Oka / Koshizuka / Ishiwatari Super Light Water Reactors and Super Fast Reactors

Supercritical-Pressure Light Water Cooled Reactors
1. Auflage 2010
ISBN: 978-1-4419-6035-1
Verlag: Springer
Format: PDF
Kopierschutz: 1 - PDF Watermark

E-Book, Englisch, 651 Seiten

Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain the understanding of the conceptual design elements and specific analysis methods of supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference to engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology.

Autoren/Hrsg.

Oka, Yoshiaki

Koshizuka, Seiichi

Ishiwatari, Yuki

Yamaji, Akifumi

Inhaltsverzeichnis

1;Preface;8
2;Acknowledgements;10
3;Contents;12
4;Chapter 1: Introduction and Overview;20
4.1;1.1 Industrial Innovation;20
4.2;1.2 Evolution of Boilers;20
4.3;1.3 Overview of the Super LWR and Super FR;25
4.3.1;1.3.1 Concept and Features;25
4.3.2;1.3.2 Improvement of Thermal Design Criterion;29
4.3.3;1.3.3 Core Design Criteria;31
4.3.4;1.3.4 Improvement of Core Design and Analysis;32
4.3.5;1.3.5 Fuel Design;35
4.3.6;1.3.6 Plant Control;38
4.3.7;1.3.7 Startup Schemes;41
4.3.8;1.3.8 Stability;47
4.3.9;1.3.9 Safety;56
4.3.9.1;1.3.9.1 Safety Principle;56
4.3.9.2;1.3.9.2 Plant and Safety Systems;58
4.3.9.3;1.3.9.3 Safety Criteria;59
4.3.9.4;1.3.9.4 Safety Analysis at Supercritical Pressure;61
4.3.9.5;1.3.9.5 LOCA Analysis;66
4.3.9.6;1.3.9.6 Summary of Safety Analysis;69
4.3.9.7;1.3.9.7 Simplified Probabilistic Safety Assessment;69
4.3.10;1.3.10 Super FR;73
4.3.10.1;1.3.10.1 Fuel, Core and Plant System;73
4.3.10.2;1.3.10.2 Zirconium Hydride Layer Concept for Negative Void Reactivity;77
4.3.11;1.3.11 Computer Codes and Database;80
4.4;1.4 Past Concepts of High Temperature Water and Steam Cooled Reactors;81
4.5;1.5 Research and Development;82
4.5.1;1.5.1 Japan;82
4.5.2;1.5.2 Europe;87
4.5.3;1.5.3 GIF and SCWR;87
4.5.4;1.5.4 Korea, China, US, Russia and IAEA;87
4.6;References;88
5;Chapter 2: Core Design;98
5.1;2.1 Introduction;98
5.1.1;2.1.1 Supercritical Water Thermophysical Properties;99
5.1.2;2.1.2 Heat Transfer Deterioration in Supercritical Water;101
5.1.2.1;2.1.2.1 Background;101
5.1.2.2;2.1.2.2 Numerical Computations;103
5.1.2.3;2.1.2.3 Determination of Deteriorated Heat Flux;104
5.1.2.4;2.1.2.4 Heat Transfer Deterioration at High Flow Rates;106
5.1.2.5;2.1.2.5 Heat Transfer Deterioration at Low Flow Rates;107
5.1.3;2.1.3 Design Considerations with Heat Transfer Deterioration;109
5.2;2.2 Core Design Scope;111
5.2.1;2.2.1 Design Margins;111
5.2.2;2.2.2 Design Criteria;115
5.2.2.1;2.2.2.1 Neutronic Design Criteria;115
5.2.2.2;2.2.2.2 Thermal Design Criteria (Thermal Limit for Normal Operations);116
5.2.3;2.2.3 Design Boundary Conditions;117
5.2.3.1;2.2.3.1 Core Pressure, Inlet Temperature and Average Outlet Temperature;117
5.2.3.2;2.2.3.2 Determination of the Core Size;117
5.2.3.3;2.2.3.3 Fuel Discharge Burnup and Enrichment;118
5.2.4;2.2.4 Design Targets;119
5.2.4.1;2.2.4.1 Flat Coolant Outlet Temperature Distribution;119
