ML20086N136

Alabama Power Co Jm Farley Nuclear Plant Unit 2,Cycle 3 Startup Test Rept

ML20086N136
Person / Time
Site:	Farley
Issue date:	02/10/1984
From:	Clayton F ALABAMA POWER CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 8402170243
Download: ML20086N136 (23)

Text

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ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER 2, CYCLE 3 STARTUP TEST REPORT PREPARED BY PLANT REACTOR ENGINEERING GROUP APPROVED:

C O bM 2-1#'/. Technical Superintendent MM J>r' Plant Manager DISK: CYCLE 2 8402170243 840210 h POR ADOCK 05000364 '

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J TABLE OF CONTENTS PAGE i

1.0 Introduction 1 2.0 Unit 2 Cycle 3 Core Refueling 2 3.0 Control Rod Drop Time Measurement 7 4.0 Initial Criticality 9

5.0 All-Rods-Out-Isothermal Temperature Coefficient, Boron Endpoint and Flux 4 Distribution 10 6.0 Control and Shutdown Bank Worth Measurements 13 7.0 Power Ascension Procedure 15 8.0 Incore-Excore Detector' Calibration 18' 9.0 Reactor Coolant System Flow Measurement 20 l

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1.0 INTRODUCTION

The Joseph M. Farley Unit 2 Cycle 3 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 2 of the Joseph M. Farley Nuclear Plant is a Three Loop Westinghouse pressurized water reactor rated at 2652 MWth. The Cycle 3 core loading consists of 157 17 x 17 fuel assemblies.

The Unit began commercial operations on July 30, 1981, completed Cycle 1 on October 22, 1982 with an average core burnup of 15350.5 MWD /MTU, and completed Cycle 2 on September 17, 1983 with an average core burnup of 10371.2 MWD /MTU.

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2.0 UNIT 2 CYCLE 3 CORE REFUELING REFERENCES

1. Westinghouse Refueling Procedure FP-APR-R2

2. Westinghouse WCAP_10410 (The Nuclear Design and I

Core Management of the Joseph M. Farley Unit 2 Power Plant Cycle 3)

.: The refueling commenced on 9/28/83 and was completed in

-! 6 days on 10/4/83. The as-loaded Cycle 3 core is shown in Figures 2.1 through'2.4, which give the location of each fuel

assembly and insert, including the burnable. poison insert i locations and configurations. The cycle 3 core has a naminal j design lifetime of 14360 MWD /MTU and consists of 41 region 3 assemblies,'52 region 4 assemblies, and 64 region 5 assemblies.

- Fuel assembly inserts include 48 full length control rod clusters, 64 burnable. poison inserts, two secondary sources, and 43 thimble plug inserts.

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WESTINGilOUSE PROPRIETAl1Y CLASS 2

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, WESTINGHOUSE PROPRIETARY CLASS 2 FIGURE 2-2 Control Rod Locations RPNM LKJHG F EDC8 A

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0 E S 0 E- O E O E 8 Fresh BP 12 Fresh BP Configuration Configuration g E O E g E O E O E O E O E D E O E O E E O E Core. Center 16 Standard Core Center 1 Fresh BP Configuration gO E O E O O O O E O O O O O O O E O O O E. O O O O O O O O O O E O O O O 0

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

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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.

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SUMMARY

OF RESULTS For the Hot-full flow' condition (T > 541*F.and-i all' reactor coolant pumps operating) Te8KHical Speci-fication 3.1.3.4: requires that the rod drop time from i '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

-times were measured to be less than 2.2 seconds. The longest drop time' recorded was 1.48 seconds.for rod

B-10. The' rod drop time results for both dashpot

entry and dashpot bottom are. presented in Figure 3.1.

l Mean drop times are summarized below:

! TEST MEAN TIME TO MEAN TIME.TO-l CONDITIONS DASHPOT ENTRY- DASHPOT BOTTOM 4

Hot-full Flow 1.359 sec 1.876 sec

! To- confirm nonnal rod mechanism operation prior i' 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, Land the functioning of-the-Digital Rod. position indicator and the bank: overlap unit-l was checked.- Rod' stepping speed measurements were con-

-ducted. .All results were satisfactory, i

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m UNIT 2 CYCLE 3 NORTH 900 R

1.35 1.35 1.38 1.85 I

1.91 1.90 1.37 1.34-  ;

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K 1.82 1.87 1.88 1.90 1.37 '1.35 1.83 .1.83 \ (

j 1.45 1.37 1.35 1.32. 't.35 1.33' 1.38 .

