Super Light Water Reactors and Super Fast Reactors (eBook)

Preface 8
Acknowledgements 10
Contents 12
Chapter 1: Introduction and Overview 20
1.1 Industrial Innovation 20
1.2 Evolution of Boilers 20
1.3 Overview of the Super LWR and Super FR 25
1.3.1 Concept and Features 25
1.3.2 Improvement of Thermal Design Criterion 29
1.3.3 Core Design Criteria 31
1.3.4 Improvement of Core Design and Analysis 32
1.3.5 Fuel Design 35
1.3.6 Plant Control 38
1.3.7 Startup Schemes 41
1.3.8 Stability 47
1.3.9 Safety 56
1.3.9.1 Safety Principle 56
1.3.9.2 Plant and Safety Systems 58
1.3.9.3 Safety Criteria 59
1.3.9.4 Safety Analysis at Supercritical Pressure 61
1.3.9.5 LOCA Analysis 66
1.3.9.6 Summary of Safety Analysis 69
1.3.9.7 Simplified Probabilistic Safety Assessment 69
1.3.10 Super FR 73
1.3.10.1 Fuel, Core and Plant System 73
1.3.10.2 Zirconium Hydride Layer Concept for Negative Void Reactivity 77
1.3.11 Computer Codes and Database 80
1.4 Past Concepts of High Temperature Water and Steam Cooled Reactors 81
1.5 Research and Development 82
1.5.1 Japan 82
1.5.2 Europe 87
1.5.3 GIF and SCWR 87
1.5.4 Korea, China, US, Russia and IAEA 87
References 88
Chapter 2: Core Design 98
2.1 Introduction 98
2.1.1 Supercritical Water Thermophysical Properties 99
2.1.2 Heat Transfer Deterioration in Supercritical Water 101
2.1.2.1 Background 101
2.1.2.2 Numerical Computations 103
2.1.2.3 Determination of Deteriorated Heat Flux 104
2.1.2.4 Heat Transfer Deterioration at High Flow Rates 106
2.1.2.5 Heat Transfer Deterioration at Low Flow Rates 107
2.1.3 Design Considerations with Heat Transfer Deterioration 109
2.2 Core Design Scope 111
2.2.1 Design Margins 111
2.2.2 Design Criteria 115
2.2.2.1 Neutronic Design Criteria 115
2.2.2.2 Thermal Design Criteria (Thermal Limit for Normal Operations) 116
2.2.3 Design Boundary Conditions 117
2.2.3.1 Core Pressure, Inlet Temperature and Average Outlet Temperature 117
2.2.3.2 Determination of the Core Size 117
2.2.3.3 Fuel Discharge Burnup and Enrichment 118
2.2.4 Design Targets 119
2.2.4.1 Flat Coolant Outlet Temperature Distribution 119
2.2.4.2 Flat Core Power Distribution 120
2.2.4.3 Burnup Reactivity Compensation 121
2.3 Core Calculations 121
2.3.1 Neutronic Calculations 121
2.3.1.1 Calculation Codes and Data Libraries 121
2.3.1.2 Cell Burnup Calculations of Normal Fuel Rods 122
2.3.1.3 Cell Burnup Calculations of Fuel Rods with Gadolinia 124
2.3.1.4 Assembly Burnup Calculations 124
2.3.1.5 Core Burnup Calculations 126
2.3.1.6 Handling of Control Rods in ASMBURN and COREBN 127
2.3.1.7 Branching Burnup Calculation 127
2.3.1.8 Summary of the Neutronic Calculations 129
2.3.2 Thermal-Hydraulic Calculations 131
2.3.2.1 Radial Heat Conductions and Transfers 133
2.3.2.2 Heat Transfer Correlation for Supercritical Water Cooling 136
2.3.2.3 Axial Heat Transport 136
2.3.2.4 Outline of the Single Channel Thermal-Hydraulic Analysis 137
2.3.2.5 Applying the Single Channel Model to Core Thermal-Hydraulic Calculations 138
2.3.3 Equilibrium Core Calculations 139
2.3.3.1 Two- and Three-Dimensional Core Calculation Models 139
2.3.3.2 Coupling of Neutronic and Thermal-Hydraulic Calculations 140
2.3.3.3 Equilibrium Core Calculations 140
2.4 Core Designs 141
2.4.1 Fuel Rod Designs 141
2.4.1.1 Fuel Rod Heated Length 142