5.2.4.2;2.2.4.2 Flat Core Power Distribution;120
5.2.4.3;2.2.4.3 Burnup Reactivity Compensation;121
5.3;2.3 Core Calculations;121
5.3.1;2.3.1 Neutronic Calculations;121
5.3.1.1;2.3.1.1 Calculation Codes and Data Libraries;121
5.3.1.2;2.3.1.2 Cell Burnup Calculations of Normal Fuel Rods;122
5.3.1.3;2.3.1.3 Cell Burnup Calculations of Fuel Rods with Gadolinia;124
5.3.1.4;2.3.1.4 Assembly Burnup Calculations;124
5.3.1.5;2.3.1.5 Core Burnup Calculations;126
5.3.1.6;2.3.1.6 Handling of Control Rods in ASMBURN and COREBN;127
5.3.1.7;2.3.1.7 Branching Burnup Calculation;127
5.3.1.8;2.3.1.8 Summary of the Neutronic Calculations;129
5.3.2;2.3.2 Thermal-Hydraulic Calculations;131
5.3.2.1;2.3.2.1 Radial Heat Conductions and Transfers;133
5.3.2.2;2.3.2.2 Heat Transfer Correlation for Supercritical Water Cooling;136
5.3.2.3;2.3.2.3 Axial Heat Transport;136
5.3.2.4;2.3.2.4 Outline of the Single Channel Thermal-Hydraulic Analysis;137
5.3.2.5;2.3.2.5 Applying the Single Channel Model to Core Thermal-Hydraulic Calculations;138
5.3.3;2.3.3 Equilibrium Core Calculations;139
5.3.3.1;2.3.3.1 Two- and Three-Dimensional Core Calculation Models;139
5.3.3.2;2.3.3.2 Coupling of Neutronic and Thermal-Hydraulic Calculations;140
5.3.3.3;2.3.3.3 Equilibrium Core Calculations;140
5.4;2.4 Core Designs;141
5.4.1;2.4.1 Fuel Rod Designs;141
5.4.1.1;2.4.1.1 Fuel Rod Heated Length;142
5.4.1.2;2.4.1.2 Fuel Rod Diameter;142
5.4.1.3;2.4.1.3 Fuel Rod Cladding Materials;143
5.4.1.4;2.4.1.4 Evaluating Method and Limits for Cladding Stress;144
5.4.1.5;2.4.1.5 Design Conditions;145
5.4.1.6;2.4.1.6 Stress Evaluations and Determination of the Cladding Thickness;145
5.4.1.7;2.4.1.7 Initial Gap Size;146
5.4.1.8;2.4.1.8 Initial Pellet Density;147
5.4.2;2.4.2 Fuel Assembly Designs;147
5.4.2.1;2.4.2.1 Requirements for the Fuel Assembly Design;147
5.4.2.2;2.4.2.2 Hexagonal Fuel Assembly;148
5.4.2.3;2.4.2.3 Square Fuel Assembly;150
5.4.2.4;2.4.2.4 Other Designs (Solid Moderator and Water Rods);155
5.4.3;2.4.3 Coolant Flow Scheme;156
5.4.4;2.4.4 Low Temperature Core Design with R-Z Two-Dimensional Core Calculations;159
5.4.4.1;2.4.4.1 Design Criteria;159
5.4.4.2;2.4.4.2 Fuel Design;159
5.4.4.3;2.4.4.3 Core Characteristics Evaluations with R-Z Two-Dimensional Core Calculations;160
5.4.5;2.4.5 High Temperature Core Design with Three-Dimensional Core Calculations;164
5.4.5.1;2.4.5.1 Core Size;164
5.4.5.2;2.4.5.2 Fuel Loading and Reloading Patterns;164
5.4.5.3;2.4.5.3 Coolant Flow Distributions;165
5.4.5.4;2.4.5.4 Control Rod Design and Control Rod Patterns;167
5.4.5.5;2.4.5.5 Radial Core Power Distributions and Radial Core Power Peaking Factor;169
5.4.5.6;2.4.5.6 Axial Core Power Distributions and Axial Core Power Peaking Factor;170
5.4.5.7;2.4.5.7 Local Power Distributions for a Homogenized Fuel Assembly;171
5.4.5.8;2.4.5.8 Total Power Peaking Factor and MLHGR;171
5.4.5.9;2.4.5.9 Coolant Outlet Temperature Distribution;173
5.4.5.10;2.4.5.10 Maximum Cladding Surface Temperature Distribution;173
