ML20195G637
| ML20195G637 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/31/1988 |
| From: | Hunsader S COMMONWEALTH EDISON CO. |
| To: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| 4718K, NUDOCS 8806280079 | |
| Download: ML20195G637 (88) | |
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COMMONWEALTH EDISON COMPANY BRAIDWOOD UNIT 1 CYCLE 1_.
STARTUP REPORT SUPPLEMENT MAY 1988 4
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0906200079 OB053f PDR ADOCK 05000456 P.
PDF 2149m(052388)
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TABLE-OF CONTENTS Section Title Page List of Tables 2
List of Figures 3
1.0 Introduction 4
2.0 Discussion of Braidwood Startup Program 5
3.0 Discussion of Braidwood Startup Tests 18 3.1 System Testing at Various Power Levels 20 3.2 Physics Testing 46 3.3 Transient Testing 67 3.4 Instrumentation and Calibration Testing 75 4.0 References 86 i
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m LIST OF TABLES TABLE TITLE PAGE 2.0-1 Braidwood Unit 1 Major Milestone 6
2.0-2 Braidwood Operational Modes 7
2.1 HZP ARO Flux Map Results 8
2.2 30% Power Flux Map Results 10 2.3 50% Power Flux Map Results 12 2.4 75% Power Flux Map Results 14 I
2.5 90% Power Flux Map Results 16 3.0 List of Test Summaries 18 i
3.1.2 Steam Generator Level Control Summary 21 3.1.5-1 Primary Makeup Water Chemistry Summary 28 3.1.5-2 Reactor Coolant Water Chemistry Summary 29 3.1.5-3 Steam Generator Water Chemistry Summary 30 3.1.5-4 Main Feedwater Chemistry Summary 31 3.2.3 Measured vs. Predicted ITC and MTC 48 3.2.4 Measured vs. Predicted Power Coefficient Verification Factors 50 3.2.7 Measured vs. Predicted Bank Worths 54 3.2.8 Measured vs. Predicted Boron Endpoint Concentrations and Boron Reactivity Worths 64 3.3.2-1 10% Load Decrease at 35% Power Summary 70 3.3.2-2 10% Load Increase at 35% Power Summary 71 3.3.2-3 10% Load Decrease at 75% Power Summary 72 3.3.2-4 10% Toad Increase at 75% Power Summary 73 3.4.2 Thermal Power Measurement and Statepoint Data Collection Summary 76 2
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LIST OF FIGURES FIGURE TITLE PAGE 3.2.7-1 Control Bank D Integral Worth vs. Steps Withdrawn 57 3.2.7-2 Control Bank C Integral Worth vs. Steps Withdrawn 58 3.2.7-3 Control Bank B Integral Worth vs. Steps Withdrawn 60 3.2.7-4 Control Bank A Integral Worth vs. Steps Withdrawn 61 3.2.7-5 Integral Worth of Control Banks in Overlap 62 3.2.7-6 Differential Worth of Control Banks in Overlap 63 3.2.7-7 Differential Boron Worth over the Control Banks 64 P
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1.0 - INTRODUCTION; 1
This supplemental report describes the required testing at Braidwood Station Unit 1 from the time of initial criticality.
through the 90%' power testing plateau.. EIt satisfies the requirement.of the Braidwood Unit 1 Technical Specifications that a Startup Report be submitted to the NRC every 90 days until completion-of the Startup Testing Program.
Braidwood Station, located in northeastern Illinois, utilizes a-four loop Westinghouse pressurized water reactor system.
' Westinghouse'_ Electric Corporation, Sargent & Lundy, and the Commonwealth Edison Company jointly participated in the design and construction of Braidwood Station. The plant is cyerated by Commonwealth Edison. Company with Sargent & Lundy as the Architect - Engineer.
The Nuclear Steam Supply System is designed for a power output of 3411 MWt..The equivalent warranted gross and approximate net electrical output are 1175 MWe and 1120 MWe, respectively.
Cooling for the plant is provided by a large man-made cooling pond of approximately 2500 acres constructed over a previously strip-mined area.
Essential service cooling is provided by a 99-acre auxiliary cooling pond which is integral with the main pond.
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F 2.0 - DISCUSSION OF BRAIDWOOD STARTUP PROGRAM The Braidwood Unit 1 startup testing program consists of single and multi-system tests that occur commencing with initial fuel loading and continuing through full power. These tests demonstrate overall plant performance and include such activities as precritical testing, low power tests, and power ascension tests. Testing sequence documents are utilized for each plateau to coordinate the sequence of testing activities at that plateau.
In the subsections that follow, a description of the completed sequences of testing is provided. Also included as a part of Section 2.0 is a table showing major milestones for Braidwood Unit 1 which have occurred so far during the startup program and a list of operational modes as defined by the plant Technical Specifications.
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TABLE 2.0 - 1 BRAIDWOOD UNIT 1 MAJOR MILESTONES MAJOR MILESTONES DATE Precritical License Issued 10/17/86 Fuel Load Commenced 10/25/86 5% Power License Issued 05/21/87 Initial Criticality 05/29/87 5% License (Low Power) Tests Completed 06/23/87 Full Power License Received 06/30/87 Entered Mode 1 07/07/87
' Initial Synchronization to Grid 07/12/87 30% Tests Completed 09/19/87 50% Tests Completed 10/23/87 Plant In Service
- 11/20/87 75% Tests Completed 12/02/87 90% Tests Completed 12/10/87 Surveillance Outage 01/01/88 thru 03/21/88
- Illinois Comerce Cot.rnission definition 6
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4 TABLE 2.0 - 2 OPERATIONAL MODES REACTIVITY.
% RATED AVERAGE COOLANT MODE
. CONDITION, K,gg THERMAL FOWER*
TEMPERATURE 1.
POWER OPERATION
> 0.99
> 5%
> 350*F 2.
STARTUP-
> 0.99 15%
> 350*F J.
HCTI STANDBY
< 0.99 0
> 350*F
. 4.
HOT SHUTDOWN
< 0.99 0
350'F>Tavg>200*F 5.
COLD SHUTDOWN
< 0.99 0
1 200'F 6.
REFUELING **
i 0.95 0
1 140*F Excluding decay heat.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
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2.1 - INITIAL CRITICALITY & LOW POWER TEST SEQUENCE (IC & LPT)
The IC & LPT Csquence Document was utilized to define the sequence of tests and operations, beginning with initial criticality, which constituted the low power testing program.
This document ensured that post core loading precritical testing had been completed and results approved prior to continuation of the. testing program.
Prior to commencement of dilution to initial criticality, source range nuclear instrumentation channels were verified to have a signal to noise ratio greater than 2 and power range high level trip setpoints were conservatively set to 20 i 1% of full power.
Plant operating procedures were utilized where appropriate to establish plant conditions.
This sequence document performed a comparison of selected control room indication with the process computer to demonstrate computer monitoring capability. This document obtained a full core flux map with All Rods out (ARO), a flux map with Control Bank D fully inserted with remaining banks withdrawn and a flux map with the control banks at their Hot Zero Power rod insertion limits.
(Refer to Table 2.1 for a tabulation of the flux map results obtained.)
procedure also measured the worth of all control and shutdca., banks less RCCA F-10, the predicted most reactive stuck rod. This is discussed in Section 3.2.9.
Results of individual tests completed during the Initial Criticality and Low Power Test sequence are discussed primarily in Section 3.2 of this report.
Upon completion of this testing phase, the plant was aligned as directed by the Shift Engineer.
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s TABLE 2.1 HZP ARO FLUX MAP RESULTS.
ACTUAL-LIMIT Peak Linear He'at Rate (KW/ft) 0.4028 12.6 Extrapolated to 100%
12.471-12.6 FDHN 1.5447 2.000 t
Fxy - rodded N/A-
- N/A Fxy - unrodded 1.6349 1.8500 Quadrant Power Tilt Ratios 0.9843 1.0102 1.02
. 0028 1.0028 i
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2.2 - TEST SEQUENCE AT 30% POWER The Test Sequence at 30% Power was' utilized to define the activities which constituted the startup testing program from Hot Zero Power, after-completion of low power testing, up to and including the 30% power testing program. This document ensured that the Initial Criticality and Low Power Test Sequence had been completed and results approved prior to entry into Mode 1.
Prior to increasing power for this test sequence, power range high level trip setpoints were conservatively set to 49 1 1% of full power ar.d reactor core flux map results from a low power map were. verified acceptable. The.results were also extrapolated to 49% power to ensure parameters indicative of DNBR and linear heat rate were acceptable for power ascension to the 30% testing plateau.
Refer to Table 2.2 for a tabulation of flux map results.
Until a precision calorimetric could be performed in the 20-30%
power range, reactor power level was monitored by using reactor coolant system delta temperature indication where 100% power was conservatively equated to a 51'F core temperature difference.
Plant operating procedures ware utilized where appropriate to establish plaut conditions and to change reactor power.
During ascension to the 30% plateau, power was stabilized near the 5%,
10%, and 20% levels to accommodate testing at these plateaas.
Results of individual tests completed during power ascension to and while at the 30% plateau are discussed in Sections 3.1 through 3.4 of this report.
At the 30% powat cesting plateau, this sequence document performed a comparison of selected control room indications with the process computer to demonstrate computer monitoring capability.
Data was also taken at the 30% plateau and reviewed to demonstrate the operability of the inadequate core cooling monitoring system. Reactor cavity sump leakage data was obtained and a leckage rate of 0 gpm was determined. The plant was then aligned as directed by the Shift Engineer.
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TABLE 2.2 30% FLUX MAP RESULTS ACTUAL' LIMIT Peak: Linear Heat Rate (KW/ft) 3.358 12.6 Extrapolated to 100%
11.364 12.6 FDl!N.
_ 1.4728 1.8776 Fxy.- rodded N/A N/A Fxy - unrodded 1.5660 1.7684 Quadrant Power Tilt Ratios 0.9890 1.0060 1.02 0.9993 1.0057 i
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2.3 - TEST SEQUENCE AT 50% POWER The Test Sequence at 50% Power was utilized to define the activities which constituted the startup testing program between 30% and 50% power and at approximately 50% of rated thermal power. This document ensured that the Test Sequence at 30%
Power had been completed and the results approved prior to increasing power above the 30% testing plateau.
Prior to increasing power for this test sequence, power range high level trip setpoints were conservatively set to 69 i 1% power and reactor core flux map results from a 30% powar baseline map were verified acceptable.
The flux map results were also extrapolated to 60% power to ensure parameters indicative of DMBR and linear heat rate were acceptable for power ascension to the 50% testing plateau.
Refer to Table 2.3 for a tabulation of flux map results.
Plant operating procedures were utilized where appropriate to establish plant conditions and to change reactor power.
Durirg this testing sequence, power was stabilized near the 35% and 40%
levels to accommodate testing at those power levels.
Results of individual tests completed up to and while at the 50% plateau are discnised in Sections 3.1 through 3.4 of this report. At the 50% testing plateau, a comparison was made of selected control room indications with the process computer to demonstrate computer monitoring capability.
Data was also taken at the 50% plateau and reviewed to demonstrate the operability of the inadequate core cooling monitoring system. Reactor cavity sump leakage data was obtained and a leakage rate of 0 gpm was determined.
Primary containment ventilation system data, auxiliary building ventilation system data, and steam tunnel ventilation system data was obtained and analyzed to verify the adequacy of these systems to provide the necessary cooling and ventilation in their respective areas. Upon completion of this testing phase, the plan. was restored as directed by the Shift Engineer.
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TABLE 2.3 50% FLUX MAP RESULTS ACTUAL LIMIT Peak Linear' Heat Rate (KW/ft) 5.~2853 12.6
. Extrapolated to 100%
11.245 12.6 FDHN 1.4529 1.7965 Fxy - rodded N/A~
N/A Fxy - unrodded 1.5316 1.7143 Quadrant Power Tilt Ratios 0.9885 1.0074 1.02 0.9983 1.0058 13 i
2.4 - TEST SEQUENCE AT 75% POWER The Test Sequence at 75% power was utilized to define the activities which constituted the startup testing program during escalation from 50% to 75% power and at approximately 75% of rated thermal power.