2.4.1.2 Fuel Rod Diameter 142
2.4.1.3 Fuel Rod Cladding Materials 143
2.4.1.4 Evaluating Method and Limits for Cladding Stress 144
2.4.1.5 Design Conditions 145
2.4.1.6 Stress Evaluations and Determination of the Cladding Thickness 145
2.4.1.7 Initial Gap Size 146
2.4.1.8 Initial Pellet Density 147
2.4.2 Fuel Assembly Designs 147
2.4.2.1 Requirements for the Fuel Assembly Design 147
2.4.2.2 Hexagonal Fuel Assembly 148
2.4.2.3 Square Fuel Assembly 150
2.4.2.4 Other Designs (Solid Moderator and Water Rods) 155
2.4.3 Coolant Flow Scheme 156
2.4.4 Low Temperature Core Design with R-Z Two-Dimensional Core Calculations 159
2.4.4.1 Design Criteria 159
2.4.4.2 Fuel Design 159
2.4.4.3 Core Characteristics Evaluations with R-Z Two-Dimensional Core Calculations 160
2.4.5 High Temperature Core Design with Three-Dimensional Core Calculations 164
2.4.5.1 Core Size 164
2.4.5.2 Fuel Loading and Reloading Patterns 164
2.4.5.3 Coolant Flow Distributions 165
2.4.5.4 Control Rod Design and Control Rod Patterns 167
2.4.5.5 Radial Core Power Distributions and Radial Core Power Peaking Factor 169
2.4.5.6 Axial Core Power Distributions and Axial Core Power Peaking Factor 170
2.4.5.7 Local Power Distributions for a Homogenized Fuel Assembly 171
2.4.5.8 Total Power Peaking Factor and MLHGR 171
2.4.5.9 Coolant Outlet Temperature Distribution 173
2.4.5.10 Maximum Cladding Surface Temperature Distribution 173
2.4.5.11 Water Density Reactivity Coefficient 174
2.4.5.12 Doppler Reactivity Coefficient 176
2.4.5.13 Core Shutdown Margin 176
2.4.5.14 Scram Reactivity Curve 177
2.4.5.15 Alternative Shutdown System 178
2.4.5.16 Summary and Design Issues of the ``First Trial Design´´ 179
2.4.6 Design Improvements 180
2.4.6.1 Coolant Flow Scheme: Outer Core Downward Flow Cooling 181
2.4.6.2 Power Distributions and MLHGR 184
2.4.6.3 Coolant Outlet Temperature Distribution 185
2.4.6.4 Improvements of the Neutron Economy 186
2.4.7 Summary 189
2.5 Subchannel Analysis 192
2.5.1 Subchannel Analysis Code 192
2.5.1.1 Governing Equations 192
2.5.1.2 Iterative Procedure 194
2.5.1.3 Heat Transfer Coefficient 195
2.5.2 Subchannel Analysis of the Super LWR 196
2.5.2.1 Computational Conditions 196
2.5.2.2 Subchannel Analysis 197
2.6 Statistical Thermal Design 200
2.6.1 Comparison of Thermal Design Methods 201
2.6.2 Description of MCSTDP 203
2.6.2.1 Design Criteria 203
2.6.2.2 Philosophy of the Design Procedure 204
2.6.2.3 Uncertainties Considered 205
2.6.2.4 Details of the Design Procedure 208
2.6.3 Application of MCSTDP 209
2.6.3.1 Statistical Characteristics of Uncertainties 209
2.6.3.2 Subfactor of Subchannel Area 212
2.6.3.3 Results and Discussion 214
2.6.4 Comparison with RTDP 217
2.6.4.1 Introduction of RTDP 217
2.6.4.2 Results and Comparison 218
2.6.5 Summary 219
2.7 Fuel Rod Behaviors During Normal Operations 219
2.7.1 Evaluation of the Maximum Peak Cladding Temperature 219
2.7.2 Fuel Rod Analysis 220
2.7.2.1 Calculation Code 220
2.7.2.2 Irradiation History of the Fuel Rod 221
2.7.2.3 Basic Fuel Rod Behaviors 222
2.7.3 Fuel Rod Design 224
2.7.3.1 Sensitivity Study 224
2.7.3.2 Mechanical Strength Requirement of Cladding 226
2.8 Development of Transient Criteria 227
2.8.1 Selection of Fuel Rods for Analyses 228
