ML20151V699

Snoc Jm Farley Nuclear Plant Startup Test Rept Unit 2 Cycle 13. with

ML20151V699
Person / Time
Site:	Farley
Issue date:	09/11/1998
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9809150055
Download: ML20151V699 (16)

Text

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, Dave Morey Southern Nucl:ar

. Vice President Op: rating Comp:ny

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'farley Project P.O. Box 1295 Birmingham. Alabama 35201 Tel 205.992.5131 September 11, 1998 SOUTHERN h gg Energyto ServeYourWorld" Docket No.: 50-364. 10 CFR 50.4 i

U. S. Nuclear Regulatory Commission ,

ATTN: Document Control Desk Washington, DC 20555 j Joseph M. Farley Nuclear Plant - Unit 2 Cvele 13 Startuo Test Report Ladies and Gentlemen: 1

- 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 2 Cycle 13.

Should you have any questions, please advise.

Respectfully submitted,  !

I Dave Morey RDR/JAC/hl"sE:marpwrup45. doc Enclosure l cc: Mr. L. A. Reyes, Region II Administrator Mr. J.1. Zimmerman, NRR Project Manager 1 Mr. T. P. Johnson, Plant Sr. Resident Inspector  !

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PDR ADOCK 05000364 l' P PDR ,

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SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT Startup Test Report Unit 2 Cycle 13 l

l SOUTHERN NUCLEAR OPERATING COMPANY  ;

JOSEPH M. FARLEY NUCLEAR PLANT I Startup Test Report Unit 2 Cycle 13 TABLE OF CONTENTS 1

Section Subic.c.t has 1.0 Introduction .........................................1 l 2.0 Unit 2 Cycle 12 - 13 Core Refueling.. .. . . . . . . . . . . . . . . . . . ....2 l

3.0 Control Rod Drop Time Measurement ................. ................3 1 4.0 Initial Criticality..... ... . .... . . . . .. .. . . . .. . .. .. . . . . . . . . .. . .. . . 5 4

5.0 All-Rods-Out Isothermal Temperature Coeflicient and Boron Endpoint.. .... .. ......... . . . . . . . . . . . . . . . . . . .........6 6.0 Control and Shutdown Bank Worth Measurements. ... . . . . . . . .7 7.0 Power Ascension Activities... . ...... . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. 8 l I

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l 1.0' INTRODUCTION l L

The Joseph M. Farley Unit 2 Cycle 13 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 l l Limits Report, or values in the FSAR safety analysis.

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[ The Unit 2 Cycle 13 core has been uprated to increase the NSSS power to 2785 l l MWth (core full power of 2775 MWth plus 10 MWth pump heat). The uprate i design change was accomplished under 10 CFR 50.59 and associated Technical l Specifications Amendment Number 129. The Cycle 13 fresh fuel has also been i designed to provide: 1) improved fuel skeleton stability under irradiation; 2) j

( improved corrosion performance; and 3) additional measures to control fuel rod ,

l internal pressures at high burnups. The additional Vantage + fuel assembly  !

l design features adopted for Cycle 13 to obtain improved fuel performance include l ZIRLO Mid and IFM grids, ZIRLO guide thimble and instrument tubes, annular  :

i fuel pellets in the top six inches ofIFBA rods, and 1.25X IFBA at 100 psig )

Helium backfill pressure. Also to reduce corrosion and the propensity for axial ,

l ' offset anomaly (AOA), insertion of thimble plugs was re-introduced for Cycle 13. I The reload design for this cycle utilizes 68 fresh feed ZIRLO clad VANTAGE +

l assemblies with the above design features,65 once burned ZIRLO clad l VANTAGE + fuel assemblies and 24 twice burned Zircaloy-4 clad VANTAGE 5 l fuel assemblies. ZIRLO cladding is designed to provide better corrosion  ;

L resistance than the Zircaloy-4 cladding and has, based upon oxide measurements l - as discussed in Section 2.0, performed as expected. 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  ;

l will not exceed 40,000 MWD /MTU burnup at EOL. The design depletion of {

l reactivity of the Cycle 13 core is 18,300 MWD /MTU with an allowed power  :

l coast down of up to 19,500 MWD /MTU. I l

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9 a 2.0 . UNIT 2 CYCLE 12 - 13 CORE REFUELING

. Unloading of the Cycle 12 core into the spent fuel pool commenced on 4/4/98.