This document ensured that the Test Sequence at 50% power had been completed and the results approved prior to increasing power above the 50% testing
- plateau, prior to increasing power for this test sequence, power range high level trip setpoints were conservatively set to 94 1 1% power and reactor core flux map results from a 50% power baseline map were verified acceptable. The flux map results were also extrapolated to 94% power to ensure parameters indicative of DNBR and linear heat rate were acceptable for power ascension to the 75% testing plateau. Refar to Table 2.4 for a tubulation of flux map results.
Plant operating procedures were utilized where appecgriste to establish plant conditions and to change reactor power. Results of i dividual tests completed while at the 75% plateau are discussed in Sections 3.1, 3.3 and 3.4 of this report. At the 75% testing plateau, a comparison was made of selected coatrol room indications with the process computer to demonstrate computer monitoring capability. Data was also taken at the 75%
testing plateau and reviewed to demonstrate the operability of the inadequate core cooling monitoring system.
Reactor cavity sump leakage data was obtained and a leakage rate of 0 gpm was determined.
Also whila at the 75% testing plateau, an axial xenon oscillation was induced so that flux map data could be obtained to be used for calibration of the excore axial flux difference instrumentation.
Indicated excore delta flux (WI) was driven to approximately -16.5% by insertion of Control Bank D concurrent with a boron dilution. A full core flux map was then obtained and after approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Bank D was borated out to its original position of greater than 180 steps.
This induced an axial xsaon oscillation and WI was allowed to drift to approximately +8% at which time Control Bank D was reinserted by boron dilution to drive WI back to its target value of approximately -5%.
WI was than allowed to drift another 4-5% in the negative direction to approximately -10% at which time Control Bank D was withdrawn to reposition WI back to the target value of approximately -5%.
This maneuver induced an axial xenon oscillation and subsequently dampened the oscillation thereby successfully demonstrating axial xenon oscillation suppressien control.
Upen completion of this testing phase, the plant was aligned as directed by the Shift Enginear.
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TABLE 2.4
-75% FLUX MAP RESULTS ACTUAL LIMIT Peak Linear Heat. Rate ~(KW/ft) 8.245 12.6
- Extrapolated to 100% Power 11.310-12.6 FDHN--
1.4298 1.6760 l
-Fxy - rodded-N/A N/A Fxy - unrodded 1.5152 1.6340
- Quadrant Power Tilt Ditios 0.9887 1.0056 1.02-0.9967 1.0090 15 l
2.5 - TEST SEQUENCE AT 90% POWER The Test Sequence at 90% Power was utilized to define the activities which constituted the startup testing program during escalation from 75% to 90>. power and at approximately 90% of rated thermal power. This document ensured that the Test Sequence at 75% Power had been completed and the results approved prior to ir. creasing power above the 75% testing plateau. Prior-to increasing power for this test sequence, power range high level trip setpoints were set to 108 i 1% power and reactor core flux map results from a 75% power baseline map were verified acceptable. The flux map results were also extrapolated to 108% power to ensure parameters indicative of DNBR and linear heat rate were acceptab.e for power ascension to l
the 90% testing plateau.
Refer to Table 2.5 for a tabulation of flux map results.
Plant operating procedures were utilized where appropriate to establish plant conditions and to change er, actor power.
Results of individual tests-completed while at the 90% platea.a are discussed in Sections 3.2 and 3.3 of th.s report. While at the 90% testing plateau, a baseline flux map was obtained for use in satisfying the initial conditions of the upcoming 100>. sequence document.
Upon completion of this testing phase, the plant was restored as directed by the Shift Engineer.
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_g TABLE 2J
.90% FLUX MAP RESULTS ACTUAL LIMIT Peak Linear Heat Rate (KW/ft) 9.713 12.6' Extrapolated to 100%.
10.926 12.6
'FDHN 1.4245 1.6016
-Fxy - rodded-N/A N/A Fxy - unrodded 1.5211 1.5844 Quadrant Power Tilt Ratios
-0.9869 1.0058 1.02 0.9981 1.0092 b
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r TABLE 3.0 3.1 SYSTEM TESTING AT VARIOUS POWER LEVETE 3.1.1 Pipe Vibration, EM-30 B/C 3.1.2 Steam Generator Level Controller Response, FW-30 A/B/C/D 3.1.3 Thermal Expansion - Feedwater, FW-32B 3.1.4 Main Feedwater (Performance Verification of Waterhanner Prevention System - Lower Nozzle), FW-33B 3.1.5 Chemistry and Radio Chemistry Criteria for Monitoring h3ter Quality During Startup and Power Ascension, PS-32 3.1.6 Radiation Surveys During Power Ascension, PS-33 3.1.7 Pro:ess and Effluent Monitors and Failed Fuel Monitor Checks, PS-34 3.1.8 Rea: tor Coolant System Flow Measurement, RC-31B 3.1.9 Accomatic Reactor Control, RD-36 3.1.10 Heat Capacity Verification for Control Room HVAC System, VC-30 3.1.11 Heat Capacity Verification for Primary Containment Ventilation System, VP-30 3.1.12 Heat Capacity Verification for Switchgear Heat Removal Ventilation System, VX-30 3.1.13 Heat Capacity Verification for Miscellaneous Electrical Equipment Room Ventilation System, VE-30 3.1.14 Dynamic Automatic Steam Dump Control, MS-82 3.1.15 Thermal Expansion - Secondary EM-80 3.1.16 Heat Capacity Verification for Safety Valve Enclosure Room Ventilation System, VV-80 3.1.17 Precritical Checkout of Post Accident Neutron Monitoring, NR-80 3.1.18 Waterhammer Prevention System Dynamic Monitoring (Data Collection During Main Feedwater Isolation Valves Closing Transient) FW-80 3.1.19 Waterhammer Prevention System Dynamic Monitoring (Data Collection During Main Feedwater Isolation Valves Opening Transient) FW-81 3.2 PHYSICS TESTING 3.2.1 Incore Flux Mapping at Low Power, IC-31 3.2.2 Incore Moveable Detector and Incore Thermocouple Mapping at Power, IC-32C 3.2.3 Isothernel Temperature Coefficient Measurement, IT-30 3.2.4 Power Coefficient Determination, IT-31 3.2.5 Determination of Core Power Range for Low Power Physics Testing, NR-32 3.2.6 Reactivity Computer Checkout, NR-33 3.2.7 Bank Worth Measurement at Zero Power, RD-34 A/B 3.2.8 Borca Endpoint Determination, RD-35 3.2.9 Miscellaneous Physics Test Results, FH-32 3.' TRANSIENT TESTING 3.3.1 Ioss of Offsite Power, AP-30 3.3. 2 Load Swing Test, NR-36 3.3.,
Shutdown From Outside the Control Room, RC-35 18
TABLE 3.0 (Continued) 3.4 INSTRUMENTATION AND CALIBRATION TESTING 3.4.1 Calibration of Steam and Feedwater Flow, FW-31 3.4.2 Thermal Power Measurement and Statepoint Data Collection, IT-32B/C/D 3.4.3 Operational Alignment of Excore Nuclear Instrumentation (During Power Escalation), NR-34D 3.4.4 Axial Flux Difference Instrumentation Calibration, NR-35A 3.4.5 Reactor Loose Parts Monitoring, LM-30 3.4.6 Startup Adjustments of Reactor Control Systems, RD-80 3.4.7 Process Computer Verification - Thermal Power, CX-80 3.4.8 Process Computer Verification - Incore Thermocouple, CX-81 3.4.9 Process Computer Verification - Boron Follow, CX-82 3.4.10 Process Computer Verification - Flux Mapping, CX-83 1
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i p-3.1 - SYSTEM TESTING AT VARIOUS POWER LEVELS' 3.1.1 - PIPE VIBRATION (AT APPROXIMATELY 20, 30, 50, AND 75%
REACTOR POWER), EM-30B/C OBJECTIVE
-The Pipe Vibration test procedures demonstrated that the. peak stresses resulting from steady' state flow induced vibration were within allowable design limits. The tests were performed in an effort to qualify ~or accept the piping associated with the Main Feedwater system.
TEST METHODOLOGY The pipe Vibration tests utilized normal operating conditions on the Main Feedwater system at various power levels in order to
' inspect and monitor the effects of flow induced vibration. The inspection consisted of walking down the affected piping and through the~use of portable vibration analyzers. Readings were taken on each sele'eted piping system portion and anywhere else
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levels were deemed excessive.- A simple beam analogy or finite element computer analysis was then applied in the areas of concern arriving atlan allowable deflection. limit'and a comparison'was'made between the theoretical limit and-the actual reading.
Based on-th, outcome, vibration levels less than-the allowable limit met t'..
Acceptance Criteria. Vibration monitoring was also performed using temporarily mounted accelerometers on piping located in areas inaccessible and the
. readings stored in a data collection system. The data was then forwarded to offsite engineering for their analysis and approval.
SUMMARY
OF RESULTS l
The tests performed at the various power levels generated no i-calculations where the levels of vibration were questionable with respect to the piping subsystems under the inspection process. The vibration readings obtained through system walkdowns were analyzed to have met their respective allowable limits and therefore the Acceptance Criteria was met.
Results from the remote vibration monitoring process met all Acceptance
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Criteria limits and was confirmed by offsite engineering.
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3.1.2 - STEAM GENERATOR LEVEL CONTROLLER RESPONSE, FW-30A/B/C/D OBJECTIVE Steam generator level control stability was demonstrated throughout power ascension. Changing feedwater flow configurations and major power changes necessitated the need for multiple performances of these tests.
Level control stability of the four steam generators was demonstrated while operating on the feedwater bypass regulating valves and the main feedwater regulating valves. Level control stability was also demonstrated while transferring feedwater flow between the feedwater bypass regulating valves and the main feedwater regulating valves and while swapping main feedwater pumps.
TEST METHODOLOGY In order to verify level control stability while operating on the bypass or main feedwater regulating valves, a 5% level deviation was manually established in each steam generator.
The control system was chen transferred to the automatic control position.
The actual steam generator level was monitored to determine if it returned to the programmed level of 66% within a specific time frame.
In order to verify level control stability while transferring between the feedwater bypass regulating valves and the main feedwater regulating valves, steam generator level was monitored while transferring from one to the other and then back again.
This was done to ensure stability when transferring to the main feedwater regulating valves in the case of power ascension or to the bypass regulating valves in the case of power descension.
A swapping maneuver among feedwater pumps was performed to verify the pressure stability of the level control system.
Every combination of the three main feedwater pu=ps was tested to ensure stability with any pump supplying feedwater.
SUMMARY
OF RESULTS When given a 5% level deviation (high or low), the bypass regulating valves returned steam generator level to the programmed level within 37.5 minutes as expected. This was done at approximately 6% power.
When given a 5% level deviation (high or low), the main feedwater regulating valves returned the steam generator level to the programmed level within 93 minutes as expected.
This was done approximately 35, 48 and 75% power.
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3.l'.2 - STEAM GENERATOR LEVEL CONTROLLER RESPONSE, FW-30 A/B/C/D (Continued)
Tuning adjustments of all four main feedwater regulating valves were performed when determined appropriate. This activity involved stroke setting and positioner calibration. The corrective actions rectified the periodic operational problems.
After transferring from the feedwater bypass regulating valves to the-main regulating valves, 5% cteam generator level deviations returned to and remained within 2.0% of the programmed level _within 93' minutes. This was done at approximately 23% power.
After transferring from the main feedwater regulating valves to the feedwater bypass regulating valves, the steam generator level returned to and remained within 2.0% of the programmed level within 37.5 minutes, as expected. This was done at approximately 23% power.
