W3P85-3218, Startup Rept for Us Nrc
| ML20133H300 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/10/1985 |
| From: | Barkhurst R, Cook K, James Smith LOUISIANA POWER & LIGHT CO. |
| To: | Martin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| W3P85-3218, NUDOCS 8510170010 | |
| Download: ML20133H300 (420) | |
Text
{{#Wiki_filter:- _ i LOUISIANA ,42 OuAnONOE staur P O W E R & L i G H T! P O BOX *6008 NEW OnLEANS LOUIStANA 70174
- 1504) 366 2345
$ ON vsNU October 10, 1985 W3P85-3218 A4.05 QA Robert D. Martin Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011
Dear Mr. Martin:
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 STARTUP REPORT
,a Enclosed are two copies of the Waterford 3 Startup Report which provides the results of the recently completed startup test program conducted under the subject license. This report is submitted pursuant to Section 6.9.1.1 in the Waterford 3 Technical Specifications (NUREG-ill7) of Appendix A to Facility Operating License No. NPF-38.
Very truly yours, alOwk_ K.W. Cook Nuclear Support & Licensing Manager KWC:GEW:sms Enclosure cc (w/ enclosure): NRC Director, Of fice of I&E (Document Control Desk) (6) NRC Resident Inspectors Office cc (w/o enclosure): G.W. Knighton, NRC-NRR k J.H. Wilson, NRC-NRR 9 B.W. Churchill W.M. Stevenson I s 8510170010 851010 PDR ADOCK 05000382 P PDR L
\l LOUISIANA POWER & LIGitT COMPANY WATERFORD STEAM ELECTRIC STATION. UNIT 3 ;
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J i i STARTUP REPORT
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LICENSE NUMBER NPF-38 DOCKET NUMBER 50-382 , I i October 10; 1985 - n s I
2 APPROVAL FOR ISSUE The Waterford Steam Electric Station Unit 3 Startup Report has been approved for issue by the members of the Plant Operations Review Committee (PORC) and the Plant Manager-Nuclear. PORC MEMBER SIGNATURE DATE Maintenance Superintendent -
/0//P / ff Operations Superintendent /d -/# -V[
Radiation Protection Superintendent ""- /0!/0!i5' U ' Plant Quality Manager Ou/e ,-i l, - //////pf Technical Support Superintendent ua ,. m ,, /6 o J-o .i.t - Pian a .. cr 'bue 4,;ur PORC Chairman . /O-/p-R(~~ PLANT MANAGER-NUCLEAR ID[IO/S C 1 O U .
I 3 V ACKNOWLEDGEMENTS The Phase III Tect Program was conducted and contributions were made to this Startup Report by the following individuals. Louisiana Power & Light Company appreciates the professional manner in which the test program was conducted and hereby expresses its gratitude to the contributing members of the program. Tom Andrews Hamid Mahdavy Dennis Barr Steve Matlock Russ Brian John McCauley Marcia Brisson Glenn McCloskey Howard Brodt Matt Melancon Roger Carr George Miller Mark Constable Jack Mosely [_ ( j i Charles DeDeaux Jerry Moyers Tom Earle Chuck Mitchell Jim Edwards Rudy Poulos Dave Evans Bob Ryan Ed Fiegler Dave Shannon Tom Firestone Rob Starkey Steve Johnson Rick Thomas Roger Ju Bernie vonKutzleben William Karras Carl Whitaker Mark Konya Reid Wolf John Lewis (h
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4 7-~s ( (_-) TABLE OF CONTENTS EADE
1.0 INTRODUCTION
AND
SUMMARY
17
1.1 INTRODUCTION
18 1.1.1 The Startup Report 18 1.1.2 The Facility 18 1.1.3 The Test Program 29 1.2
SUMMARY
50 1.2.1 Initial Fuel Load 50 1.2.2 Post Core Hot Functional Testing 50 1.2.3 Initial Criticality 51 1.2.4 Low Power Physics Testing 51
,- 1.2.5 Power Ascension Testing Through 20% Power 51
(),,, 1.2.6 Power Ascension Testing From 20% Through 50% Power 52 1.2.7 Power Ascension Testing From 50% Through 80% Power 53 1.2.8 Power Ascension Testing From 80% Through 100% Power 53 2.0 INITIAL FUEL LOADING 55 2.1 Preparations 56 2.2 Reactivity Monitoring 60 2.3 The Fuel Loading Sequence 63 2.4 Fuel Movement 65 2.5 Fuel Load Verification 67 2.6 Delays, Problems and Resolutions 71
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5 TABLE OF CONTENTS s (continued) V EAGE 3.0 POST-CORE HOT FUNCTIONAL TESTING 74 3.1 INSTRUMENTATION TESTING / CALIBRATION 75 3.1.1 Intercomparison of Plant Protection System (PPS), Core Protection Calculator (CPC), and Plant Monitoring Computer (PMC) Inputs 75 3.1.2 Incore Instrumentation Baseline Data 79 3.1.3 Moveable Incore Instrumentation Operation Verification 81 3.1.4 Post Core Vibration and Loose Parts Monitoring System 85 3.2 REACTOR COOLANT SYSTEM TESTING 87 3.2.1 Reactor Coolant System Flow and Flow Coastdown Measurement 87 3.2.2 Reactor Coolant System Leak Rate Measurement 118 (,,) 3.2.3 Reactor Coolant System Heat Loss 121 3.2.4 Reactor Coolant System Expansion Measurement 125 3.2.5 Control Element Drive Mechanism (CEDM) and Control Element Assembly (CEA) Tests (CEDM Performance) 133 3.2.6 Pressurizer Spray Valve and Control Adjustment 140 3.3 OTHER TESTING 145 3.3.1 Post-Core Test Data Record 145 3.3.2 Heated Junction Thermocouple Operation Verification 147 3.3.3 RCS and Steam Generator Parameters 150 3.3.4 Determination of Auxiliary Spray Flow Split 152 3.3.5 Post-Core Thermal Expansion Testing 154 4.0 INITIAL CRITICALITY 4.1 CEA Withdrawal 158 lm) 4.2 RCS Dilution , 159
. . . . . - _ - . - -. . . _ _ . . ~
4 6 3 IABLE OF CONTENTS (continued) a i EADE 5.0 LOW POWER PHYSICS TESTING 169 5.1 CEA Symmetry Checks 171 4 5.2 Shutdown CEA and Regulating CEA Worth Measurements -174 5.3 Isothermal Temperature Coefficient 174 l 5.4 Critical Boron Concentration Measurement 179
- 5.5 Boron Worth Measurements 180 i
1 6.0 POWER ASCENSION TESTING 182 6.1 POWER LEVEL DETERMINATION 183 6.1.1 RCS AT Power Determination 183 6.1.2 NSSS Calorimetric 185 6.2 INSTRUMENTATION TESTING / CALIBRATION 191 6.2.1 Nuclear and Thermal Power Calibration 191 6.2.2 Process Variable Intercomparison 195 6.~2.3 Linear Power Subchannel Calibration 203
- 6.2.4 Vibration and Loose Parts Monitoring System 209 6.2.5 Control Systems Checkout 212-6.2.6 Incore Detector Signal Verification 218 1
6.3 CPC/COLSS TESTING 224 i 6.3.1 COLSS Power / Flow Verification Data Record 224 6.3.2 Adjustment of COLSS Secondary Pressure Loss Terms 229 6.3.3 CPC/COLSS Verification 241 6.3.4 Radial Peaking Factor and CEA Shadowing Factor
-Verification 247
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6.3.5 Temperature Decalibration Verification 254 6.3.6 Shape Annealing Matrix Measurement 263 l l I
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7 f 1%GE L 6.4 PHYSICS TESTING 289 6.4.1 Core Performance Record 289 6.4.2 Variable Tavg 310 4 6.5 REACTOR COOLANT SYSTEM TESTING 318 i 6.5.1 kCS Calorimetric Flow Measurement 318 6.5.2 Natural Circulation Demonstration Testing 329 j 6.6' TRANSIENT TESTING 337
- 6.6.1 Remote Reactor Trip With Subsequent Remote Cooldown 337 j 6.6.2 Load Changes 348 -
l 6.6.3 Loss of Offsite Power Trip 350 6.6.4 80% Total Loss of Flow Test / Natural Circulation 356
- 6.6.5 100% Turbine Trip 373 6.7 PLANT TESTING 381 6.7.1 NSSS Plant Data Record 381 6.7.2 Transient Data Record 383 6.7.3 Biological Shield Effectiveness Survey 385
- j. 6.7.4 Power Ascension Testing Ventilation capability ' 388 i 6.7.5 Atmospheric Steam Dump and Turbine Bypass Valve
-Capacity Checks 392 6.7.6 Initial Turbine Startup. 401 6.7.7 B0P Data Record -405 6.7.8 Level 2 Piping Vibration Testing - 407 6.7.9 Pipe Whip Restraint Monitoring 409 l 6.7.10 Inspection of Mechanical Snubbers and Spring Supports. 411 l 7.0- APPENDIX A - List of Acronyms 414 b
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8 O . LIST OF TABLES PAGE 1.1 Design Parameters of Waterford 3 SES 26 1.2 List of .Startup Test Procedures and FSAR Chapter 14 Test Commitments Filled 32 1.3 Power Ascension Milestones 34 1.4 Post-Core Hot Functional Test Plateaus and Tests Performed at Each Plateau 35 1.5 Power Ascension Test Plateaus and Tests Performed at Each Plateau 36 2.2.1 . Detector Count Rates.(Uncorrected for Background) 62 3.2.1.1 4-RCP Steady State PCHFT RCS Flow Rate Measurement Test Results- 94 3.2.1.2 As-Left PCHFT CPC Flow Constants 95
) 3.2.1.3 As-Left PCHFT COLSS Flow Constants 95 3.2.2.1 RCS Leak Rate Test Results (Gallon; per Minute) '120 3.2.3.1 Final RCS Heat Loss Test Results 123 3.2.5.1 CEA Drop Times to 90% Inserted 136 3.2.5.2 Average Drop Times to 90% Inserted of Three Drops of CEAs Outside 20 138 3.3.3.1 RCS and Steam Generator Parameters 151 4.1 1/M Summary for the Approach to Initial Criticality- 162 4.2 Verification of Startup and Log Power Channel Overlap 164 5.0.1 .Waterford 3 SES LPPT Results 181 6.1.1.1 RCS Delta-T Power Determination Test Results 184 6.1.2.1 NSSS Calorimetric desults 189 6.2.1.1 Nuclear and Thermal Power Calibration 194 6.2.3.1 As-Left Signals and Signal Fractions at 20% 206 o.2.3.2 As-Left Signals and Signal Fractions at 50% 207 6.3.1.1 Calibrated Turbine First Stage Pressure (BTFSP) and Secondary Calorimetric Power (BSCAL) for all Power Plateaus 226
() 6.3.2.1 Adjustment of COLSS' Secondary Pressure Loss Terms, 20% Power Test Results 232
9 i O LIST OF TABLES (continued) EAGE ]
-6.3.2.2 Adjustment of COLSS Secondary Pressure Loss Terms, 50% Power Test Results 232 6.3.2.3 Adjustment of COLSS Secondary Pressure Loss Terms, 80% Power Test Results 232 6.3.2.4 Adjustment of COLSS Secondary Pressure Loss Terms, 100% Power Test Results 233 j 6.3.2.5 Adjustment of COLSS Secondary Pressure Loss Terms, Installed i 2 COLSS Constant Values 234 6.3.2.6 Verification of Installed COLSS Secondary Pressure Loss Constants' Adequacy 235 2 6.3.3.1 CPC/CEDIPS Comparisons 244 6.3.4.1 CPC Planar Radial Peaking Factors 252 j 6.3.4.2 COLSS Planar Radial Peaking Factors 253 6.3.4.3 CEA Shadowing Correction Factors 253 6.3.5.1 Summary of Temperature Decalibration Verification Test Results 256 -
l 6.3.6.1 Comparison of Measured and Design SAM and BPPCC Values 271 6.4.1.1 20% Peaking Factors 296
- 6'.4.1.2 50% Peaking Factors 297 i
1 6.4.1.3- 80% Peaking Factors 297 6.4.1.4 100% Peaking Factors 297 6.4.2.1 50% Variable Tavg Test Results 317 6.4.2.2 95% Variable Tavg Test Results 317 6.5.1.1 20% Plateau Results, RCS Calorimetric Flow 323
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6.5.1.2 l 50% Plateau Results, RCS Calorimetric Flow 324 6.5.1.3 80% Plateau Results, RCS Calorimetric Flow 325 6.5.1.4 100% Plateau (First Run) Results,'RCS Calorimetric Flow 326 6.5.1.5 100% Plateau (Second Run)~Results, RCS Calorimetric Flow 327 , 6.6.4.1 CESE Single Value Acceptance Criteria. Parameters During the i First-60 Seconds F.ollowing Loss of Flow - 361 6.6.5.1 Turbine Trip Single Value Acceptance Criteria and Results 375
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i i i 10 O l LIST OF TABLES (continued) i
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l 6.7.4.1 Summary of IIVAC Test Results 391 i 6.7.5.1 Measured ADV and TBV Capacities 399 1 i O . , i I ! 6 t t 1 l i P O
- . 11 t '
i . O LIST OF FIGURES 4 4 PME i
- 1.1 The Facility 19 1.2 The Region Within 50 Miles of WSES-3 21 i 1.3 The Region Within 5 Miles of WSES-3 22 1.4 Isometric View of the NSSS 23 1.5 Reactor Coolant System Arrangement Elevations 24 1.6 WSES-3 Cycle 1 Power History Through Completion of Test -
g Program 38 1 1.7-1 Major Events During Power Ascension (March 22 - 31, 1985) 39 1.7-2 Major Events During Power Ascension (April 1 - 10, 1985) 40
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1.7-3 Major Events During Power Ascension (April 11 - 20, 1985) 41 i 1.7-4 Major Events During Power Ascension (April 21 - 30, 1985) 42 1.7-5 Major Events During Power Ascension (May 1 - 10, 1985) 43 1.7-6 Mafor Eventa During Power Ascension (May 11 - 20, 1985) 44 i 1.7-7 . Major Events During Power Ascension (May 21 - 30, 1985) 45 } 1.7-8 Major Events During Power Ascension (May 31 - June 9,1985) 46 i 1.7-9 Major Events During Power Ascension (June 20 - 29, 1985) 47 1.7-10 Major Events Durins Power Ascension (June 30 - July 9,1985) 48 1 1.7-11 Major Events During Power Ascension (July 10 - 19, 1985) 49
- 2.1.1 Location of Temporary Fuel Loading Neutron Detectors 'A' and 'B' 58 l 2.1.2 Schematics ofiTemporary Neutron Counting Station Setup 59 2.3.1 FuelLoadingSequence 64 2.4.1
. Elapsed Time Between Fuel Assemblies Placed in Core 66 l 2.5.1 WSES-3 Cycle 1 Core Map (Fuel) 68 2.5.2 WSES-3 Cycle 1 Core Map (CEAs) 69 2.5.3 WSES-3 Cycle 1 Core Map (Neutron Sources) 70 I
- 3.2.1.1 RCS Flow and Flow Coastdown Measurement Test Sequence 89 l 3.2.1'.2 RCP 1A Differential Pressure Transmitter PDT-0110 a
Calibration Curves . 96 , 3.2.1.3 RCP 1A Differential Pressure Transmitter PDT-0111 I
. Calibration Curves 97 i
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LIST OF FIGURES (continued) D195 , 3.2.1.4 RCP IB Differential Pressure Transmitter PDT-0112 Calibration Curves 98 3.2.1.5 RCP IB Differential Pressure Transmitter PDT-Oll3 Calibration Curves 99 3.2.1.6 RCP 2A Differential Pressure Transmitter PDT-0120 Calibration Curves 100 3.2.1.7 RCP 2A Differential Pressure Transmitter PDT-0121 Calibration Curves 101 3.2.1.8 RCP 2B Differential Pressure Transmitter PDT-0122 Calibration Curves 102 3.2.1.9 RCP 2B Differential Pressure Transmitter PDT-0123 Calibration Curves 103 3.2.1.10 Reactor Vessel IA Differential Pressure Transmitter PDT-0124W Calibration Curves 104 3.2.1.11 Reactor Vessel IB Differential Pressure Transmitter PDT-0124X Calibration Curves 105 3.2.1.12 Reactor Vessel 2A Differential Pressure Transmitter PDT-0124Y Calibration Curves 106 3.2.1.13 Reactor Vessel 2B Differential Pressure Transmitter PDT-0124Z Calibration Curves 107 3.2.1.14 SG #1 Differential Pressure Transmitter PDT-9116-SMA . Calibration Curves 108 3.2.1.15 SG #1 Differential Pressure Transmitter PDT-9116-SMB Calibration Curves 109 3.2.1.16 SG #1 Differential Pressure Transmitter PDT-9116-SMC Calibration Curves 110 3.2.1.17 SG #1 Differential Pressure Transmitter PDT-9116-SMD Calibration Curves () 3.2.1.18 SG #2 Differential Pressure Transmitter PDT-9126-SMA Calibration Curves 111 112 I ( l l
13
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G' LIST OF _ FLGURES (continued) EA06 3.2.1.19 SG #2 Differential Pressure Transmitter PDT-9126-SMB Calibration Curves 113 3.2.1.20 SG #2 Differential Pressure Transmitter PDT-9126-SMC Calibration Curves 114 3.2.1.21 SG #2 Differential Pressure Transmitter PDT-9126-SHD Calibration Curves 115 3.2.1.22 Measured Flow Coastdown vs. FSAR Assumed Flow Coastdown 116 3.2.4.1 Reactor Vessel Support Lateral Restraint Gaps 129 3.2.4.2 Steam Generator Sliding Base X-Direction Gaps 130 3.2.4.3 Reactor Vessel Support Anchor Bolt Grillage-to-Washer Gaps 131 3.2.4.4 Steam Generator Anchor Bolt Nut-to-Washer Caps 132 O ( ,/ 3.2.5.1 Histogram of CEA Drop Times to 90% Inserted at 545*F d 2250 psia 139 3.2.6.1 Location of Temporary Thermocouples 143 3.2.6.2 Pressurizer / Reactor Coolant System Depressurization versus Time Curve 144 3.3.2.1 Approximate Location of Heated Junction Thermocouple Levels 148 4.1 W3 Initial Criticality - Inverse Multiplication 165 4.2 W3 Initial Criticality - Boron Concentration 166 4.3 W3 Initial Criticality - 1/M versus Boron 167 4.4 W3 Initial Criticality - Overlap Verification 168 5.1.1 CEA Symmetry Test 173 5.2.1 Integral CEA Group Worth Regulating Groups 6 and 5, No Overlap BOL, HZP (measured values) 175 5.2.2 Integral CEA Group Worth Regulating Groups 4 and 3, No Overlap BOL, HZP (measured values) 176 5.2.3 Integral CEA Group Worth Regulating Groups 2 and 1, No Overlap BOL, HZP (measured values) 177 (-~)s 5.2.4 Integral CEA Group Worth Part Length CEAs (Group P) BOL, HZP (measured values) 178
14 I l (_ s) LIST OF FIGURES (continued) EaDE 6.2.3.1 Excore Signal Paths 204 6.2.6.1 Incore Detector System Layout 219 6.2.6.2 Incore Detector Test Circuits (TS-3 and TS-4) 221 6.3.1.1 Calibrated Turbine Power 227 6.3.2.1 SG1 Feedwater to Steam Header Pressure Loss versus Steam Flow 236 6.3.2.2 SG2 Feed <ater to Steam Header Pressure Loss versus Steam Flow 237 6.3.2.3 SGI Generator Steam Header Pressure Loss versus Steam Flow 238 6.3.2.4 SG2 Generator Steam Header Pressure Loss versus Steam Flow 239 y 6.3.5.1 CPC Temperature Decalibration, Channel A 258 x 6.3.5.2 CPC Temperature Decalibration, Channel B 259 6.3.5.3 CPC Temperature Decalibration, Channel C 260 6.3.5.4 CPC Temperature Decalibration, Channel D 261 6.3.6.1 W3 Cycle 1 Shape Annealing, Axial Shape Index 272 6.3.6.2 W3 Cycle 1 Shape Annealing, Upper Excore Responses 273 6.3.6.3 W3 Cycle 1 Shape Annealing, Upper Peripheral Powers 274 6.3.6.4 W3 Cycle 1 Shape Annealing, Middle Excore Responses 275 6.3.6.5 W3 Cycle 1 Shape Annealing, Middle Excore Response - Expanded Scale 276 6.3.6.6 W3 Cycle 1 Shape Annealing, Middle Peripheral Powers 2/7 6.3.6.7 W3 Cycle 1 Shape Annealing, Middle Peripheral Powers - Expanded Scale 278 6.3.6.8 W3 Cycle 1 Shape Annealing, Lower Excore Responses 279 6.3.6.9 W3 Cycle 1 Shape Annealing, Lower Peripheral Powers 280 6.3.6.10 W3 Cycle 1 Shape Annealing, Channel A Excore Values 281 6.3.6.11 W3 Cycle 1 Shape Annealing, Channel A Peripheral Powers 282 O O l L
15
\pl N.s LIST OF FIGURES (continued)
EAGE 6.3.6.12 W3 Cycle 1 Shape Annealing, Channel B Excore Values 283 6.3.6.13 W3 Cycle 1 Shape Annealing, Channel B Peripheral Powers 284 6.3.6.14 W3 Cycle 1 Shape Annealing, Channel C Excore Values 285 6.3.6.15 W3 Cycle 1 Shape Annealing, Channel C Peripheral Powers 286 6.3.6.16 W3 Cycle 1 Shape Annealing, Channel D Excore Values 287 6.3.6.17 W3 Cycle 1 Shape Annealing, Channel D Peripheral Powers 288 6.4.1.1 Waterford -3 SES, 20% Axial Power Comparison (CECOR vs CPC A) 298 6.4.1.2 Waterford -3 SES, 20% Axial Power Comparison (CECOR vs CPC B) 299 7-s 6.4.1.3 Waterford -3 SES, 20% Axial Power Comparison (, (CECOR vs CPC C) 300 6.4.1.4 Waterford -3 SES, 20% Axial Power Comparison (CECOR vs CPC D) 301 6.4.1.5 Waterford -3, Radial Power Distribution Comparison, 20% Rated Thermal Power 302 6.4.1.6 Waterford -3, Radial Power Distribution Comparison, 50% Rated Thermal Power 303 6.4.1.7 Waterford -3, Radial Power Distribution Comparison, 80% Rated Thermal Power 304 6.4.1.8 Waterford -3, Radial Power Distribution Comparisor., 100% Rated Thermal Power 305 6.4.1.9 Waterford -3, Axial Power Distribution Comparison, 20% Rated Thermal Power , 306 6.4.1.10 Waterford -3, Axial Power Distribution Comparison, 50% Rated Thermal Power 307 6.4.1.11 Waterford -3, Axial Power Distribution Comparison,
,, 80% Rated Thermal Power 308 6.4.1.12
(_ ) Waterford -3, Axial Power Distribution Comparison, 100% Rated Thermal Power 309
_. . - - -.- .- - -. _ .. . . - . . . _ . - . . . . . _ - . - . . - . . ~ . . . . . . . - - 10 l LIST OF FIGURES (continued) i PAGE l i 6.4.2.1 Variable Tavg Test Sequence 314 i l t 6.5.1.1 Adjustment Flowchart During Calorimetric Flow Measuremert 319 t 6.5.2.1 Natural Circulation Demonstration, Reactor Coolartt Hot ' and Cold Leg Temperatures 334 i 6.5.2.2 Natural Circulation Demonstration, Steam Ger.erator Pressures 335 6.5.2.3 Natural Circulation Demonstration, Pressurizer Pressure 336 6.6.1.1 WSES-3 Remote Reactor Trip, RCS Cold Leg Temperatures 3 '+ 1 6.6.1.2 WSES-3 Remote Reactor Trip, RCS Hot Leg Temperatures 342 6.6.1.3 WSES-3 Remote Reactor Trip, Pressurizer Pressure 343- f 6.6.1.4 WSES-3 Remote Reactor Trip, Pressurizer Level 344 6.6.1.5 WSES-3 Remote Reactor Trip, Steam Generator Pressures 345 l 6.6.1.6 WSES-3 Remote Reactor Trip, Steam Generator Levels 346 i
- 6.6.3.1 Loss of Off-Site Power Test, Reactor Coolant Hot and
i Cold Leg Temperatures 353 , 6.6.3.2 Loss of Off-Site Power Test, Steam Generator Levels 354 i 6.6.4.1 W3 Loss of Flow Test, Steam Generator Average Pressure 364 I 6.6.4.2 W3 Loss of Flow Test, RCS Temperatures 365 6.6.4.3 W3 Loss of Flow Test, Pressurizer Pressures 366 6.6.4.4 W3 Loss of Flow Test, Pressurizer Level 367 l 6.6.4.5 W3 Loss of Flow Test, Steam Generator 368 6.6.4.6 W3 Loss of Flow Test, RCS Temperatures 369
- 6.6.4.7 W3 Loss of Flow Test, Pressurizer Pressures 370 f
- 6.6.4.8 W3 Loss of Flow Test, Pressurizer Level 371 4
6.6.4.9 Natural Circulation Demonstration, Reactor Coolant Hot l and Cold Leg. Temperatures 372 ' 6.6.5.1 ~W3 Turbine Trip Test - Hot Leg Temperatures 377 6.6.5.2 .W3 Turbine Trip Test - Pressurizer Pressures 378
- 6.6.5.3 W3 Turbine Trip Test - Pressurizer Levels 379 ,
) 6.6.5.4 W3 Turbine Trip Test - Steam Generator Pressures 380 [
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SUMMARY
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1.1 INTRODUCTION
1.1.1 The Startun Report
'The issuance of this Startup Report for the Waterford Steau l Electric Station Unit 3 (WSES-3) is in ecmpliance with the
, United States Nuclear Regulatory Commission's (US NRC's) Regulatory Guide 1.16, Revision 4 (Reporting of Operating Information - Appendix A, Technical Specifications), as outlined in, and required by Sections 6.9.1.1 through f.9.1.3
! of the Station Technical Specifications.
