Text

Snoc Jfnp Startup Test Rept Unit 1 Cycle 16. with
	ML20205S964
Person / Time
Site:	Farley
Issue date:	04/20/1999
From:	Dennis Morey; SOUTHERN NUCLEAR OPERATING CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NEL-99-0170, NEL-99-170, NUDOCS 9904270100
	Download: ML20205S964 (15)
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D:ve Morey S uthern N:cle:s Mce President ~ Operating Compaq,Inc.

Farley Project Post Ofhce Box 1295 Birmingham, Alabama 35201 Tel 205992 5131 SOUTHERN April 20, 1999 COMPANY Energy to Serve nurWorld*

Docket No.: 50-348 NEL-99-0170 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit 1 Cycle 16 Startun Test Report Ladies and Gentlemen:

In accordance with the reporting requirements of Technical Specification 6.9.1.1, Southern Nuclear Operating Company is submitting a Startup Test Report for Farley Nuclear Plant Unit 1 Cycle 16.

Should you have any questions, please advise. There are no NRC commdments in this submittal.

Paanac*kily submitted, Dave Morey j JAC/MGE/maf: pwrup99. doc /

Enclosure:

Startup Test Report Unit 1 Cycle 16 9904270100 990420 PDR

'I P ADOCK 05000348 PDR t

f,5 ,

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l Page 2 i U. S. Nuclear Regulatory Commission l cc: Southern Nacient Operating Company Mr. L M. Stinson, General Manager - Farley i U. S. Nuclear Regulatory Commissiost Washington. D. C.

Mr. J. I. T _ - . laconsing Project Manager- Farley U. S. Nm '--- Rei-% C---:--6 Region H Mr. L A. Reyes, F=pn' Adnumstraser Mr. T. P. Johnson, Senior Resident laspector- Farley

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SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT Startup Test Report O' nit 1 Cycle 16 l

SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT Startup Test Report Unit 1 Cycle 16 TABLE OF CONTENTS Section Subiect Eags 1.0 Introduction .. .. .... . ......................................................I 2.0 Unit 1 Cycle 15 - 16 Core Refueling ................................... ........... 2 3.0 Control Rod Drop Time Measurement ......... ............... ....... ... . ...... 3 4.0 Initi al Criti cality. . . . . . . . . . . . . . . . . . . . . . ... . .... . . . . .. . . . .. . . . . . . . . .. . . . . .. . . .. . . . . . . 5 5.0 All-Rods-Out Isothermal Temperature Coefficient and B oron Endpoint . . . .. . . . . . . . . . . . . . . . . . . . . . .. . .. .. . . . . . . .. .. . . . . . . .. . .. . . . . . . . . . . . . . . 6 6.0 Control and Shutdown Bank Worth Measurements...... ..... ............. 7 7.0 Power Ascension Activities........ ........... ........... . ...... ... .. ............. . 8 i

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1.0 INTRODUCTION The Joseph M. Farley Unit I Cycle 16 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, Core Operating Limits Report, or values in the FSAR safety analysis.

The Unit 1 Cycle 16 core has been uprated to increase the NSSS power to 2785 l MWth (core full power of 2775 MWth plus 10 MWth reactor coolant pump heat). The uprate design change was accomplished under 10 CFR 50.59 and associated Technical Specifications Amendment Number 137. The Cycle 16 fresh fuel has also been designed to provide 1) improved fuel skeleton i stability under irradiation; 2) improved corrosion performance; and 3) l additional measures to control fuel rod internal pressures at high burnups. The additional VANTAGE + fuel assembly design features adopted for Cycle 16 to obtain improved fuel performance include ZIRLO Mid and IFM grids, ZIRLO i guide thimble and instrument tubes, annular fuel pellets in the top six inches of IFBA rods, and 1.25X IFBA at 100 psig Helium backfill pressure. Also to reduce corrosion and the propensity for axial offset anomaly (AOA), insertion of thimble plugs was re-introduced for Cycle 16. The reload design for this cycle utilizes 68 fresh feed ZIRLO clad VANTAGE + assemblies with the above design features, 61 once burned ZIRLO clad VANTAGE + fuel assemblies 28 twice burned VANTAGE 5 Zircaloy clad fuel assemblies. The secondary sources are located at D-08 & M-08 within once burned assemblies, as was the case with the previous cycle. The loading pattern places RCCAs into fuel assemblies which will not exceed 40,000 MWD /MTU burnup at EOL.