When transferring among the-feedwater pumps, the feedwater header pressure oscillations were less than 3% of operating pressure. These ne.euvers were performed at the approximate 35%
and 48% power levels.
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m TABLE 3.1.2 STEAM GENERATOR LEVEL CONTROL
SUMMARY
BYPASS REGULATING VALVE LEVEL CONTROL RESPONSE PERFORMED AT 6% POWER FW-30A TEST ACCEPTANCE ACTUAL TIME STEAM LEVEL CRITERIA IN
RESPONSE
GENERATOR DEVIATION MINUTES IN MINUTES A
5% 0 1 37.5 10.6 5% o i 37.5 15.7 B
5% 0 1 37.5 9.8 5% o i 37.5 13.7 C
5% 0
<-37.5 8.5 5% o i 37.5 15.1 D
5% 0 1 37.5 14.2 5% o i 37.5 14.3 TRANSFER RESPONSE TRANSFER RESPONSE FROM BYPASS TO MAIN FROM MAIN TO BYPASS FEEDWATER REGULATING VALVE FEEDWATER RECULATING VALVE AT 23% POWER AT 23% POWER FW-30B TEST W-30B TEST ACCEPTANCE ACTUAL TIME ACCEPTANCE ACTUAL TIME STEAM CRITERIA IN RESPONSE IN STEAM CRITERIA IN RESPONSE IN GENERATOR MINUTES MINUTES GENERATOR MINUTES MINUTES A
i 90 1.7 A
i 37.5 17.8 B
i 90 4.3 B
i 37.5 12.5 C
i 90 5.5 C
i 37.5 16.7 D
i 90 4.0 D
1 37.5 15.2 FW HEADER PRESSURE W HEADER PRESSURE AT 35% POWER FW-30C TEST AT 48% POWER FW-30D TEST ACCEPTANCE MAXIMUM ACCEPTING MAXIMUM CRITERIA IN PRESSURE CRITERIA IN PRESSURE PUMP PSIG (3% OF FW OSCILLATION PUMP PSIG (3% OF W OSCILLATION SWAP HEADER PRESSURE)
IN PSIG SWAP HEADER PRESSURE IN_PSIG B to C 33.8 15 ALL 35 10 C to A 34.2 10 23
TABLE 3.1.2 (CONTINUED)
MAIN FEEDWATER REGULATING VALVE LEVEL CONTROL RESPONSE AT VARIOUS PCHER LEVELS ACCEPTANCE ACTUAL TIME RESPONSE IN MINUTES STEAM LEVEL
-CRITERIA IN W-30C TEST W-30D TEST W-30D TEST
-GENERATOR DEVIATION MINUTES AT 35% POWER AT 46% POWER AT 75% POWER A
5% O i 93 30.3 37.7 35.0 5% o 1 93 37.3 29.2 33.8 B
5% 0
'I 93 26.3 32.2 39.5 5% o i 93 31.7 39.2 38.7 C,
5% 0 1 93 31.7 26.7 42,0 5% o i 93 27.3 35.5 32.5 D
5% 0 1 93 14.7 26.7 31.0 5% o 1 93 30.1 31.3 28.8 l
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3.1.3 - THERMAL EXPANSION-FEEDWATER, W-32B OBJECTIVE Thermal expansion testing of the feedwater system was conducted to verify that components and piping could expand without restriction of movement uoon system heatup.
It was also conducted to confirm the correct functioning of component supports, piping supports and restraints.
TEST METHODOLOGY At feedwater system hot conditions, system walkdowns were performed.
Piping and components were visually examined and specific snubber positions recorded.
Interferences were identified and dispositioned by the design engineers.
When necessary, system walkdowns were again conducted following the resolution of interferences. All piping movements were evaluated by the design engineers.
SUMMARY
OF RESULTS The piping and components were not to be constrained from expanding and actual thermal expansion movements could not vary from predicted thermal movements by more than i 25% or i 1/4 inch, whichever was greater. During the course of system walkdowns, several minor interferences were determined. These interferences were evaluated by the design engineers and determined to be acceptable as is, or specific corrective action was recommended. All recommended corrective actions were performed.
Some portions of the feedwater system were again examined and measured following the removal of interferences.
Movement of components not within the 1 25% or i 1/4 inch criterion were evaluated by the design engineers on a case-by-case basis.
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3.1.4 - MAIN FEEDWATER (PERFORMANCE VERIFICATION OF WATER HAMMER PREVENTION SYSTEM - LOWER NOZZLE), FW-33B OBJEC:TIVE The purpose of this test was to observe and record the transient following the admission of feedwater into the prehester region of the steam g)nerator. This test was perfcrmed at 28% reactor power.-
TEST METHODOLOGY Fourteen accelou ";ers were attached at various points along the main feedwater lines to record pipe vibration. The admission of f eedwater into the preheater region of the steam generator was 3ccomplished by opening the feedwater isolation valves in accordance with station normal operating procedures.
Offsite engineering performed an evaluation of damaging water hammer using the accelerometer data taken during the performance of the test.
SUMMARY
OF RESULTS The data from the test was successfully recorded and analyzed.
It showe6 is damaging water hammer in any portion of the four main feedwater lines. This test demonstrated water hammer prevention as outlined in NUREG 1606.
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3.1.5 - CHEMISTRY AND RADIOCHEMISTRY CRITERIA FOR MONITORING WATER LALITY DURING STARTUP AND POWER ASCENSION, PS-32 OBJECTIVE 1
This test was performed to verify that the water quality within the primary water make-up system, the reactor coolant system, the steam generator blowdown system, and steam generator feedwater system met the chemistry requirements specified in the Technical Specifications and/or the Westinghouse NSSS guidelines. The test was performed at Pre-Heatup (Mode 5),
Heatup Prior to Criticality (Mode 3), at criticality (Mode 2),
30%, 50%, and 75% Power.
TEST METHODOLOGY The testing was performed by obtaining samples of the primary make-up water, reactor coolant, steam generator blowdown, and main feedwater systems from the appropriate sample panels throughout the plant. Chemical analyses where then performed on every sample and tabulated.
SUMMARY
OF RESULTS During the six executions of this test, all required Acceptance Criteria were adequately met for each system that was sampled.
No corrective actions in plant operation were needed to meet the Acceptance Criteria. On occasions, one of the samples had to be repeated or reanalyzed because a result was not consistent with the sptcification. Upon reanalysis the cample was shown to be within specifications.
Tables 3.1.5-1 through 3.1.5-4 contain a sammary of the results for each system sampled along with the Acceptance Criteria or guidelines stated within the test. The values given in these tables are the results of the samples obtained from each system at each execution of the test.
27
TABLE 3.1.5-1
SUMMARY
OF PRIMARY MAKEUP WATER CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST PS-32 CHEMISTRY CRITERIA /
MODE MODE MODE PARAMETER GUIDELINE 5
3 2
30%
50%
75%
3.83 1.38 1.6
<1.0 1
<1 I
Chloride (ppb)
(LLD*
4.19
<1
<l.0
<1
<1 Fluoride (ppb)
Total Chloride and Fluoride $100 ppb 3.83 5.6
<2.6 (2.0
<2
<2 Silica
$100 ppb
<5 9.3
<5
<5.0
<5 12 pH 0 25 C 6.0-8.0 6.73 6.06 6.17 6.22 6.07 6.15 Specific Conductivity Micromho/cm 8 25'C 11.0 0.493 0.4714 0.706 0.7775 0.998 0.296 Total Solids <l.0 ppm **
N/A N/A N/A 0.54 0.51 0.57 Suspended Solids 10.05 ppm
<0.03
<0.05
<0.05
<0.04
<0.1 0.02 Sodium 110 ppb 2.21
<5.0
<1
<l.0
<1 3.22 Aluminum 120 ppb
<0.01
<l.0
<1
<1.0 3
<2.0 Calcium 15 ppb 1.860
<l.0
<1
<1.0 1
1.35 Magnesium 15 ppb 2.083
<2.0 1.4
<1.0 2
2.23 Potassium 110 ppb 3.06
<2.0
<1
<1.0
<1
<1
- LLD = Lower Limit of Detectability
- Total Solids Criteria added to test beginning with 30% powee execution 28
TABLE 3.1.5-2
SUMMARY
OF REACTOR COOLANT WATER CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST PS-32 CHEMISTRY CRITERIA /
MODE MODE MODE PARAMETER GUIDELINE 5
3 2
30%
50%
75%
<l50 ppb 27
<5
<10 5.3 5
5.46 Fluoride
<l50 ppb.
<49.8
<10-
<49.8
<1
<1
<l.0 Dissolved
<100 ppb 200 9.2 7
20
<1 0
Silica
<1000 ppb 70.3 613 222 557 85 62 Lithium 0.7-2.2 ppm 1.02 1.24 0.75 1.41 1.34 1.6 Hydrogen ** 25-50 cc/kg H O
<LLD
<LLD 41.7 35 31.4 31 2
Suspended Solids (0.200 ppm
<0.16
<0.2 0.05 0.04 0.18 0.04 Aluminum (50 ppb 2.1 7.8 17
<l.0 4
8.46 Calcium (50 ppb 3.41 2.9 1.9
<1.0 1
<1.0 Magnesium
<25 ppb 2,19 1.9
<1
<1.0
<1
<l.0
- When Tave > 180*F
- When RCS > 1% Reactor Power 39
TABLE 3.1.5-3
SUMMARY
OF STEAM GENERATOR WATER CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST PS-32 CHEMISTRY CRITERIA /
MODE MODE MODE PARAMETER GUIDELINE
- O 3
2 30%
50%
75%
Cation Conductivity Micromho/cm 9 25'C
<0.8 N/A N/A N/A
.<0.76
<0.7
<0.62 pH 9 25 C
>9.0 N/A N/A N/A 9.4 9.3
>9.1
- Guidelines not applicable until > 30% power 30
TABLE 3.1.5-4
SUMMARY
OF FEEDWATER CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST PS-32 CHEMISTRY.
CRITERIA /
MODE MODE MODE PARAMETER GUIDELINE
- 5 3
2 30%
50%
75%
Hydrazine
>20 ppb N/A N/A N/A 1303 166.05 79 Dissolved Oxygen 13 ppb N/A N/A N/A 2.36 0.09 0.51 Guidelines not applicable until > 30% Power l
4 d
d 31
l 3.1.6 - RADIATION SURVEYS DURING POWER ASCENSION (3-10% and 48-52%
Power), PS-33 OBJECTIVE The Radiation Surveys During Power Ascension test was performed to determine dose levels at specified points throughout the plant and to verify the effectiveness of rad 2ation shielding to gamma and neutron radiation.
TEST METHODOLOGY Gamma radiation dose rate values were established by surveying with portable survey instrumentation in the Service, Radwaste, Fuel Handling, Turbine, and Auxiliary Buildings and the Unit 1 Containment. Neutron radiation dose rate values were established in the Unit 1 Containment and certain penetration areas.
SUMMARY
OF RESULTS The effectiveness of neutron radiation shielding in Unit 1 Containment was found to be inadequate during performances of this test.
Radiation surveys for selected base points would be repeated after any neutron shielding modifications are completed.
l s
Th6 effectiveness of gamma shielding and lhe general determination of dose levels were found adequate during performances of the test. At nominally one half of full power, gamma dose rates ranged from (0.1 mR/hr in the Service and Turbine Buildings to 3.2 mR/hr at one Auxiliary Building location. Auxiliary Building Neutron dose rates ranged from <5 mrem /hr to 5.0 meem/hr.
In the Containment Building gamma dose rates ranged from 0.1 mR/hr to 46 mR/hr. Only six points exceeded 5.0 mR/hr; two at 6 mR/hr, one each at 12, 17, 40 and 46 mR/hr. Neutron dose rates in the Unit 1 Containment Building ranged from <5 mrem /hr to 300 mrem /hr.
Only four points exceeded 20 mrem /hr: one each at 32, 35, 250, and 300 mrem /hr.
