L-2013-226, Extended Power Uprate Cycle 26 Startup Report - Supplement 3

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Extended Power Uprate Cycle 26 Startup Report - Supplement 3
ML13234A338
Person / Time
Site: Turkey Point NextEra Energy icon.png
Issue date: 08/12/2013
From: Kiley M
Florida Power & Light Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2013-226
Download: ML13234A338 (22)
v • d • e


Text

August 12, 2013 L-2013-226 10 CFR 50.36 FPL.

U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, D.C. 20555-0001 Re:

Turkey Point Unit 3 Docket No. 50-250 Renewed Facility Operating License No. DPR-31 Extended Power Uprate Cycle 26 Startup Report - Supplement 3

References:

(1)

J. Page (NRC) to M. Nazar (FPL), Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Extended Power Uprate (TAC Nos. ME4907 and ME 4908), Accession No. ML11293A365, June 15, 2012.

(2)

M. Kiley (FPL) to U. S. Nuclear Regulatory Commission (NRC) (L-2012-426), Renewed Facility Operating License No. DPR-31 Extended Power Uprate Cycle 26 Startup Report, Accession No. ML12341A097, December 4, 2012.

(3)

M. Kiley (FPL) to U. S. Nuclear Regulatory Commission (NRC) (L-2013-073), Renewed Facility Operating License No. DPR-31 Extended Power Uprate Cycle 26 Startup Report-Supplement 1, Accession No. ML13071A345, February 26, 2013.

(4)

M. Kiley (FPL) to U. S. Nuclear Regulatory Commission (NRC) (L-2013-180), Renewed Facility Operating License No. DPR-31 Extended Power Uprate Cycle 26 Startup Report-Supplement 2, Accession No. ML13182A007, May 23, 2013.

Pursuant to Turkey Point Unit 3 Technical Specification 6.9.1.1, Florida Power & Light Company (FPL) is providing supplement 3 to the Unit 3 Cycle 26 Startup Report submitted on December 4, 2012 via Reference 2. The Startup Report and supplemental reports are required due to the implementation of Unit 3 Amendment No. 249 for the Extended Power Uprate (EPU)

Project that was issued via Reference 1. Supplement 1 to the initial Startup Report was submitted on February 26, 2013 via Reference 3.

Supplement 2 to the initial Startup Report was submitted on May 23, 2013 via Reference 4. At the time that supplement 2 to the initial Startup Report was issued, the remaining 10% transient ramp tests from 100% rated thermal power (RTP) to 90% RTP, and then from 90% RTP back to 100%

RTP had not been completed due to three unscheduled outages. The final tests have been completed and the results of the completed 100% RTP power ascension transient and steady state testing results for Unit 3 are being submitted in the attached final report.

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Page 2 of 2 Should you have any questions regarding this submittal, please contact Mr. Robert Tomonto, Turkey Point Licensing Manager, at 305-246-7327.

Sincerely, Michael Kiley Site Vice President Turkey Point Nuclear Plant Attachment cc:

USNRC Regional Administrator, Region II USNRC Project Manager, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 1 of 20 TURKEY POINT UNIT 3 CYCLE 26 SUPPLEMENT 3 EXTENDED POWER UPRATE STARTUP REPORT

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 2 of 20 IV V

Table of Contents Introduction Power Ascension Test Program Results Summary References 4

10 19 19

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 3 of 20

1.

Introduction The purpose of this Startup Report is to provide a summary description of the plant startup and power ascension testing performed at Turkey Point following the implementation of License Amendment No. 249 on the Extended Power Uprate (EPU) for Renewed Facility Operating License DPR-31 for Turkey Point Unit 3 that was issued on June 15, 2012

[Reference 1]. The amendment increased the authorized maximum steady-state reactor core power from 2300 megawatts thermal (MWt) to 2644 MWt. This Unit 3 Cycle 26 Startup Report is being submitted in accordance with Turkey Point Technical Specification 6.9.1.1, Items (2) and (4).

Technical Specification 6.9.1.1, Startup Report, states:

"A summary report of plant startup and power escalation testing shall be submitted following:... (2) amendment to the license involving a planned increase in power level,...

and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions of characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other conmmitments shall be included in this report. Subsequent Startup Reports shall address startup tests that are necessary to demonstrate the acceptability of changes and/or modifications."

All required testing per the Startup Test Program has been completed for Turkey Point Unit 3 Cycle 26.

The plant startup and power escalation testing verifies that key EPU core and plant parameters are operating as predicted. The major parts of this testing program include:

1) Cycle 26 core design summary,
2) Low power physics testing, and
3) Power ascension testing.