The report describes the initial fuel load, the postcore hot functional testing, the initial criticality, the low power physics testing and the power ascension testing performed following the receipt of a Low Power Operating License (NPF-26) on December 18, 1984, and an Operating License (NPF-38) on March 16, 1985. It addresses each of the post , core-load tests described in the FSAR, and includes a descrip-I tion of the measured values of the operating conditions or f characteristics obtained during the test program and a 4 comparison of these values with design predictions and specifications. Any corrective actions that were required to t achieve satisfactory operation of the plant are also described, as are all other specific details required in license
. conditions based c., other commitments (e.g. the Safety Evaluation Report (SER) NUREG-0787).
i 1.1.2 The Facility WSES-3 (Figure 1-1) is a nuclear generating station utilizing a Combustion Engineering (C-E) 3410 MWth (including 20 MWth reactor coolant pump (RCP) heat) pressurized water nuclear () steam supply system (NSSS) and a Westinghouse Electric e t
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20 Corporation turbine generator outputting 1153 MWE gross (1104 MWE net). Ebasco Services was the architect-engineer and managed construction services. The unit is located adjacent to two fossil fueled generating units, Waterford SES-1 and -2, cn the west bank of the Mississippi River between Baton Rouge and New Orleans, Louisiana. The site is in the northwestern section of St. Charles Parish, near the towns of Killona and Taf t (Figures 1.2 and 1.3). The Louisiana Power & Light (LP&L) Company is its owner-operator, and was responsible for the design and construction of the facility. Construction commenced on November 19, 1974, and was essentially completed by May 1984. The NSSS is a closed cycle, two loop system consisting of a g-s reactor vessel, two steam generators, four reactor coolant
's _) pumps and a pressurizer (Figures 1.4 and 1.5). It is similar to the systems utilized at the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. The nuclear core consists of 217 fuel assemblies each containing 236 fuel rods. The fuel rods contain slightly enriched uranium dioxide pellets (1.87 - 2.91 wt % U235 for cycle 1) clad in zircaloy tubes with welded end caps. Ninety-one control element assemblies (CEAs), consisting of NiCrFe alloy-clad boron carbide neutron absorber rods, are located in select fuel assemblies throughout the core.
The turbine generator consists of a tandem compound, six flow exhaust, 1800 rpm turbine using steam at 526.6 F and 860 psia. The generator is an 1800 rpm, three phase, 60 cycle hydrogen and water cooled unit, rated at 1,333,200 KW. The generator output feeds LP&L's 220 KV transmission system. Condensate rm cooling is provided by the Mississippi River, and will be (v) pumped throu'gh the plant at a rate of 1,400,000,000 gallons
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1 l Table 1.1 lists some of the major design parameters of WSES-3. i l [ l e r i l 1 - l I O i i l 6 l I
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26 TABLE 1.1 [a Part 1 of 3 DESIGN PARAMETERS OF WATERFORD-3 SES Hydraulic and Thermal Design Parameters, RCS Rated core heat output 3,390 GTh RCP heat input to RCS 20 %Th Total thermal power 3,410 ETh System pressure (nominal) 2,250 psia Reactor coolant flow rate 148 x 108 lb/hr Average coolant flow velocity along fuel rods 16.4 ft/sec Nominal core inlet temperature 553 F Nominal core exit temperature 611 F Average operating temperature (100% power) 582 *F V Fuel center temperature (maximum at 100% power) 3,420 *F Total reactor coolant system volume (without pressurizer) 10,300 ft 3 Core Mechanical Design Parameters Number of fuel assemblies 217 Fuel weight (as UO 2) 223,900 lb Total weight 310,744 lb Number of fuel rods 49,580 Number of control element assemblies (full / part length) 83/8 e f
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27 TABLE 1.1 O- (Continued) Part 2 of 3 7 DESIGN PARAMETERS OF WATERFORD-3 SES 1 Nuclear Design Data Core diameter-(equivalent) 136 in
. Core height (active fuel) 150 in Fuel enrichment Region 1 (cycle 1) 1.87 wt%
Fuel enrichment Region 2 (cycle 1) 2.38 wt% Fuel enrichment Region 3 (cycle 1) 2.88 wt% Total control element assembly worth (net) 11.35 %Ak/k t. Steam Generator Design Data (Each Generator, Full- Power)
\l
. Heat-transfer rate '
5.819 x-108 BTU /hr Steam-pressure 900 psia Steam flow rate 17.565 x 108 lb/hr Steam temperature 532 *F Fee'dwater temperature 445 *F Blowdown flow (maximum) 250 gpm Reactor Coolant Pump Design Data ' Flow 99,000 gpm Head 310 ft-Motor rating 9',700 hp
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\~ '1 (Continued)
Part 3 of 3 DESIGN PARAMETERS OF WATERFORD-3 SES Pressurizer Design Data Operating temperature 653 F Operating pressure 2,250 psia Internal free volume 1,500 ft 3 Normal operating water volume (100% power) 800 fta Normal operating steam volume (100% power) 700 ft3 Installed heater capacity 1,500 KW Maximum spray flow 375 gpm Continuous spray flow 1.5 gpm Containment Design Data *
'es' Inside diameter 140 ft Height 240.5 ft Free volume 2,677,000 ft 3 Reference accident pressure 44 psig Electrical Design Data Electrical power (gross) 1,153 MWE Electrical power (net) 1,104 MWE Diesel generator rating (each) 4,400 KW NOTE: The above are all desig+ 4. c3 r values only and do not necessarily reflect actual as-built values.
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29 ('% t ! LJ 1.1.3 The Test _Ersgram
\,-
The power ascension test program at WSES-3 was developed by LP&L and designed to fulfill the requirements of the NRC's Regulatory Guide 1.68, Revision 2, as detailed in Chapter 14 of the WSES-3 Final Safety Analysis Report (FIAR). The-objective of the test program was to determine the as-built plant characteristics during steady state and transient operation from cold shutdown conditions to 100% power, to; confirm certain design bases, and to demonstrate the plants' , ability to withstand those anticipated transients and postulated failures analyzed in the FSAR. The test program commenced with the initial fuel loading, and continued through the 100% power-test plateau. It culminated p) with the satisfactory completion of testing at 100% power. \_/ The program was divided into three categories, each of which is described below: a) Precritical (Post-Core Hot Functional) Testing 1 s g This consisted of a series of tests performed af ter the '
; fuel had been loaded, but before the reactor sustained its first critical operation, to allow a final evaluation of those systems requiring the core to be in place 5 4 (Examples are: i) CEA testing; ii) RCS flow and flow coastdown measurement). The plant was brought to hat' standby conditions (545 F, 2250 psia, k,gf < 0.99, and 0%
of rated thermal power) using RCP heat. Testing was
't performed at various plateaus of increasing temperature and pressure, with the bulk of the testing occurring at hot standby.
(D U
30 0 The Post-Core Hot Functional phase of the test program is summarized in Section 1.2.2, and detailed in Section 3.0 of this report. b) Low Power Physics Testing 1
' This consisted of a series of tests performed af ter the reactor was taken critical and sustained critical opera-tion without producing measurable nuclear heat. Core physics parameters were measured, and similarity between the WSES-3 and SONGS-2 cores was demonstrated. Based in
. part on this similarity (additional similarity was demon-strated during the power ascension test program), WSES-3 qualified as a follow-on plant to SONGS-2 (the C-E 3410 class reactor prototype plant), and was able to eliminate the following tests from its test program:
J-
- Pseudo-ejected CEA
- Dropped CEA
- PLCEA Xenon Control The Low Power Physics phase of the test r,ogram is summarized in Section 1.2.4, and detailed in Section 5.0 of this report, c) Power Ascension Testing
.This consisted of a -series of test performed at increasing power levels to make final adjustments / calibrations to equipment, to demonstrate satisfactory at-power operation of the plant, and to verify its ability to withstand operational transients. This phase of the test program demonst' rated catisfactory operation'of all plant systems
.f (x as an integral unit, and verified adequacy of plant operating and off-normal procedures.
31 v) The Power Ascension phase of the test program is summarized in Sections 1.2.5 through 1.2.8, and detailed in Section 6.0 of this report. The test program was conducted under strict adherence to test procedures, which directed the individual tests and documented all test data and results. Table 1.2 lists by title and number the test procedures used during the test program, and identifies the FSAR Chapter 14 commitments satisfied by a given procedure. The testing function ful-filled by the individual procedures is described in detail in the individual test descriptions of Sections 2.0 through 6.0 of this report. Table 1.3 lists major milestones of the power ascension test program, and Tables 1.4 and 1.5 list the PCifFT and the power I ascension tests and the plant conditions / power levels, respec-tively, at which each test was performed. Figure 1.6 shows the WSES-3 Cycle I power history from initial criticality through completion of the test program, while Figures 1.7-1 through 1.7-11 show significant events that affected the test program. t > x.J L
32 TABLE 1.2 [~'}/
- x. ,
Part 1 of 2 LIST OF STARTUP TEST PRECEDURES AND FSAR CilAPTER 14 TEST COMMITMENTS FILLED Seq. Procedure FSAR Ch. 14 No. Number Procedure Title Commitment
- 1. SIT-TP-400 Initial Fuel Load 14.2.10.1
- 2. SIT-TP-500 Post-Core Hot functional Controlling Document 14.2.10.1.3
- 3. SIT-TP-501 Intercomparison of PPS, CPC, and PMC Inputs 14.2.12.3.4
- 4. SIT-TP-502 RCS Flow and Coastdown Measurement 14.2.12.3.2
- 5. SIT-TP-503 CEDM Performance 14.2.12.3.1
- 6. SIT-TP-505 Pressurizer Spray Valve and Control Adjustment 14.2.12.3.5
- 7. SIT-TP-506 RCS Leak Rate Measurement N/A
- 8. SIT-TP-507 Incore Instrumentation Baseline Data 14.2.12.3.3
- 9. SIT-TP-508 RCS Heat Loss 14.2.12.3.6
- 10. SIT-TP-509 RCS Expansion Measurements 14.2.12.3.17
- 11. SIT-TP-511 Post-Core Test Data Record N/A
- 12. SIT-TP-512 Moveable Incore Instrumentation Operation Verification 14.2.12.3.3
- 13. SIT-TP-513 Post-Core Vibration and Loose Parts
(} \s_- 14. SIT-TP-600 Monitoring System Initial Criticality 14.2.12.3.40 14.2.10.2
- 15. SIT-TP-650 Low Power Physics Test
- 16. SIT-TP-700 14.2.12.3.10/11/12/13/14 Power Ascension Test Controlling Document 14.2.12.3
- 17. SIT-TP-701 NSSS Plant Data Record N/A
- 18. SIT-TP-702 Transient Data Record N/A
- 19. SIT-TP-704 RCS aT Power Determination 20.
14.2.12.3.27 SIT-TP-705 Nuclear and Thermal Power Calibration 14.2.12.3.27
- 21. SIT-TP-707 SBCS Capacity Check 22.
14.2.12.3.29 SIT-TP-708 Initial Turbine Startup N/A
- 23. SIT-TP-709 NSSS Calorimetric 24.
14.2.12.3.27 SIT-TP-710 RCS Calorimetric Flow Measurement 14.2.12.3.2
- 25. SIT-TP-711 Linear Power Subchannel Calibration 14.2.12.3.28
- 26. SIT-TP-712 Process Variable Intercomparison 27.
14.2.12.3.30 SIT-TP-714 Vibration and Loose Parts Monitoring System 14.2.12.3.40
- 28. SIT-TP-715 Biological Shield Effectiveness Survey 29.
14.2.12.3.15 SIT-TP-716 Core Performance Record 14.2.12.3.27
- 30. SIT-TP-717 CPC/COLSS Verification 14.2.12.3.27
- 31. SIT-TP-718 Variable Tavg 14.2.12.3.26
- 32. SIT-TP-721 Load Changes (Control Systems Checkout)
- 33. SIT-TP-723 14.2.12.3.31/39 Shape Annealing Matrix Measurement 14.2.12.3.28
- 34. SIT-TP-724 Temperature Decalibration Verification 14.2.12.3.28
- 35. SIT-TP-725 Radial Peaking Factor Verification 14.2.12.3.28
- 36. SIT-TP-726 Remote Reactor Trip with Subsequent
~m Remote Cooldown 14.2.12.3.33 U
33
/ 'N TABLE 1.2 klm (continued)
Part 2 of 2 LIST OF STARTUP TEST PRECEDURES AND FSAR CHAPTER 14 TEST COMMITMENTS FILLED Seq. Procedure FSAR Ch. 14 No. Number Procedure Title Commitment
- 37. SIT-TP-727 80% Total Loss of Flow Test / Natural Circulation 14.2.12.3.34
- 38. SIT-TP-728 Loss of Offsite Power Trip 14.2.12.3.35/41
- 39. SIT-TP-735 Incore Detector Signal Verification 14.2.12.3.3
- 40. SIT-TP-739 COLSS Power / Flow Verification Data Record N/A
- 41. SIT-TP-740 100% Turbine Trip 14.2.12.3.37
- 42. SIT-TP-741 Adjustment of COLSS Secondary Pressure Loss Terms N/A
- 43. SIT-TP-743 Ventilation Capability 14.2.12.3.32
- 44. SIT-TP-748 BOP Data Record N/A
- 45. SIT-TP-749 RPCS 50% Loss of Load Test 14.2.12.3.38
- 46. SIT-TP-750 RPCS 70% Loss of Feed Test 14.2.12.3.42
- 47. SIT-TP-751 RPCS 80% Loss of Load Test 14.2.12.3.38
- 48. SIT-TP-752 RPCS 100% Loss of Load Test 14.2.12.3.38
- 49. SIT-TP-753 RPCS 100% Loss of Feed Test 14.2.12.3.42
[~'h 50. SIT-TP-755 Natural Circulation Demonstration 14.2.12.3.25
\m l 51. SIT-TP-900 Pipe Whip Restraint Monitoring 14.2.12.3.17 nm
34 [ ') TABLE 1.3 8
": POWER ASCENSION MILESTONES EVENT TIME DATE Received Low Power Operating License NPF-26 1300 12.18.84 Commence Initial Core Load 2140 12.18.84 Mode 6 Declared for First Time 2155 12.18.84 Completed Initial Core Load 1400 12.24.84 Mode 5 Declared for First Time 2100 12.30.84 Commenced Post Core Hot Functional Testing ~1500 12.31.84 Mode 4 Declared for First Time 0226 1.23.85 Mode 3 Declared for First Time 1847 2.1.85 Completed Post Core Hot Functional Testing 1800 2.20.85 Mode 2 Declared for First Time 2013 3.4.85 Initial Criticality Achieved 2148 3.4.85 Commenced Low Power Physics Testing 0145 3.5.85 io) \_/
Completed Low Power Physics Testing Received Operating License NPF-38 1215 3.10.85 3.16.85 Commenced Initial Power Ascension 0345 3.17.85 Mode 1 Declared for First Time 1748 3.17.85 Initial Synchronization to Grid (@ ~10% power) 1813 3.18.85 20% Power Attained for First Time ~0750 4.12.85 50% Power Attained for First Time 2337 4.19.85 80% Power Attained for First Time 1845 5.7.85 100% Power Attained for First Time 1844 7.1.85 Declared Commercial Operation 0001 9.24.85 (O-
,-~ p Q
) N. O TABLE 1.4 POST-CORE !!OT FUNCTIONAL TEST PLATEAUS AND TESTS PERFORMED AT EACil PLATEAU TEST PLATEAU SEQ. m m NO. TEST (PROCEDURE NUMBER) p pc w r
- e si G - O w w w vi -
gh 3E O $ 8 0 $" 0
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4 4 1 Intercomparison of PPS,CPC and PMC Inputs (SIT-TP-501) X X X 2 RCS Flow and Coastdown Measurement (SIT-TP-502) X 3 CEDM Performance (SIT-TP-503) X X 4 Pressurizer Spray Valve and Control Adjustment (SIT-TP-505) X , 5 RCS Leak Rate Measurement (SIT-TP-506) X 6 Incore Instrumentation Baseline Data (SIT-TP-507) X X X X 7 RCS Heat Loss (SIT-TP-508) X 8 RCS Expansion Measurements (SIT-TP-509) X X X X X 9 Postcore Test Data Record (SIT-TP-Sil) X X X X 10 Movable Incore Instrumentation Operation Veri f. (SIT-TP-512) X 11 Post-Core Vibration and Loose Parts Monit. Sys.(SIT-TP-513) X 12 Ifeated Junction Thermocouple Operation Verif. (SIT-TP-500) X 13 RCS and Steam Generator Parameters (SIT-TP-500) X X X 14 Determination of Auxiliary Spray Flow Split (SIT-TP-500) X 15 Postcore Thermal Expansion Testing (SPO-99P-003) X X X X 16 Adjustment of COLSS Second. Press. Loss Terms (SIT-TP-741) X 17 Ventilation Capability (SIT-TP-743) X NOTE: An RCS Ileatup/Cooldown and Pressurization History (per SIT-TP-500) was recorded during plant heatup and pressurization. O
. _ - - _ _ ~ . - - . . - - . . . - O O TABLE 1.5 Part 1 of 2 POWER ASCENSION TEST PLATEAUS AND TESTS PERFORMED AT EACH PLATEAU TEST PLATEAU ~ SEQ. TEST (PROCEDURE NUMBER) o " ~ 20%- " * * *
~ "
NO. " M 50% 80% 100% S M S S $ $ $ M
$ E -
$ E -
$ E -
E E 4 i i 4 2 ? 2 ? 2 W 2 W W $ 5 5 2 1 Low Power Physics Test (SIT-TP-650) NSSS Plant Data Record (SIT-TP-701) X X X X X
]X ]X ] ]X ]X ]X ]X X
X _ ]X ]X ]X ]X X 3 kCS Delta-T Power Determination (SIT-TP-704) X X X X 4 Nuclear and Thermal Power Calibration (SIT-TP-705) X X X X X X X X X X X X X X X X 5 SBCS Capacity Checks (SIT-TP-707) . X 6 Initial Turbine Startup (SIT-TP-708) X X X X X X X _X X 7 NSSS-Calorimetric (SIT-TP-709) X X X X X X 'X X X X X X X X 8- RCS Calorimetric Flow Measurement (SIT-TP-710) X l X X
-9 Linear Power Subchannel Calibration (SIT-TP-711) X X 10 Process Variable Intercomparison (SIT-TP-712) X X X X 11 Vibration and Loose Parts Monit. Sys. (SIT-TP-714) X X X X X 12 Biological Shield Effectiveness Survey (SIT-TP-715) X X X X 13 Core Performance Record (SIT-TP-716) X X X X 14 CPC/COLSS Verification (SIT-TP-717) X X X X X
-15 Variable Tavg (SIT-TP-718) X X 16 Load Changes (Control Systems Checkout) (SIT-TP-721) X X X X X 17 Shape Annealing Matrix Measurement (SIT-TP-723) X X 18 Temperature Decalibration Verification (SIT-TP-724) X 19 Radial Peaking Factor Verification (SIT-TP-725) X 20 Remote Reactor-Trip with Subsequent Remote Cooldown (SIT-TP-726) X 21 80% Total Loss of Flow / Natural Cire. (SIT-TP-727) X 22 Loss of Of fsite Power Trip (SIT-TP-728) X 23 Incore Detector Signal Verification (SIT-TP-735) X X X X 24- COLSS Power / Flow Verif. Data Record'(SIT-TP-739) X X X X X X X X X X 25 100% Turbine Trip (SIT-TP-740)
X
\
TABLE 1.5 (continued) Part .' of 2 POWER ASCENSION TEST PLATEAUS AND TEST. PERFORMED AT EACH PLATEAU TEST PLATEAU SEQ. . TEST (PROCEDURE NUMBER). o u * *
~ ~
20% 50%
- NO. Load Changes (Control Systems Checkout (SIT-TP-721) " "
S M S" S S S" 80% S $ 100%
$ 2-5 2 -
$ 2-E E -
1 1 4 4 2 ? 2 ? 2 W 2 ? 5 5 5 5 26 Adjustment of COLSS Secondary Pressure Loss Terms (SIT-TP-741) X X X X X 27 Ventilation Capability (SIT-TP-743) X X X X 28 BOP Data Record (SIT-TP-748) X X X X X X X X X X X X 29 Natural Circulation Demonstration (SIT-TP-755) X 30 Pipe Whip Restraint. Measurements (SIT-TP-900)727) X _X X X 31- Thermal Expansion (SPO-99P-003) X X X X NOTES: a) Transient Data Record (SIT-TP-702) performed during every initial power increase.
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50 f% 1.2 SUMW RY 1.2.1 Initial EneLLaafd The initial fuel loading of WSES-3 was performed in accordance 4 with test procedure SIT-TP-400, " Initial Fuel Load". Fuel loading commenced on December 18, 1984, at 2140 (CST), approx-imately 8 hours af ter the facility received its low power e license from the NRC. The first fuel assembly (B077),
, containing the first of two neutron sources, was seated at core location X-11 at 2331. Fueling operations lasted 4 days, 17 hours and 20 minutes. The last fuel assembly was placed in the core at 1500 on December 23, 1985. The subsequent fuel i loading verification took until 1400 on December 24, 1984; its satisfactory completion marked the end of the fuel loading operation.
\_.)
