NOC-AE-03001470, Unit 2 Cycle 10 Startup Testing Summary Report
| ML030700584 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/05/2003 |
| From: | Jordan T South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NOC-AE-03001470, STI:31560478 | |
| Download: ML030700584 (32) | |
Text
Nuclear Operating Company Soth A=s ft* ES &nP.289Mbds.,4A TPras 77483 * ,
March 5, 2003 NOC-AE-03001470 File No.: G02.06 10CFR50.36 STE: 31560478 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Washington, DC 20852 South Texas Project Unit 2 Docket No. STN 50-499 Unit 2 Cycle 10 Startup Testing Summary Report South Texas Project Technical Specification 6.9.1.1 requires a summary report of appropriate plant startup and power escalation testing results following: a) amendment to the license involving a planned increase in power level, b) the installation of fuel that has a different design, and c) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. During the recent Cycle 9 to Cycle 10 refueling outage, South Texas Project Unit 2 installed 68 feed fuel assemblies, each with reduced-enrichment annular axial blanket pellets in the top and bottom seven inches of the fuel stack. In addition, all four Model E Steam Generators were replaced with Model Delta 94 Steam Generators, and the full power Reactor Coolant System average temperature was raised from 590°F to 592°F. The installation of the Model Delta 94 Steam Generators allowed South Texas Project Unit 2 to implement a power uprate of 1.4% from 3800 Megawatts Thermal to 3853 Megawatts Thermal in accordance with Amendment 127 to the Operating License.
Attachment A to this letter is a summary report of the startup physics test results obtained during startup and power ascension. Attachment B is a summary report of the Power Uprate.
Attachment C to this letter is a summary report of the specific tests performed for the Replacement Steam Generators. No corrective actions were required to obtain satisfactory operation.
One test of the Replacement Steam Generators has not been performed at the time of the preparation of this report due to plant conditions unrelated to the Replacement of the Steam Generators. A supplement to this report will be submitted as required by Technical Specification 6.9.1.1
NOC-AE-03001470 Page 2 There are no new licensing commitments contained in this letter. If there are any questions, please contact Mr. M. E. Kanavos at (361) 972-7181 or me at (361) 972-7902.
Vice President, Engineering & Technical Services
/KJT Attachment A: South Texas Project Unit 2 Cycle 10 Startup Physics Testing Summary Report Attachment B: South Texas Project Unit 2 Cycle 10 Power Uprate Summary Testing Report Attachment C: South Texas Project Unit 2 Cycle 10 Steam Generator Replacement Return-To-Service Testing Summary Report
NOC-AE-03001470 Page 3 cc:
(paper copy) (electronic copy)
Ellis W. Merschoff A. H. Gutterman, Esquire Regional Administrator, Region IV Morgan, Lewis & Bockius LLP U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 L. D. Blaylock/W. C. Gunst Arlington, Texas 76011-8064 City Public Service U. S. Nuclear Regulatory Commission Mohan C. Thadani Attention: Document Control Desk U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike R. L. Balcom Rockville, MD 20852 Texas Genco, LP Richard A. Ratliff A. Ramirez Bureau of Radiation Control City of Austin Texas Department of Health 1100 West 49th Street C. A. Johnson Austin, TX 78756-3189 AEP Texas Central Company Cornelius F. O'Keefe Jon C. Wood U. S. Nuclear Regulatory Commission Matthews & Branscomb P. 0. Box 289, Mail Code: MN116 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704
NOC-AE-03001470 Attachment A Page 1 of 5 ATTACHMENT A SOUTH TEXAS PROJECT UNIT 2 CYCLE 10 STARTUP PHYSICS TESTING