The test data collected during EPU power ascension summarized in this report concludes that all major systems, structures, and components (SSCs) performed as predicted and there was no adverse impact to the performance of the unit. All startup testing results related to core performance were found acceptable and previously reported in Reference 5.

Therefore, they are not repeated in this report. The EPU power ascension test data was found acceptable and previously reported in Reference 5. Copies of the completed EPU power ascension test procedures are available on site for review.

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 4 of 20 II.

Power Ascension Test Program The EPU power ascension test program consisted of a combination of normal startup and surveillance testing, post-modification testing, and power ascension testing deemed necessary to support acceptance of the proposed EPU. During the EPU start-up, power was increased in a slow and deliberate manner, stopping at pre-determined power levels for steady-state data gathering and formal parameter evaluation. These pre-determined power levels are referred to as test plateaus. The typical post-refueling power plateaus were used until the previously licensed full power condition (2300 MWt) was attained (approximately 87% of the EPU full power level of 2644 MWt). Above 2300 MWt, 3% intervals between test plateaus were established for data acquisition and evaluation. A summary of the power ascension test plan from LAR-205 [Reference 2] for power levels beginning at 2300 MWt is provided in Tables 2.12-1 and Table 2.12-2.

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 5 of 20 Table 2.12-1 PTN Extended Power Uprate Power Ascension Test Plan Test Test Prior Rated Thermal Power - % of 2644 MWt (Allowance +0%, -5%)

(Allowance +0%, -1%)

Modification Description To Startup 0

5 10 15 20 25 30 40 45 50 55 60 65 70 75 80 87 89 92 95 98 100 Nuclear Steam Supply Data System Collection X

X X

X X

X X

X X

(NSSS) Data Cleto Record Balance of Data Plant Data Collection X

X X

X X

X X

X X

X Record Transient Data Data Collection X

X X

X X

X X

X X

Recordl6 C

Power Distribution X09 X

XOl X

XO)

XO)

XO)

XO)

X Core Map and COLR Parameters NSSS Calorimetric Verify Thermal and Power Power and Adjust X

X X

X X

X X

X X

X Range Nuclear Channel Instrumentation Adjustment Reactor Coolant RCS Flow System (RCS)

Calorimetric X

Flow Measurement

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 6 of 20 Table 2.12-1 PTN Extended Power Uprate Power Ascension Test Plan Test Test Prior Rated Thermal Power - % of 2644 MWt (Allowance +0%, -5%)

(Allowance +0%, -1%)

Modification Description To Startup 0

5 10 15 20 25 30 40 45 50 55 60 65 70 75 80 87 89 92 95 98 100 Calibrate Excore Incore-Excore ntueaio Axial Offset Instrumentation X(2)

X(3)

X(4)

Calibrations to Incore Axial Offset 10% Ramp to X

Load Changes Verify System X

Response

OST Turbine Turbine Trip to Verify X

System

Response

Turbine Stop

Valve, Standard Governor turbine valve Valve, and tests w/post-X(6)

Intercept modification Valve tests Testing.

Steam Generator Manually inserted level Level setpoint step-Feedwater changes in the X

X X

Flow steam

Response

generator.

Testing Monitor Vibration Vibration in X

X X

X X

Monitoring Plant Piping and Supports

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 7 of 20 Table 2.12-1 PTN Extended Power Uprate Power Ascension Test Plan Test Test Prior Rated Thermal Power - % of 2644 MWt (Allowance +0%, -5%)

(Allowance +0%, -1%)

Modification Description To Startup 0

5 10 15 20 25 30 40 45 50 55 60 65 70 75 80 87 89 92 95 98 100 Monitor Thermal Thermal X

X X

X X

X X

Expansion Expansion in Plant Piping Plant Verify Radiation Expected X

X Surveys Dose Rates Plant Verify Temperature Expected X

X Surveys Temperatures NOTES:

(1)

If required.

(2)

Incore flux map for data acquisition will be performed at less than 50% of 2644 MWt or when annunciator B2/2 or B2/3 alarms, whichever comes first.

(3)

Incore flux map for data acquisition will be performed at approximately 87% of 2644 MWt, if required.

(4)

At steady state equilibrium Xenon conditions.

(5) Not in LAR-205, performed as part of normal startup procedures (6)

Test moved from 35% to 50% to ensure that ESFAS high steam flow and AMSAC arming setpoint are not adversely affected by steam flow imbalance.

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 8 of 20 Table 2.12-2 Large Plant Transient Tests in Turkey Point EPU Power Ascension Test Plan Proposed Test Description Expectation Turbine Overspeed Trip The turbine will be, with the This test will verify proper operation of from 5% EPU Power reactor at approximately 5%

the overspeed mechanism for the new power, automatically tripped EHC turbine control, and proper as speed exceeds the electronic operation of the new turbine control overspeed trip setpoint.

valves.