'Ihe design depletion of reactivity of the Cycle 16 core is 18,100 MWD /MTU j with an allowed power coast down of up to 19,300 MWD /MTU. ;

2.0 UNIT 1 CYCLE 15 - 16 CORE REFUELING Unloading of the Cycle 15 core into the spent fuel pool commenced on 10/24/98. During the omond, each ' fuel assembly was inspected with binoculars for indications of damage or other problems. No indications of physical damage were found. White or grayish deposits were observed but to a much lesser degree than seen in the previous cycle core omond Fuel oxide measurements were not performed for the cycle 15 core omond assemblies.

Since the fuel inspecdons revealed no fuel damage or defects, no revisions to the )

original Cycle 16 core loading pattern were required. Cycle 16 Core reload x=- = +i on 11/30/98 and was completed on 12/02/98. ;

REFERENCES

1. Procedure FP-APR-RIS, J. M. Farley Unit 1 Cycle 15-16 Refueling.

2. Westinghouse Wr'AP 15110, The Nuclear Design and Core Management of the. Joseph M. tarley Unit 1 Power Plant, Cycle 16.

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4 3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-1-STP-112)

FURPOSE j The purpose of this procedure was to measure the drop time of all control rods under hot full-flow conditions in the reactor coolant system in order to ensure l compliance with Technical Specifw.ations requirements.

SUMMARY

OF RESULTS For the hot full-flow condition (Tavg 2 541*F and all reactor coolant pumps operating), Technical Specification 3.1.3.4 requires verifw.ation that each ,

control rod will insert in s 2.7 seconds when tripped from the fully-withdrawn position. For this test, an automatic data acquisition system provided by the Analysis and Measurement Services Corporation (AMS) was used to obtain drop time data from an entire rod bank (8 rods) at a time. Individual control rod drop times are shown in Figure 3.1. All control rod drop times were measured to be less than 2.7 seconds. The longest drop time recorded for time i to dashpot entry was 1.86 seconds for rod B6. Mean drop times are I l

summarized below.

RCS Conditions Mean Time to Mean Time to Dashpot Entry Dashpot Bottom Hot full-flow 1.62 sec. 2.29 sec.

l To confirm normal CRDM operation prior to conducting the rod drop test, the !

Verification of Rod Control System Operability (IWP 0-ETP-3643) was also performed using the AMS System to acquire stepping data for an entire rod i bank at a time. In this test, the stepping waveforms of the stationary, !! fit and movable gripper coils were examined for anomalies; rod speed was measured; ami the functioning of the Digital Rod Position Indicator (DRPI) and bank i overlap unit were checked. In addition, the bank overlap unit settings for the fully withdrawn rod position to 226 steps were verified to be correct. Timing measurements were also performed on the stepping waveforms for CRDM Logic Cabinet performance testing. All results were satisfactory.

Figure 3.1 FNP Unit 1 Cycle 16 ControlRod Drop Times R P N M L K J H O F E D C B A 1

14,~5 eja L,68

.y@ 14,8a; LNj$ LISP Laph 2 1483r 14de LW L36Y'i 3 145e idow Les iden j 14y L38)@ L34$ L3Bih - 4

@f 14 9 L#b 5 4 873 142a;". l6t( L61s u 12"4 c IJB g.:" L969

.x kJ. 6 f, -

L38V L3th L34 A L3Ts; L35M L34& LS6s 6 B.stp; L57::g; L57f? L67g Ll?P L13j L18 fly L397 7 1.58 1 LS5;:, LJ72 LWV L38k L49$. L33% (40N 8 t 884' 145y LJS,y 1404 i

LSTj f L355 L33 ) ' BM 9 i l e t,;' I.A.L., . L6e. , ~ L66; ,

12. ,4 ..

- - t # L..

t L61n.+ - . -~

RJ7h LISP L3fg L33% L404 L33M BJ53 10 Lap:i- '1 imu L11 e L35Ii 11 140. . V.

