ML20154J427

Jm Farley Nuclear Plant Unit 1,Cycle 9 Startup Test Rept

ML20154J427
Person / Time
Site:	Farley
Issue date:	09/19/1988
From:	Hairston W ALABAMA POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NT-88-0430, NT-88-430, NUDOCS 8809230023
Download: ML20154J427 (21)

Text

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i AIABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PIRfr ,

UNIT NUMBER 1, CYCLE 9 1 L STARIUP TEST REPORT L 1

P PREPARED BY PLANT REACTOR E'NGINEERI!G GROUP i i

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- d d. k Technical Manager I

J N M General Manager-Nuclear Plant 1

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TABLE OF CONTDRS t 1

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l 1.0 Introduction 1 i

2.0 Unit 1 Cycle 9 Core Refueling 2 i 3.0 control Rod Drop' Time Measurement 7 4.0 Initial criticality 9  !

r 5.0 All-Rods-Out-Isothermal Temperature Coefficient and Boron Endpoint Measurement 10 4  !

l 6.0 control and Shutdown Bank Worth Measurements 12  !

l 7.0 Power Ascension Procedure 14 f t

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, 8.0 Incore-Excore Detector Calibration 16  ;

i 9.0 Reactor Coolant System Flow Measurement 18 [

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1.0 INTRODUCTION

l The Joseph M. Farley Unit 1 Cycle 9 Startup Test Report addresses the core refueling and the startup tests required by plant procedures following .

the refueling. The report provides a brief synopsis of each test and gives j a comparison of measured parameters with design predictions, Technical [

specifications, or values assumed in the FSAR. I Unit 1 of the Joseph M. Farley Nuclear Plant is a Westinghouse three ,

loop pressurised water reactor rated at 2652 MWth. The core loading consists of 197 17 x 17 fuel assemblies. The Unit began commercial i i operations on December 1, 1977.  !

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' r i Cycle 8 power operation began on December 2, 1986, and ended on March  :

25, 1988, with an average core burnup of 16,190 MWD /MTU.

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2.0 UNIT 1, CYCLE 9 CORC RE WELIff, REFERENCES

1. Westinghouse Refueling Procedure FP-AIA-R8

2. Westinghouse WCAP 11755 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant Cycle 9)

The fuel shuffle commenced on 4/3/88 and was completed in 24 days on 4/27/88. The as-loaded Cycle 9 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.

The burnable poison inserts used for Cycle 9 are wet annular burnable absorber rods (MABAs). The Cycle 9 core has a nominal design lifetime of 17500 mD/MW and consists of 5 region 6 assemblies, 22 region 9A assemblies, 4 region 9B assemblies, 58 region 10 assemblies, 40 region 11A assemblies, and 28 region 11B assemblies.

Fuel assembly inserts include 48 full length control rod clusters, 52 wet annular burnable poison inserts, two t.econdary sources, and 55 thimble plug inserts.

During core unload, a detailed fuel inspection program was conducted to eliminate leaking and defective fuel assemblies. Each fuel asserbly was visually inspected with binoculars and was ultrasonic-ally leak tested with the Bro'11 Bovari Failed Fuel Rod Detection Systera (FFRDS) . Assemblies having observed or suspected visual defects were re-examined with an underwater 'IV camera, and assemblies considered suspect on the initial leak test were subjected to re-examinationwiththeFTRESSystem. Five assemblies were found to have leaking fuel rods, of whic.) J18 and K44 had been scheduled for core reload. Nine assemblies had visual defects, of which J18, K44, J05, J17, K19, K21, and J25 were scheduled for core reload. These assemblies were rejected from core reload, which required a revised Cycle 9 core design to be obtained from Westinghouse.

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FIGURE 2.1: UNIT 1 CYCLE 9 REFERENCE LOADING PATERN R P N  % L K J H G F E D c B A 3:e 290 32 ,,,. ,

1 J21 J24 J14 64 133 4W160 827 4Wi&O til 73 _ 2 J29 un D43 E09 ulo Dil J59 23e 37 1 M 90 m37 1 M 40 821 i M 10 34 48 _3 J02 u18 2A2s of uS2 ca 2A29 uct J49 18 047 1 M 40 t19 M 250 tSOS 16ws40 e23 1 M 79 029 50 ,,

4 a4 rer Zul c0 Zuy tas 2AQs tal Zu2 c22 423 4e 179 12v544 438 1N32D 14e 710 13 towoe ac3 i M 3D 410 65

-. ---5 all 2AZ3 2A&2 735 2A24 03 c25 K12 D20 F44 2M7 2A21 1 1M00 sto

) ?' ..'..l- 6 R17 i M ac 1Mac 132 i M 7D toi 135 1 M 70 139 .