10% Ramp Load Change Ramp load change limited by These ramps will test NSSS and BOP at new 30% and 100%

station license conditions and control system operation, to ensure EPU Power fuel pre-conditioning that no unanticipated aggregate effects considerations.

have been produced by interaction of the plant modifications.

Turbine Stop Valve, Standard turbine valve testing Validate dynamic performance of new Governor Valve, and augmented by post-governor valve design to ensure Intercept Valve Testing at modification tests associated adequate transient response. Verify 35% EPU Power with EHC Turbine control acceptable dynamic performance of and Governor Valve the new HP turbine rotor during Replacement.

changes in individual arc steam flows.

Steam Generator Level /

Verify response to manually Verify SG level control system Feedwater Flow Dynamic inserted level setpoint step-response and acceptability of over-testing at 30%, 87% and change of 5% in the steam shoot, damping and steady-state limit 95% EPU Power generator. Both up-going and cycling at the new licensed power down-going setpoint changes level. Verify acceptable operation of of different magnitudes will the feedwater control system.

be inserted.

FPL provided specific acceptance criteria in response to NRC request for additional information [Reference 3] for the 10% ramp load change list in Table 2.12-2 above as follows:

Reactor coolant system (RCS) average temperature, pressurizer pressure, and pressurizer water level will be controlled to the programmed values.

Steam generator water level will demonstrate good feedwater level control and maintain acceptable margin to the trip level setpoint.

" Nuclear power peak overshoot/undershoot should be less than 3 percent reactor thermal power.

Steam generator water level should return to programmed level setpoint within +/-2 percent narrow range with dampening oscillations within 15 to 20 minutes.

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 9 of 20 Prior to exceeding the previous licensed core thermal power of 2300 MWt, the data gathered at the pre-determined power plateaus, as well as observations of the slow, but dynamic power increases between the power plateaus, allowed verification of the performance of the EPU modifications. The steady-state data collected at approximately 87% power was especially significant because this test plateau corresponded to the previous licensed core power level of 2300 MWt. Data collected at this plateau formed the basis for comparison of data collected at higher plateaus.

Once testing was completed at the 2300 MWt plateau, power was slowly and deliberately increased through four additional test plateaus, each differing by approximately 3% of the EPU rated thermal power. Both dynamic performance during the ascension and steady-state performance for each test plateau were monitored, documented and evaluated against pre-determined acceptance criteria and expected values. The acceptance criteria for the power ascension test plan were established as discussed in Regulatory Guide (RG) 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants and NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 14.2.1. Criteria were provided against which the success or failure of the test was judged.

In some cases, the criteria were qualitative where applicable. Quantitative criteria had appropriate tolerances.

Specific acceptance criteria and expected values were established and incorporated into the power ascension test procedures. Level 1 acceptance criteria are values for process parameters assigned in the design of the plant that are safety significant. If a Level I criterion is not satisfied, the power ascension will be stopped and the plant will be placed in a condition that is judged to be safe based upon prior testing. Resolution of the issue that resulted in exceeding the Level 1 criterion must be resolved by equipment changes or through engineering evaluation, as appropriate. Level 2 acceptance criteria are values that relate to plant functions or parameters that are not safety significant. If Level 2 criteria are not met, the Power Ascension Test Plan may continue. Investigation of the issue that resulted in exceeding the Level 2 criterion may continue in parallel with the power escalation. These investigations would be handled by existing plant processes and procedures.

Following each increase in power level, test data was evaluated against its performance acceptance criteria (Level 1 and 2 and expected values, i.e., prediction targets for power level). If the test data satisfied the acceptance criteria, then system and component performance were considered to have complied with their design requirements. Predicted values are used for optimizing SSC performance only and are not acceptance criteria.

In addition to the steady-state parameter data gathered and evaluated at each test condition, the dynamic parameter response data gathered during the ascension between test plateaus was also evaluated and demonstrated overall stability of the plant.

Hydraulic interactions between the new main feedwater pumps and the steam generator flow control valves, as well as the impact of the higher main feedwater flow, were monitored and evaluated. Individual control systems, such as steam generator level control and feedwater heater drain level control, were optimized for the new EPU conditions, as required. The power ascension testing adequately identified any unanticipated adverse system interactions and allowed them to be corrected in a timely fashion prior to operation at higher power levels.

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 10 of 20 Vibration Monitoring A piping and equipment vibration monitoring program, including plant walkdowns and monitoring of plant equipment pre-and post EPU, was established to ensure that any steady-state flow induced piping vibrations following EPU implementation were not detrimental to the plant, piping, pipe supports, or connected equipment.