1.56 '

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- BA W. . <

L36lifi L308- L331 BR3, 12 LMg idem, 2

L35)g L35j - 13 4 North L78f? K44s 146 L4U m L38d' L35E < 14 X.XX 4-Breakar"opeams"to Dashpot entry X.XX e-Breaker"openms"to Dashput hanam 15 RCS Tageka 544 'F RCS Pressure. 2250 nsin

% Flow: 100 %

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4.0 INITIAL CRITICALITY (FNP-1-STP-101)

FURPOSE h purpose of this evolution is to achieve initial criticality under carefully controlled conditions, establish the upper flux limit for conducting zero power physics tests, and operationally verify the calibration of the reactivity computer.

SUMMARY

OF RESULTS Initial reactor criticality for Cycle 16 was achieved during dilution mixing on 12/27/98. W reactor was allowed to stabilize at the following conditions.

RCS Pressure 2235 psig RCS Temperature 548.4 'F Intermediate Range Power 1.89 x 10" amps RCS Boron Concentration 1582 ppm Bank D position 197 Steps Once criticality was achieved, the point of adding nuclear heat was determined in order to define the flux range for physics testing. W point of adding nuclear heat was determined to be 2.65x10# amps on Powa Range Nuclear Instrumentation (PRNI) channel N-44 that was coaw'M to the reactivity computer. Low power physics testing reactivity measurements were performed with flux level at least 30% below the point of adding nuclear heat.

h reactivity computer calibration was verified by making reactivity changes and comparing the reactivity indicated by the reactivity computer with values calculated from the Inhour Equation.

. 5.0 ALIrRODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT (FNP-1-STP-101)

PURPOSE The objectives of these measurements were to detennine the hot zero power (HZP) isothermal and moderator temperature coenicients for the all-rods-out (ARO) configuration and to measure the ARO, HZP witical boron (boron endpoint) concentration.

SUMMARY

OF RESULTS The ARO, HZP temperature coeffsients and the ARO boron endpoint concentration are tabulated below.

ARO, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT Rod Boron Measured ITC Design Calculated Configuration Conc. ITC Acceptance Criterion MTC (ppm) (pcmPF) (pcmPF) (pem*F)

All Rods Out 1552 +0.245 +0.08

2 +3.718 i Where:

ITC = loothermal Temperature Coefficient: (includes Doppler Coefficient of-1.587 pcmPF), and MTC = Cycle maximum Moderator Temperature Coefficient.

The MTC calculated frorn testing (+1.918 pcmPF) was normalized to the ARO design-predicted critical boron concentration (1554 ppm) and was correded for the predicted MTC increase with burnup (+1.8 pcmPF) to obtain the +3.718 pcmPF maximum MTC for Unit I, Cycle 16.

ARO, HZP BORON ENDPOINT CONCENTRATION Rod Coafiguration Measured Ca Design-Predicted Cs (ppm) (ppm)

All Rods Out 1551 1554 i 50 Since the maximum Cycle 16 MTC (+3.718 pcmPF) was less positive than the Technical Specifications limit of +7.0 pcmPF, no rod withdrawal limits were required The design review criterion for the ARO critical boron concentration was also satisfuul.

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6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-1-STP-101)

FURPOSE N objective of the bank worth measurements was to determine the integral reactivity worth of anch control and shutdown bank for comparison with the values predicted by design.

SUMMARY

OF RESULTS The rod worth measurements were performed using the Dynamic Rod Worth Measurement (DRWM) method. During this inessurement each bank was driven continuously from the fully withdrawn position to the fully inserted position at the maximum attainable stepping speed, without changing the boron concentration. The integral of the' reactivity change for each bank was measured using the reactivity computer. The measured bank worths satisfient the review criteria both for the banks measured individually and for the total worth of all banks combined.

Summary Of Control And Shutdown Bank Worth Measurements Centrol or l Predicted Bank Standard Review Measured Bank Percent i

Shutdown Worth (pcm) Criteria (pcm) Worth (pcm) Difference From Bank Predicted A 386.7

100 387.4 0.2 B (Ref) 1257.9

188.7 1303.8 3.7 C 865.3

129.8 925.5 7.0 D 1107.1

166 1098.4 -0.8 SD-A 1026.3

153.9 1003.8 -2.2 SD-B 976.6

146.5 1027.6 5.2 All banks 5619.9

561.9 5746.5 2.3 l

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7.0 POWER ASCENSION ACTIVITIES Upon completion of HZP physics testa, the following activities were performed during power a-alaa, or at full power. Sequencing of these activities was controlled under FNP-1-ETP-4441, Power Ascension Following Unit Uprate

1. Measurement of NIS intermediate range channel currents in order to scale the IR high flux trip and rod stop setpoints; Incore Excore AFD channel recalibration at 48% power.