The initial fuel loading is further discussed in detail in Section 2.0 of this report. I i 1.2.2 Pos_t Core Hot FuntLlonal Testing i The post-core hot functional test program was performed in accordance with test procedure SIT-TP-500, " Post Core flot Functional Test Controlling Document", and other test procedures of the SIT-TP-500 series, as listed in Tables 1.2 and 1.4 Testing commenced on December 31, 1985, and lasted approximately 52 days, until February 20, 1985, i Post-core hot functional testing is further discussed in detail in Section 3.0 of this report. 1 I , t_-) ,
51 /"'Tl 1.2.3 Init taLCtiticali1Y The approach to initial criticality was performed in accor-dance with test procedure SIT-TP-600, " Initial Criticality". Withdrawal of the CEA's commenced at 0328 on March 4, 1985. RCS dilution followed the CEA withdrawal until initial criti-g cality was satisfactorily achieved at 2148 on the same day. Initial criticality is further discussed in detail in Section 4.0 of this report. 1.2.4 Law _fAwcr_EhyJics Testing Low power physics testing was performed in accordance with test procedure SIT-TP-650, " Low Power Physics Test". Testing commenced, after initial criticality had been achieved, at (- / 0145 on March 5, 1984, and lasted approximately 5.4 days, until 1215 on March 10, 1985. Low power physics testing is further discussed in detail in Section 5.0 of this report. 1.2.5 Power Astension Tc1 Ling,.IhtaugL221fowgg Power ascension testing through 20% power was performed in accordance with test procedure SIT-TP-700, " Power Ascension Test Controlling Document", and other test procedures of the SIT-TP-700 series, as shown in Tables 1.2 and 1.5. Testing commenced at 0345 on March 17, 1985 with the initial power increase above the power levels maintained for low power physics testing. /% ( ) x_/
I 52 e O O The turbine generator was synchronized to the grid at 1813 on March 18, 1985, with the reactor at approximately 10% power. Twenty percent power operation was reached at approximately 0750 on April 12, 1985; this marked the first major test plateau. Testing at 20% power was completed at 1908 on April 18, 1985, when the initial increase to 50% reactor power commenced. The test results of the power ascension tests performed through 20% power are further discussed in detail in Section 6.0. 1.2.6 Power A:iggnsion Taline From 2.01.Thr0Eth_5fGwcI Power ascension testing through 50% power was performed in i fw
\
accordance with test procedure SIT-TP-700, " Power Ascension (_ / Test Controlling Document", and other test procedures of the e SIT-TP-700 series, as shown in Tables 1.2 and 1.5. Testing commenced with an increase in reactor power from the 20% power test plateau at 1908 on April 18, 1985. Hinor test plateaus were established at 30% and 40% po'ver and maintained for 9.25 and 9 hours respectively. Fifty percent power was achieved at 2337 on April 19, 1985. Testing at 50% power was completed at 1120 on May 6, 1985. , The test results <>f the power ascension tests performed from 20% through 50% power are further discussed in detail in Section 6.0. v [. l _ _ _ - _ _ - _ _ _ _ _ _ - _ _ _ _ . __ _ _ _ _ _ .
l 53
.i i
1.2.7 Pawar Ascension Testine From 50t Through 801 Po.ier i i .i 1 Power ascension testing through 80% power was performed in , j accordance with test procedure SIT-TP-700, " Power Ascension
; Test Controlling Document", and other test procedures of the i SIT-TP-700 series, as shown in Tables 1.2 and 1.5. Testing commenced with an increase in reactor power f rom the 50%
{ power test plateau at 1120 on May 6, 1985. Minor test plateaus were established at 60% and 70% power and maintained I for 10.5 and 9 hours respectively. Eighty percent power was ] achieved at 1845 on May 7, 1985. Testing at 80% power was completed at 0800 on June 26, 1985. r 4 The test results of the power ascension tests performed from 50% through 80% power are further discussed in- detail in i Section 6.0, i i ) 1.2.8 Power Ascension Testine From 801 Throuah'1001 Power 4 j Power ascension testing through 100% power was performed in accordance with test procedure SIT-TP-700, " Power Ascension i Test Controlling Document", and other procedures of the j SIT-TP-700 series, as shown in Tables 1.2 and 1.5. Prior to * [ commencing the power increase to 100% power, the 80% Total l Loss of Flow test, followed by the Loss of Offsite Power test j at 20% power were performed. The initial power escalation l from 80% to 100% power commenced at 0800 on June 26, 1985 l follow'ing completion of " post 80% power plateau testing" and - a return to criticality and 80% power. Minor test plateaus were es'tablished at 90% and 95% power and maintained for 1 approximately 8.2 and 4.9 hours respectively. One hundred l percent power was achieved at 1844 on July 1,1985. Testing l at 100% power was completed at 1730 on July 12, 1985. ).
1 i i ' i 4
- 54 '
i i ! The test results of the power ascension tests performed from j 80% through 100% power are further discussed in detail in ! i Section 6.0. . i l [ i ! i 1 r 1 j. t 1 5 4 i l J 1 t l i I h i 5 I i I l i i ! i l 1a i { b O . I I e S
. - . .2__ __ __- , _.mma e ___ _ ___..m -. _m _ _2 _a 55 6
l SECTION 2.0
. INITIAL FUEL LOADING J
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56 2.1 Erepa rations Initial preparations for fuel load commenced in the first quarter of 1983 with the receipt on site of the 217 fuel assemblies destined to make up the WSES-3 initial core. As the capacity of the new fuel storage racks is insuf ficient to accommodate an entire core, the fuel ' was stored dry in the spent fuel storage racks in a checker-board j arrangement. Following their receipt, the 5-fingered CEA's were ' j loaded into the fuel assemblies designated to host them during the first cycle (the 4-fingered CEA's, which straddle the fuel assemblies, were placed into the upper guide structure (UGS) when this was ready to receive them). Each core component was inspected during the process of removal from its shipping container and 3 placement in storage. The selected storage location also helped minimize the amount of handling / transferring required for each core i component. 4 O Prior to commencing fuel load, the containment refueling pool deep end, the transfer canal, and the. transfer pit in the fuel storage building were filled with borated water, at a concentration slightly 1 in excess of 2000 ppm, to a level of approximately one foot above the top of the transfer. canal. This was-accomplished utilizing the fuel pool purification system return line, thereby eliminating the need to overflow the reactor vessel. The chance of leakage around the seal ring, and a clean-up of the upper cavity floor to permit unrestricted access to the reactor vessel flange by fueling , . observers were avoided in this manner. Filling of the refueling pool, as described, assured containment integrity as required by the i Station Technical Specifications, and provided lubrication for the fuel transfer equipment. 3 O - p
^
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57 s
! w The reactor vessel was filled to approximately one foot above the ! top of the RCS houlegs with borated water, also at a concentration A
slightly above 2000 ppm. The shutdown cooling system was subsequent- [ ly maintained in operation as required by the Station Technical Specifications. Two temporary incore neutron detectors and associated electronics provided by Combustion Engineering were set up and calibrated. The detectors were placed in detector housings and se't in pla$e at< core locations V-7 and V-15 as channels "A and "B", respectively (Figuie 2.1.1). The elect;onics were set up as a neutron counting station (Figure 2.1.2) at the plant southeast corner of the refueling pool on the +46 foot level of containment, from where the reactor vessel and fuel loading operations could be closely observed. A strip chart recorder was connected to one of the channels to provide a continuous 7~ ' visual display of the neutron countrate in addition to the audible
\s countrate provided in containment by the other temporary channel, and one of the permanent plant start-up channels. The two permanent plant start-up channels had neutron' counting equipment connected to them that was set up within the control room, such that a total of four detectors would provide information on the neutron multiplica-tion throughout the core load. One of the start-up channels also
- provided an audible countrate in the contiol room, as required by the i
Station Technical Specifications. Following' satisfactory set-up and 4 checkout of both temporary and permanent plant start-up neutron detectors, a background countrate was determined for each detector without fuel or start-up neutron sources in containment. The response check of the neutron detectors required by the Station Technical Specifications was performed using the first fuel assembly (BO77),to be loaded with a start-up neutron' source'in one of its CEA guide tubes. The fuel assembly was . lowered into the reactor vessel adjacent to the permanent start-up and the temporary neutron g~)s g detectors, remaining grappled to the fuel handling machine at all
'. times. A neutron countrate significantly above the previously_
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58 N-O. A B C D E F GHJKLMMPR S T V W X Y l l l 1 i 2 3 4 5 8 7
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.j %
l LOCATION OF TE!?ORARY FUEL LOADING NEUTRON DETECTORS 'A' AND 'B' l FIGURE 2.1.1 l .
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60 m measured background countrate indicated a response to neutrons and verified operability of all four detectors. Upon completion of the response check fuel assembly B077 was placed in core location X-11. 2.2 Reactivity Monitoring The neutron multiplication of the core was closely monitored at all times during core load by means of an inverse multiplication versus number of fuel assemblies loaded ("1/M") plot for each of the four neutron detection / counting channels. After the first fuel assembly (B077) containing start-up neutron source "B" was inserted into the reactor vessel, at core location X-11, a base countrate, C,, was determined by averaging at least five individual counts over a 100 second period. All subsequent countrates, C g, determined by averaging at least three individual counts over a 100 second period, fg were divided into this base countrate; the resultant C /C salues 9 1 () were plotted against the number of fuel assemblies loaded for the four channels being monitored, providing the 1/M plots. A new base countrate was redetermined for the temporary detectors when these were moved. The new base countrate was normalized to the old one to assure continuity of the 1/M plots. Countrates were corrected for background if the background contributed greater than 5% of the countrate. Following the insertion of every fuel assembly into the core, a neutron count was taken on each of the four detectors. These counts were translated into inverse multiplication (1/M) numbers, and plotted against the number of assemblies loaded into the core to assure nuclear safety. A
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For approximately the first fifteen and the last five fuel assemblies to be inserted into the core, the 1/M value was determined prior to the refueling machine ungrappling from the fuel assembly. This was done because of the core coupling changes, that dramatically increase the neutron countrate during the first fif teen assemblies to be loaded into the core, and as a precaution while i loading the last five. i While loading the 16th through 212th assemblies a strip chart trace of the countrate off a temporary neutron detector channel was used , to determine visually any variations in suberitical neutron multipli- l cation. Based on the trend of this trace, it was possible to permit i the refueling machine to ungrapple from the fuel assembly placed into
-the core prior to the 1/M value having been determined. This method of monitoring the neutron multiplication provided additional safety and shortened the fuel loading operation by at least 12 hours.
Throughout the core loading, the neutron multiplication behavior was as expecte'd. Table 2.2.1 lists the countrates for all four neutron detector channels for the first. 20, and selected subsequent assemblies l inserted into the core. . I i
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. 62 '
TABLE 2.2.1 DETECTOR COUNT RATES (CPS) (Uncorrected for Background) NO ASSYS. ' TEMPORARY TEMPORARY STARTUP STARTUP IN CORE DET.A DET. B DET. 1 DET.2 BACKGROUND 0.0067 0.005 0.08 0.08 1 1.40 5.25 1.10 0.01 2 1.45 5.39 1.18 0.07 3 1.59 5.54 2.08 0.05 4 1.62 7.83 2.12 0.09 5 1.57 14.35 2.14 0.10 6 2.26 17.29 2.31 0.08 7(1) 3.27 17.37 2.34 0.10 8' 5.24 17.91 2.32 0.08 9 5.71 17.80 2.55 0.10 10 6.84 18.09 2.44 0.08 11 18.80 18.28 2.70 0.10 12 22.24 18.36 2.64 0.08 i 13 22.60 18.35 2.71 0.07 14 22.58 24.29 2.71 0.08 15 22.52 37.72 2.80 0.11 h)'% 16 17 22.93 23.47 122.52 48.07 55.48 2.65 2.92 0.05 0.27 18(2) 253.61 3.08 0.06 19 122.62 261.52 3.13 0.09 20 122.00 272.54 3.03 0.07 131 143.57 283.50 3.05 0.07 132(3) 142.25 282.72 2.99 1.25 133(4) 98.07 64.55 3.31 1.16 215 97.37 104.41 3.06 3.97 216(5) 97.90 -- 3.09 3.14 217(6) -- -- 3.05 3.29 ' NOTES: (1) Temporary detector cables relocated; new base countrate determined
-(2) Temporary detector cables relocated;-new base countrate determined (3). Startup neutron source "A" placed into core (4) Temporary detector "B" moved to core location D-15: new base countrate determined for both detectors (5) : Temporary detector "B" removed from core before placement of .this ? assembly -
O- -
)~ -(6) . Temporary detector "A" removed from core before placement of this assembly 1 .'
63 v i 2.3 The Fuel Loadine Sequrac.c The fuel loading sequence started at core location X-11 (reactor south side) where fuel assembly B077 containing neutron startup source "B" was placed as a free-standing assembly. Subsequent assemblies were loaded around the first essembly and, as loading progressed, around the temporary neutron detectors located at core locations V-7 and V-15, until a closely coupled slab nine assemblies wide had been formed. This slab was continued to the reactor north side, with loading alternating between an east and west direction. 4 Core location D-15 was lef t vacant to accommodate temporary neutron detector "B" following insertion of the fuel assembly (B031; the 132nd assembly.to be loaded) containing the second start-up neutron . source, "A". Core location Y-15, vacated by the relocation of i temporary neutron detector B", was filled af ter the slab had been completed on the reactor north side. After the slab was complete, the east side of the core was loaded,- followed by the west side, with leading occurring alternatingly ~in a north and south direction. 1 After the east and west sides of the core had been loaded, temporary neutron detector "B" was removed from the core and the hole filled by fuel assembly B011. Finally temporary neutron detector "A" was removed from the core and the last assembly (B055) placed in its ! location. A'two part core loading. verification verifying a) correct fuel and_ component location and orientation, and b) proper alignment , of the fuel assemblies was then performed. This completed the core loa' ding. Figure 2.3.1 depicts the loading sequence. . i \; l l
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21 ' 2 215 S/G #1 FUEL LOADING SEQUENCE FIGURE 2.3.1
65
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(O. 2.4 Fuel Movement L Fuel loading was executed by plant operations personnel. It was a supported by Reactor Engineering Department and Combustion Engineer-ing personnel. Fuel was loaded around the clock by three shifts per day. All personnel involved in activities involving the spent fuel handling machine, transfer systems or refueling machine were required , to wear paper shoe covers, paper coveralls, cotton glove liners and
- head covers to maintain cleanliness requirements. Those individuals who functioned as fueling observers in the refueling pool upper level (vessel flange), were required to wear a full complement of anti-contamination clothing, (i.e., shoe covers, cloth coveralls, cloth hood and cotton glove liners under rubber gloves). Access to refueling pool was governed by a Radiation Work Permit (RWP). All personnel' exiting the refueling pool area were monitored for
% contaminations by Health Physics.
v Fuel' loading officially began at 2140 on December. 18, 1984, when the the spent fuel tool was latched (grappled) to fuel assembly B077 containing Startup Source "B", in spent fuel rack GG-12. This assembly was utilized to perform the neutron response check as required by the Station Technical Specifications on the two startup-detectors and the two temporary detectors. The assembly was ungrappled from the refueling machine in core' location X-11 at 2331
- where it remained free standing until the second and third assemblies were placed in location Y-10 and Y-12 (Figure 2.3.1). Fuel loading i
then continued as described in section 2.3. Throughout-the fuel load the reactor water level was maintained between the top of the hot legs and the vessel flange. No major fuel-related problems or delays
. occurred. However, several equipment problems resulted in delays to fuel loading; these are' discussed in detail in section 2.6. 'Figu're 2.4.1 shows the time elapsed between individual fuel assemblies placed in the_ core.
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1 L \ . 3. Tha refueling machine ungrappled from the last ' fuel assembly to be I loaded into the core at 1500 on December 23, 1984. This completed the fuel movement portion of the initial fuel loading of WSES-3. A i fuel placement' and positioning verification followed. 2.5 Fuel Load Verification- , j The fuel load verification that followed the completion of fuel
- l movement utilized the refueling machine fuel hoist TV system, and consisted of: -
i a) Verifying all fuel assemblies, CEA's and start-up neutron sources were loaded into their pre-assigned core location and were i
- oriented correctly. To do this the core was scanned twice
- once to verify all fuel assemblies in their correct location with their serial numbers oriented to the southeast (SE), and to verify the CEA's and neutron sourcer in their correct host fuel 1 5 assemblies; the second scan verified CEA serial numbers and '
double-checked their core locations. These verifications were recorded on video tape. With the exception of the 4-fingered CEA's which were loaded as -an integral part of the UGS, all fuel t 4-assemblies and core. components were verified correctly loaded. ! This verification required 11.3 hours. Figures- 2.5.1, 2.5.2 and 2.5.3 show the as-loaded core. I i- ,
- b) Verifying the position of the fuel assemblies to assure *
- j. alignment with the fuel' alignment plate of'the UGS. The position of fuel assemblies with respect to the centerline of selected rows (6, 16, C, F, L, S, and W) in the core was measured, using- ,-
[ the refueling machine fuel hoist TV camera. The data showed the fuel to have been loaded acceptably to allow the UGS to be t i installed into the reactor vessel. This~ verification required 11.5 hours. -i i e f g
+ 68
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N A B C D E F GHJKLMNPR S T V W X Y l 1 C012 C039 CO22 C013 2 C002 C011 C216 C201 B065 C214 C202 C017 C039 3 C005 C102 B047 A058 B015 A071 8061 A010 B009 C107 C037 4 C001 A063 B078 A023 B003 A036 B001 A029 B064 A026 B074 A004 CO23 5 C032 C104 B014 A064 B049 A033 B042 A046 B019 A024 B062 A013 B010 C101 CO26 6 C020 B037 A034 B002 A031 B056 AC50 5054 A056 B032 A022 B068
. A03' B008 CO25 7 C215 A001 3024 A040 B037 A028 B041 A054 B026 A038 B020 A035 B055 A005 C204 9 C019 -
C038 9 C213 B033 A041 B027 A055'B022 A067 B051 A066 B048 A025 B021 ,A020 B059 C205 10 C034 C018 11 B031 A027 B030 A043 B073 A032 B070 A039 B076 A062 B017 A068 B080 A016 B077 12 CO27 C009 13 C209 B066 A070 B025 A072 B067 A045 B053 A061 B018 A047 B016 A021 B043 C203 14 C014 C031 15 C212 A012 B011 A042 B072 A053 B071 A049 B044 A052 B069 A065 B079 A018 C207 16 C007 B063 A006 B045 A017 B075 A051 B052 A050 B023 A057 B036 A011 B050 C024 17 C033 C103 B029 A073 B034 A059 B058 A044 B004 A069 B012 A015 B039 0108 C035 IE CO21 A002 B005 A014 B007 A009 B046 A019. B035 A007 B028 A003 C040 19 C036 C106 B013 A048 B006 A008 B063 A030 B038 C105 C016 20 C004 C006 C210 C211 B040 C208 C206 C003 C010 21 C028 C030 C008 C015 S/G #1 o WSES-3 CYCLE 1 CORE MAP (?UEL1 FIGURE 2.5.1
..-m- - - - - , . , , , . - - - - - - - , , , . - .
69 O N A B C D E F GHJKLMNPR S T V W X Y i 1 8 1 2 10 43 3 76 38 8 58 72 4 62 42 4 24 2 34 5 37 77 C 13 H 46 57 6 61 33 71 50 31 18 3 7 7 B 73 70 23 D 5 8 9 51 52 54 75 6 32 68 10 11 v 1 40 41 44 78 63 16 ? 13 39 15 74 84 45 28 29 14 15 22 A 71 65 25 C 48 i 16 Si 19 17 67 69 64 17 36 60 F 26 E 11 49 19 35 12 53 - 9. 14 47 i
- 19 56 66 21 30 3 I i
20- 55 27 21 83 S/G #1 o WSES-3 CYCLE I CORE MAP (CEAsl . O FIGURE 2.5.2 l
f 70 0
-N A B C D E F GHJKLMNPR S T V W X Y 1
2 j l 3 4
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7 e i 9 10 i 11 13 14 15 18 17 _ 1 18 l 19 l 20 21 S/G #1 1
-A = Neutron source 'A' in northeast guide tube of fuel assembly B031
- B = Neutron source 'Bin southwest guide tube of fuel assembiv B077 o WSES-3 CYCLE 1 CORE MAP (NEUTR0! SOURCES)
FIGURE 2.5.3 I
~
l
71
\
f 2.6 Delavs. Problems and Resolutions 2.6.1 Refueling machine fuel hoist underloads were experienced during fuel assembly insertion into the core at various times. 4-In particular, fuel assembly B043 required repositioning of
- the refueling machine, and 29 minutes were required to finally seat the assembly in the core. Fuel assembly A016 required 14 minutes of effort and C207 took 48 minutes. Other 1
fuel assemblies generated underloads, however none required appreciable time to correct. i Total Time Lost: ~1 hour 31 min. 2.6.2 Relocation of the handling / tie off ropes and detector cabling on the two incore detector assemblies consumed approximately r- 40 minutes time, in addition to the " normal" relocation and/or _f g
\% ,/ removal.
Total Time Lost: ~ 40 min. 2.6.3 The containment audible count rate speaker for the temporary counting station failed, resulting in a suspension of core alterations. In addition to the time required to replace the speaker - I hour -and 41. min. -' it was determined that the spent fuel handling machine operability checks would have to be reperformed, resulting in an additional delay of 3 hours 15 min. Total Time Lost: 4-hours 56 min.
-2.6.4 As a result of an overheated power cord on a "T" bar I
' underwater light assembly, the unit was removed from within
(~~N the-reactor. vessel. ' Loading activities continued, utilizing ( ,) : the fuel hoist TV camera lights, with no delays or problem. i r , ,- , e n ,-n- - , ,- , , , , a-,.
4 72 3 Later, however, it was discovered that a (-inch nominal hex nut was missing from the T-light pivot brackets. A search of the core and lower plate was made with an underwater TV system. The nut was not fosnd. Total Time Lost: 53 min. 1 2.6.5 A rigid coupling on the spent fuel handling machine bridge e drive failed. A replacement coupling was obtained off-site, machined and installed and the' spent fuel machine returned to j se rvice . After less than 8 hours of operation another drive shaft. coupling failed. Closer inspection disclosed that the C" flange mount for the : gear reducer / motor unit had worked '
- loose from'the gear reducer, allowing the gear reducer / motor and output shaft to become badly misaligned with respect to l
the two drive shafts. The decision was made to remove the s complete bridge drive train for repair. In the-meantime, i attempts would be made to move the bridge manually (pushed / pulled by two fuel loading personnel). The bridge was propelled this way for the remainder of the fuel load, approximately 124 fuel assemblies.
~
Total Time Lost: 16 hours 12 min.- ' 2.6.6 The 24-volt control voltage to the refueling machine control microprocessor failed when a fuse blew, while a fuel assembly
- was being lowered into the core,.in the lower slow zone. The
! fuel. hoist' continued to operate in the "down" mode. Main [ power to the machine was turned off, but not before the-mechanical programmer sustained damage from being 'overdriven. ; i-
73 r A cam operated switch was damaged, a coupling shear pin sheared, and cams and bearings were displaced. A replacement switch was removed from the CEA hoist mechanical programmer, , a new shear pin was fabricated, the bearings and cams were repositioned and adjusted, and the machine was returned to service. Total Time Lost: 6 hours 8 min. Total delay, during the core load, based upon the above identified problems was just over 29 hours.
! f
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r 4 74 P i i i l t I i SECTION 3.0 1 i POST-CORE HOT FUNCTIONAL TESTING , 8 h i 8 1 4 l i 1 i 1 i i r i I e i f i 4 l r F ( t F i e O O _ _ ..,7,,,_e,-m. _.w.., .--__,--r__., ,.%.. ,,m__,,, , . . , , ,._ ..__.,-_,_,-__~._,---.._.m._ .. . . . . , . . , _ . . _m.m_.e.,,y,- _ . , _ _ _ _ . . - . . ~ .