The predominant way of assessing piping and equipment vibrations was to monitor the piping during the plant heat-up and power ascension. The methodology used for monitoring and evaluating vibration was in accordance with ASME OM-S/G-2007, Standards and Guides for Operation and Maintenance of Nuclear Power Plants, Part 3, Requirements for Preoperational and Initial Startup Vibration Testing of Nuclear Power Plant Piping Systems.

The scope of the piping and equipment vibration monitoring program included accessible piping that experienced an increase in process flow rates. Branch lines attached to this piping that experienced increased process flows were also monitored as operating experience has shown that branch lines are susceptible to vibration-induced damage. The scope of the program included the following systems:

  • Main steam (outside of containment) including Reheater Inlet,
  • Main Steam modified supports (inside containment),

Condensate,

  • Heater Drains,
  • Moisture Separator Drains, and
  • Turbine Gland Steam and Drains.

III. Results NSSS Data Collection The Turkey Point Unit 3 nuclear steam supply system (NSSS) significant parameters were observed at the 20%, 30%, 50%, 75%, 87%, 89%, 92%, 95%, 98%, and 100% EPU power plateaus. During power ascension to 100%, the NSSS significant parameter values at the various power plateaus met all Level 1 and Level 2 criteria with the exception of RCS Tref and steam flow.

RCS Tref is affected by the turbine inlet pressure (TIP). The Tref circuit was normalized to the actual TIP value at approximately 95% power for the expected value at 100% power. At 99.3% power the TIP value is only 5 psi below the predicted value which results in the actual Tref only 0.2 degree Tref below the predicted value. This minor deviation is well within the expected range and meets Level 1 and Level 2 criteria. The main steam flow and main feedwater flow safeguards signals experienced excessive flow oscillation above 99% power as previously reported in Reference 4. Single channel Hi steam flow and feed flow/steam flow mismatch alarms were received at approximately 99.7% power which exceeded test acceptance criteria. Unit 3 power was lowered to approximately 99.5% power to reduce these oscillations to normal levels while the cause of the alarms was determined and corrective

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 11 of 20 actions implemented. The alarms were found to be caused by instrument signal spiking as a result of flow induced turbulence in the feedwater and main steam lines. FPL has taken corrective actions to modify the steam flow and feedwater flow circuits which eliminated the alarms such that Unit 3 was able to reach 100% rated thermal power (RTP).

Below provides a brief summary of major control parameters observed during the power ascension test program. In addition, Table 1 provides a summary of the NSSS significant parameters at the 100% power plateau.

  • RCS temperatures - As can be seen from Table 1, the maximum measured average RCS temperature at the100% EPU power plateau is 579.5°F which is below the EPU full power limit of 581.5°F.
  • Pressurizer pressure - remained at 2235 psia +/- 5% within normal operating band throughout the power ascension and transient test.

" Pressurizer level - The pressurizer level program changes from 22.5% at RCS average temperature 547 OF to 56.9% at full load RCS average temperature of 580'F. Pressurizer level was maintained on program throughout the power escalation within normal control system tolerances.

  • Containment temperature - temperature ranged from 99.5°F to 1 14.7°F throughout the power ascension well below the 125 OF limit.

" Steam generator level - remained constant at 50% narrow range scale within normal control system tolerances throughout the power ascension.

Balance of Plant (BOP) Data Collection The Turkey Point Unit 3 balance of plant (BOP) significant parameters were observed at the 20%, 30%, 50%, 75%, 87%, 89%, 92%, 95%, 98% and 100% EPU power plateaus. As the majority of the EPU hardware changes were made to BOP equipment, extensive monitoring of the secondary side was performed during the EPU power ascension. Major systems and components monitored included:

  • High pressure turbine, low pressure turbine, main generator and exciter vibration, High pressure turbine, low pressure turbine, main generator and exciter bearing temperatures, High and low pressure turbine steam pressure and temperature,
  • Turbine digital controls,
  • Main generator gas temperatures,
  • Turbine cooling water system performance,
  • Condensate, main feedwater, and heater drain system pressure and temperature,
  • Condensate, main feedwater, and heater drain pump performance,
  • Heater drain valve performance,
  • Isolated phase bus cooling performance, and

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 12 of 20

  • Main generator electric output.

There were no BOP parameters that exceeded level 1 criteria at 100%. There were no BOP parameters that exceeded level 2 criteria other than those previously reported in Reference 5 and all were found acceptable upon further evaluation. The BOP data collected during the EPU power ascension testing is too extensive to include in this summary report but Table 1 provides a summary of major control parameters at 100% power. The completed test procedure and all BOP data collected at the 20%, 30%, 50%, 75%, 87%, 89%, 92%, 95%,

98% and 100% EPU power plateaus are available for review on-site, if required.