2. Core hot channel factor surveillance and Excore Detector calibration confirmation at 48% power.

3. Secondary side walk downs and system msponse comparisons at approximately 48%,70%,80%,90%,95% and 100% power.

4. Steam generator level control testirig at approximately 95% power.

5. Tavg optimization testing, Equilibrium Flux Map, B anciar Coolant System flow measurement and evaluation of the need to rescale OPAT and OTAT protection loops to the loop ATs measured during the RCS flow test at 100% power.

SUMMARY

OF RESULTS In order to invoke the Technical Specification 3.10.3 test exceptions for HZP physics tests, intermediate and power range trip setpoints ofless than or equal to 25% power were used for initial reactor startup and physics testing Following completion of physics tests, the NIS power range high flux trip setpoint was increased to 80% to allow power escalation above 25%. The 80% setpoint (vice 109%) was administratively imposed to address the possibility that the power range channels initially could be indicating nonconservatively. Intermediate Range detector currents measured at approximately 48% power were used to ,

calculate Rod Stop and High Flux Trip Setpoint data for Intermediate Range i channels N35 and N36. !

4 At approximately 48% power, a full core flux map was obtained for the " base case" map for the Incore-Excore calibration test. Five additional (quarter-core) !

flux maps were performed at various positive and negative axial offsets in order to develop equations relating detector current to incore axial offset.

While holding at this power level, Maintenance recalibrated the power range NIS Channels N41 through N44, and a full core flux map at equilibrium Xenon conditions was obtained for evaluation of hot channel factors and confirmation of excore detector calibration. These results were satisfactory and are summarized in Tables 7.1 and 7.2.

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e Table 7.1 Detector Current Versus Axial Offset Equations Obtained From Incore-Excore Calibration Test CHANNELN41:

I-Top = 0.8846

AO + 174.4934 A I-Bottom = -1.0933

AO + 167.3067 A CHANNEL N42:

I-Top = 0.8%1

AO + 171.6566 A I-Bottom = -1.1154

AO + 163.0855 A CHANNELN43: !

I-Top = 0.8883

AO + 177.7344 A I l

I-Bottom = -1.2155

AO + 183.8158 pA CHANNEL N44: l I-Top = 0.8924

AO + 170.8624 A I-Bottom = -1.1490

AO + 167.8359 pA i

Table 7.2 Summary Of Power Ascension Full Core Flux Map Data PARAMETER M &_121 MAP 399 Avg. % Power 49% 100 %

Max. Power Tilt

1.0063 1.0051 Avg. Core AO 0.969 -1.289 Max. FAH 1.5884 1.4958 FAH Limit 1.957 1.70 FMZ) Steady State 2.0801 1.9118 FdZ) SS Limit 5.0000 2.5000 FfZ) Transient 2.0757 1.8318 FfZ) Tran. Limit 4.0832 2.0470

Calculated Power Tilts based on assembly FAH from all assemblies.

l Secondary side walkdowns, instrument scaling data recording, MCB indication !

and alarm evaluations, plant computer indication and alarm evaluations, and l controls systems dynamic response and stability evaluations were performed at various power level plateaus as controlled by FNP-1-ETP-4441. These evaluations confirmed =, wad plant response to uprated power level and design parameter predictions.

At approximately 95% power, FNP-1-ETP-4445, " Steam Generator Water Level Control Testing," was performed. This procedure placed ~ the level control system in " Manual," removed the last NCB card in the loop and placed it on an extender board and placed the loop back to " Auto ". With the NCB card on an extender board, a 5% setpoint change wa6 introduced at the "setpoint" input. W system pesformance was monitored and data was collected for review by SNC and SCS pwsc,c! present during testing. W acceptance criteria was that the steam generator water level control system would retum steam generator median level to the desired setpoint i 2% with dampening oscillations within approximately 3 time constants. ;

l Testing showed that the steam generator ester level control system was stable 1 i

and did not require tuning. Overall, the steam generator medial level overshot the setpoint by 1 to 2% and d=,vaad out to the setpoint (* 2%) in about 12.5 l minutes or 3 time constants (based on 250 second time constant). j l

At 100% power, turbine generator VWO and RCS T , optimization testing was performed to verify design predictions for T and turbine steam flow margin. h optimum T , was determined to be 575 'F, one degree higher than predicted due to steam generator plugging being higher than anticipated.