SUMMARY
REPORT
NOC-AE-03001470 Attachment A Page 2 of 5 I. Hot Rod Drop Time (seconds):
Acceptance Criteria (AC): _*2.8 seconds
- Maximum value for 57 control rods II. Rod Worth Measurements (Dynamic Rod Worth Measurement Method Used):
Design Review Criteria (DRC): Each bank within 15% or 100 pcm of the predicted value (whichever is greater)
Total rod worth within 8% of predicted Acceptance Criteria (AC): Total rod worth > 90% of Predicted RCCA Bank Measured Predicted Delta Percent Pass/Fail Pass/Fail Worth Worth (M-P) Difference DRC AC (pcm) (pcm) (pcm) (%)
Shutdown A 241.2 243.9 -2.7 -1.1 P Shutdown B 690.0 715.7 -25.7 -3.6 P Shutdown C 381.7 377.3 4.4 1.2 P Shutdown D 378.1 371.2 6.9 1.9 P Shutdown E 479.7 472.2 7.5 1.6 P Control A 903.9 890.5 13.4 1.5 P Control B 599.8 586.2 13.6 2.3 P Control C 797.7 792.2 5.5 0.7 P Control D 494.1 479.3 14.8 3.1 P Total 4966.2 4928.5 37.7 0.8 P P ARO: All Rods Out
% Difference = 100 x (M - P) / P
NOC-AE-03001470 Attachment A Page 3 of 5 HI. Hot Zero Power (HZP) Critical Boron Concentration (ppm):
Design Review Criteria (DRC): +/-50 ppm Acceptance Criteria (AC): +/-1000 pcm (152.7 ppm)
Measured Predicted (M-P) Pass/Fail Pass/Fail (M) (P) DRC AC 2001 2032 -31 P P IV. HZP, ARO Isothermal Temperature Coefficient (ITC) (pcm/*F):
Design Review Criteria (DRC): +/-2 pcm/0 F Acceptance Criteria (AC): none Measured Predicted (M-P) Pass/Fail Pass/Fail (M) (P) DRC AC
-2.43 -3.12 0.69 P V. Inferred HZP, ARO Moderator Temperature Coefficient (pcmfOF)*:
Design Review Criteria (DRC): none Acceptance Criteria (AC): < +5 pcmPF, or rod withdrawal limits established Measured Predicted Adjusted Pass/Fail Pass/Fail DRC AC
-0.63* -1.32 -0.01 ** - P Measured MTC is actually an inferred MTC obtained by subtracting the design Doppler Temperature Coefficient (-1.8 pcm/0 F) from the measured Isothermal Temperature Coefficient.
- Adjusted MTC includes measurement uncertainty and Integral Fuel Burnable Absorber burnout correction.
NOC-AE-03001470 Attachment A Page 4 of 5 VI. POWER DISTRIBUTION MEASUREMENTS:
Design Review Criteria (DRC): Incore Quadrant Power Tilt < 1.02 Assembly Power Error (M-P) < +/-0.1 Acceptance Criteria (AC): FDHN < Technical Specification (TS) 3.2.3 Limit Fxy < TS 3.2.2 Limit Largest Reactor Power Incore Quadrant Limiting FDHN Limiting F.yLimit Assembly FDHN Limit Fxy Power Power Tilts Error Low Power 1.006 0.995 1.4599 1.8886 1.6700 2.1778 0.053 (29.0%) 1.006 0.992 Intermediate Power 1.005 0.998 1.4070 1.6644 1.6372 1.9947 0.052 (77.0%) 1.003 0.994 Full Power 1.006 1.002 1.4129 1.557 1.6048 1.9070 0.051 (100.0%) 1.000 0.993 FDHN: Nuclear Enthalpy Rise Hot Channel Factor Incore Tilt: Measured Incore Tilt in Excess of Designed Core Asymmetry
NOC-AE-03001470 Attachment A Page 5 of 5 VII. Reactor Coolant System Flow Measurement (gpm):
Design Review Criteria (DRC): none Acceptance Criteria (AC): > 403,000 gpm VIII. Full Power Critical Boron (ppm):
Design Review Criteria (DRC): +/-50 ppm Acceptance Criteria (AC): +/-1000 pcm (159 ppm)
Bumup Measured Predicted (M-P) Pass/Fail Pass/Fail (EFPD) (M) (p) DRC AC 5.34 1363 1385.8 -22.8 P P
NOC-AE-03001470 Attachment B Page 1 of 2 ATTACHMENT B SOUTH TEXAS PROJECT UNIT 2 CYCLE 10 POWER UPRATE
SUMMARY
TESTING REPORT
NOC-AE-03001470 Attachment B Page 2 of 2 Introduction South Texas Project implemented Amendment 127 to the Unit 2 Operating License for a power uprate on May 13, 2002. Because the amendment required Model D94 Steam Generators to be installed, reactor power was actually raised to the new maximum power level of 3853 Megawatts thermal on December 10, 2002, following the Cycle 10 Refueling Outage in which the Steam Generators were replaced.