D14 Dif G7 DC1 E13 2A45 E45 2A36 t&O 2A27 , t01 1M3f 2A13 241 4MSc tt3 1 M 10 39r, I M SD the lac til 1 M 50 3?D 1 M 90 t35 4W11l 360 '

t J39 2A57 E47 ZA34 d3 2A53 G8 07 09 u34 C3 2A37 K11 u49 J20 la 804 iM@ 410 1CC B24 ** 540 650 til 510 47 16 600 t20 690 g L54 (C3 2A51 toe t&4 c1 CA C4 01 t49 (14 02 2A41 EM E18 640 4W130 tot 1 M 75 200 1M70 826 450 002 i M 50 24 i M 60 13 0 4Wt 3 4M g J44 2MO E60 use E24 2A&S (10 EOS (58 2A47 (42 2, sac [97 IMS JJ3 EL4  % 45 t&& iM4 t31 1 M 30 t43 1 M 10 845 i M 80 413 i M OD G14 *g U23 D22 O2 ZA31 C0 2A64 t02 b59 t19 2A09 sto 2A34 2A07 ,

34 8 1296Ce 134 1 M 90 420 110 180 ' 6.m e07 ' M 50 190 75 .,

J44 ZA16 Z?44 F43 100 154 K55 EOS D19 f50 D58 2A33 .45 3 en 1 M 60 tot 1 M 2D 1s04 i M 60 t25 iM 7: 822 150 12 Jia rtt ZA55 E20 2A35 E29 2A39 E15 Du L34 J13 250 54 1 M ao E06 i M 50 m28 1 M 40 31 4 13 J'4 2A04 ZA38 E17 2A44 E41 9 05 E11 J28 14 804 4WO90 e40 6w 00 Re2 44 34 J64 ZA04 D64 d5 2A44 2A12 450 45 60 12 '5 J01 407 J15 q-- FUEL ASSEMBLY INSERT SERI AL hTMBER

-cr-- FUEL ASSEMBLY SERIAL NUMBER v

The original w/o U-235 enrichments were:

Region 6 (F) assemblies ..... 2.9951 Region 9A (J) assemblies .... 3.597%

Region 9B (J) assemblies .... 3.906%

Region 10 (K) assemblies .... 3.597%

Region llA (2A) assemblies . . 3.805%

Region llB (2A) assemblies .. 4.207%

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- FIGURE 2.2 CONTROL ROD LOCADONS P N M L K J H G F E D C 8 A R

1 2 A D A 3 SA SA SP e

4 C B SP S C 8 S3 SP S8 A B D C D B A 8

7 SA SS SS SP SA 8 so' D SP C SP C SP D 9 SA Sa SS SS SA 10 A B D 0 D B A f

11 SS SP SS SP,

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12 C 5 SF B C 13 SP SA SA 14 A D A 15 Absorber Mglet:

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) PUNCT10N NUMBER OF CLUSTERS f Control Bank D 4 Control Bank C 8 Control Bank B 8 Control Bank A 8 Shutdown Bank SE 8 Shutdown Bank SA 8 SP (Spare Rod Locations) 13

• LOCATION NS & C11 = CORE WATER LEVEL THEIO10 COUPLE PROBES J

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FIGURE 2.3 I

l BURNASLE ABSORSER AND SOURCE ASSEMBLY LOCATIONS l

G F E D C 8 A N M L K J H M P r I

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16 16 16 3

16 SS 16 12 12 4

16 12 12 14 5

to 16 16 16 16 16 6

16 16 4 4 14 14 7

16 6 so- 16 16 16 4 9 4 16 to to 16 16 to 14 16 to 12 is to 12 11 I l

12 is SS to 12 12 14 16 16 13 4 4 14 15 o' l 704 WASAs in

1. 1. Number of WAGAs Secondary Source 52 Clusters SS (WABA = WET ANNULAR BUR 5ABLE ABSORBER) 1 5

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FIGURE 2.4 BURNABLE ABSORBER AND SECONDARY SOURCE ROD CONFIGURATIONS l l

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, 0 0 0 5 O O O O O O O E O E O E O O O O O E E O O O O O O E O E O I O O O E O I 4 BA Configuration 12 BA Configuration

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, E O E, E O E O E O O O O O O E E O C O O O E O E O E C O O O O E O E O O O 16 BA Configuration Secondary Source Rods 6

3.0 COtEROL ROD DROP TIME MEASUREMDE (RIP-1-STP-112) l 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 ensure compliance with Technical Specification requirements.