75 ' :() s/ s 3.1- INSTRUMENTATION TESTING / CALIBRATION , 3.1.1 'Intercomoarison of PPS. CPC. and PMC Inputs (SIT-TP-SQll PURPOSE: j The purpose of this test was to demonstrate that the inputs and appropriate outputs of the Plant Protection System (PPS), the Core Protection Calculators (CPC's), and the Plant Monitoring Computer (PMC) were in satisfactory agreement with
- one another. Permanent plant instruments (meters and ,
recorders) were also intercompared. ; j- This test satisfied the commitments of FSAR section
~
14.12.12.3.4. METHOD: i' Plant conditions were stabilized'at each of the three test
- . plateaus -- 120 F, 345 F, and 545*F -- during the heatup
, following initial fuel load Data from each of the four f sources (PPS, CPCs, PMC and meters) were simultaneously gathered for each of the following parameters:
- 1. RCS cold leg temperature
-2. RCS hot leg temperature
- 3. RCP differential pressure
- 4. RCP speed-
- 5. RCS pressure-
- 6. Pressurizer level -
-7.= Steam generator.-level
- 8. - Steam generator. pressure I
- q
) 9. Steam generator primary side differential pressure ;
- 10. Reactor. vessel differential pressure l
9 l
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- 11. Containment pressure
- 12. Refueling water storage pool level Based upon the data gathered for each parameter, a target value was calculated as the average of the readings from the most reliable source; the order of reliability of data sources, from most reliable to least, was as follows:
- 1. Core Protection Calculator data
- 2. Plant Protection System data
- 3. Plant Monitoring Computer data
- 4. Control Board Instrumentation Data The deviation of each recorded value from this target value was calculated and compared to the specified tolerance to g determine acceptability. If the deviation exceeded the speci-N,,) . fied tolerance, recalibration of the loop was initiated and a test deficiency was generated. The deficiency was cleared only when subsequent testing revealed that. the parameter de-viation fell within the specified tolerance.
RESULTS: At the 120*F plateau, four deficiences were generated repre-senting forty-nine parameters' failure to meet specified criteria. Of these forty-nine, twelve were attributable to the inoperability of the Qualified Safety Parameter Display System #1 (QSPDS #1) - PMC data link, and eight were attri-
. butable to the fact that the feedwater control systems were de-energized during the performance of the data collection.
. Evaluation of the remaining erroneous indications was performed, and recalibration was initiated where necessary;'
some parameters' -specific tolerances did not reflect the [) V actual loop accuracies, and were changed accordingly.
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, 77 - f)
U At the 345'F plateau, a total of twenty-two parameters failed to meet their specified tolerances. Fourteen of these had also been deficient at the 120 F plateau. Four new deficien-cies were written to document the eight new failures. Trcubleshooting and recalibration of the problem indications continued. At :the 545*F plateau, a total of fif ty parameters failed to meet their specified tolerances. The fifty parameters fell into the following three categories:
- 1. RCS Hot and Cold Leg RTD indications (22) - The safety-related RTD's which provide hot and cold leg temperature input to the Core Protection Calculators were all offscale at. the two earlier temperature fs . plateaus. Thus, data recorded at the 545*F plateau
(,,/ provided the first indication of problems with these indications. Extensive troubleshooting, recalibra-tion, and rework of these RTD's continued throughout the power ascension ' test program, and a detailed
' history of this problem is given -in section 6.2.2 of this report.
- 2. Remote Shutdown Panel Instrumentation (2) - Two-
-indicators located at LCP-43, the temote shutdown panel, require transfer of pressurizer pressure and level control from the control room to LCP-43. At the time-the 545'F: test data were recorded, operations was unable to support this transfer. Data for these two .
instruments, RC-ILI-0110-1 and RC-IPI-0100-1, were successfully recorded during the performance of SIT-TP-712,.the equivalent of this procedure which was t-
,, performed- during power ascension -(see section 6.2.2).
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- . . - .- -. - . - ~ - - . .. . .. - - . - ,
78 m n 3. Miscellaneous Indications (26) - Thirteen PMC points, x ten control board meters, and three PPS inputs failed to meet their specified tolerances. Troubleshooting of these parameters continued while low power physics testing was conducted and during the wait for the unit's full power operating license. All 26 para- , meters were successfully tested in accordance with SIT-TP-712. I Twelve deficiencies were written at the 545 F plateau to document the fifty out-of-tolerance parmeters; thirteen of these fifty had failed previously at either the 120'F or 345'F plateau. 1 CONCLUSIONS: Of the fif ty out-of-tolerance indications remaining at the completion of the test, twenty-three were safety-related. Only four of these were not related to the CPC hot-and cold I leg RTD's; these four' indications were resolved and retested satisfactorily prior to initial criticality. i
/ An evaluation of the impact of the out-of-tolerance CPC RTD's , was performed, and it was determined that power operation at levels up to 20% would not be restricted. Evaluation of the RTD problems continued with the performance of SIT-TP-712 '
(see section 6.2.2). e i. i e
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i L
79
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a . ;(J) . n E 3.1.2 Incore Instrumentation Baseline Data (SIT-TP-507) PURPOSE: 1 This test was performed to verify that the resistance of each incore detector and background detector and their associated cabling at operating conditions was equal to or greater than 1x107 ohms. The test also collected baseline data for core exit thermocouple temperature readinge during plant heatup. l This test satisfied in part the commitments of FSAR section , 14.2.12.3.3. i METHOD: ' b -The test was. performed from January 9 through February 9, 1985. With the reactor coolant system at. normal hot standby
- operating conditions ( >525*F and 2250 1 15 psia) the detector resistances were measured. Each detector cable was i
removed at the input of the amplifier card and, using a high potential (<50V) resistance meter, the individual detector and background detector resistances were measured. 2 i At various times during the heatup, both at stable temperature plateaus and during heatup transients, the core exit temperatures were recorded using the thermocouple in each detector string. Data was recorded via computer f printouts. ( I E R 6
)
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RESULTS: All incore detector resistance values were greater than 1x107 ohms. Cable connectors at the input to amplifiers E-4 and E-6 required rework before they could be measured. Baseline core exit thermocouple data was collected at various temperatures throughout the heatup. CONCLUSIONS: i The resistance reading for each incore detector and background detector was satisfactorily verified to be greater than 1x107 ohms, thereby indicating negligible impact on the incore signals from current leakage. Adequate baseline data was collected for the core exit thermocouples during plant !
,.s heatup for future reference. All test objectives and
_, ) acceptance criteria were met. , 4 r h b 4 i h
. i r \~-)\ i
81 t I
\j 3.1.3 Moveable Incore Instrumentation Operation Verification (SIT-TP-512)
PURPOSE: The purpose of this test was to: a) measure the movable incore detector guide tube path lengths with the reactor coolant system cold ( < 120 *F) for paths 18 and 23 only b) to measure the guide tube path lengths with the reactor coolant system hot ( > 525 F) for all 56 paths using drive machines 1 and 2
,es (s,) c) to operate the movable incore detector system (MICDS) from the control room using the plant monitoring computer (PMC) as the controller d) to demonstrate the mechanical operation of the movable incore detector system.
This test satisfied in part the requirements of FSAR Sections 14.2.12.2.58 and 14.2.12.3.3. METHOD: Measurements of the guide tube path lengths for cold and hot RCS conditions were performed using the manual control box (MCB). with a dummy detector cable installed in the drive machine being tested. The dummy detector cable was inserted p_ - in the selected guide tube path until the encoder reading (v ) stopped changing. The encoder reading was recorded as step
#1. The dummy detector cable was then withdrawn approximately W
82' a 20 inches and reinserted until the encoder reading stopped
-changing. This second encoder reading was recorded as step
#2. If the difference between step #1 and-step #2 encoder readings was greater than 0.3 inch, the withdrawal / reinsertion was repeated and a third encoder reading was recorded as step
#3. The average of the two or three encoder readings was recorded. The-average reading was taken as the guide tube path length.
Cold guide tube path measurements were taken for paths 18 through 23 using drive machine 1 in the normal and alternate configurations through transfer machines 1 and 2 (a total of 12 measurements). These-measurements were taken to clear a deficiency from preoperational test SP0-65C-001. Hot guide tube path measurements were taken for all paths !
- \m / using drive machines 1 and 2 in the normal and alternate configurations through transfer machines 1 and 2 (a total of 112 measurements). Data from the hot guide tube-path measure-
-ments was incorporated into- the MICDS software on the (PMC).
The MICDS was operated from'the control room using the PMC MICDS software as the controller in the manual mode using drive machines.1 and 2 and in the semiautomatic mode using drive machines 1 and 2. Proper operation of the MICDS and detector positioning to within 0.3 inch of the desired I position were verified. .Due to the availability of only one 1 good dummy detector cable, PMC MICDS software operation was performed using both real detector cables l installed in drive machines 1 and 2. I L k -) m l g
. . , , . , . _ . . . - _ . . , _ , - - - , - _ ,, .- - r,,
83 5
. s.
(, /m) - RESULTS: I 1 Data from the hot MICDS guide tube path measurements yielded the following results:
- 1) The encoder readings were repeatable generally to within O.1 inch, and at worst to within 0.4 inch.
- 2) Hot path measurements were 0.1 to 0.5 inches longer than the cold path measurements taken for paths 18 through
- 23. This difference can be attributed to thermal expansion.
~
- 3) Hot path measurements were consistent with the cold path measurements taken in preoperational test SP0-65C-001.
.The only exception was path 6 for transfer machine B with s,/- readings which differed by-about 20 inches. Based on the repeat' ability of the data taken during this test, the preoperational \;. test data is deemed to be incorrect.
- 4) The difference between the normal and alternate configurations.was small and can be considered to be zero.
5)
~
The MICDS-hardware operated satisfactorily using the manual control' box. j The MICDS software operated in the manual and semiautomatic modes, as required by this test. Two problems which affect operation using the MICDS software were discovered: a) The Transfer Enable switch did not indicate a "Not Enable" state when the detector was inserted past.the switch. This problem would affect operation in the () i automatic mode only. s, [ T , l' i \ l xx ' t.. * . l . , 9 .- - __ _ .. _ . _ - - ~
l i 84 l l b) Path verification alarm messages occurred when there were no apparent failures. This problem would affect MICDS operation in all modes. L Although these two problems were not entirely resolved, operation of the MICDS in manual and semiautomatic modes was satisfactorily demonstrated. CONCLUSION: The MICDS operated as required by this test. All test objectives and acceptance criteria were satisfactoril / met. O J 9 9 e
85 f*"'s
)
s~ - 3.1.4 Post-Core Vibration and Loose Parts Monitoring System (SIT-TP-513) PURPOSE: To establish steady state vibration and loose parts moniUJr,ing baseline data for the four reactor coolant pumps (RCPs), the two steam generators, the reactor lower vessel and reactor upper vessel under various RCP configurations. This test satisfied, in part, the commitments of FSAR Chapter 14, Section 14.2.12.3.40, Baseline Vibration and Loose Parts Monitoring. MITHOD:
) \--
Data was recorded on cassette tapes via the vibration and loose parts monitoring system's (V&LPMS) tape recorders, durtrg t stable RCP configurations established per SIT-TP-502, Postcore RCS Flow and Coastdown Measurements. Each channel of the recorded data was then analyzed using a spectrum i analyzer, and plotted using an X-Y Plotter to generate the power spectral density (PSD) signatures. l
. l RESULTS:
1 This test was performed on February 10 and 11, 1985 during the performance of SIT-TP-502,- as discussed above. A total of thirteen cassette tapes were used to record this data. There are two sections of recorded data, each containing four different channels, for a total of eight channels of data on the:first eight tapes. Several tapes were played back through gy the audio monitor during the test,. and it was discovered that i
\_,I several channels have very high levels of background noise.
i k s. e e
i 86 1 N ( An evaluation of this problem by a technician produced no resolution. To circumvent this problem, the noisy channels were switched to other tape tracks and recorded on separate tapes while the problem evaluation continued. This strategy was successful, and good results were obtained. CONCLUSION: Baseline data was recorded for all RCP configurations specified in SIT-TP-502, and the data acquisition acceptance criterion was thus satisfied. Evaluation of the PSD's will be performed following the installation and calibration of a new spectrum analyzer. /"%
- \
V e
87
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l )
\s_ s' 3.2 REACTOR COOLANT SYSTEM TESTING 3.2.1 Reactor Coolant System Flow and Flow Coastdown Measurement (SIT-TP-502)
PURPOSE: The purpose of this test was to: a) Determine the as-built post-core reactor coolant system (RCS) flow rate b) Determine the post-core flow coastdown characteristics f'~~') t
'/ and to verify that the flow.coastdown is consistent or conservative with respect to the coastdown characteristics assumed in the safety analysis c) Verify the validity of the flow-related algorithms and constants in the core protection calculator (CPCs) and the core operating limits supervisory system (COLSS) d) Establish reference post-core differential pressures (APs) within the RCS This test satisfied the requirements of FSAR section 14.2.12.3.2 METHOD:
,- y This test was. performed at nominal hot standby conditions of
(_-) 545 F and 2250 psia. The measurements were made through a sequential combination of steady state and transient flow 8 t .
88
'CA v i t
conditions as depicted in Figure 3.2.1.1. Steady state measurements were those made with a stabilized RCS flow rate provided by either '1, 2, 3, or 4 reactor coolant pumps (RCPs) running. These configurations provided data for the determination of the RCS flow rate, the verification of the CPC and COLSS constants, and the establishment of reference post-core RCS APs. Transient measurements were those made following the trip of 1, 2, or 4 RCPs, while the RCS flow was changing from 4 one steady state configuration to another. These configurations provided data for the determination of the flow coastdown characteristics. Within 15 days of commencing the test, all twenty RCS AP 2 transmitters (8 for the RCPs, 4 for the reactor vessel (RV), and 8 for the steam generators (SG)) were calibrated to
-% provide accurate pre-icst calibration data. Following N,/ completion of test data collection the transmitters were all calibration checked to provide instrument drift data. Any drift data was subsequently figured into the flow
- calculations.
- The four-RCP steady state RCS flow rate was determined by two different methods
i i) using RV APs ii) using RCP APs The RV AP method result, being the more accurate of the two, was used to meet the RCS flow rate acceptance criteri'n. o The RCP AP method value was required for adjustment of COLSS flow constants, and was compared to the RVAP method value for info rmation. Data collection for both methods consisted of i recording RCS AP, RCS temperature and RCS pressure data ' ('~) ( ,j concurrently on a high speed test data acquisition system _ (TDAS) at a rate of I sample per second. Backup data was j -- _ - - _ - - - - - - - - - - - - - - . _ _ _ _ _ _ _ _ - - - - - _ _ - _ - - _ - - _ _ _ - - _ - - - - - - - _ _ _ - - - - - - . , , - - - , , - - - - - --,,--a.-------._--.u,_- . - - - - _ _ . - - --_---_.---___u.--
Ot . O - O
= DATA COLLECTION PolNT/
TEST RUN NUMBER f ALL ALL ALL ALL i . 4 i b
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- IA IA IB 2A d 2B IB 28 2B g
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- RCS FLOW AND FLOW COASTDOWN MEASUREMENT TEST SEQUENCE FICIIRE 3. 2.1.1 i
90 e j L ("')N
\m, recorded using the plant monitoring computer (PMC) and strip chart recorders. Averaged-AP data was calibration corrected '
before the RV AP data was normalized to 545*F and 2250 psia, and'the RCP AP data was normalized to a reactor coolant medium j specific gravity of 0.75, such that all measured aP data was , compatible with the respective flow vs. differential pressure ,- curves from which the RCS flow rate was determined. During every steady state configuration not previously established, process noise data was recorded on. each of the three data collection devices for information and possible
- application during test data evaluation. This was accomplished 1
by recording data at high speed (20 samples per second on the TDAS, I sample per second off the PMC, and at ~ 10mm/second on the strip charts) simultaneously on all recording devices for
- -x a predetermined period of time.
l \--
- Three flow coastdown measurements were performed
1 i) a 1-RCP trip flow coastdown
- - ii) a 2-RCP trip flow coastdown iii) a 4-RCP trip flow coastdown The 1-RCP trip flow coastdown was initiated from a 4-RCPs running steady state configuration by turning the RCP 2A f switch to the "STOP" position. RCP 2A was selected based on -
.the requirement to investigate the loss of the strongest RCP, as determined during the pre-core RCS flow measurement. This trip test' collected data to. verify the coastdown due to a
- locked rotor.
1 d
_ -. .~ .. . 91
'Q The 2-RCP trip flow coastdown was initiated from a 4-RCPs running steady state configuration by simultaneously turning the RCP 1A and 2A switches to the "STOP" position. This trip test collected data to verify the coastdown due to a loss of power from a two pump bus.
-The 4-RCP trip flow coastdown was initiated from a steady state configuration by simultaneously tripping all four RCPs 4
from a previously temporarily installed special
" total-loss-of-flow" (TLOF) trip switch. This trip test collected data to verify the coastdown due to a total loss of
. forced reactor coolant flow.
Dats collection for all three coastdowns consisted of record-ing RCS AP, RCS temperature, RCS pressure, RCP shaft speed, (~~g and RCP breaker status data concurrently on the TDAS at a rate
'\- , of'20 samples per second. Backup data was recorded off the PMC and on strip charts. The TDAS was then used to calculate ; the flow coastdown at 50 msec intervals, using the data that it has previously collected. The calculation resul'ts provided the . input for the flow coastdown curves, whose plotting was optional for the 1- and 2-RCP coastdowns, but required for the 4-RCP coastdown (Figure 3.2.1.22). The plotted curve (s) allowed evaluation of the shape of the measured curve (s) with respect to the'~one(s) assumed in the safety analysis to assure conservatism. An evaluation of the time (Teo) required by the tripped RCPs to reach 90% of rated speed (<1070 rpm) was also performed as part of the verification for conservatism. For each transient test Too was determined for every RCP that had been tripped. The largest Teo for a given transient was compared-to a table of Tso vs. COLSS EPOLI (Constant for power operating limit uncertainty) penalty factors specific for that 7-s transient, to determine the magnitude of the COLSS penalty
(, f - factor required to be implemented into COLSS to assure con-servatism for that parti,cular flow coastdown. Af te r . all i
92 n,
\_
transient testing was complete the previously determined COLSS penalty factors were compared to each other, and the largest, enveloping all others, selected as the one to be implemented into COLSS until satisfactory completion of the 80% total loss of flow test (see section 6.6.4). This approach of determin-ing a COLSS penalty factor in place of the CPC core coolant mass flow rate calibration constant FC2 (CPC PID 061) to assure conservatism was the result of investigations made by Combustion Engineering to facilitate and expedite the post-core flow measurement based on previous performance of this test at Arkansas Nuclear One Unit 2, and San Onofre Nuclear Generation Station Units 2 and 3. Following the satisfactory completion of the test sequence, the measured RCS flow rate was used to make the initial (N adjustments to the CPC and COLSS flow constants. The CPC \- core coolant mass flow rate calibration constant FCI (CPC PiD 060) was adjusted for each CPC channel such that the base core coolant mass flow rate constant MDBAR (CPC PID 265) for that channel reflected the calculated normalized measured flow rate value +0.000, -0.005. The COLSS positive flow bias constants D15(1) through D15(4) were adjusted such that for each RCP the dif ference, AF(j) (with j = l-4), between the COLSS calculated individual RCP average volumetric flow rate and the measured individual RCP volumetric flow rate normalized to the total vessel flow rate was -396 gpm $ AF(j) gpm i +396 gpm, ,ND 3 the difference, AF(RV), between the COLSS calculated average RCS flow rate and the measured RCS flow rate as a percentage of design flow was -0.2% 1 AF(RV) $ +0.2%. A kv )
, ~._ -, . . - .. .-
i ; i i 93 3
\
RESULTS: i This test was performed twice. Its first execution was , . terminated following completion of the steady state and transient flow calculations prior to. adjusting the CPC and COLSS constants, because data evaluation revealed a significant difference in'the pre-test calibration and post-test calibration check AP data, making the 4-RCP steady state and 4-RCP flow coastdown test results highly questionable. The 1-RCP and 2-RCP flow coastdown portions were unaffected by this deficiency, because only Tso, which is independent of
- the AP transmitter calibration, was used in the determination i
a of flow coastdown conservatism. Furthermore, the establishment of baseline post-core differential pressures l within the RCS does not require repeatability to the same i fr s degree of accuracy required for the valid determination of a
- 'A_s 4-RCP steady state RCS flow rate and validation of the 4-RCP i
flow coastdown curve assumed in the safety analysis, and was therefore not required to be repeated either. Thus only the 4-RCP steady state and flow coastdown portions of the test were reperformed, following a recalibration of the AP l transmitters. i i The steady state RCS flow rate was satisfactorily determined 2
-during the retest. The test results are shown in Table 3.2.1.1.
i 5 i i I L (~~T v . g . r l
- 94
/~'s t ) %.J TABLE 3.2.1.1 4-RCP STEADY STATE PCHFT RCS FLOW RATE MEASUREMENT TEST RESULTS RCS FLOW RATE ACCEPTANCE CRITERIA Normalized to [gpm)
METHOD gpm 148x106 lbm/hr RV AP 449648.6 1.1495 418400 $ FLOW $ 452800 RCP AP 447458.9 1.1439 N/A The values in Table 3.2.1.1 were calculated using only pre-test AP calibration data in order to expedite completion of testing and data evaluation. When post-test AP calibration check data became available, this was compared to pre-test data in the form of plots included in this report A () for documentation as Figures 3.2.1.2 through 3.2.1.21. The satisfactory repeatability eliminated the need to reperform the flow calculations to compensate for instrument drift. It should be noted that RCP AP transmitter PDT-110 for RCP 1 A (Figure 3.2.1.2) was not used in performance of the flow calculations due to its erratic response. The actual four-RCP trip flow coastdown was much more conser-vative than the flow coastdown assumed in the safety analysis. The conservatism of the shape of the measured coastdown (Figure 3.2.1.22) was, however, questionable due to the electronic noise components of the measured parameters incor-porated into the flow calculations. Although the best fitted flow coastdown curve was satisfactory not all of its data points lay above the FSAR flow coastdown curve. Thus, to assure that COLSS be conservative and the plant be operated in its analyzed operating space, the value of -4.7175 determined i i
/S
- G) l I
95 m (v ) for EPOLI during all coastdown testing, was increased to
-7.0000. This value was implemented and retained until later power ascension test results allowed a decrease to this penalty factor to be made.
