Advances in Light Water Reactor Technologies (eBook)

Chapter 1: Application of Probabilistic Safety Analysis in Design and Maintenance of the ABWR 14
1.1 ABWR Design 14
1.1.1 ABWR Development 14
1.1.2 ABWR Technical Features 16
1.1.2.1 Reactor Pressure Vessel and Internals 16
1.1.2.2 Reactor Internal Pumps 18
1.1.2.3 Fine Motion Control Rod Drive 18
1.1.2.4 Improved Core 19
1.1.2.5 Emergency Core Cooling System 19
1.1.2.6 Reinforced Concrete Containment Vessel 19
1.1.2.7 State-of-the-Art IandC Technologies 19
1.1.2.8 Turbine System 20
1.1.2.9 Radioactive Waste Treatment System 20
1.1.2.10 Features of ABWR General Arrangement 20
1.2 Application of PSA in Design and Maintenance of ABWR 22
1.2.1 Safety Features of Conventional BWRs 22
1.2.1.1 Conventional ECCS Design 22
1.2.1.2 Characteristics of the Conventional BWR Risk Profile 23
1.2.2 Philosophies of ABWR Safety Design 25
1.2.2.1 The Constant Risk Philosophy 26
1.2.2.2 The Positive Cost Reduction Philosophy 27
1.2.3 Concrete Measures to Enhance Safety 29
1.2.3.1 Approach to Enhance Safety 29
1.2.3.2 PSA Performance of the ABWR 32
1.2.3.3 ABWR Design Related to Safety Enhancement and/or Cost Reduction 33
1.2.3.3 Simplification of the Primary System 34
1.2.3.3 Primary Containment Vessel Innovations 35
1.2.3.3 Adoption of Fine Motion Control Rod Drive 36
1.2.3.3 ECCS Initiation Level Separation Between Transient and LOCA 36
1.2.3.3 Adoption of New Design Main Control Panel and Instrument and Control (IandC) Technologies 37
1.2.3.3 Accident Management of the ABWR 38
1.2.4 Application in Maintenance 39
1.3 Supplemental Notes on PSA 39
1.4 Notes on PSA at Conceptual Design Stage 39
1.5 Notes on Comparing PSA Results 40
References 43
Chapter 2: The Advanced Accumulator: A New Passive ECCS Component of the APWR 44
2.1 Overview of the APWR 44
2.1.1 Large Capacity Core 46
2.1.2 Neutron Reflector 46
2.1.3 Advanced Safety Systems 47
2.1.4 Advanced Main Control Board and Integrated Digital Control and Protection System 47
2.1.5 Main Components with Increased Capacity 48
2.2 Development of the Advanced Accumulator 49
2.3 ECCS of the APWR 50
2.4 Characteristics of the Advanced Accumulator 54
2.5 Development of the Advanced Accumulator 56
2.5.1 Structure of the Flow Damper 56
2.5.2 Design of the Flow Damper 57
2.5.3 Principle of the Advanced Accumulator 58
2.5.4 Theoretical Consideration of the Flow Damper 60
2.5.5 Transition of Water Level in the Standpipe 72
2.6 Confirmation Tests of the Advanced Accumulator 77
2.6.1 Purpose of Scale Testing 78
2.6.2 Principle of the Advanced Accumulator 79
2.6.3 Confirmation of Quick Changeover 81
2.6.4 Performance of the Flow Damper 82
2.6.5 Total Performance of the Advanced Accumulator 86
2.7 Structure of Flow in the Flow Damper 93
2.8 Conclusion 96
References 96
Chapter 3: Severe Accident Mitigation Features of APR1400 97
3.1 Introduction 97
3.2 Description of Nuclear Systems and Safety Systems 100
3.2.1 Safety Injection System 101
3.2.2 In-Containment Refueling Water Storage Tank 102
3.2.3 Auxiliary Feed Water System 102
3.2.4 Containment Spray System 102
3.3 Design Features against Severe Accident 103
3.3.1 Robust Containment 103
3.3.2 Safety Depressurization and Vent System 104
3.3.3 Evaluation of Hydrogen Mitigation System 105
3.3.4 Cavity Flooding System 105
3.3.5 Reactor Cavity Design 107
3.4 Severe Accident Management and External Reactor Vessel Cooling 108
3.5 Probabilistic Safety Assessment (PSA) for the APR1400 Design 110
3.6 Resolution of Severe Accident Issues 112
3.6.1 Severe Accident Progression 112
3.6.2 Identification of Severe Accident Issues 113
3.6.3 Hydrogen Control 114
3.6.3.1 AICC Pressure Calculation 115
3.6.3.2 Hydrogen Distribution 115
3.6.3.3 Hydrogen Mitigation System 117
3.6.4 Direct Containment Heating 118
3.6.5 Steam Explosion 119
3.6.5.1 Integrity of Reactor Cavity Structure 120