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1.35 o 1.91 1.88 1.87 180 -N 1.87 1.34 1.32 1.35 1.37 g 1.85 1.82 1.84 1.91 1.35 1.34 1.35 1.33 1.35 1.37 1.37 l 1.83 1.88 1.84 1.85 1.83 1.90 1.85 1.35 1.38 1.84 1.90 E 1.37 1.33 1.33 1.32 1.85 1.87 ,1.83 1.83 0 1.37 1.35 1.88 C

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i 15' I4 13' 12 ll 10 9 8 7 6 5 4 3 2 ORIVE LINE " DROP TIME" TA9ULATION l

TEMPERATURE - 547 F PRESSURE - ??1s wh 5 FLOW - inn ~

X. XX BREAKER "0PENING" TO DASHP0T ENTRY - IN SECONDS DATE 10-22-83 X.XX BREAKER "0PENING" TO DASHPOT BOTTOM - IN SECOMOS FIGURE 3.1

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4.0 INITIAL CRITICALITY (FNP-2-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 reactivity computer.

SUMMARY

OF RESULTS Initial Reactor Criticality for Cycle 3 was achieved during dilution mixing at 2120 hours0.0245 days <br />0.589 hours <br />0.00351 weeks <br />8.0666e-4 months <br /> on October 22, 1983 The reactor was allowed to stabilize at the following critical conditions: RCS pressure- 2234 psig, RCS _g temperature 548.1*F, intermediate range power 9 x 10 amp, RCS boron concentration 1592 ppm, and Control Bank D position- 180 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 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT, BORON ENDPOINT AND FLUX DISTRIBUTION (FNP-2-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 (+0.46 pcm/ r).

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. The maximum incore tilt was 1.033. Westing-house was notified so that the hot zero power safety analyses and core modeling could be reviewed. This review was for information purposes only and the plant commenced power escalation immediately on completion of physics testing. The next two flux distribution maps, taken at 31% and 48% power, demonstrated that incore tilt had decreased to below 1.02.

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TABLE 5.1 ARO, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIEN_T Rod Configuration Boron Measured Calculated a Design Acceptance T

Concentration a T " mod Criterion ppm pcm/ F pcm/ F pcm/ F All Rods Out 1605 -2.039 +0.461 -2.3 1 3 a - Isothermal temperature coefficient, includes -2.5 pcm/ F doppler coefficient T

a - M derator only tcmperature coefficient mod ARO, HZP BORON ENDPOINT CONCENTRATION 1

Rod Configuration Measured CB (Ppm) Design predicted CB (ppm)

All Rods Out 1597 1629 1 50 11 l

TABLE 5.2 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 +10.8% 0.848 1 15% for Pi < 0.9 Maximum negative.

percent error -7.9% 1.055 i 10% for Pi > 0.9 B. Incore Quadrant Tilt:

Maximum Westinghouse Incore Tilt Review Criteria 1.0334 > 1.02*

• Westinghouse review performed with satisfactory results.

No review necessary if <l.02.

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6.0 CONTROL AND' SHUTDOWN BANK WORTH MEASUREMENTS (FNP-2-ETP-3601)

PURPOSE 4

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 i 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 i individually and for.the combined worth of all banks.

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TABLE 6.1 I

SUMMARY

OF CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS l

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l Predicted Baii Measured Worth & Review Bank Percent Bank Criteria (pcm) Worth (pcm) Difference L Control A 713 i 107 749.3 +5.1 l

. Control B (Ref.) 1425 1 143 1403.2* -1.5 l

Control-C. 713 1 107 631.2 -11.5 f'

Control D 1030 1 155 1011.9 .-l.8 l

l Shutdown A -1208 1 181 1231.8 +2.0-L L Shutdown B 799 i 120 719.8 -9.9 l

All Banks Combined 5888 1 589 5747.2 -2.5 i

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• Measured by dilution method 14

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i 7.0 POWER ASCENSION PROCEDURE-(FNP-2-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 4 3. Flux map at less than 50% power

! 4. Adhering to the delta flux band during ascension j to 75% power-4 5. Incore/Excore calibration at 75% power.

SUMMARY

OF RESULTS 1

j In compliance with Westinghouse recommendations and fuel warranty provisions, the power ramp-rate was i 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.

, Alignment of the incore movable detector system

!. normal, calibrate and emergency paths was accomplished j during power ascension (at power levels above 5%)- u

prior to performing the flux map;at 31% power.

4 Due to the lower neutron. leakage of the Cycle 13' core, design-predicted power range'NIS detector currents equal to about 75% of the corresponding' Cycle 2 values'were used for initial AFD channel scaling. As described below,.

these preliminary settings were checked-during subsequent

{ power ascension and final settings were developed from l the incore-excore recalibration performediat approximately j 75% power.