SUMPARY OF RESULTS For the hot full-flow condition (T 1 541' and all reactor coolant pumps operating) Technical Specificalion 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 times were measured to be less *.han 2.2 seconds. The longest drop time recorded was 1.85 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 MFAN TIME TO CONDITIOtiS DASHPOT ENIRY DASHPOT BCTTIOM Hot full-flow 1.63 sec 2.20 see To confirm normal rod mechanism operation prior to conducting the rod drops, a Control Rod Drive Test (RTP-0-ETP-3643) was performed. In

the test, the stepping waveforms of the stationary, lift and movable i 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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l 4 i NORTH l' SIT 1 CYCLE 9 900

-R 1.64 1.60 1.63 y 2.14' 2.12 2.14 P 1.64 1.62 a N

2.17 2.14 [

1.60 1.63 1.63 1.63 2.14 2.19 M 2.12 2.12 1.60

'2.17

,1.62 2.18 [L s

1.73 1.59 1.62 1.57 1.63 1.65 1.65 . g 2.15 2.23 2.19 N\ ' 27 1.61 2,15 2.14 1.56 2.12 1.55 1.01 2.13 2.05 2.07 2.13 J 8 1.60 1.60

) 0 1.62 1.60 2.12 2.10 2.08 180* -M 2.20 1.68 1.55 1.56 1.64 2.17 2,06 2.05 2.15 -G 1,72 1.67 1.62 1.57 1.58 1.64 1.60 2.22 2.22 2.13 2.08 2.10  :

2.17 2.13 F 1.60 1.61 2.18 -

2.18 E 1.64 1.61 1.56 1.62 g 2.21 2.13 2.08 2.17 0 1.64 1.66 2.18 2.17 .

1.67 1.70 1.85 2.23 2.21 2.3'A B 1

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% 1 15 14 13 12 ll 10 9 8 7 6 5 4 3 2 i ORIVE LINE "0 ROP TlHE" TABULATION 549 F PRESSURE - 2235 psig 100 TEMPERATURE . 1. FLOW -

l I.II BREAKER "0PENING" TO DASHPOT ENTRY - IM SECOMOS D ATE . 5-18-88 l X.II BREAXER "0PENING" TO DASHPOT BOTTOM - IN SECOMOS l

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

PURPOSE The purgwe 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

Or RESULTS  ;

Initial reactor criticality for Cycle 9 was achieved during dilution mixing at 1855 on May 18, 1988. The reactor was allowed to stabilize at the following critical conditions:

RCS pressure 2244 psig l RCS temperature 547'r j Intermediate range power 1.0 x 10', amp i RCS boron concentration 1977 ppm Control Bank D position 182 steps l rollowing 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 wtta taken during the flux increase preceding and immediately follow- '

ing initial criticality to demonstrate that adequate overlap existed.

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5.0 ALL-RODS-CUT ISO 1HERMAL TEMPERNIURE COEFFICIENT AND BORON ENDPOINT '

MEASUREMENT (FWP-1-ETP-3601) r FURPOSE  ;

The objectives of these measurements were to determine the hot, zero r power isothermal and moderator temperature coefficients for the  ;

all-rods-out (ARO) configuration and to measure the ARO boron end-  ;

point concentration.  ;

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SUMMARY

OF RESULTS i Ihe measured ARO, hot zero power temperature coefficients and the ARO  !

boron endpoint concentration are shown in Table 5.1. The isothermal

temperature coefficient was measured to be -0.32 pcm/'r which meets the design acceptance criteria. This gives a calculated moderator l temperature coefficient (corrected to ARO) of +1.954 pc W 'r which is within the Technical Specification limit of +5.0 pcm/'r. Thus, no rod withdrawal limits were needed to satisfy the +5.0 pcW'r limit.