During the offload, each fuel assembly was inspected with binoculars for l

i indications of damage or other problems. No indications of physical damage l

were found. White or grayish deposits were observed but to a much lesser degree  ;

than seen in the previous cycle core offload.

As follow-up to previous oxide measurements related to zinc addition, oxide  !

measurements were again performed. The inspections utilized a moveable eddy )

current probe in a special underwater fixture mounted on the fuel racks in the ,

spent fuel pool. In order to provide baseline data and to validate the accuracy of !

the eddy current equipment, a discharged fuel assembly which had previously i been subjected to oxide thickness measurements was re-tested. Following core i unload, oxide thickness measurements were performed on 12 additional fuel 1 assemblies. These measurements showed that the oxide deposits on the fuel rods j were within the expected, normal tolerances. Crud scrapings were not performed.

Since the fuel inspections revealed no fuel damage or defects and oxide f

measurements were within acceptance criteria, no revisions to the original Cycle 13 core reload pattern were required. Cycle 13 Core reload commenced on 4/26/98 and was completed on 4/28/98.

REFERENCES

1. Procedure FP-APR-R12, J. M. Farley Unit 2 Cycle 12-13 Refueling.

2. Westinghouse WCAP 15035, The Nuclear Design and Core Management of the Joseph M. Farley Unit 2 Power Plant, Cycle 13.

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

PURPOSE 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 compliance with Technical Specification requirements.

SUMMARY

OF RESULTS For the hot full-flow condition (Tavg ;t 541 F and all reactor coolant pumps  ;

operating), Technical Specification 3.1.3.4 requires verification 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 was 2.14 seconds for rod B6. Mean drop times are summarized below.

RCS Conditions Mean Time to Mean Time to Dashpot Entry Dashpot Bottom Hot full-flow 1.43 sec. 1.95 sec.

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

Verification of Rod Control System Operability (FNP-0-ETP-3643) was performed also using the AMS System to acquire stepping data for an entire rod bank at a time. In this test, the stepping waveforms of the stationary, lift and movable gripper coils were examined for anomalies; rod speed was measured; and the functioning of the Digital Rod Position Indicator (DRPI) and bank 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.

In additic;n to control rod drop time measurement, RCCA Eddy Current Testing was performed on each RCCA once the fuel had been offloaded to the Spent Fuel Pool. The purpose of the eddy current testing was to identify and, if necessary, l characterize excessive wear on the cladding of the RCCA rodlets. The testing was performed to continue trending previously identified wear scars. All RCCAs met inspection criteria for continued use. Based upon these inspection results, the fully withdrawn position was retained at 226 steps for Cycle 13.

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Figure 3.1 FNP Unit 2 Cycle 13 Control Rod Drop Times R P N M L K J H G F E D C B A )

1 1.46 1.44 > 1.42-2.04' l.95 1.95 2 <

l 1.40- 1.45 1.94 ': 2.83( 3 1.44 1.40 1.40 . 1.41..

1.97 '. 1.97 -: 1.93 1.92 ; 4 1.41. 1.42:

1.93 2.00: 5 I 1.45 1.38 1.41- 1.40 1.42 1.39 1.57 2.00 1.90 1.92 1.89 1.90 1.88 : 2.14' 6 1.41 . 1.36 : 1.38 1.46 l 1.90 1.83- 1.94- 1.98 7 ,

1.42; 1.38 1.38 1.52 l

1.96 1.95- 1.91' 2.01 8 1.42 . 1.39 1.36' l.43 .