The initial adjustments to the CPC and COLSS flow consients were satisfactorily determined and implemented into the respective data bases, as required. Tables 3.2.1.2 and 3.2.1.3 list the as-left CPC and COLSS constants, respectively. TABLE 3.2.1.2 AS-LEFT PCIET CPC FLOW CONSTANTS s/ CPC CHANNEL FLOW CONSTANT FCI FC2 MDBAR A 1.1213 0.0 1.1488 B 1.1212 0.0 1.1'482 C 1.1218 0.0 1.1490 D 1.1209 0.0 1.1489 f TABLE 3.2.1.3 - AS-LEFT PCIET COLSS FLOW CONSTANTS CONSTANT CONSTANT DESCRIPTION 'VALU2 D15(1) - RCP 1A -410.1 D15(2) - RCP IB -1392.4 D15(3) - RCP 2A -109.1 D15(4) - RCP 2B -227.9 EPOL1 -7.0000 k (v) -
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APPLIED DP (PSID) 2 0 2-27 UP + 2-27 DN o 3-13 UP a 3-13 DN
~
STEAM GENERATOR #2 DIFFERENTIAL, PRESSURE TRANSMITTER PDT-9126-SMB CALIllRATION CIIRVES
],
FIGURE 3.2.1.19
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- -10 10 30 50 i
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APPLIED DP (PSID) > 2-27 DN o 3-13 UP A 3-13 DN I d. , STEAM GENERATOR #2 DIFFEitENTIAL PRESSURE TRANSMITTER PDT-9126-SMC CALIBRATION CllRVES h j FIGURE 3.2.1.20
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117
/T CONCLUSION:
The post-core RCS flow rate was satisfactorily determined and found to be within the acceptance criterion. The RCS flow coastdown characteristics were also satisfactorily de te rmined. Although conservatism of the 4-RCP coastdown was found to be questionable with respect to the flow coast-down assumed in the safety analysis, this possible non-conservatism was compensated for by increasing the COLSS penalty factor, EPOLl. The magnitude of this penalty was based on the time for the RCPs to reach 90% of rated speed during the coastdown. The CPC and COLSS flow constants were determined and satisfactorily input into their respective data bases to assure a conservative operation of the plant. The reference post-core AP data base was ade'quately established. 7s All test objectives were achieved and all test acceptance ( ,)_ criteria were satisfactorily met. 4
\
v
118-Oi V. 3.2.2 RCS Leak Rate Measurement (SIT-TP-506) o PURPOSE: The post-core RCS leak rate measurement was performed to demonstrate that the RCS leakage at normal operating temperature and pressure is within the limits of the Technical Specifications. The test also demonstrated that - the plant operations procedure gives acceptable results and provided independent calculations of leak rate in the event that unacceptable results were obtained from this leak rate q procedure. . Finally this test demonstrated that a known leak rate (~1 GPM) can be accurately detected using the operations procedure. i V METHOD: This test was performed _twice during hot functional testing. The first run on February 5, 1985 yielded unacceptable results and was repeated satisfactorily on February 7, 1985. With the RCS and CVCS in steady state conditions, a plant computer snapshot of plant conditions was obtained. The snapshot consisted of information on water levels, pressures i and temperatures for the following components.
- Reactor Coolant System
- Pressurizer-
- Reactor Drain Tank
- Volume Control Tank u.) .
e m y. y e+,., -> e v -P 4*" - ' ' ' - - ' ' ^' - - * ' ' -'#"* *-i ' '#
119 N )i
- Quench Tank
- Containment Sump
- Equipment Drain Tank
- Holdup Tanks
- Safety Injection Tanks Plant conditions were maintained steady for at least one hour after which time a second plant snapshot was obtained. Using predetermined constants relating change in volume to change in temperature or level of the associated components, the operations procedure calculated a leak rate using a change in s volume during the one hour time period.
%/
If these results were unacceptable, then an independent calculation using this test procedure, which performs a mass balance on the system using existing thermodynamic conditions, was to be performed. The procedure was repeated after establishing a 1 GPM sample flow rate to demonstrate that the above methods can detect this leak rate. RESULTS: Both calculations are extremely sensitive to changing plant conditions which led to the unacceptable results during the first test run. Close review of the operation procedure generated changes which corrected and simplified some of the calculations. f3 b
120 k u) The rerun of the operations procedure yielded acceptable results, as listed below in Table 3.2.2.1, so no verification using this test procedure was required. TABLE 3.2.2.1 RCS LEAK' RATE TEST RESULTS (GALLONS PER MINUTE) WITH KNOWN LEAK SOURCE BASELINE' SUPERIMPOSED
' Identified 0.05 0.05 Unidentified 0.09 1.11
. Total 0.14 1.16 CONCLUSIONS:
The Reactor Coolant System leak rate at normal operating temperature and pressure was within the limits of the Technical Specifications. The operations surveillance procedure accurately measured leakage from the reactor coolant system in the range of allowable leak-rates, and a 1 gpm leak rate was detectable. All test objectives and acceptance criteria were met.
121 i [N
\~-]
i 3.2.3 Postcore Reactor Coolant System Heat Loss (SIT-TP-508) PURPOSE: This test was performed to measure the heat loss from the entire reactor coolant system (RCS), and from only the pressurizer with spray and without spray. These measured values were then implemented into the plant monitoring computer (PMC) data base to be used in various Core Operating Limit Supervisory System (COLSS) calculations. For use in calculating the heat loss, this procedure also measured the heat input to the RCS from the reactor coolant pumps (RCPs) and the pressurizer heaters. . , z This test satisfied the commitment of FSAR section 14.2.12.3.6. METHOD: This test was performed three times during postcore hot functionals on February 6, March'3 and March 13, 1985. The measurement of heat loss was performed by means of heat balance on the RCS. Heat input to the system was from the RCPs and pressurizer heaters. RCP heat input was calculated using the measured voltage and current to each pump and an ! assumed efficiency. Pressurizer heat input was calculated using the measured voltage and current and the time each j ' heater was energized. -Heat loss from the CVCS was calculated
- ~
using charging and letdown flows and enthalpies. Heat output from the RCS was calculated by " steaming down" the generators. This was accomplished by' raising the levels es above normal then securing feedwater and blowdown. Using the (%) volume of water that " steamed down" during a one hour period, I F.
< , ,, , , , e.. . ~ . . ~ . - . - -~ . . , - . , , , , . -
r I 122 73 ( l wi the heat removed from the RCS was calculated. The RCS heat loss was simply the difference between the measured heat inputs and outputs. RESULTS: NOTE - The final test results are tabulated in Table 3.2.3.1; all other data presented in this discussion of test results is given to document the evolution of the derivation of the final test values. The first performance of this test provided the following results:
/
k,s/ s Pa rameter Results Acceptance (BTU /hr) Criteria (BTU /hr)
- Pressurizer heat loss without spray 3.56x10s- , 4.30x10s
- Pressurizer heat loss with spray 8.43x105 5 5.10x10 5
- Total RCS heat loss 2.84x107 N/A
- RCP heat input 7.79x107 N/A
- Pressurizer heater input 2.42x105 N/A The pressurizer heat loss.without spray met the acceptance criteria. The pressurizer heat loss with spray exceeded the acceptance criteria; however, based on an evaluation by Combustion Engineering of this data considering the impact of the cooler than expected spray temperature, as measured in the pressurizer spray valve and control adjustment test per SIT-TP-505 (see section 3.2.6), the measured heat loss value was found to be acceptable. .
N
-( ) .
Both the RCP heat-input and pressurizer heater input were satisfactorily determined; neither had an acceptance criterion to meet.
. . ~ .-. -
123 l (~'/ s,-
/ ,
The total RCS heat loss was almost twice that measured during pre-core hot functional testing (1.435x10 BTU 7 /hr). Based on this large difference in measurement values, this portion of the test was repeated to check for repeatability of the post-core value. The second measurement of total RCS heat loss. yielded a value of 2.149x107 BTU /hr. This value, although relatively close to the initial one, was still significantly larger than both the pre-core value, and the magnitude of the expected post. core value. A subsequent walkdown of the RCS found a large section of pressurizer surge line insulation removed. Following reinstallation of the removed insulation, the total RCS. heat loss measurement was performed for a third time, and gave acceptable results of 1.58 x 107 BTU /hr. ! '"g ('d TABLE 3.2.3.1 FINAL RCS HEAT LOSS TEST RESULTS T RESULTS ACCEPTANCE PARAMETER (BTU /hr) CRITERIA (BTU /hr) Pressurizer heat loss without spray 3.56 x 106 54.30 x los Pressurizer heat loss with spray 8.43 x 106 15.10 x 106 Total'RCS heat loss 1.518 x 107 N/A RCP heat input 7.79 x 107 N/A ' Pressurizer heater input 2.42 x 106 N/A 4 1 G 1, .
124 1 CONCLUSIONS: The RCS heat loss and heat input parameters were satisfactor-ily measured and' installed into the PMC data base foi use in the COLSS calorimetric calculations. The heat loss and heat input values tabulated in Table 3.2.3.1 were those used throughout the initial test program by other tests utilizing heat loss / input terms. All acceptance criteria of this test were sr.tisfied or the test results were determined acceptable. Y l l l l l i l-O e y -,.r.gy- n ,n- -,, , , , - rr r- ++-r,,--e<+%-er-+-ewr**----,-e-***--
-r%-- =-4,-- -*
- 125
! I xs 3.2.4 RC3 Expansion Measurements (SIT-TP-509) DATES PERFORMED: Prerequisite baseline data was taken on 1/8/85. Measurements at the 120*F plateau were made on 1/9/85. The 260 F plateau data was taken on 1/23/85, followed on 1/25/85 by the 345cf plateau measurements. Initial measurements at the 545 F plateau were performed on 2/4/85; final measurements at this plateau were taken on 2/8/85, following a required 72 hour soak. Resolution of all out-of-tolerance clearances was achieved by 3/3/85. PURPOSE: s (J 'N
\
The purpose of this test was to demonstrate the unobstructed thermal expansion of RCS components during plant heatup, and optionally during plant cooldown. Verification that shims installed during precore hot functional testing (in accordance with SIT-TP-302) were correctly sized was a second objective of this test. This test satisfied the requirements of FSAR section 14.2.12.3.17. METHOD: Baseline data was taken prior to filling and venting the RCS. Subsequently, plant conditions were stabilized at each of four RCS temperature plateaus during the heatup to hot standby conditions; the specified plateaus were at 120*F, 260*F, 345 F, and 545 F. Reactor vessel support latera1 t'~'; ( / restraint gap, reactor vessel anchor bolt grillage-to-washer v
126 7y ,. gap, and steam generator sliding base x-direction gap measure-ments were taken at each plateau to verify unobstructed thermal expansion of these components. Following a 72 hour soak at 545 F, precise measurements of the above gaps were performed to verify not only that the clearances were sufficient, but also that they were not excessive. RESULTS: Throughout the heatup, all gaps were verified to be large enough so that thermal growth of RCS components was unobstruc-ted. The specific checks performed are detailed below:
- 1. The reactor vessel support lateral restraint gaps were N verified to be greater than 0.020 in, at each measurement point.
- 2. - The minimum gap between each steam generator sliding base and its x-direction stop was verified to be greater than <
0.080 in.
- 3. Each reactor vessel support anchor bolt grillage-to-washer gap was verified to be greater than 0.005 in.
Following the 72 hour soak at 545 F, precise measurements-were taken with the following results:
- 1. The reactor vessel support lateral restraint gaps were found to be acceptable; that is, the minimum total clear-1 ance between the support block and the shim pack mounted on the lateral restraint was measured to be greater thaa 0.040 in. and less -than 0.100 in, for each reactor vessel-
[\ ( ,/ ' cold leg support. The actual test data is presented in Figure 3.2.4.1. , p 5
- - - - r q -- - - , a,, , e,--.., ,r ,-s, -ry + , , , ~ , . , , , , , ~ < , - - , - , -..-.p-g p ww ggw, -v e- , , ,-g,-n ,,
127
,-m.
l ) v
- 2. The steam generator sliding base x-direction gaps were found to be unacceptable on both generators; that is, the measured gaps fell outside the acceptable range of 0.183 in, to 0.215 in. The shim packs were replaced to bring the gaps into tolerance. Figure 3.2.4.2 provides the measured clearances, both prior to and following rework of the shim packs.
- 3. The reactor vessel support anchor bolt grillage-to-washer gaps were also found to Le unacceptable at 545*F; that is, the measured grillage-to-washer gaps fell outside the acceptable range of 0.005 in, to 0.015 in. Shims were fabricated and installed to bring these gaps into toler-ance. The measured grillage-to-washer gaps, both prior to and following installation of the shims, are shown on e ~x Figure 3.2.4.3.
( ) G! A steam generator #1 anchor bolt nut installed during precore hot functional testing was checked during this test to ensure that the nut-to-washer gap had been properly set. The measured gap at 545 *F exceeded the 0.010 in. to 0.020 in. specification. All anchor bolts on both steam generators were then checked and several were found to be out of tolerance. Those found out of tolerance were subsequently adjusted to within specifications. As-left gap data is shown in Figure 3.2.4.4. CCNCLUSION: The RCS components were determined to be free to expand thermally during plant heatup to the normal operating
,,s temperature of 545*F. -
i v
)
- -r
.g 128 4
O - (u, The shim packs installed during precore hot functional testing on the reactor vessel support lateral restraints were ) found to be sized properly.
- The clearance measured at the steam generator- sliding base x-direction stops were found to be inadequate to accommodate i
, the hot leg thermal growth anticipated from zero power to 100%
- power. The shim packs were removed and reworked to provide l the requisite clearances.
l The reactor vessel support-anchor bolt grillage-to-washer gaps were found to be too large to adequately restrain ' . vertical. motion of the vessel during design seismic events. i Installation of shims between the grillage and the washers i l was accomplished to bring the gaps, with one. exception, to i f-r within the allowable tolerances. The out-of-tolerance gap l k (.016" at the A2 gap on loop IB) was deemed acceptable based
; .upon a review performed by Combustion Engineering.
d
, Steam generator-anchor-bolt nut-to-washer gaps were measured for all eight bolts on eachesteam generator. Those nuts i ; whose gaps were found out-of-tolerance were tightened to I
achieve the- requisite clearance, f All test objectives and acceptance criteria were satisfactorily met. 4 t' s f 4
-+ ,,- , - > %+ --c -,, w r , y,-- *, -e- g--e=-,- g m p- w- 'y-f,-- , ew----ee---- g -w m-y e
, i- ,-- c--
3 * * - - +t--.ev----vv--
129
.( )., FIGURE 12Al REACTOR VESSEL SUPPORT LATERAL RESTRAINT CAPS lateral r straint g y i i 33 I
81 82 ACCEPTANCE CRITERION: (FOR EACH LOOP) 0.040"4hin(A1.A2,A3)+ min (Bl.32,B3)G 0.100 " where min (Al, A2, A3) represents the minimum value RV ' among Al, A2 and A3 for the loop, and
/N min (B1,B2,B3) represents the minimum value
- k. among 31, B2 and B3 for the Icop.
reactor vessel support Al A2 A3 1 1 1 i T i lateral restraint LOOP Al 6AP A3 GAP A3' G A P 81 GAP B3 GA PlB3' GA P lA .022 " .037 " .043 " .041 " .055 " .055 " 1B .028 " .043 " .040 " .042 " .043 " .044 " 2A .029 " .042 " .057 " .029 " .042 " .057 " 20 .021 " .043 " .048 " .044 " .041 " .046 " (A3' and B3' Gaps are measured directly below A3 and B3 Gaps.) V l i 1
l 130 k l FIGURF 32A2 STEAM CENERAT_0R SLIDING BASE X-DIRECTION GAPS cz4 - SG SLID lNG BASE]CI-.+ 4.-
> Rv Q
r's AS-FOUND AS-LEFI (545*) (545') C1 C2 C1 C2 SGI .155 " .15 8 " SGI .195 " .192 " SG2 .125 " .125 " SG2 .209 " .190 " 9 The acceptance criteria for 'the steam generator sliding base x-direction gaps are: 0.183 " 6 C1 1 0.215 " 0.183 " i C2 6 0.215 " O
l i FIGURE 3,2A3__ 131 (', , REACTOR VESSEL SUPPORT ANCHOR BOLT GRILLAGE-TO-WASHER CAPS
/
O 9 2A 0
/0 2B
#De o O
+p REACTOR VESSEL g Acceptance Criterion:
@ O 0.005 " s GAP $ 0.015 " G O 8/
O lB o IA p ,8* 9O o AS-FOUND LOOP l Al Gap A2 Gap B1 Gap B2 Gap I 1A .030 " .039 " .040 " .075 " 1B .040 " .033 " .034 " .033 " 2A .016 " .016 " .025 " .034 " y7 2B l .030 " .030 " .021 " .026 " AS-LEET 1hWYfire&'$ LOOP Al Cap A2 Gap B1 Gap B2 Gap 1A .006 " .015 " .013 " .013 " 1B .013 " .016 " .007 " .006 "
.015 i' 2A .007 " .006 " .013 "
\p > U 2B .005 " .010 " .012 " .013 "
l 132 o U FIGURE 3.2A4 STEAM CENERATOR ANCHOR BOLT NUT-TO-WASHER CAPS SG SLIDING BASE > av O
@ @ se ANCHOR BOLT NUT
(
%./
AS-LEFT NUT-TO-WASHER CAPS bolt 1 bolt 2 bolt 3 bolt 4 bolt 5 bolt 6 bolt 7 bolt 8
.015 "
SG1 .020 " .020 " .020 " .020 " .012 " .014 " .012 " . 017 " l SG2 .020 " .016 " .013 " .014 " .015 " .019 " .020 " . 016 " : i 4 The acceptance criterion for each anchor bolt's nut-to-washer gap is: [ l 0.010 " n gap n .020" l I i
+y ,a ~m , , , ,y , , - ,- , - , , , . s-----,---y-
133
)
(w) 3.2.5 Control Element Drive !!echanism (CEDM) and Control Element Assembly (CEA) Tests (CEDM Performance; SIT-TP-503) PURPOSE: The objectives of this test were to verify proper operation of the control element assemblies (CEAs), their respective control element drive mechanisms (CEDMs) and associated indications and alarms under hot shutdown and hot standby conditions. The test consisted of the following: A demonstration of the proper operation of the control element drive mechanisms (CEDMs) and control element assemblies (CEAs) under hot shutdown and hot standby conditions.
/N i i
\s / - A check of the CEA position indication systems and a verification that the indications by core protection calculators (CPCs).and CRT are within 3 inches. A verification of the proper functioning of the CEDM upper and lower electrical limits. A measurement of CEA withdrawal and insertion rates. , A verification that each of the 91 individual CEAs has the proper drop time from a fully withdrawn position to its 90% insertion position at hot shutdown and hot standby conditions. A verification by inspection of CEA position versus time recorder trace, that the dropped CEA decelerates as it
,, ' approaches the fully inserted position.
Is._ /l ,
134 ,e A demonstration of the proper operation of the CEDM holding bus. A verification of proper operation of the CEDM Control System (CEDMCS) and its associated computer alarms and limits under hot standby conditions. METHOD: At hot shutdown the CEAs were withdrawn and inserted in manual individual. Careful checks were made of the CEDM position indicating systems as well as verifying proper CEDM operation by analyzing CEDM coil traces. The CEAs were again withdrawn and dropped to measure 90% and 100% insertion times. Those 7-~s CEAs outside two standard deviations were drooned three more k,) times. While at hot shutdown, the CEDM holding busses were tested by placing each subgroup on the bus and verifying it would not drop when its subgroup breaker was opened. At hot standby, each CEA was again tested in manual individual, and 90% and 100% insertion times were recorded. Those CEAs outside two sigma were dropped three more times. At hot standby the CEDMs were tested in manual group and manual sequential to verify functions such as Upper and Lower Group Stop, Upper and Lower Sequential Permissives, Exercise Limits, Power Dependent Insertion Limits, Minor and Major Deviation Alarms and Out of Sequence alarms by moving groups of CEAs to the proper location and verifying proper control or alarm function. 4
135
. ,-m RESULTS:
CEA slipping and sticking was experienced on a few CEAs during the test. These problems were attributed to sluggish gripper action and misalignment of the CEA extension shaf t and the upper gripper. These problems were accommodated with modified CEA timing and voltage adjustments. Minor problems were experienced with the CEA processing software in the plant monitoring computer (PMC). Also, the Out of Sequence and Power Dependent Insertion Limit (PDIL) alarms were not satisfactorily verified during this test. They and all other test deficiencies.were, however, reverified during subsequent CEA movements (e.g. during initial criti-cality, low power physics testing, etc.). CEA drop times (to 90% inserted) and reed switch functional testing were both satisfactory for satisfying Station Technical Specifications 3/4.1.3.3 and 3/4.1.3.4' for entering mode 2. All 90% insertion times were less than 3.0 seconds-(2.74 seconds maximum) at hot standby, as shown in Tables 3.2.5.1 and 3.2.5.2, and Figure 3.2.5.1. Reed switch position transmitters were always within 4.5 inches of each other (no deviations greater than 2.0 inches); also the reed switch position transmitters were within 3.0 inches of the CEA pulse counting system. CONCLUSION: With the satisfactory retest- of all deficient test items, all acceptance criteria of this test were met, and the CEAs and CEDMs were shown to work as expected. ( h I i t
136 i l TABLE 3.2.5.1 Part 1 of 3 , CEA DROP TIMES TO 90% INS'ERTED CEDM 90% Insertion Times (seconds) 100% Insertion Times (seconds)
# 320*F 545 F 320 F 545*F 1 2.42 2.63 2.73 2.93 '
2 2.56 2.62 2.89 2.91 ! 3 2.52 2.51 2.88 2.80 4 2.53 2.64 2.87 2.96 a 5 2.21 2.59 2.51 2.88 6 2.39 2.61 2.74 2.91
,. 7 2.12 2.55 2.42 2.84 i 8 2.36 2.57 2.70 2.88
- 9 2.20 2.63 2.51 2.93 1 10- 2.00 2.57 2.28 2.85 l 11 1.73 2.63 2.'1 2.93 1 12 -2.16 2.69 2.49 3.01 1'
13 2.65 2.63 3.05 2.94
. 14 2.58 2.59 2.86- 2.84
- j. 15 2.59 2.61 2.91 2.91 l 16 2.52 2.65 2.80 2.93-17 2.42 2.58 2.71 2.88 18 2.55 2.56 2.87 2.85 19 2.45 2.66 2.84 .2.97 20 2.55 2.61 2.87 2.95 PI 21- 2.59 2.67 2.95 3.01 i 22 2.43 2.69 2.79 3.01 L 23 2.68 2.64 3.04 2.90
$ 24 2.39 2.64 -2.72 2.94
- 25 2.30 2.67 2.60 2.95
{ 26 2.20 2.64 2.48 2.90 j ~27. 2.40 2.59 2.74 2.92 28 2.25- 2.40 2.54 2.70 i 29 2.37- 2.40 2.65 2.68 30 2.00 2.42 2.28 2.71 31 2.35 2.38 2 65
. 2.67 ,
32 2.30 2.45 2.57 2.74_ t
- - 33 2.32 2.38 L2.61 2.67
, 34 2.32 2.39 2.58 2.68 , 35 2.40 2.42 2.69 2.71 36 2.49 2.54 2.77 2.80 37 2.65 2.67 2.99 2.97 38 2.46 2.66 2.76 '2.92 39 2.40 2.60 2.71 2.91 ' '
- . .40 2.67 2.65 2.99 2.93 41 .2.69 2.65 3.03 2'.93
~
,D
[ t i
. - ~ _ _ , _ _ . - . _ _ . . _ _ _ _ . . . _ , . - . _ _ . . - _ . . _ . . _ _ . _ . . _ _ _ . _ . . . . _ . _ . . _
137 TABLE 3.2.5.1 (continued) Part 2 of 3 CEA DROP TIMES TO 90% INSERTED . CEDM 90% Insertion Times (seconds) 100% Insertion Times (seconds)
# 320*F 545*F 320 F 545*F 42 2.58 2.62 2.94 2.91 43 2.63 2.61 3.00 2.89 44 2.63- 2.67 2.95 2.95 45 2.70 2.72 3.04 3.01 46 2.65 2.66 3.00 2.95 47 2.04- 2.63 2.35 2.94 48 2.55 2.66 2.89 2.95 49 2.47 2.72 2.78 2.96 i 50 2.52 2.62 2.86 2.93 1 51 2.60 2.66 2.91. 2.94 52 2.60 2.68 2.92 2.94 53 2.61 2.70 2.91 2.99 54 2.45 2.60 2.77 2.90
] 55 2.52 2.67 2.85 2.97 1 56 2.60 2.67 2.91 2.96
.,, 57 2.70 2.68 3.04 2.97 4
58 2.58 2.56 2.90 2.86 1 59 2.54 2.62 2.84 2.92 j 60 2.56 2.66
' 2.89 2.98 61 12.61 2.74 2.89 3.02 i 62 2.62 2.60 2.94 2.90 4
63 2.54 2.65 2.82 2.92 64 2.53 2.64 2.83 2.91 4 65 2.38 2.70 2.69 2.99 j 66 2.32 2.62 2.63 .2.90 67 2.10 2.60 2.40 2.88 68 2.28 2.67 2.58 2.96 69 2.29 2.65 2.58 2.94 70 2.54 2.64 ' 2.85 2.92 71 2.58 2.58 2.86 2.85 72 2.56 2.58 2.86 2.86
~73 2.56 2.62 2.87 2.90 74 2.58 2.64 2.89 2.92 75 2.61 2.66 2.93 2.97 76 2.60 2.62 2.85 2.89 77 2.56 2.63 2.87 2.91-78 2.62 2.64 2.93 2.93 79 2.55 2.58 2.86 2.90 80- 2.54 2.59- 2.85 2.89 81 2.58 2.62 2.90 2.94 4,,
82 2.57 2.55 2.88 2.87 q.