5.4.5.11;2.4.5.11 Water Density Reactivity Coefficient;174
5.4.5.12;2.4.5.12 Doppler Reactivity Coefficient;176
5.4.5.13;2.4.5.13 Core Shutdown Margin;176
5.4.5.14;2.4.5.14 Scram Reactivity Curve;177
5.4.5.15;2.4.5.15 Alternative Shutdown System;178
5.4.5.16;2.4.5.16 Summary and Design Issues of the ``First Trial Design´´;179
5.4.6;2.4.6 Design Improvements;180
5.4.6.1;2.4.6.1 Coolant Flow Scheme: Outer Core Downward Flow Cooling;181
5.4.6.2;2.4.6.2 Power Distributions and MLHGR;184
5.4.6.3;2.4.6.3 Coolant Outlet Temperature Distribution;185
5.4.6.4;2.4.6.4 Improvements of the Neutron Economy;186
5.4.7;2.4.7 Summary;189
5.5;2.5 Subchannel Analysis;192
5.5.1;2.5.1 Subchannel Analysis Code;192
5.5.1.1;2.5.1.1 Governing Equations;192
5.5.1.2;2.5.1.2 Iterative Procedure;194
5.5.1.3;2.5.1.3 Heat Transfer Coefficient;195
5.5.2;2.5.2 Subchannel Analysis of the Super LWR;196
5.5.2.1;2.5.2.1 Computational Conditions;196
5.5.2.2;2.5.2.2 Subchannel Analysis;197
5.6;2.6 Statistical Thermal Design;200
5.6.1;2.6.1 Comparison of Thermal Design Methods;201
5.6.2;2.6.2 Description of MCSTDP;203
5.6.2.1;2.6.2.1 Design Criteria;203
5.6.2.2;2.6.2.2 Philosophy of the Design Procedure;204
5.6.2.3;2.6.2.3 Uncertainties Considered;205
5.6.2.4;2.6.2.4 Details of the Design Procedure;208
5.6.3;2.6.3 Application of MCSTDP;209
5.6.3.1;2.6.3.1 Statistical Characteristics of Uncertainties;209
5.6.3.2;2.6.3.2 Subfactor of Subchannel Area;212
5.6.3.3;2.6.3.3 Results and Discussion;214
5.6.4;2.6.4 Comparison with RTDP;217
5.6.4.1;2.6.4.1 Introduction of RTDP;217
5.6.4.2;2.6.4.2 Results and Comparison;218
5.6.5;2.6.5 Summary;219
5.7;2.7 Fuel Rod Behaviors During Normal Operations;219
5.7.1;2.7.1 Evaluation of the Maximum Peak Cladding Temperature;219
5.7.2;2.7.2 Fuel Rod Analysis;220
5.7.2.1;2.7.2.1 Calculation Code;220
5.7.2.2;2.7.2.2 Irradiation History of the Fuel Rod;221
5.7.2.3;2.7.2.3 Basic Fuel Rod Behaviors;222
5.7.3;2.7.3 Fuel Rod Design;224
5.7.3.1;2.7.3.1 Sensitivity Study;224
5.7.3.2;2.7.3.2 Mechanical Strength Requirement of Cladding;226
5.8;2.8 Development of Transient Criteria;227
5.8.1;2.8.1 Selection of Fuel Rods for Analyses;228
5.8.2;2.8.2 Principle of Rationalizing the Criteria for Abnormal Transients;229
5.8.2.1;2.8.2.1 Principle of Ensuring the Fuel Integrity at Abnormal Transients;229
5.8.2.2;2.8.2.2 Classification of Abnormal Transients and Modeling;231
5.8.2.3;2.8.2.3 Evaluations of Allowable Maximum Cladding Temperature and Fuel Rod Power;232
5.9;2.9 Summary;236
5.10;References;237
6;Chapter 3: Plant System Design;240
6.1;3.1 Introduction;240
6.2;3.2 System Components and Configuration;241
6.3;3.3 Main Components Characteristics;242
6.3.1;3.3.1 Containment;243
6.3.2;3.3.2 Reactor Pressure Vessel;245
6.3.3;3.3.3 Internals;246
6.3.4;3.3.4 Turbine;247
6.3.5;3.3.5 Steam Lines and Candidate Materials;249
6.4;3.4 Plant Heat Balance;249
6.4.1;3.4.1 Super LWR Steam Cycle Characteristics;249
6.4.2;3.4.2 Thermal Efficiency Evaluation;251