1.95 : 1.94 1.86 ~ 1.92 ' 9 1.42 1.40 1.39 1.41 . 1.40 1.39 1.44 . j 1.95 1.89 l.96 1.% . 1.92 1.88 2.00 10 1.42 1.43 ' ,

1.91 1.95. 11 l.41 1.39- 1.43 1.42 -

1.91 1.91' l.40 1.92 12 '

I 1.42 1.42 - 1 l

l 1.95' l.89 , . 13 i l 4 North 1.52 1.47 1.43

l. 2.06' l.97 2.01: 14 X.XX .-Breaker " opening" to Dashpot entry j X.XX *-Breaker " opening" to Darhpot bottom 15 l

- RCS Temperature: 548 *F I RCS Pressure: 2235 osin i

% Flow: 100 %

i 4.0 INITIAL CRITICALITY (FNP-2-STP-101)

PURPOSE The 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. j

SUMMARY

OF RESULTS  !

Initial reactor criticality for Cycle 13 was achieved during dilution mixing on i 5/15/98. The reactor was allowed to stabilize at the following conditions.

RCS Pressure 2235 psig RCS Temperature 547.4 F )

Intermediate Range Power 1.1 x 10'" amps RCS Boron Concentration 1499 ppm Back 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. The point of adding nuclear I 4

- heat was determined to be 2.263x10 amps on Power Range Nuclear

! Instrumentation (PRNI) channel N-44 that was connected to the reactivity l computer. Low power physics testing reactivity measurements were performed l with flux level at least 30% below the point of adding nuclear heat. The 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.

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5.0 ALI-RODS-OUT ISOTiiERMAL TEMPERATURE COEFFICIENT AND

. BORON ENDPOINT (FNP-2-STP-101)

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

SUMMARY

OF RESULTS The ARO, HZP temperature coefficients 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) (pcmFF) (pcmFF) (pcmFF)

All Rods Out 1508 -0.285 --0.199 i 2 +3.3 8 Where:

ITC = Isothermal Temperature Coeflicient: (includes Doppler Coefficient of-1.595 pcmFF), and MTC = Cycle maximum Moderator Temperature Coefficient.

The MTC calculated from testing (+1.31 pcmFF) was normalized to the ARO design-predicted critical boron concentration (1524 ppm) and was corrected for the predicted MTC increase with burnup (+2.2 pcmFF) to obtain the +3.38 pcmFF maximum MTC for Unit 2, Cycle 13.

ARO, HZP BORON ENDPOINT CONCENTRATION Rod Configuration Measured Cn Design-Predicted Cn (ppm) (ppm)

All Rods Out 1508 1524 i 50 Since the maximum Cycle 13 MTC ( .38 pcmFF) was less positive than the Technical Specification limit of +7.0 pcmFF, no rod withdrawal limits were required. The design review criterion for the ARO critical boron concentration was also satisfied.

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6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-2-STP-101)

PURPOSE The objective of the bank wonh 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 Dynamic Rod Wonh Measurement (DRWM) method. During this measurement 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 wonhs satisfied the review criteria both for the banks measured individually and for the tctal worth of all banks combined.

Summary Of Control And Shutdown Bank Worth Measurements Control or ' Predicted Bank Standard Review Measured Bank Percent Shutdown Wonh (pcm) Criteria (pcm) Worth (pcm) DifTerence From Bank Predicted A 450.0

• 100 475.4 5.6 B (Ref) 1287.8

• 193 1409.7 9.5 C 790.4

• 119 812.8 2.8 D 1110.8 167 1164.2 4.8 SD-A 1166.2 175 1260.9 8.1 SD-B 939.3 141 952.3 1.4 All banks 5744.5 574.4 6075.3 5.76

u 7.0 POWER ASCENSION ACTIVITIES Upon completion of HZP physics testing, power ascension testing was conducted.

Unit ramping and testing proceeded very smoothly with plant control systems exhibiting very stable performance. The sequence of these activities was controlled under plant procedures FNP-0-SOP-103, Return To Service Checklist and Return To Service Systems Lineup, and FNP-2-ETP-4441, Power Ascension Following Unit Uprate, and completed on June 15,1998. Key activities performed during power ascension or at full power included the following.

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Correlation of auxiliary reactor power indications to reactor power level determined plateaus. from calorimetric measurements at various reactor power 2.

Verification low power. ofproper steam generator level control dynamic response at 3.

Power Range Nuclear Instrumentation (PRNIS) Axial Offset Calibration

4. and initial core hot channel factors surveillance.

