ML20049J333
| ML20049J333 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 02/15/1982 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 8203150090 | |
| Download: ML20049J333 (400) | |
Text
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STARTUP REPORT 79' . ,, . O g jf ..[ tW3[fiS 9 ggg((Da l 3 %,g'[3 6 1 . i E POWER 1 g 1 l g QOSI I FEBRUARY 15,1982 0 YD g I e8 m 88P a888 m P PDR
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l Duke Power Company McGuire Nuclear Station Unit No. 1 I Docket No. 50-369 License No. NPF-9 I I "^""" ~"" I I I I I I I I I February 15, 1982 I I
TABLE OF CONTENTS Section Page LIST OF TABLES iv LIST OF FIGURES vii
1.0 INTRODUCTION
1.0-1 2.0
SUMMARY
2.0-1 3.0 ITIAL FUEL LOADING 3.0-1 4.0 TESTING PRIOR TO INITIAL CRITICALITY 4.0-1 4.1 REACTOR COOLANT SYSTD1 FLOW TEST 4.1-1 4.2 REACTOR COOLANT SYSTEM FLOW COASTDOWN TEST 4.2-1 4.3 RESISTANCE TEMPERATURE DETECTOR (RTD) BYPASS LOOP FLOW 4.3-1 VERIFICATION TEST 1 4.4 PRESSURIZER FUNCTIONAL TEST 4.4-1 4.5 MOVABLE INCORE DETECTOR FUNCTIONAL TEST 4.5-1 4.6 FULL LENGTil ROD DRIVE TIMING TEST 4.6-1 I 4.7 ROD POSITION INDICATION ALIGNMENT CllECK 4.7-1 4.8 ROD DROP TIME MEASURDIENT 4.8-1 5.0 INITIAL CRITICALITY 5.0-1 6.0 ZERO POWER PilYSICS TESTING 6.0-1 6.1 ISOTilERMAL TDiPERATURE COEFFICIENT OF REACTIVITY MEASURDIENT 6.1-1 6.2 BORON ENDPOINT MEASUREMENT TEST 6.2-1 6.3 ZERO POWER FLUX MAP TEST 6.3-1 I 6.4 ROD WORTil AND BORON WORTil DETERMINATION 6.4-1 6.5 STUCK ROD WORTil MEASURDIENT TEST 6.5-1 6.6 PSEUDO EJECTED ROD TEST 6.6-1 7.0 NATURAL CIRCULATION TESTING 7.0-1 7.1 NATURAL CIRCULATION WITil SIMULATED LOSS OF ALL 7.1-1 5 ONSITE AND OFFSITE AC POWER 7.2 NATURAL CIRCULATION VERIFICATION TEST 7.2-1 I 7.3 7.4 NATURAL CIRCULATION WITil SIMULATED LOSS OF OFFSITE POWEB EFFECT OF STEAM CENERATOR ISOLATION ON NATURAL CIRCULATION 7.3-1 7.4-1 8.0 POWER ESCALATION TESTING - CORE PERFORMANCE / PLANT RESPONSE 8.0-1 8 8.1 UNIT LOAD STEADY STATE 8.1-1 8.2 TilERMAL POWER OUTPUT MEASUREMENT 8.2-1 8.3 CORE POWER DISTRIBUTION TEST 8.3-1 I I
I TABLE OF CONTENTS Section Page 8.4 PSEUDO ROD EJECTION TEST 8.4-1 8.5 DOPPLER ONLY POWER COEFFICIENT VERIFICATION 8.5-1 8.6 UNIT LOAD TRANSIENT TEST 8.6-1 I 8.7 LOSS OF OFFSITE POWER TEST 8.7-1 8.8 PRELIMINARY INCORE AND NUCLEAR INSTRUMENTATION SYSTDI 8.8-1 CORRELATION TEST I 8.9 BELOW BANK ROD TEST 8.9-1 8.10 TARGET FLUX DIFFERENCE CALCULATION 8.10-1 8.11 UNIT LOSS OF ELECTRICAL LOAD TEST 8.11-1 8.12 DYNAMIC ROD DROP TEST 8.12-1 8.13 INCORE AND NUCLFAR INSTRUMENTATION SYSTDI CORRELATION 8.13-1 9.0 POWER ESCALATION TESTING - INSTRUMENTATION AND CONTROLS 9.0-1 9.1 PRESSURIZER SAFETY RELIEF VALVE ACOUSTIC LEAK 9.1-1 DETECTION TEST i 9.2 STARTUP ADJUSTMENTS OF REACTOR CONTROL SYSTEM 9.2-1 9.3 CALIBRATION OF STEAM AND FEEDWATER FLOW 9.3-1 INSTRUMENTATION AT POWER I 9.4 9.5 9.6 STEAM DUMP CONTROL SYSTDI DYNAMIC TEST NUCLEAR INSTRUMENTATION SYSTDI CALIBRATION AT POWER 9.4-1 9.5-1 STEAM CENERATOR LEVEL CONTROL TEST 9.6-1 lgW 9.7 9.8 PRESSURIZER PRESSURE AND LEVEL CONTROL TEST ROD CONTROL SYSTEli AT POWER TEST 9.7-1 9.8-1 9.9 OPERATIONAL ALIGNMENT OF PROCESS TDiPERATURE INSTRUMENTATION 9.9-1 10.0 POWER ESCALATION TESTING - MISCELLANEOUS 10.0-1 10.1 INITIAL ROLL OF Tile MAIN TURBINE AND GENERATOR 10.1-1 10.2 LOSS OF CONTROL ROOM TEST 10.2-1 I 10.3 FEEDWATER PIPING TilERMAL EXPANSION TEST 10.3-1 10.4 STEAli GENERATOR WATER liAMMER TEST 10.4-1 10.5 EFFLUENT RADIATION MONITOR TEST 10.5-1 I 10.6 LOOSE PARTS MONITORING 10.6-1 10.7 NEUTRON NOISE ANALYSIS TESTING 10.7-1 10.8 I 10.9 10.10 10.11 REACTOR COOLANT SYSTEM PRDIARY LOOP FLOW MEASURDiENT TEST PROCEDURE TO VERIFY PRIMARY AND SECONDARY CilDilSTRY RADIOLOGICAL SHIELD SURVEY 10.8-1 10.9-1 10.10-1 PIPING DYNAMIC RESPONSE FOLLOWING LOSS OF E'.ECTRICAL 10.11-1 LOAD TEST I 10.12 COLD LEG ACCUMULATOR PIPING ACOUSTIC DilSSION TEST 10.12-1 10.13 COMPARISON OF BORON CONCENTRATION - CHDiICAL ANALYSIS 10.13-1 VERSUS TllE BORONOMETER 11.0 POWER ESCALATION TESTING - TO BE PERFORMED 11.0-1 11.1 1 11.2 LOSS OF OFFSITE POWER TEST EFFL.UENT RADIATION MONITOR TEST 11.1-1 11.2-1 11.3 RADIOLOGICAL SilIELD SURVEY 11.3-1 11
1 , TABLE OF CONTENTS , Section Page 11.4 TEST PROCEDURE TO VERIFY PRlf!ARY AND SECONDARY CllE}ilSTRY 11.4-1 11.5 COLD LEG ACCUMULATOR PIPING ACOUSTIC EMISSION TEST 11.5-1 11.6 UNIT LOAD STEADY STATE 11.6-1 11.7 NEUTRON NOISE ANALYSIS TESTING 11.7-1 11.8 DOPPLER ONLY POWER COEFFICIENT VERIFICATION 11.8-1 11.9 UNIT LOAD TRANSIENT TEST 11.9-1 I 11.10 LARGE LOAD REDUCTION TEST 11.10-1 11.11 UNIT LOSS OF ELECTRICAL LOAD TEST 11.11-1 11.12 PIPING DYNAMIC RESPONSE FOLLOWING LOSS OF ELECTRICAL 11.12-1 l LOAD TEST 8 11.13 TURlilNE TRIP TEST 11.13-1 11.14 STEAM GENERATOR MOISTURE CARRYOVER TEST 11.14-1 I I I I g I I I I i lii i 1
I LIST OF TAllLES - I Table 2.0-1 Monthly Summary - August 1981 2.0-2 fionthly Summary - September 1981 2.0-3 !!onthly Summary - October 1981 5 2.0-4 Monthly Summary - November 1981 2.0-5 Monthly Summary - December 1981 2.0-6 Monthly Summary - January 1982 3.0-1 Fuel Loading Summary 4.1-1 Calculated Reactor Coolant Loop Flows I 4.3-1 llo t Leg RTD Bypass Loop Data 6.0-1 Reactivity Computer Checkout 6.1-1 1sothermal Temperature Coefficient Summary 6.2-1 IlZP Boron Endpoint Test Results 6.3-1 Summary of Zero Power Flux Map Configurations 6.4-1 lizP Integral Bank Worths and Differential Baron Worths 6.6-1 Maximum Peaking Factors and Worth of Ejected Rod 6.2-1 Primary Power Determination 8.2-2 Secondary Power Determination
- 8. 2 -3 lies t Estimate Thermal Pouer Determination 8.3-1 Core Power Distribution Tests - Sammary of Results 8.4-1 Results of the Pseudo Ejected Rod (D-12) Worth Determination 8.4-2 Excore Detector Response - Pseudo Rod Ejection Test 8.4-3 Axial Offset Ratios from Assemblage Powers - Rod D-12 at 167 Steps 8.4-4 Tilting Factors from Assemblage Powers for Entire Core lleight - Rod D-12 at 167 Steps I
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E I LIST OF TABLES Table 8.4-5 Nuclear Peaking Factors for Enthalpy Rise for Assemblages in the Power Normalization - Rod D-12 at 167 Steps 8.4-6 Axial Offset Ratios from Assemblage Powers - Rod D-12 at 228 Steps 8.4-7 Tilting Factors from Assemblage Powers for Entire Coro Height - Rod D-12 at 228 Steps 8.4-8 Nuclear Peaking Factors for Enthalpy Rise for Assemblages in the Power Normalization - Rod D-12 at 228 Steps 8.5-1 Results of Doppler Only Power Coefficient Verification 8.6-1 10% Load Decrease from 30% Power 8.6-2 10% Load Increase from 30% Power 8.6-3 10% Load Decrease from 75% Power 8.6-4 10% Load Increase from 75% Power 8.6-5 Control System Setpoint Data I 8.8-1 Preliminary Incore and Nuclear Instrumentation Systems Correlation Results 8.9-1 Excore Detector Data Sheet - Below Bank Rod Test E 8.9-2 Axial Offset Ratios from Assemblage Powers - Rod H-4 at 228 Steps j s 8.9-3 Tilting Factors from Assemblage Powers for Entire Core Height - Rod H-4 at 228 Steps 8.9-4 Nuclear Peaking Factors for Enthalpy Rise for Assemblages in the Power Normalization - Rod H-4 at 228 Steps 8.9-5 Axial Offset Ratios from Assemblage Powers - Rod H-4 at 205 Steps 8.9-6 Tilting Factors from Assemblage Powers for Entire Core Height - Rod H-4 at 205 Steps I I V I
I LIST OF TABLES Table 8.9-7 Nuclear Peaking Factors for Enthalpy Rise for Assemblages in the Power Normalization - Rod 11-4 at 205 Steps 8.9-8 Axial Offset Ratios f rom Assemblage Powers - Rod 11-4 at 0 Steps 8.9-9 Tilting Factors from Assemblage Powers for Entire Core lleight - Rod 11-4 at 0 Steps I 3.9-10 Nuclear Peaking Factors for Enthalpy Rise for Assemblages in the Power Normalization - Rod 11-4 at 0 Steps 8.13-1 incore and Nuclear Instrumentation Systems Correlation 9.5-1 Nuclear Instrumentation System Overlap Data 9.6-1 Bypass Valve Controller Settings (Typical of Four) 9.6-2 Control Valve Controller Settings (Typical of Four) 9.6-3 Peedwater Pump Speed Controller Settings I 9.7-1 Pressurizer Pressure and Level Control System Final Controller Setpoints 9.9-1 100% Full Power AT Settings I 10.6-1 Loose Parts Monitoring Areas 10.9-1 Typical Chemistry: Reactor Coolant 10.9-2 Typical Chemistry: Secondary Systems 10.9-3 Major Transients Experienced During Power Escalation au Exhibited in Blowdown Water Quality I 10.13-1 Boron Determination - Chemical Analysis Versus Boronometer Readings During Initial Criticality 10.13-2 Boron Determination - Chemical Analysis Versus Boronometer 5 Readings During Zero Power Physics Testing 10.13-3 Boron Determination - Chemical Analysis Versus Boronometer I Readings During Steady State Conditions of Zero Power Physics Testing 10.13-4 Boron Determination - Chemical Analysis Versus Boronometer 8 Readings, January 1982 I vi I
I LIST OF FICURES i i Figure 2.0-1 Power llistory - August 1981 2.0-2 Power llistory - September 1981 2.0-3 Power llistory - October 1981 2.0-4 Power llistory - November 1981 2.0-5 Power llistory - December 1981 2.0-6 Power llistory - January 1982 2.0-7 Reactor Core Map - Excore Detector Locations 2.0-8 Reactor Core Map - Control Rod Locations 2.0-9 Reactor Core Map - Movable Incore Detector Thimble Locations 2.1-10 Exit Thermocouple Locations 3.0-1 Core Loading Sequence - Initial Fuel Loading 3.0-2 Core Loading Pattern 3.0-3 Core Assembly Insert Pattern 4.2-1 NC Flow Coastdown Test - 1/4 Coastdown Transient 4.2-2 NC Flow Coastdown Test - 4/4 Coastdown Transient 4.2-3 NC Flow Coastdown Test - 1/3 Coastdown Transient 4.2-4 NC Flow Coastdown Test - 3/3 Coastdown Transient 4.2-5 Reanalysis of Worst Case 4/4 Coastdown - Nuclear Power and Core Flow Data 4.2-0 Reanalysis of Worst Case 4/4 Coastdown - llaat Flux and DNBR Data 4.4-1 Pressure Response to Opening of Both Pressurizer Spray Valves 4.4-2 Pressure Response to Activation of all Pressurizer seaters g 5 vii I
I LIST OF FIGURES Figure 4.8-1 Typical A and B Rod Drop Trace 4.8-2 Typical A Plus B Rod Drop Trace I 4.8-3 Sample of Combined Traces l l W 4.8-4 Rod Drop Time Tabulation - Temperature - 120 F Pressure - 360 PSIG % Flow - 0 I 4.8-5 Rod Drop Time Tabulation - Temperature - 160 F Pressure - 355 PSIG % Flow - 100 4.8-6 Rod Drop Time Tabulation - Temperature - 168 F to 180 F Pressure - 355 PSIG % Flow - 100 4.8-7 Rod Drop Time Tabulation - Temperature - 550 F Pressure - 2240 PSIG % Flow - 0 4.8-8 Rod Drop Time Tabulation - Temperature - 555 F Pressure - 2233 PSIG % Flow - 100 5.0-1 ICRR vs. Time - N31 5.0-2 ICRR vs. Time - N32 5.0-3 ICRR vs. Water Addition - N31 5.0-4 ICRR vs. Water Addition - N32 6.0-1 Determination of Nuclear Heat 6.1-1 Isothermal Temperature Coefficient of Reactivity Control Bank D at 188 Steps Withdrawn (Heatup) 6.1-2 1sothermal Temperature Coefficient of Reactivity Control Bank D at 188 Steps Withdrawn (Cooldown) 6.1-3 Isothermal Temperature Coefficient of Reactivity Control Bank D in, Control Jank C at 220 Steps Withdrawn (Heatup) 6.1-4 1sothermal Temperature Coefficient of Reactivity Control Bank D at 22 Steps Withdrawn (Cooldown) 6.2-1 Boron Endpoint Measurement - All Rods Out Configuration 6.3-1 HZP, ARO, BOL Flux Map, Map A - Relative Powers 6.3-2 HZP, D in, BOL Flux Map, Map B - Relative Powers viii
I LIST OF FIGURES Figure I 6.3-3 IlZP, D & C in., BOL Flux Map, Map C - Relative Powers 6.3-4 IlZP, D, C, & B at Insertion Limits - BOL Flux Map, I Map D - Relative Powers 6.3-5 llZP, D, C & B at Insertion Limits, D-12 Ejected, BOL Flux Map, Map E - Relative Powers I 6.4-1 Dif ferential and Integral RCC Bank (RCCA) Worth - Control Bank D 6.4-2 Differential and Integral RCC Bank (RCCA) Worth - Control Bank C 6.4-3 Differential and Integral RCC Bank (RCCA) Worth - Control Bank B I 6.4-4 Differential and Integral RCC Bank (RCCA) Worth - Control Bank A 6.4-5 Differential and Integral RCC Bank (RCCA) Worth - Shutdown Bank E 6.4-6 Differential and Integral RCC Bank (RCCA) Worth - Shutdown Bank D 6.4-7 Differential and Integral RCC Bank (RCCA) Worth - Shutdown Bank C 7.1-1 S/G A, B, C &D Steam Pressure vs. Time 7.1-2 NC Loop liighest Average Temperature & NC System Wide Range Pressure vs. Time i 7.1-3 S/G A, B, C & D Narrow Range Level vs. Time 7.1-4 Aux. Feedwater Flow to S/G A, B, C & D vs. Time 7.2-1 Natural Circulation at 3% Power, ARO, BOL Flux Map - Relative Powers 7.2-2 NC System Wide Range Pressure vs. Time 7.2-3 Reactor Coolant Loop 1A Wide Range Temperatures vs. Time 7.2-4 Reactor Coolant Loop 1B Wide Range Temperatures vs. Time ix I
I LIST OF FIGURES Figure 7.2-5 Reactor Coolant Loop 1C Wide Range Temperatures vs. Time 7.2-6 Reactor Coolant Loop 1D Wide Range Temperatures vs. Time 7.3-1 Reactor Coolant Loop L\ Wide Range Temperatures vs. Time 7.3-2 Reactor Coolant Loop 1B Wide Range Temperatures vs. Time 7.3-3 Reactor Coolant Loop 1C Wide Range Temperatures vs. Time 7.3-4 Reactor Coolant Loop 1D Wide Range Temperatures vs. Time 7.4-1 Reactor Coolant Loop 1A Wide Range Temperatures vs. Time 7.4-2 Reactor Coolant Loop 1B Wide Range Tem aratures vs. Time 7.4-3 Reactor Coolant Loop 1C Wide Range Temperatures vs. Time 7.4-4 Reactor Coolant Loop 1D Wide Range Temperatures vs. Time 8.1-1 Steam Generator Level vs. Power (Loop 1) 8.1-2 Steam Generator Level vs. Power (Loop 2) 8.1-3 Steam Generator Level vs. Power (Loop 3) 8.1-4 Steam Generator Level vs. Power (Loop 4) 8.1-5 Steam Generator Level vs. Power (Loop 1) 8.1-6 Steam Generator Level vs. Power (Loop 2) 8.1-7 Steam Generator Level vs. Power (Loop 3) 8.1-8 Steam Generator Level vs. Power (Loop 4) 8.1-9 Steam licader Pressure vs. Power 8.1-10 Feedwater Flow vs. Power 8.1-11 Feedwater Temperature vs. Power 8.1-12 T ,T and T "# cold ""* P ) i I 8.1-13 Thot, T , and T vs. Power (Loop 2) 8.1-14 T ,T g, and T cold I vs- P wer (Io P 3) I
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I LIST OF FIGURES Figure I 8.1-15 T hot, T e, and T cold vs. Power (Loop 4) 8.1-16 Differential Temperature vs. Power (Loop 1) 8.1-17 Differential Temperature vs. Power (Loop 2) 8.1-18 Differential Temperature vs. Power (Loop 3) 8.1-19 Differential Temperature vs. Power (Loop 4) 8.1-20 Pressurizer Level vs. Power I 8.3-1 RCS Total Flowrate vs. R and R - Four Loops In Operation 2 8.3-2 Rod Bow Penalty as a Function of Burnup 8.3-3 RCS Total Flowrate vs. R and R2- Four Loops In Operation 8.5-1 Example of Load Increase and Load Decrease I 8.9-1 Below Bank Rod Test - Typical at Power Rod Worth Curve 8.12-1 Dynamic Rod Drop Test Results 8.13-1 Excore Calibration Procedure Sequence 9.5-1 Nuclear instrumentation System Overlap 10.4-l Steam Generator Water Hammer Instrumentation and Flow Path 10.7-1 McGuire Unit 1 Neutron Noise Analysis - Norm CPSD Magnitude vs. Frequency in Hertz 10.7-2 McGuire Unit 1 Neutron Noise Analysis - Coherence vs. Frequency in Hertz 10.7-3 McGuire Unit 1 Neutron Noise Analysis - Phase in Degrees vs. Frequency in Hertz I I xi I
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1.0 INTRODUCTION
The McGuire Nuclear Station, located on the southern end of Lake Norman in North Carolina, consists of a Westinghouse four loop, pressurized I water reactor rated at 3411 MWt and a Westinghouse turbine-generator rated at 1180 MWe. The design and fabrication of the initial core is also supplied by Westinghouse. Construction started at the McGuire site on June 23, 1971 under an exemption granted by the Atomic Energy Commission (AEC). Construction permits were granted on February 28, 1973. Issuance of a fuel loading and Zero Power Physics Testing license by the Nuclear Regulatory Commission (NRC) took place on January 23, 1981. Fuel loading commenced l I on January 28, 1981. On June 12, 1981, the NRC issued a 5% power licence to the McGuire Nuclear Station. On July 8,1981, the NRC issued an operating full power license for McGuire Unit 1. Initial Criticality was achieved on August 8, 1981. Following the successful completion of Zero Power Physics Testing and Natural Circulation Testing, Power Escalation Testing was started on September 11, 1981. Further testing was performed at the following power levels: Power Level (%) Date First Achieved 10 September 11, 1981 I 20 30 50 September 13, 1981 September 21, 1981 November 3, 1981 75 November 26, 1981 90 January 5, 1982 I 100 January 13, 1982 This report is prepared in accordance with the requirements of Technical I Specification 6.9.1 and addresses the results of startup testing from Initial Fuel Loading through testing at the 100% full power level, with I regard to McGuire Nucicar Station, Unit 1. At this time, testing at the 100% power level has not been completed. At the completion of all Power Escalation Testing, a supplement to this report will be filed. I I I 1.0-1 I