6.4.3;3.4.3 Factors Influencing Thermal Efficiency;254
6.4.3.1;3.4.3.1 Core Outlet Temperature;254
6.4.3.2;3.4.3.2 Core Inlet Temperature;255
6.5;3.5 Summary;257
6.6;References;258
7;Chapter 4: Plant Dynamics and Control;259
7.1;4.1 Introduction;259
7.2;4.2 Analysis Method for Plant Dynamics;259
7.3;4.3 Plant Dynamics Without a Control System;264
7.3.1;4.3.1 Withdrawal of a Control Rod Cluster;266
7.3.2;4.3.2 Decrease in Feedwater Flow Rate;266
7.3.3;4.3.3 Decrease in Turbine Control Valve Opening;268
7.4;4.4 Control System Design;270
7.4.1;4.4.1 Pressure Control System;271
7.4.2;4.4.2 Main Steam Temperature Control System;273
7.4.3;4.4.3 Reactor Power Control System;274
7.5;4.5 Plant Dynamics with Control System;276
7.5.1;4.5.1 Stepwise Increase in Pressure Setpoint;277
7.5.2;4.5.2 Stepwise Increase in Temperature Setpoint;279
7.5.3;4.5.3 Stepwise Decrease in Power Setpoint;280
7.5.4;4.5.4 Impulsive Decrease in Feedwater Flow Rate;280
7.5.5;4.5.5 Decrease in Feedwater Temperature;282
7.5.6;4.5.6 Discussion;283
7.6;4.6 Summary;284
7.7;References;284
8;Chapter 5: Plant Startup and Stability;286
8.1;5.1 Introduction;286
8.2;5.2 Design of Startup Systems;287
8.2.1;5.2.1 Introduction to Startup Schemes of FPPs;287
8.2.1.1;5.2.1.1 Constant Pressure Supercritical Boiler;288
8.2.1.2;5.2.1.2 Sliding Pressure Supercritical Boiler;289
8.2.2;5.2.2 Constant Pressure Startup System of the Super LWR;290
8.2.3;5.2.3 Sliding Pressure Startup System of the Super LWR;296
8.3;5.3 Thermal Considerations;299
8.3.1;5.3.1 Startup Thermal Analysis Code;299
8.3.1.1;5.3.1.1 Heat Transfer Correlations;301
8.3.1.2;5.3.1.2 Determination of Critical Heat Fluxes;304
8.3.2;5.3.2 Thermal Criteria for Plant Startup;305
8.3.3;5.3.3 Thermal Analyses;306
8.3.3.1;5.3.3.1 Power Increase Phase in Constant Pressure Startup or Sliding Pressure Startup;306
8.3.3.2;5.3.3.2 Pressurization Phase in Sliding Pressure Startup;307
8.3.3.3;5.3.3.3 Temperature Increasing Phase in Sliding Pressure Startup;311
8.3.3.4;5.3.3.4 Design of Startup Curves Based on Thermal Considerations;312
8.4;5.4 Thermal-Hydraulic Stability Considerations;312
8.4.1;5.4.1 Mechanism of Thermal-Hydraulic Instability;312
8.4.2;5.4.2 Selection of Analysis Method;314
8.4.3;5.4.3 Thermal-Hydraulic Stability Analysis Method;315
8.4.3.1;5.4.3.1 Mathematical Model;315
8.4.3.2;5.4.3.2 Steady-State Calculation;318
8.4.3.3;5.4.3.3 Frequency Domain Analysis;319
8.4.3.4;5.4.3.4 Decay Ratio;320
8.4.3.5;5.4.3.5 Stability Criterion;321
8.4.4;5.4.4 Thermal-Hydraulic Stability Analyses;321
8.4.4.1;5.4.4.1 Thermal-Hydraulic Stability at Full Power Normal Operation;323
8.4.4.2;5.4.4.2 Thermal-Hydraulic Stability at Partial Power Operations at 25MPa;326
8.4.4.3;5.4.4.3 Thermal-Hydraulic Stability at Pressurization Phase;327
8.4.4.4;5.4.4.4 Parametric Studies of Thermal-Hydraulic Stability;329
8.5;5.5 Coupled Neutronic Thermal-Hydraulic Stability Considerations;333
8.5.1;5.5.1 Mechanism of Coupled Neutronic Thermal-Hydraulic Instability;333