Secondary side walk-downs to confirm expected system response at plateaus

5. ofapproximately 48%,70%,80%,90%,95% and 100% reactor power.

Verification ofinstrument scaling projections for specified control and protection loops at various reactor power plateaus.

6 Confirmation ofproper main control board (MCB) indications, MCB

7. annunciation and plant computer response at various reactor power plateaus.

Verification high power. ofproper steam generator level control dynamic response at 8.

Determination ofoptimum Median Tavsand corresponding main turbine governor valve position.

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Reactor coolant system flow measurement, core hot channel factors

10. surveillance and Excore Detector calibration surveillance at 100% power.

Evaluation of the OPAT and OTAT protection loops scaling based upon the 100% loop ATs measured during the RCS flow test.

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Turbine performance testing and final confirmation of proper MCB indications, MCB annunciation and plant computer response at 100%

reactor power.

SUMMARY

OF RESULTS During power escalation, stabilization plateaus ofapprcximately 18 %, 33 %,48 %

70%,80%,90%,95% and 100% were selected to confirm agreement ofdiverse reactor power level indication and for conducting of selected system performance testing or required surveillance testing. Reactor power level indications from PRNIS, RCS loop ATs, turbine first stage impulse i pressure (P mp), and calorimetric power were compared under stable conditions at each of the above plateaus. Agreement of these diverse indications ofpower level was within the review criteria for continued power escalation without any instmment channel re-calibrations being required except for the Excore PRNIS axial ofTset calibration normally performed at approximately 33% to 48% power.

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7.0 POWER ASCENSION ACTIVITIES l Upon completion of HZP physics testing, power ascension testing was conducted. I Unit ramping and testing proceeded very smoothly with plant control systems exhibiting very stable performance. The sequence of these activities was controlled under plant procedures FNP-0-SOP-103, Return To Service Checklist and Return To Senice Systems Lineup, and FNP-2-ETP-4441, Power Ascension Following Unit Uprate, and completed on June 15,1998. Key activities performed during power ascension or at full power included the following.

1. Correlation of auxiliary reactor power indications to reactor power level determined from calorimetric measurements at various reactor power  ;

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2. Verification of proper steam generator level control dynamic response at i Iow power.

3. Power Range Nuclear Instrumentation (PRNIS) Axial Offset Calibration and initial core hot channel factors surveillance.

4. Secondary side walk-downs to confirm expected system response at plateaus of approximately 48%,70%, 80%,90%,95% and 100% reactor power.

5. Verification ofinstmment scaling projections for specified control and protection loops at various reactor power plateaus.

6. Confirmation of proper main control board (MCB) indications, MCB annunciation and plant computer response at various reactor power plateaus.

7. Verification of proper steam generator level control dynamic response at high power.

8. Determination of optimum Median Tavg and corresponding main turbine governor valve position.

9. Reactor coolant system flow measurement, core hot channel factors ,

suneillance and Excore Detector calibration surveillance at 100% power. l

10. Evaluation of the OPAT and OTAT protection loops scaling based upon the 100% loop ATs measured during the RCS flow test.

11. Turbine performance testing and final confirmation of proper MCB  ;

indications, MCB annunciation and plant computer response at 100% )

reactor power.

SDMMARY OF RESULTS During power escalation, stabilization plateaus of approximately 18%,33%,48%,

70%,80%,90%,95% and 100% were selected to confirm agreement of diverse reactor power level indication and for conducting of selected system performance testing or required surveillance testing. Reactor power level indications from PRNIS, RCS loop ATs, turbine first stage impulse pressure (P mp),

i and calorimetric power were compared under stable conditions at each of the above '

plateaus. Agreement of these diverse indications of power level was within the review criteria for continued power escalation without any instrument channel re-calibrations being required except for the Excore PRNIS axial offset calibration normally performed at approximately 33% to 48% power.