I I 2.0 SUte1ARY Significant startup milestones and events for McGuire Nuclear Station Unit 1 are listed below. Receipt of Fuel Loading and Zero Power Testing License January 23, 1981 Start of Fuel Loading I Completion of Fuel Loading January 28, 1981 February 2, 1981 Receipt of Low Power (5%) License June 12, 1981 I Receipt of Full Power License Initial Criticality Start of Zero Power Physics Testing July 8, 1981 August 8, 1981 August 8, 1981 i l I Completion of Zero Power Physics Testing Start of Natural Circulation Tests Completion of Natural Circulation Tests-August 20, 1981 August 29, 1981 September 1, 1981 l I Power Escalation to 10% September 11, 1981 Power Escalation to 20% September 13, 1981 Power Escalation to 30% September 21, 1981 Power Escalation to 50% November 3, 1981 I Power Escalation to 75% Commercial Operation Power Escalation to 90% November 26, 1981 December 1, 1981 January 5, 1981 Power Escalation to 100% January 13, 1982 McGuire Unit 1 startup and power escalation testing as addressed in this report are summarized below. I (a) Initial Fuel Loading I Initial Fuel Loading began on January 28, 1981, and was completed on February 2, 1981. The major problems encountered were failure of the three temporary neutron detectors used for fuel loading and failure of the air pump on the spent fuel pool side fuel I transfer conveyor system. These problems resulted in delays of 25 and 11 hours respectively. In general, however, fuel loading proceeded in a safe and smooth manner. (b) Test Prior to Initial Criticality I Several tests were performed after fuel loading but prior to initial criticality. These included the Reactor Coolant System Flow and Flow Coastdown Tests, the RTD Bypass Flow Test, and the Pressurizer I Functional Test. In addition, Rod Drive Timing tests, Rod Position Indication checks and Rod Drop Time tests were performed. A thorough checkout of the Movable Incore Detector System was also made l at this time. No major problems were encountered except for the 1 l W Reactor Coolant System Flow Test where the total core flow was less l than the acceptance criteria. This problem is still unresolved at this tine. I I "~' I
I (c) Initial Criticality Initial Criticality was achieved on August 8, 1981, at 0928 hours with Control Bank D at 120 steps withdrawn. All other control rods were fully withdrawn. The Reactor Coolant System boron cancen-tration was 1316 ppm. These conditions were within the acceptance criteria for initial criticality. The acceptance criteria was a reactivity equivalent of 50 ppm of Control Bank D at 145 steps and a boron concentration of 1301 ppm, g 3 (d) Zero Power Physics Testing Zero Power Physics Testing began on August 8, 1981, at 1120 hours and completed on August 20, 1981, at 1900 hours. Parameters measured included isothermal temperature coefficients, boron endpoints, and l rod worths. In addition, flux maps were taken at various rod W configurations. The J oint of adding nuclear heat was determined to start at y x 10 amps on the reactivity computer picoammeter, and 2.0 x 10 amps on the two intermediate range channels. The all rods out moderator temperature coefficient at beginning of life was +1.36 pcm/ F based on an isothermal temperature coefficient of .57 pcm/ F and a Doppler Coefficient of -1.93 pcm/ F. (e) Natural Circulation Testing Natural Circulation Testing began on August 29, 1981, at 0435 hours and completed on September 1, 1981, at 0722 hours. It was performed to demonstrate satisfactory reactor operation in the natural circu-lation mode. During these tests, the reactor was run at 1 to 3% F.T. to simulate decay heat conditions. The reactor coolant pumps were tripped and the resulting plant conditions were monitored. All tests were completed with no difficulty and all acceptance criteria were met. (f) Power Escalation Testing The power escalation testing program was designed to provide initial startup data in areas of core physics, controls and instrumentation, ! plant transients, chemical control, and behavior of the plant's radiation environment. Testing began on September 11, 1981, at 0020 hours and reached the 100% full power plateau on January 13, 1982. I The 100% power level testing has not been completed yet, pending i resolution of the reactor coolant low flow problem and the steam generator tube vibration problem. l During the 10% power testing, the turbine generator was first put on line. This occurred on September 12, 1981, at 0246 hours. At 20% power, the Loss of Control Room Test initial 7 failed due to l the inability to reset the Auxiliary Feedwater Sy tem to regulate I flow to the steam generators. This problem was ccrrected and the test was rerun successfully. At 30% power, the first core physics l tests were performed. These tests included flux mapping, rod i 2.0-2 I
I ejection tests, and Doppler Only Power Coef ficient tests. The physics tests as well as various instrument and control tests continued up through the 90% full power level with no significant I problems. At 75% and 90% full power, tests were made to determine the
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Reactor Coolant System Flow Rate. Results of testing again indi-
- W cate possible low flow in the system. This problem has not been resolved at this time.
In order to verify the full power capability of McGuire Unit 1 and l the performance of the steam generators, the NRC issued Amendment No. 11 to the Facility Operating License NPF-9. This amendment I concerning the low reactor coolant system flow permitted operation above 90% rated thermal power for no more than 48 hours. The unit spent approximately 22 hours at 100% thermal power during which time testing was performed. Throughout power escalation, radiation surveys were performed. I No unexpected radiation levels were encountered. In addition, neutron noise baseline data was obtained for the reactor. Tables 2.0-1 through 2.0-6 provide a detailed monthly summary of operations during power escalation testing for McGuire Nuclear Station Unit 1. Figures 2.0-1 through 2.0-6 show the monthly power history for McGuire Unit 1 during power escalation testing.
- Figures 2.0-7 through 2.0-10 show general information for McGuire Unit 1 excore dete< or locations, control rod locations, incore thimble loca-tions, and enermocouple locations referred to in this report.
I I I I I I I
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l l I l MONTHLY
SUMMARY
McGuire 1 August 1981 1 , Initial criticality was achieved on August 8, 1981. Zero power physics gl testing commenced on the same date and was completed on August 20, 1981, 5! with no significant problems encountered. The low power test program was begun on August 29, 1981, and completed on September 1, 1981. Again, no significant problems were encountered. ' i Listed below is a sequence of significant events for the remainder of i the month. ! 8-16-81 0558 Manually initiated reactor trip per Zero Power Physics j { Test Controlling Procedure g, 8-17-81 0701 g1 Manually initiated reactor trip due to Urgent Failure l Alarm for SD-A power cabinets; bad batch of fuses caused power failures , 8-18-81 0416 Manually initiated reactor trip per Zero Power Physics Test Controlling Procedure 8-18-81 1716 Manually initiated reactor trip per Stuck Rod Worth ; i Measurement Test l 1 8-31-81 0712 Low steam pressure automatic reactor trip during Natural ' Circulation testing. Main Steam Isolation valves were opened before the l, ' steam pressure had completely equalized across the W valve. As the pressure equalized, the rate calcu-lator (x 10 amplifier) initiated the trip signal. , I: 3l I I. I: I Table 2.0-1 I! l- - - - -.. - - --- - - ..-
I I MONTilLY
SUMMARY
McGuire 1 September 1981 Unit 1 Low Power Test program (Natural Circulation) was completed on Sep-tember 1, and the Unit was taken to Cold Shutdown. Listed below is a sequence of significant events for the remainder of the month. 09-01-81 Unit in Cold Shutdown for maintenance and preparation through for Power Escalation Testing 09-10-81 09-11-81 0030 Reactor critical, began Power Escalation Testing. 0040 Began esca.lation to 10% F.P. 2300 Reached 10% F.P. 09-12-81 0246 Closed generator breakers, began generating power 09-13-81 1650 Stabilized at ind'cated 20% F.P. 09-14-81 0113 Manual Reactor Trip per Loss of Control Room Test I 0130 Safety Injection on low steam line pressue due to feedwater control system malfunction 09-17-81 0050 Reactor critical I 0255 Generator breakers closed, =8% F.P. 0405 Began escalation to 20% F.P. 0523 Manual Reactor Trip per Loss of Control Room Test 09-18-81 1128 Reactor critical 1346 Generator breakers closed, 210/. F.P. 09-19-81 0015 Stabilized at indicated 30% F.P. 09-20-81 2300 Began power reduction to e20% F.P. to perform testing 09-21-81 0200 Stable at 20% F.P. 0630 Began power escalation 0910 Stable at 30% F.P., began initial operation of feedwater heaters Table 2.0-2
I 09-22-81 2015 Began power reduction to <20% F.P. to repair / adjust feedwater control valves. 2330 Stable at 16% F.P. 09-23-81 0001 Began power escalation 1530 Stable at 30% F.P. 09-26-81 0000 Began power reduction to repair drain valve on steam chest. 0100 Generator breakers open. 0428 Reactor shutdown, began cool down for maintenance on valves. During the month of September the generator was on line for 236 hours 57 minutes, producing a gross of 43,259 FNH and a net of 15,930 FMI. The net core burnup for the month was 2.26 EFPD. I I I I I I I I I I Table 2.0-2 (Continued) I
I MONTilLY
SUMMARY
McGuire 1 October 1981 I Unit 1 power escalation testing was in progress at the 30% power level. The unit was shutdown from the first to the fifth, and from the tenth to l the twenty-first, to repair a leaking primary system valve. Listed below is a sequence of significant events for the month. 10-1-81 Cold shutdown to repair valves and instruments. Ileat through up to 230 F. 10-4-81 I 10-5-81 1739 Reactor critical 2043 Generator on line 10-6-81 0207 Steady at 30% power, began testing I 1845 Inadvertent valve operation caused condenser vacuum loss with reactor power transient to 35% F.P. 10-8-81 0815 Began shutdown to repair steam drain line 0854 Generator off line 0915 Zero power 1340 Leak repaired, began power escalation 1500 Generator on line 1650 Stable at 30% power 10-9-81 0441 Condenser vacuum leak with reactor power transient to 33% F.P. I 10-10-81 0350 Began power reduction to repair feedwater regulator valve which stuck cused I 0554 Cenerator off line 10-11-81 I through 10-20-81 Shutdown to repair feedwater valves, primary system valve, and repair condenser boot seal. 10-21-81 0455 Reactor critical 0709 Cenerator on line I Table 2.0-3 I
I 1138 Steady at 30% power 1415 Decreased power to repair instrument tube leak and un-stick feedwater valve
! 2310 Steady at 30% power j 10-22-81 0743 Power swing for test I
1204 Power swing for test 10-24-81 0318 Turbine and Reactor trip caused by turbine hydraulic fluid ll leak up 2042 Reactor critical 2240 Generator on line 10-25-81 0215 Power decrease to open stuck feedwater regulator valve and investigate high NC pump bearing temperature 1556 Steady at 30% power 10-26-81 0036 Reactor trip on low steam generator water level; instru-ment malfunction 0800 Generator back on line 1414 Steady at 30% power 10-29-81 1911 Began series of small power transients for testing control i systems 10-30-81 1146 Reactor trip on low steam generator level, instrumentation malfunction 2040 Reactor critical 2114 Reactor trip on low steam generator level, insufficient l auxiliary steam for FWP = 10-31-81 Hot standoy while working on stuck containment isolation valve. During the month of October, the generator was on line for 316 hours 23 minutes, producing a gross of 75,302 MWH and a net of 43,660 MWH. The net core burn-up ' for the month was 3.62 EFPD. I I, Table 2.0-3 (Continued) I I;
I MONTilLY SURIARY McGuire 1 November 1981 . f l Unit 1 power escalation testing continued at the 50% power plateau, through the 14th. The unit was shutdown at that point to inspect the tubes in steam generator A. The unit was then brought to the 75% power plateau and testing resumed, continuing through the end of the month. Listed below is a sequence of significant events for the month. 11-1-81 0340 Reactor Critical 0516 Generator on line 1333 Began 4 cycles of 10% load reduction and return to power for the Unit Load Transient Test I 11-2-81 2320 Feedwater Pump Recirc. valve failed open, caused partial loss of load 11-3-81 2215 Moisture Separator Rehm.ter safety valve lifted causing rapid load reduction of 27% F.P. 11-5-81 0014 Reduced load 5% F.P. to swap feedwater pumps 11-6-81 1606 Load reduction of 3% F.P. due to swapping KG pumps 1830 Slow load reduction due to xenon build-up in core 11-7-81 0917 Pressurizer relief valve on sampling line lifted 11-9-81 1020 3% F.P. load reduction for test 1037 Reactor trip. Faulty relay allowed both feedwater pumps to trip during a periodic test. 1400 Reactor critical 1515 Generator on line 11-11-81 1519 Began shutdown when a containment isolation valve caused a loss of bearing cooling water for the NC pumps. Tripped reactor manually when pumps began to overheat. 1745 Roar ter a rtical 1850 Generator on line I Table 2.0-4 I
1 l 11-13-81 0552 Il Began series of 3% F.P. load reductions for testing 3 11-15-81 1630 Reactor trip per test procedure 2040 Reactor critical 2230 Reactor trip per test procedure 11-16-81 Unit shutdown to inspect steam generators through 11-23-81 11-24-8L 2311 Reactor critical 11-25-81 0223 Cencrator on line 11-30-81 1713 Began series of 10% F.P. load reductions for testing 2225 Began load reduction when feedwater pump bearings began to overheat. Power variations occurred while attempting to recover feedwater pumps and maintain core flux balance. ' i l During the month of November, the generator was on line for 481 hours W 21 minutes, producing a gross of 268,420 MWil and a net of 241,061 MWil. The 3 l net core burn-up for the month was 10.22 EFPD. ' l i i I I; I I i l I I I I Table 2.0-4 (Continued) I I-
I ' MONTilLY
SUMMARY
McGuire 1 December 1981 l Unit 1 power escalation testing at the 75% FP plateau was in progress on the 1st and 2nd. During the Loss of Offsite Power test, the Turbine Generator Governor valves over controlled, and caused a plant trip on underfrequency to the Reactor Coolant Pumps. Subsequently, the generator hydrogen cooler tubes failed and flooded the generator with cooling water during prepara-tions to return to power. The unit was down for the remainder of the month, returning to power January 1,1982. During the two days of operation, the gcccrator was on line for 45 hours I 36 minutes, producing a gross of 35,476 FMI and a net of 18,088 FMI. net core burn-up for the month was 1.04 EFPD. The I I I I I I I I I I I I Table 2.0-5
I MONTilLY
SUMMARY
McGuire 1 January 1982 Unit I returned to operation on the 2nd, after the generator was dried out. l
=
Testing was completed for the 75% and 90% F.P. plateaus, and 24 hours of the 100% F.P. testing was completed. On the 15th, the unit began normal power operation at 50% F.P., and continued at this level through the end of the month. Listed below is a sequence of significant events for the month. 1/02/82 0605 Reactor critical 0740 Cencrator on line 1/03/82 1732 Reactor trip due to Lo-Lo S/G level. MXII motor control center tripped and shut off power to main feedwater pump controls.
=2230 I
Reactor critical 1/04/82 ~0330 Generator on line 1/05/82 0330 Reached 90% F.P., began testing 0902 Reactor trip caused by erroneous trip signal from SSPS during Periodic Test i 1 1420 Reactor critical l 1552 Generator on line 1/09/82 1435 Reactor trip due to turbine trip, both caused by reversed polarity on a test pot installed to adjust governor valves 1/10/82 =0130 Reactor critical
=0215 Generator on line l 1/11/82 1206 Reactor trip with safety injection when steam pressure l
instrument line froze up and simulated a steam line break. l Frozen instrument lines on both steam and feedwater systems = l had caused problems from the startup on the 10th until the trip. l Table 2.0-6 I
I 1/11/82 2230 Reactor Critical
~2315 Generator on line 1/13/82 1505 Reached 100% power, began testing I 1/15/82 0842 Manually tripped Reactor Coolant Pump from 50% F.P.
hi-hi vibration alarm. This caused a low-flow reactor trip. Vibration monitor had malfunctioned. for
=1330 Reactor critical
=1415 Generator on line 1/18/82 Power increased to 75% F.P. for demand peak and returned to 50% F.P.