Once the plant had stabilized to equilibiium xenon conditions at the uprated 100% power level (2775 MWt), the RCS Flow Measurement (FNP-1-STP-115.1) and a. full-core flux map were performed concurrently to permit evaluation of the effects of hot-leg thermal streaming. h map was also used to initiate monthly surveillance of core hot channel factors, incore thermocouples, and excore NIS channel (Incoro-Excore) calibration. In >

addition, calorimetric data and RCS spare RTD data were used to ddermine the RCS loop 100% ATs for determination of the need to perform additional rescaling of the OPAT and OTAT protection channels for Cycle 16. Rescaling of affected instmment loops used the optimum Tavg value of $75 'F and measured 100% AT values. .

i Following completion ofinstmment loop rescaling the unit wu ramped down to appmximately 95% core power to the pre-uprate 100% thermal power rating of 2652 MWt. Data was taken at this power level and compared to the pre-uprate tuit>ine baseline test data to verify that no unexplained performance degradation had occurred and also to establish the desired Tu at which to perform the remainder of the testing.

The unit was then ramped back to 100% rated thermal power, maintaining the desired Tu, and turbine uprate guarantee testing was performed. This testing determined that the unit uprate achieved an increase in electrical output of l approximately 25 MWe. On January 21,1999, following completion of the l uprate guarantee testing evolutions, final assessment ofunit main control board I indications and plant computer alarms was conducted and confirmed no of- l normal indications or alarms that were related to the unit uprate. This I completed the testing activities specified by FNP-1-ETP-4441, " Power Ascension Following Unit Uprate."

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D:ve Morey S uthern N:cle:s Mce President ~ Operating Compaq,Inc.

Farley Project Post Ofhce Box 1295 Birmingham, Alabama 35201 Tel 205992 5131 SOUTHERN April 20, 1999 COMPANY Energy to Serve nurWorld*

Docket No.: 50-348 NEL-99-0170 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit 1 Cycle 16 Startun Test Report Ladies and Gentlemen:

In accordance with the reporting requirements of Technical Specification 6.9.1.1, Southern Nuclear Operating Company is submitting a Startup Test Report for Farley Nuclear Plant Unit 1 Cycle 16.

Should you have any questions, please advise. There are no NRC commdments in this submittal.

Paanac*kily submitted, Dave Morey j JAC/MGE/maf: pwrup99. doc /

Enclosure:

Startup Test Report Unit 1 Cycle 16 9904270100 990420 PDR

'I P ADOCK 05000348 PDR t

f,5 ,

, l i

l Page 2 i U. S. Nuclear Regulatory Commission l cc: Southern Nacient Operating Company Mr. L M. Stinson, General Manager - Farley i U. S. Nuclear Regulatory Commissiost Washington. D. C.

Mr. J. I. T _ - . laconsing Project Manager- Farley U. S. Nm '--- Rei-% C---:--6 Region H Mr. L A. Reyes, F=pn' Adnumstraser Mr. T. P. Johnson, Senior Resident laspector- Farley

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SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT Startup Test Report O' nit 1 Cycle 16 l

SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT Startup Test Report Unit 1 Cycle 16 TABLE OF CONTENTS Section Subiect Eags 1.0 Introduction .. .. .... . ......................................................I 2.0 Unit 1 Cycle 15 - 16 Core Refueling ................................... ........... 2 3.0 Control Rod Drop Time Measurement ......... ............... ....... ... . ...... 3 4.0 Initi al Criti cality. . . . . . . . . . . . . . . . . . . . . . ... . .... . . . . .. . . . .. . . . . . . . . .. . . . . .. . . .. . . . . . . 5 5.0 All-Rods-Out Isothermal Temperature Coefficient and B oron Endpoint . . . .. . . . . . . . . . . . . . . . . . . . . . .. . .. .. . . . . . . .. .. . . . . . . .. . .. . . . . . . . . . . . . . . 6 6.0 Control and Shutdown Bank Worth Measurements...... ..... ............. 7 7.0 Power Ascension Activities........ ........... ........... . ...... ... .. ............. . 8 i

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1.0 INTRODUCTION The Joseph M. Farley Unit I Cycle 16 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, Core Operating Limits Report, or values in the FSAR safety analysis.