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The low power portion of FNP-2-ETP-4440, Steam Generator Water Level Control Testing, was performed during the chemistry hold at approximately 31%

power. The purpose of this procedure was to verify proper dynamic response of the steam generator level control system. The sequence of this procedure placed the level control system in " Manual." The last NCB card in the loop was removed and placed on an extender board and the level control system was then returned to

" Auto." With the NCB card on an extender board, a 5% setpoint change was  ;

introduced at the "setpoint" input. The level control system performance was  !

monitored and performance data was collected and reviewed by SNC and Westinghouse personnel who were present during testing. The acceptance criteria [

were that the steam generator water level control system would return steam j generator median level to the desired setpoint i2% with dampening oscillations i occurring within approximately 3 time constants. Testing at this power level -

showd that steam generator water level control system was stable and did not  !

require tuning. Overall, the S/G median level overshot the setpoint by I to 2%

and dampened out to the setpoint i2% in approximately 6 minutes or 1.5 time constants (based on a 250 second time constant).

i Also at this plateau, a full core flux map was obtained as the " base case" map for ;

the Incore-Excore cross-calibration test. Five additional (quarter-core) flux maps j were performed at various positive and negative axial offsets in order to develop  ;

equations relating detector cunrent to incore axial offset.

While awaiting a chemistry hold at this power level, the PRNIS channels were  ;

calibrated. Also a full core flux map at equilibrium Xenon conditions was obtained for evaluation of het channel factors and confirmation of PRNIS detector axial offset calibration. These results were satisfactory and are summarized in Tables 7.1 and 7.2.  ;

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Table 7.1 )

Detector Current Versus Axial Offset Equations Obtained From '

Incore-Excore Calibration Test l

CHANNEL N41:

I-Top = 0.9691

• AO + 173.2089 uA I-Bottom = -1.0480

• AO + 169.2889 uA ,

CHANNEL N42:

i I-Top = 0.9539

• AO + 174.2237 uA l

• AO I-Bottom = -1.0970 + 167.8269 uA l

CHANNEL N43:

i I-Top = 0.9843

• AO + 177.9756 uA  !

. I-Bottom = -1.0744

• AO + 171.6638 uA l t

l CHANNEL N44: )

I-Top = 1.0736

• AO + 191.5263 uA ,

I-Bottom = -1.2757

• AO + 190.7116 uA j Table 7.2 Summary Of Power Ascension Full Core Flux Map Data PARAMETER MAP 328 M AP 329 Avg. % Power 31% 100 %

Max. Power Tilt

• 1.0090 1.0077-Avg. Core AO 1.056 -3.463 Max. FAH 1.5274 1.475 FAH Limit 2.049 1.70 Fn(Z) Steady State 2.1179 1.9052 Fn(Z) SS Limit 5.0000 2.4975 Fn(Z) Transient 2.1568 1.9410 Fz(Z) Tran. Limit 4.2598 2.2342

• Calculated Power Tilts based on assembly FAH from all assemblies.

At approximately 48%,70%,80%,90%,95% and 100% reactor power, secondary side walk-downs, instrument scaling data comparisons, hiCB indicator response comparisons, and ccotrol systems response and stability evaluations were performed. Assessments of RCS loop T hot and AT scaling and margin to OPAT and OTAT reactor trip and high steam flow ESF setpoints were also performed during power ascension as designated by FNP-2-ETP-4441. Systems evaluated included: main steam; feedwater and feedwater heater vents and drains; steam generator level control; steam generator feed pump speed control; rod control; pressurizer level control; and pressurizer pressure control. Evaluations performed at the designated reactor power plateaus confirmed that plant instrument loop scaling, h1CB indicator response and control systems performance criteria were met. Ample margins to OPAT, OTAT and high steam flow setpoints were also confirmed at designated plateaus. The initial calibration of the steam flow instrument channels incorporated conservative scaling with respect to the high steam flow setpoint. Based on operating full power performance data, the steam flow instrument channels were rescaled to the specified setpoint and were renormalized to provide closer agreement with feedwater flow indication. Secondary side walkdowns identified various dump controllers and drain valves that were contributing to thermal losses and affecting secondary side thermal efficiency. These deficiencies were investigated and determined to be equipment maintenance issues and not related to process parameter changes associated with the Unit uprate. hiaintenance repairs were initiated and implemented to correct the secondary side thermal efficiency losses identified.