1/26/82 Power modulated several times to allow set-up/ calibration of turbine Deli system. I NOTE: Power level currently restricted to 50% by the vendor until Steam Generator tube vibration problems are resolved. , During the month of January, the generator was on line for 666 hours 28 minutes, producing a gross of 450,940 MWit and a net of 424,022 MWil. The total core burn-up is now 32.69 EFPD, 15.56 EFPD for the month. I I I I I I I Table 2.0-6 (Continued) I
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- - - - . , - - - - - - - , - - - - _ ,..%. - - , -,-_----r - - . -
i I lI Reactor Core Map Excore Detector Locations I Source Range N 31 Intermediate Range N 35 I Power Range N 41 Power Range N43 180 E 135 o 225 l 3 i h l QUADRANT 4 QUADRANT 1 5 Spare l Spare i 270 l 90 I QUADRANT 3 QUADRANT 2 s 315 45 I Power Range N 42 Power Range N 44 ! Source Range N 32 Intermediate Range N 36 North Figure 2.0-7 I
Reactor Core Map l Control Rod Locations E (53 Control Rods) R P N M L K J 11 C F E D C B A 1 2 SA-4 CB-4 CC-1 CB-1 SA-1 GR-2 GR-2 GR-1 GR-1 GR-1 3 SD-4 SB-4 SB-1 SC-1 GR-1 GR-2 GR-1 GR-1 SA-4 CD-2 SE-1 CD-1 SA-1 4 GR-1 GR-2 GR-1 GR-1 GR-2 SC-4 SD-1 5 GR-1 GR-1 CB-4 CC-4 CA-1 CC-1 CB-1 GR-1 GR-2 GR-1 GR-2 GR-2 7 SB-4 SB-1 270 GR-1 GR-2 CC-4 SE-4 CA-2 CD-3 CA-1 SE-2 CC-2 N GR-1 GR-1 GR-2 GR-2 GR-2 GR-1 GR-1 SB-3 SB-2 9 GR-2 GR-1 CB-3 CC-3 CA-2 CC-2 CB-2 10 GR-2 GR-2 GR-1 GR-2 GR-1 11 SD-3 SC-2 GR-1 GR-2 12 SA-3 CD-2 SE-3 CD-1 3A-2 GR-2 GR-1 GR-1 3h-2 1R-1 13 SC-3 SB-3 SB-2 SD-2 GR-1 GR-1 GR-2 GR-1 14 SA-3 CB-3 CC-3 CB-2 SA-2 GR-1 GR-1 GR-1 GR-2 GR-2 15 0 XX - Y XX - Bank Name; Y - RCC No. GR - A A - Group Number S - Shutdown Bank C - Control Bank Figure 2.0-8
E I Reactor Core Map Movable Incore Detector Thimble Locations (58 Thimbles) R P N M L K J 11 C F E D C B A C-7 B-4 2 B-8 A-2 F-2 1! A-9 C-9 D-8 F-3 4 A-4 D-5 E-3 5 A-10 D-7 C-10 C-3 B-6 F-5 B-3 A-5 A-7 6 7 E-4 D-1 B-5 E-7 8 C-2 F-9 C-8 E-6 E-8 F-1 E-9 D-3 270 9 F-8 C-1 A-8 F-7 I 10 A-3 CP B-1 11 D-2 D-10 C-5 B-10 B-2 12 E-1 F-6 A-6 13 E-5 D-4 B-9 C-4 14 B-7 C-6 E-2 D-6 5 F-4 A-1 15 0 1 I Figure 2.0-9
I Exit Thermocouple Locations (65 Thermocouples) R P N M L K J 11 G F E D C B A T-53 T-48 T-43 2 T-62 T-21 T-16 T-11 T-7 3 T-50 T-49 T-39 T-34 4 T-26 T-22 T-12 T-3 5 T-63 T-54 T-44 T-35 6 T-31 T-27 T$7 T-8 T-4 T-1 7 T-64 T-59 T-50 T-40 g T-32 T-23 T-18 T$3 T-5 270" 9 T-65 T-55 T-45 T-36 10 T-33 T-28 T-19 T-9 T-2 11 T-60 T-51 T-41 T-37 12 T-29 T-24 T-14 T-6 13 T-61 T-56 T-46 4 T.-42 T-38 14 T-30 T-25 T-20 T-15 T-10 15 T-57 T-52 P-47 0"
- NOTE: These thermocouples are located in the upper head.
I Figure 2.0-10 I
3.0 INITIAL FUEL LOADING Core loading operations began at 2329 hours on January 28, 1981, with the installation of fuel assembly C4, containing a primary source, into core 5 location 11-15, and ended with the installation of fuel assembly C23 into core location M-14 at 1926 hours on February 2, 1981. A total elapsed time I of 115.95 hours was required to load the 193 fuel assemblies. The average assembly loading time was 36 minutes per assembly. Factoring out delays due to two major problems encountered during fuel loading (temporary detector problems and fuel transfer system problems), the total core loading Ii time of 80.0 hours is estimated. This corresponds to an average assembly loading time of 25 minutes per assembly. Figure 3.0-1 shows the initial loading sequence for McGuire Unit 1. prior to the start of fuel loading and throughout core loading operations, reactor coolant circulation and water temperature was maintained using the Residual lleat Removal System by taking suction from Loop C llot Leg and i discharging to all four Cold Legs. The boron concentration in the Reactor Coolant System and the Residual IIcat Removal System was maintained above the Tech Spec limit of 2000 ppmb and below 2150 ppmb to ensure greater 'I than 2 couiiu; ner second on the source range instrumentation. Transfer tube boron concex ration was also maintained above the 2000 ppmb limit. Initial Reactor Vessel boron concentration was 2066 ppmb. Residual lleat [ Removal Loop boron concentration was sampled and analyzed every eight hours is during core loading operations. The average boron concentration was 2081 ppmb with a range between 2061 and 2092 ppmb. Transfer tube boron concentration was also sampled and analyzed every eight hours. The average transfer tube I boron concentration was 2089 ppmb, with a range between 2014 and 2118 ppmb. The Residual Heat Removal temperature, taken every eight hours, averaged 115 F with a range between 112 and 120 F. I Four significant problems were encountered during the loading sequence. These are described as follows: (a) After the first two assemblies had been loaded, it was discovered at 0200 on January 30, 1981 that containment integrity had never been properly established, when a containment isolation valve and an associated vent valve were found open to atmosphere. They were immediately closed. (b) 9 Several times air was automatically added to the containment due to low pressure in the Reactor Building. A flow path through the Containment Purge and Ventilation System corrected the pressure problems and ensured all air was being filtered. (c) Containment humidity was very high during fuel loading and sweating was evident on the surface of the containment walls. Containment I Ventilation systems could not lower building humidity. I 3.0-1 I i
. . . - _ . - - - ~ . . - - .- - --__ - ._. _ . - . - _ - . _ . - - _ - . . _ - _ - . _ - . _
t ) (d) Radiat ion Monitor IDIF39(L), Containment Gas Monitor, went , into alarm st.ite several times during fuel loading. This caused 4 a Containment Evacuation Alarm. Problems with this monitor stem f rom particulate buildup on radiation monitor 1DiF38 causing it j to alarm. In order to clean IEMF38, its chart paper was turned to fast speed. This caused a spike in the check source circuitry ll
=
of IEMF39(L) causing it to alarm. Installation of a capacitor across the timer clutch coil filtered off the spikes and corrected the problem. New fuel was delivered to M;Cuire Nuclear Station during the period of July 12, 1978 to October 17, 1978. The 17 x 17 array fuel complete with i inserts was inspected and placed in dry storage in the Unit One Spent Fuel pit. This shipment of fuel contained 193 assemblies including the secondary source assemblies, but did not include the two primary source rods. A survey of these assemblies indicated approximately 3 - 5 mR, B, y on contact. l E I On July 23, 1980, the two primary source rods were delivered to the McGuire 5 site and inserted Californium 252 (CF i g )the appropriate assemblics. The source material was g WhenmanfacguredonNovember30, 1978, the source strengths were approximately 8.1 x 10 neutrons /second. When these primary sources were inserted into ghe core on January 28, 1981, the source ' utrengths were approximately 4 x 10 neutrons /second. The dose from ' i these source rods when they were in the fuel mast was measured on the fuel handling crane to be 20 mR neutron and imR 8, y. E I Fuel llandling was accomplished by following written, approved procedures, 5 t l In general it proceeded smoothly. There were a vari m y of minor problems ; which affected the fuel transfer equipment but there was only one major l
- delay in this area. The air pump on the spent fuel pool side conveyor l
system failed and had to be replaced. The equipment vendor had a spare pump on site and this saved a lot of cime. The transfer canal had to be drained to do the work and thin resulted in a total delay in fuel loading ' of 11. hours. A summary of the fuel loading sequence complete with the problems encountered is shown on Table 3.0-1. The operations group at the station was responsible for the handling of the f ue.1 Staffing of personnel consisted of three crews of seven people each, wirking on eight hour shifts, 24 hours a day. Each crew was made , up of one SHO (Senior Reactor Operation) in the Control Room in charge of l fuel loading. Also an R0 (Reactor Operator), a crane operator, and an assembly number verifier were assigned to both the Reactor Building and the Spent Fuel Pool. One representative from Westinghouse and one from hl W Stearnes Rogers were present at all times; however, these personnel I worked two twelve hour shifts. 3.0-2 I I I.
Throughout fuel loading, neutron count rate was monitored using two permanent nuclear instrumentation source range channels, N31 and N32, with readouts in the Control Room. There were no problems experienced with this instru-mentation with the exception of a few "high flux at shutdown" alarms which I required the trip setpoint to be readjusted. An audible count rate from the source range instrumentation was heard both in the Control Room and ,l the Containment. Three temporary neutron sensitive detectors (Reuter Stokes lM B n mp rti nal Counters) were installed in the reactor vessel prior I to10fuel loading. These detectors were purchased by Duke and were used with Duke-built detector housings and associated Duke supplied electronics. .l Prior to fuel loading, this equipment was checked for proper functioning l j of electronics using a radioactive source, as well as being submerged under-water to check for watertightness of detector housings. While this r l initial checkout turned up no abnormalities, it was found that upon loading of the first two source assemblies, the temporary detectors would not function properly as no signals were received on the readout equipment. Sub- )l l IW t sequent inspection of the inside of the detector housings showed water seepage through the top of the detect 3r housing was the problem. i After several attempts to make the housings watertight, it was decided to l use the Westinghouse temporary incore detector system which had been shipped to the site. This problem caused fuel loading to be delayed for a period of 25 hours while the Westinghouse equipment was flown to Charlotte, North Carolina and brought to the site. The Westinghouse equipment consisted of three Br neutron detectors and all associated electronics for obtaining i diditalandgraphicalreadoutsofneutroncount rate. After the equipment was inserted into the reactor vessel, it was checked out and found to be working perfectly. There were no other electrical problems with these temporary detectors during fuel loading. Readouts for these three lI i temporary detectors were in the Containment Building. New background and baseline readings were taken using the new detectors and fuel loading proceeded. 5 Staf fing of personnel f or data analysis and monitoring of neutron count rates consisted of three crews of four people working eight hours each, j'k M 24 hours a day. The crews included a Shif t Coordinator Engineer in the
' Control Room, a Reactor Building Engineer, a Control Room Engineer, and a i Data Analyst in the Containment Building. These people were responsible for taking and analyzing all data. In addition, there were two Westinghouse l'
j Physicists on shift working two twelve hour days. These people were here i to assist fuel loading and answer any questions that would arise during fuel loading. i !I 1 i 3.0-3 4
I As stated earlier, neutron count rate was monitored throughout fuel loading using the two permanent source range detectors, N31 and N32 and three g 3 temporary source range detectors. Each detector was used to record initial background before fuel loading started (six five-minute counts). Following the loading of the initial nucleus of eight assemblies, initial baseline . count rates were obtained for each detector. New baseline data (three five-minute counts) was taken throughout fuel loading following any primary source assembly move or any detector move. Following the loading l of the initial nucleus of eight assemblies, count rates were taken for each 5 assembly loaded (three one-minute counts) and inverse count rate ratio (ICRR) was calculated. All data was verified acceptable by using the CHI squared method of statistical analysis. Data acceptability was performed smoothly and quickly by using a calculator program written specifically for use during fuel loading. A plot of ICRR vs. the number of fuel assemblies loaded was plotted for each detector. Results indicate that detector outputs were as expected in response to movement of source assemblies or loading toward or away from a given detector. One thing that was not initially expected was the 5a low values of ICRR for the temporary detectors. Throughout fuel loading there was no renormalization of ICRR and this led to very low ratios. Throughout fuel loading the detectors and associated electronics worked perfectly and no problems were experienced with this equipment. There were, however, two minor problems with this system that were caused by l human error. On February 2, 1981, at 0802 hours, power from both N31 and = N32 source range monitors was interrupted for about 30 seconds which led to a loss of indication from these monitors for this amount of time. This was caused by Instrument and Electrical personnel who, while trying to g W troubleshoot an unrelated alarm, inhibited all input signals to the Solid State Protection System. Fuel loading had been stopped prior to this g time to work on fuel transfer system problems. Another problem concerned the sensitivity of the temporary detectors. Inadvertent bumping of the 5 temporary detectors by personnel caused the count rates to change signi-ficantly necessitating the retaking of baseline data for the affected detectors. Upon completion of loading fuel assembly 190, Temporary Detector A was removed from tlw core. Temporary Detector B was then removed from the ' core after loading assembly 191, and finally Detector C was removed after loading assembly 192. Fuel loading was then completed with the loading of assembly 193 on February 2, 1981 at 1926 hours. The temporary detector equipment was then returned to Westinghouse. The average count rates upon completion of core loading on Source Range Channels N31 and N32 were 5.87 cps and 3.46 cps respectively. I i 3.0-4
l, lt ' I At the completion of fuel loading, core verification took place. It was i started on February 2, 1981, at 2345 hours and completed on February 3, 1981,
, at 1051 hours. It was initially planned to tape core verification using video tape equipment. Ilowever, there were numerous p oblems with the l
i camera equipment and the video tape equipment - the most significant being jl noise problems that prevented a clear tape from being taken. finally decided to forgo the video tape method and proceed with core If was lW verification using a camera only and recording the insert and assembly l numbers by hand on a data sheet. This method proceeded smoothly with no l further problems. The final configuration of the assemblies and the inserts
- in these assemblies are shown on Figures 3.0-2 and 3.0-3.
in summary, initial fuel loading at McGuire Nuclear Station, Unit 1, was iI l completed in approximately five days and in general proceeded smoothly and efficiently. l l l ) ,I 4 a i 4 i 1 3.0-5
I I Fuel Loading Summary ; 1 Date Ti .ne l I 1/28/81 2329 First assembly loaded into core (C4 into location H-15, Primary Source Assembly) 2350 Second assembly loaded into core (C30 into location J-1, ; Primary Source Assembly) 2355 Duke built temporary detectors not responding (Source Range OK), fuel loading suspended g< g! 1/29/81 0042 Radiation monitor IFRF39 alarms - Containment Evacuation Alarm 0040 Arrange to get Westinghouse detectors f rom Pittsburgh, PA 0735 Discovered water in Duke temporary detectorn 9, 1406 Containment pressure of .6 psig - venting with containment g air release and Addition System 3 1/30/81 0030 Westinghouse detectors and data acquisition system arrive and detectors are placed in the core ! l 0042 Radiation monitor 1DIF39 alarms - Containment Evacuation Alarm ' 0215 Containment Integrity discovered breached when PT/1/A/4200/02C l was run. IVX40, IVX41 found open on sample blower line. g' l Reclosed immediately. E l l l 0400 Fuel loading resumed; loaded Assembly 3 0700 Loaded Assembly 9 I: 0701 Reshuffle fuel 0807 Containment Evacuation Alarm due to high flux at shutdown setpoint exceeded on N31 (10 cps) 0852 Reset N31 nigh flux at shutdown trip setpoint from 10 cps to 30 cps 1019 Reset N31 high flux at shutdown trip setpoint from 30 cps to 17 cps after reshuffle of fuel. 1048 Loaded Assembly 10 ' 1258 Loaded Assembly 15 Table 3.0-1 (1 of 6)
; Date Time l
t 'g 1309 Neutron count rate jumps by a factor of 6 on temporary g Detector B after loading Assembly 15. Fuel loading stopped to discuss this with Westinghouse. Determine jump expected
- due to core coupling.
4 l 1359 Loaded Assembly 16 1 1530 Loaded Assembly 21 l 1531 Relocating Detector A 1649 Loaded Assembly 22 1907 Loaded Assembly 26 l jW 1908 Minor problems with manipulator crane latching mechanism ! in Spent Fuel Pool 2050 Debris discovered on core plate at core location L-5 2120 Loaded Assembly 27 )5 j 2347 Paint chip removed from core plate Location L-5 1/31/81 0007 Loaded Assembly 28 0606 Loaded Assembly 44 0607 Relocate Detector C 0706 Minor problems with containment manipulator crane hoist 3 0729 Loaded Assembly 45 0740 Stearns-Rogers assists in minor repairs to manipulator crane 0820 Loaded Assembly 46 0904 Loaded Assembly 48 0905 Intermediate load trip out on containment manipulator crane hoist 0918 Loaded .*.ssembly 49 0927 Reset N-31 high flux at shutdown trip setpoint from 17 cps to 30 cps 0943 Loaded Assembly 50
$ 1 Table 3.0-1 I (2 of 6)
I Date Time ! 0956 Radiation monitor 1DiF39(L; alarms - Containment Evacuation Alarm during loading of Assembly 51 = 1005 Loaded Assembly 51 1007 Radiation monitor 1E!!F39(L) declared inoperable i 1015 Loaded Assembly 52 1119 Loaded Assembly 56 5 1131 Radiation monitor 1DiF39(L) declared operable 1132 Loaded Assembly 57 1441 Loaded Assembly 69 1506 Relocation of fuel 1650 Loaded Assembly 70 1715 Loaded Assembly 71 g i 1728 Relocation of Detector C 5-1809 Loaded Assembly 72 2248 Loaded Assembly 88 2320 While lowering Assembly 89 into core location 11-13, there was an intermediate low load alarm. Assembly 89 was then lowered out of position until it was six inches from the core plate. Then it was moved to the correct location - and lowered to the core plate. 2328 Loaded Assembly 89 2/1/81 0027 Loaded Assembly 92 l 0050 Relocation of Detector C 0111 Loaded Assembly 93 0202 Loaded Assembly 96 0219 Relocation of fuel 0237 Loaded Assembly 97 0444 Loaded Assembly 104 < Table 3.0-1 I (3 of 6)
i 1 i Date Time !g 0452 Radiation monitor 1DIF39(L) alarmed - Containment Evacuation j Alarm f 0501 Radiation monitor 1FMF39(L) cleared j 0533 Loaded Assembly 105 I 0555 Relocation of fuel 0608 Reset N-32 high flux at shutdown trip from 17 cpm to 30 cps 0659 Loaded Assembly 106 'l 0937 Loaded Assembly 117 i
- 0949 Relocation of Detector A I
1012 Loaded Assembly 118 1401 Loaded Assembly 135 1406 Fuel loading suspended due to automatic opening of valve IVQ6 to relieve negative containment pressure 1432 NRC in Atlanta, GA, is contacted to discuss automatic relief of containment pressure. No Tech Specs violated. Fuel loading started. 1436 Loaded Assembly 136 2034 Loaded Assembly 157 2056 Relocation of fuel 2129 Loaded Assembly 158 2142 Reset N32 high flux to shutdown trip setpoint from 30 cps to 17 cps 2146 Loaded Assembly 159 2254 Loaded Assembly 163 2335 Problem : develop with Spent Fuel Pool Conveyor System Air l8 Driven Motor. Decision is made to replace motor with Westinghouse spare. Secured fuel loading and pumped fuel transfer canal water to Refueling Water Storage Tank, 5 I I Table 3.0-1 (4 of 6) 1 - . _ _ --_ _ _ - _ _ . -. . . - . . _ . . - - - - - __ - _ - - -
Date Time I 2/2/81 0801 Lost N31 and N32 detector voltages and signals for about 30 seconds. Problem occurred when Instrument and Electrical personncl inhibited SSPS Train A and B inputs to work on . an unrelated alarm problem. 1020 Operating containment purge system to equalize containment pressure with outside atmosphere. kW 1049 Resumed fuel loading. Loaded Assembly 164. 1105 Loaded Assembly 165 1120 Problems with manipulating crane in containment as Assembly 166 is lowered. The scale weight decreases but there is no trip. Bridge realigned. 1143 Loaded Assembly 166 1246 Loaded Assembly 171 1301 Detector C is inadvertently bumped and moved causing the counts to decrease by a factor of three. Redoing baseline on Detector C. 1323 Loaded Assembly 172 1423 Loaded Assembly 177 1433 liad a Train A blackout with diesel generator lA start, g Train B operable. Fuel loading continuing. 5 1441 Loaded Assembly 178 1448 Stopped fuel loading to determine the reason for the Train
, A blackout.