1 The design acceptance criterion for the ARO critical boron 1 concentration was also satisfactorily met.  ;

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TABLE 5.1 '

ARD, HZP ISODIERMAL AND MODERA10R TEMPERATURE COEFFICIENT Rod Configutation Boron Measured Calculated ay Design Acceptance Concentration ay a, , , Criterion ppa pcuV*F pcuV'F pcW'F Bank D at 209 Steps 1991 -0.32 +1.954 -0.52 + 2 <

ay - Isothermal temperature coefficient, includes -2.19 pcW'F doppler coefficient O a,,, - Moderator only temperature coefficient, corrected to the all-rods-cut conditicn ARO, HZP BORCN ARO BORCM DOPOINF ENCENIRATICE Rod Configuration P Msured C, (ppm) Design - predicted C, (ppm)

Bank D at 228 Steps 19 % .4 1998 ! 50 t

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s 6.0 cottrROL NiD SHWLOhU BNiK WORM MEASURFKNrS (IUP-1-ETP-3601)

FURPOSE 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 We 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.

he control and shutdown bank worth measurement results are given in Table 6.1. % e measured worths satisfied the review criteria both for the banks measured individually and for the corrhined worth of all

) banks.

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IABLE 6.1 SLNEARY OF COPTIBOL AIO SIRTIDGN BAMC WORTH MEASURDENTS s

Predicted Bank Measured Worth & Review Bank Percent Bank Criteria (pca) Worth (pcm) Difference Control A 559 100 591.8 5.9 Control B (Ref.) 1227 123 1182.4* -3.6 Control C 787 118 739.3 -6.1 Control D 1049 ! 157 1001.7 -4.5 l C l Shutdown A 1039 156 1028.2 -1.0 '

l Shutdown B 919 ! 138 849.1 -7.6 l All Banks en=hined 5580 558 5392.5 -3.36 l

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• Measured by dilution method .

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7.0 POWER ASCENSICN PROCEDURE (ntP-1-ETP-3605)

PURPOSE 2 e purpose of this procedure was to provide controlling, instructions for:

1. NTS intermediate and pwer range setpoint changes, as required prior to startup and during power ascension.

2. Ramp rate limitation during power ascension.

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

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

b. incore movable detector systen alignment (n@-1-ETP-3606),

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

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

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

SLROtARY 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 tanga setpoint was increased to 80% to enable power escalation (above 25%) for calorimetric recalibration of the power range channels. (The 80% setpoint was used instead of 109% in case the uncalibrated power range channels were indicating nonconservatively.)

At approximately 30% power, the power range channels were recali-brated, the high-range trip setpoint was restored to 109%, and set-point currents were determined for the intermediate range channels.

The 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. his ramp rate was observed during the ascension to 100%

power.

We startup and power ascension test program controlled by ETP-3605 comenced with the zero-power physics tests described in Sections 4.0

- 6.0 of this report, rollowing physics testing, incoce movable detector system core limit settings were determined for all modes of operation during the ascension to 30% power. h e Incore-Excore recalibration test (described in Section 8.0) was performed at 30%,

48% and 80% power, and the reactor coolant system flow measurement (Section 9.0) was perforned at 100% power. Surveillance of reactor core hot channel factors was accomplished using data from the full-e core flux m ps taken during the Incore-Excore procedure. As suntna-rized in Table 7.1, all results were within Technical Specification limits.

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) M LE 7.1 SLAWIARY OF POWER ASCENSIG4 FIBK MAP DATA Parameters Map 209 Map 215 Map 216 Date 05/23/88 05/25/88 05/27/88 l Time 17:40 02:33 03:12 t

Avg. % Power 31.86 48.02 80.16 i

Max FaH 1.5680 1.5563 1.5020 1

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! Max. Power Tilt

• 1.0262 1.0262 1.0176  :

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G Avg. Core 1 A. O. +6.825 +9.244 +4.652 naxim m FQ(Z) 2.1543 2.1499 1.9581

! FQ Limit 4.6052 4.5240 2.8783

, Xenon Conditions Non-Equilibritan Equilibrium Equilibrium

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• Calculated (incore) power tilts based on assembly FaHN from all assemblies.