The Unit 1 Cycle 16 core has been uprated to increase the NSSS power to 2785 l MWth (core full power of 2775 MWth plus 10 MWth reactor coolant pump heat). The uprate design change was accomplished under 10 CFR 50.59 and associated Technical Specifications Amendment Number 137. The Cycle 16 fresh fuel has also been designed to provide 1) improved fuel skeleton i stability under irradiation; 2) improved corrosion performance; and 3) l additional measures to control fuel rod internal pressures at high burnups. The additional VANTAGE + fuel assembly design features adopted for Cycle 16 to obtain improved fuel performance include ZIRLO Mid and IFM grids, ZIRLO i guide thimble and instrument tubes, annular fuel pellets in the top six inches of IFBA rods, and 1.25X IFBA at 100 psig Helium backfill pressure. Also to reduce corrosion and the propensity for axial offset anomaly (AOA), insertion of thimble plugs was re-introduced for Cycle 16. The reload design for this cycle utilizes 68 fresh feed ZIRLO clad VANTAGE + assemblies with the above design features, 61 once burned ZIRLO clad VANTAGE + fuel assemblies 28 twice burned VANTAGE 5 Zircaloy clad fuel assemblies. The secondary sources are located at D-08 & M-08 within once burned assemblies, as was the case with the previous cycle. The loading pattern places RCCAs into fuel assemblies which will not exceed 40,000 MWD /MTU burnup at EOL.

'Ihe design depletion of reactivity of the Cycle 16 core is 18,100 MWD /MTU j with an allowed power coast down of up to 19,300 MWD /MTU. ;

2.0 UNIT 1 CYCLE 15 - 16 CORE REFUELING Unloading of the Cycle 15 core into the spent fuel pool commenced on 10/24/98. During the omond, each ' fuel assembly was inspected with binoculars for indications of damage or other problems. No indications of physical damage were found. White or grayish deposits were observed but to a much lesser degree than seen in the previous cycle core omond Fuel oxide measurements were not performed for the cycle 15 core omond assemblies.

Since the fuel inspecdons revealed no fuel damage or defects, no revisions to the )

original Cycle 16 core loading pattern were required. Cycle 16 Core reload x=- = +i on 11/30/98 and was completed on 12/02/98. ;

REFERENCES

1. Procedure FP-APR-RIS, J. M. Farley Unit 1 Cycle 15-16 Refueling.

2. Westinghouse Wr'AP 15110, The Nuclear Design and Core Management of the. Joseph M. tarley Unit 1 Power Plant, Cycle 16.

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4 3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-1-STP-112)

FURPOSE j The purpose of this procedure was to measure the drop time of all control rods under hot full-flow conditions in the reactor coolant system in order to ensure l compliance with Technical Specifw.ations requirements.

SUMMARY

OF RESULTS For the hot full-flow condition (Tavg 2 541*F and all reactor coolant pumps operating), Technical Specification 3.1.3.4 requires verifw.ation that each ,

control rod will insert in s 2.7 seconds when tripped from the fully-withdrawn position. For this test, an automatic data acquisition system provided by the Analysis and Measurement Services Corporation (AMS) was used to obtain drop time data from an entire rod bank (8 rods) at a time. Individual control rod drop times are shown in Figure 3.1. All control rod drop times were measured to be less than 2.7 seconds. The longest drop time recorded for time i to dashpot entry was 1.86 seconds for rod B6. Mean drop times are I l

summarized below.

RCS Conditions Mean Time to Mean Time to Dashpot Entry Dashpot Bottom Hot full-flow 1.62 sec. 2.29 sec.

l To confirm normal CRDM operation prior to conducting the rod drop test, the !

Verification of Rod Control System Operability (IWP 0-ETP-3643) was also performed using the AMS System to acquire stepping data for an entire rod i bank at a time. In this test, the stepping waveforms of the stationary, !! fit and movable gripper coils were examined for anomalies; rod speed was measured; ami the functioning of the Digital Rod Position Indicator (DRPI) and bank i overlap unit were checked. In addition, the bank overlap unit settings for the fully withdrawn rod position to 226 steps were verified to be correct. Timing measurements were also performed on the stepping waveforms for CRDM Logic Cabinet performance testing. All results were satisfactory.