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Since both intermediate range NIS detectors were' I replaced during~the' outage,.it was necessary to startup-

using ~ estimated preliminary detector currents - for the j rod stop and reactor trip setpoints. 'Following entry into Mode 1, the 20% intermediate range rod stop was en-countered at 7;5% indicated power, su) new IR setpoint

-currents were extrapolated from measured data to permit i power ascension to continue. At'approximately 30% power, i

the first restart. calorimetric revealed that the reactor-trip.setpoint for-intermediate range channel N36.had been set .at 32.4% of rated thermal power, which was -in excess

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of the Tech. Spec. limit of 30%. This event was reported l in LER 83-053 and the N36 trip setpoint was immediately  !

reduced to the: current equivalent of 25% power. The N35 setpoint currents were also refined using the data measured at 30% power, and additional current data was taken during

. the ascension above 30% power to check the settings on both intermediate range channels.

Full flux maps were taken at 31%, 48%, 58%, 71%,

and 78% power. As summarized in Table 7.1, all results were within Technical Specification Limits.

An incore/excore calibration check was performed at 48% power with satisfactory results. However, quad-rant power tilt calculations peformed using FNP-2-STP-7.0 with the design-predicted NIS detector currents gave tilt ratios in excess of 1~.02 on one channel. Therefore, power escalation above 50% power was conducted using Technical Specification special test exception 3.10.2, which requires flux map verification of hot channel factors every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. At approximately 75% power, a complete incore-excore recalibration was performed to comply with Technical Specification requirements. The incore-excore recalibration is described in Section 8.0.

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TABLE 7.1

SUMMARY

OF POWER ASCENSION FLUX MAP DATA Parameters Map 51 Map 52 Map 53 Map 54 Map 55C Date 10/26/83 10/30/83 10/30/83 10/31/83 10/31/83 Time 11:46 00:04 15:33 00:54 12:51

Avg. % Power. '31%. 48% 58% 71% 78%

Max FAH 1.5289 1.5037 1.5322 1.4936 1.4744 Max. Power. Tilt

• 1.0161 1.0143 1.0162 1.0132 1.0134 Avg.' Core % A. O. -1.439 -3.791 -3.705 -4.968 +0.427 l

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

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8.0 INCORE-EXCORE DETECTOR CALIBRATION (FNF-2-STP-121)

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 systen, 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 that an AFD target band could be defined for ascension to 78% power. Flux maps for incore-excore recalibration were run at approximately 75% power at average percent core axial offsets of - 24.559, -12.475, -4.968 (using surveillance map 54 as supplemental data) and +0.593, as determined from the INCORE code 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.

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 Table 8.1) 18

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

I-Top = 1.2437*AO + 204.9130 pa I-Bottom = -1.1557*AO + 205.0941 pa CHANNEL 42:

I-Top = 1.0627*AO + 191.6829 pa I-Bottom = -1.1215*AO + 193.8551 pa CHANNEL N43:

I-Top = 1.2009*AO + 205.5291 pa I-Bottom = -1.2185*AO + 203.5039 pa CHANNEL N44:

I-Top = 1.1006*AO + 187.5665 pa I-Bottom = -1.1377*AO + 196.6878 pa 19-

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

PURPOSE j The purpose of this procedure was to measure the flow rate in each reactor cool &nt loop in order to confirm that the total core flow met the minimum flow requirement given in the Unit 2 Technical Specifications.

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 the average of six calorimetric heat balance measure-ments, the total core flow was determined to be 286,684.5 gpm, which meets the above criterion.

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. . . Alabama Power Company I

600 North 18th street Pcit Offics Box 2641 Birmingham. Alabama 35291 Telephone 205 250-1000 F. L CLAYTON, JR.

Senior Vice President Alaban1a Nw'er the southem electic system February 10, 1984 Docket No. 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 30555 Attention: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Unit 2 Cycle 3 Startup Report Gentlemen:

Enclosed is the Startup Report for Unit 2 Cycle 3 as required

, by the September 2,1983 letter from F. L. Clayton, Jr. to Mr. S. A. Varga.

If you have any questions, please advise.

Yours very truly, h N'" A

. L. Clayton, r. /

FLC,Jr/MDR:cl Enclosure cc: Mr. R. A. Thomas Mr. G. F. Trowbridge Mr. J, P. O'Reilly Mr. E. A. Reeves Mr. W. H. Bradford Dr. I. L. Myers 4