- (_ / -
a t
138 y _(J TABLE 3.2.5.1 (continued) Part 3 of 3 CEA DROP TIMES TO 90% INSERTED CEDM 90% Insertion Times (seconds) 100% Insertion Times (seconds)
# 320 F 545 F 320*F 545'F 83 2.55 2.61 2.85 2.89 1
84 2.55 2.64 2.87 2.96 85 2.60 2.67 2.92 2.98 86 2.59 2.60 2.88 2.87 87 2.55 2.59 2.87 2.87 88 2.38 2.42 2.68 2.69 89 2.38 2.46 2.67 2.73 90 2.44 2.43 2.74 2.69 91 2.44 2.50 2.73 2.74 i TABLE 3.2.5.2 AVERAGE DROP TIMES TO 90% INSERTED OF THREE DROPS OF CEAs OUTSIDE 20 320'F 545 F RETEST # CEA # TIME CEA # TIME 1 7 2.50 -3 2.52 2 7 2.52 3 2.52 3 7 2.52 3 2.51 1 10- 2.55 36- 2.53
- 2 10
- 2.56 36 2.54 1 3 10 2.56 36 2.53 1 11 2.58 45 2.71 i 2 11 2.57 45 . 2.66 3 11 2.57 - 45 2.71-1 30 2.38 49 2.68 2 30 2.41 49 2.66 3 30 2.37 49 2.65 1 47 2.53 61 2.69 2- 47 2.57 61- 2.72 3 47 2.56 61 2.71 1 67 2.56 -- --
l 2 67 2.60- -- -- 3 67 2.57 -- -- 1 ! \_/
, . . . . . - . - . - . . , . ~ . . . . _ _ _ , _ , . . _ . , , . . _ _ , _ , - . .
139 (- V FIGURE 3.2.5.1 HISTOGRAM OF CEA DROP TIMES TO 90% INSERTED AT 545 F AND 2250 PSIA p . _ _4_ . _ .._ _ . _ __..._p_ .- . - - ..
,g_ ,
t-- -
-r
~
30 A 5 m e ,4 . h 20 -
-/
y . z , i 10
, , . r ,- ,
3 4
,%.h-- . . , , , , , "'*
c ' 0 2.3 2.4 2.5 2.6 2!7 -1 2.8 TIZE FOR 90% INSERTION (SECONDS 1 4-FING = 4 Fingered CEA , PL = Part-Length CEA CEA TYPE- MEAN(X) 2X STD. DEV. (20) PART LENGTH (8) 2.40 0.04 FOUR. FINGER (4) 2.45 0.06 FULL LENGTH (79) 2.63 . .-
.--, 0.08 \~J .
140-A l i
'V 3.2.6 - Pressurizer Spray Valve and Control Adjustment (SIT-TP-505)
PURPOSE: 1 The objectives of this test were two-fold:
- 1. To establish the proper flow settings for the pressurizer
= continuous spray valves (RC-302A and RC-302B) at steady-state conditions, so as to minimize the temperature differential between the RCS cold legs and the pressurizer spray nozzle.
- 2. To measure tne rate at.which the pressurizer / reactor coolant system pressure could be reduced, utilizing pressurizer spray flowing through the pressurizer main
%) spray valves (RC-301A and RC-301B) in parallel.(required) s- / and individually (for information only).
This test satisfied the requirements of FSAR Section , 14.2.12.3.5. l METHOD: i Seven temporary thermocouples were mounted on the spray piping at various locations upstream of the pressurizer main
' spray valves as shown in Figure 3.2.6.1, to provide spray line temperature data. The reactor coolant system was stabilized- '
at approximately 550 F and 2250 psia. The temperature reading of thermocouple #7 was compared to the' averaged RCS cold leg temperatures. Both continuous spray valves (RC-302A and RC-302B) 'were adjusted fully open, to minimize the temperature differential.between the pressurizer spray line (__- and the average temperature of the'RCS cold legs. l I l-4 - .. _ , . .-. . - __,_ , . ~ - . _ . . . . _ _ _ _ . . . . . . , - . _ _ _ . _ . _
s 141 S 3 I ) x,/ The pressurizer spray effectiveness was demonstrated by manually controlling the pressurizer main spray valves to the full open position, with all heaters off, while measuring the time required to reduce pressurizer pressure f rom 2250 to 2100 t psia, with the RCS at hot, zero power conditions. The accep-tance criteria were based on both valves operating in parallel. However, depressurization times for~ each valve separately were -also to be determined for information purposes. To assure that the spray valves were fully open while pressure was decreased through the test range, the initial pressure was raised to 2330 t 10 psia. The valves were then actuated collectively and singly. The pressure / time data was recorded via computer printout. RESULTS: (]
\ ; ' _,/
The acceptance criterion of maintaining the pressurizer spray line temperature at no more than 25 to 30 colder than the average RCS cold leg temperature, at steady-state conditions, could not be met. With the continuous spray calves fully open, the actual temperature difference was approximately 52 F. A reevaluation of the acceptance criterion by~ Combustion Engi-neering indicated that the spray nozzle portion of the system would not be adversely affected if the temperature dif ference
., r was maintained at 85 F, or less. A reevaluation of the spray piping by Ebasco, utilizing the revised AT value of 85 F, indicated that the system would also not be adversely af fected.
The pressurizer spray effectiveness was' demonstrated to be well within the 93 second time limit for both valves. The r actual depressurization time was 79 seconds. The individalu
,. depressurization- times for valves RC-301A and RC-301P were i
s
,) 107 and 108.6 seconds, respectively. The pressure vs. time
', curves are shown on Figure 3.2.6.2.
t 1
& ~
__.-..- - _.- - - - _ - _ _ - - ~ . . . . ~ --
's p -
(;; 142 J CONCLUSION: 4 i 1 Following a reevaluation of the acceptance criterion for the i maximum pressurizer spray line temperature differential, all objectives of this test were met and all acceptance criteria " satisfied. t 'h , I i t f ) i- ) i s .i l i l 9 e l
~ ~ . - - - , , . , - , - ~ . ~ . . _
.~ . . . . . - - . . . . .. . . _ _ - . . . _ _ - - .
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=
7 i 143 FIGURE 3.2.6.1 j LOCATION OF TEMPORARY THERMOCOUPLES
*7 f
I' 8'-8" 7'-9" 4 n *2 l'-0" 7'-0" ( N - 1O - 1
- 3 i
j PRESSURIZER j- Auxiliary Spray 3'-3"
/
; o
/ >
i 2'-0"
, "/
/
4 i *6 } 8 ' r i ! /\ 4 a Main ' l SP ray
* = Approximate Thermocouple Locations
. O. *
/\-
Main Spr y
- O 4
- , - , . . . - -._,.-,--,-ry,y .rr - - c-, , , , , . . ,.,,,,.c.,y p..,, . , , .,-%,,,.w_,.,y,.w,, ,,,....,ww%,.m,..,,we...mp,p.,7,, ...m.,wy.,,--.m, ,...,..,,w,,.w,m,,,y,
144 (^'s/ FIGURE 3.2.6.2
*/
PRESSURIZER / REACTOR COOLANT SYSTEM DEPRESSURIZATION VERSUS TIME CURVE Case Valve (s) Time (sec) 1 RC-301A luy 2 RC-301B 108.6 3 RC-301A&B 79 00:0 ' i l i L 1 w n
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(_..__Z:Z _L. _ ~. _~q y' i . . s - x - , h 9 20 40-~ 7_~_J^60 3 0 ~_Z.~_-~_10 0 r ~ _ _1; t o p-i g p_ _3g,,g _ a l g/ - - - ~ ~ -_ .
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- 145 O
3.3 OTHER TESTING 3.3.1 Post-Core Test Data Record (SIT-TP-511) PURPOSE: The purpose of this test was to provide a permanent baseline data record of plant parameter indications during the post-core hot functional: test program. METHOD: After completion of the RCS fill and vent following the
. initial. fuel loading, and prior to commencing system heatup
\ and pressurization, data collection consisting of the following plant systems parameters was initiated:
- RCS temperatures and pressures
- Charging and letdown
- Pressurizer
- Steam generator
- Secondary
- Safety injection
- PPS-
- RCPs and RCP motors Containment atmosphere
- Leak detection Data was collected in the form of computer snapshots using the plant monitoringLeomputer (PMC), once per hour or once <
per shift, depending on the system. Parameters not expected sto change much over the course of a shift (i.e., containment
- atmosphere;' leak detection; PPS; safetyjinjection) were
= l
t; a t- 146 i-t .A i
- - recorded at the lower ' frequency, all others at the higher
[- frequency. Data collection continued throughout the post-core t
- j. hot functional test program (see also section 6.7.1). ,j
- i
- j. .
1 j RESULTS: '
- t. :i The required. data was gathered at the specified intervals. !
i i CONCLUSION: y A substantial data base of significant. plant parameters was 1 established for plant conditions corresponding to RCS i { ( conditions' ranging from about 120'F/350 psia.(Mode 5) to g 545*F/2250 psia'(Mode 3). This data was placed in the plant historical file for future reference. All test objectives and acceptance criteria were. satisfactorily met. i . t i I ) 1-4 i_ 9 l -i
- - l i
i f. e i i i i; I' i' I.
..=-...,-,-e-..-.~.....-. ..,,,---.,m*+wb - .._,_.,+.,,-e.w,,. _,,.-w,-w+,*,r,w,,,ew
147 [~T Q' ^ i f
; 3.3.2 Heated Junction Thermocouple Operation Verification
- (SIT-TP-500, Attachment 8.2.6)
PURPOSE:
- The purpose of this test was to demonstrate the sequential '
change of state from uncovered to covered of the heated junction thermocouples (HJTCs) during the reactor coolant system (RCS) fill and vent. METHOD: Reactor vessel water level was established at approximately ,
- 14 ft. (i.e., between the hot leg center-line and the top of f-- the hot leg nozzle), as shown in Figure 3.3.2.1. During the
\s / RCS fill and vent, the reactor vessel water level- indications on the qualified safety parameters display system (QSPDS) channel 1 and 2 were compared to each other and to actual levels as indicated on a temporarily installed tygon tubing level gage.
1
- RESULTS
At the start of the test, sensors 8, 7, and 6 were already covered due to their depth within the reactor vessel and the requirement to have the RCS water level to at least the center-line of the hot leg to permit coolant circulation by the shutdown cooling system. The remaining sensors, 5 through' 1, changed state sequentially starting with-sensor #5 and ending with sensor #1 as the water level within the reactor- {. i
- s e
d O
1 I 148 (
. IN-CORE INSTRUMENTATION SUPPORT PLATE
======== -CEOM N0ZZLI 9 , =M =# '
1 , bl NSTR UA'.ENTATION IN-CORE INSTRUMENT LEVEL 1 (23.2*) GUIDE TUBEx Q} ,,, ., ;h N0ZZLE LEVEL 2 (20. 8' ) j AllGNMENT CONTROL g_I l PIN LEVEL 3 (18.3') LEVEL 4 (17.3'} FlFMENT QS BLY Q N jtIT Z{ 3={ -
/ UPPER
,,,, GUl0E WITHDRAWN l i J -
LEVEL 5 (15.3') z_ s u . 1 J [ , j STRUCTURE LEVEL 6 (13.5') p.l J
) [
i I
)
j i l 30" I D LEVEL 7 (11.8') C" 3 0 T7
/
.' I \INLIT (10.1')
LEVEL 8 OUTLET q l i N0ZZLE N0ZZLE "l l l Tl .t l h .
!f Ii E !II -
CORE SURVEILIANCE HOLDER II '
' E li h(({RT 150" t gj f -
CORE ACTIVE - CORE II l Bi l J~' N SHROUD LENGTH !i i T"-7% Ii E i i FUEL g { ASSEMBLY
'ii i a R 7-g"p."
~ -
SNUBBER fN ER
\ STRUhT RE
..i!!!!!!i:!"
CORE STOP L l l l l APPROXIMATE LOCATION OF HEATED JUNCTION THER>f0 COUPLE LEVELS FIGURE 3.3.2.1 f
\
1 149 i O was raised. Both QSPDS channels indicated the respective
. sensors' change of state from uncovered to covered almost simultaneously, and the time of the sensors' change corresponded well with the level indications shown by the tygon tubing level gage.
CONCLUSION: The heated junction thermocouples satisfactorily changed state during the RCS fill and vent, to provide an accurate indication of the coolant level within the RCS. i ( l [ . l [ \ . s t
, - , , , - - - - ,,w- - , , , , , , - - , - e--,,v,,n . ,m->,-~w ~-m, ,-e
, ,. , , . . , . , - ,,,.-e- ,, ,..--e,
150 i{ ) 3.1.3 RCS and Steam Generator Parameters (SIT-TP-500, Attachment 8.2.2 - 8.2.4)) PURPOSE: The purpose of this measurement was to provide baseline data correlating RCS temperature and pressure data with steam generator pressures during RCS heatup, METHOD: During the RCS heatup data was recorded off the plant monitoring computer (PMC), or control panel indications if the PMC was not operational, for the following parameters: O
- Reactor coolant loop 1B cold leg temperature
- Reactor coolant loop 2A cold leg temperature
- Pressurizer temperature
- Pressurizer pressure
- Steam generator #1 pressure -
- Steam generator #2 pressure Data was recorde,d at the 260*F/350 psia, the 345*F/<392_ psia, and the 545'F/2250 psia plateau, s
151 s iq,'t RESULTS: The required data was satisfactorily collected and is shown in Table 3.3.3.1. TABLE 3.3.3.1 RCS AND STEAM gel.TRATOR PARAMETERS TEST PLATEAU PARAMETER 260*F/ 345 F/ 545 F 350 psia <392 psia 2250 psia RC Loop 1B Cold Leg Temperature, F 261.4 341.6 543.6 RC Loop 2A Cold Leg Temperature, *F 257.7 340.9 544.4 ('~') Pressurizer Temperature, *F 430.3 428.0 652.6 Pressurizer Pressure, psia 354.0 344.4 2246.2 Steam Generator #1, Pressure, psia 37.3 120.3 978.9 Steam Generator #2, Pressure, psia 35.0 118.0 9
77.4 CONCLUSION
All data required to satisfactorily establish a data base correlating RCS temperature and pressure data with steam generator pressures during RCS heatup was obtained. The measurement objectives were met. 6 t 1
\_. /
l l r , , - - ---- -- - - -- .- , - - - , . , . . _ , -
152 i A. 3.3.4 Determination of Auxiliary Spray Flow Split (SIT-TP-500, Attachment 8.4.4) PURPOSE:
.The purpose of this test waa to collect data for a response to an NRC question concerning the RCS depressurization capability using auxiliary spray provided by the charging pumps with a failed open loop charging valve.
l METIIOD: With the RCS stable at approximately 545*F and 2250 psia,
-g both auxiliary spray isolation valves were verified closed, \ ,/ while makeup was supplied to the RCS through at least one charging loop isolation valve. Care was taken to minimize heat removal from the RCS by securing blowdown and minimizing steam demand. A second charging pump was started, loop 2 charging isolation valve was verified open and loop I charging isolation valve was verified closed before commencing data collection.
Before initiating the transient the spray valve controller was placed in manual with 0% output and all pressurizer heaters were secured. All four reactor coolant pumps (RCPs) were then stopped, and both auxiliary spray valves were opened while the open charging isolation valve was closed. When pressurizer prest.ure had decreased to 2150 psia the loop 2 charging isolation valve was reopened and pressure decreased further to 2000 psia. Both auxiliary spray valves were then closed and the RCS returned to normal hot. standby ( configuration at app::oximately 545*F and 2250 psia. e
A 153 . RESULTS: The failure during 'the test of a clamp-on sonic flow meter I attached to the auxiliary spray line to measure its flow rate , caused this data to have been lost. Additionally, evaluation of charging flow data collected per the plant monitoring # computer (PMC) showed that the. output of the two charging pumps.run during the test dropped from approximately 88 gpm " just prior to the test to approximately 44 gpm for most of the test only to return to about 88~gpm. With the above exceptions the test progressed smoothly and - sufficient data was collected to allow an evaluation of the depressurization capability using auxiliary spray flow provided by the two charging pumps with a failed open loop ' charging valve to be made. sm i CONCLUSION: e i The data collection sufficed to permit a response to the j subject NRC question-to be made. This response,was 1 transmitted to the NRC via LP&L let'ter W3P85-2115, dated June 13, 1985, from K. Cook to G. Knighton. 4 o l i 1 I i ( . l
154 ,a t 1 xv 3.3.5 Post Core Thermal Expansion Testing (SPO-99P-003) PURPOSE: The purpose of this test was to verify that piping anel component expansions are free, unrestrained and within tolerance (during plant heat-up and normal operation) as predicted by analysis. This test satisfied the commitments of FSAR Chapter 3, Section 3.9.2.1, Preoperational Vibration, Thermal Expansion and Dynamic Testing on Piping, and FSAR Chapter 14, Sect. 14.2.12.3.17, Piping Thermal Growth, Vibration and Shock (see also section 3.2.4). METHOD: \_ / The systems selected for testing were identified by engineering as a result of re-analysis, and to clear test deficiencies from testing performed during pre-core hot functional testing. Rigid restraints or building steel were utilized as reference points to measure movements at various locations by establish-ing a bench mark on the restraint and/or pipe. Spring hangers and snubbers were also used to measure pipe movements independently of the built-in scale provided. Ca rpente r squares, plumb bobs and steel rulers were used for measure-ment. Piping temperatures were measured using hand held digital pyrometers and thermocouples probes. Prior to the beginning of hot functional testing, systems involved in thermal expansion monitoring were walked down to /
,_s
\
ensure that piping and piping components were free to expand (x-j) in an unrestrained direction. <
155
,\
'm Insulation was temporarily removed at measuring points to allow clearance observation / measurement, and pipe temperature measurement.
Problems encountered during testing, which caused inappropriate pipe movements, were resolved by engineering and corrective action was implemented before proceeding. RESULTS: All thermal movements measured during testing were acceptable based on the following criteria:
- 1) Thermal movements were within the acceptable range of 20%
7-~g- or k" (whichever was greater) of the calculated movement.
'd "
- 2) Because of heat losses, actual piping temperatures were slightly less than maximum operating temperatures.
Measured movements were declared acceptable based on
, interpolation between actual and maximum operating 2
temperature values. All piping systems monitored for thermal movements fell within acceptable limits. The following significant events occurred during testing, and corrections were made as necessary:
- 1) The main steam line to the emergency feedwater pump turbine and the blowdown from steam generator to blowdown tank experienced fluid transients during testing.
Damaged restraints and piping were replaced per design. (NI
\.
v r , 1
, , =-- , , . , , .
156
- 2) Main steam line (SMS 40-15) restraints were modified per SMP-306 to bring thermal movements within acceptable range.
- 3) Sealant material in some of the sleeves were replaccd or removed to allow pipe movements as required.
- 4) Clearances provided between piping and restraints at several locations were inadequate to allow for thermal growth. These deficiencies were corrected prior to further heat-up.