1555 The cause of the Train A blackout was Instrument and Electri-cal work on Breaker 1TAS which tripped the normal supply feeder breaker. , 1600 Loaded Assembly 179 1833 Loaded Assembly 190 1844 Removed Detector A from core 1850 Loaded Assembly 191 E Table 3.0-1 (5 of 6) I
Date Time 1900 Removed Detector B from core 1902 Loaded Assembly 192 l 1913 Removed Detector C from core 1926 Loaded Assmebly 193 2345 Started core verification 2/3/81 1051 Completed core verificacion i I I I I I 5 I I I I I Table 3.0-1 I (6 of 6)
W -2 -u- - - - - - - -- - - - - - 1 Core Loading Sequence Initial Fuel Loading
- =
i Denotes temporary detector; A, B or C as applicable II l \3 . ) Source bearing assembly !I -l Z ^ssc=bly loaded during current loading sequence step #z l 1 y Assembly previously loaded into permanent position I ./ Assembly previously loaded into temporary position i !3 Not yet loaded Secondary Source l NOTE: Lines with arrows depict relocation of assemblies or lg detectors as well as final removal of the three temporary [g detectors and loading of the last three assemblies Figure 3.0-1 (1 of 9)
core Loading Sequence Initial Fuel Loading Ob lg go I 0: R P N ti - L K J H G f E D C 8 A Oa a l i I I I I ' 1 5 y 4 , / I N> 2 3 2 ; h l5 6 7 h 3 .9$ 8 g 4 5-6- I 7-90 - -I - 270 l 9-10 - II-12 13 14
. I 15 3 oo ,
. N32
- Core loading Sequence Step I to 9a I I Figure 3.0-1 g (2 of 9) g B
*l 1
B i N31 . I Core Loading Sequence Initial Fuel Loading IB Oo l R P N M L K J H G F E D C B A I I I I s I MM 2 ee / W@Wx @ 3 f / Y[/ hh 4 ,
/
'- /
B 6- f 7- [ I 900 [ -
- 270 g_ .
'o - / ! =
B II - l 12 [
'3 l
14 9c
/ ,,
15 {\Y 9b I Oo
, 8 Core Loading Sequence Step 9b to 9d l
Figure 3.0-1 (3 of 9)
I core Loa ing Sequence Initial Fuel Loading iB O I R P N M L K J H G F E D C B A I I i i i i i I
- 15 12 10 /[ 18 21 2 ,
@@ 16 2 13(((( h :19
'22 h21t 3 k 17 14 11 h,[/g 20 23 g 4 '441s . 24 25 26 27 28 29 30 5- \' 31 32 33 34 35 36 37 6- @ 44 38 39 l 40 41 42 43 g
7- 45 46 47 48 49 50 51 90 71b 52 53 54' 55 56 57 58 - 270o l 9- 64 65 10 - @ 59 66 60 67 61 62 68 59a _
'63 70 71a I
ri -
/ \k 3 12 j7 \ 59b .
13 59[\ \ , ia SM N\ g 15 h Oo I 8 : Core Loading Sequence Step 10 to 71b
. I l .
rigure 3.0-1 (4 of 9) I
l B l Core Loading Sequence Initial Fuel Loading 18 0 l R P N M L K J H G F E O C B A l l i l i i MMERM l 2 ! i@ MM M v/AR @ I s WMffMK6;M
< c W//MsM@@
l s- i MMMM s-V/A M M M M 7- V/lfllffflf/llf/)f/ll/g 9O* ' f f l/lf f f /l f /llll f/ l l - 270* s-GY/AfM/AD M
\O - hf/lflffflllhfflllg i1 - 72 73 74
) 75 76 77 78 I i 12 '
79 80
; 81 82 83 84 85 .
13 86, k 87 88 ,89, 90 91 92a 14 92b 93 94 95 964 96b 97 98
^'
15 I h Oo . 8 Core Loading Sequence Step 72 to 98 s Figure 3.0-1 (5 of 9)
Cora Lording S qu:nca I Initial Fuel Loading ON31 - g 18 0 5 R P N M L K J H G F E D C 8 A g WM/AWAN 2
@ VffM7/A % D E a
V/#ffsPMAM s-hhhhy g
.- war //ma 7-fffllffll}Yh h 9Oo /
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'2 WMME5Wrd I o GREGMM I
>< o VM VA VB A % M
; 'IS 99 100 101 102 103 104 105a 1
Oo . l O I Core i.oading Sequence Step 99 to 105c l . I l 8 l l n aure 3.o-1 l i (6 of 9) l l I
I 1 E N31 Core Loading Sequence Initial Fuel Loading 1800 R P N M L K J H G F E D C B A I I I e i I .1 1 2 DMWJV/lMMVi o a . WFf4%W9MVA \ S~ ff/hhh/f/h//h V Yl17t > s- WAVfff/ffA R \
! 7- ](([ [ h 110113h16 107 i
' -*~ N # #2 N '" "' ~'
9- ! ((fj//4h109 112 115 ll7a l0 ' f ff 30G //) h n- kN1 Vf##4V/A M i2 fffffffffffll Y f f l3 ffh fIffffffff f f4 { /ffllffffffff I5 fffflllfffflkff/l \ Oo 8 - Core Loading Secuence step 105 to 117b - Figure 3.0-1 (7 of 9)
I Core Loading Sequence ON31 g Initial Fuel Loading 18 0o W s R P N M L K J H G F E D C 8 A g i i I e i i I g I " ' . A 2 -- @ [/[ [ [ [ (( [ [ 147 149 3 (((([// [/ h143 146 148 l hk[/h139 14N145 4 /(((( 5- N I(([,/f[/g' (([(h 135 13k 141 [/ 144 6- E(((((( (( h134 135 137 140 7- 158 156 153 150[h [ [ [ [ [ / E[MI/[/[f- [' ' 90o 159 l57a 154 151(((([ ((f((((((Jk((/((h - 270o l 9~ 160 155 152 ?ff f f f f f f fff f h h f f f} fllg 10 ~~ 57 3 / 115 '119 121 124 11 - ( (((( (((( ,.T20122 125 128 l 12 '(((((((((( [ fI 12D126 129 13 /((/[ / 127 130 132 14 @ (( (( k[ h h 131 133 g VfffffffffkfkYh IS-oo 8 118 to 160 Core Loading Sequence Step I E
,1 .m 3 (8 of 9)
I
Core Loading Sequence Init.ial Fuel Loading
~
N31 l 191b 192a y ISO 5 R P N l ( M L K J H G F E O C B A I I l l l 8 I I
\ ?7 , 9 l
2 o 1gogl [ [ h k[jh h,hIh f _ h h ff ((/) //f,[ 3 189 188 185 ,h, l 4 i 187 184 181 rf f, j (([, '/) / ((f[ h 5- 186 183 180,176 hh'f/,9ffff/)ff/f/))fh{1[ff , 6- 182179177176ffh,f,h/ f/h'Y/llf)lVlf9lll,)YlY///ffhff/ffff/, { ) / 7- ffh,fl}ff/l fff),fhfffffl/h) l 7 I 90* k lflhhffflfll,/,llff,f/hff,/fff/)& f ~27O* s- % 9,- SVARMMWM/fM///Mf4%MW)nV& l l0 ~ 167 164 162 161ff,yff/hfh)/f/flfffhflff II ~ 1 71 168 165 16ff/fffk,ffl,/hff/)lllflff//h I2 17216h 16flllflh/ff,fflflf//flffffll/), t3 .174173 17Nfl '$' f)l?hffffff,f/lll)f/ll l l4 175 h f/,fI,f/ h;fll/hf/kff/f/hh l l5 ?\ fff,f/ jQ'h?fffff,/l ( . 1
; Oo 193 192b Core Loading Sequence Step __161 to 193 F i r,u re 3.0-l (9 4 r 4)
W.B. !!cGuire Unit 3, Cycle ! Core Loading Pattern Key A - 2.1 w/o enriched ' B - 2.6 w/o enriched C - 3.1 w/o enriched R P N F1 L K J H G F 1: D C B A I C27 C43 C60 C12 C07 C50 C26 2 C28 Co3 = C36 C32 C54 A38 C01 A22 C04 A16 CO2 3 C56 C51 B46 A30 B03 A40 B45 A42 B19 A33 B47 CIS C17 4 B24 B48 A43 B34 A52 .B43 A01 B49 B59 B39 C05 C47 B28 C06 C42 A49 B16 A51 B20 A32 B61 A23 B09 A14 B22 A28 C25 C45 5 6 C38 A06 B37 A05 B62 A4. B10 A26 B31 ALO B63 A46 B14 A62 C34 C41 C03 A13 B12 A53 B15 A15 B07 A03 B13 A63 B06 A61 C55 C40 7 8 C22 A37 B18 A36 B38 A60 B11 A29 B53 A35 B32 A21 B57 A24 Cl3 270"F 9 C20 C61 A58 B50 A64 B33 A47 B17 A44 S02 A56 B64 A48 Cll C24 10 C09 A02 A17 B35 AL8 B23 All B01 B01 A55 B26 A04 B05 A39 C39 11 C53 C08 A25 B04 A50 B42 A2, B58 A20 B52 A59 B40 A65 C64 (C49 s 12 C19 B54 B36 B51 A31 B30 A12 B60 A07 B21 B27 B55 C18 i 13 C35 C59 B29 A08 B41 A57 B44 A45 B25 AS4 B08 C33 C46 14 C16 C23 C21 A09 C31 A34 C30 A19 C62 C57 C48 C44 C14 C52 C10 7 258 C29 15 I Fly,ure 3.0-2 8 E
B W.B. McGuire Unit 1, Cycle 1 Core Assembly Insert Pattern Key RC - Control Rod j TP - Thimble Plug P - Burnable Poison Rod t SS - Secondary Source Rod PS - Primary Source Rod R P N M L K J 11 G F E D C B A y TP A10P TP A10P TP A10P TP l 32 6 35 4 29 3 26 l A9P RC 17/ RC 20P RC BPS RC 12P RC B9P 2 1 12 o 14 8 46 1 7 3 24 4 E 3 A9P TP 20P RC 16P RC SS RC 16P RC 20P TP B9P g 4 18 25 13 3 29 1 53 26 4 11 138 2 4 RC 20P RC 20P TP 16P RC 16P TP 20P RC 20P RC 49 29 41 33 8 11 6 25 11 24 9 12 5 5 TP 12P RC 20P TP 16P TP 16P TP 16P TP 20P RC 12P TP 23 2 16 31 19 19 27 27 34 12 4 23 23 1 5 I 6 B10P 4 RC 10 16P 23 TP 2 16P 22 RC 15 20P 5 RC 33 20P 18 RC 36 16P 24 TP 13 16P 28 RC 34 B10P 1 7 TP 20P RC 16P TP 20P TP 20P TP 20P TP 16P RC 20P TP 39 10 11 15 30 3 25 9 38 19 10 16 28 28 16 g g B10P RC 16P RC 16P RC 20P RC 20P RC 16P RC 16P RC B10P 270 y 3 47 2 2 9 43 4 44 1 50 10 26 14 42 5 9 TP 20P RC 16P TP 20P TP 20P TP 20P TP 16P RC 20P TP 9 2 1 5 36 13 15 21 128 6 12 4 8 20 3 B10P RC 16P TP 16P RC 20P RC 20P RC 16P TP 16P RC B10P 10 15 51 30 38 30 6 1 19 2 6 39 8 20 29 20 TP 12P RC 20P TP 16P TP 16P TP 16P TP 20P RC 12P TP 11 14 35 8 102 21 4 22 27 93 7 33 13 24 17 1 RC 20P RC 20P TP 16P RC 16P TP 20P RC 20P RC p~ 27 17 21 34 28 6 18 18 31 32 45 22 40 B9P TP 20P RC 16P RC SS RC 16P RC 20P TP A9P 1 1 41 26 52 21 3 2 48 20 31 7 22 3 14 B9P RC 12P RC 20P RC APS RC 12P RC A9P 3 32 5 25 16 17 1 37 7 30 2 TP A10P TP A10P TP A10P TP 15 17 1 7 2 14 5 37 0 Figure 3.0-3
l l 4
! l 5 4.0 TESTING PRIOR TO INITIAL CRITICALITY 1
ig Following initial fuel loading of McGuire 1, various tests were performed 15 prior to initial criticality. This testing included the following. ) l Reactor Coolant System Flow Test l Reactor Coolant System Flow Coastdown Test i Resistance Temperature Detector Bypass Flow Verification Test l Pressurizer Functional Test ) Movable Incore Detector Verification Test l Full Length Rod Drive Timing Test l Rod Position Indication Alignment Check Test j Rod Drop Time Measurement Test i l These tests were performed in a time period between February 1981 and ' 1 August 1981. i
- l l
!s !I i 4 !I 4 Il 4 4 f k 4 i 4.0-1 3 i _-
E 4.1 REACTOR COOLANT SYSTEM FLOW TEST The purpose of this test was to obtain data on reactor coolant pump input power and reactor coolant loop elbow tap dif ferential pressure in order to calculate the Reactor Coolant System flow rate at no-load operating tempera-ture and pressure with various operating reactor coolant pump (NCP) con-figurations The procedure called for instrumenting a loop for NCP speed, elbow tap AP, l reactor coolant loop temperature and reactor coolant pump power. W'th w the loop instrumented, the pumps were run in various combinations, from four pumps running to one pump running. Data was recorded at one minute g intervals for ten minutes on each pump combination. The same procedure 3 was used to collect data on the other three loops. A total of 16 pump configurations were run. Multiple problems were encountered in data g collection. These included leaking valves, NC pump vibration problems, seal leakoff problems, and RTD temperature problems. Because of these 5 problems, it took 2 1/2 months to collect all the data. Once the data had been collected, the reactor coolant system loop flow was calculated. From the raw data, the average of various parameters for all combinations of pumps was determined. Then for each loop with a specific g pump configuratica, the specific gravity was calculated. Next, the rela- g tive coolant flow for each loop, under farious pump configurations, was calculated using the following definition:
- ~
1/2 FI = (DPI)(SG4)/(DP4)(SGI) where: FI - Relative loop flow for a given number of pumps running DPI - The differential pressure for a given number of pumps running DP4 - The differential pressure for a specific loop when all pumps are running SC4 - The specific gravity in a given loop when all pumps are running SGI - The specific gravity for a given number of pumps running The base case for the relative coolant flow was the four pump operating g condition. The relative loop flow in this configuration was set equal to one. As pumps were secured, the flow in the operating loops increased 5 because of back flow through the idle loops. I I 4.1-1 I I
5 For each loop, a gallon / minute flow rate for the four pump configuration was assumed and using the relative loop flow previously determined, the
- g flow rate for the three, two and one pump configuration was calculated.
5 ""2"a '"" """"""" ' " r"'" ' ' " ' "r- '"r""- '" """ "" 9""" " "- dition, the input power required for each combination was determined from the pump data supplied by the manufacturer. The difference in ll power between the normal input power measured in the four pump configura-
;W tion and the input power prediction from the manufacturer's data was calculated.
This difference in power, called the offset value, was added to the input power prediction calculated in the three, two, and one pump con-figuration. Tuis revised input predicted power was then subtracted from the normal input power actually measured in the plant. The difference was recorded as power deviation (the deviation could have been either negative or position). The deviation for the three, two, and one pump
- h configurations were added together for a total deviation. This process sw was repeated for other trial and error flow rates until the deviation was exactly 0.0. This then became the flow rate for a given loop.
Results of this test are shown on Table 4.1-1. Two of the acceptance criteria were met. All flows were within +10% of the average flow rate, and all calculated flow rates were between 97,500 and 106,275 GPl!. The
- l third acceptance criteria was not met.
EU did not exceed 403,650 GPl!. The total calculated flow rate lt was decided to hold off any action until a more accurate method of I calculating flow (based on an energy balance across the steam generators) was performed during power escalation testing. I lI l5 I I .I I I 4.1-2 3
I I I Calculated Reactor Coolant Loop Flows Loop 1A 100,500 GPM I Loop 1B 98,374 GPM g Loop IC Loop 1D 102,195 GPM 100,110 CPM 5 TOTAL NCS FLOW RATE 401,179 GPM S. I I I I E l I I I I I I Tabic 4.1-1 I,
l 4 3 4.2 REACTOR COOLANT SYSTD1 FLOW COASTDOWN TEST I I The purpose of the Reactor Coolant Flow Coastdown Test was to determine ! reactor coolant flow versus time for various, specified reactor coolant ) pump trip combinations; and to compare these test results with minimum i acceptable flow coastdown criteria in the FSAR. i. l Various combinations of reactor coolant pumps were operated and steady-( state data acquired. Subsequently, all or a portion of the operating ! pumps were tripped and data were recorded during t he reactor coolant flow transient. Steady-state data were again taken following the flow t ransient . 1/4 Coastdown l For this coastdown, besides verifying the validity of the safety analysis, three other quantitles were calculated: (a) Low flow time delay (b) Undervoltage trip delay
.m (c) Underfrequency trip delay
'E l The low flow time delay is defined as the time from the opening of the reactor coolant pump breaker to the time of the first motion of the Rod jl j Position indication (RPI) signal. The acceptance criteria called for the ]W low flow time delay to be less than or equal to 2.49 seconds. The actual j time was 1.16 seconds. The undervoltage trip delay was defined as the undervoltage relay delay l time measured in the Reactor Protection System Functional Test plus the ' time from the opening of the reactor trip breaker to the time of the first l motion of the RP1 signal. The acceptance criteria was that the undervoltage ( trip delay be less than or equal to 1.50 seconds. The actual time was j { 0.898 seconds. lW l The underfrequency trip delay was defined as the underfrequency relay delay j
- time measured in the Reactor Protection System Functional Test plus the 1 lg time from the opening of the reactor trip breaker to the first motion of the
- g RPI signal. The criteria was that the underfrequency trip delay be less than i
- or equal to 0.60 seconds. The actual time was 0.403 seconds.
. l l When the actual reactor coolant system (NC) flow, corrected for flow sensor j delay, was compared to the flow assumed in the Final Safety Analysis Report, the actual flow did not meet the acceptance criteria. (See Figure 4.2-1).
'g i
Flow sensor delay is defined as the time at which the best straight line l 3 approximation to the inverse flow curve drawn in the 4/4 coastdown intersects the inverse flov value of 1.0. 4.2-1 l l lI
I
,4/4 Coastdown For this coastdown, the actual flow (corrected for flow sensor delay) was compared to the flow assumed in the Final Safety Analysis Report. The actual flow again did not meet the assumed flow in the safety analysis. l (See Figure 4.2-2). Also, the slope of the Inverse Total Core Flow curve *'
for the 4/4 Coastdown should have been <0.08048. The actual value was 0.09835. 1/3 Coastdown In this case, the actual flow (corrected for flow sensor delay) was compared to the flow assumed in the Final Safety Analysis Report. The actual flow always equaled or exceeded the flow predicted in the safety analysis. (See Figure 4.2-3) . 3/3 Coastdown in this case, the actual flow (corrected for flow sensor delay) was compared to the flow assumed in the Final Safety Analysis Report. The actual in this case did not meet the flow predicted in the safety analysis. (See Figure 4.2-4). RESULTS The actual recorded flow coastdown was non-conservative with respect to the flow coastdown curves presented in the McGuire FSAR Chapter 15 loss of flow analysis for the 4/4, 1/4 and 3/3 cases. A reanalysis was conducted on the most limiting loss of flow transient presented in the FSAR using the actual recorded flow coastdown. The limiting transient was found to be the complete loss of forced reactor coolant flow. The recorded flow coastdown (as compared to the flow coast-down utilized in the FSAR) that was used in the reanalysis is shown in Figure 4.2-2. The results of the reanalysis are shown in Figures 4.2-5 and 4.2-6. The analysis demonstrated that incorporating the recorded flow l W coastdown also yielded results which continued to meet the licensing basis; 1.e., the DNBR remained above 1.30 during the incident. As a l g I result, all test data was deemed acceptable. g l l I I l B l l . 8 l l 4.2-2 l
I NC Flow Coastdown Test I 1/4 Coastdown Transient I I I I . N m,
\ .