8.5.2;5.5.2 Coupled Neutronic Thermal-Hydraulic Stability Analysis Method;335
8.5.2.1;5.5.2.1 Neutron Kinetics Model;335
8.5.2.2;5.5.2.2 Fuel Rod Heat Transfer Model;337
8.5.2.3;5.5.2.3 Water Rod Heat Transfer Model;340
8.5.2.4;5.5.2.4 Excore Circulation Model;340
8.5.2.5;5.5.2.5 Stability Criteria;341
8.5.3;5.5.3 Coupled Neutronic Thermal-Hydraulic Stability Analyses;341
8.5.3.1;5.5.3.1 Coupled Neutronic Thermal-Hydraulic Stability at Full Power Normal Operation;344
8.5.3.2;5.5.3.2 Coupled Neutronic Thermal-Hydraulic Stability at Partial Power Operations at 25MPa;344
8.5.3.3;5.5.3.3 Coupled Neutronic Thermal-Hydraulic Stability at Pressurization Phase;347
8.5.3.4;5.5.3.4 Parametric Studies of Coupled Neutronic Thermal-Hydraulic Stability;348
8.6;5.6 Design of Startup Procedures with Both Thermal and Stability Considerations;352
8.7;5.7 Design and Analysis of Procedures for System Pressurization and Line Switching in Sliding Pressure Startup Scheme;355
8.7.1;5.7.1 Motivation and Purpose;355
8.7.2;5.7.2 Redesign of Sliding Pressure Startup System;356
8.7.3;5.7.3 Redesign of Sliding Pressure Startup Procedures;357
8.7.3.1;5.7.3.1 Procedures Before Nuclear Heating;359
8.7.3.2;5.7.3.2 Start of Nuclear Heating and Feedwater Warming;359
8.7.3.3;5.7.3.3 System Pressurization to the Operating Point;360
8.7.3.4;5.7.3.4 Switch to Once-Through Mode;360
8.7.4;5.7.4 System Transient Analysis;360
8.8;5.8 Summary;362
8.9;References;364
9;Chapter 6: Safety;365
9.1;6.1 Introduction;365
9.2;6.2 Safety Principle;365
9.3;6.3 Safety System Design;366
9.3.1;6.3.1 Equipment;366
9.3.1.1;6.3.1.1 Reactor Shutdown System;367
9.3.1.2;6.3.1.2 Coolant Supply System;368
9.3.1.3;6.3.1.3 Valves for Coolant Discharge and Isolation;369
9.3.2;6.3.2 Actuation Conditions of the Safety System;371
9.4;6.4 Selection and Classification of Abnormal Events;373
9.4.1;6.4.1 Reactor Coolant Flow Abnormality;374
9.4.2;6.4.2 Other Abnormalities;376
9.4.3;6.4.3 Event Selection for Safety Analysis;377
9.4.4;6.4.4 Uniqueness in the LOCA of the Super LWR;378
9.5;6.5 Safety Criteria;379
9.5.1;6.5.1 Criteria for Fuel Rod Integrity;380
9.5.2;6.5.2 Criteria for Pressure Boundary Integrity;381
9.5.3;6.5.3 Criteria for ATWS;381
9.6;6.6 Safety Analysis Methods;382
9.6.1;6.6.1 Safety Analysis Code for Supercritical Pressure Condition;382
9.6.2;6.6.2 Safety Analysis Code for Subcritical Pressure Condition;387
9.6.3;6.6.3 Blowdown Analysis Code;388
9.6.4;6.6.4 Reflooding Analysis Code;393
9.7;6.7 Safety Analyses;396
9.7.1;6.7.1 Abnormal Transient Analyses at Supercritical Pressure;398
9.7.1.1;6.7.1.1 Partial Loss of Reactor Coolant Flow;398
9.7.1.2;6.7.1.2 Loss of Offsite Power;399
9.7.1.3;6.7.1.3 Loss of Turbine Load;400
9.7.1.4;6.7.1.4 Isolation of Main Steam Line;402
9.7.1.5;6.7.1.5 Pressure Control System Failure;402
9.7.1.6;6.7.1.6 Loss of Feedwater Heating;402
9.7.1.7;6.7.1.7 Inadvertent Startup of AFS;403
9.7.1.8;6.7.1.8 Reactor Coolant Flow Control System Failure;404