All measured dose rates have been dispositioned as acceptable for continued plant operation.
i i
i 32 i
{
3.1.7 - PROCESS ANL EFFLUENT MONITORS AND FAILED FUEL MONITOR CHECKS (5% and 30% Power), PS-34 OBJECTIVE The Process and Effluent Monitors and Failed Fuel Monitor Checks startup test was performed to verify proper responses of all process and effluent monitors and the failed fuel monitor to known sources of radiation.
TEST MEIllODOLOGY Radioactive sources of known activities were exposed to the process and effluent monitor and failed fuel monitor detectors.
The observed detector responses were compared to the calculated expected detector responses to the radioactive sources to determine percent differences.
SUMMARY
OF RESULTS All except 10 tested procsss and effluent monitor and failed fuel monitor detector responses were within ! 20% of the expected detector responses during the performances of this test, thus meeting the test Acceptance Criteria of i 20%
agreement.
Several errore, discrepancies, and equipment problems discovered in post test review for the two executions of the test will be closed following recalibrations and will be confirmed in the final test execution at full power.
1 l
l 33
3.1.8 - REACTOR COOLANT SYSTEM (RCS) FLOW MEASUREMENT (30,. 50, and 75% POWER), RC-31B OBJECTIVE The RCS Flow Men
" test procedure was performed at various reactor power levet t 2S, 50, and 75% power) to determine the RCS flowrate for each.
4 loops and also the total RCS flowrate.
TEST METHODOLOGY While the plant was at 30, 50, and 75% reactor power level data was obtained to determine the RCS flowrate. This data consisted of RCS hot and cold leg RTDs both nominal and reverse pclarity resistance readings. These RTD resistance readings weto converted to temperatures (*F) and then using Steam Tables, hot
& cold leg enthalpies and cold leg specific volume were determined.
This data combined with calorimetric power values from the appropriate Startup Test IT-32C, or IT-32D, were used to calculate the flowrates. This method of calculation was used each time the test was performed at the various power levels.
The actual calculated flowrater for the various power levels were as follows:
TOTAL RCS
% POWER LOOP 1 LOOP 2 LOOP 3 LOOP 4 FLOWRATE 30%
102,954 97,802 102,912 100,698 404,366 50% (first 102,002 99,509 103,582 101,704 406,797 execotlon) 75%
101,878 98,906 103,099 99,769 403,652 All the results were consistent with the expected values and met the minimum flow requirements for total RCS flowrate.
SUMMARY
OF RESULTS The required total RCS flowrate as determined by calorimetric measurement varied for the different power levels and were as follows:
% POWER TOTAL FLOWRATE
% THERMAL DESIGN VALUE 30%
> 339,840 gpm 90%
50%
1 377,600 gpm 100%
75%
2 390,400 gpm Technical Specification minimum value 34
3.1.9 - AUTOMATIC REACTOR CONTROL, RD-36 OBJECTIVE This procedure demonstrated the capability of the automatic reactor control system to maintain reactor coolant system average temperature within acceptable tolerance of Trof under steady state and transient conditions.
TEST METHODOLOGY With reactor power stabilized at approximately 30* and Tave 4
matched to Tref, rod control was placed in automatic to monitor for oscillations. After approximately thirty minutes, Tave was increased.to 6*F higher than Tref ty manual withdrawal of Control Bank D.
Rod control was then placed in automatic and Tave allowed to stabilize. After Tave stabilized, rod control was placed in manual to decrease Tave 6'F lower than Tref by insertion of Control Bank D.
Rod control was placed back in automatic and Tave was allowed to stovilize.
SUMMARY
OF RESULTS During steady state operation, it was found that Tave maintained Tref with no problems being encountered. When Tave was increased by 6
- F, it took approximately 2.6 minutes to return Tave to within 1.5'F of Tref.
When Tave was decreased by 6
'F, it took approximately 3.1 minutes to return Tavo to within 1.5'F of Tref.
Both of these times were well within tne Acceptance Criteria time of '.0 minutes.
i l
1 i
l l
I l
l 35 l
l l
3.1.10 - HEAT CAPACITY VERIFICATION FOR CONTROL ROOM HVAC SYSTEM, VC-30 OBJECTIVE The Heat Capacity Verification for Control Room HVAC System procedure was performed to provide heat removal capacity data for-the Control Room ventilation system. -The data obtained in the procedure was analyzed to verify that the design heat loads could be removed from the Main Control Room and associated areas.
TEST METHODOLOGY In order to obtain data for the heat removal capacity of each Control Room ventilation system train, each train was tested independently.
The temperatures, heat loads, cooling air flow rates and water flow rates for each train were measured. The data for each train was then analyzed to determine that the design heat loads could be removed.
sUMHARY OF RESULTS The test room temperature data did not satisfy the Acceptance Criteria limits for several rooms adjacent to the Control Room.
The test data values were analyzed and verified to meet the appropriate design heat load removal requirements.
Both trains satisfied design heat load removal requirements.
l 36
3.1.11 - HEAT CAPACITY VERIFICATION FOR PRIMARY CONTAINMENT VENTILATION, VP-30 OBJECTIVE The Heat Capacity Verification for Primary Containment Ventilation procedure was performed to provide heat removal capacity data for the primary containment ventilation system.
The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.
TEST METHODOLOGY In order _to obtain data for the heat removal capacity of the primary containment ventilation system each independent train of primary containment ventilation was tested with the cther train off. The temperatures, heat loads, and cooling air and water temperatures for each train of primary containment ventilation were measured at 0% reactor power with the Reactor Coolant System ) 550*F.
The data for each train of primary containment vr :tilation was then analyzed and extrapolated to 100% reactor power to verify that the design heat loads could be remo* red.
SUMMARY
OF-RESULTS Offsite engineering analyzed the data obtained and concluded that the s'rstem was capable of removing the design containment heat load.
Offsite engineering also requested selected primary containment ventilation system parameters be monitored over a period of time to provide additional confirmation of analysis asrumptions.
This data collection is in progress.
37
3.1.12 - HEAT CAPACITY VERIFICATION FOR SWITCMGEAR HEAT REMO &
VENTILATION VX-30 OBJECTIVE The Heat Capacity Verification for Switchgear Heat Removal Ventilation procedure was performed to provide heat removal capacity data for the switchgear heat removal ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be remov'd.
TEST METHODOLOGY In order to obtain data for the heat removal capacity of the switchgear heat removal ventilation system, the system was divided into subsystems and tasted independently. The temperatures, heat loads, and cooling air flowrates for each subsystem (Div 11 and Div 12 ESF switchgeat, power cable spreading room Div 12, and the non-essential switchgear room) were measured. The data for each subsystem was then analyzed to determine that the design heat loads could be removed.
SUMMARY
OF RESULTS Test data was submitted to offsite engineering for analysis.
The analysis concluded that extrapolated room temperatures for the subsystems were acceptable. The test Acceptance Criteria of room temperatures less than or equal to 104*F was met as the maximum room temperature found was 92*F.
38
1 3.1.13 - HEAT CAPACITY VERIFICATION FOR MISCELLANEOUS ELECTRICAL EQUIPMENT ROOM VENTILATION SYSTEM, VE-30 OBJECTIVE The Heat Capacity Verification for Miscellaneous Electrical Equipment Room Ventilation System procedure was performed to
' provide heat removal capacity data for the miscellaneous electrical equipment room ventilation system. The data obtained in the -procedure was analyzed to verify that the design heat loads could be removed.
TEST METHODOLOGY In order to obtain dato for the heat removal capacity of the miscellaneous electrical equipment room ventilation system, the system was divided into subsystems.and tested independently.
The temperatures for each subsystem were measured. The data for each subsystem (Div 11 and Div 12) was then analyzed to determine that the design heat loads could be removed.
SUMMARY
OF RESULTS The data, as analyzed by offsite engineering showed that each subsystem of miscellaneous electrical equipment room (MEER) ventilation was able to remove tne design heat loads. The Test Acceptance Criteria of room temperatures less than or equal to 104*F was met as the maximum room temperature found was less than 96*F.
39
3.1.14 - DYNAMIC AUTOMATIC STEAM DUMP CONTROL, MS-82 OBJECTIVE To. demonstrate that each bank of condenser steam dump valves is capable of relieving their designed capacity of steam.
TEST METHODOLOGY With reactor power between 20% and 52% one bank of three steam dump valves is opened and the change in reactor power due to the newly imposed steam demand is recorded.
Each three valve bank is expected to pass steam flow equivalent to 10 1 2% of reactor power.
SUMMARY
OF RESULTS The four banks of steam dump valves (3 valves per bank) were found to relieve steam flows equivalent to range of between 10.6% and 11.75% reactor power '<hich satisfied the 10 t 2%
Acceptance' Criteria. The test demonstrated that the steam dumps are capable of supporting load rejection testing.
l l
l l
40 V
y w
3.1.15 - THERMAL EXPANSION-SECONDARY, EM-80 OBJECTIVE Thermal expansion testing of secondary systems was conducted to verify that components and piping could expand without restriction of movement upon system heatup.
It was also conducted-to confirm the correct functioning of component supports, piping supports and restraints.
TEST METHODOLOGY At secondary-systems ambient and hot conditions, system walkdowas were performed.
Piping and components were visually examined and specific snubber positions recorded.
Interferences were identified and dispositioned by the design engineers. When necessary, system.walkdowns were again conducted following the resolution of interferences. All piping movements were evaluated by the design engineers.
SUMMARY
OF RESULTS The piping and components were not to be constrained from expanding and actual thermal expansion movements could not vary from predicted thermal movements by more than 1 25% or 1/4 inch, whichever was greater. -During the course of system walkdowns, several minor interferences were determined. These interferences were evaluated by the design engineers and determined to be acceptable as is, or-specific corrective action was recommended. All recommended corrective actions were initiated.
Some portions of the secondary systems were again
_ examined and measured following the removal of interferences.
Hovement of components not within the i 25% or i 1/4 inch criterion were evaluated by the design engineers on a case-by-case basis, g
i 1
41 1'
l-l
3.1.16 - HEAT CAPACITY VERIFICATION FOR SAFETY VALVE ENCLOSURE ROOM VENTILATION SYSTEM, VV-80 OBJECTIVE The Heat Capacity Verification for Safety Valve Enclosure Room Ventilation System procedure.was performed to provide heat removal capaci' y data for the safety valve enclosure room a
ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.
TEST METHODOLOGY The heat removal capacity procedure was performed at approximately 75% reactor power and extrapolations were made to 100% power conditions.
The temperatures,-heat loads, and cooling air flowrates for the safety valve enclosure room ventilation system were measured.
The data obtained was then analyzed to determine that the design heat load could be removed.
SUMMARY
OF RESULTS The data obtained was submitted to offsite engineering for analysis.
Extrapolations to design ceaditions determined that temperatures would be acceptable.
The Test Acceptance Criteria of room temperatures being less 122*F was met as the maximum room temperature found was 114*F.
42 V
3.1.17'- PRECRITICAL CHECKOUT OF POST ACCIDENT NEUTRON MONITORING SYSTEM, NR-80 OBJECTIVE To verify proper operation of both channels of the Post Accident Neutron Monitoring System (PANMS) alarms, indications, and test functions. To verify that the PANMS detector high voltages and discriminators'were in their plateau regions.
TEST METHODOLOGY Using a neutron source in close proximity to the detector under test, the high voltage and discriminator voltages were varied and detector responses recorded and plotted to determine the voltage plateau regions. The high voltage was also reduced to verify trouble alarm operation.
The discriminator and high voltages were then returned to their initial values.
Test function switches were actuated and various test point's voltages were monitored.
SUMMARY
OF RESULTS The as found discriminator and high voltages for each detector were demonstrated to be within the plateau region.
Alarm and internal test functions were found to operate in an acceptable manner.
43
{
t
3.1.18 - WATERHAMMER PREVENTION SYSTEM DYNAMIC MONITORING (D?TA COLLECTION DURING MAIN FEEDWATER ISOLATION VALVES CLOSING TRANSIENT), FW-80
+.