3.6.5.2 Steam Spike Analysis 120
3.6.6 Molten Core Concrete Interaction 121
3.6.7 Equipment Survivability 124
3.7 Conclusions 126
References 128
Chapter 4: Development and Design of the EPR Core Catcher 130
4.1 Introduction 130
4.2 Overview of the EPR Severe Accident Mitigation Features 131
4.2.1 High-Pressure Core Melting 131
4.2.2 Hydrogen Deflagration/Detonation 131
4.2.3 Energetic Steam Explosions 132
4.2.4 Basemat Penetration 132
4.3 Challenges for the Development of a Core Catcher System and Approach Followed for the EPR 133
4.3.1 Challenges for the Design of Measures for Melt Stabilization 133
4.3.2 Characteristic Features of the Approach Proposed to be Followed 133
4.3.3 Design Principles for the Core Melt Stabilization Measures 134
4.4 General Strategy for Core Melt Stabilization 135
4.5 Description of Components 136
4.5.1 Components used for Temporary Melt Retention in the Reactor Pit 136
4.5.1.1 Sacrificial Concrete 136
4.5.1.2 Protection Layer 138
4.5.1.3 Melt Plug 139
4.5.2 Components used for Melt Spreading 140
4.5.2.1 Melt Discharge Channel 140
4.5.2.2 Melt Plug Transport System 142
4.5.2.3 Core Catcher Assembly 142
4.5.2.4 Core Catcher Cooling Structure 147
4.5.2.5 Interface with the CHRS 148
4.5.3 Severe Accident IandC 151
4.6 Conclusions 152
References 152
Chapter 5: Nuclear Power Development and Severe Accident Research in China 154
5.1 Introduction 154
5.2 LWR Development in China 156
5.2.1 Selection of Technology Lines 157
5.2.2 Self-Reliant Technology 158
5.2.3 Nationwide Coordination 159
5.3 Severe Accident Research in China 160
5.3.1 IVR 161
5.3.2 Passive Containment Cooling 167
5.3.2.1 Low Reynolds k-epsi Turbulence Model 172
5.3.2.2 Advanced Models for Thermal Radiation 173
5.3.3 Hydrogen Safety 177
5.3.4 Steam Explosion 181
5.4 Summary 186
References 186
Chapter 6: Full MOX Core Design of the Ohma ABWR Nuclear Power Plant 188
6.1 Introduction 188
6.2 Outline of the Ohma NPP: A Full MOX Core ABWR 188
6.3 Design of the Full MOX Core ABWR 189
6.3.1 Design Principles 189
6.3.2 Plutonium Characteristics Needing Consideration in the Design: Large Neutron Absorption 190
6.3.3 Plutonium Characteristics Needing Consideration in the Design: Variation in the Amounts of Pu Isotopes 190
6.3.4 Plant Design Modifications for the Enhancement of the Ohma NPP 190
6.4 Core Design 191
6.4.1 Design Conditions of MOX Fuel 191
6.4.2 MOX Fuel and Core Basic Specifications 192
6.4.3 MOX Fuel Rod Specifications 192
6.4.4 MOX Fuel Lattice Design 192
6.4.5 Full MOX Fuel Loading Pattern 193
6.5 Core Characteristics 194
6.5.1 Void Coefficient and Dynamic Parameters 194
6.5.2 Control Rod Worth in MOX Core 194
6.5.3 Excess Reactivity 195
6.5.4 Shutdown Margin 196
6.5.5 Thermal Hydraulic Margins 197
6.5.6 Fuel Temperature and Internal Pressure 197
6.5.7 Initial Plutonium Composition 198
6.6 Core Dynamics and Safety Analyses 200
6.6.1 Stability Analysis 200
6.6.2 Abnormal Transients During Operation 201
6.6.3 Accident Analyses 202
6.7 Design Methods and Verifications 203
6.7.1 Core Design Methods 203
6.7.2 Verifications of Design Methods 205
6.7.3 Verification for MOX Lead Test Assembly in Tsuruga-1 205
6.7.4 Verification for Void and Absorber Worth in EOLE (EPICURE) 207
6.8 Summary 208
References 209
Chapter 7: CFD Analysis Applications in BWR Reactor System Design 210
7.1 CFD Analysis Application to BWR-5 Recirculation System 210
7.1.1 Introduction 211
7.1.2 Symbols 212
7.1.3 Analytical Study of Flow Stabilization 213
7.1.3.1 Flow Conditions at Cross Branch Pipe Region 213
7.1.3.2 Countermeasures for Flow Stabilization 213
7.1.3.3 Analytical Method 214
7.1.3.4 Analytical Results 215
7.1.4 Verification of Flow Stabilization by Tests 216
7.1.4.1 Test Rig and Method 216
7.1.4.2 Test Results 218
7.1.5 Conclusions 220
7.2 CFD Analysis Application in an ABWR 220