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I : c S l 0 5 3 E c.s l I i I 1
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l l l 1 Time (Seconds) Figure 4.2-1 1 E 1
NC Flow Coastdown Test 4/4 Coastdown Transient i 1.0
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9 10 3 Time (Seconds) Figure 4.2-2 3
I NC Flow Coastdown Test 1/3 Coastdown 1ransient I I I I ._ n.u a y I i A 3 i NN g I j " ~ NN \ \ E N h
.MFAS1TR FD 8
C x - u
\ \ $ Al CEPTnNCE :RITERIb l $
H U.6 I I :) 1 , 3 4 i b 7 F 9 11) I I Time (Seconds) Figure 4.2-3 I
SC Flow Coastdown Test I 3/3 Coastdown Transient I I.3 - I k 3 0.9 - =
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a h a d g 2 % -0 . 3 I
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0.t 0 1 2 3 4 5 6 7 8 9 10 Time (Seconds) Figure 4. 2-4 I
.I . Reanalysis of Worst Case 4/4 Coastdown i Nucicar Power and Core Flow Data I 1 2000 1.0000 - 8
= 80000 -
E 50000 - W ' sz E .40000 - l ' 20000 - 0.0 I l5 1 l 2000 1 0000 - 80000 - r S I t w 60000 - - 8 v I 1 40000 - - E I 20000 - J 1 0.0 j ko $ Eh o o oo a ~ ' e e l TIME (SEC) 'I
~ _ ,. . s ,
1 I
Reanalysis of Worst Case 4/4 Coastdown I lieat Flux and DNBR Data 1 2C00 I 1.0000 - 80000 - 5 d -
.60000 -
7 Y 40000 - 20000 - 0.0 2.2000 2 1000 - 2.0000 - 1.9000 - - 1 8000 - - M E 1 7000 -
! 6000 -
t 5000 - 1 4000 - - 1.3000 - - t 2000 , , , o c o 8 o 8 o 8 o 8 o 8 o ~ , e m - T!*E (SEC) B Figure 4.2-6 I
l l i 4.3 RESISTANCE TEMPERATURE DETECTOR (R1D) BYPASS LOOP FLOW VERIFICATION TEST I l The purpose of this startup test was to measure the actual flow rate and transport time for each RTD Loop and to verify the low flow alarm setpoints in each loop. The minimum required flow rate for each bypass loop was based on transport time for the hot leg bypass loops of 1.0 seconds and actual installed , i piping lengths from the bypass loop connection on the reactor coolant loop l to the last downstream RTD. There was no required transport time for the l cold legs since the cold leg temperature does not change appreciably with ] power. Table 4.3-1 shows the minimum flow rates required and actual flow l rates measured for the hot lei;s. I I All RTD low flow alarm setpoints were et and checked to trip within +2% of 90% of the total measured RTD loop flow rates. All acceptance criteria B was met. I 1 i I I l l I 1 I 1 I m M l 4.3-1 l I L
I I I I! Ilot Leg RTD Bypass Loop Data l W 1 Actual Bypass Flow I, Reactor Coolant Loop Minimum Required Flow, GPM Measured, GPM A 104 130 B 74 121 I C 73 118 I D 89 137 I I I I I i I i I Table 4.3-1 g I I
i 1
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4.4 pkESSURIZER FUNCTIONAL TEST The purpose of this test was to establish a continuous spray flow rate )l l lW by adjustment of the manual bypass valves which parallel the power operated spray control valves. Also to determine the effectiveness of the pressurizer normal control spray, and to determine the effectiveness g of the pressurizer heaters. .g During the Continuous Spray Flow Rate Adjustment nortion of the test, r the continuous spray flow valves (INC28 and INC30) were adjusted while ) the power operated pressurizer spray valves (INC27 and INC29) remained j closed. 'I he adj us tment was made until a minimum spray flow was achieved 3 such that line temperatures remained above their low tempera- }g the spray"F) ture setpoint (eS30 while the dif ference between the average 1 3 pressurizer temperature and each average spray line temperature was maintained at less than 320 F. The only problem encountered was the 4 I poor accessibility of the valves to be throttled. During the Control Spray Effectiveness portion of the test, pressurizer hackup and control heaters were de-energized and the pressurizer control 1l W spray valves (INC27 and INC29) were fully opened until the pressurizer pressure decreased to approximately 2000 psig, at which time the systems were returned to normal and the transient was terminated. The pressurizer I pressure response was recorded during the transient. The spray valves caused a pressure decrease of 250 psia in about 90 seconds. The results of the test are shown in Figure 4.4-1. The pressure response was outside I the tolerance early in the transient. This was due to the operation of the spray valves and the initialization of spray flow. This was evaluated and found to be acceptable. I During the Heater Effectiveness portion of the test, pressurizer power operated relief valves and pressurizer spray valves were closed. The pressurizer level control and pressurizer spray control were then put in I manual. All pressurizer heaters were then energized until the pressurizer pressure increased about 60 psia nefore returning all systems to normal. The test was run four times. The first time it was determined that the data was taken incorrectly. Test number two failed due to blown fuses which caused heater response to be too slow. Test number three failed due to a leaking press,rfzer spray valve. Finally, test number four produced acceptable results. Figure 4.4-2 shows the results of test number tour. I 'I ll 4.4-1 I I
I Pressure Response to Opening of Both Pressurizer Spray Valves I 225's - I h \ I
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i i, ll 13 4.5 MOVABLE INCORE DETECTOR FUNCTIONAL TEST j
}g The purpose of this test was to verify proper operation of all five and !g ten path transfer devices and to check all previously measured path lengths using a dummy cable. In addition all alarms and indicator lights 1
were checked for proper operation, as were the leak detection and gas purge systems associated with the movable incore detector system.
! During the early phase of the test, sticking of the detectors was a major
)g problem. Investigation of the problem revealed that the thimbles !g
- installed had been manufactured using an old annealing proce3s. New incore thimbles were obtained from Duke Power Company Catawba Nuclear Station Unit 2 in South Carolina. These were manufactured using the new I annealing process. These new thimbles were cleaned and then installed at McGuire Unit 1. This adequately corrected the detector sticking i
problem. Testing continued with only minor problems arising. These i were easily corrected. The incore leak detection system was tested by pouring water into the !g header in each ten path transfer and observing it actuate the level switch l5 and correspond 1n;; alarms. The incore gas purge system was checked, and a positive pressure between .02 and .04 inches of H O was measured on all ten path transfers 2 II j All acceptance criteria was met. I I I I I I 4.5-1 I L __ - --- - ._ _ - - - - - - - -- --
I 4.6 FULL LENGTil ROD DRIVE TIMING TEST The purpose of this procedure was to verify proper timing of each ! slave cycler and to check the operation of each control rod drive l mechanism (CRDM) prior to using the CRDM's in both the cold and hot condition. ,g 5 Several faulty circuit boards throughout the rod control system caused some delay in initial testing while troubleshooting the system. I Some problems were encountered with the temporary microphones connected to each CRDM during the remainder of the testing, but these proved to be faulty connections. No rod drive timing problems were encountered while performing this test. CRDM stepping problems occurred while I performing Rod Drop Time Measurement Test on Rod M-2 (See Section 4.8). Delayed release of the movable gripper on withdrawal was attributed to a " sludge" type buildup in the mechanism. An exercise program recommended by Westinghouse along with plant heatup solved this I problem. All acceptance criteria was met. I I I I I I I I I I I 4.6-1 I
!I i l l i 1 4.7 ROD POSITION INDICATION ALIGNMI:NT CllECK i {g The purpose of this test was to verify that the Digital Rod Position lW ludication (DRPI) system correctly indicates the position of all rods j over the full range of travel, that the rod position alarms perform as
- g designed, and that each rod operates satisfactority over its range g of travel.
l One coil stack (for rod D-8) was discovered to be bad and was replaced. = Throughout the rod testing (Rod Drive Timing and Rod Drop Timing) probler.s were encountered with bad Detector-Encoder printed circuit cards and with the cables between the coil stacks and the two data f cabinets. The printed circuit cards were repaired and the cables replaced as they failed. A complete changeout of the cables to j another type of cable is planned. The DRPI system indicated the position of each rod within +4 steps i atter the above problems were resolved. All alarms and rods performed I satisfactorily. All acceptance criteria was mer. il i lI !I 4 l lI l l l i l I .I I i, I I . . ,4 4 I E . _ ._ _ . _ _ _ _ _ . . . . _ . _ _ _ _ _ .
I 4.8 ROD DROP TI!1E MEASUREMENT l 3 The purpose of this test was to determine the drop time of each rod g from loss of stationary gripper voltage to dashpot entry and to rod bottom under both no flow and full flow conditions, first with the plant cold and again with the plant hot. Cold plant condition rod drops were performed with some problems. Because of the inaccessability of the head area with the missile shields in place, a microphone had to be installed on each Control I Rod Drive Mechanism (CRDM). Many problems were encountered in maintaining good connections on the microphones. The rod in position 11-2 did not drop to the bottom during the cold f ull-flow drops, but I stuck at the entrance to the dashpot. Subsequent investigation pointed to a small bit of debris between one of the rodlets and the guide tube. This caused the rod cluster to cock and jam at the entrance of the dashpot. Subsequent multiple redrops of rod H-2 I showed no further signs of problems. The rod in location M-2 started misstepping during the cold full-flow rod tests. (see Section 4.6, Full s Length Rod Drive Timing Test). After the cold plant rod drops, the reactor head had to be removed to replace the head seal, so it was decided to perform the cold full-flow rod drops again to confirm proper operation of the system prior to heatup. Similar problems with microphone traces were encountered that were seen on the original cold rod drops described above. Ilo t f ull flow rod drops proceeded with a minimum of problems. Before starting the hot rod drops, a new type of microphone mounting I bracket was fabricated for each of the 53 microphones. This minimized problems with microphones during the hot plant rod drops. The hot no-flow rod drops were performed utilizing a test box which combined I the A and B coil traces and the A plus B coil traces on ona visicorder trace, thus reducing the required number of rod drops by one-half. Figures 4.8-1 and 4.8-2 show the A and B traces and the A plus B traces while Figure 4.8-3 is a sample of the combined trace. I Figures 4.8-4 through 4.8-8 show the times of the five sets of rod drops. All times were well within the acceptance criteria of 2.2 seconds to dashpot entry in the hot, full-flow conditions. I I I I 4.8-1 I
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.E" T AEUL AT I ON - T E MPEF All'RE - 555"F PRESSURE - 2233_ psi n ___. I FLCd _100__ _ ._ hIXX EFElrER "OPEhlhG" 10 DLSMP0T ENTRY-!N SECOSDS DATE _Zl23/81 _ .. y.XX EF.ELKER OPENING" TO DISHPOT E0TICM-lH SECCSDS PLLhI _ IDENTIFICATION um Figure 4.8-8 I !I 1 i i 1 5.0 INITIAL CRITICA!.ITY l }g Initial Criticality w s achieved on August 8, 1981, at 0928 hours with all rods out except Control Ilank D, which wis at 120 steps out, and )3 a reactor coolant boron concentration of 1316 ppm. Criticality was i g achieved by deboration. 5 After establishing baseline counts, the shutdown and control banks were withdrawn at 50 step intervals. Inverse Count Rate Ratio (ICRR) I was plotted at each interval until Control Bank D was approximately 145 steps withdrawn. I i ILc.eline counts for the ICitR plots that began with boron dilution were then obtained. Boron dilution commenced on August 7, 1981, at 2310 hours ' by the addition of approximately 50 gpm demineralized water. Initial horon concentration was 2090 ppm. Boron was sampled every 30 minutes I and plots of ICIIR vs. Water Addition and ICRR vs. Time were maintained for both source range channels. I When the 1CRR from Source Range Channel N32 equaled O.1, the ICRR for al1 Source Range channels was re-normalized to 1.0; the dilution rate was changed about to 30 gpm, and ICRR data was taken at 5 minute intervals c I instead of 10 minute intervals. Dilution continued until the ICRR was about 0.2 then the reactor coolant l system was allowed to mix. Criticality was achieved about three etinutes later at 0928 hours. Bank D was then inserted to 120 steps to maintain the reactor just critical. The total time to reach criticality from the start of dilution was 10 hours and 30 minutes. The just critical conditions of Control Bank D at 120 steps and a boron concentration of 1316 ppm were within the acceptance criteria of a reactivity equivalent of +50 ppm ! of Control Bank D at 145 steps and a boron concentration of 1301 ppm. See l' inures 5.0-1 through 5.0-4 for plots of ICRR vs. Time, and ICRR vs. Water Addition during the test. i l l l l ' l I i I I 5.0-1 i .i f l ; i i 4 I r I ~ 1 4 1 !I s i N I .I 1 b i .. - t . - 4 i, . I . . -
- i. .
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- 0 . f0 . 0, 5 m . + j0 nU , -- 2 n'- - , . O. a ' ,, ) ,1 ,' 3 R R E C I m:d"sc u'oit - O ! I' 1 ii , '1 ;; I:;,lIi1l l !!sll i ,i l ' E 6.0 ZERO POWER PliYSICS TESTING I, Zero Power Physics Testing (ZPPT) began on August 8, 1981, at 1120 hours and ended on August 20, 1981, at 1900 hours. The purpose of this test program was as follows: (a) To perform nuclear instrumentation overlap verification (b) To establish the point of nuclear heat and the upper limit of the neutron flux level during ZPPT (c) To perform a checkout of the reactivity computer (d) To provide a sequence of testing to gather data to verify core design parameters. Data included isothermal temperature i coefficient measurement, boron endpoint measurements, flux map data, and rod worth determinations. E- Following Initial Criticality, a minimum of one decade overlap between the source rant;c and intermediate range detectors before the source range was i blocked was verifieg When the intermediate range (N35 - N36) first came onscaleaIOl x 10 amps, the source range (N31 - N32) read 550 cps. At 1xIg amps on the intermediate range, the source range read I 1.5 x 10 cps. Thesougerangemaybeblockedwhentheintermegiaterange is greater than 1 x 10 the source range. amps but must be blocked before 1 x 10 cps on The upper limit of neutron flux for zero power physics measurements was set below the point of adding nucicar heat. Atorabovethepgtof adding heat, the doppler broadening of capture resonances in U causes h u feedback effects which tend to mask the values of reactivity. The point of adding heat can be observed as an exponential decay of the reactivity trace or an increase in Tave. The controlling bank was withdrawn until a startup rate of approximately .25 DPM was obtained. The flux level was al1 wed to increase and signs 9 ofnuclearheatingwereopservedat3.5x10 amps on the reactivity computer and at 2.0 x 10 amps on both intermediate range detectors. The range for zero power physics testing was defined as the next lowest whole decade such that the upper end of that decade is not within 60 g of nuclear heat. Based on this_griteria, the ZPPT range on the reactivity g y computer was set between 1 x 10 amps and 10 x 10 amps. All zero power physics testing took place within this range. Figure 6.0-1 shows the traces for IR amps and Tave used to determine when nuclear heat occurred. I 6.0-1 8 4 I During reactor physics measurements, the core reactivity was monitored via an analog reactivity computer. This computer provided a solution to h the delayed neutron precursor decay rate equation for the six groups of 11 5 delayed neutrons. Proper computer operation was dependent upon amplifier and potentiometer settings corresponding to properties of the delayed neutron precursors. g 3 To initially set up the computer, one power range channel (NI43) was taken out of service and its signal was input to the reactivity computer. In order to verify that the conputer was working properly, a checkout was performed by making reactivity changes of approximately +25 and +50 pcm and comparing the output of the reactivity computer to reac-tivity inferred from doubling time measurements. The results of this j 5 check out are presented in Table 6.0-1. The reactivity computer indicated reactivity was within 2.9% of the reactivity inferred from doubling time measurements for all reactivity insertions checked. This value was within the 4% acceptance criteria averaged over three measurements. . I I I I I i 1 i I 6.0-2 8 I . _ _ _ _ _ . _ _ . . . . _ ___..__.~.._._ ___ _ __ ________ _______ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _.____ ._. E O O NI E N $ M E E S M' S M MN O E E 1 Reactivity Computer Checkout i Reactivity inserted i Doubling it.dicated on Reactivity Average Initial Flux Level Time Period y Reactivity Computer calculated Percent Percent N-35 N-36 Sec. Sec. PCM PCM Difference Error i 9 x 10 -7 1.0 x 10~ 171.5 247.5 30.5 29 5.2 1.75 x 10- 1.5 x 10~ 178.2 257.1 29 29 0.0 2.9 i .9 x 10- 1.0 x 10~ 179.0 258.2 30 29 3.5 i t 1.0 x 10- 1.2 x 10- 92.0 132.8 51 51 0.0 1.1 x 10- 1.2 x 10~ 95.1 137.2 49.5 50 1.0 0.3 1.0 x 10~ 1.2 x 10~ 90.8 131.0 51 51 0.0 ; i c2 E C l I Determiriation of Nuclear lleat 9 ._ .1__ y_.__ _ _._ w ,.:'~^~~~G I c. _-_-s. _. _- _,4 . _ . . . - - - - -- _1 _--- _ _ _ .l _- _ LqI - - - - L ~ - _ _ _ . . . _..m 4 ;' n- - -g- -. _ . __-j-- ' ~, _ l-- -- 4 ;- q; --]