OBJECTIVE To verify that the closure of the Main Feedwater Isolation Valves does not result in primary or secondary plant transients.
TEST METHODOLOGY Selected plant parameters were trended as power was reduced and feedwater flow was transferred from the main nozzle to the auxiliary nozzle. A verification was performed as to whether or not operator intervention was required to mitigate any transient caused by this flow transfer.
SUMMARY
OF RESULTS Data was obtained snowing excellent plant stability throughout the flow transfer. No operator intervention was required.
44
3.1.19 - WATERHAMMER PREVENTION SYSTEM DYNAMIC MONITORING-(DATA COLLECTION DURING MAIN FEEDWATER ISOLATION VALVES OPENING TRANSIENT), FW-81 OBJECTIVE To-verify that the opening of the Main Feedwater Isolation Valves does not result in primary or secondary plant transients.
TEST METHODOLOGY Selected plant parameters were trended as power was increased and feedwater flow was transferred from the auxiliary nozzle to the main nozzle. A verification was performed as to whether or not operator intervention was required to mitigate any transient caused by this flow transfer.
SUMMARY
OF RESULTS Data was obtained showing excellent plant stability throughout the flow transfer.
No operator intervention was required.
Purge flows and preheater flows were verified to increase steadily to their required values.
45
3.2 - REACTOR PHYSICS TESTING 3.2.1 - INCORE FLUX MAPPING AT LOW POWER, IC-31 OBJECTIVE The objectives of this test were to obtain or verify high voltage plateaus for all six detectors required to perform an incore flux map, to verify the performance of the incore movable detector. flux mapping (IC) system and to cbtain a flux map at low power (less than 5%). This included an initial operational alignment and checkout of the IC system and associated automatic-data logging devices.
TEST METHODOLOGY The detectors were inserted into the core to the approximate maximum flux location in the core, and the applied voltage was incremented to determine the platea 1 curve (defined as the range of applied detector voltage over which the detector output is nearly constant)..The operating voltage selected was the mid-range of the plateau region. With the detectors set at their operating voltages, the sequence for obtaining a full core flux map was performed. Thirteen passes through the core, each with one detector in the calibrate position, J10, were obtained (one pass was repeated to acquire previously missed data due to a temporarily blocked detector path). The data collected during the flux map was processed using an offsite INCORE program.
SUMMARY
OF RESULTS The plateau regions were determined and the operating voltages were established at 90V for all detectors. All 58 core locations were mapped, meeting the Technical Specification minimum of 44 locations. Analysis of the flux map data was as follows:
Parameter Flux Map-Result Acceptance Criteria l
Incore Flux Tilt 1.0102 i 1.04 FDHN 1.5447 1 1.87 l
Pg(Z) 2.4197 5 3.0 i
l 46 l
l l
3.2.2 - INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING AT POWER (QUARTER CORE), IC-32C OBJECTIVE The purpose of this procedure was to provide a method for
- obtaining quarter cure flux m3ps during transient reactor xenon conditions.
TEST METHODOLOGY In the'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> previous to the test a voltage plateau procedure had been executed to determine the operating voltaaes for the incore detectors. A full core flux' mapping procedure had been performed less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> previous to the quarter core flux mapping (QCFM) to provide a baseline for detector calibration factors. The test was performed as part of the xenon transient data acquisition for Axial Flux Difference calibration at approximately 75% power. The flux maps were analyzed using the offsite INCORE computer code.
SUMMARY
OF RESULTS The flux map required to support AFD calibration was obtained.
The results of the quarter core flux map were determined acceptable.
47 L
3.2.3 - ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT, IT-30 OBJECTIVE This test was performed to determine the Isothermal Temperature Coefficient (ITC) of reactivity, and to derive the Moderator Temperature Coefficient (MTC) of reactivity at the beginning of core life for the initial fuel cycle.
TEST METHODOLOGY The ITC was determined by measuring the change in reactivity induced by changing the temperature of the moderator, clad and fuel and dividing by the temperature change. The MTC is obtained by applying a precalculated Doppler fuel coefficient correction factor to the ITC value to eliminate the fuel temperature change portion of the ITC.
This test was performed at Hot Zero Power conditions starting with a Reactor Coolant System (RCS) cooldown of approximately 3*F at a rate of approximately 10'F/hr. After a stabilization period, an RCS heatup was initiated for an appropriate 3'F increase at a rate of approximately 10*F/hr.
Plots of reactivity vs. temperature were maintained for each cooldown and heatup portion of the test.
Each cooldown and heatup was performed two times for each of the following control rod configurations: Control Banks D &
C in (with B > 200 steps), Control Bank D in (with C 2 200 steps) and All Rods Out (Control Bank D 2 200 steps).
SUMMARY
OF RESULTS All values of the measured ITC were within 1 pcm/*F of one another and all were within i 3.0 pcm/*F of the acceptance criteria values.
During the All Rods Out execution of the test the MTC was determined to be -0.45 pcm/*F meeting the acceptance criteria and the Technical Specification requirement of All Rods Out MTC < 0 pcm/*F.
Table 3.2.3 summarizes the results of the ITC measurements.
48
_ =
P
.?l
- Acceptance Criteria Measured Acceptance Criteria MTC ARO
-2.65 i 3.0 pcm/'F
-2.36 pcm/*F
<0 pcm/*F
-0.45 pcm/'F CBD'In
-4.21 1 3.0 pcm/*F
-3.92 pcm/*F N/A
-2.01-pcm/'F CBD&C In
-8.00 1 3.0 pcm/'F
--7. 98 pcm/
- F N/A
-6.07 pcm/'F l
49
+8 7
3.2.4 - POWER COEFFICIENT Df:ERMINATION (30%, 50%, 75%,, 90%), IT-31 OBJECTIVE The objactive of this procedure was to verify the design predic'. ions of the Doppler Only Power Coefficient during the power ascension startup testing sequences at 30%, 50%, 75% and 90% pcwer.
TEST KETHODOLOGY The Doppler power coefficient verification was performed by varying plant load by approximately 2% (1% at 30% power). Three successive load swings were performed, allowing Tave and WT to stabilize prior to each load swing. No changes in rod position or RCS boron concentration are allowed during the load changes.
The change in WT due to the load swing reflects the change in power and the change in Tave reflects the change in core reactivity.
SUMMARY
OF RESULTS Af ter the load swings were performed, values for Tave and WT were taken from the strip chart recordings.
Using this data, and the power readings obtained prior to and after the load swings, the average measured Doppler coefficient verification factor (Cm) was calculated, where Cm is effectively the measured ratio of change in Tave vs change in reactor power.
This value of Cm was thcn compared with the predicted Doppler coefficient verification factor (CP). CP was calculated using the Doppler Only Power Coefficient and Isothermal Temperature Coefficient predictions from the core design report. The absolute values of Cm and CP were compared to determine if the acceptance criteria of not greater than a 0.5'F/% power difference were met. See Table 3.2.4 for the actual results.
50
TABLE 3.2.4 Measured vs Predicted Power Coefficient Verification.
Factor Absolute Values Test m
p Power (C )
(C )
Difference 30%
1.98 *F/% Power 2.06 *F/% Power 0.08 F/% Power.
50%
1.46 *F/% Power 1.59 *F/% Power 0.11 *F/% Power 75%-
0.99 'F/% Power 1.10 *F/% Power 0.11 *F/% Power' 90%
0.90 *F/% Power 0.98 *F/% Power 0.08 *F/% Power F
4 51 L
3.2.5 - DETERMINATION OF CORE POWER RANGE FOR LOW POWER PHYSICS TESTING, NR-32 OBJECTIVE This procedure was used to determine the power level (neutron flux level) at which detectable reactivity feedback effects from nuclear fuel heating occurred and to establish the range of neutron flux in which zero power reactivity measurements were performed.
TEST METHODOLOGY Initial conditions were established with the RCS at an average temperature of 555.4'F, RCS pressure at 2235 psig and the RCS boron concentration at 1062 ppm.
The reactor was critical with the flux at 1.1 X 10-8 amps on both Intermediate Range channels.
Initially, the reactivity computer was verified to be set up using power range channel N-44 which was taken out of service.
Reactivity computer outputs of reactivity and flux along with RCS Tava were displayed on a four~ pen strip chart recorder.
The Tave input taken was from the 7300 process instrumentation racks.
A reactor period measurement using actual core reactivity changes was then made on the reactivity computer and compared to the predictions from the core design report (WCAP 10935).
Next, the determination of the power range for low power physics testing was made. Control Bank D was withdrawn to achieve a positive reactivity addition of 25 1 3 pcm.
Reactivity and flux level were then observed to determine the point of adding nuclear heat as indicated by negative reactivity addition from the Doppler fuel temperature coefficient. RCS Tave was also monitored for changes as an indication of nuclear heating.
Tave was expected to increase as a result of the nuclear heating effects in the fuel.
SUMMARY
OF RESULTS The reactor period measurement was four.d to be within 0.1% of the design value, which was within the allowed 4% tolerance.
The reactor power level at which detectable reactivity feedback l
effects from nuclear heating occurred was determined to be 8 X 10-7 amps on both Intermediate Range (IR) NIS channels.
This was within the expected range of 5 X 10-7 to 5 X 10-6 amps.
The neutron flux level range at which zero power reactivity measurements were to be performed was detarmined to be 6 X 10-9 j
l l'
to 6 X 10-8 amps as indicated on the reac tivity computer.
The l
l range of neutron flux levels allowed for physics testing that is l
actually trended and used for physics testing applies to the j
reactivity computer flux output. An analogous range for the IR NIS channels can be determined but is not used during subsequent testing.
52 l
3.2.6 - REACTIVITY COMPUTER CHECKOUT, NR-33 OBJECTIVE The Reactivity Computer Checkout test procedure was performed to demonstrate proper operation of the NSSS Vendor supplied (Westinghouse) reactivity computer through dynamic tests using actual neutron flux signals and core reactivity changes.
TEST METHODOLOGY An approximate 25 pcm reactivity increase as shown on the reactivity computer strip chart recorder was initiated by withdrawing Control Bank D.
A stopwatch was used to determine the reactor flux doubling time. This doubling time was used to determine the theoretical reactivity increase using design predictions. The predicted reactivity increase was compared to the reactivity indicated on the reactivity computer strip chart recorder.
This was repeated for reactivity increases of approximately 50.and 75 pcm.
SUMMARY
OF RESULTS The acceptable average deviation between indicated and predicted reactivity was required to be less than 4%.
The absolute deviations between indicated and theoretical reactivity were 0.2%, 0.0% and 0.6% for reactivity gains of 25.3 pcm, 49.5 pcm and 77.3 pcm respectively.
The average absolute deviation between indicated and predicted reactivity was 0.3%.
53
F l
3.2.7a - BANK WORTH M QSUREMENT AT ZERO POWER, RD-34A OBJECTIVE The Bank Worth Measurement at Zero Power test procedure was
' performed to determine the differential and integral worth of selected individual control rod banks and to determine the differential boron worth over the control banks for comparisons to the design predictions made in the NSSS vendor supplied core design report (WCAP 10935).
TEST METHODOLOGY In order to perform the bank worth measurement a reactor coolant system dilution was initiated and the resulting reactivity change compensated for by inserting the bank to be measured.
During the course of the boron concentration change, flux and reactivity signals were recorded using the reactivity computer.
From the reactivity traces a total reactivity change for the movement of the bank was determined as well as differential reactivity changes per step of bank motion.
The value of integral bank worth was then combined w!.th data from the appropriate boron endpoint measurement to determina the total bank worth.
In order to determine the differential boron worth over the conttol banks, a least squares fit was applied to :he plot of totalized integral bank worths versus selected ban: endpoint boron concentrations. The slope of the the line w.s the differential boron worth.
SUMMARY
OF RESULTS For the bank worth measurements the acceptable variation from the design predictions was 110%.