- 5) Steam generator space sampling system thermal movements were out of tolerance due to the excessive weight of pipe support components. This deficiency was cleared by
,r y redesigning pipe supports. \w. - )
CONCLUSION: This thermal expansion testing of piping systems performed in accordance with procedure SP0-99P-003, satisfactorily demonstrated that piping and component expansions were free, unrestrained and acceptable during plant heat-up and normal operation. a r ) u-
1- + 157 SECTION 4.0 1 t i
- INITIAL CRITICALITY I
t i l d i I I i. 1 i i l l I i I e
-w.n-,., .v,
)
1 1 158 ) I fO: The initial criticality of the WSES-3 reactor occurred on March 4,1985 at 2148, CST. 'The criticality was controlled by procedure SIT-TP-600, In#.tial Criticality, and was attained in a safe, orderly manner by first 4 gulling the CEAs to a predetermined configuration (in a predetermined sequence) and then' diluting-the RCS-Boron concentration until criticality was achieved. - Concurrent with the CEA withdrawal and boron dilution, inverse countrate ratios (also . called inverse multiplication ratios, or 1/M's) were calculated and used to estimate when criticality would occur. i
-The basis for using 1/M's was that countrates increase to large values as criticality is approached. If a base countrate taken when the reactor
-is subcritical is divided by the countrate measured as the reactor approaches criticality, the ratio will approach zero. Extrapolation to
~
zero. provides an estimate of criticality. 4.1 CEA Withdrawal The CEA withdrawal ~ portion of the approach to criticality commenced on March 4, 1985 at 0328. The reactor was in mode 3 with the RCS temperature and pressure at about 545*F and 2250 psia. All CEAs were fully inserted and the RCS boron concentration was approximately 1780 ppm. -VCT, pressurizer and letdown line boron samples were within 10 ppm of the RCS sample value. 4 Hourly boron sampling was initiated and a plant monitoring computer 1 i (PMC) collect log was started on a five-minute trend. The CEAs were pulled in a controlled sequence that was chosen such that each step j in the sequence resulted in about a 0.5-1,0% 6p reactivity addition. }- The shutdown and part-length CEA groups were withdrawn in the Manual Group (MG) mode while regulating groups 1-6 were withdrawn in the Manual Sequential (MS) mode. The CEA withdrawal sequence took thirteen steps and resulted in all CEAs being fully withdrawn,
- except CEA group 6, which was left at 75 inches withdrawn to provide j
reactivity control when criticality was achieved. At the completion I l of each step of the sequence a CEAC snapshot was taken, which was f: used as a CEA position record. i a 1
159 r\ b Prior to the initiation of the CEA withdrawal and following each of a the steps in the withdrawal sequence, 1/M's were calculated and used to estimate criticality. Before the first CEAs were withdrawn the base countrate consisting of the average of five one-hundred second coun' was determined for the two startup neutron detectors. Subsequent countrates used for the 1/M's were determined from the average of three one-hundred second counts. The 1/M's for the CEA withdrawal portion are summarized in Table 4.1 and presented in Figure 4.1. Using a figure similar to Figure 4.1, estimates of criticality were made. In no case was criticality estimated to occur for the next immediate CEA pull, which was as expected. Problems encountered during this phase of the approach were minimal and easily fixed. CEAs 49, 37, 81 and 46 slipped during CEA pulls, and CEAs 34, 72 and 23 did not initially move when required. i A 4 (_,/ During the CEA withdrawal, a deficiency from post-core hot functional test procedure SIT-TP-503, CEDM Performance (see section 3.2.5) concerning the Out-of-Sequence (00S) alarm and interlock was successfully cleared. The last CEA withdrawal step was completed at 0832. The 00S testing was completed at about 1114. 4.2 RCS Dilution The RCS dilution portion of the approach to criticality began on March 4, 1985 at 1124. The RCS was at 545*F, 2250 psia and 1780 ppm boron. Three charging pumps were running and dilution was via the VCT. The PMU charging rate to the VCT was 130 gpm. The RCS boron sampling frequency was changed to once every 30 minutes. At 1136 charging pump A/B was secured due to back pressure regulator valve oscillations. At 1440, the charging pump was returned to service. At 1628, the dilution was halted to allow boron mixing in the RCS (boron samples were as follows: RCS-1090 ppm, Pzr-1255 ppm, and *
- (,_) VCT-55 ppm). At 1730 countrates greater than cps above background
r 1 160 (d
\
t were verified for both startup channels (39.30 and 38.35 cps, respec-tively, for startup 1 and startup 2). At 1900, the RCS boron was sufficiently mixed (boron samples were as follows: RCS-990 ppm, Pzr-1010 ppm, and VCT-990 ppm) to allow for the dilution to resume. Direct dilution was then employed by manually manipulating PMU-140. At this time the boron sampling frequency was increased to every 15 minutes. Criticality was declared at 2148. The critical configura-tion was as follows: 545*F, 2260 psia, 820 ppm boron and CEA group 6 at 75 inches withdrawn. In diluting from 1780 to 990 ppm boron 37,590 gallons of PMU were required. In diluting from 990 to 820 ppm boron 13,550 gallons of PMU were required. At 2220, CEA group 6 was withdrawn to 81.5 inches withdrawn and power was stabilized at 5.0x10-5% power, in anticipation of Low Power Physics Tests. Coincident with the RCS dilution phase was the calculation of 1/M's. 7-~ The calculations were performed every 30 minutes through 1900 and (m / every 15 minutes thereaf ter until criticality was attained. A new base count rate for each startup channel was determined prior to commer.cing the RCS dilution from the average of five one-hundred second counts. Subsequent countrates used for the 1/M's were determined from one 120-second count. A summary of the 1/M's is found in Table 4.1 and Figure 4.1. Estimations of criticality using the 1/M's were as expected. Boron concentrations for the entire approach to criticality are shown on Figure 4.2. Figure 4.3 shows inverse multiplication versus boron concentration. During the dilution portion of this approach, log power channel data was also recorded in order to verify overlap between the startup channels and log power channels. The overlap verification showed that the log power channels behave reliably before the startup channels reach their upper limit. The startup detectors are highly sensitive detectors designed to monitor low power (<10-6% power)
,,_ conditions, while the log power detectors are much less sensitive, fv)
161
- i. O j large, or logarithmic range detectors. Overlap was verified as ;
i shown in Table 4.2 and Figure 4.4. Startup channels 1 and 2 ' ! automatically deenergized at 2218 and 2225, with log power at i 1 1.0x10-6% power. ' i i 4 i 4 l 1 ) i i 1 i [- i l h ! r O
162 e~ L.) TABLE 4.1 Part 1 of 2 1/M EUMMARY FOR THE APPROACH TO INITIAL CRITICALITY TIME CEA GP/ POSITION BORON CONCENTRATION INVERSE MULTIPLICATION STARTUP 1 STARTUP 2 0328 ARI 1.00 1.00 0330 1800 0337 A/25 1.00 1.02 0346 A/40 1.02 1.00 0427 A/60 .95 .98 0430 1800 0446 A/150 .80 .79 0505 B/30 .79 .80 0513 B/50 .81 .82 0530 1800 0547 B/150 .84 .80 0609 P/150 .81 .82 0629 1/90 .82 .80 0630 1800 0641 2/105 .68 .69 0657 3/105 .70 .70 ~ 73 ) g 0723 5/30 .68 .68 0730 1800 0830 1800 0832 6/75 .69 .68 0945 1790 1045 1790 1114 6/75 .67/1.00* .68/1,00* 11249 6/75 - 1780 1.01 1.01 1154 6/75 1770 .98 .99 1230 6/75 1710 .95 .97 1300 6/75 1660 .87 .85 1330 6/75 1560 .82 .81 1400 6/75 1440 .74- .73 1430 6/75 1380 .69 .67 1500 6/75 1230 .64 .62 1530 6/75 1210 .56 .58 1600 6/75 1150 .50 .52 1630 6/75 1090 .44 .45 1700 6/75 1030 .43 .42 1730 6/75 980 .42 .44 1800 6/75 990 .41 .43 1815 6/75 990 .42 .43 1830 6/75 990 .43 .45 1845 6/75 990 .44 .44 1900 6/75 990 .43 .44 p 1915 6/75 980 .41 .42 i U S RCS Dilution commences
- 1/M's renormalized to pre-dilution base countrate
163 f G'O TABLE 4.1 (continued) Part 2 of 2 1/M
SUMMARY
FOR THE APPROACH TO INITIAL CRITICALITY TIME CAA GP/ POSITION BORON CONCENTRATION INVERSE MULTIPLICATION STARTUP 1 STARTUP 2 1930 6/75 970 .40 .39 1945 6/75 960 .39 .38 2000 6/75 940 .37 .36 2015 6/75 920 .34 .33 2030 6/75 900 .31 .31 2045 6/75 880 .28 .28 2100 6/75 860 .23 .24 2115 6/75 840 .17 .18 2130 6/75 820 .12 .12 2145 6/75 820 .03 .03 G l O o i l
164 TABLE 4.2 VERIFICATION OF STARTUP AND LOG POWER CllANNEL OVERLAP LOG POWER CHANNELS TIME STARTUP 1 STARTUP 2 1 l 2 l 3 l 4 (counts) (counts) (% Power) 1124 1986 2015 3.0x10 8 3.0x10-8 3.0x10-8 7.0x10 8 1154 2035 2051 3.0x10-8 3.0x10-8 3.0x10-8 6.0x10-8 1230 2095 2102 3.0x10-8 3.0x10-8 4.0x10-8 5.0x10-8 1300 2293 2396 3.0x10-8 3.0x10-8 4.0x10-8 5.0x10-8 1330 2443 2516 3.0x10-8 3.0x10-8 4,oxio-a 6.0x10-8 1400 2700 2765 3.0x10-8 3.0x10-s 4.0x10-8 5.0x10-8 1430 2908 3029 3.0x10-s 3.0x10-8 4.0x10-s 6.0x10-8 1500 3128 3258 3.0x10-s 3.0x10-s 3.0x10-8 5.0x10-8 1530 3552 3473 3.0x10-8 3.0x10-8 3.0x10-8 5.0x10-8 1600 3968 3915 3.0x10-8 3.0x10-8 3.0x10-s 5.0x10-8 1630 4491 4498 3.0x10-8 3.0x10-8 3.0x10-8 5.0x10-8 1700 4685 4852 3.0x10-8 3.0x10-8 3.0x10-8 3.0x10-8 1730 4716 4662 3.0x10-8 3.0x10-8 3.0x10-8 3.0x10-8 1800 4852 4684 3.0x10-8 3.0x10-s 3.0x10-8 4.0x10-8 1815 4723 4768 3.0x10-8 -3.0x10-s 3.0x10-8 3.0x10-8
- J 1830 4610 4559 3.0x10-8 3.0x10-8 3.0x10-8 3.0x10-8 1845 4570 4659 3.0x10-8 3.0x10-s 3.0x10-8 4.0x10-s. l 1900 4669 4650 3.0x10-8 3.0x10-8 3.0x10-8 3.0x10-8 1915 4855 4815 3.0x10-s 3.0x10-s 3.0x10-s 3.0x10-8 1930 5016 5191 3.0x10-8 3,oxio-s 3.0x10-8 3.0x10-8 1945 5161 5408' 3.0x10-8 3.0x10-8 3.0x10-8 4.0x10-8 2000 5393 5640 3.0x10-s 3.0x10-8 3.0x10-8 4.0x10-s 2015 5883 6105 3.0x10-8 3.0x10-8 3.0x10-8 4.0x10-8 2030 6484 6632 3.0x10 3.0x10-8 4.0x10-8 5.0x10-8 2045 7226 7148 4.0x10-s 3.0x10-8 4.0x10-8 5.0x10-8 2100 8521 8536 4.0x10-s 4.0x10-8 4.0x10-8 6.0x10-s 2115 11606 11451 4.0x10-8 4.0x10-8 5.0x10-8 7.0x10-8 2130 16982 16485 4.0x10-8 4.0x10-8 7.0x10-8 9.0x10-s 2145 65351 62327 1.0x10 1.0x10-7 2.0x10-7 2.0x10-7 2150 240000 180000 3.0x10-7 3.0x10-7 7.0x10-7 6.0x10-7 O
co-->n-r,--rca mamm<a-o o o o o P 9 ~ ~ o.P T. Y Y. t _ Y. . .i.. .7. ?. . .f. . .
. 9. . .J CEA POSITION *~
Bank / Position Le . ARI A/25 ggg i g g A/60 .. A/150 .+ " u-3/$o B/30 . 3 H
*~
B/150 . + g B/150
- 1/90 ,, r
,. 2/105 3/105 + O 5/30 +
3 i
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- 6/75 .+ Re. ~~ ma r o ' $5H~ 3n o - 22,< e nn o zz I N c E ,E-3~~ .o Dilution to VCT +. + r-P Z ~ N g. Initiated
- u<
* ..t'3 9 *
. . P0 Cn
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+,
e- +- Dilution Resumed - Direct Injection
~ **
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++
+
. M
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i I l i l s', so-->=-amnmon uomo. l
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+
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FIGURE 4.3 4
- W3 INITIAL GRITICALITY - 1/M VERSUS. BORON ,
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_ u _s.---,,a --u +a w- x- - - -- u- ----, , .x a as--+-- m.-- - n9.a a-..- a -L- - 1-. . - 169 SECTION 5.0 d LOW POWER PHYSICS TESTING j r I r I I h P I l 1 I
/
e 1 1 l
170 C\ U The Low Power Physics Test (LPPT) program at Waterford 3 SES was conducted between March 4 and March 10, 1985, to verify the physics
- parametere pertinent to the Waterford 3 SES reactor by comparing measured J
results to predicted values. Specifically, the following physics parameters were determined and verified: l 1) CEA Symmetry Checks, i s ', 2) Shutdown CEA and Regulating CEA Worth Measurements,
- 3) Isothermal Temperature Coefficient Measurements, 4)' Critical Boron Concentration Measurements
- 5) Boron Worth Measurements.
O The measurement of.these parameters is discussed in more detail in following sections. The results of the LPPT measurements are summarized in Table 5.0.1. The LPPT satisfied the commitments of the following FSAR Chapter 14 sections:
- 1) 14.2.7.13.3, Psuedo-Ejected CEA,
- 2) 14.2.12.3.10, Isothermal Temperature Coefficient,
- 3) 14.2.12.3.11, Critical Boron Cor. centration, 4). 14.2.12.3.12, Shutdown and Regulating CEA Group Worth,
- 5) 14.2.12.3.13, ~ Inverse Boron Worth
- 6) . '14'. 2.12. 3.14, CEA Symmetry.
1
- I 171 l
O It should be noted that the psuedo-ejected CEA test was not performed as this test was (,nly to be performed if the remainder of the' low power physics tests demonstrated that the Waterford 3 SES core was significantly different from the San Onofre Nuclear Generating Station, Unit #2 (SONGS 2) core. Since the Waterford 3 SES core was demonstrated not to be significantly different from the SONGS-2 core, this test was not performed. The only major problem encountered during the LPPT were related to slipping / dropping CEAs and failures of the plant monitoring computer
.-(PMC). ' Most of the CEA problems were only minor slips and/or drops which were quickly recoverable. However, on March 6,1985 at 1348, problems were encountered with CEA #38 when it dropped from approximately 75" withdrawn. The problems with CEA #38 were not resolved until 0230 on March 7, 1985', creating a delay in the LPPT of almost 13 hours. The PMC failures encountered during the LPPT n's occurred primarily during the CEA symmetry check portion of the LPPT. Most of the PMC failures were easily resolved by rebooting the computer. The PMC failures usually were not major problems, simply a hindrance to continued testing, generally creating only
- minor delays.
All Low Power Physics Test objectives and acceptance criteria were satisfied. 5.1 CEA Symmetry Cheeks The CEA symmetry check was performed to verify the proper, symmetric loading of CEAs within symmetric CEA groups, the proper coupling of each CEA to its extension shaft, and to verify that no core loading or fuel fabrication errors had occurred. j \. -
I l 172 l i t ) QJ Each CEA should have approximately the same reactivity worth as its symmetric counterparts. This was verified by trading the withdrawal of a CEA with the insertion of a symmetric CEA and measuring the reactivity worth difference between the fully inserted CEA and the fully withdrawn CEA. The CEAs were divided into 16 symmetric groups. The smallest symmetric group contained two CEAs, while the largest groups were made up of eight CEAs. The first CEA inserted from each symmetric group was the reference CEA. 'At the completion of the measurement of a symmetric group, the reference CEA was again fully inserted. Reinserting the reference CEA was per formed in order to collect data required to correct individual worth measurements for any drif t which occurred during the measurements of the symmetric group.
-~
The center CEA, which has no symmetric counterpart, was inserted to
\m ,/ produce a negative deflection in the reactivity computer output trace in order to verify that the CEA was coupled to its extension shaft.
All 16 symmetric groups were successfully tested for symmetric CEA worth. All CEAs within a symmetric CEA group were demonstrated to be within 11.5 cents of the symmetric CEA group average deviation, as shown in Figure 5.1.1. All CEAs were shown to be properly coupled to their extension shaft, and no evidence was found which would indicated the presence of either a core loading or fuel fabrication error. All CEA symmetry check acceptance criteria were satisfied. l l I f~%.
- N~Y
WATERFORD-3 173 FIGURE 5.1.1 l CEA SYMMETRY TEST ' KEY:
-N s.22 - CEA Numbe r M -Deviation from Symmetric Group Average, A-@
.55 2-73 2-74 i
.88 . 64 9-83 3-63 4-2 3-64 9-94
.55 .M . 94 45 25 1-72 A-51 8-3ll 9-40 A-52 1-75 74 .51 .87 .55 .24 .83 9-62 5-45 P2-31 6-22 P1-32 5-46 B-85
. 40 .48 .32 .5B .15 +.15 .18 A-50 A-25 9-15 B-16 A-28 A-53
[ .14 .29 . 42 . 44 .28 .23 \ 3-62 P1-E 4-9 1-6 4-10 P2-33 3-65
.00 .00 21 .28 .33 .12 +.02 2-71 9-30 B-14 A-2 9-17 9-41 2-78
+.23 +.24 +.15 .03 . 43 .24 .20 A-91 4-57 6-21 1-5 1-7 6-23 4-59 _ gg
^~88 +. 2 +.42 +. 2 .38 .40 .22
., y 2-70 &-37 B-13 A-3 9-18 9-42 -W
+.51 +. 5 +. E +.03 37 . 42 .17 3-61 P2-29 4-6 1-4 4-11 P1-34 3-68
+.51 +.24 +. 48 +. 33 +. 00 .10 .08 A-49 A-24 &-12 9-19 A-27 A-54
+. 78 .+.82 +.83 +.37 .28 .15 9-61 5-44 P1-28 9-20 P2-35 5-47 9-e8
+.87 +.37 +. 28 +. 2 +.24 .04 .12 1-M A-48 9-M B-43 A-55 1-78
+1.07 +. M +. 82 +. 53 +.13 +.31 B-E 3E 4-2 3-67 9-67
+. 61 +.52 +. 2 +.54 +.19 2-S 2-79
+.571 +. 5 A-M n
'd 35 S/G #1 e
- , ~~ . . . _. - ~ . _ . - - - - .
174 , i' 5 '. 2 Shutdown CEA and Regulating CEA Worth Measurements J I i To measure the group worth of the various CEA groups, regulating ' groups 6 through 1,. shutdown group B and the part length CEA group 1 were diluted into the core in the manual group (MG) mode and the
. magnitude'of the reactivity change was measured from the reactivity !
traces produced by the reactivity computer. The nonoverlap group worths are tabulated in Table 5.0.1. The resulting CEA group worth curves are shown in Figures 5.2.1 through 5.2.4. All CEA group worth acceptance criteria were satisfied.
-The total inserted worth (nonoverlap) of CEA groups 6 through 1, group P and group B was calculated from the measured data as 11.252%
4 Ak/k. The predicted' total worth was 11.327% Ak/k. The measured-i total worth differed from the predicted . total worth by only -0.66%, l- ~ wel1 within the 10% acceptance criteria. i i The shutdown margin at the zero power insertion limit (ZPIL) was 4 calculated to be 9.398% Ak/k, verifying that the CEA. Insertion Limit
; was acceptable since >5.15% Ak/k margin was available.
e l 5.3 -Isothermal Temperature Coefficient Measurements l
; The isothermal temperature coefficients (ITCs) were measured by slowly raising and then lowering RCS temperature approximately five
' to, ten degrees F while maintaining constant boron concentration and CEA position. The resulting reactivity changes were calculated by the reactivity computer and recorded on an x-y plotter as a function of RCS' temperature. The slop of.the line produced on the x y plotter -was .the .ITC.
~
The ITC.was measured for the following CEA configurations: T
,,.-.,,e - .ym. - ,-,..., ,_,,rm ,m.v, + - . . , _ . , , , ,,.,e-~. -. - . . . ~ - .,.~,,e., .-u- . - _ _ . , , - ,
Figure 5.2.1 175 INTEGRAL CEA WORTH, NO OVERLAP O' Group 6
-0.35 -
- 0.3 -
7
<3 -0.25 -
5 l 0
-0.2-3 g -0.15 -
8
- 0.1 -
- 0.OS -
O , , , , , , , , , , , , , , O 20 40 60 80 100 120 140 4 0 n CEA POSmON (inches inserted) BOL HZP (mecoured) Group 5
- O.35 -
- 0.3 -
2 1
<3 -0.25 -
E i h -0.2-O 5 l g -0.15 - i O 5
- 0.1 -
l l
-0.05 -
O O , , , , , , , , , , , , , , O 20 40 60 80 100 120 140 l CEA POSmON (inches Inserted) D BOL.HZP (mecoured)
l l Figure 5.2.2 176 r INTEGRAL CEA WORTH, NO OVERLAP Group 4
-1
- 0.9 -
l
- 0.8 -
i
- 0.7 -
d 3 - 0.6 - h - 0.5 - O E g - 0.4 - 8 g -0.3-
- 0.2 -
- 0.1 -
O , , , , , , , , , , , , , , O 20 40 60 80 100 120 140 O CEA POSmON (Inches Inserted) BOL.HZP (measured) Group 3
- 0.9 -
g -0.8 -
}
q - 0.7 - U
- 0.6 -
t 0 - 0.5 - R
- 0.4 -
E o - O.3 -
- 0.2 -
- 0.1 -
(] O , , , , , , , , , , , , , () O 20 40 60 80 100 120 140 l l D CEA POSmON (Inches Inserted) BOL.HZP (measured)
I Figure 5.2.3 177 ('j~' INTEGRAL CEA WORTH, NO OVERLAP Group 2
- 0.7 I
- 0. 6 -
2 - 0.5 - 1 U -0.4-5 8 3 - O.3 - a. 3 O
$ - O.2 -
- O.1 -
O , , , , , , , , , , , , , , O 20 40 60 80 100 120 140 0 o CEA POSmON (Inches Inserted) BOL.HZP (mecoured) Group 1
- 1.2 - '
- 1.1 -
k.x - O.9 - 4 - O.8 - E g - 0.7 - 8 - O.6 - 3
- a. - 0.5 -
8 g - 0.4 - 0
- O.3 -
- O.2 -
- 0.1 -
ft
\j O
O 20 40 60 100 80 120 140 o CEA POSmON (Inches Insert'ed) BOL.HZP (measured)
1 178 O lisure 5.2.4 INTEGRAL CEA WORTH (GROUP P)
- - 0.45 -
- 0.4 -
- 0.35 -
7 1 - 0.s -
<3 M
-O.25 -
- O.2 -
(L o -0.15 - E o
- O.1 -
-0.05 -
O , , , , , , , , , , , , , , O 20 40 60 80 100 120 140 O CEA POSITION (Inches inserted) EOL.HZP (rneosured) I
~
O
,,...,_,c_,. . - , . . . . _ . , , _ . , .,-.,-.,-,,,,,_,,.,,.,.,,,y._
179 m
- 1) CEA Group 6 >135 inches withdrawn (essentially all rods out (EAR 0))
- 2) CEA Groups 6 through 3 at the lower electrical limit (LEL)
- 3) CEA Groups 6 through I at the LEL The ITC is the sum of the fuel temperature coefficient (FTC) and the moderator temperature coefficient (MTC). The FTC is a negative constant (supplied by the reactor vendor), while the MTC may be slightly positive at high boron. concentrations and large and negative for low concentrations. The MTC was calculated by subtracting the FTC from the measured ITC.
The results of the ITC/MTC tests are summarized in Table 5.0.1. All ITC/MTC acceptance criteria were satisfied. .(- 5.4 Critical Boron Concentration Measurements The critical boron concentrations (CBCs) were measured by stabilizing'the plant in the desired configuration and maintaining temperature, pressure and CEA position as constant as possible. Once the RCS conditions were stable, RCS boron samples were collected and analyzed to determine the actual CBC. During the LPPT, it was not always practical to establish the CEAs at exactly the required position; rather the CEAs were positioned near the desired point and then briefly withdrawn or inserted to the required position and the residual reactivity worth measured. The measured residual reactivity worth was then converted to an' equivalent boron concentration and added to or subtracted from the measured CBC. JCritical-boron concentrations were measured-in the same p) q, configurations as, and just previous - to, the ITC/MTC measurements. The CBC test results are summarized in Table 5.0.1. All critical
-boron concentration acceptance criteria were satisfied.
180 [d'T 5.5 Boron Worth Measurements The inverse boron worth (IBW) was calculated using CBCs and CEA group worths measured previously. The IBW was calculated by dividing the difference in CBCs by the difference in CEA group worths between two CEA positions. Inverse boron worths were calculated for the following configurations:
- 1) CEA Groups 6 through 3 fully inserted, and
- 2) CEA Groups 6 through I fully inserted.
! The results of the IBW calculations are summarized in Table 5.0.1. All inverse boron worth acceptance criteria were satisfied. x--) , l
. 1 I
, . - - - - ~ , , - - - , , , e,.as, n., , - - - - - - - a-,-,- . - - - - , . - - - , , . . - - - - -n , , , - , . .-- -e.-.~--._-.. .,
4 181 I p i TABLE 5.0.1 WATERFORD 3 SES LPPT RESULTS i MEASURED ACCEPTANCE PARAMETER UNITS VALUE CRITERIA 5.1 CEA SYMMETRY CHECKS C 1.07C (max) within 11.5C (see Figure 5.1.1) of group average 5.2 CEA GROUP WORTHS Group 6 Mk/k 0.392 0.409 10.05 Group 5 Mk/k 0.383 0.390 10.05 l Group 4 Mk/k 0.913 0.913 10.09 l Group 3 Mk/k 1.008 1.020 10.10 Group 2 Mk/k 0.640 '0 662 10 07
-Group 1 %ak/k 1.230 1.214 10.12 Group B Mk/k 3.008 3.076 0.31 >
Group P Mk/k 0.397 0.369 10.05 Total Inserted Worth 7,ak/k 11.252 11,327 tl.13 Shutdown Margin- Mk/k 9.398 >5.15 at t e- ZPIL i i 5.3 ITC MEASUREMENTS ITC @ EARO Ak/k/*F -0.391x10-4 -0.393x10-4 !0.3x10-4 MTC @ EARO Ak/k/*F -0.235x10-4 -0.237x10-4 10.3x10-4 ITC w/6-3 @ LEL Ak/k/*F -1.025x10-4 - 1.302x10-4 10.3x10-4 I. MTC w/6-3 @ LEL Ak/k/*F -0.870x10-4 -1.142x10-4 10.3x10-4 i
'ITC w/6-1 @ LEL Ak/k/*F -1.530x10-4 --1.810x10-4 10.3x10-4 MTC w/6-1.@ LEL Ak/k/ F -1.370x10-4 -
1.650x10-4 10.3x10-4 5.4 CBC MEASUREMENTS
-CBC @ EARO ppm 829.10 832 150 CBC.w/6-3 @.LEL ppm 619.52 629 150 CBC w/6-1 @ LEL ppm 506.80 499 150 5.5 INVERSE BOP 0N WORTH MEASUREMENTS IBW w/6-3 @ LEL ppm /Mk/k -77.7 -74.3 110-IBW w/6-1 @ LEL - pps/Mk/k -60.0 -69.7 10 O
-,,--,-+-g v,v--rev- ,,-,-w ae-y- -
---w-- ve *w ree.,s.-e. - w,m- -g--=+ ,g* y ---ws.y
e _ -- , --
. 182 l
l SECTION 6.0 POWER ASCENSION TESTING I l . i l l i t
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183' m I. i 6.1 POWER LEVEL DETERMINATION
- 6.1'.1 Reactor Coolant System Delta-T 4 Power Determination (SIT-TP-704)
PURPOSE: The purpose of this test was to determine the thermal output of the reactor at power- levels. up to approximately 20% by means of a primary system calorimetric. The power level m a
- calculated in this test was then used as the standard for calibrating the core protection calculators and the excore
~
nuclear instrumentation (see section 6.2.1). r METHOD: RCS hot leg and cold leg temperatures were recorded and averaged. (The hot-leg temperatures were corrected for temperature bias, previously determined during hot zero power plant conditions.) From these temperatures the enthalpy rise across the core (Ah) was determined. The core thermal power 1 (Q) was then calculated by multiplying Ah by the core .nass flow rate (M). .The core thermal power (Q), when divided by the rated thermal power (RTP) of the core (3390 MWth), and i multiplied by 100%, yielded the percent of rated thermal-power at.which the reactor was operating.
