ML20207R711

Jm Farley Nuclear Plant Unit 1,Cycle 8 Startup Test Rept

ML20207R711
Person / Time
Site:	Farley
Issue date:	03/13/1987
From:	ALABAMA POWER CO.
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ML20207R709	List: ML20207R708 ML20207R711
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NUDOCS 8703180052
Download: ML20207R711 (22)

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ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT UNIT NUMBER 1, CYCLE 8 STAR'IUP TEST REPORT PREPARED BY PLANT REAC'IOR ENGINEERING GROUP

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h Technical Manager

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<. 5-General Manager-Nuclear Plant DISK: CYCLES /8 l

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8703180052 870313 PDR ADOCK 05000348 P PDR ,

TABLE OF CONIDTTS PAGE 1.0: Introduction 1 2.0 Unit 1 Cycle 8 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 Distribution 10 6.0 Control and Shutdown Bank Worth Measurements 12 7.0 Power Ascension Procedure 14 8.0 Incore-Excore Detector Calibration 17 9.0 Reactor Coolant System Flow Measurement 20 l

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1.0 INTRODUCTION

h e Joseph M. Farley Unit 1 Cycle 8 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 Westinghouse three loop pressurized water reactor rated at 2652 MWth. he Cycle 8 core loading consists of 157 17 x 17 fuel , assemblies.

h e Unit began commercial operations on December 1, 1977, and completed cycle 7 on October 3, 1986, with average core burnup of 17,231.33 MND/MTU.

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2.0 UNIT 1, CYCLE 8 CORE REPUELING REFERENCES

1. Westinghouse Refueling Procedure FP-ALA-R7

2. Westinghouse WCAP 11291 (he Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant Cycle 8)

The refueling commenced on 10/12/86 and was completed in 16 days on 10/28/86. The as-loaded Cycle 8 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 locations and configurations. We burnable poison inserts used for Cycle 8 are wet annular burnable absorber rods (WABAs). We Cycle 8 core has a nominal design lifetime of 16000 MND/MIU and consists of 4 region 7 assemblies, 18 region 8A assemblies, 50 region 9A assemblies, 24 region 9B assemblies, and 61 region 10 assemblies. Fuel assembly inserts include 48 full length control rod clusters, 49 wet annular burnable poison inserts, two secondary sources, and 58 thimble plug inserts.

During unload of the Cycle-7 core, each fuel asembly was visually inspected for damage using binoculars (procedure FNP-0-ETP-3636).

Assemblies J67, Hil, J25, H10, J38, and H32 were found to have grid damage, and assemblies ZD4 and CO3 had noticable rod bow due to fiVe cycles of operation. Of these, Hil, H32, ZD4 and CO3 were discharged from the core as scheduled. Since the grid damage to asemblies J25 and H10 was relatively minor, these assemblies were reloaded for Cycle-8 as recommended by Westinghouse. Assemblies J67 and J38, which had more severe grid damage, were rejected for reload and were replaced by assemblies HOl and H34 in the Westinghouse core redesign.

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.. . . FIGURE 2.1 ALA Cycle 8_ Loading Pattern i R P N M L K J H G F E D C B A 7217 707 TI7 I Fil El3.- Kil H09 33+ KOS T21 k25 723 H40 ,

E14 Mr F' F2 F DS Ll+ -

2 H I3 K+3 KS7 7 74- K45 757 kSt K03 HOG -

A9 F F Gl2. F TI2 F F R4 r HO7 739 K27 7 71 k48 33I. k40 7 72 Kol TIG H35 g Gir cil F' k3 F 'HS F F3 F Alli TIS H37 K3r kro 623 kl2 73S Trf 732 MH G3r KGl; k22 H21 Bil F F Tl F 87 Hi F7 F GI F F Pil 702 K47 756 N2+ 727 K2s 74o krc 3'o 9 k42 TG4 Kil 749 g L4 F NG F D9 F M4 F M9 ^ F CG F E+

TSo k3G 77G KIS TIz Kl+ 7Gs 703 7t9 k39 74 5 K52 TG8 KLG 74G E.l o F 71+ F 76 F 87 k2 Ll3 F' GG F Gl+ F L.lo 7 Hol 74+ KSI TlI 76+ 330 737 K4G 73G TIO 759 747 K49 70 5 HS4 8

MS Plo F L. 8 R8 Ml2 PG F SG D4 A8 Et F Slo bl3 JAS kcl 7GO KOG TSI KIQ 353 70G' TGI K44 TO4 KS1 7G2 Kl9 tit a E.G F 72 f* Tio F 89 Kl+ P9 F' G IO F G2 F' LG