At approximately 90% power, the high power ponion of FNP-2-ETP-4440, Steam Generator Water Leve', Control Testing was conducted with results again demonstrating that the steam generator water level control system was stable and did not require tuning. Overall, the S/G median level overshot the setpoint by 1 to 2% and dampened out to the setpoint ( 2%) n approximately 6 minutes or 1.5 time constants (based on a 250 second time constant).

Following the high power level portion of the steam generator level control testing, the unit was ramped to 2652 h1Wth (approximately 95% core power level). Evaluation ofindicated main turbine #4 governor valve position versus expected and projection of the #4 governor valve position for 100% power were performed. Design projections were 0% and 50% open at 95% and 100% power, respectively, versus measured values of 10% and 43%, respectively.

The unit was then ramped to 2775 hiWth at approximately 1% reactor power per hour with the unit being stabilized during the ramp for confirmation of power level by calorimetric. Testing was then conducted to determine optimum median Tavsand turbine flow margin. Based on these measurements, the optimum median Tavswas determined to be 575 F. Since this was also the design value specified for initial calibration of program Trer, re-calibration of the Pressurizer Level, Rod Control, and Steam Dumps control loops was not required.

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After reaching full power equilibrium Xenon conditions, procedure FNP-2-STP-115.1, RCS Flow Measurement, was performed. 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 Technical Specifications. In addition, the RCS loop 100% ATs measured during this test I i

were used to evaluate the need to rescale the OPAT and OTAT protection channels. A full-core flux map was also concurrently performed to support surveillance for core hot channel factors, incore thermocouples, and PRNIS channel calibration (procedures FNP-2-STP-110, FNP-2-STP-108 and FNP-2-STP-121, respectively). RCS flow was determined by precision calorimetric method and for comparison purposes, by measurement of the loop elbow tap flow meter voltages. Results are given below in Table 7.3. Though lower than the previous cycle 12 measurement, flow results were very close to expected when re-insertion of thimble plugs and an increase in the equivalent l steam generator plugging level are considered. I Table 7.3 Unit 2 Cycle 13 RCS Flow Measurement Comparison 1

Cycle 12 Cycle 13 Cycle 13 l l Calorimetric Method Elbow Tap Method Basdine Calorimetric Elbow Measured IMfference Change Measured IMfference G ange

! Hows (gpm) Flows (gpm) Tap Flows tiow Data From From flow Data From From (gpm) (gpm) Basellae Preslaus (gym) Baseline Previous Cycle C)cle I mop-1 (A) 99169 94019 9785R 92753 -6.5% -I .3% 96486 2.7% -1.4%

l I mop-2 (H) 93364 88568 92744 87864 -5 9% -0.8% 90838 -2.7% -2.1%

! Imop-3 (C) 95997 93391 95298 93107 3.0% 0.3% 93400 -2.7% 2.0%

Total RCN 2kR530 27597N 285900 273725 5. l

• 6 -0 8% 280725 -2.7% -1.8%

l Notes: 1) Baseline flow was determined by precision calorimetric carly in plant life. (Determined for Elbow Tap Methodology comparisons.)

2) Stea.n generator plugging equivalent increased from previous cycle by 0.6% to a total of 7.3% equivalent plugged (plugged and sleeved).

1 3) %c Cycle 13 core design provided for thimble plugs to be reinstalled. Design predictions regarding thimble plug impact to RCS flow predicted a total flow decrease of 0.6%

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l Following completion of median T vg determination and RCS flow measurement, the unit was ramped down to approximately 95% core power (2652 MWth).

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' Turbine baseline testing at pre-uprate conditions was performed as a part of the turbine uprate contract provisions. This confirmed that no unexplained degradations in unit performance had resulted from outage work. j The unit was then ramped back to 100% rated thermal power, and turbine uprate l guarantee testing was performed. This testing determined that the unit uprate i uchieved an increase in electrical output of approximately 26 MWe. On June 15, l 1998, following completion of the uprate guarantee testing evolutions, final i assessment of unit main control t sard indications and plant computer alarms was l conducted and confirmed no off-normal indications or alarms that were related to the unit uprate. This completed the testing activities specified by FNP-2-ETP- ,

4441, Power Ascension Following Unit Uprate. i I

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