Only certain portions of the following tests identified in Section 14.2.12.3 of the Updated Final Safety Analysis Report were applicable to the power uprate. These affected tests were:
- 14. Radiation Survey Test
- 15. Nuclear Instrumentation Calibration Test
- 22. Evaluations of Core Performance Test Radiation Survey Test The objective of this test was to verify shielding effectiveness by measuring neutron and gamma radiation dose levels at selected points throughout the plant. For the power uprate, neutron and gamma surveys were performed at selected locations in the Reactor Containment Building after power was raised to 3853 megawatts in accordance with normal radiation survey procedures. The surveys indicated no change in dose rates from surveys performed before the power uprate.
Nuclear Instrumentation Calibration Test The objective of this test was to calibrate the power-range channels to reflect actual power levels. During the refueling outage, all software and procedure changes were incorporated for the uprated thermal power level of 3853 megawatts. Normal plant procedures were used to calibrate power range channels to reflect actual power levels (% of 3853 megawatts) at specified power levels during power ascension and after reaching 3853 megawatts.
Evaluations of Core Performance Test The objective of this test was to verify that core performance margins are within design predictions for expected normal and abnormal rod configurations. During the refueling outage and power ascension, adjustments were made to delta-temperature amplifier gains to ensure delta-temperature power reflected actual power (% of 3853 megawatts). Delta-temperature amplifier gains were evaluated after raising power to 3853 megawatts and no further adjustments were required. Incore flux maps were obtained during the power ascension to ensure there was sufficient margin to raise power to 100%. An Incore flux map was obtained after the power uprate to 100% to verify margin. Since Incore flux maps are performed as part of Startup Physics Testing, test results are tabulated in Attachment A of this letter and not repeated here.
NOC-AE-03001470 Attachment C Page I of 22 ATTACHMENT C SOUTH TEXAS PROJECT UNIT 2 CYCLE 10 STEAM GENERATOR REPLACEMENT RETURN-TO-SERVICE TESTING
SUMMARY
REPORT I
NOC-AE-03001470 Attachment C Page 2 of 22 I. Thermal Expansion Test The Objective of this test was to verify by visual observation, measurement, and evaluation that specified Replacement Steam Generator components and connected piping are free to expand without restriction of movement.
The Acceptance Criteria were that the equipment, piping and components addressed in the procedure are verified to expand during heat-up without obstructions or restrictions. All piping and components shall not cause interferences with surrounding equipment, supports, restraints, or structures. Thermal movements for each support, restraint, and/or component shall be within the anticipated ranges or evaluated as acceptable.
Observations, measurements, and evaluation of specified Replacement Steam Generator components and connected piping were made at ambient conditions prior to heatup of the Reactor Coolant System, at a Reactor Coolant System temperature of approximately 180 TF on November 29, 2002 and at a Reactor Coolant System temperature of approximately 567 TF on December 1, 2002. All Acceptance Criteria were met.
II. Vibration Monitorine Test The Objective of this test was to demonstrate that vibration of specified Replacement Steam Generator components and connected piping are within acceptable limits at operating conditions.
The Acceptance Criteria were that equipment, piping and components addressed in the procedure have vibration levels within limits specified in applicable codes.
Observation and evaluation of vibration of Steam Generator Blowdown System piping was performed on December 1, 2000 while operating each Steam Generator Blowdown subsystem at its normal flowrate. All Acceptance Criteria were met.
Measurement and evaluation of vibration of each Steam Generator's Feedwater piping was performed on December 12, 2002. All Acceptance Criteria were met.
III. Steam Generator Blowdown Recirculation Test (OTEP04-SG-0007)
The Objective of this test was to demonstrate that the Steam Generator Blowdown Recirculation system operates as designed following the changes in piping made due to Steam Generator Replacement.
The Acceptance Criteria was that the Steam Generator Blowdown Recirculation system operate as designed.
Data was collected during operation of each Steam Generator's Blowdown Recirculation system on November 24, 2002 and evaluated to verify that the system can be operated as designed. All Acceptance Criteria were met.
NOC-AE-03001470 Attachment C Page 3 of 22 IV. Reactor Coolant System Flow Verification (OTEP04-SG-0001)
The Objective of this test was to measure the Reactor Coolant System flow rate prior to criticality using data obtained from installed elbow tap differential pressure (AP) instrumentation.