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748 k&O 770 kC3 741 KI8 784 KS8 752 K23 752 KO7 T28 ,O Ll2. F Nio F DT F DI2 F M7 F CIO F EIL '

hit K3+ k2o GS3 KoS 722 7GS' 74) kr+440 Kir ko9 His .,U 85* F F TIS F' M9 H15 F*l F 4If F F' PS HSG ToI K32 T73 K30 72G Ic19 7G9 klG 753 Hl+ ,g GI C S~ F' Kls F HIl F F13 F NC 71 Ho2 kof kl7 775 Ko2 75 4 K41 K21 H33 ,

A7 F F G4 F 7+ F F R7

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H10 708 KSC 7IS Xc7 '713 H38 ,,

E2 Mll F F'l4 P' bil L2 720 J2+ 729 ._

Fr 13 kS' In each core location the assembly identification is given at the top and the previous cycle core location is given below (except for assemblies G23, G33, G35 and G46, for which the Cycle 6 locations are given).

Assembly letter designator G H J J K Fuel enrichment region 7 8A 9A 9B 10 From core cycle 6 7 7 7 Feed w/o U-235 (original): 3.002 2.999 3.597 3.906 3.597 3

FIGURE 2.2 CONTROL ROD LOCATIONS R P N M L K J H G F E D C B A 1

2 A D A 3 SA SA SP 4 C B SP B C 3

5 SP l SB SP SB J f 6 A B D C D 8 A 7 SA SB SB SP SA 8 D SP C SP C SP D 9 SA SP SB SB SA 10 A B D C D B A 11 SB SP SB SP 12 C B SP B C 13 SP SA SA 14 A D A 15 Absorber Mgtgrial:

FUNCTION NUMBER OF CLUSTERS Control Bank D 8 Control Bank C 8 Control Bank 8 8 Control Bank A 8 Shutdown Bank SB 8 Shutdown Bank SA 8 SP (Spare Rod Locations) 13 LOCATIONS N5 & Cl1 = CORE WATER LEVEL THERMOCOUPLE PROBES 4

". FIGURE 2.3 BURNABLE ABSORBER AND SOURCE ASSEMBLY LOCATIONS R P N M L K J H G F E D C B A 1

2 3 12 16 12 4 12 16 SS 16 12 5 12 20 8 20 12 6 12 20 20 20 20 12 7 16 20 20 16 8 16 8 20 8 16 9 16 20 20 16 10 12 20 20 20 20 12 11 12 20 8 20 12 12 12 16 SS 16 12 13 12 16 12 14 15 1

1. 1. Number of WABAs 756 WABAs in SS Secondary Source 49 Clusters (WABA = WET ANNULAR BURNABLE ABSORBER) l l

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FIGURE'2.4 BURNABLE ABSOR8ER CONFIGURATIONS

,0 0 D E O 'O O O E O O E O E D E O E E O O E E O O O E O O E O E O E O O O E O E l

l 8 BA Configuration 12 BA Configuration l

E E,

,5 O B, g 5 5 O E O E E O E O E O E E O E B B B l 5 O E O E E O E O E E D E E E E 16 BA Configuration 20 BA Configuration l

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

PURPOSE h e purpose of this test was to measure the drop time of all full length control rods wxler hot-full conditions in the reactor coolant system to ensure compliance with Technical Specification requirements.

SUMMARY

OF RESULTS For the Hot-full flow condition (T 2 541' and all reactor coolant pumps operating) Technical Spe M ication 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 coltage decay until dashpot entry. All full length rod drop times were measured to be less than 2.2 seconds. %e longest drop time recorded was 1.79 seconds for rod B-6. We rod drop time results for both dashpot entry and dashpot bottom are presented in Figure 3.1. Mean drop times are sumarized below:

TEST !EAN TIME 'IO MEAN TIME 'IO CONDITIONS DASHPOT ENTRY DASHPCfr BOTIOM Hot-full Flow 1.61 sec 2.15 see To confirm nonal rod mechanism operation prior to conducting the rod drops, a control Rod Drive Test (FNP-0-ETP-3643) was performed. In the test, the stepping waveforms of the stationary, lift and moveable gripper coils were examined, and the functioning of the Digital Rod position indicator and the bank overlap unit were checked. Rod stepping speed measurements were also conducted. All results were satisfactory.