For each of the measuremtits the bank worth was found to be within this limit.
The average of the absolute values of percent error between measured and predicted worths was 2.9%.
The best agreement was found in the measurement of Control Bank B which was within 0.2%
of its predicted value. The worst agreement was found in the measurement of Shutdown Bank D which was within 6.7% of its predicted value. A summary of the bank worth measurements and the predicted values appears in Table 3.2.7 and plots of selected integral and differential bank worths are found in Figures 3.2.7-1 through 3.2.7-4.
For the determination of differential boror. worth over the control banks the acceptable variation from the design prediction was 110%. The determined value of -11.54 pcm/ ppm is within 10% of the predicted value of -11.34 pcm/ ppm (See Figure 3.2.7-7).
54
3.2.7b - BANK WORTH WITH OVERLAP MEASUREMENT AT ZERO POWER, RD-34B OBJECTIVE The Bank Worth With Overlap Measurement test procedure was performed ?r determine the integral and differential worths of the control banks in an overlap configuration and to compare these values with Westinghouse design predictions.
TEST METHODOLOGY The bank worth measurements were made in this procedure by initiating a reactor coolant system boration and compensating for this insertion of negative reactivity by manual withdrawal of the control banks in normal overlap.
Flux and reactivity were monitored on the reactivity computer.
Boration was continued until Control Bank D was positioned at 126 stsps withdrawn after which no further changes in reactivity were made. The remaining data for Control Bank D was obtained from the executed test RD-34A, Bank Worth Measurement at Zero Power for Control Bank D.
Integral worth was then calculated by adding up the increments of reactivity over the entire range of bank travel.
Differential worth was calculated by dividing the reactivity increments by the number of control bank steps the increment represented.
Both types of bank worths were plotted to illustrate their position dependence and the core properties they demonstrate (See Figures 3.2.7-5 and 3.2.7-6).
SUMMARY
OF RESULTS The integral reactivity worth of the control banks with overlap was expected to be 3639 pcm i 10%. The value obtained was 3770.3 pcm, which was within 0.9% of the predicted value, t
l l
l l
l 1
l l
l l
55 l
l
TABLE 3.2.7 Measured Versus Predicted Bank Worths MEASURED PREDICTED BANK WORTH [pcm)
. WORTH [pcm)
% DIFFERENCE-SBC' 1152.7 1211 i 121
-4.8%
SBD 815.5 764 1 76
+6.7%
SBE 568.2 588 1 59
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CBA 626.9 619 1 62'
+1.2%
CBB 1254.0 1256 126
-0.2%
CBC 1120.0 1130 1 113
-0.9%
CBD 653.3 634 1 63
+3.0%
56
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D I F F E? R E NTI AL.
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63 1
3.2.8 - BORON ENDPOINT DETERMINATION, RD-35 9
OBJECTIVE This test determined the rod worths at the extreme ends of bank-travel (at the near fully withdrawn or near fully inserted position) for use in the Bank Worth Measurement at Zero Power test, RD-34A.
In addition, the just critical Reactor Coolant
-System (RCS) boron concentration wts determined at selected RCC bank configurations along with the All Rods Out critical boron concentration and the boron reactivity worth for each control bank.
TEST MImiODOLOGY With RCS loop and pressurizer boron concentration having been alternately sampled at least twice each ct approximately thirty minute intervals and within 20 ppm of each other, RCC banks were positioned to within 50 pcm of the desired endpoint configuration with the reactor just critical in the zero power physics testing range. The controlling bank was then inserted or withdrawn to reach the desired endpoint configuration while neutron flux, reactivity, RCS Tave and pressuriser level were monitored on a strip chart recorder.
When the reactivity trace had stabilized, the controlling bank was repositioned to re-establish the initial flux level. This process was performed threa times for each desired RCC configuration.
SUMMARY
OF RESULTS The rod worths at the extreme ends of travel were determined and added t3 the worths measured in the bank worth measurement test procedure. The All Rods Out just critical RCS boron concentration was found to be 1082.2 ppm which mat the acceptance criterion of 1080 1 50 ppm.
Boron reactivity worths are shown in Table 3.2.8.
4 -
l 64 i
TABLE 3.2.8 MEASURED VERSUS PREDICTED BORON ENDPOINT CONCENTRATIONS AND BORON REACTIVITY WORTHS ACTUAL BEP PREDICTED BEP BORON REACTIVITY PREDICTED CONCENTRATION CONCENTRATION WORT 9 [ ppm)
[ ppm]
[ ppm]
REACTIVI1f WORTH [ ppm]
_SBE out (All CB in) 766.3 759 1 76 N/A N/A CBA out (CBB, C,D in) 821..
814 i 81 55 55 1 6 CBB out (CBC, D in) 933.8 922 1 92 112,5 108 1 11 CBC out (CBD in) 1027.9 1024 1 102 94.1 102 1 10 CBD out (All Rods Out) 1082.2 1080 t 50 54.3 56 6
N-1 (All Rods fully 481.8 469 i 47 N/A N/A inserted except F-10 fully withdrawn) i i
i 65 l
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3.2.9 - MISCELLANEOUS PHYSICS TEST RESULTS OBJECTIVES Additional testing covered in the Initial Criticality r.nd Low Power Test Sequence included measuring the N-1 rod worth, the worth of all rods fully inserted except RCCA F-10, the predicted most reactive stuck rod.
METHODOLOGY To determine the N-1 rod worth a rod swap was done with Shutdown Bank B.
RCCA F-10 was withdrawn while Shutdown Bank B was inserted to maintain just critical conditions. When RCCN F-10 was fully withdrawn, a dilution was initiated to insert the remaining worth of Shutdown Bank B and the worth of Shutdown Bank A.
A Boron Endpoint Determination (RD-35) was performed for the last 25 steps of Shutdown Bank A.
SUMMARY
OF RESULTS All criteria were satisfied. A tabulation of the measurement results, along with associated predicted values and acceptance criteria, appears below:
Parameter Measured Measured Value Acceptance Criterion N-1 Rod Worth 6987.2 pcm
>6510 pen (7230 1 723 pcm predicted)
N-1 Critical Boron 481.8 ppm 469 1 47 ppm predicted N-1 Rod Bank. Boron Worth 600.4 ppm 611 1 61 ppm predicted 66
r:-
3.3 - TRANSIENT TESTING 3.3.1 - LOSS OF OFFSITE POWER, AP-30 0_BJECTIVE The' Loss of Offsite Power test was performed to verify the plant's ability to safely sustain a loss of offsite power condition for -thirty minutes.
TEST MEzHODOLOGY In order to establish the unit's ability 6.o safely sustain a loss of offsite power the following RCS a.id steam generator parameters were recorded:
RCS loop wide range temperatures, auxiliary feedwater pumps discharge pressures, presrurizer (PZR) pressure and level, steam generator (SG) level, anti SG pressure.
The emergency diesel generators ween monitored along with the Engineered Safeguards Features (ESF) loecs to verify proper diesel starting, ESF bus load shedding, and proper sequenced starting of the required ESF loads. The operating personnel used their regular Station Emergency Procedures while responding to the 30 minute transient. The event was initiated by electrically isolating the Unit 2 power feeds from the offsite power grid and from Unit 1.
Multiple offsite power sources and Unit 1 crosstics were always available if needed.
SUMMARY
OF RESULTS The test procedure specified six acceptance criteria parameters to evaluate plant performance during the transient. Four of the parameters dealt with plant component operations and were satisfied as follows:
- 1) All required loads on the ESF buses were shed prior to the diesel generators energizing the ESF buses.
both trains of centrifugal charging pumps, both trains of component cooling pumps, both trains of essential service water pumps, both trains of control room chillers, and the motor-driven auxiliary feedwater pump. Station procedures called for a manual start of the motor-driven auxiliary feedwater pump prior to its normal sequenced start. A separate test section independently verified proper load sequencing of bus 141.
- 3) The diesel-driven auxiliary feedwater pump automatically started following the initiation of the loss of offsite power. The pump tripped on low suction pressure because the low suction pressure trip circuitry was de-energized by the loss of offsite power. This circuitry has been modified and retested successfully.
67
3.3.1 - LOSS OF OFFSITE POWER, AP-30 (Continued)
- 4) Was met by having at least one safe shutdown train operable on emergency power during the 30 minute transient. The test was terminated 35 minutes after initiating the loss of offsite power, s
The remaining two acceptance critoria dealt'with the voltage and frequency parameters of the diesel generators and the minimum / maximum values for the RCS parameters mentioned earlier.
.5)
The diesel generators started and energized the ESP buses to 4160 ! 420 volts and 60 i 1.2 HZ in 11.1 (Bus 141) and 11.6 (Bus 142) seconds after receiving the undervoltag_ start signal.
- 6) The acceptance criteria for the RCS parameters were as follows:
RCS Tave less than a 100*F/hr cooldown rate, PZR pressure 1840-2315 psig, PZR level less than 90%, SG level greater than 42% and SG pressure less than 1160 psig. All acceptance criteria were met during the transient.
While the instantaneous RCS Tave cooldown rate immediately after the initiation of the transient exceeded 100*F/hr, this rate slowed considerably and the RCS cooled down ler1 than 50*F over the 1/2 hour duration of this test.
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68 1
l
3.3.2 - LOAD SWING TEST (35% AND 75% POWER), NR-36 OBJECTIVE This test was performed to demonstrate the dynamic response of the Reactor Coolant System (RCS) and the Rod Control System to automatically bring the plant to steady state conditions following a 10% reduction in turbine load, and to a 10% increase in turbine load.
TEST METHODOLOGY With plant conditions stable at the appropriate power icvel (35%
or 75%) a 10% step load decrease was programmed into the turbine's Digital Electro-Hydraulic Control (DEHC) System at a rate of 2350 MW per minute. The 10% decrease was initiated and plant parameters were allowed to stabilize. After stabilization, a 10% step load increase was programmed into the DEHC System and initiated. Plant parameters were again allowed to stabilize.
During the course of the test, strip chart recordings of essential plant parameters were taken so that their response could be analyzed. The parameters monitored included RCS Tave, Tref, pressurizer pressure and level, steam generator pressure and levels, steam and feedwater flows, rod position and rod speed, OTWT and OPWT, reactor power, feedwater pump speed and discharge pressure and feedwater regulating valve demand position.
SUMMARY
OF RESULTS Tha load decreases and increases did not cause the reactor to trip nor the turbine to trip.
The steam generator safety valves and pressurizer safety valves did not lift during any of the load swings and nuclear power over/undershoot was less than 3%
in all of the executions. No manual intervention was required.
From the accelerometer data, it was determined that no damaging water hammer occurred during any of the load swings at nny of the power levels. See Tables 3.3.2-1 through 3.3.2-4 for additional details.
1 l
69 1
TABLE 3.3.2-1 10% LOAD DECREASE AT 35% POWER
SUMMARY
INITIAL FINAL CONDITION CONDITION Genecator I,oad (MWe) 375 240 Nuclear Power (%)
35.5 24.5 Tave Loop 1A (*F) 567 563 Tave Loop 1B (*F) 567 564 Tave Loop 1C ('F) 567 564 Tave Loop l') (*F) 565 563 Tref (*F) 567 563 Delta T Loop 1A (*4) 44 33.5 OPWI Loop 1A (%)
109 109 OTWI Loop 1A (%)
>150
>150 Pressurizer Pressure (psig) 2230 2250 Pressurizer Level (%)
35 31 Steam Generator Level Loop 1A (%)
64 64 Steam Generator Level Loop 1B (%)
67 67 Steam Generator Level Loop 1C (%)
65 65 Stsa.n Generator Level Loop 1D (%)
67 67 Steam Header Pressure (psig) 1038 1038 Steam Flow Loop 1A (pph) 1.3E6 0.9E6 Steam Flow Loop 1B (pph) 1.2E6 0.8E6 Steam Flow Loop 1C (pph) 1.lE6 0.7E6 Steam Flow Loop 1D (pph) 1.lE6 0.7E6 Feedwater Flow Loop 1A (pph) 1.2E6 0.8E6 Feedwater Flow Loop 1B (pph) 1.2E6 0.9E6 Feedwater Flow Loop 1C (pph) 1.1E6 0.7E6 Feedwater Flow Loop 1D (pph) 1.2E6 0.8E6 Feedwater Temperature Loop 1A (dF) 350 325 Feedwater Temperatare Loop 1B ('F) 350 324 i
Feedwater Temperature Loop IC (*F) 350 325 Feedwater Temperature Loop 1D (*F) 350 324 Feed Pump Discharge Pressure (psig) 1131 1125 Control Bank D Position (steps) 190 145 Control Bank C Position (steps) 228 228 l
l Boron Concentration (ppm) 833 833 l
70 l
l L _.