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I i i 'i i . ' f , t t=4 -. 1 ! -' i 6 6 . . . I, ' . i , Nji i i > lil * ^ ', Nuclear ", t ' i - I , . , .>i . . , i i . .I _ i Heat ' L t=3 . , i * ~ ; . 6 t t I' ' ' I i .! , i .. .! i! , i i , . J ,6 . i i.si i 2. , . . 1 i .i . . . i.. . . . , , t . '
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1 i n i ., i . . . i , . L. . . } , i ; - .j . --+-- i l , , . m- - E. , t l I ! .. , t i - i ' t=0 ), 1 1 I I 3 3 l i 1 _ fl, , i . . ! __ ___ p i ! . . , dTime e i . 1 l i . i i i (min.) i I 'I 1 __ _ - ,i- ' - ! 1 e I t t ' , i , . , . ; , . -7 -7 -7 - 2 x 10 3 x 10 4 x 10 555 55n IR (Amps) T AVE Figure 6.0-1 5 3 6.1 ISOTilERMAL TEMPERATURE COEFFICIENT OF REACTIVITY MEASUREMENT The isothermal temperature coefficient is defined as the change in reac-I tivity for a unit change in the moderator, clad, and fuel pellet tempera-ture. The isothermal temperature coefficient can be considered as being made up of two parts; the moderator temperature coefficient and the doppler coefficient. The doppler component of the isothermal temperature coefficient isalwaysnegativeduetotb38cffect of temperature on the resonance absorption cross section in U . The moderator component, however, !s only negative for undermoderated cores. When so1uble boron is used in the coolant (moderator), the possibility exists that the I reactor could become overmoderated thus causing a positive moderator temperature coefficient. The isothermal temperature coefficient was measured at hot zero power conditions under various control rod configurations and soluble boron concentrations. The doppler coef ficient was used to determine the moderator temperature coefficient. The moderator temperature coefficient I is required to be negative under all operating conditions. A summary of the data gathered is shown on Table 6.1-1. The measurement was done by monitoring core reactivity while changing Tavg between approximately 553 F and 557 F. The temperature change was accomplished by regulating the amount of steam being dumped to the condenser. The temperature and reactivity change were plotted on an X-Y plotter. The temperature coefficient is determined by calculating the slope of the plot. Actual plots of the ARO and D in configuration tests are shown in Figures 6.1-1 to 6.1-4. During the zero power testing program, an attempt was made to determine I if the moderator temperature coefficient became positive. In the all rods out (D at 188) configuration, the isothermal temperature coefficient was determined to be -0.57 pcm/ F. The Doppler contribution was -1.93 pcm/ F, giving an ARO moderator coef ficient at BOL of +1.36 pcm/ F. In order to ensure a negative moderator temperature coefficient at all times, rod withdrawal limits were established. These temporary limits specified the rod withdrawal limits and corresponding boron concentra-I tions for various power levels. These limits will be removed when sufficient reactor poisons have built in to preclude a positive moderator temperature coefficient (about 40 efpd). 5 I I g e.1-1 I i < l } 1 l The isothermal temperature coefficient values were obtained by averaging the values from the cooldowns j and heatup. i i i i Isothernal Temperature Coefficient Su=cary Isothermal Temperature Design j Bank Positions Coefficient Average Value ! .i Configuration (steps withdrawn) CB ppm Heatup/Cooldown pcm/ F pcm/ F pcm/ F ARO D at 188 1308 cooldown -0.60 nRO D at 188 1314 heatup -0.57 -2.39 13.0 -0.53 , I D in D at 22 1251 cooldown -1.83 -2.02 -3.63 13.0 D in C at 220 1258 heatup -2.20 I C&D in B at 193 1122 cooldown -5.97 -5.86 P C&D in 1122 -6.93 13.0 B at 193 heatup -5.75 l Y B,C&D in A at 195 1030 cooldown -7.00 1 -6.83 -7.38 13.0 B,C&D in A at 195 1034 heatup -6.75 l A,B,C&D in ! SB E at 187 967 cooldown -8.90 -9.72 A,B,C&D in SB E at 187 960 heatup -10.35 13.0 l -10.53 l k I M M M M ' I E E E E E E E E E U E E O M E E E E W ) Isothermal Te=perature Coefficient of Reactivity Control Bank D at 188 Steps Withdrawn (Heatup) ~. .; ;- ' ' . j. i ~ ~ . _I ~ ~. ~ ~ ] . 42 ncm . j , ,. . . _ . _ ... . _ . . . . . . 7- _ _ __ _ % :x = =: := := px x x y , i : 5 h1 7 ' .yi IH -
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o ncm i ,.1 . . , ,1 r3 T .; t y% , in a I;r ! ~ ~ ~ ~ j lk - T ui g - 7 7 . . .. . . . . . . . . . Q =_. _,1 -2 pcm ;; , , li ,,l, .,- l 554 555 556 l 552 553 ave 9 Isothernal Temperature Coefficient of Reactivity Control Bank D at 188 Steps Withdrawn (Cooldown) \\, \\ \ \ \ ,,,, -
- , e #
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_ ^ _ _ . :.~ - --- - ~ ~ +4 PC 'l 11 1 ' \ ! s , {' \ - ' ~ - c \\ ' 2 , . - - - ~ o +2 vc* $ _ . - ~ .- .e \\\ S , -, ,t \, \a, y ; \\ .\j,; . \ - ~ ~ \)s) . ^ \ \ - - . . ~ \ ; i \.\ \ \n g pcm r . \ , j, ij , - ~ 7 _.,. _ lk ~ . . - ^ . . . . , . . - ,_ -. - - - ~ ' -.. 1.. \ \. ~ ~ ~ , ,. - _ .- . L -O ~ ~ '^~ ' ~ - - - 553 55h 555 S5b T ave g ,n SS # as 888 g N E E M M M E 'M M W M M M O E W NE E M Isothermal Temperature Coefficient of Reactivity Control Bank D in Control Bank C at 220 Steps Withdrawn (Heatup) }l l lI I f 1
- !l i ll1 j i
,!I i - - i; . _ - I i' !;;l i: l ,.!j,l t..i l. .. - - g 3 I I 1 i 1 1,.. i I 0 pcm . <f !. ; (A ' ,r T ,: m. fgj g f l J I. . g , - y.: x : } r r, s: t a , i m.t u ,L, , * -5 pcm ,,ji ;. l} [r fl q jl j.j. g .t- l-7 4:L t > !.}t ,1 i 2 .1 ; : 7 Oi i i i- - i'l I_ i F ' ( .o 1 -t ;i _. ,4; :. . . . . . . . . _ e$ ;_-.j- - I l'
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, , , 558 556 557 553 554 555 ave l l 1 Isothermal Temperature Coefficient of Reactivity Control Bank D at 22 Steps Withdrawn (Cooldown) I-i 'l l . ~ ~ ' ;,t _ _ - ~ ll{ . +10 pcm 'l,' , j- . ~ 7 3 i . IlI ~ y i ' rk ;; !f i n i ,g g j - ~ ' ~ .~ .~ . . . - ! l i '? ..-. . - -- +5 pcm i1 }. . __ - . , ij ~ ~ ~ ~ I j '{ , li ~ i u . 1 nyg~ h ' (([ O pcm ; _ . . - i i, - 3 -g m x I.j }; 554 555 556 557 553 T,y, 9 l M E E @ g g g g g g 8 6.2 BORON ENDPOINT MEASt'R12 TENT TEST The purpose of the borcn endpoint measurement test was to determine the just -l= critical boron concentration for a particular rod configuration. The just critical boron concentration was measured with the controlling bank near the fully withdrawn or tully inserted position. The controlling bank was then withdrawn or inserted completely and the resulting reactivity change was
- g E measured. This reactivity change was converted to an equivalent amount of horon and added to the just critical boron concentration to get the boron endpoint concentration. The following configurations were measured
l (a) alI rods out (h) Control Bank D at 0 steps I (c) (d) (c) Control Bank C and D at 0 steps Control Bank B, C and D at 0 steps Control Bank A, B, C and D at 0 steps (f) all Control Banks at 0 steps and Shutdown Bank E at 0 steps I (g) all Control Banks at 0 steps and Shutdown Banks D and E at 0 steps (h) all Control Banks at 0 steps and Shutdown Banks C, D and E at 0 steps (1) N-1 configuration - F-10 in Control Bank C withdrawn The rods were positioned approximately 30 steps f ro: the fully inserted or withdrawn limits prior to obtaining boron endpoint data. The reactor coolant system was allowed to mix until 3 samples taken at 15 minute intervals were I within 110 ppm and the pressurizer was within 120 ppm of the reactor coolant system. All measurements were repeated at least twice. The results are shown on Table 6.2-1. All acceptance criteria were met. A typical trice I obtained for the all rods out endpoint is shown on Figure 6.2-1. I I I I I I I 6.2-1 I I l l i l i HZP Boron Endpoint Test Results l I Vendor Predicted Test Acceptance
- Measured Percent Difference Critical Baron Critical Boron Critical Boron from Test Acceptance l Banks Inserted ppm ppm ppm %
ARO 1311 1310 I< 1311 150 -0.08 1 D 1247 1246 112 1248 +0.16 D,C 1124 1128 +0.27 i 1125 118 I g D,C,B 1029 1033 115 1029 -0.39 ! k e. D,C,B,A 961 960 112 967 +0.73 a 2 D,C,B,3,SE 873 879 +13 891 +1.37 7 D,C,B,A,SE,SD 805 823 +11 819 -0.49 I D,C,B,A,SE,SD,SC 712 726 113 723 -0.41 l N-1 (F-10 out) N/A 723 150 702 -2.90 'I 4 l *This value is adjusted f rom the original vendor prediction to account for differences between the previous 1 measurecent and its predicted value. i i l 1 l - .l l 1 i m M M M W W W W W W M M M M M M M M M 8 Boron Endpo.'nt Measurement All Rods Out Configuration ,, -4r - - ;,. . . ; i , , , , , i 8 , , . . . . . ... , . . i i i i .. m.4 ,, , i ! , . i . . . . . i ,! 6 . . t i t = i 6 e i i ! . i ' ' t ' i' ,,, . . i . , , , e i , i I 1 i i i 1 t '.jg , ,,, , , . - ~. I s . 3 6 . . . . . e . t * * * ' I*i ,,, , . . . . i e i e.# 4 . . .
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, e i 1 t __ L . t - _ _ I i . . e . . i * -10 pcm 0 pcm +10 pcm +20 pcm 8 I t n r.u r e 6.2-1 I !I: lI 6.3 ZERO POWER FLUX MAP TEST To determine the power distribution at zero power, flux maps were taken j utilizing the incore movable detector flux mapping system. These maps ! were taken at various control rod configurations. Data obtained from the incore system was input to the SNA-CORE computer code. The output l of the code was used to verify various design calculations and verif y a correct core loading pattern. Table 6.3-1 gives a summary of the l j various configurations when flux maps were run. A y comparison of l j5 predicted vs. measured values of relative power (Fhli) as well as quadrant l l power till ratios and axial offset values are shown on Figures 6.3-1 j through 6.3-5. In general, relative powers were underpredicted on the
- periphery of the core and overpredicted on the interior of the core.
1 .l l 1 l !I lI i i I ~ !I i i 4 5 6.3-1
- I I
i i Summary of Zero Power Flux Map Configurations i i l Rod position Boron Concentrations Map Number Configuration Steps ppa ] ' A ARO D 0 198 1310 ! . B D in C @ 215 1250 ' i \ C D,C in B 0 198 1115 .i D HZP Insertion Limit B @ 166 1145 C 0 51 i Y ! 7, E HZP Insertion Limit B 0 167 1190 l "- C 0 51 l
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.ar .o11. s.o ta t.s s 3 I.tof .8 k .GGo .%9 1. tor. h:31 11+a 1.oNs ht t.o ts t.073 Lo2r 1.146 , l. Ne h!8Y Lloa 997 .'&o _-0. 8 +0.7 + o.7 +2 7 + 3.S* +f3 +55 + 5,1 + (,. 2 +7.sl+Y.1 +0.'l - o.7 Y, 5
- t. o7s hiol 1.12 8 t,orf ,59t, 551 f. coo f*ott 1. 04 L t it 7 hto7 1434 f.tos I.Its I ors L .n1 s.co9 1. lot s 075 1 191 Lif8 f.tT9 Lif0 /.171 h /98 h!9/ 4o75 blo7 L*ot <357 I +og to,9 +1./ + 3.5 + 3. 0 +37 F31 # 3,3 + 5,2. +7 3 + /. 5 + 1,9 -1. 9 - 4. 3 -/,.7
.N lA W ^t o1(o I,14] l 1.Is7 fo llt f.I o f 1.03 4 l 1.'2,7] .9]o ~ . 9 t. I 1,13 5 * . 87l ld - -- .Bs; .4s1 1 313 1.075 h /Bf I hibr h 1tz. 1, tfor I.!81 f.oTS*. s z75 . f7 8s1 + 3,7 +2 9 + 3.o l +V.4 # t,9 #27 + 1. o f ht :-o.t -3. 3 -33 8 40.1 ,4 8c3 + /.7 1o29 .930 1.oss 4097l 1459 b lT? h l8(o t.itt 1,tio $1,oty hizo , q eo N 481 10/7 957 1.16 7 t,to r,.1 /. tt,f 4131 4 tt,5 4 10/, /,/o7 .9f7 , /.o y7 .YS9 +0.3 + l.9 +2 1 +22 + L 1 l t/.2 -y.I - f. 7 -bS -fo.9 - fo T I -f a t - 0. 4 . V3a .845 . 99r 9'l4 ' ho6i f.c50 ! 1.ott l 1.019 t ^ CO ; ,9sc; . 513 l @ n . Y6 2. oYlf l T ~~ . 6(, L , /.009 .919 . 1.04f A o 3] l-l. 04f .9ff I/se04 slif;B[t-17*I ).491+ 1.1 + 7. o +L4 s o.'f + -b 5 27 i -/. S -2. 0 =- 2. 0 I p' 5f4 , fe t. f 93 7 ,777 +0,1 . (,6 o -0,7 ,-2 7 .79 f .74 7 7'S .S/o . .c73 785 ,6&o l/. 5~ ; - 3.0 - 2.7 57s g .5571 -3./ l QPTR (Full Core lleight) 1.004 1.016 Total cor """ Axial Offset (-3*7%) I 1 Predicted 4 3 Maximum Error (+8.9%) At F07 % Difference 0.984 0.996 (p-m)/m x 100 liighes t Relative (1.2/4) Power at M-12 Fim are b.3-I ilZP, Din, BOL Flux !!ap . Boron 1250 ppm
- P"
- Relat e Powers 270 N F 0 7 E 9 'O !! p 13 P,t Is*
I 7-A .sn n,zr s. n.1 .n2 .w . rar .su l .s71 . .rus . set . 91 3 .rz e . s rf s.1 - r.2. -e.i -c.o r.s .W7 .1IT 4:23 L 244 /.2 11 42&s /./? T /.cG7 J'? * .'Ist g *393 .731 Loor 177 1.010 f.not 1.t04 1.3cl 1,ago 477 131 *393 -14.5 -to,G -30 3. 8 -94 - /.. c - $~ o -73 - 8. T ~/ o.1 -//. 5 * , Wro ,7/2 /. coy /,193 g t ylg L353 h t& 3 ,924 .45G ,qly .ft1
- o. ,3 .7,94
. ss 17Sy a.sor iite . sat . <.or . a s.s c .n3 .7sr . csi . rar i.,ss - 2.0 e o.1 is1r - c. 4 + &. o + .T. 8 -4.4 -7 2. _-4.7_ -6.4 - 4.1 - 2.1 -L7 - 2.o o 71 2 .GM .551 .R81 h2Vo ( 3 31 /,431 41/9 b 225 829 4dI 461 770 D . 7 31 , t* 81 54 0 . F18 Asso 1.aos 1.src ngos t.22 a .g73 .ssa .sst .ist -* 2.'l + 1. 7 ~ /. I -/.4 -At 21 -c.1 - o, ' +4./
- t4.4 + 1. C - T. I
+2 4- __.542 .4 T7 .45Es ' .Els1 4018 4/84 L292 41Br /* 302. h/97 /.083 77G .?!9 /.o/6 .177 , 5 11 .514f .177 ,tsj 118 /.ofV 4114 43OL 1. 2 74 /.Sak t.If1 Loff 51T .fft [ .371 +h r +2.1 _ + 2. </ .* /. / -d. i +c. 9 40 9 ~ d. & 0.o -02 + f.f +13. 2 +/. 2 ~3 9 39 .718 LOGO I.131 /, pot latif fol?/ hero LI70 f.oSt I.got t.ty/ I./sy List / /./?1 f./of *72 9 7tJ9 t.23e !. !C8 f,cao /~ .128 403o 4451 r.nno t.r ey /.210 /.08 9 bist 4058 f.nsa ** G ~ 2. G +3*Q + 3.f - 0. 9 +/. 4 +hf +2 4 +/ o -0. y +h4 + 3.'1 -Ao -0.1 .+0.3
- f. cts f.b73 f.2L 7 f. 249 1 119 f* 1.S l .13%
920 f.208 f 3It L 211 fo 3g, ho!5 / sot, 3 42 .1ys s.et t L es1i 4soz & tos Asny 12 of *913 G- 4'3 /,2e/ 4327 4 20t= A. sal l.ott, + o, 3 u.out + 0. 8 +0.9 +b Y + 2.7 + 3.3 +A o -2Af - 2. / -02 -o. 5 +hl +0.7 - 0. 3 +b 4 O 370 4267 /.2&7 /Voy /.284 flL*/ .521 .470 321 f./2a f,199 h ss3 y,2yg, y, gq 7 AST 0 NYY b2o& 4 317 f.3/& h176 hts? . t yS. 5"o f *EVA' h tS'A 1.274 1. 3 ?& 4.317 L2.ob .844 ll +L9 n.3 v 3, s' 42,$ .o.18 + $,2 4 o, g - 1,7 -30 -14 t
- 8. 9 - o. L -c. /o o.2 +h2 hfoy ho% Lot) , fg3 ,'f6 7 /,26 0 L2fo /,29f hff; ,904
.9L 3 4220 /.32sj f,19] .ggy f,0 3l Loss tsan o.tos 1 327 b zot .9ss .ft3 1. tot 4121 g zog, t, sog. says host 1.o tt { - l.1 -/. C + 0.1. *o.1 -02 + 0. 2 + 0.1
- S.7 tC4 t 5~. T + 3. 3 *b3 + 1* 5 t 3.'l + 0 2' f.!z s /. o 92. bl&/ l./C 7 /s2tL h!Vr /027 .786
.13 1 /.01: j,t r o /. 2s y /.tgt A z a, 1070 I.fre 1220 L'K9 !.a to .ine
- t. cts /.o e s h 2ia 4tri g 72 8 bog 2 Lise L12o I.o 94 Asm
-bo -ht -L o + /. 2 +bt + 0. 2 +/. 7 +2.1 + 4a f
- V. 3 -th 3 10,7 +hl -oo7 -S;L
/ /Yi /.o18 261 .f76 fro 2L .621 *576 *18G .190 o* ff 5 ./.b79 f.lTS 1.].70 l.251 /.2[oo 978 .Teo . 977 . ril
- l. ' .917 .99/ . Ir re 1.o94 f. t e g 13o1 strG t. 202. 1.194 tory e47 +/y +bo .* o. 4 - V. T - f.o
~ho -b 3 - 4.9 +1. T +sy +/. & +). 5 < 2.o + 3. 3 f./ 78 oGTG .543 478 14/ .74Y .6f1 551 ',ff,o 119I /,2TB h370 /.278 /,220 *S7V , Sto 6 trl .139 y 730 6 21 , $l,o gyf 1 22 o 13os 1.3/4 blef 'E o - 2. s * *. o +2.1 +2.4 + 1. t +/. 9 +2. 2 + 1.4 +2. 7 +3.f
- 0. C -4.0 hl32 . 72 4 4t3 o'llS .1 74 ,s s(, .4 TLC LIVO l.30Y /,3ll Is 3/4 /.1571 i , , s rl.