9.7.1.9;6.7.1.9 Uncontrolled CR Withdrawals;404
9.7.1.10;6.7.1.10 Summary;406
9.7.2;6.7.2 Accident Analyses at Supercritical Pressure;407
9.7.2.1;6.7.2.1 Total Loss of Reactor Coolant Flow;407
9.7.2.2;6.7.2.2 Reactor Coolant Pump Seizure;409
9.7.2.3;6.7.2.3 CR Ejections;409
9.7.2.4;6.7.2.4 Summary;411
9.7.3;6.7.3 Loss of Coolant Accident Analyses;411
9.7.3.1;6.7.3.1 Large LOCA;411
9.7.3.2;6.7.3.2 Small LOCA;416
9.7.3.3;6.7.3.3 Summary;416
9.7.4;6.7.4 ATWS Analysis;417
9.7.4.1;6.7.4.1 ATWS Analysis with Alternative Action;418
9.7.4.2;6.7.4.2 ATWS Analysis Without Alternative Action;420
9.7.4.3;6.7.4.3 Sensitivity Analyses in ATWS Events;423
9.7.4.4;6.7.4.4 Summary;426
9.7.5;6.7.5 Abnormal Transient and Accident Analyses at Subcritical Pressure;428
9.8;6.8 Development of a Transient Subchannel Analysis Code and Application to Flow Decreasing Events;431
9.8.1;6.8.1 A Transient Subchannel Analysis Code;431
9.8.2;6.8.2 Analyses of Flow Decreasing Events;433
9.8.2.1;6.8.2.1 Partial Loss of Reactor Coolant Flow;434
9.8.2.2;6.8.2.2 Total Loss of Reactor Coolant Flow;436
9.8.3;6.8.3 Summary;439
9.9;6.9 Simplified Level-1 Probabilistic Safety Assessment;439
9.9.1;6.9.1 Preparation of Event Trees;439
9.9.2;6.9.2 Initiating Event Frequency and Mitigation System Unavailability;447
9.9.3;6.9.3 Results and Considerations;448
9.9.4;6.9.4 Summary;451
9.10;6.10 Summary;452
9.11;References;453
10;Chapter 7: Fast Reactor Design;456
10.1;7.1 Introduction;456
10.2;7.2 Design Goals, Criteria, and Overall Procedure;456
10.2.1;7.2.1 Design Goals and Criteria;456
10.2.2;7.2.2 Overall Design Procedure;458
10.3;7.3 Concept of Blanket Assembly with Zirconium Hydride Layer;460
10.3.1;7.3.1 Effect of Zirconium Hydride Layer on Void Reactivity;460
10.3.2;7.3.2 Effect of Zirconium Hydride Layer on Breeding Capability;465
10.3.3;7.3.3 Effect of Hydrogen Loss from Zirconium Hydride Layers on Void Reactivity;466
10.4;7.4 Fuel Rod Design;468
10.4.1;7.4.1 Introduction;468
10.4.2;7.4.2 Failure Modes of Fuel Cladding;469
10.4.2.1;7.4.2.1 Melting of Fuel Pellets;469
10.4.2.2;7.4.2.2 Overheating of Cladding;470
10.4.2.3;7.4.2.3 Cladding Collapse;470
10.4.2.4;7.4.2.4 Rod Overpressure;470
10.4.2.5;7.4.2.5 Pellet Cladding Interaction;471
10.4.2.6;7.4.2.6 Other Failure Modes;471
10.4.3;7.4.3 Fuel Rod Design Criteria;471
10.4.3.1;7.4.3.1 Thermal Design Criteria;471
10.4.3.2;7.4.3.2 Hydrodynamic Design Criterion;472
10.4.3.3;7.4.3.3 Thermo-Mechanical Design Criteria;473
10.4.4;7.4.4 Fuel Rod Design Method;474
10.4.5;7.4.5 Fuel Rod Design and Analysis;477
10.4.5.1;7.4.5.1 Admissible Design Area;478
10.4.5.2;7.4.5.2 Nuclear Performance of Candidate Fuel Rod Designs;479
10.4.6;7.4.6 Summary of Fuel Rod Design;480
10.5;7.5 Core Design Method and 1,000MWe Class Core Design;482
10.5.1;7.5.1 Discussion of Neutronic Calculation Methods;482
10.5.2;7.5.2 Core Design Method;483
10.5.2.1;7.5.2.1 Nuclear Design Method;485
10.5.2.2;7.5.2.2 Thermal-Hydraulic Design Method;491