7.2.1 ABWR Lower Plenum CFD Analysis and Reactor Internals FIV Stress Evaluation 220
7.2.1.1 Application to 60 Sector 1/5-Scaled Test 220
7.2.1.2 Evaluation of FIV Performance of Lower Plenum Structure 221
7.2.2 Conclusion 224
7.3 CFD Analysis Application to Development of a Thicker RIP Nozzle for Seismic Performance Improvement 225
7.3.1 Introduction 225
7.3.2 Method of CFD Analysis 225
7.3.3 CFD Analysis Qualification with 1/5-Scale Tests 226
7.3.4 Evaluation of Influence in the Actual ABWR 227
7.3.5 Conclusion 232
7.4 CFD Analysis Application to the Next Generation Reactors: Some Concluding Remarks 232
References 232
Chapter 8: Next Generation Technologies in the Digital IandC Systems for Nuclear Power Plants 233
8.1 Overview of IandC Systems for NPPs 233
8.2 Chronicle of Digitalization 235
8.2.1 The First Generation 235
8.2.2 The Second Generation 236
8.2.3 The Third Generation System for the ABWR 237
8.3 Advantages and Evaluation of the Third Generation Digital System 239
8.3.1 Advantages of Digitalization 239
8.3.2 Issues and Solutions 241
8.4 Field Programmable Gate Array Technology 242
8.4.1 Overview as a Technical Solution 242
8.4.2 Design Concept of FPGA-Based Systems 243
8.4.3 Overview of FPGA-Based Systems 244
8.5 Development Process of FPGA-Based System 246
8.6 Logic Qualification of FPGA-Based Systems 249
8.6.1 Qualification Process 249
8.6.2 FPGA-Specific Issues 250
8.6.3 Hazards Specific to FPGA-based systems 251
8.6.3.1 Small Logic Errors 251
8.6.3.2 Timing Errors 252
8.6.3.3 Logic Synthesis Errors and Place and Route Errors 253
8.6.3.4 Logic Embedding Errors 253
8.6.4 Logic Qualification of the Systems 254
8.7 Next Generation IandC systems what the Future Holds
8.7.1 Design Scope and Concept 254
8.7.2 Elemental Technologies 256
8.7.2.1 Measurement and Monitoring Tools 256
8.7.2.2 Data Mining and Diagnostic Tools 258
8.7.3 Toward the Next Generation 259
References 260
Chapter 9: Advanced 3D-CAD and Its Application to State-of-the-Art Construction Technologies in ABWR Plant Projects 261
9.1 3D Integrated Engineering System 261
9.1.1 Introduction 261
9.1.2 Outline of the Plant Integrated CAE System 262
9.1.3 Plant Layout Design Using 3D-CAD 263
9.1.3.1 3D Design System and Database System 263
9.1.3.2 Review and Evaluation System 264
9.1.3.3 Application of the 3D-CAD Data to Production, Design, and Fabrication 264
9.1.3.4 Application to Construction Planning and Management 266
9.1.3.4 Construction Planning System 266
9.1.3.4 Construction Management System 267
9.2 Advanced Construction Technologies 267
9.2.1 Introduction 267
9.2.2 Applied Construction Technologies 269
9.2.2.1 Broader Application of Large Module/Block Construction Method 269
9.2.2.2 Open-Top and Parallel Construction Method 270
9.2.2.3 Application of Floor Packaging Construction Method 271
9.2.2.4 Full Application of Information Technology to Quality Plant Engineering and Construction 271
9.2.3 Development of Advanced Technologies 272
9.3 Conclusion 272
References 273
Chapter 10: Progress in Seismic Design and Evaluation of Nuclear Power Plants 274
10.1 Outline of Seismic Design of NPPs 275
10.1.1 Investigation of Earthquakes 276
10.1.1.1 Plate Tectonics Around Japan 276
10.1.1.2 Earthquake Source Patterns 276
10.1.1.3 Active Faults in the Japanese Archipelago 277
10.1.1.4 Surveys of Active Faults 278
10.1.2 Determination of Design Basis Earthquake Ground Motions 282
10.1.2.1 Estimation of Earthquake Ground Motion 282
10.1.2.2 Estimation of Design Basis Earthquake Ground Motions 285
10.1.3 Classification of Importance in Seismic Design 285
10.1.4 Seismic Design of Buildings and Equipment 287
10.1.5 Events Accompanying an Earthquake 288
10.1.6 Ensuring Safety in the Seismic Design 290
10.2 Assessment of Residual Risk 293
10.3 Recent Seismic Topics at Kashiwazaki-Kariwa NPP 296
References 297
Index 298