Acceptance Criteria was that Reactor Coolant System flow rate is greater than the minimum required.
Reactor Coolant System flow rate was determined to be 454,405 gallons per minute on December 1, 2002. This was greater than the Thermal Design flow rate of 392,000 gallons per minute in FSAR Table 5.1-1. In addition, this flow rate was greater than the Reactor Coolant System flow determined using the same method during Cycle 1, which was expected. All Acceptance Criteria were met.
V. Low Power Steam Generator Water Level Control Test (OTEP04-SG-0003)
The Objective of this test was to demonstrate the ability of the low power steam generator level control system to control at steady state power and to demonstrate the ability of the low power steam generator level control system to respond to a mismatch between steam generator level and setpoint.
The Acceptance Criteria was that the actual steam generator levels remain within specified limits of the programmed values, and that steam generator levels automatically returned to and remained within design limits of the level setpoint following a level setpoint change.
This test was performed on December 5, 2002 at a reactor power level of approximately 12%. Data was collected and evaluated during steady state operation. For each steam generator, a -5% level setpoint change was initiated and response of the level control system was monitored. This was followed by a
+5% level setpoint change and response of the level control system was monitored. Figure 1 shows a typical response of Steam Generator level and Low Power Feedwater Regulating Valve position demand. All Acceptance Criteria were met.
VI. Calibration of Steam Flow Transmitters (OTEP04-SG-0001)
The Objective of this test was to verify the calibration of steam flow transmitters.
The Acceptance Criteria was that the difference between transmitter steam flow and actual steam flow is within the specified limits.
Data was collected and used to verify proper scaling of steam generator steam flow instrumentation at a reactor power level of approximately 77% on December 7, 2002, and at 100% power on December 10, 2002. One transmitter required calibration at 77% power. At 100% power, two transmitters required calibration and three other transmitters were calibrated to more closely normalize steam flow with feed flow. All calibrations were completed on December 11, 2002.
NOC-AE-03001470 Attachment C Page 4 of 22 VU. Steam Generator Water Level Control Test (OTEP04-SG-0004)
The Objective of this test was to demonstrate proper operation of the turbine-driven feedwater pumps and the pump's speed controllers at steady state power, to demonstrate the ability of the steam generator level control system to control at steady state power and to demonstrate the ability of the steam generator level control system to respond to a mismatch between steam generator level and setpoint.
The Acceptance Criteria were that actual steam generator levels and feedwater to steam header delta pressure are within specified limits of the programmed values, main feedwater regulating valve positions are between the maximum and minimum valve position curves specified for the test, and steam generator level automatically returns to and remains within design limits of the level setpoint following a level setpoint change.
This test was initially performed on December 8 and 9, 2002 at a reactor power level of approximately 77%. Data was collected and evaluated during steady state operation. For each steam generator, a -5%
level setpoint change was initiated and response of the level control system was monitored. This was followed by a +5% level setpoint change and response of the level control system was monitored.
Figure 2 shows a typical response of Steam Generator level and Main Feedwater Regulating Valve position demand. Figure 3 shows a typical response of Steam Generator Feedwater and Steam Flow.
All Acceptance Criteria were met.
Adjustments to control settings were made to the Main Feedwater Regulating Valve for Steam Generator A to improve performance and the valve was retested until satisfactory response was achieved. The final control settings for the Steam Generator A Main Feedwater Regulating Valve were then set in the controllers for the other three Main Feedwater Regulating Valves and all valves were tested satisfactorily.
The steady state operation portion of this test was performed again on December 12, 2002 at a reactor power level of 100%. Data was collected and evaluated during steady state operation. All Acceptance Criteria were met.
NOC-AE-03001470 Attachment C Page 5 of 22 VIII. Load Swing Test (OTEP04-SG-0005)
The Objective of this test was to demonstrate the ability of the plant to sustain an approximate 10%
power load reduction.
The Acceptance Criteria was that response of plant systems to the step load change is as follows:
No reactor trip.
No safety injection initiation.
No steam line safety or relief valve operation.
No pressurizer safety valve operation and no pressurizer relief valve operation.
Nuclear power undershoot is less than 3 percent for load decrease.
No manual intervention required to stabilize plant systems.