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O 8.0 INCORE-EXCORE DETEC'!OR CALIBPATION (f'NP-1-STP-121)  ;

PURPOSE i i

I %e objective of this procedure was to determine the relationship  ;

between power range upper and lower excore detector currents and j incore axial offset for the purpose of calibrating the delta flux j penalty to the overtemperature AT reactor trip, and for calibrating  ;

the control board and plant computer axial flux difference (ArD) i channels. .

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! A full Incore-Excore recalibration, consisting of one full-core i j base-case flux map and five quarter-core flux maps, was performed at  !

approximately 30% power. %e six maps were obtained at axial offsets t of +6.8%, +23%, +12.4%, -5.4%, -12.0%, and -26.0% and, during each j

] map, detector currents and calorimetric data were taken. he .

detector currents were normalized to 100% power and a least-squares  !

fit was performed to derive an output current vs. axial offset  ;

equation for each top and bottom detector. Calibration currents i

! derived from these equations were used to recalibrate the nuclear  !

) instrumentation system (NIS) delta flux channels and the [

l overtemperature AT delta flux input.

I i 1 Power was then increased to 48% and stabilized in preparation for a j full core flux map for hot channel factor surveillance. However, i excore quadrant power tilt ratic (Q m ) calculations performed during the power ascension disclosed that a high indicated o m had ,

developed. Since the flux map indicated that core hat channel factors were satisfactory and that the high excore o m was not 3,

associated with the actual incore tilt (which was in a different i quadrant), a detector equation I-zero current renormalization was  !

) performed to correct the problem and the NIS delta flux "

j instrumentation was recalibrated.

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) Nhen power escalation was resumed, the Q m again begin to increase, '

reaching 1.02 at just above 99% power. %erefore, following a

! planned power reduction and stabilization at 80% for maintenance purposes, a full-core flux map was performed. Again, the flux map

indicated that core bot channel factors were satisfactory and that -

the indicated high om was not a reflection of actual incore tilt.

herefore, the detector equation I-zero current renormalization was ,

j. repeated and the NIS delta flux instrumentation was recah brated. No  !

4 further calibration problems were experienced during a subsequent  !

further power reduction for turbine DEN checkout, nor during the  !

, return to full power. l l

) We finalized incore-excore equations used to derive delta flux l channel calibration data are given in Table 2.1.

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DEITC70R CURADfT VERSUS AXIAL OFFSET EQUATIONS 7 CSTAINED FROM INCORE-EXCORE CALIBRATION TEST j l

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4 CHANNEL N411

. I-Top = 0.6902*Ao + 178.92 va '

I-Bottom = -1.2255*Ao + 175.36 va

) CHANtEL 42:

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I-Top = 0.7932*Ao + 173.46 pa -

! I-Bottom = -1.2903*AC + 174.42 a  :

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J CHANtEL N43:  ;

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! I-Top = 0.7254*Ao + 180.01 va I-Bottom = -1.3113*Ao + 189.48 va  !

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f I-Top = 0.7355*Ao 4 167.49 va f I-Bottcm = -1.3021*Ao + 170.73 va J t  !

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9.0 REACTOR COOIR E SYSTEM FLOW PIASUREMC C (RIP-1-STP-ll5.1)

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 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 gprn for three loop operation.

From the average of ten caloriw tric heat balance measurements, the i total core flow was determined to be 282,826 gre, which meets the above criterion.

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NT-88-0430 Alabama Power Company 600 North 18th Street Post Oft:ce Box 2641 B,rmingham, Alabama 35291-0400 Te!ephone 205 2%1837 W. G. Hairston. lit senor Vice Prevdent Nuclear Operatons Alabama Power 1% sourwn emrc re em Septemb9r 19,1988 Docket No. 50-348 U. S. Nuclear Regulatory Commission Attention iccument Control Desk Washington, D. C. 20555 Joseph M. Earley Nuclear Plant - Unit 1 Cycle 9 - Startup Report Gentlemen:

Enclosed is the Startup Report for Unit 1 Cycle 9 as outlined in the April 19, 1988 letter from fir. R. P. Mcdonald.

If you have any questions, please advise.

Yours very truly, W.J. /4.22 -

W. G. Hairston, III hG!A TR:emb Enclosure cc: Mr. L. B. Irng Dr. J. N. Grace Mr. E. A. Reeves

!!r. G. F. Itaxwell f

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