10.5.2.3;7.5.2.3 Neutronic Thermal-Hydraulic Coupled Equilibrium Core Calculation Method;492
10.5.3;7.5.3 Materials Used in Core Design;494
10.5.4;7.5.4 Fuel Assembly Design;495
10.5.5;7.5.5 Core Arrangement;496
10.5.5.1;7.5.5.1 In-Vessel Flow Path;496
10.5.5.2;7.5.5.2 Fuel Loading Pattern for Negative Void Reactivity;497
10.5.6;7.5.6 Design of 1,000MWe Class Core;498
10.6;7.6 Subchannel Analysis;506
10.6.1;7.6.1 Introduction;506
10.6.2;7.6.2 Temperature Difference Arising from Subchannel Heterogeneity;508
10.6.2.1;7.6.2.1 Introduction of Fuel Assembly and Subchannel Parameters;508
10.6.2.2;7.6.2.2 Rise of Maximum Cladding Surface Temperature by Subchannel Heterogeneity;508
10.6.3;7.6.3 Evaluation of MCST over Equilibrium Cycle;510
10.7;7.7 Evaluation of Maximum Cladding Surface Temperature with Engineering Uncertainties;514
10.7.1;7.7.1 Treatment of Downward Flow;514
10.7.2;7.7.2 Nominal Conditions and Uncertainties;516
10.7.3;7.7.3 Statistical Thermal Design of the Super FR;520
10.7.4;7.7.4 Comprehensive Evaluation of Maximum Cladding Surface Temperature at Normal Operation;521
10.8;7.8 Design and Improvements of 700MWe Class Core;523
10.8.1;7.8.1 Design of Reference Fuel Rod and Core;524
10.8.2;7.8.2 Core Design Improvement for Negative Local Void Reactivity;524
10.8.2.1;7.8.2.1 Principles for Reducing Local Void Reactivity;525
10.8.2.2;7.8.2.2 Sensitivity Analyses for Negative Local Void Reactivity;529
10.8.2.3;7.8.2.3 Example of Improved Core with Negative Local Void Reactivity;532
10.8.3;7.8.3 Core Design Improvement for Higher Power Density;533
10.8.3.1;7.8.3.1 Principle of Improving Power Density;533
10.8.3.2;7.8.3.2 Sensitivity Analyses for Higher Power Density;534
10.8.3.3;7.8.3.3 Example of Improved Core with Higher Power Density;536
10.9;7.9 Plant Control;537
10.9.1;7.9.1 Plant Transient Analysis Code for the Super FR;538
10.9.2;7.9.2 Basic Plant Dynamics of the Super FR;538
10.9.3;7.9.3 Design of Reference Control System;540
10.9.4;7.9.4 Improvement of Feedwater Controller;542
10.9.4.1;7.9.4.1 Feedback from Power to Flow Rate Ratio;542
10.9.4.2;7.9.4.2 Feedback from Reactor Power;543
10.9.4.3;7.9.4.3 Feedback from Derivative of Power;545
10.9.5;7.9.5 Plant Stability Analyses;545
10.9.5.1;7.9.5.1 10% Decrease in Power Setpoint;546
10.9.5.2;7.9.5.2 1% Increase in Pressure Setpoint;546
10.9.5.3;7.9.5.3 4C increase in Steam Temperature Setpoint;547
10.9.5.4;7.9.5.4 Impulsive Decrease in Feedwater Flow Rate by 5%;547
10.9.5.5;7.9.5.5 10C Decrease in Feedwater Temperature;548
10.9.6;7.9.6 Comparison of Improved Feedwater Controllers;549
10.9.7;7.9.7 Summary of Improvement of Feedwater Controller;550
10.10;7.10 Thermal and Stability Considerations During Power Raising Phase of Plant Startup;551
10.10.1;7.10.1 Introduction;551
10.10.2;7.10.2 Calculation of Flow Distribution;552
10.10.3;7.10.3 Thermal and Thermal-Hydraulic Stability Considerations;554
10.10.4;7.10.4 Sensitivity Analyses;562
10.11;7.11 Safety;565
10.11.1;7.11.1 Introduction;565