Plant variables (i.e., Tavg, pressure, feed flow, steam flow, etc.) do not incur sustained or diverging oscillations.
On December 12, 2002, a turbine step load decrease of approximately 10 percent power was initiated at 200 percent per minute from a reactor power level of approximately 95%. Control setting adjustments were made to the Main Feedwater Regulating Valves to Steam Generators A and C to enhance response.
Figures 4 through 9 show the response of plant parameters to the step load decrease. Plant variables were stable 18 minutes after initiation of the step load decrease. All Acceptance Criteria were met.
NOC-AE-03001470 Attachment C Page 6 of 22 IX. Larme Load Reduction Test (OTEP04-SG-0006)
The Objective of this test was to demonstrate the ability of the plant to sustain an approximate 25%
power load reduction.
The Acceptance Criteria was that response of plant systems to the step load change is as follows:
No reactor trip.
No safety injection initiation.
No steam line safety or relief valve operation.
No pressurizer safety valve operation.
Nuclear power undershoot is less than 3 percent for load decrease.
No manual intervention required to stabilize plant systems.
Plant variables (i.e., Tavg, pressure, feed flow, steam flow, etc.) do not incur sustained or diverging oscillations.
On December 13, 2002, a turbine step load decrease of approximately 25 percent power was initiated at 200 percent per minute from a reactor power level of approximately 95%. Figures 10 through 15 show the response of plant parameters to the step load decrease. As allowed by the test procedure, the Reactor Coolant System was borated to maintain control rods above the control rod insertion limit. Plant variables were stable 23 minutes after initiation of the step load decrease. All Acceptance Criteria were met.
X. Steam Generator Thermal Performance Test (OPEP07-SG-0003)
The Objective of this test is to verify the performance of the Replacement Steam Generators at or near full power.
The Acceptance Criteria is that measured parameters meet or exceed the values specified in the test procedure.
This test has not yet been performed due to plant conditions unrelated to Replacement of the Steam Generators. A supplement report will be made as required.
NOC-AE-03001470 Attachment C Page 7 of 22 List of Figures Figure 1 LPFRV B Level Swing Test SG B Level and LPFRV B Position Demand Figure 2 MFRV C Level Swing Test SG C Level and MFRV C Position Demand Figure 3 MFRV C Level Swing Test SG C Steam Flow and SG C Feed Flow Figure 4 10% Step Load Reduction SG B Level and MFRV B Position Demand Figure 5 10% Step Load Reduction SG B Steam Flow and SG B Feed Flow Figure 6 10% Step Load Reduction Loop C Delta-T Figure 7 10% Step Load Reduction Tref and Auct. High Tavg Figure 8 10% Step Load Reduction Pressurizer Pressure PT-457 Figure 9 10% Step Load Reduction Pressurizer Level LT-465 Figure 10 25% Step Load Reduction SG C Level and MFRV C Position Demand Figure 11 25% Step Load Reduction SG C Steam Flow and SG C Feed Flow Figure 12 25% Step Load Reduction Loop C Delta-T Figure 13 25% Step Load Reduction Tref and Auct. High Tavg Figure 14 25% Step Load Reduction Pressurizer Pressure PT-457 Figure 15 25% Step Load Reduction Pressurizer Level LT-465 Abbreviations:
LPFRV Low Power Feedwater Regulating Valve SG Steam Generator MFRV Feedwater Regulating Valve Tref Reference Average Coolant Temperature Auct. High Tavg Auctioneered High Average Coolant Temperature
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NOC-AE-03001470 Attachment C Page 16 of 22 FIGURE 9 10% Step Load Reduction 58 57 56 55 54 530
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NOC-AE-03001470 Attachment C Page 17 of 22 FIGURE 10 25% Step Load Reduction 80 78 76 74 72 ig 70 68 66 64 62 60 Time I MFRV C Position Demand - SGC Level I
NOC-AE-03001470 Attachment C Page 18 of 22 FIGURE 11 25% Step Load Reduction 4.5 4.3 4.1 3.9 3.7 3.5 S3.3 E
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NOC-AE-03001470 Attachment C Page 19 of 22 FIGURE 12 25% Step Load Reduction 100 95 90 85 80 ag 75 70 65 60 Time
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NOC-AE-03001470 Attachment C Page 20 of 22 FIGURE 13 25% Step Load Reduction 594 592 590
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