Figure 3.1 FNP Unit 1 Cycle 16 ControlRod Drop Times R P N M L K J H O F E D C B A 1

14,~5 eja L,68

.y@ 14,8a; LNj$ LISP Laph 2 1483r 14de LW L36Y'i 3 145e idow Les iden j 14y L38)@ L34$ L3Bih - 4

@f 14 9 L#b 5 4 873 142a;". l6t( L61s u 12"4 c IJB g.:" L969

.x kJ. 6 f, -

L38V L3th L34 A L3Ts; L35M L34& LS6s 6 B.stp; L57::g; L57f? L67g Ll?P L13j L18 fly L397 7 1.58 1 LS5;:, LJ72 LWV L38k L49$. L33% (40N 8 t 884' 145y LJS,y 1404 i

LSTj f L355 L33 ) ' BM 9 i l e t,;' I.A.L., . L6e. , ~ L66; ,

12. ,4 ..

- - t # L..

t L61n.+ - . -~

RJ7h LISP L3fg L33% L404 L33M BJ53 10 Lap:i- '1 imu L11 e L35Ii 11 140. . V.

1.56 '

14 t*

- BA W. . <

L36lifi L308- L331 BR3, 12 LMg idem, 2

L35)g L35j - 13 4 North L78f? K44s 146 L4U m L38d' L35E < 14 X.XX 4-Breakar"opeams"to Dashpot entry X.XX e-Breaker"openms"to Dashput hanam 15 RCS Tageka 544 'F RCS Pressure. 2250 nsin

% Flow: 100 %

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4.0 INITIAL CRITICALITY (FNP-1-STP-101)

FURPOSE h purpose of this evolution is to achieve initial criticality under carefully controlled conditions, establish the upper flux limit for conducting zero power physics tests, and operationally verify the calibration of the reactivity computer.

SUMMARY

OF RESULTS Initial reactor criticality for Cycle 16 was achieved during dilution mixing on 12/27/98. W reactor was allowed to stabilize at the following conditions.

RCS Pressure 2235 psig RCS Temperature 548.4 'F Intermediate Range Power 1.89 x 10" amps RCS Boron Concentration 1582 ppm Bank D position 197 Steps Once criticality was achieved, the point of adding nuclear heat was determined in order to define the flux range for physics testing. W point of adding nuclear heat was determined to be 2.65x10# amps on Powa Range Nuclear Instrumentation (PRNI) channel N-44 that was coaw'M to the reactivity computer. Low power physics testing reactivity measurements were performed with flux level at least 30% below the point of adding nuclear heat.

h reactivity computer calibration was verified by making reactivity changes and comparing the reactivity indicated by the reactivity computer with values calculated from the Inhour Equation.

. 5.0 ALIrRODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT (FNP-1-STP-101)

PURPOSE The objectives of these measurements were to detennine the hot zero power (HZP) isothermal and moderator temperature coenicients for the all-rods-out (ARO) configuration and to measure the ARO, HZP witical boron (boron endpoint) concentration.

SUMMARY

OF RESULTS The ARO, HZP temperature coeffsients and the ARO boron endpoint concentration are tabulated below.

ARO, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT Rod Boron Measured ITC Design Calculated Configuration Conc. ITC Acceptance Criterion MTC (ppm) (pcmPF) (pcmPF) (pem*F)

All Rods Out 1552 +0.245 +0.08

2 +3.718 i Where:

ITC = loothermal Temperature Coefficient: (includes Doppler Coefficient of-1.587 pcmPF), and MTC = Cycle maximum Moderator Temperature Coefficient.

The MTC calculated frorn testing (+1.918 pcmPF) was normalized to the ARO design-predicted critical boron concentration (1554 ppm) and was correded for the predicted MTC increase with burnup (+1.8 pcmPF) to obtain the +3.718 pcmPF maximum MTC for Unit I, Cycle 16.

ARO, HZP BORON ENDPOINT CONCENTRATION Rod Coafiguration Measured Ca Design-Predicted Cs (ppm) (ppm)

All Rods Out 1551 1554 i 50 Since the maximum Cycle 16 MTC (+3.718 pcmPF) was less positive than the Technical Specifications limit of +7.0 pcmPF, no rod withdrawal limits were required The design review criterion for the ARO critical boron concentration was also satisfuul.

1 r ]

9 0

+

6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-1-STP-101)

FURPOSE N objective of the bank worth measurements was to determine the integral reactivity worth of anch control and shutdown bank for comparison with the values predicted by design.

