The design acceptance criterion for the ARO critical boron concentration was satisfactorily met.

(See Table 5.1.)

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 dif-ferences between measured and design-predicted relative assembly power satisfied the design criteria for the maximum positive percent error, but not for the maximum negative percent error.

The design criteria states that the percent error between measured and expected relative fuel assembly powers should be within i 10% for assemblies with relative powers > 0.9, and within 115% for assemblies with relative powers T O.9.

Two assemblies failed to meet these criteria:

in core position L-ll (relative power = 0.955) showed anThe assembly error of -10.7%, and the errer for the assembly in location L-12 (relative power = 1.123) was also -10.7%.

In addition the HZP, ARO flux map indicated that the incore tilt exceeded the design criterion of 1.02. (See Table 5.2.)

Westinghouse was notified of the assemblies that failed the relative power criteria and of the incore tilt exceeding 1.02.

Westinghouse agreed that power escalation could continue up to 75% power. In addition, a rod insertion limit of D at 150 steps was recommended as long as the incore tilt exceeded 1.02.

Subsequent flux maps at 35% and 44% full power indicated i

that the percent difference between measured and expected assembly power decreased to well within the design acceptance criteria in both assemblies.

These two full core flux maps also indi-cated that incore tilt had decreased to below 1.02.

t 9

TABLE 5.1 ARO, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT Rod Configuration Boron Measured Calculated Design Concentration a

a a

(ppm)

(hem /*F)

(h8N/*F)

(h8M/*F)

All Rods Out 1799

+1.94

+4.55

+4.05 7 - Isothermal temperature coefficient 0

Mod - Moderator only temperature coefficient a

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

Design-predicted CB (ppm)

All Rods Out 1804 1792 1 50 1

4 1

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TABLE 5.2 RESULTS OF HZP,-ARO FLUX DISTRIBUTION MAP A.

FAH percent error between measured and design-predicted values versus relative assembly power Pg of assembly i.

P Item value i

Design Criterion Maximum positive

+15.0%

0.564 1 15% for Pg < 0.9 percent error Maximum negative

-10.7%

1.123 1 10% for Pg > 0.9 percent error B.

Incore Quadrant Tilt:

Maximum Incore Tilt Design Criterion 1.0528*

< 1.02

• The measured incore tilts at 35% and 44% power were 1.0183 and 1.0136, respectively.

i l

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6.0 CONTROL AND SHUTDOWN 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 comparision 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.

a 12

1 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 513 1 77 517.4

+0.9 Control B (Ref.)

1401 1 140 1353.5*

-3.4 Control C 1018 i 153 956.0

-6.1 s

Control D 1056 i 158 1006.4

-4.7 to Shutdown A 959 i 144 949.0

-1.0 i

Shutdown B 1079 i 162 991.9

-8.1 i

All Banks Combined 6026 i 603 5774.2

-4.2 I

• Measured by dilution method l

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7.0 POWER ASCENSTOM PROCEDURE (FNP-1-ETP-3605)

Purpose The parpose of this procedure was to provide control-ing instructions for:

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1.

Ramp rate and control rod movement limitations 2.

Incore movable detector system final alignment t

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 t

i 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 ammelative hours out of any seven-day operating period.

Control cod motion during the initial return'to power was minimized, and the startup was

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conducted with the rods withdrawn as far as poEsible.

In accordance with Westinghouse recompeudutions a rod insertion limit of 150 steps was established on Control Bank D.

This was necessary due to the 1.0528 incore tilt indicated by the HZP flux map.

The 35% power flux map incore tilt was below 1.02 and the insertion limit 5

was discontinued.

Design-predicted NIS detector currents equal to i

80% of the Cycle - 5 values were used for initial reactor i

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.

i Full core flux maps were taken at 35 The results for the first two finx maps m%, 44% and 78% power.

et all Technical Specification Limits.

The 78% power map was performed while under the exception of Technical Specification 3.2.1.a.2.b for Incore-Excore recalibration.

The results of these maps are summarized in Table 7.1 i

An incore/excore calibration check at 35% power indi-cated that a preliminary redetermination of the incore/excore intercept currents was necessary.