Vibration Monitoring The Turkey Point Unit 3 piping and equipment within the scope of the EPU vibration monitoring program were observed at several different plant operating conditions, namely the 30%, 50%, 87%, 89%, 92%, 95%, 98% and 100% EPU power plateaus. The first observations were conducted prior to the shutdown in which the EPU modifications were implemented. Data from these observations was used to develop the list of priorities and baseline data for observation during the EPU power escalation. By comparing the observed pipe vibrations / displacements at various power levels with previously established acceptance criteria, potentially adverse pipe vibrations were identified, evaluated and resolved.

Engineering has reviewed the vibration and thermal expansion data from each of the applicable plateaus through 100% and determined that all lines in the monitoring program have met their acceptance criteria.

A screening criteria was established where piping/tubing/supports with vibration levels with displacements of 1/16" to 1/8" required further engineering analysis. For the piping, tubing, and supports within the program that experienced vibration levels exceeding the screening criteria, nine items showed low margin but were still below the acceptance criteria at the 100% power level. Modifications were implemented to improve margin for these items.

Thermography Checks and Temperature Profiles Temperature monitoring of the Main and Auxiliary Transformers and the Isophase Bus duct using thermography was performed at the 30%, 50%, 75%, 87%, 89%, 92%, 95%, 98% and 100% power levels. The thermography checks were performed to ensure none of the equipment was overheating due to the EPU power increase. The test data was evaluated and found that all temperatures were below the alert range and within the expected values at 100%

power.

10% Load Ramp Two 10% load changes are required at a ramp rate of 1 %/min. The first test was a down power equivalent to a 10% turbine load change starting at 100% power +0% - 1%. The second load change was a load increase equivalent to a 10% turbine load change starting at approximately 90%.

As stated above in Section II, the purpose of these tests are to provide additional confidence in the validity of the analytical models and assumptions used in the analysis of

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 13 of 20 plant modifications and integrated plant response to transients, and also verify that no new thermal hydraulic phenomena or adverse system interactions are created by the EPU.

The dynamic behavior of the various plant control systems are observed and evaluated against Level 1 and Level 2 acceptance criteria to ensure that the combination of increased EPU power and changes to the plant configuration (EPU modifications) have not resulted in an unacceptable aggregate impact. Test acceptance criteria are:

RCS average temperature, pressurizer pressure, and pressurizer water level will be controlled to the programmed values.

Steam generator water level will demonstrate good feedwater level control and maintain acceptable margin to the trip level setpoint.

Steam generator water level should return to programmed level setpoint within +/-2 percent narrow range with dampening oscillations within 15 to 20 minutes.

  • Nuclear power peak overshoot/undershoot should be less than 3 percent reactor thermal power.

10% Down Power Due to performance issues with a feedwater pump, Unit 3 performed a rapid down power on 5/4/13 from approximately 98.3% indicated nuclear power to 85.4% over a 13 minute period reducing turbine load by 118 MWe. This rapid down power achieved a ramp rate greater than 1 %/min and reduced power greater than the prescribed 10% turbine load (13.6%). However, the initial power level was 0.7% less than 99% rated thermal power identified in Table 12.2-1 of the test program.

The initial condition for the ramp test is specified in table 2.12-1 above which covers both steady state plateau data gathering done at 3% intervals and the dynamic 10% ramp tests.

Since the tolerance covers both types of test and the steady state plateaus are only 3%

apart, a 1% tolerance is needed to ensure an adequate power increase that will result in observable changes in parameters.

Table 2.12-1 has "(down)" applied to the box for the 10% ramp test under the 100%

column. The intent of this note is to indicate that the 1% tolerance on initial power does not apply to the ramp up from 90% to minimize the potential for an overpower condition to occur due to turbine control system tolerance. As stated in References 1, 2 and 3 the intent of the test is to determine there are no adverse system interactions or new thermal hydraulic phenomena introduced to plant systems from EPU changes and to sufficiently test NSSS and BOP control systems.