~
1 TABLE 3.3;2-2 10% LOAD-INCREASE AT 35% POWER
SUMMARY
INITIAL FINAL CONDITION CONDITION Generator Load (MWe) 240 380 Nuclear Power (%)
25 36 Tave Loop 1A (*F) 564 568 Tave Loop 1B ('F) 564 566 Tave Loop 1C ('F) 564 566 Tave Loop 1D (*F) 561 567 Tref (*F) 561 565 Delta T Loop 1A (%)
34 44 OPWT Loop 1A (%)
109 109 OTWT Loop 1A (%)
>150
>150 Pressurizer Pressure (psig) 2250 2230 Pressurizer Level (%)
31 38 Steam Generator Level Loop iA (%)
64 64 Steam Generator Level Loop 1B (%)
67 67 Steam Generator Level Loop 1C (%)
65 65 Stean Generator Level Loop 1D (%)
67 67 Steam Header Pressure (psig) 1040 1038 4
Steam Flow Loop 1A (pph) 0.9E6 1.3E6 Steam Flow Loop 1B (pph) 0.8E6 1.2E6 Steam Flow Loop 1C (pph) 0.7E6 1.lE6 Steam Flow Loop 1D (pph) 0.7E6 1.lE6 Feedwater Flow Loop 1A (pph) 0.8E6 1.lE6 L
Feedwater Flow Loop 1B (pph) 0.9E6 1.2E6 Fe3dwater Flow Loop iC (pph) 0.7E6 1.lE6 Feedwater Flow Loop 1D (pph) 0.8E6 1.lE6 Feedwater Temperature Loop 1A (*F) 325 350 l
Feedwater Temperature Loop 1B (*F) 324 349 l
Feedwater Temperature Loop 1C (*F) 324 349 Feedwater Temperature Loop 1D (*F) 324 350
' Feed Pump Discharge Pressure (psig) 1119 1113 I
Ccntrol Bank D Position (steps) 145 211 Control Bank C Position (steps) 228 228 i
Boron Concentration (ppm) 833 833 71 l
1
TABLE 3.3.2-3 10% LOAD DECREASE AT 75% POWER
SUMMARY
INITIAL FINAL CONDITION CONDITION Generator Load-(MWe) 880 760 Nuclear Power.(%)
74.5 63.7 Tave Loop 1A ('F) 580 577 Tave Loop 1B (*F) 579 577 Tave Loop 1C (*F) 579 576 Tave Loop 1D ('F) 578 575 Tref (*F) 578 576 L
Delta T Loop 1A (%)
78 69 OPWT Loop 1A (%)
108 108 OTWT Loop 1A (%)
140 150 Pressurizer Pressure (psig) 2230 2220 Pressurizer Level (%)
50 46 l
Steam Generator Level Loop 1A (%)
65 64 Steam Generator Level Loop 1B (%)
66 65 Steam Generator Level Loop IC-(%)
66 64 l
Steam Generator Level Loop 1D (%)
66 66 l
l Steam Header Pressure (psig) 993 1007 Steam Flow Loop 1A (pph) 2.9E6 2.5E6 Steam Flow Loop 1B (pph) 2.8E6 2.3E6 Steam Flow Loop 1C (pph) 2.8E6 2.4E6 Steam Flow Loop 1D (pph) 2.8E6 2.4E6 Feedwater Flow Loop 1A (pph) 2.8E6 2.4E6 Feedwater Flow Loop 1B (pph) 2.9E6 2.5E6 Feedwater Flow Loop 1C (pph) 2.9E6 2.5E6 Feedwater Flow Loop 1D (pph) 2.7E6 2.3E6 Feedwater Temperature Loop 1A (*F) 412 403 Feedwater Temperature Loop 1B (*F) 412 402 Feedwater Temperature Loop 1C (*F) 412 403 s
Feedwater Temperature Loop 1D (*F) 412 402 Feed Pump Discharge Pressure (psig) 1159 liS6 Controi Sank D Position (steps) 182 150 Control Bank C Position (steps) 220 228 Boron Concentration (ppm) 740 740 72
7._
= - _ _ _
TABLE 3.3.2-4 10% LOAD INCREASE AT 75% POWER
SUMMARY
INITIAL FINAL CONDITION CONDITION Generator Load (MWe) 760 890 Nuclear Power (%)
63.7 76 Tave Loop 1A ('F) 576 580 Tave Loop 1B (*F) 577 579 Tave Loop 1C ('F) 575 579 Tave Loop 1D (~F) 574 578 Tref (*F) 576 579 Delta T Loop 1A (%)
69 78 OPWT Loop 1A (%)
108 108 OTdT Loop-1A (%)
150 136 Pressurizer Pressure (psig) 2220 2230 F
Pressurizer Level (%)
47 52 Steam Generator Level Loop 1A (%)
64 64 Steam Generator Level Loop 1B (%)
65 66 Steam Generator Level Loop 1C (%)
64 65 Steam Generator Level Loop 1D (%)
66 66 Staam Header Pressure (psig) 1001 988 Steam Flow Loop 1A (pph) 2.5E6 2.9E6 Steam Flow Loop 1B (pph) 2.3E6 2.7E6 Steam Flow Loop 1C (pph) 2.4E6 2.8E6 Steam Flow Loop 1D (pph) 2.4E6 2.8E6 Feedwater Flow Loop 1A (pph) 2.4E6 2.8E6 Feedwater Flow Loop 1B (pph) 2.5E6 2.9E6 Feedwater Flow Loop 1C (pph) 2.5E6 2.9E6 Feedwater Flow Loop 1D (pph) 2.3E6 2.7E6 Feedwater Temperature Loop 1A (*F) 403 415 Feedwater Temperature Loop 1B (*F) 402 414 Feedwater Temperature Loop 1C (*F) 402 414 Feedwater Temperature Loop 1D (*F) 402 415 Feed Pump Discharge Pressure (psig) 1151 1155 Conttol Bank D Position (steps) 152 198 Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 745 745 73
)
3.3.3 - SHUTDOWN FROM OUTSIDE THE CONTROL ROOM, RC-35 OBJECTIVE Shutdown From Outside The Control Room verified that the unit could be taken from approximately 20% reactor power to Hot Standby conditions from outside the control room with a minimum shift crew. The potential to safely cool the unit to cold shutdown conditions from outside the control room was also demonstrated.
TEST METHODOLOGY The Unit 1 reactor was manually tripped from 21% reactor power.
Utilizing abnormal operating procedure 1BwCA PRI-5, Control Room Inaccessibility Unit 1, the Reactor Operator and Station Control Room Engineer (SRO) proceeded to the Remote Shutdown Panel (RSP) to assume remote control of plant equipment.
Cooldown then proceeded without incident. Upon completion, control of the plant was transferred back to the Main Control Room.
A standby operations staff remained in the Main Control Room throughout the test to assume unit control if needed.
SUMMARY
OF RESULTS Shutdown From Outside the Control Room to Hot Standby conditions was accomplished with the minimum shift crew and maintained for at least 30 minutes. The actual time recorded was 35 minutes.
The unit was cooled down to <350'F for the residual heat removal system to be placed into operation.
The Residual Heat Removal (RHR) system was placed in operation from the RSP and it demonstrated that the plant could be cooled at a rate of 28.9'F/ hour which meets the criterion of < 50*F/ hour.
The RHR system was able to cool the Reactor Coolant System (RCS) by 55'F which meets the criterion of demonstrating the ability to cool the RCS by 50*F using RHR. Minimum shift manning guidelines were maintained by those personnel performing functions for the test.
During performance of the test, three unexpected items occu. red. The 1A Reactor Trip Breaker local external trip if r.kage failed to function properly.
The 1B Reactor Trip Breaker (RTB) was used to perform the test and the 1A RTB has since been repaired. The ID Steam Generator Power Operated Relief Valve (S/G PORV) failed to open properly to aid in plant cooldown. The lA, 1B and 1C S/G PORVs operated pecperly to cooldown the plant and the ID S/G PORV hca since been repaired.
The 1B Auxiliary Feedwater Pump automatically started and was manually shut down following restoration of steam generator levels.
74
3.4 - INSTRUMENTATION AND CALIBRATION TESTING 3.4.1 - CALIBRATION OF STEAM AND FEEDWATER FLOW, W -31 OBJECTIVS The Calibration of Steam and Feedwater Flow test was performed to verify _that the feedwater flow and steam flow instrumentation was properly calibrated.
TEST METHODOLOGY With the plant stable at approximately 0%, 30%, 50%, 75%, and 90% power, plant parametern concerning steam and feedwater flows were recorded. These parameters included steam flows, feedwater flows, feedwater tempering flows, feedwater temperatures, and steam generator pressure. Differential pressure gauges were installed on the precision main feedwater flow venturies to accurately measure feedwater flows.
Steam generator blowdown was isolated for the duration of the test. The precision feedwater flow measurement was then compared to the electrical output of the steam.and feedwater flow transmitters and square root extractors. The acceptance of this startup test was based on the comparisons.
SUNMARY OF RESULTS During the. executions of this test a number of the comparisons failed to meet the stringent test acceptance criteria. These differences were either corrected through transmitter and instrument loop.ascaling_and recalibrations or determined acceptable b) engineering review. The test to be executed during the 100% power sequence will provide the firal calibration verifications based on actual full power flow data and additional data taken at lower power levels. Tests conducted at lower power power levels serve to scale steam and feedwator flow instrumentation progressively durir.g powe; ascension using extrapolated values.
75
3.4.2 - THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION 0%, 30%, 40%, 50%, 75%, and 90%) IT-32B/C/D OBJECTIVE The three basic objectives of this test series were to periodically determine thermal power using calorimetric data, collect control and protection instrumentation data at steady state power levels (statepoints), to verify and align WT instrumentation and to check Tave alignments. There w>ere no acceptance criteria for the tests in this series. Table 3.4.2 lists the various tests and intent of each.
TEST METHODOLOGY Statepoint data was collected at the approximate 0%, 30%, 50%,
75%, and 90% power levels while WT instrumentation alignment was done in the 75% sequence. Thermal power calorimetrics were performed at each execution mentioned above and at 40% power.
Calorimetric data included feedwater temperature and main feedwater flow venturi WP, steam pressure, atmospheric pressure, f 6edwater tempering flow and steam generator clowdown flow.
Three readings were taken within a 20 minute time period for each of the parameters to assure good quality of calorimetric l
calculation input values.
61.ignment of the WT process instrumentation was accomplished by plotting these values gathered at various power levels as a function of the respective calculated calorimetric power. A trend line fitted through the points was extrapolated to 100% power to predict the 100% WT and Tave values.
SUMMARY
OF RESULTS Power ascension testing between 0% power and 90% power progressed well with only minor instrument anomalies which were documented and resolved.
These out-of-range anomalies were in general due to iterative calibrations, hysteresis effects and an overly conservative initial scaling of full power WT to 51'F.