10.11.2;7.11.2 Analyses of Abnormal Transients and Accidents at Supercritical Pressure;566
10.11.3;7.11.3 Analyses of Loss of Coolant Accidents;571
10.11.4;7.11.4 Analyses of Anticipated Transient Without Scram Events;578
10.12;7.12 Summary;579
10.13;References;582
11;Chapter 8: Research and Development;585
11.1;8.1 Japan;585
11.1.1;8.1.1 Concept Development;585
11.1.2;8.1.2 Thermal Hydraulics;589
11.1.3;8.1.3 Materials and Water Chemistry;591
11.2;8.2 Other Countries;595
11.2.1;8.2.1 Europe;595
11.2.1.1;8.2.1.1 HPLWR-I Project;595
11.2.1.2;8.2.1.2 HPLWR-II Project;596
11.2.2;8.2.2 Canada;597
11.2.3;8.2.3 Korea;598
11.2.4;8.2.4 China;598
11.2.5;8.2.5 USA;599
11.3;8.3 International Activities;601
11.3.1;8.3.1 Generation-IV International Forum;601
11.3.2;8.3.2 IAEA-Coordinated Research Program;601
11.3.3;8.3.3 International Symposiums;602
11.4;References;604
12;Appendix A: Supercritical Fossil Fired Power Plants - Design and Developments;612
12.1;Introduction;612
12.2;Improvement of Steam Conditions;612
12.3;Boiler Design Features;614
12.3.1;Natural Circulation Boilers;614
12.3.2;Once-Through Boilers (UP: Universal Pressure Boiler for Constant Pressure Operation);614
12.3.3;Once-Through Boilers (Benson Boilers for Sliding Pressure Operation);617
12.3.4;Sliding Pressure Operation;617
12.3.5;Typical Arrangement of a Benson Boiler;618
12.4;Water Chemistry Guidelines;619
12.4.1;Characteristics of Water Chemistry in Boilers;619
12.4.2;Application of Low pH Coordinated Phosphate Treatment for Natural Circulation Boilers;620
12.4.2.1;Water Treatment Methods in Actual Circumstances;620
12.4.2.2;Chemical Analysis Results of the Scale;620
12.4.2.3;Zinc Compounds in Ammonia Water;622
12.4.2.4;Reaction of Zinc Compounds in Sodium Phosphate Solution;622
12.4.2.5;Research Conclusions;623
12.4.3;Doing CWT on Once-Through Type Boilers;623
12.4.3.1;Observation of Pressure Drop in Boiler;623
12.5;Pressure Parts Materials;623
12.6;Materials for Conventional Super Critical Boilers;623
12.6.1;Materials for the Advanced Super Critical Boiler;626
12.7;Summary;631
12.8;References;631
13;Appendix B: Review of High Temperature Water and Steam Cooled Reactor Concepts;632
13.1;Introduction;632
13.2;Supercritical Pressure Reactors;632
13.2.1;Water Moderated, Supercritical Steam Cooled Reactor (WH, 1957);633
13.2.2;Heavy Water Moderated, Light Water Cooled, Once-Through Pressure-Tube Type Reactor (GE Hanford, 1959);636
13.2.3;SCOTT-R, Once-Through, Graphite Moderated, Light Water Cooled Tube Reactor (WH, 1962);639
13.2.4;SC-PWR: Indirect-Cycle, Supercritical-Pressure PWR (WH);640
13.2.5;SCLWR and SCFR: Light Water Cooled (Moderated) Once-Through Reactor with RPV (the University of Tokyo, 1992);641
13.2.6;B500SKDI, Natural Circulation Integrated SCPWR (Kurchatov, Institute 1992);645
13.2.7;CANDU-X, Supercritical-Pressure CANDU (AECL, 1998);648
13.3;Nuclear Superheaters (GE, 1950s-1960s);649
13.4;Steam Cooled Fast Breeder Reactors;651
13.5;Summary;655
13.6;References;631
14;Index;658