~
y Thus: - Q(MWth) = M(lb'/hr) a Ah(BTU /lb )/3412141.(BTU m /MWth-hr)
, and-l l
l=
/"\ % RTP = Q(MWth)x100(%)/3390(MWth)
V l l l 4
,, .,. , + , -m,
184 O,q g RESULTS: The reactor coolant system delta-T power determination was performed ten times during the power ascension test program. There were no significant difficulties encountered in the
- performance of this test. ,5-
_ The results are summarized in Table 6.1.1.1 below. s,
'., r TABLE 6.1.1.1 '
RCS DELTA-T POWER DETERMINATION TEST RESULTS
-Date RCS Mass Flow Rate Ah Reactor Power 6
(10 1b,/hr) (BTU /lb,) (%) 3/17/85
~
169.71 3.04 4.46 3/17/85 169.86 2.86 4.20 3/19/85 169.42 8.57 12.55 3/20/85 168.69 14.58 21.26 4/12/85 170.16 11.92 17.54 4/13/85 169.92 14.18 20.83 4/14/85 169.86- 14.32 21.02 4/14/85 170.05 14.48 21.29 4/18/85 170.08 13.73 20.19-4/18/85 170.04 14.24 20.93
. CONCLUSION:
- Reactor power was satisfactorily determined, thereby providing a reliable standard for calibration of the core protection calculators..and excore nuclear instrumentation. -All test objectives were. met and acceptance criteria satisfied.
O
- v h
m
185 4
\.
y s.,) 6.1.2 NSSS Calorimetric-(SIT-TP-709) PURPOSE: , The NSS$ calorimetric power measurement provided an acenrate deternination of reactor power based on a secondary plant energy balance. This power measurement was used to meet the following objectives:
- 1. Verify that the secondary calorimetric based core power calculated by the Core Operating Limits Supervisory System (COLSS) is an accurate determination of core power.
J
- 2. Calibrate the reactor coolant system delta-T power
~~
determined by COLSS to-achieve satisfactory agreement with the secondsry calorimetric power.
- 3. Verify the secondary heat balance calculations performed in one of the nuclear engineering procedures.
- 4. Verify the secondary heat balance calculations performed by an off-line applications computer program.
METHOD: Stable initial conditions for reactor power, reactor coolant system temperatures, pressure, feedwater temperature and steam generator levels were established. Data was then collected at 30 second. intervals for one hour and averaged. The average values'of the various parameters were used to calculate reactor power from the following equation:
?\
Q.
l l i 186 . l j\ - ( .
% Reactor Power =
'2
[{ I (M f -M.bd )(h s -hg )+Mbd (hbd -h f }}+Mch(hId-hch)+Qloss-Qrep -Qpar} xKx100% i (=1 g g g g i g 3390 MWt , Where: Mf,. = Mass flow of feedwater supplied to steam generator i 1 M bd.
= Mass flow of blowdown from steam generator i 1
h,, = Enthalpy of steam from steam generator i (corrected for 99.8% quality) 1 hg = Enthalpy of feedwater supplied to steam generator i 1 hid. = Enthalpy of blowdown from steam generator i 1 l
-MehL= Mass flow of charging to'the reactor coolant system
- ' -h ld = Enthalpy of letdown from the reactor coola'nt system
~
- h = Enthalpy of charging to reactor coolant system ch Ql oss = RCS Heat Loss Q rep = Heat input from reactor coolant pumps .
JQ = Heat input from pressurizer pzr K = Conversion factor from BTU /hr.to MW = 1/(3.412 x 106 ) 100% = Conversion factor from fraction of full-power to % of full power 3390 MWt = Rated full core power level NOTES All steam and water properties (i.e. enthalpy,' specific volume) were obtained from 1967 ASME Steam Tables. I g e e 8
+
w +e w~ -,*- % e ,-v .*t-,-5 .e-= y, e4g e +ge-+- e v e - e-+ -w>w-++ r ,--oe e.+g-ww<--c,
187 7i i .J For both equilibrium and non-equilibrium xenon conditions the hand calculated calorimetric power was compared to the average COLSS calculated secondary calorimetric power (BSCAL). At equilibrium xenon conditions, these values were to agree within 0.2% of rated thermal power (2 0.5% at 20% power); at non-equlibrium xenon conditions, they were to agree within 2.0% at all test plateaus. If necessary the test data and COLSS constants were evaluated and this procedure repeated until the desired agreement was obtained. For both equilibrium and non-equilibrium xenon conditions the hand calculated calorimetric power was compared to the average core delta-T power (BDELT). At equilibrium xenon conditions these values were to agree within 1 0.2% of rated thermal power (t 0.5% at 20% power); at non-equilibrium xenon condi- . -g tions,.they were to agree within i 2.0% at all test plateaus. N,,) If this agreement was not achieved, a new delta-T power gain, E19,.was calculated from the following relationship: E19 = Average BSCAL (Average Static Delta-T Power - E20) where E20 is the delta-T power bias term.
.After setting the delta-T power gain into the plant computer, a new set of data were taken at 30-second intervals for 5 minutes. The average value of BDELT was then' compared with the hand calculated calorimetric power or an average value of BSCAL and the difference verified as acceptable.
d( -
188 b When the plant was at equilibrium xenon, NE-5-201, Heat Balance Calculation, and NE-72-03, POWER Program were executed ; using the average data collected for this procedure. The
. power. levels calculated by means of these two procedures were compared to the calorimetric power calculated for this test. !
If these power levels agreed within 0.5% ( 1.0% at 20% power) 'of the calorimetric power, then NE-5-201 and NE-72-03 were considered acceptable. RESULTS: The NSSS calorimetric procedure was performed thirteen times between April 13 and July 3, 1985. No significant problems were encountered in performing this procedure. Results of
~~
s each test are summarized in Table 6.1.2.1.
- \s -
I CONCLUSION: The power levels calculated from NE-5-201 and NE-72-03 were in good agreement with the hand calculated calorimetric power, l thereby verifying acceptability of the methodology and accuracy of both procedures. k The COLSS delta-T power (BDELT) and COLSS calorimetric power j '(BSCAL) were within the required tolerance of the hand calcu-lated power, although on several occasions it' was necessary to adjust the delta-T power gain and show acceptability of the - i BDELT value by comparison to BSCAL. All acceptance criteria l 7 for non-eqpilibrium and equil.ibrium xenon were satisfactor,ily met. U . e a w . -p.+ .- - y-.p-- ,.m,wy,, . -'w,
O O O TABLE 6.1.2.1 NSSS CALORIMETRIC RESULTS (I Date Calorimetric Power ( BSCAL( BDELT I) BDELT( BSCAL( NE-5-201 Power ( Power Program ) E<1uilibrium (before) (a f ter) (a f ter) -(Yes/No) 4/13 18.73 18.73 21.04 18.77 - - - NO 4/14 19.20 19.35 21.34 19.44 19.48 19.31 19.30 YES 4/18 29.97 29.99 27.27 30.39 - - NO 4/19 41.14 41.00 41.08 41.08(3) - - - N0 4 4/20 49.97 49.97 50.08 50.08(3) - - - NO 4/21 50.57 50.63 50.12 50.11 50.11 50.34 50.30 YES 5/6 59.22 59.22 59.63 59.63(3) - - - NO 5/7 70.21 70.25 70.73 70.73(3) - - - NO i 5/8 79.72 79.70 81.06 81.06(3) - - - NO 5/9 80.64 80.65 82.31 79.44 79.40 80.63 80.60 YES 7/1 94.11 94.16 94.67 94.67(3) - - - NO 7/2 98.63 99.07 100.80 101.58 100.67 - - NO 7/3 99.29 99.32 98.69 99.56 99.44 99.15 99.30 YES (NOTES to table on next page)
~
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~. . . . - . - - . -.- - . . -
r 190 l Notes to-Table 6.1.2.1 (1)' Criteria used to determine the acceptability of a parameter were as follows: i Non-Equilibrium Xenon
- a. -2.0% 5(Calorimetric Power)-(BSCAL) 52.0% and (before) ,
i i
- b. -2.0% 5(Calorimetric Power)-(BDELT) 52.0% or (before)
-2.0% $(BSCAL)-(BDELT) 12.0%
(after) (after) Equilibrium Xenon 4
- a. -0.2% 5(Calorimetric Power)-(BSCAL) 10.2% (10.5% at 20% power) and p (before)
- b. -0.2% 5(Calorimetric Power)-(BDELT) 50.2% ( 0.5% at 20% power) or (before)
[ -0.2% 5 (BSCAL)-(BDELT) 5 0.2% ( 0.5% at 20% power) (after) (after)
- c.
-0.5% 5(Calorimetric Power)-(NE-5-201 Power) 10'5% .
d ., -0.5% 5(Calorimetric Power)-(POWER Program) 10.5% i (2) All power values reported in units of % RATED THERMAL POWER
- j. .(3) Cal'ibration not required
- 6
, ,, , , ., .,, _,,......,.m.. . . , . . , , , , _ , , , , _, , . . , , , , . _ _ _ . _ _ _ _ , , _ , , _ _ , . , ., ,,,,,a,_ , , , . _ _ _ , , ,
4 191 5
' (-sI N_s' ,
6.2 INSTRUMENTATION TESTING / CALIBRATION 6.2.1 ' Nuclear and Thermal Power Calibration (SIT-TP-705) J i
; PURPOSE:
i The objective of the nuclear and thermal power calibration , test' was to calibrate excore linear power, CPC thermal power , .. and CPC nuclear power to a standard measurement .of core power. I 4 FSAR Chapter 14, Section 14.2.12.3.27, Steady-State Core
- Performance, was partially satisfied by performance of this
, test. E
-s s METHOD:
)
- Initial conditions were established with the reactor at l steady-state conditions. A standard measurement of core i
. power was then determined by one of the following methods:
1 .
- t. A. Up to approximately 20%~ rated thermal power, reactor power was calculated from a primary system calorimetric measurement (i.e., RCS delta-T power measurement; see
;- 'section 6.1.1).
4 1 B. Above 20% rated thermal power, reactor power was obtained l-
- from a secondary energy balance . calculation performed by .;
j the Core Operating Limit Supervisory. System. I i (~~ 1 1
_ . _ . ._ .- _ m ~ _ _ _ l l 192 l' i
.f~'$
4 V. - 4
-Utilizing the standard power, a new voltage output f rom the excore linear amplifier was determined. The amplifier gain
.of one channel was adjusted to obtain the new voltage. After performing the adjustment for one channel and verifying acceptable agreement between excore linear power and standard power, the remaining three channels were. adjusted to agree with the first.
Calibration of CPC nuclear power (PHICAL) and thermal power (BDT)"was accomplished by changing the respective values of
; the addressable calibration constants. These constants were computed from the following:
4 i - New Constant = Old Constant x Standard Power
; Indicated Power
() A'f ter installing the new constants in the CPCs, both the thermal and nuclear power indications for all channels were verified to be in agreement with the current value of standard
- , power.
l RESULTS: The nuclear and thermal power calibration. procedure was performed thirty-four times between March 17 and. July 10, 1985. In all but three cases,^ the PPS excore linear power and t CPC powers were calibrated within the required tolerance. In , two of these cases, a reactor trip occurred before the cali-bration procedure could be completed. Subsequent performance - of the procedure ensured that the parameters of interest were 4 calibrated within specification. l
.(-~\
, s_/ - 6 l l L .
193 l Q} During the performance of this procedure at a nominal 5% power on March 17, 1985, both CPC addressable calibration constants were required to be adjusted for all four CPC channels. However, the new nuclear power calibration con-stants for channels B, C and D and the new thermal power calibration constant ~ for channel A were found to exceed the maximum values allowed by Technical Specifications. Conse-quently, the maximum allowable values were installed in place of the calculated values. This resulted in acceptable cali-bration of channels B and D. Channels A and C remained out-of-tolerance. Similarly, the channel C and D excore linear power _ indications remained out-of-tolerance af ter adjusting the amplifier gain potentiometers to maximum. Two
- days later, on March 19, 1985 these discrepancies were cleared by reperformance of the test at approximately 15% power. All o
g PPS and CPC power indications were then calibrated to the s_,) required tolerance.
- Table 6.2.1.1 summarizes the date and power level at which each procedure was performed.
d k [ p. me,
194 v. TABLE 6.2.1.1 NUCLEAR AND THERMAL POWER CALIBRATION i DATE PERFORMED STANDARD POWER (%) DATE PERFORMED STANDARD POWER (%) 3/17/85 4.20 5/6/85 59.57 ! 3/19/85 12.71 5/7/85 71.50 3/20/85 21.26 5/8/85 79.13 " 4/12/85 17.54 5/9/85 79.55 4/14/85 19.35 5/19/85 24.14 4/14/85 21.29 5/20/85 61.77
'4/18/85 20.20 5/27/85 80.37 4/18/85 19.98 6/25/85 79.42 4/19/85 29.76 6/26/85 89.85 4/19/85 41.50 7/1/85 93.79 4/20/85 49.31 7/1/85 98.44 4/21/85 50.94 7/2/85 100.00 4/23/85 50.23 7/3/85 99.98 5/2/85 50.69 7/5/85 60.18 i
5/4/85 19.50 7/7/85 88.97 5/6/85 50.09 7/9/85 99.30 5/6/85 49.88 7/10/85 99.35 3 CONCLUSIONS: . With the above noted exceptions,-the PPS and CPC power
; indications were successfully calibrated to the standard indication of reactor power. For the cases in which the calibration was out of tolerance or not completed, associated deficiencies were cleared and the procedure was reperformed
, successfully.
t-k ) l
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i . . _. ,m,., . , - , - . ,, _ - , , , . _ _ _ _ , _ , . _ . . _ _ ~ .- ~ _ . , , , , , . . . _ - - ~ ~ , . . , . . . . _ , - . .
195 i J 6.2.2 Process Variable Intercomparison (SIT-TP-712) s PURPOSE: E E
.The purpose of this test was to demonstrate that the inputs i
and appropriate outputs of the Plant Protection System (PPS), the Core Protection Calculators (CPCs), and the Plant
~
Monitoring Computer (PMC) were in satisfactory agreement with one another. Permanent plant instrumentation (meters and recorders) were also included in the intercomparison. { This test satisfied the commitments of FSAR section
; 14.2.12.3.30.
i METHOD:
. ( ~)
V Plant conditions were stabilized at each of the four test
= plateaus -- 20%, 50%, 80%, and 100% power -- during the.
- initial power ascension following core load. Data from each -
of the four sources (PPS, CPCs, PMC, and permanent plant instrumentation)were simultaneously gathered for each of the following parameters: J 4
- 1. RCS cold leg temperature
- 2. RCS hot leg temperature
- 3. RCP differential pressure
- 4. RCP speed
'S. RCS~ pressure j m x).
7
, r. _ . .-. ,v , , - . , _ . . . . _ .,.-_.m__. ..m .
I 196
- s_/
(O 4
- 6. Pressurizer level
- 7. Steam generator level
- 8. Steam generator pressure
- 9. Steam generator primary side differential pressure
- 10. Reactor vessel differential pressure
- 11. Containment pressure
- 12. Refueling water storage pool (RWSP) level.
, Based upon the data gathered for each parameter, a target value was calculated as the average of the readings from the
\~s/ most reliable source; the' order of , reliability of data sources, from most reliable to least, was as-follows:
- 1. Core Protection Calculator data
- 2. Plant Protection System data
- 3. Plant Monitoring Computer data 4
- 4. Control Board Instrumentation data The deviati.on of each recorded value from this target value was calculated and compared to the specified tolerance to determine acceptab'ility. If the deviation exceeded the specified tolerance and a test deficiency was generated,
, recalibration of the loop was initiated. The deficiency was cl~ eared only when subsequent testing revealed that the
() parameter deviation fell within the specified tolerance. O .
I I 197
,/ 5 RESULTS:
At the 20% power plateau, thirty-two deficiencies were
. generated; seventeen of these were written against-RCS hot and cold leg temperature indications. Four RCP speed sensors were out-of-tola-ins: as well es . "r RWGP level indications. Of the remaining a..ven ouc-of-specification indications, six were PMC or control board-related.
At the 50% power plateau, eight additional deficiencies were generated; seven of these were RCS RTD-related. Of the fifteen deficiencies written at 20% power that were not related to RCS-RTDs, all but two had.been reworked to meet the specified tolerances. Thus, the total number of outstanding deficiencies following completion of 50% power
-s testing was twenty-eight; twenty-four of these were RCS x_ ,e temperature indications.
At the 80% power plateau, ten new' deficiencies were generated; seven of these were RCS temperature indications.
~
All deficiencies outstanding from the 50% plateau which were not related to RCS RTDs had been reworked and found acceptable at 80%. Thus, the total number of outstanding deficiencies following completion of the 80% power testing was thirty-four; thirty-one of. these were RCS temperature indications. At the 100% power plateau, eleven additional deficiencies were written; seven of these were RCS temperature indications. One of the three non-RTD-related deficiencies
- from 80% had been successfully reworked, leaving a total of six non-RTD-related deficiencies. Thirty-eight deficiencies
~
relating to RCS temperature indication remained outstanding.
/~'i L-
%s 9
y m .., -
, , - - - - , -- - , .r- - - , _ - . - - - , .----
198 p , '\ , / Two of the deficiencies not related to RCS temperature were 4 reworked and satisfactorily retested; the other four deficiencies have also been reworked, but have not yet been retested. Three of the four deficiencies were written against PMC parameters, while the fourth was written 'against a control board instrument; no PPS or CPC parameters, with the exception of the RTDs, were out-of-tolerance at the completion of 100% power testing. t Discussion of RCS RTD Problems Problems with the RCS RTDs were first noted during the per-formance of SIT-TP-501, Pre-critical Intercomparison of PPS, CPC, and PMC Inputs (see also section 3.1.1). An evaluation of the magnitude of the RTD errors found that operation at the g 20% power plateau would not represent an unsafe condition. _ s_,) - Analysis and troubleshooting 'of the RTD problems continued. Troubleshooting of the problem revealed the following:
- 1. The RTDs which provide hot and cold leg temperature indication to CPC channels A and B were significantly more in error.than those providing indication to channels C and D. The former RTDs were manufactured by Weed, while the latter RTDs were manufactured by Rosemount.
Initially, all four CPC channels were driven by Rosemount detectors, but the requirement for dual-channel RTDs to accommodate the Qualified Safety Parameter Display System (QSPDS) led to the installation of-the Weed RTDs. T 'J f% - (v! . i - i
- I l
. ' , .-. n , - -
199
. ;O V
- 2. The primary components of the Weed RDT error were identified as inaccuracies in the RTD curve used to relate resistance to degrees Fahrenheit, normal ,
calibration inaccuracies, and a thermocouple effect which was discovered at- the RTD junction; this thermocouple effect involved a potential difference induced *across the RTD leads which biased the input to the temperature
- transmitters.
- 3. Corrosion at the various terminations between the RTD and the process analog control (PAC) cabinets and noise induced within the PAC cabinets due to inadequate shielding also contributed to the errors, but to a : sser extent than those items detailed in (2) above.
' p)
\s_-
Corre-tive 3ction was initiated as follows:
- 1. Four Weed RTDs were replaced, two each in channels A and B.
-:2. All RTDs, both Rosemount and Weed, were carefully recalibrated.
- 3. All terminations .were cleaned of corrosion and ensured tight.
- 4. Evaluation of the impact of the greater-than-anticipated RTD inaccuracies was performed. As additional RTD data became available, this impact was continuously reassessed. The results of the. analyses,~ performed by Combustion Engineering, Inc., are detailed below:
4 v) . . F
. , -- . . _ , ~ . _ _ ., - - - , , - , . - . - . . _ . . - .
. . , . - . . _ . . . ~ .- . - .
200-
.u() . ,
- The impact of the greater RTD inaccuracies on the CPC calculations was twofold. First, the inaccuracy in measured RCS flowrate was increased; second, the accuracy of core thermal power BDT was adversely affected.
- Operation at reactor power levels of up to 50% was deemed acceptable, on the condition that the PPS high linear power trip setpoint was adjusted to 65% instead of 70% power. This encured that sufficient margin remained in the CPC DNBR and LPD calculations to account for the increased uncertainty associated with ,
the RCS flowrate; si o, induced errorr, in BDT were of greatest concern only above 65% power' d.-ing a CEA deviation event. i -%g s ,) - A^ preliminary analysis was conducted during plant testing .at 50% power to determine whether ' full power operation would be permissible given the magnitude of the RTD errors. The limited availability of RTD data
- did not provide adequate assurance that safe operation at 100% power could be achieved. However, operation 'at reactor power levels of-~up to 90% was approved,- on the condition that the following penalties were applied to.all four CPC channels:
BERR1 - increase by 8.3% (above original value). BERR3 - increase by 4.2% (above original value) PFMLTD - increase,by.17.4% (above original ~value)
,. PFMLTD - increase by 17.4% (above original value)'
) -
~
201
,/3 I )
%,) These additional penalties were incorporated into the CPC channels prior to increasing power above 50%. Extensive RTD data collection was then initiated to support a more accurate analysis of the problem.
- Based upon the extensive RTD data collected, a more detailed analysis was performed. It was determined that the channel C and D RTDs (manufactured by Rosemount) exhibited errors within the allowable range. Also, channel A and B RTD errors were better quantified, and more appropriate penalty factors were developed to assure conservative full power
!cration. Operation it power levels up to 100% was approved, on the cot..!: Lion that the following penalties were applied to CPC channels A and B only:
/o (j 1 BERR1 - increase by 3% (above original value) BERR3 - increase by 4% (above original value) PFMLTD - increase by 17.8% (above original value) PFMLTL - increase by 17.8% (above original value) These modified penalties were incorporated into CPC channels A and B prior to operation above 90%. The penalty factor addressable constants in channels C and D were restored to their original values.
- Continuous monitoring of the performance of the CPC RTDs was made a prerequisite of continued power operation, to ensure that the temperature errors remain within the bounds of the Combustion Engineering
()
,..m analvsis. This monitoring will be performed by a Nuclear Engineecing procedure whenever four reactor
202 (")
.o coolant pumps are in operation. The procedure also provides for the calculation of new penalty factor addressable constants in the event the temperature errors increase beyond current values.
In summary, the problem with CPC RTDs was discovered via the performance of SIT-TP-501 and SIT-TP-712. Troubleshooting conducted by the plant Instrument and Controls Department-revealed the primary causes The most erroneous RTDs were replaced, and recal}}