Each of the system parameters impacted by the test is affected by a change in either Tref or Tavg. The turbine, control rod, pressurizer level, pressurizer pressure and steam generator

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 14 of 20 level control system will respond to the magnitude and rate of either Tavg or Tref change either through thermodynamic/hydraulic changes. The maximum Tavg change is a function of the turbine ramp rate, reactivity feedback from the core (power defect) and control rod system response to Tref. Given the same ramp rate and control system setting for a ramp test done at two different initial power conditions, only power defect would affect Tavg response. A review of the power defect curves for Unit 3 shows that between 100% power and 80% power, the power defect is linear with reactor power. Therefore, the amount of negative reactivity added to the core per degree change of Tavg will be the same anywhere between 100% power and 80% power. Therefore, starting the test at a slightly lower initial power will not change the magnitude or rate of RCS temperature change and the challenge to the control systems will be the same. Secondary system response is a function of feedwater flow which is linear with turbine load/steam flow. There is a slight increase in the differential pressure across the feedwater regulating valve as FW flow decreases which make it more challenging for steam generator level control. Therefore, starting the test at a slightly lower initial power level will not change the magnitude or rate of feedwater and steam flow changes. Since it has been concluded that the slightly lower initial power level will not reduce the challenge to control systems and thermal hydraulic responses, the power reduction performed on 5/4/13 is acceptable for demonstrating integrated plant response to transients and verifying that no new thermal hydraulic phenomena or adverse system interactions are created.

A review of system transient data for the 5/4/13 10% load reduction shows that:

  • RCS average temperature, pressurizer pressure, and pressurizer water level were controlled to their programmed values.

" Steam generator water level demonstrated good feedwater level control and maintained acceptable margin to the trip level setpoint.

" Steam generator water level returned to programmned level setpoint within +/-2 percent narrow range with dampening oscillations within 15 to 20 minutes.

" Nuclear power peak overshoot/undershoot was less than 3 percent reactor thermal power.

In addition, level 1 and 2 criteria for each of the above control systems were met since each controlled their respective process parameter within technical specification limits with margin. There were no new thermal hydraulic phenomena or adverse system interactions observed.

All acceptance criteria were met for the 10% load down power test.

10% Power Increase The 10% power increase was performed on 6/25/13 starting at 88.5% of RTP. Starting 1.5% below 90% power provided ample margin to avoid over power at the end of the

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 15 of 20 ramp. As stated above in the 10% down power discussion, starting slightly lower than 89%

power does not impact control system response.

Over a 10 minute period, the turbine load was increased by 78 MWe which is equal to a 9.2% load increase. The ramp test did achieve the full 84 MWe 10% load change over an additional 4 minute period. The turbine control system (TCS) response is designed to reduce the ramp rate as turbine load approaches its target. However, the TCS decreased the ramp rate at more than expected as the target was approached. LAR-205 Section 2.12.1.2.3.5 and Table 12.2-2 do not specify tolerances for the ramp rate or the turbine load. The following system responses were observed for the ramp increase test:

RCS temperature was increased with manual rod control since the automatic out rod motion feature has been removed at Turkey Point. Operators maintained Tavg within 1.5 degrees of Tref for the entire transient.

Pressurizer pressure reached a maximum of 2238.6 psig and was reduced by increased spray to a minimum value of 2217.6 psig and then returned to to the program setpoint of 2235 psig +/- 3 psi.

Pressurizer level followed RCS Tavg and maintained level within 0.6% of the programmed value throughout the transient demonstrating excellent control.

The load increase achieved over the 10 minute ramp was 78 MWe which equates to 9.2% of 100% turbine load. Average Nuclear power increased from 88.3% to 96.9 for a 8.6% change which is well with the 3% undershoot criteria.

  • All 3 steam generator level control systems demonstrated very good feedwater level control maintaining within 2% of the setpoint from the beginning of the transient.

This also demonstrated the ability to maintain adequate margin to trip.

The power increase ramp test met all the control system acceptance criteria identified above in the power decrease ramp test. In addition, level 1 and 2 criteria for each of the above control systems were met since each controlled their respective process parameter within technical specification limits with margin. There were no new thermal hydraulic phenomena or adverse system interactions observed.

However, the 10% load change and 1% ramp rate test conditions were not met. Given the observed control system response and the available margin to acceptance criteria and the control system response observed during the down power when a 1 %/min ramp rate and 10% load change were achieved, the test results adequately demonstrate that the test acceptance criteria would also be met for the slightly higher load increase and ramp rate.

Therefore, the test deviation for the 10% load and 1% ramp rate are acceptable.

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 16 of 20 Plant Radiation Surveys The purpose of the survey is to verify plant areas remain as low as reasonable achievable (ALARA) per 10 CFR 20 and doses remain within current limits. 10 CFR 20 criteria is met by confirming doses remain low in normally accessible areas, and dose rate postings within each area are maintained. Turkey Point also has design basis dose limits. Per UFSAR Chapter 11.2 and Chapter 8A normal operation doses are considered in shielding design and environmental qualification of electrical equipment. Normal operational doses limits are based on the radiation zones identified in UFSAR Table 11.2-1. In addition, the secondary shield wall inside containment is designed to attenuate normal operational doses such that the dose rate at the outside of the containment wall is less than 1 mrem/hr.