SUMMARY

OF RESULTS The rod worth measurements were performed using the Dynamic Rod Worth Measurement (DRWM) method. During this inessurement each bank was driven continuously from the fully withdrawn position to the fully inserted position at the maximum attainable stepping speed, without changing the boron concentration. The integral of the' reactivity change for each bank was measured using the reactivity computer. The measured bank worths satisfient the review criteria both for the banks measured individually and for the total worth of all banks combined.

Summary Of Control And Shutdown Bank Worth Measurements Centrol or l Predicted Bank Standard Review Measured Bank Percent i

Shutdown Worth (pcm) Criteria (pcm) Worth (pcm) Difference From Bank Predicted A 386.7

100 387.4 0.2 B (Ref) 1257.9

188.7 1303.8 3.7 C 865.3

129.8 925.5 7.0 D 1107.1

166 1098.4 -0.8 SD-A 1026.3

153.9 1003.8 -2.2 SD-B 976.6

146.5 1027.6 5.2 All banks 5619.9

561.9 5746.5 2.3 l

a

7.0 POWER ASCENSION ACTIVITIES Upon completion of HZP physics testa, the following activities were performed during power a-alaa, or at full power. Sequencing of these activities was controlled under FNP-1-ETP-4441, Power Ascension Following Unit Uprate

1. Measurement of NIS intermediate range channel currents in order to scale the IR high flux trip and rod stop setpoints; Incore Excore AFD channel recalibration at 48% power.

2. Core hot channel factor surveillance and Excore Detector calibration confirmation at 48% power.

3. Secondary side walk downs and system msponse comparisons at approximately 48%,70%,80%,90%,95% and 100% power.

4. Steam generator level control testirig at approximately 95% power.

5. Tavg optimization testing, Equilibrium Flux Map, B anciar Coolant System flow measurement and evaluation of the need to rescale OPAT and OTAT protection loops to the loop ATs measured during the RCS flow test at 100% power.

SUMMARY

OF RESULTS In order to invoke the Technical Specification 3.10.3 test exceptions for HZP physics tests, intermediate and power range trip setpoints ofless than or equal to 25% power were used for initial reactor startup and physics testing Following completion of physics tests, the NIS power range high flux trip setpoint was increased to 80% to allow power escalation above 25%. The 80% setpoint (vice 109%) was administratively imposed to address the possibility that the power range channels initially could be indicating nonconservatively. Intermediate Range detector currents measured at approximately 48% power were used to ,

calculate Rod Stop and High Flux Trip Setpoint data for Intermediate Range i channels N35 and N36. !

4 At approximately 48% power, a full core flux map was obtained for the " base case" map for the Incore-Excore calibration test. Five additional (quarter-core) !

flux maps were performed at various positive and negative axial offsets in order to develop equations relating detector current to incore axial offset.

While holding at this power level, Maintenance recalibrated the power range NIS Channels N41 through N44, and a full core flux map at equilibrium Xenon conditions was obtained for evaluation of hot channel factors and confirmation of excore detector calibration. These results were satisfactory and are summarized in Tables 7.1 and 7.2.

8

e Table 7.1 Detector Current Versus Axial Offset Equations Obtained From Incore-Excore Calibration Test CHANNELN41:

I-Top = 0.8846

AO + 174.4934 A I-Bottom = -1.0933

AO + 167.3067 A CHANNEL N42:

I-Top = 0.8%1

AO + 171.6566 A I-Bottom = -1.1154

AO + 163.0855 A CHANNELN43: !

I-Top = 0.8883

AO + 177.7344 A I l

I-Bottom = -1.2155

AO + 183.8158 pA CHANNEL N44: l I-Top = 0.8924

AO + 170.8624 A I-Bottom = -1.1490

AO + 167.8359 pA i

Table 7.2 Summary Of Power Ascension Full Core Flux Map Data PARAMETER M &_121 MAP 399 Avg. % Power 49% 100 %

Max. Power Tilt

1.0063 1.0051 Avg. Core AO 0.969 -1.289 Max. FAH 1.5884 1.4958 FAH Limit 1.957 1.70 FMZ) Steady State 2.0801 1.9118 FdZ) SS Limit 5.0000 2.5000 FfZ) Transient 2.0757 1.8318 FfZ) Tran. Limit 4.0832 2.0470

Calculated Power Tilts based on assembly FAH from all assemblies.