1 f.327
- tf l a&SI .7s*/ . 33 3 ll .383 119 .s et .981 h r8' /. 2 3.7 /.317 i
+13 l +4.f., + 4.C - 7. 'l ~7. ! - 2.1
- v. o +/ 7 +nc v47 - 0. 2 + *. 2 +L7 l 29/ 744 I .9 71 /.ot o i h tf4 1; 42sS~lL/2/ A&& b/ ort ! .'143 ' .393 1,8 L 6 6
.353 731 ,977 4080 ' A zo/ h s 04, r.101. 4 otra 1.,97 7 : .731 f -L 1- -2*I to.4 -+ /. 9 < +0.L t +0.1 ' 4A 7 :-/.9 i-7 / 4 -/.c= +ca.- b . S% 7/2 . 43 7 .E52 ,9af ! 12 ,Glaf . 57' .7M 931 91 2 .72 3 .sn I ( 0 .tvY + o. ? +2 3 ,-oo,C -h o + 0. 4 -4. 4 -7./l QPTR (Full Core lleight) Total Core Axial Offset (-2*6%) . .006 Flaximum Error (+14.4%) tieasured at D-12 Predicted ff 3 liighese Relative g l Power at D08 % Dif ference .995 .939 (p-m)/m x 100 Figure 6.3-2 liZP, D & C in. BOL Flux Map Map C B ron 1115 ppm Relative Powers B 0 19 8 s t e ps sid . 270 3 t/ T & 7 8 9 M H 12 is M IT l , , O. 7/ 8 r.301 0.11: 0. b Ol, o.b): 0.?21 C.737 Ch8f 0.7&& o.777 .7. 5! 7 C.777 0.7 4 O.S?/ -S2 - S.o - 1.1 - 2. s - t, to - t.o. s - 7. c E O.511 p,0$p g,N [ 283 f,093 , , , , o.563
- p oys 1.01u G35 s ss5 2%
- f. t 8 3
- 2
- c. 9 t t 0.S37 0 52'-
0.126 0.9 y; I,l83 I. I 5 5 2.osk - 81.1 - B.T - 4.1 'f . B - 2. t.o + 1. 4... . 2. it - le. f - 7.7 - 4. / - 2.0 o.5 72 l.12 */ D.9 44 !. 2 't b t.335 t.356 sju, ),303 l.1 18 l, t 9 G.5 ? S l Ciao C. 5 SI E O o.526 f.ogs o.983 /./ 8 7 s. 2 t. 8 /. 21: t, f s*/ #.218 1.2 t 6 /.387 :7.883 s.03F O. Y2 6 - 7.1 - 8. 0 - 6. V - 4. 8 - 5. 0 - 4. 8 - /. I - 0.7 - I. 8 - C. 5 - 3. 2 - 2. 3 */. 5 I.o o l 0.12 b o.7cs I,o 2 3 p,333 1,375 1.437 1.3SE I.327 .o.111 G. M: 0.407 f.ord 1.335 f.332 a.i f e 'A b " O.635 O. i ts i 0 003 o.u o 1.335 E 0.i t,8 '/. I - 4. 8,
- 0. b 7 's
-19 - 3,2 s.3o2 - 7. 8 - 2. 9 t.9tv - 1. /, - 1. */ - /.? - 0.1 ' + 0. 6 . - 2 . t, - 5. 6 0.707 1.210 c. 2 Olr 1,00b I.e lo I.oys g,qo3 p,ggo j,149 f,07 sj j, p o p D.?tt t .12 ] y,2 sg 4 0. 71 5 O.k 8 l I.883 g, o g y o,99 0 t. 09 s" g. cs1 s.239 g,25; y,13g ,,g5, g_,,5 o,pa g, g g 7 g,s e s 0.688 - 9. 9 - v. 8 E - 3. Ce - 1.1 - i. to - /. 5 - 1. 2 + i. .f +2., 0.833 - - o. S I,0 Eb o. 8T5 ! Om - p. y - i. 3 1.047 I.252 - o. 9 - 3. 3 f,2 98 I,104 0.800 0.193 1. Iso g. 2.y g I.216 1,064 ; u Dt use o. w - F omu 1.> rs 1.302 i.o 5s ' o.s,e o.s n 1.o78 v.s d o. no iss9 a.3n i. a r e I i.2se - 3. 3 -09 + s.b + 0. 'l - 0. 5 ; +'il + 7. 7 - 0. 7 - 0. 2 o '#. + 1.1 + 4. 0 - I.i */. O ~ 4.1 3.14 5 1.olo 1.177 I.338 f.2 3 2 U.d k D. 855 0. 7(s i IL 815 O.be8 I,l bi 1.2 Si 1.235 1.0 210 0.717 o.777 8 'b777 I. ot t, 1. 2.9 s i.3 35 f.238 0,813 o. 8 8 h o.772 0,886 o.813 J. 2,3 8 f. 31t + s. 2 + 2.9 + le.la + ( 7 , s.2il + V. 6 s.os ta - O.T - 2. 5 - 2. 2 + l.1 - 0. 2 + o. 4 + #. 9 + 1. 5 + e . 4. I + 0. 6
- 0. 5 8 t* 2 0. s e t ' t.lil I.'t Co 2 1.24,8 1.C68 4 7taQ 9,4 5 c o.7 42 1.0I6 f.16 2 f.323 f.07'l 0.550; O.591 g
k o.18't ' O.S te l t o. 6 +0.3l I.s s y 4 3,7
- o. try
- 3. 7 n,259 - 0. 7 j,ol g + 8.0 o.7px + 8. O g,9 yg +7 < 0,pyg g.07 8 + 3. 6 + 6. 7 p,259 + 0. 6 g.9 p sg +t>.8 g.lS9 + 7. Y 0, 5b2 +14 0.589 - 1. 4. C. 2 7 2 1.01'l n.2 90 1.350 1,222 0. M (3 875 0. ]'t % 0.03 5 O.E2*l I,l te l i.2*/ l.247 0.911 0,7 7t, E ( o.171 + 0.7 p.opu + 0.1
- 1. 2 9,.
+ o. i g.33 5 - l.1 j,33 g + f.3 g,993 + 2,9 g,ggy + 1. 2
- c. 7 72 0.88k
- 3. 7 + te.1 C. 69 5
- 4. 9 f.133
+ 5. 9
- 8. 33T
+ '5. 5 1.191 + 7. fa 1.006 0 777 + 8.f +0.1 J-) U I,l 73 l.3Ch I . 2 6;' f.021 l o.:J S 0.8s7 8037 0 643 C. 5 t 2 l 1.o23 I.243 f.2 53 f.I 2 7 0.146 )(' .]. ~17 t,o n .l SS !. M 8 8 302 s.059 l o. S U O.893 1.032 0 893 0. W l.CS1 13:A l. 2 rc a t. s 5S" 0. 7 W 6 8 - 0.7 696 - /. 9 1.214 - 2.1 + I. t 1.123 a.9 71
- 3.8 l
- to . 7 I.089 ).0l7
+ 4. 2 + 4. 0 r 5.4 I.17D I.89 1 I.110 + (c 't h + 3. 5 t.CGD + o. 3 I.052 0.973 1.167 e s. 2 + 2. 'J + 2 . 'l I. t loo 0.64r
- 1. g e 3 1,138 c.0T9 1.04S 1 ?'C 1197 I.883 a. t*91 L O. to B I f. t a y a,9 9 o I,o93 , ,o g9 p,33g 9,259
, + 1. 5
- l. 8 + 9. 7 +2.1 + 2. 3
- 2. 't - 1. 5 - 2.9 + I. 2 + s. o '+q.I + 5. 8 +s.7 4 5. ? + $. 9 E 0. 9 ti 3 0.10s u.we: 0.414 1. 2 3 b a t. 2 te : ,.3o2 1 8.3 55 1.353 ' t. 2 76 bl38 0.1'l1 l u.k.1 s O.03I C. b 7 GE03 O.9 *ll g 0 108 o.833 0.t 79 a990 1.viv s.335 /.3o2 0190 0.9 808 - 2. 2 - 2. 4 ' -11 < + r.1 + 5. 3 + 5. s +ss +s1
- s. 2 <yy *oM + 0. 3 *2.1 0.531 s.oa 7 o.1cs 1.177 1.227 1.12 5 /.288 /.23 0 s.2.29 c.9 # 5 /.072 o,991 I s 73
- o. S at, o.s a o s.035 o.893 i,t B 7 i 298 1,2 /o 8 o. 8 3 3 /.c35 g - 3.0 - 2.9
- 1. 2 t,8 s.29t 1.15'I
+ 5. 3 + 2. 5 + 0. 3 - s. 7 p.s 97 - 3 . 'i - 3.5 -3. 5 + 5*. /, ~J 3 + o. 9 e 7. 9 0.s4s I. otu l 1.t <ts o.so2 i 0.99: 0.1 a I.030 ; 1.16 3 \ 1.109 ; c. 9e ? s 0.53oe
- o. sn. o.4 ut 8 p - 3.1 - 5. 3,.- - 1,0
} I.tes 1.ssS J s.colo + 4. 3 - + 2. to
- o. r;. ?
4o8 1.oss - /, 9 .1.rSS - 0. 7 , -l 11.1e3 7 l - 1,[un7'. ? , -05_26i
- 0. 6 .
o.u15 c. 7 te a 0.171 1s91 a.,3 9 i o. 7 u O. w 2 ,
- 0. WI 0 777 o H1 C.121 C. 7 t.+
f - I. S o.7(n. + 0. s - o.: - c.7 - 0. 9 - i.1 C. GC EI - l. to E _ QPTR (Full Core Height) Total Core g Axial Oftset 1.020 1.010 Measured I Maximumat B03 Error (-11.1%) 1 2 Predicted 4 1 li b;he s t Relative ' Difference (1.902) I . 'N 1 .979 Power at HO '. (p-n)/m x 100 Etgure 6.3-3 lizP, D, C 6 B at Insertion Limits Baron 1145 ppm BOL Flun !!ap B 0 166 steps wd. Map D C 0 51 steps wd. Relativo Powers O 5 270 l 1 3 4 6 7 8 1 10 N n n g g 110 .Det is .<** . to & B 8 84 .u3 < .><2 1's .s!, :>, .14,. A . e 4 ,, - 3,3 *l. 4 - z.2 't. s - f. 's - 5. 7 - 7. f 531 . *:n-- I . I 'fl 1.t : $ 08' ,/9) 1.%f 1. f f t : 'r * , 11 3C1 0.101 " !2 3 5l *'l9 ' **015 *13 .la h 4.041 . f - ,. e - 9.1 1.02 3 - 3.3 1 121 -9o 1,24 t -s5 -c 2 -l. a - r. 2 -tu -vo - 1. s , Sif IAtt .ce: 1. o t t o.ast ~ .1 s :.a s. t a s I.301 1.jst ..p?, .502 .tro .sor ,ilf ,839 f.!3 7 s.a tu 4.2tw I.612 s. tsy t.n3s s.h? .ht *1ri .Vtf ( .415' - 2 ,1 5 -9. 7 - /. O - /. 3 e '## + r.0 + 0.(c - 3.1 - 4.1 - 4. (. - 3. 7 - 4,0 ~. . I ,, y, , ,.,, , , s. s , ggi ,tga .sc1 .61S ,swa ,ts, p. 3 9 p.yg3 ,_4, g ,,3,, fa9s .fi: .SG .&f t ,10s . i C t* 831 . re ftr 1 12 s.211 s.331 f 9:2 p.339 g - 0.1 - 2 .1 - 3,1 - 2,1 - s.1 -8.8 + >. a
- 0. 2 _ + a. 3 - 2 . t, 0.4 - 0. 7 c o.3 1,01$ 1,218 1.215 1210 1.02 0 *.015 121 I.I30 t.15 3 .w
. Vl1 f.llio 1.02 7 .411 1.11 g s.t O* I. H '.* .111 1.s 2r t.. L 3
- W 'l
- E .le.He I.123 1.I31 .493 p,gow ,,p oo I. s et t.202 s. s te t
- 1 le - 2. 3 + I. 0 + 4. 8 + 0. le = 2 . f.* = 2.7 - 0.4
- O.lo + 0.1 "0. 0 - 1. 9 + 0. 2 + l.1 -I. 0
- f. 5 5 .111 I. L ll I.22? 9.238 g,g .g 3 . lM
.154 4..o A 1.2 d 2 1.l a t ./81 404 ,g g)
- 1. 2I9 1.I)) s.2 3 b I.s t' 7 .14 L I.800 .733 .153 p.18 8 .153 133 1. .::1o
.14 t= 1. tO7 1.22l* 1,L%D f* - 2, I -to.2 + S. 'i - 5. 2 - 2. 8 - 8.0 + 0. 8 * *4. 7 -0. 2. - 3.1 v -77 -1.to + 0. 5 +. l. 8 + 0. 7 13 1 .128 .630 .131 e V*l I.215 1.241 8. Z re.4 I.U(e6 .GI3 .601 f.012 1.2 16 1.315 l L 11 f. 2 t te , s.cy & .713 ,9 3e .?f3 . J 's 1.377 .11 3 f.oy a J. 2 % s.331 f. 2 /* 7 .ISI 91 & .399 - 2. s { -17 + 0.9 el.u - o. 4 -0. u -0.ci +laa e o. I *a3 r o.e + 4. 5
- S. 7 + 2.4 - 2. S
.GSS l 300 182 I4*6 e.2:3 1.353 f.lif ,135 o .643 IJss 1.402 1.111 f. ll a .*lo 132 3. u z r.m sa+o .w > u3u 0 H .ss- .ss. I a.>u *.v=> i.=sa >. . e .m .,sa . con l.n o -%0 - 1, 3 . o.o + l. ; o. . J # 3.3 *%9 e 4.8 + 5.1 + 9. o - r. 3 l - 3.o - 4.1 + 3.0 ] 111 1.013 1,285 1.355 1. 2 (2 4!'o .133 .713 .ddl .D *. k e d. 3* C l.J 5S 4Uw 149 .D3 e 261 8 331 s.29* < ovt .717 715 s.ov t- s.2% s.339 s.t.et .453 .tik D3 93 te ] - 0.7 - 0.1 + l. 0 - l.1 v 1.3 *lB ' O. 3 v 4 fo + S. C + U 'S *,'t *l8 + 3.3 + 2. l* * !. I IAE I.32' l.Aos s.cg3 ,118 141 1. 2 to2 s.2 51 n.oss .rar 123 1. C Ef 11 1 . ): 4 o.123 *133 9.291 *
- 23f* l.'07 .74%
11L 1.107 o.23u I 2%I g,soa 133 .9S 3 s.118 M3 ' 'o0 }( - 1.4 , + 0.1 + 3. 2 + 3. 5 + 3.8 - 2 .7 e l.1 - 0. 8 + 1. 4 + 3.2 - o.B + 0.1 - 2 ,1 e,1,5 +4 3
- 4. 4 0. U 3 :fb t.1/2 f.121 . fo& 2 633 1.131 1.16 G 112 4.089 f.669 ' d.241 1.2 34 1. 2 10
*b1W I.la s .9 7L 1.137 4.833 .U % p.231 .41L 1.1050 f.soo I. a lo t s.t et 1. 2co t s, loo [. t. s 2 2 + I, to + 3.7 r 4. 2 +.4.6 v S,2 n .1 +- I. 4 +2.I - 0.1 - 2. 5 - 0.1 -01 -I.I + 3.1 t 3.4 .%i n A A2 .%O .tu .010 . *w .t il 1.rul I.3I3 o.3is t. 2 02 s.aos e ts .831 .T M8 s.a ll .3 M .K M .i09 8 31 , t,5 ce 912 1.Z97 . 1.331 1, q u I,331 , y,y + 5,9 4.0 e I. 3 + 0.1 0.0 - 3.1 - 1. o + l.u i , 2,2 + 2. 2 +2.o e 2., !.135 f.2/b l.111 .510 9 IE *416 _.505 ,967 . fly J,p2 g j, 2 01 p.211 1.181 N . .1,s .e s t o.. v o nsu 1.21s- i.ito o.2sv .>'<- isi .s39 . . i - 0.0 l m - 2.1 - l.1 + 1.ls to.6 +2. to . tl.4 + 0, 3 t 0.9 * /. 7 - 0. 3 - 0.1 + 0. 3 .And 4.051 ~ 4 .Gie J.4 54 l 126 413 l . 5o5 .413 f. e l l e.wd 1 l I.s? 49 l 8.107 "I,'s Z $ %35fe ' 541.[- l .'4 9 f .1 014 llL3 l*E O7 ' J.0 4 2 .44U l.048 , f f 2.1 1 - 0.1 - 9. 2 , - 1. 0 + 1. 0 -1.3. -O;l f 0. ]' = 2.0 - 3 l1 , - o.*t .6Y5 .13t 3M .a52 ,71s I 7%f .p52i , ts y t, .741 713 ,23h .743 .79 t ,gyf m - 0. 3 t 0. 5 -l. 3 g ~ 2,1 - 0.*t -l .0 - 1. 2 QPTR (Full Core lleight) Total Core (-38. 37. ) Axial Offset 1.016 1.006 " "" U
- Measured
(+8. 2 37. ) at D-12 1 ' l W liighest Relative (1.482) Predicted 4 3 Power at 11 - 4 0.995 .934 % Difference g (p-m)/m x 100 g Fip,ure 6.3-4 I llZP, D, C 6 B at Innertion Limits, D12 Ejected !!OI. Flun Map Boron 1190 ppm Map E B C 167 steps wd. C0 51 steps wd. I Relative Powers 270" 7 8 9 /0 12- 13 14 i s* 3 1/ 6~ G // , g o.w g y ~.kon .np ,1ss .hso s. ws 1. m
- T*/ .ift . S ait f.14 f4D j, y sa
.1]l "b f ,,' . / - ). L - /, .) t 3 f' -/J. ? l' s. L I.115 E ,37) ,1x1 r, 0 s 1 [ ,, v ;; ; o, s ,. , ,9 3, ,.3, u n3 1.132 r.m 2.v.
- 2. ora 3.229 2 uv
- 2. S he 2.rsg i.c,g B .3a . ris 1. a o. , .. m
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- Bank Bank Worth Bank Worth from Predicted Boron Worth Boron Worth from Predicted
)
- Idenfitication pcra pcm % pcm/ ppm pcm/ ppm .
l Control D 637 164 669 +5.02 -10.00 -10.79 -7.90 [ ! Control C 1242 +124 ~ 1250 +0.64 -10.01 -10.42 -4.09 ll I Control B 991 199 996 +0.50 -10.40 -10.06 +3.27 l ' j g Control A 697 170 695 -0.29 -10.13 -11.21 -10.76
- l fu l $ Shutdown E 892 189 840 -5.83 N/A -11.05 N/A !