The four RCS loop full power WTs were rescaled as follows:
WT (*F)
LOOP 1A 56.9 LOOP IB 56.1 LOOP 1C 57.1 LOOP 1D 56.5 76 P
e.
t v
TABLE 3.4.2 Thermal Power Measurement and Statopoint Data Collection Sunnary
- TEST APPROXIMATE-TEST SEQUENCE THERMAL PWR STATEPOINT DATA INST. ALIGNMENT
'IT-32C 30%
0%
X
'IT-32C 30%
30%
X IT-32B' 50%-
40%
IT-32C 50%
50%
X IT-32D-75%
75%
X
'X 1T-32C 90%
90%
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77
3.4.3 - OPERATION!.L ALIGNMENT OF EXCORE NUCLEAR INSTRUMENTATION (DURING POWER ESCALATION) NR-34D OBJECTIVE This test was performed to verify that the excore nuclear instrumentation system was functioning per design and. capable of detecting, alarming and mitigating reactivity excursions.
TEST METHODOLOGY Selected parameters and alarms were evaluated monitored, and determined during various testing phases.
During Power Escalation the overlap between the SR and IR channels was determined. The-high level trip setpoints for the IR and PR channels were verified.
Finally, the PR detector response versus coce power was checked for linearity by aligning the channels with the calorimetric thermal power calculated in the thermal power measurement tests.
SUMMARY
OF RESULTS The setpoints of the IR channel and PR channel trips were verified as meeting all associated acceptance criteria. A minimum overlap of 1.5 decades was observed on all SR/IR and channel combinations.
Specifically, the overlaps for the four SR/IR channel combinations were observed to be more than 1.91
- decades, f
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3.4.4
- AXIAL FLUX DIFFERENCE INSTRUMENTATION CALIBRATION AT 50% OR 75% POWER, NR-35A OBJECTIVE This test was performed to assure that a linear relationship exists between the excore neutron detector currents and tha incore Axial Fiux Difference (AFD). Once established, this excore current /AFD relationship was uced to accomplish various calibrations of the excore channels, the OTWT AFD inputs, axial flux difference indications and plant process computer inputs.
TEST METHODOLOGY For the 50% power execution of this test excore detector current and flux map data at various Control Bank D positions previously obtained at approximately 50% power were used.
For the 75%
power execution of this test the same data was obtained during a planned, controlled, 75% power axial xenon transient where several flux maps were taken. The incore axial offsets determined from the flux map data were graphically compared to the excore channel normalized top and bcttom detector currents.
Equations were developed for the least squares fit lines through the observed data-points. These equations and their constants (i.e. slope, y intercept} were subsequently used to determine expected currents for calibrations, and voltages to be measured at various equipment test points. The measured outputs were then verified proper or adjusted as necessary to meet the expected values. The test performed a very detailed channel test and alignment partially using normal instrument calibration procedures and verified proper comp 4etion of the instrument procedures used.
SUMMARY
OF RESULTS The excore detector data and incore flux map results were successfully used in properly calibrating the OTWT inputs, process computer inputs and Axial Flux Difference indications at both the 50% and 75% reactor power levels.
Voltages to the pen recorders and computer points which were calculated from measured detector currents, AFD indications, and the OTWT setpoint inputs were verified to be within the required accuracies.
During test performance at 50% power several circuit boards were found to have circuit board links improperly installed. This occurrence was evaluated as having no adverse impact on plant safeti, the boards were corrected and there was no impact on test results.
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.. g 3_.4.5 REACTOR LOOSE PARTS MON _I_TORING, G-30 OBJECTIVE The test objective was to gather background noise frequency response data at 0%, 30%, 50%, 75%, and 90% reactor power. This data will be used-as a reference baseline for_ alarm setpoints and when analyzing suspected loose parts in the NSSS.
- TEST METHODOLOGY -
' At each power level a background noise recording of each of the 24 loose parts accelerometer channels was made.
SUMMA _RY OF RESULTS The data collected in the test will provide a baseline for each of the 24 accelerometer channels. Several sensor problems were identified during testing.
During the test performed in the initial critical sequence at 0%
power channels 2, 11 and 20 were discovered to be failed.
Channels 2 and 11 were repaired and background data was retaken for these channels at 0% power under an Action item Record.
Data was also not obtained for channel 20 during the testa conducted in the 30%, 50%, 75% and 90% test sequences. A separate Action Item Record has been written to verify that the appropriate background data will be retaken following repair of channel 20.
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c-3.4.6 - STARTUP ADJUSTMENTS OF REACTOR CONTROL SYSTEMS, RD-80 OBJECTIVE
'Tha objective of this test was to determine the Tave program which would result in the design steam pressure at full load and therefore optimum plant efficiency without exceeding pressure limitations for the turbine or temperature limits for the maximum allowable Tave. This was accomplished by making adjustments to the reference Tave (Tref) program and rescaling of turbine impulse pressure instrumentation as necessary.
TEST ME1HODOLOGY Data for this test was obtained from executions of the Thermal power Measurement and Statepoint Data Collection test, 2.47.82A-E at various power levels.
Data utilized for this test included RCS Tave, Thot, Tcold, steam generator pressure, and turbine impulse chamber pressure.
At 75% power, the available data was extrapolated to 100% to determine if any Tref or impulse pressure rescaling would be necessary. The same data will be reviewed at 100% power for any additional required rescaling.
If any scaling changes were made, additional plant data would be taken to verify the adequacy of the scaling.
SUMMARY
OF RESULTS At 75% power, Tave extrapolated to 585.8'F which was verified to be below the design maximun. ;f 588.4*F.
Steam generator pressure was extrapolated to 985 psia which matched the design value or 990 1 10 psia. Turbine impulse chember pressure was extrapolated to 792 psig at 120% power.
Tref did not require rescaling, however, turbine impulse chamber pressure was rescaled to approximately 792 psig at 120% power The original full range impulse pressure scaling was approximately 822 paig at 120% power.
Following the scaling changes, additional statepoint data was obtained at 75.7% power which verified the adequacy of the changes.
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3.4.7 - PROCESS COMPUTER VERIFICATION - THERMAL POWER, CX-80 OBJECTIVE, The Process Computer Verification of Thermal Power was performed to verify the ability of-the process computer to receive and reduce calorimetric data.
TEST METHODOLOGY The process computer verification thernal power was performed by performing a precision secondary plant calorimetric measurement executing while executing the computer calorimetric program and comparing the to outjuts.
SUMMARY
OF RESULTS The acceptance criteria for this test consisted of a 1 2% power agreement between the precision calorimetric performed by hand and the computer calorimetric power results.
The computer results failed to meet the acceptance criteria in comparison with the hand calorimetric results at 30%, 40% and 50% power.
During the course of power ascension the test was run at 30%,
40%, 50%, and 75% power. The 75% power computer results were within 2% power from the hand calorimetric and results generally improved with increasing power as the calorimetrics become more accurate at higher feedwater flows.
Based on the results of this test at 75% power, this computer program has been satisfactorily demonstrated, s
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3.4.8 - PROCESS COMPUTER VERIFICATION - INCORE THERMOCOUPLE, CX-81 OBJECTIVE
.The process Computer Verification of Incore Thermocouple was performed to verify the ability of the process computer to receive and reduce data from the incore thermocouples.
TEST MimiODOLOGY The process computer verification incore thermocouple was performed by taking a core exit thermocouple map from a flux map.
Thermocouple temperature distributions were than used to hand calculate fuel assembly relative powers based on enthalpy changes and to calculate core power tilts. These hand calculated thermocouple based results were compared against similar thermocouple calculations performed by the process computer.
SUMMARY
OF RESULTS The acceptance criteria for this test consisted of a 1 0.01 (10.1 at 0% power) agreement between the hand calculated values and the computer calculated values. The greater tolerance at 0%
power is needed due to the nearly non-existent core WT at 0%
power.
During the course of power ascension the test was run at 0% and 75% power. Acceptance criteria was met at both power levels. This computer program has been accepted as satisfactorily demonstrated.
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3.4.9 - PROCESS COMPUTER VERIFICATION - BORON FOLLOW, CX-82 OBJECTIVE The Process Computer Verification of Boron Follow was perfortred to verify the ability of the process computer to monitor Reactor Coolant System (RCS) Boron and Lithium Concentrations.
TEST METHODOLOGY The process computer verification boron follow was performed by taking RCS chemistry samples, analyzing them for boron and lithium and inputting these initial values into the computer program.
Following an RCS chemistry change samples were again taken and analyzed in the laboratory and these final results compared to the final values f rom the computer program.
SUMMARY
OF RESULTS The acceptance criteria for this test consisted of a 1 5% boron and lithium concentration agreement between the laboratory data and the computer data. The computer data failed to meet the 1 5% agreement criteria in comparison with the laboratory data for lithium at 30%.
This was attributed to the very low lithium concentrations where a 5% tolerance represents fractions of a ppm and the laboratory results are accurate to only ! 10%.
During the course of power ascension the test was run at 0%, 30%
and 50% power.
Based on the performance of the test at 50% this computer program has been satisfactorily demonstrated.
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3.4'.10 - PROCESS COMPUTER VERIFICATION - FLUX MAPPING,CX-83 OBJECTIVE The Process Computer Verification of Flux Mapping was. performed to verify the ability of the process computer to receive and reduce data from the incore movable detectors.
TEST METHODOLOGY The process computer verification flux mapping was performed by taking one pass of a flux map and comparing the voltages received at the process computer.versus the voltages recorded by a strip chart recorder at the movable detector panel.
For the selected pass the strip chart recordings for all six detectors were hand digitized using mylar overlays.. These digitized values were compared against the raw voltages recorded by the computer movable detector sof tware. - The outputs were then compared for selected parameters.
SUMMARY
OF RESULTS The acceptance criteria for this test consisted of a 3% of full scale voltage agreement between the hand digitized data and the computer data.
During the course of power ascension this test was run at 0%, 30%, 50% and 75% power. Acceptance criteria was met at all power levels. Based on these results the accuracy of this computer program has been satisfactorily demonstrated.
During performance of the test at 50% power, Incore Detector Drive E was not used in the flux mapping.
The acceptance criteria were satisfied for the other five detectors at the 50%
power plateau.
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REFERENCES
- 1) Braidwood Station Final Safety Analysis Report
- 2) Regulatory Guide 1.68, Revision 2
- 3) Braidwood Station Technical Specifications
- 4) Braidwood Station Operating Licent.; NPF-70
- 5) Braidwood Station Operating License NPF-72
- 6) WCAP 10935, Core Physics Parameters and Plant Operations Data for the Braidwood Generating Station Unit 1, Cycle 1
- 7) Westinghouse NSSS Startup Manual
- 8) Byron Unit 1, Cycle 1 Startup Report
- 9) Byron Unit 2, Cycle 1 Startup Report
- 10) Braidwood Unit 1, Cycle 1 Startup Report (Partial Report February 1989) h o
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p-CN Commonwestth Edison
(
_ ) One First Nabonal P' ara. Chicago. Ilknois Address Reply to: Post Omco Box 767
's Chicago, Illinois 60690 0767 May 31, 1988 MiositY Mn!50 TDL]t l.g ---k __ 2M j
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Mr. A. Bert Davis Regional Administrator Region III U.-S. Nuclear Regulatory Commission 799 Roosevelt Road
,_,j
,[, _~WCT Glen Ellyn, IL 60137 eg a
Subject:
Braidwood Nuclesr Power Station Unit 1 Cycle 1 Startup Report Supplement NRC Docket No. 50-456 References (a):
February 29, 1988 S.C. Hunsader letter to A.B. Davis
Dear Mr. Davis:
Reference (a) provided the Startup Report for Braidwood Unit 1, Cycle 1 in accordance with Technical Specification 6.9.1.3 that covered initial criticality. This letter provides the supplementary report covering the balance of Startup Testing in accordance with Technical Specification 6.9.1.3.
Very truly yours,
,d C WK S. C. Hunsader Nuclear Licensing Administrator
/klj att.
cc:
S. Sands - NRR NRC Document Control Desk p
A encl. Braidwood Unit 1. Cycle 1 Startup Report
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