Radiation surveys were performed at the 87% power plateau for a baseline at the pre-EPU 100% equivalent power level and at the EPU 100% power level in accessible areas that were expected to have increased dose rates. The expected increase in dose rates should be approximately 15% for those areas that are subject to nitrogen-16 gamma or neutron flux.

Inside Containment The dose rates measured just inside the secondary shield wall (RCS piping area) and outside the secondary shield walls showed measured dose rate increases in some locations that are larger than expected. A review of historic survey maps for the same areas from 2005 prior to EPU show readings higher than at EPU 100% power. The surveys in the primary piping loop area are performed by mounting a detector on the end of a pole and inserting the pole into the loop area while the RP technician stands behind the secondary shield wall. Using this type of measurement cannot reproduce the same survey dose point with any precision. A review of the dose rates inside the secondary shield wall shows significant differences based on minor changes in physical locations. Measurements outside the secondary shield wall have similar challenges due to the streaming affect from shield wall penetrations. A small change from target survey locations will change dose significantly and the dose rates at 100% power do not allow for a precision map to be made.

Containment is classified as a locked high radiation area. The measured dose rates did not increase the level of the posting beyond a locked high radiation area. The controls imposed by the Technical Specifications and 10 CFR 20 for locked high radiation areas will maintain personnel dose ALARA.

Inside the secondary shield wall (RCS loop area) the Environmental Qualification (EQ) program assumed a dose rate of 60 R/hr and the measured value at the entrance to the RCS piping area was 1.8 R/hr. The EQ program assumed a dose rate of 1.125 R/hr outside the secondary shield wall and the maximum measured dose was 0.6 R/hr.

The dose rates measured adjacent to the containment wall did not change from measured values prior to EPU and remain less than 1 mremnhr.

Outside Containment

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 17 of 20 Measurements outside containment did not show any appreciable increase in dose except for the Charging pump room and pipe and valve room. The charging pump room radiation measurements showed that half of the locations had decreased radiation levels and half increased in radiation level. The largest increase went from 0.5 mrem to 8 mrem in the non Regenerative Heat Exchanger room. This was the highest dose for the locations that increased. In the pipe and valve room the largest increase went from 2.0 mrem to 17 mrem near the Residual Heat Removal penetration.

While these areas exceeded the expected 15% increase, the actual dose rates are very low and do not affect the postings or radiation zone classifications for any area. The Auxiliary Building areas surveyed are controlled radiation areas. The measured dose rates do not increase the level of the posting beyond a radiation area. The controls imposed by the technical specification and 10 CFR 20 for radiation areas will maintain personnel dose ALARA. The EQ program assumed dose rates at the maximum allowed for the radiation zone classification. All surveyed areas remained below the radiation zone classifications.

The measured dose rates inside and outside containment at 100% EPU power meet the 10 CFR 20 and Turkey Point limits.

Plant Temperature Surveys Plant ambient temperature survey data was collected with data recorders in those areas potentially impacted by EPU equipment heat loads. Most areas did not see any appreciable temperature change. All area temperatures remain below the general outside area design limit of 104 'F except the condensate pump area and main feedwater pump room.

As part of the EPU design changes for the Condensate pumps general area supply and exhaust fans were added for the condensate pump area. However, temperatures were observed between 120 'F and 130 'F even with the addition of the fans. The condensate pump and motor design temperature limits are above 140 'F except for the motor lower radial bearing. The lower radial bearing does not have an oil cooler and relies on ambient cooling. The bearing has experienced alarms which actuate at a bearing temperature of 185 'F. A review of the air flow in the condensate pump area shows some portions may not be receiving adequate flow. Temporary fans have been installed to augment air flow to the low flow areas and it is maintaining bearing temperatures below their alarm point until a permanent solution can be implemented. Procedural guidance directs shutting down the pump if bearing temperatures reaches 200 'F. The high area temperature condition has been entered into the Turkey Point Corrective Action Program for resolution.

The Steam Generator Feed Pump (SGFP) Room temperature reached 11 0IF during some periods. While the EPU did increase feedwater flow the heat load to the feedwater pump room did not increase. This higher room temperature condition is a pre-EPU issue which has been compensated for by vane axial fans which provide forced circulation through the SGFP Motors. An exhaust plenum is located above each SGFP motor and is attached to 25,000 cfm exhaust fans. The additional fans maintain the feedwater pump motors below

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 18 of 20 their design maximum operating temperatures. All other equipment within the feedwater pump room is rated for temperature well above 11 0IF. Based on past plant operation, the elevated temperatures have not caused a reduction in feedwater pump reliability.