This calculation was 14 1

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-s perfor:ned and new current values were issued to calculate qttadrant power tilts., A full recalibration of the excore AFD channels was perforiaed at approximately 78% power to

' comply with Technical Specification requirements.

When 100% power was reached, the'excore, ambient tilts had to be.rezeroed due to a shif t in core ' axial tilt.

The Incore-Excore recalibration is described in section 8.0.

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TABLE 7.1

SUMMARY

OF POWER ASCENSION FLUX MAP DATA Parameter Map 137 Map 139 Map 140 Date 4/27/84 5/3/84 5/5/84 Time 18:00 05:51 04:30 Avg % Power 34.97 43.86 77.89 Max. Fg (Z) 2.1459 2.0725 2.0255 Max. FAH 1.5722 1.5181 1.5329 Max. Power Tilt

• 1.0183 1.0136 1.0201 Avg. Core % A.O.

+4.711

+7.148

+6.184 h

• Calculated power tilts based on assembly FAHN from all assemblies, i

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q INc5RE-EXCOREDETECTORCALIBRATION i

8.0 (FNP-1-STP-121)

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'1h'b objectiJe, of this-procedure.was Ao determine the rel'ation, ship between power range upper and lower excore

. detector currents and incore axiab offset for.the purpose

/ of calibrating the' delta flux penalty,to thc..overtemperature AT piotection, system, and for calibrating the control board andsplant ccaputetr axial flux - di fference,( APD) channels.

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, hindicated new 100Goormalized zero' axial' offset currentsQuadrant p

~ 9 needed toche calc 31sted.

These calculations-were completed s

according to Appendix C of FNP-1-STP-121.~ ~ Subsequent quadrant e

power tiltTcalculations were performed using the new detector

' current. values with satisfactory results.

The Power Range

. Axial offset calibration.. check STP-121, was, performed at 35%

power.' This procedure verified indicated axial offset was

" within'three yercen,teo'f the actual incere. axial offset.

Therefore, att interim incore-excore calibration was not

,'incoterexco're recalibration. required End powerMas increased to 78%

r Flux maps for incore-excore L

',recaliktation were'rpn at approximately 78% power;at average

' percent cort axial offsets of +ti.184, -12.002, -19.743, and

+16.807,~1as detei3ttined from the Incore printouts.

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i Tho/measuspf Qa.tector, currents were normalized ~to 100%

pwn; and'a least f:quares fit was-pt.rforraed to obtain the linear equation'for each top'and bottem detector current versus core axial offret.y Using these~ equations, detector, current data was generated and Titilized to recalibrate; the AFD channels and the delta flux penalty for 'the. overtemperature AT, setpoint.

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't During power, ascension,la channel. deviation alarm

' Investigation revealed that the ambient core axial occured.

tilt that was prssent when' performing' the incore-excore calibration had shifted at 100% power and a new~i A full core ~ flux map was performed

~

core axial' offset value was obtained.

Using this new value the excore detedtor equations derived a

, using% power were nor,malized to the new core axial offset at_78 the method prescribed in Appendix C of FNP-1-STP-121.

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The refinements made to the~ original recalibration equations arcipresented in Ftgure 8.1.

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

I - Top

= 1.0127

• A.O. + 192.73 I - Bottom = 1.0556

• A.O. + 190.14 CHANNEL N42:

I - Top

= 1.0263

• A.O. + 186.60 1 - Bottom = 1.0718

• A.O. + 182.01 CHANNEL N43:

I - Top 0.9892

• A.O. + 184.15

=

I - Bottom = -1.1010

• A.O. + 197.91 4

CHANNEL N44:

I - Top

= 0.9619

• A.O. + 174.73 I - Bottom = 1.0645

• A.O. + 174.79

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e 9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-1-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 requirements given in the Unit 1 Technical Specifications.

S_ummary 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 three loop operation.

265,500 gpm for heat balance measurements,From the average of six calorimetric the total core flow was determined to be 284,074.8 gpm, which meets the above criterion.

19

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