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t 4 0 UNIT 1 CYCLE 8 5*

R 1,61 2.15 1.61 2 16 1.60 2.12 kV -

P 1.60 1.60 a N

2.11 ~2.17 [

1.58 1.58 1.61 .1.63 2.18 N 2.16 2.19 2.08 1.59 2.20 1.59 2.15 [lN 1.70 1.56 1.64 1.55 1.61 ~1.62 1.62

\ 2.24 2.10 2.14 2.10 2.12 2.13 '2.18 i K

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N 1.60 1.54 1.54 1.60 2.20 2.08 2.03

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2.16 ~J kS

• go 1.62 1.59 1.58 1.61 o 2.15 180 -H 2.18 2.15 2.12 1.64 1,55 1.56 1.62 2.16 2.05 2.07 2.12 -0 1.68 1.61 1.62 1.56 1.61 1.60 1.60 I 2.12 2.14 F 2.25 2.20 2.13 2.10 2.19 1.58 1.62 2.17 2.18 E 1.61 1.59 1,56 1.62 l 0

2.11 2.11 2.11 2.18 1.61 1.64 l 2.16 2.15 C 1.67 1.72 1.79 g l 2.25 2.22 2.32 I I ,

, 270*

N 1 15 14 13 12 11 10 9 8 7 6 5 4 3 2 i ORIVE LINE "0 ROP TIME" TABULATION 544 F pggggggg , 1945psig pg , 100 TEMPERATURE -

X.XX BREAKER "0PENING" TO DASHPOT ENTRY - IM SECOMOS 11-29-86 DATE -

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

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

PURPOSE he 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.

SIBIMARY OF RESULTS Initial reactor criticality for Cycle 8 was achieved during dilution mixing at 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br /> on November 30, 1986. h e reactor was allowed to stabilize at the following critical conditions: RCS pressure, 2261 ppig; RCS temperature, 547.3*F; intermediate range power, 1.6 x 10 asp; RCS boron concentration,1807 ppm, and Control Bank D position, 188 steps. Following stabilization, the point of adding nuclear heat was determined and a checkout of the reactivity computer using both positive and aegative flux periods

, was successfully accomplished. In addition, source and intermediate range neutron channel overlap data were taken during the flux increase preceding and isenediately following initial criticality to demonstrate that adequate overlap existed.

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5.0 ALL-RODS-OUT ISOIHERMAL TEMPERATURE COEFFICIENT, BORON I!NDPOINI AND FLUX DISTRIBUTION (FNP-1-ETP-3601)-

4 PURPOSE i 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.

SIMMARY OF RESULTS he 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 -1.62 pen /*F which meets the design acceptance criteria. his gives a calculated moderator temperature coefficient of +0.75 penV'F which is within the Technical Specification limit of +5.0 pen /'F. hus, no rod withdrawal limits are needed to ensure the +5.0 penV'F limit is met.

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

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' TABLE 5.1 ARO, HZP ISO 1HERMAL AND MODERATOR TDIPERATURE COEE TICIENT Rod Configuration Boron Measured Calculated g Design Acceptance.

Concentration g a, , , Ghedm ppm pcm/*F penV'F penV'F -

l All Rods Out 1840 -1.62 +0.75 -2.11 + 2 i

g - Isothermal temperature coefficient, includes -2.32 penV'F doppler coefficient a,,, - Moderator only temperature coefficient ARO, HZP BORON ENDPOINT CONCENTRATION Rod Configuration Measured C, (ppm) Design - predicted C, (ppm)

All Rods Out 1846.0 1825, 50 ,

6.0 ' CONTROL AND SHtHDOWN BANK WORTH MEASURDENTS (FNP-1-ETP-3601)

PURPOSE 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 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.

We control and shutdown bank worth measurement results are given in Table 6.1. We 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 SIMIDOWN BANK NORTH MEASUREMENTS Predicted Bank Measured Worth & Review Bank Percent Bank Criteria (pcm)** Worth (pcm) Difference Control A 358 100 358.0 0.0 Control B (Ref.) 1245 124 1234.0* -0.9 Control C 855 128 849.2 -0.7 Control D 1020 153- 1015.9 -0.4 Shutdown A 1043 156 1018.7 -2.3 Shutdown B 1008 151 974.9 -3.3 0 '

All Banks Combined 5529 553 5450.7 -1.4

• Measured by dilution method

• • Review Criteria: Predicted worth + 10% - Reference bank and combined bank worth Predicted worth + 15% of 100 pcm which ever is greater - all other banks

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

PURPOSE he 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 reconmendations.