Leading Edge Flowmeter (LEFM) Commissioning As described in Reference 2, the Turkey Point EPU project included a 1.7% Measurement Uncertainty Recapture (MUR) thermal power increase. To achieve the MUR power increase of 1.7%, the Cameron Leading Edge Flow Meter (LEFM) CheckPlusTM ultrasonic flow measurement instrumentation was installed to improve feedwater flow measurement accuracy. An individual LEFM CheckPlusTM system flow element (spool piece) was installed in each of the three main feedwater lines and was calibrated in a site-specific model test at Alden Research Laboratories with traceability to National Standards.

The LEFM CheckPlusTM system was installed and commissioned in accordance with FPL procedures and Cameron installation and test requirements. LEFM CheckPlusTM commissioning included verification of ultrasonic signal quality and evaluated the actual plant hydraulic velocity profiles as compared to those documented during the Alden Research Laboratories testing. Final verification of the site-specific uncertainty analyses occurred as part of the LEFM CheckPlusTM system commissioning process. The commissioning process provides final positive confirmation that actual performance in the field meets the uncertainty bounds established for the instrumentation which satisfies licensing comnmitment 13 in Section 3.3 (a) of the EPU Safety Evaluation Report. In addition to the Cameron commissioning test and evaluations, FPL evaluated LEFM performance as follows:

A review of feedwater parameters for LEFM and the Venturi based measurements were completed to determine deviations and reasonableness of the Venturi correction factor.

There are no operational alarms or other deficiencies noted. Steam Generator Heat rates and calorimetric calculations were performed using the LEFM and Venturi based data. The LEFM ultrasonic flow transmitters are performing as expected. After repair of the "B" loop Venturi flow transmitter used for calorimetric calculation, the LEFM is showing a 1.3%

greater thermal power than the Venturi based value at 100%. The Venturi LEFM correction factor adjusts the Venturi flowrate down such that the 1.3% deviation is eliminated and Venturi flow matches LEFM flow.

IV. Summary The test data collected during EPU startup and power ascension and summarized in this report demonstrates that all major systems, structures, and components performed as predicted and there was no adverse impact to the performance of the unit. The EPU startup and power ascension test data satisfied all acceptance criteria. Copies of the completed EPU startup and power ascension test procedures are available on site for review.

Turkey Point Unit 3 L-2013-226 Docket No. 50-250 Attachment Page 19 of 20 V.

References

1.

J. Page (NRC) to M. Nazar (FPL), Turkey Point Units 3 and 4-Issuance of Amendments Regarding Extended Power Uprate (TAC Nos. ME4907 and ME 4908), Accession No. ML11293A365, June 15, 2012.

2.

FPL Letter, M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (NRC)

(L-2010-113), License Amendment Request No. 205: Extended Power Uprate (EPU), (TAC Nos.ME4907 and ME4908), Accession No. ML103560169, October 21, 2010.

3.

M. Kiley (FPL) to U. S. Nuclear Regulatory Commission (NRC) (L-201 1-101),

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Balance of Plant Issue, Accession No. ML11105A146, April 14, 2011.

4.

M. Kiley (FPL) to U. S. Nuclear Regulatory Commission (NRC) (L-2013-073),

Renewed Facility Operating License No. DPR-31 Extended Power Uprate Cycle 26 Startup Report-Supplement 1, Accession No. ML13071A345, February 26, 2013.

5.

M. Kiley (FPL) to U. S. Nuclear Regulatory Commission (NRC) (L-2012-426),

Renewed Facility Operating License No. DPR-31 Extended Power Uprate Cycle 26 Startup Report, Accession No. ML12341A097, December 4, 2012.

Turkey Point Unit 3 Docket No. 50-250 L-2013-226 Attachment Page 20 of 20 Table 1 BOP and Primary Parameter Summary Primary Major Parameter 100%t Avg Power %

99.33 Avg Thot 'F 613.8 AvgTcold 'F 546.2 Avg Tavg 'F 580.0 SG A Press psig 784 SG B Press psig 786 SG C Press psig 784 Pzr Avg % Level 56.2 Turbine Inlet Avg Press (TIP) psig 652.2 Tref 579.5 Containment Temp TF Highest 100.5 BOP Major Parameter SG A Level %

48.6 SG B Level %

51.0 SG C Level %

48.4 Avg Steam flow MPPH 3.78 Avg FW flow MPPH 3.91 Final Venturi FW Temp TF 438.8 Condenser Backpressure in-Hg 1.5 Generator Output Mwe 860.4 Thermal Output MWt 2624.9 (1) All feedwater flow and thermal output values are based on the Venturi flow meter except at 98% and 100% power which are based on the LEFM flow meter

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