4 e , Shutdown D 728 173 755 +3.71 N/A -10.49 N/A i 't Shutdown C 956 +96 1011 +5.75 N/A -10.53 N/A 1 Shutdown B 1200 1120 N/A N/A N/A N/A N/A Shutdown A 395 +39 N/A N/A N/A N/A N/A I N-1 (F-10 out) *6480 +648 6435 -0.69 N/A -10.55 N/A . 1 . i Control Banks ' in Overlap 14% of measured 3575 -0.97 -10.22 11.02 -10.37 -1.47 worth of CB A, l B,C,D (3610 pcm) ; 1 l
- Worth of all rods except F-10
M M M M M M M M M M M M M M M M M M M Differential and Intecral RCC Bank (RCCA) Worth Control Bank D 3 1jf f-13.0U[II}llIl hl l[!IbI I I l b d.I !:I 3l [ hill! f' [E' I i ' l I"i)} ll ly! h' "" Ifl l I ' - 1200 12.0'"[ Ti p["j~1 i i l[! I ' l , II {I l l(5 }icGuire Uni L: 1 11.0 q [ Test: Control Bank D Worth l' h l"j L f; . a. ., I. it ht 4 1000 10.0 ~ ] j , 1 ]' ~ j { ll ; } ._ i 900 Tent conditione.: E r 9.0 7t w ;+ ] { .t ! , { '. I g 1. RCC Bank Positions: ) 8.0 "} '~J" { ] ~ h SDA 228 M 7.0 !i liii kI )..).. ... _I . . - *--L 700 SDS Mu t(( }}- g "]tJ.._ U SDC 228 h b }}ff i J.- .-- -1.. 600 $ 228 SDD .[ - N g {. l~ j SDE 228 - I: l I I ,[i"l ' lg hllhbg. .. S y 5.0 500 o { g _]. CA 228 o n. J c [ 4 $ b....j .y -} 400 M 228 l .1 .? \' %z * -[ CS - ,.o t ;p' ; 0 g [...l . a[. - F- .- j - CC 228 ~[ 3.0 } ~L 1 j." 'lj CD 224 - 0 '[ ' ~ ' ~ ' ' '" ~ [j~ %~ Tl i j1c, 2. Power Level: 2.0 - -7 h r-T) i' - *T'h, _"-j-t 200 HZP %FP f._ 1.0 " 100 3. NC Temp.: 7 t 'k (( { 5 ,I} [if T p , l$itII - 0 2 0 20 60 80 Final: 556.S *F 40 100 120 140 160 180 200 220 240 ? y , RCC Bank (RCCA) Position (Steps Uithdrawn) I N !s. m C1 N o o C M ( o. [ 4 e ~ .. v m M v o y c .. ,O,4 % 2: L a N o . 4 *= ~ s m O t- a ._4 g C o o 00 co to cc co to to 5 y . a 0 J' N N N N N cJ G o o . .. h . .. N N M N N CO g g g ~g [ ' *& N CJ g H 4 == ) e c1 $4 U v4: H { "c.: ca ' d v uv4r3 c d O *:2 3 U U < c'3 0 Q . O O C v4 14 v4 ** u O sm O O m QmQmQm< oC1o uO oO m z e-. m s a e . o o . o o xv v o a . m n I (uxod) ya2on Ie2302ut o o o o o o o o o o o o o o o o o o o o o o e o o rw c n V M c4 g O o, p g :, o en g H H H # E3iEl=7 -~ ~"7M@jE={~i*Ef=t:5MiE} EJEE}gj'j: 3@EE g}Ml c4 = - =:= = =_ :.- o ,e - i-.-- - I_ m . = .=:: =t===:=_: =t = -t== r. = = . m=. t= L ._-t=i== ._ . __ t=_f==:==_:= t m a g rr.3 __+ ele. T=-- - __ ~{ = i w..__. + : L __ E_= r= _._ _..._"i= == =r -1==2E}==hz- *- ? n ed ._f_.-- o g--- =f= _ . _ _r :+ n - =j=.p=.- -..= _+m. -- t=_ ; z= _:= .. .m.. =_: ; i -- - s -- + =,. - _ =_a __.__ g..._ _=- Y,==f .. g .=2.__. . r =1. = g_ - . = ;= _1=:--q = ._.,__=r; ==;==p ,_-: ;=_,p= = =.,-=t--- -t-- == : =cz. y . .. _ s: < : - _ . = ::5 u u ===;-==.g=_._.____--+ ._: = : = r_=_ r ..- 3,===.::._.=--!.t=_===.=g = 1:t:. o m =__.:._=_ - _-=-__ g=_ .- o = r= _ , _ _ -o H .= . . . ~ _ a: =._.;-'_t. =$ =. 1 w =_={ __ _.2 = ; _ =. y_=_ _=_=.=_._--- .I_=_. . .. =. . . _ , _ z _.=_ _ %_. -_;=_5 =_ =r-- .{ _.=__.;- . r -- ==:._.; -+ .- .f_.. ; -_. =r .J' =.,. ..T= o -f== r=4; == --+ c ::: .s: c =:::1,1 = ===: ==-:-= -:=t - ==:==:=2- . ._ .__ N /M 5{E i= ==dCMI-m= :-E E p 5%.i _-- i--~~- i--Ir'.~"5 E2 ; a c ==,_t._ -t-t_._ _a._ = z =- -t- _ _ f- .r_.j_j_ . = y_4_p- , __. .. _ =_unt=g w.. o, m u u n = _.;_ .., _q.___=_._. . _ , 4 . . . v ._f.r=- .2.2.- g ;n . =1_ = = l___ - . _-._-. _ ._ 4, __q =--. . __g _ L,__ .__m a g .= g >=4 __.____m._.,.l=_.=___._ . _ m:- .. . _ ._ o --F=i__ # _ f_q rr .. }_ _ = o m c c u =_r _=. i - _. --- - + - -_ . _ _ .- - - + - . _ _ . + -4 nz 2= S u a = 2::_ ., ,L____1. _ =__=_=_=_._-_ , _ . _ ^ _x=s, n .. ,._ ._3._ - =, _ _1, __ ._ _-- =_ _ ... , _ _ a . _ .=-c==_ g=g =
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Figure 6.4-6 1 (tid ] S / u M1) t{ M Opl {U }3 GJl dj j}([ I W M M M M M M M M M M M M M M M M M M Di f ferential nnd Intenral RCC Bank (RCCA) Worth Shutdown Bank C -1300 i . ~ II I '"'"' 'I ' 1200 {]*] J 12 . 0 .*"* ] f( T~f, i i" ] l ~~]* 11.0 N1 ... .[ ! . 4 - I 1100 MCuhc M L: 1 . l! 0 tl ' l l.)l' 'l } ' l l 1 l ; p I Tcst: Shu gown Bank c Worth H l' ,'0f l 8' - I l I f { F 1000 u,.o-l I : a l 7 !i i I( ; i I k '- l k 'l 900 Tcat CondLtipnn ,c.;. 9.0 { , . ] d l hl 1. RCC Bank Pouitions: c E 8.0 m[I-]t IM k I ! ttr [m h:'f-lL T t., ~]'~ I O b 800 a E SDA __ 228 t I.il 1 d -.liil{l. ..li l s e l 1 I_ .c 7.0 +if- ~ - i-d . t[" - ~ \ -j l 700 .c sun 228 u l ll t Il [ l l l! 1 lI ~ N o,o .I.".!b .!.7k- , I , ,il. I -- 4 ~~+ 600 ~ i [ ~i 1 l h I i T+i , ! - sen 0 c: !
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3,g fd ! 4I } ' j . 1 -k 77 I q 9:4 -3 -l ]- 4 300 3 b son cA 0 0 f 400 E 4.0 'i f h 'i ( CB 0 E ~ ' IN l [] [" ~ [ l CC 0 [ . 300 f! } fdf [ .)_ j[ plI n - d ~ I 1 b .N.- CD 0 'k}% k; I}}. i 2.0 ll {l}l-- [l -l-[ t 1 j .- .[.. l- [ -M 4" g _\ 7 200 2. Power I.cvel: .. .1 . . . . L.. -- s, . 7,yp f 1.0 "- - 100 ..j j 6 n .. ( q .. .. ... _ . 3 - 0.0 i I ! I 1I! ' " ' 0 Initial: 556.5 *F g 0 20 60 80 40 100 120 140 160 180 200 22u 240 Final: 556.5 *F i' ItCC llank (ItCCA) Position (Steps Ulthdrauu)_ I 6.5 S'It'CK ROD WOR'l H MEASl'P EMENT TEST 't he purpose of this test was to verify that the insertion limits defined in the 'lechnical Specifications provide a >1.6% tK/K Shutdown Margin with the nost reactive rod " stuck" in the withdrawn position. Rod F-10 in Control Bank C was the most reactive rod. F-10 was withdrawn as Shutdown Banks B and A were inserted to compensate for the reactivity l l I changes. With Shutdown Bank A at 79 steps withdrawn and rod F-10 com-pletely withdrawn, a boron endpoint was performed on Shutdown Bank A and the corresponding reactivity change of 45.5 pcm was recorded. This valtse I was used to deterc.ine the worth of all rods minus the most reactive stuck rod. Subtracting off the worth of all rods above the r oc' insertion limit for 0% power gave a llZP shutdown margin of 4.62% t.K/K. 'I b e results of the boron endpoint measurement for the N .1 configuration of F-10 withdrawn can be seen on Table 6.2-1. The results of the worth ot alI rods except F-10 can be seen on Table 6.4-1. The test was ended I by manually tripping the reactor and observing rod F-10 drop into the core. I I I I 'I I I I I I 6.5-l l ) 6.6 PSEUDO EJECTED ROD TEST 1 i l The purpose of this test was to determine the worth of the most reactive rod, D-12 in Control Bank D, ejected from the core and to verify that that )l worth was less than the value of 860 pcm as stated in the safety analysis calculations. ) I l I With the controlling banks at the hot zero power insertion limits, D-12 l ! was borated out of the core. Rod worth data was taken. When D-12 was j about 190 steps withdrawn, a boron endpoint was performed to obtain the final rod worth. Then a flux map was taken to see if the ejected rod ; could be detected. The results of the flux map and rod worth measurement can be seen on I Figure 6.3-5 and Table 6.6-1, respectively. All acceptance criteria was met. I : l I lI i il I I I I I I 6.6-1 I Maximum Peaking Factors and Worth of Ej ected Rod Percent Percent
- Vendor Difference Vendor Difference Predicted
- Measured from Predicted Measured from Worth Worth Predicted Maximum Maximum Predicted Condition pcm pcm % pN N 2 F
Q Q BOL. IlzP , No 455 432 -5.05 5.93 6.25 +4.52 Xenon Control p Bank, D, C&b 5 at Insertion C Limits, Rod D-12 C' . Ej ec ted es I W
- Uncertainty in measurement factor of 1.04 not included.
m m m m m m M W W W W W W W W W W W l i l 3 7.0 NATURAL CIRCULATION TESTING ! g f The Natural Circulation testing program at the McGuire Nuclear Station j 1 was performed on April 25, 1981, for Test #1 and from August 29, 1981, ' f to September 1, 1981, for Tests 2, 3 and 4 to demonstrate satisfactory i reactor operation in the natural circulation moce. Onder regular plant conditions following a reactor trip (and loss of il iW norn.al forced circulation), natural circulation is established through l the reactcr core. This occ.urs due to a heatup of reactor coolant from (g decay heat, and cooldown of reactor coolant f rom stear- generators used ig as beat sinks. The density changes due to this continuove heating and cooling process causes flow in the reactor coolant system. During the performance of these tests (except for Test #1, which was , run before initial criticality), the reactor was run at 1 to 3% full f power to simulate decay heat conditions following a reactor trip. The reactor coolant pumps were tripped and the resulting plant conditions l E closely simulated natural circulation conditions which would occur l 5 after many months of operation. These tests also served to provide operator training, while in the natural circulation mode. One hundred percent of the licensed personnel at the McGuire Nuclear Station participated in at least one of the I I natural circulation tests in which the reactor was critical (Tests 2, 3, or 4). The breakdown of personnel who participated is as follows: 27 Licensed Senior Reactor Operators (SRO's) 7 Shift Supervisors 4 Shift Technical Advisors (STA's) on shift 4 Operating Engineers jul l 3 Assistant Shift Supervisors 3 Assistant Operating Engineers Training Instructors at the McGuire Technical I 3 1 Training Center 2 Associate Engineers - Operations 1 Superintendent of Operations i { "" 22 Licensed Reactor Operators (RO's) } 9 Nuclear Control Operators 5 Nuclear Control Operators currently in SRO training l 5 Assistant Nuclear Control Operators ] l 2 Assistant Shift Supervisors currently in SRO training 1 Shitt Supervisor currently in SRO training i All licensed personnel, except three (1 SRO and 2RO's) participated as a minimum in the Natural Circulation Verification Test. These I three participated as a minimum in the running of the Effect of Steam Generator Isolation on Natural Circulation Test. !,I 7.0-1 I iI 7.1 NATURAL CIRCULATION WITil SIMULATED LOSS OF ALL ONSITE AND OFFSITE AC POWER i The purpose of this test was to demonstrate that auxiliary feedwater could l' 1l 1 l 5 be controlled by manual means and remove heat ^nergy from the primary system following a loss of all onsite an.1 offsite AC power. This test l ! was performed at flo t Standby Conditions on April 25, 1981. All four reactor jg coolant pumps were in operation to simulate decay heat. The test sequence , Ig took approximately 2 hours. l 1 i The following initial conditions were established prior to performing the test: 1 l= i (a) Reactor coolant system T ~2242 psig, and pressuriSEE level ~22.5%. -550 F, reactor coolant system ll }g (b) Main steam pressure ~1036 psig I j (c) Pressurizer level and pressure controllers in AUTO Ig j (d) Steam generator levels -31% l (c) Steam generators were being fed by the two motor auxiliary l feedwater driven pumps. ! (f) Charging was maintained by one centrifugal charging pump. j (g) The steam generator power operated relief valves control ! switches were in AUTO with their manual loaders fully open. ~ (h) The main steam isolation valves were closed. 1l j jW The test was initiated by simulating the loss of all onsite and offsite AC power by doing the following: l
- m i
(a) The pressurizer spray valves were closed. (b) The pressurizer heaters were de-energized and both motor driven i auxiliary feedwater pumps were tripped and disabled by removing control power to each pump. 4 (c) The steam dumps were disabled and all main steam isolation valves were verified to be closed. iI iI 7.1-1 I - - _ - - . _ . - - _ _ _ - _ _ - _ _ _ . _ , . . _ _ I The water levels in the Steam Generators began to drop due to lack of feedwater. The turbine driven auxiliary feedwater pump started auto-matically on low-low level signal from two out of four steam generators. (The pump had previously been tested to start automatically on loss of all AC power). The turbine driven auxiliary feedwater pump was then used to feed the steam generators for a period of 1 1/2 hours. During this period, feedwater flow to each 7,enerator was controlled by manually positioning the four auxiliary feedwater pump inlet isolation l a valves to the steam generators. Plant data was taken at this time. See Figurea 7.1-1 through 7.1-4 for the results of this test. All acceptance criteria were met and the test proved that hot standby conditions could be maintained by manual centrol of the turbine driven auxiliary feedwater pump. I I I I I I I I I I 7.1-2 I I I S/C A, B C, 6 D s/G A _ _ STEN 1 PRESS. S/G B VS, TIME S /G C - +-+- S / G D -o-1 I I I I I Natural Circulation with I 1200 SI"I "" ^ Onsite and Offsite AC power Test #1 I I A ~ gk I x jf q cf n ! h c-. ~ \ o/)/ H Y f h F I 1000 E I ' n I \ O. 900 I \ \\ \Yp( ' 0- -o_.o x o s k g \ S/G B I N N S/G S/G 1 ' S/G li 800 0 b 20 4( 60 80 100 120 140 3 , TIME (MINUTES) e$ "$ < !05 am ss <m l Figure 7.1-1 NC LOOP !!IGilEST AVERAGE TEMP. 6 NC SYSTEM WIDE RANGE PRESS. VS. TIME PRESS. TEMP. ' I I I I I I I Natural Circulation with Simulated Loss of All - E Onsite and offsite AC Power 3 ' Test #1 2400 I 2300 % I MI ' 2200 \ \ 2l00 I \ as Na ~ ut c~ v 2000 I ee I I 56 & / - l
- _c \/
5 \ g c 2, A g - \ % , I y 320 400- . 120 140 60 80 100 O -j 20 4Cl
- p. '3 p. TIME (MINUTE)
$$ "' E MM uM I Figure 7.l-2 S/G A, B, C, 6 D l1 W NARROW RA?;CE LEVEL VS. TIME S /G A - - - I S/G B S/G C -h - S/G D l i I I I I I I Natural Circulation with Simulated Loss of All - Onsite and Offsite AC Power Test #1 I I ee , pe -o- o- S/G D a 70 / / o / o 60 g /f; ;; / o '/ . d/o c I 50 I -[ # # S/C ^ ci - y / / I dd 40 7 . / 1 / i l e[J/ - ~ / WM o^ , 30 c f- % o // s I ./ \ _. . . 20 -- ll ! 10 tj % I lI O O 1 20 4j 60 80 100 120 140 TIME (MI!;UTE)
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!O N Q = == F i >;u r e 7.1-3 AUX FEEDWATER FLOW TO S/G A, B, C, & D S /G A - - - - l W VS. TIME s/G E-S/G C ; ; S/G D + l l l l l l Natural Circulation with g Simulated Losa of All - g Onsite and Of fsite AC Power Te_s _t #1 A ?00 I ~ I se e g ,a ea y L _ lr 200 u 1 w _ g _3 l 5 l j; i L. . e 8'c c I .i; I,V o , r/L % r - j / \v } I l X l a 1 / , f e v 5 , g I 100 63 < {;%( $, % l l S J P; i , t v w q p t , 9 , _ _ .. - - - - S/G A ! _.Jj. 1 * .i ' ' r { ~ -- S/G B l q's ii ' H ', I .\j[ 1 s , } l t d 4 i i < [A *k, h o o 3La , - l 0 _Ji b,. m c! ' _ , , ,._- s/G o g 0 { 20 40 60 80 100 120 140 g gg TIME (MINUTE)
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_ . _ . . _. _ _ . _ _ ~. -- -. - - - -. - _ _ l I 7.2 NATURAI. CIRCUI.ATION VI;RIFICATION TEST The purpose of this test was: (a) To demonstrate that natural circulation with all Reactor Coolant Pumps tripped is adequate to remove decay heat from the core I ! (b) To demonstrate that steam generator level and feedwater flow can be controlled under conditions of natural circulation to maintain adequate coo.*ing of the reactor
- 3 (c) To demonstrate tl.e chility to maintain natural circulation and i
saturatic.,n margain with loss of pressurizer heaters (d) To determine the reactor coolant system depressurization rate when the pressurizer heaters are tripped while in natural circulation I (e) To determine the capability of the plant Operator Aid Computer Saturation Program while in natural circulation I (f) To demonstrate the effects of charging and steam flows on control of the saturation margin while in natural circulation (g) To provide Operations personnel with experience in unit operation while in natural circulation McGuire Nuclear Station Unit I was first put into natural circulation E co"ditions o" ^usust 28, 1981, at 0545. The Natural Circulation Veri-g fIcation Test was run 2 more times, August 29, 1981, and August 30, 1981, to allow as many Operations personnel as possibic an opportunity to observe or participate in the natural circulation testing. Data was taken only on the first of these three tests. All acceptance criteria were met and the test was performed smoothly except for the following minor problem: While taking a flux map in natural circulation conditions, Detector A would not operate properly. Detector F was I used as a backup to Detector A for the test of the flux map.
- The test took approximately 17 hours to complete the first time and about 8 hours the other two times. The following discussion pertains i to the first run of the test.
I ' ;I 7.2-1 I
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i I j The initial conditions of the test were: (a) The reactor was crit ical at approximately 3% reactor f ull power. l (b) All reactor coolant pumps were in operation. (c) Control Bank D was about 150 steps withdrawn. (d) Reactor Ccolant System was at normal temperature and pressure with pressurizar level at 25%. I l (e) Steam Generator narrow range levels were about 35%. From a reactor power level of approximately 3%, the reactor coolant pumps g ! were tripped and the establishment of natural circulation was verified about 30 minutes later by monitoring the wide range hot and cold leg B temgeratures. These temperatures indicated a 27 F AT across the loops (28 F AT for test run #2 and 25 F AT for run #3) and a saturation margin l l of 75 F. Plant data was taken for approximately 30 minutes after natural W circulation steady state conditions were es tablished. After taking this data, a full core flux map and a thermocouple map were taken (first g performance of the test only). See Figure 7.2-1 for FNH results f rom the Natural Circulation Flux map. 5 An auxiliary spray test was then performed to determine the position l l of the pressurizer spray valves where pressurizer pressure is first W j affected. After the auxiliary spray test, the pressurizer heaters were tripped and the reactor coolant system was slowly depressurized at the rate of about 80 psig/hr. with a pressurizer cooldown rate l of about 5"F/hr. The purpose of this depressurization was to verify l that saturation margin could be maintained without pressurizer heaters. The saturation margin was verified controllable by using charging flow and steam dump flows. The Reactor Coolant System was then repressurized and alI four reactor coolant pumps were restarted, tnus ending the test. See Figures 7.2-2 through 7.2-6 for Reactor Coolant pressures and temperature during this test. E g In general , plant responses were as expected and all acceptance criteria was met. l I I: 7.2-2 I I I Natural Circulation at 37. Power, AR0 0 BOL Flux Map Relative Powers Boron 1260 ppm D @ 185 steps ud. /0 // gg /3 l'/ /g* / g 3 Af 6~ $ 7 8 9 270 E .51 .6&1 ' . 8/ *, .74 7 .ELY 700 5/7 .5c/ .&&/ e'133 . 11 7 738 . &.',1 .557 f -v2 a. r -vo -4/ -se, -67 - o. 2 ,88 Z f. or) 492 f /.M ? A 0 l 4056 ho!S .CEG . if) E G - 49f 487 .tc o roog 1.aos isva /..; 72 t.o29 sofa /. omo 1. ca . Tao . var -/4 -2.5 - Y.Q -2. 3 -/. 5 -3.5 ~3.o -s~ S -45 - 3. o -/. / /. /.18 f.o2 I t.17L 1,/30 t,t9z l.11/ ht2t? 9C,7 /.054 . 4.0 9 .Y 9/ / 095 . is>3 ^
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-4 5 -11 , -2 1 -b? 1 5- - 2. *3 E - QPTR (Full Core licight) Total Core (-8*5%) Axial Offset 1.007 1.009 Maximum Error Measured l at All i 2 Predicted 4 3 liighest Relativ (1*257) % Difference Pw r at D-12 .995 .989 (P-M)/M x 100 E Figure 7.2-1 .. . . -_ ___. . _ _. _ _ _ . m.__. . _ __ _ _. __ __m .m _ - __m - _. .. ._. i ., , -. .- + + + + . . . . - .-+- . .. . e. . 6 6 .e . . . .-4 l - ,. . ....... .... , . . -. .-+...-.r-.. + ...-..4-n.._y-. 4 ... + .... . . . ,a. -- ..... - . , .4.,..,n... .,- . . ,e . . . . . . . . 4. . . . . <;, .s. g.. . 4, . .,
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