3. Conduct of startup and power ascension testing, to include:

a. HZP reactor physics tests (ENP-1-ETP-3601).

b. incore movable detector system alignment (FNP-1-ETP-3636).

c. incore/excore AFD channel recalibration (FNP-1-STP-121).

d. core hot channel factor surveillance (FNP-1-STP-110).

e. reactor coolant system flow measurement (FNP-1-STP-115.1).

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. ( he 80% setpoint was used instead of 109% in case the uncalibrated power range channels were indicating nonconservatively.) 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.

he 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.

Determination of the incore movable detector system core limit settings (FNP-1-ETP-3606) was accomplished for all modes of system operation during the ascension to 35% power.

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In previous Unit 1 stcrtups, the incors-excora r:cclibrction was performed at approximately 75% power. During the Cycle 8 etartup,

• however, a preliminary recalibration was performed at 35% power and

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• additional data were taken between 50% and 63% power. he incore-excore results were renormalized at 75% power to correct out-of-tolerance quadrant power tilt ratios that developed during

_ power ascension.

he revised incore-excore recalibration program resulted in twelve quarter-core, and three full-core flux maps being taken between 35% and 75% power. he results from the full-core maps were within Technical Specification Limits, and are sumarized in Table 7.1.

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TABLE 7.1

SUMMARY

OF pol 4ER ASCENSION FLUX MAP DNIA Parameters Map 178 Map 184 Map 192 Date 12/03/86 12/03/86 12/06/86 Time 03:27 19:53 20:18 Avg. % Power 34.50 33.50 75.68 Max FAH 1.5627 1.5434 1.4899 Max. Power Tilt

• 1.0133 1.0071 1.0085 Avg. Core % A. O. -0.813 +1.379 -2.672 Maximum FQ(Z) 2.0902 2.0857 1.9485 5

FQ Limit 4.5969 4.5969 3.0845 Xenon Conditions Non- Equilibrium Equilibrium Equilibrium

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

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8.0 'INCORE-IDtCORE DETECIOR CALIBRATION (FNP-1-STP-121)  ;

?4 1 1 2- .

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 (AFDJmchannels. -

SUMMARY

OF'RESULTS r

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During previous Unit 1 startups, incore-excore recalibration was perforned at approximately 75% power. However, during the Cycle 8 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. %e 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 34% power under more stable xenon conditions than the original base case map for verification of core hot channel factors.

d (c) During power escalation, seven quarter core flux maps were taken between 49% and 63% power to develop additional data for

comparison with the 35% power results.

(d). Due to the development of out-of-tolerance quadrant power tilt ratios during power ascension, the plant was stabilized at 75%

l power to achieve xenon equilibrium, and a full-core flux map was taken to normalize the calibration data at this power level.

This was.done to correct for changes in ambient incore tilts and to. compensate for temperature decalibration (the change in e reactor core neutron leakage caused by the changes in coolant temperature associated with changes in power). The power range NIS channels were recalibrated to incorporate this correction.

At 35% power, the six flux maps were performed at axial offsets of approximately -0.8%, +26%, +16%, -12%, -20%, and -7%. We 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 channels, l

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During power C:cension between 49% and 63% power, seven quartsr core flux maps were performed at axial offsets of approximately +6%,

+18%, +9%, 0%, -5%, -12%, and -18%. Detector current vs axial offset equations were derived and the resulting slopes demonstrated good agreement with the slopes in the 35% power equations. Because of this close agreement, no attempt was made to further refine the slopes by merging the 35% data with the data taken between 49% and 63% power.

No recalibration of the NIS channels was performed at this time.

At approximately 75% power, revised zero percent axial offset (I-zero) currents were determined from flux map data and were combined with the 49% - 63% power 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 DETECIOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:

I-Top = 0.8287*AO + 167.38 a I-Bottom = -0.9592*A0 + 163.0 a s

CHANNEL 42:

I-Top = 0.8682*AO + 168.81 pa I-Bottom - -0.9946*AO + 163.04 pa CHANNEL N43:

I-Top = 0.7699*AO + 163.92 pa I-Bottom - -1.0553*AO + 174.11 a CHANNEL N44:

I-Top = 0.8020*AO + 162.01 pa I-Bottom = -1.0309*AO + 162.28 pa i, 19 l.

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9.0 REAC'IOR COOIANT SYSTEM FIM MEASURDENT (ENP-1-STP-115.1)

'4 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 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 measurements, the total core flow was determined to be 283,963.3 gpm, which meets the above criterion.

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