2.8.2 Principle of Rationalizing the Criteria for Abnormal Transients 229
2.8.2.1 Principle of Ensuring the Fuel Integrity at Abnormal Transients 229
2.8.2.2 Classification of Abnormal Transients and Modeling 231
2.8.2.3 Evaluations of Allowable Maximum Cladding Temperature and Fuel Rod Power 232
2.9 Summary 236
References 237
Chapter 3: Plant System Design 240
3.1 Introduction 240
3.2 System Components and Configuration 241
3.3 Main Components Characteristics 242
3.3.1 Containment 243
3.3.2 Reactor Pressure Vessel 245
3.3.3 Internals 246
3.3.4 Turbine 247
3.3.5 Steam Lines and Candidate Materials 249
3.4 Plant Heat Balance 249
3.4.1 Super LWR Steam Cycle Characteristics 249
3.4.2 Thermal Efficiency Evaluation 251
3.4.3 Factors Influencing Thermal Efficiency 254
3.4.3.1 Core Outlet Temperature 254
3.4.3.2 Core Inlet Temperature 255
3.5 Summary 257
References 258
Chapter 4: Plant Dynamics and Control 259
4.1 Introduction 259
4.2 Analysis Method for Plant Dynamics 259
4.3 Plant Dynamics Without a Control System 264
4.3.1 Withdrawal of a Control Rod Cluster 266
4.3.2 Decrease in Feedwater Flow Rate 266
4.3.3 Decrease in Turbine Control Valve Opening 268
4.4 Control System Design 270
4.4.1 Pressure Control System 271
4.4.2 Main Steam Temperature Control System 273
4.4.3 Reactor Power Control System 274
4.5 Plant Dynamics with Control System 276
4.5.1 Stepwise Increase in Pressure Setpoint 277
4.5.2 Stepwise Increase in Temperature Setpoint 279
4.5.3 Stepwise Decrease in Power Setpoint 280
4.5.4 Impulsive Decrease in Feedwater Flow Rate 280
4.5.5 Decrease in Feedwater Temperature 282
4.5.6 Discussion 283
4.6 Summary 284
References 284
Chapter 5: Plant Startup and Stability 286
5.1 Introduction 286
5.2 Design of Startup Systems 287
5.2.1 Introduction to Startup Schemes of FPPs 287
5.2.1.1 Constant Pressure Supercritical Boiler 288
5.2.1.2 Sliding Pressure Supercritical Boiler 289
5.2.2 Constant Pressure Startup System of the Super LWR 290
5.2.3 Sliding Pressure Startup System of the Super LWR 296
5.3 Thermal Considerations 299
5.3.1 Startup Thermal Analysis Code 299
5.3.1.1 Heat Transfer Correlations 301
5.3.1.2 Determination of Critical Heat Fluxes 304
5.3.2 Thermal Criteria for Plant Startup 305
5.3.3 Thermal Analyses 306
5.3.3.1 Power Increase Phase in Constant Pressure Startup or Sliding Pressure Startup 306
5.3.3.2 Pressurization Phase in Sliding Pressure Startup 307
5.3.3.3 Temperature Increasing Phase in Sliding Pressure Startup 311
5.3.3.4 Design of Startup Curves Based on Thermal Considerations 312
5.4 Thermal-Hydraulic Stability Considerations 312
5.4.1 Mechanism of Thermal-Hydraulic Instability 312
5.4.2 Selection of Analysis Method 314
5.4.3 Thermal-Hydraulic Stability Analysis Method 315
5.4.3.1 Mathematical Model 315
5.4.3.2 Steady-State Calculation 318
5.4.3.3 Frequency Domain Analysis 319
5.4.3.4 Decay Ratio 320
5.4.3.5 Stability Criterion 321
5.4.4 Thermal-Hydraulic Stability Analyses 321
5.4.4.1 Thermal-Hydraulic Stability at Full Power Normal Operation 323
5.4.4.2 Thermal-Hydraulic Stability at Partial Power Operations at 25MPa 326
5.4.4.3 Thermal-Hydraulic Stability at Pressurization Phase 327
5.4.4.4 Parametric Studies of Thermal-Hydraulic Stability 329
5.5 Coupled Neutronic Thermal-Hydraulic Stability Considerations 333
5.5.1 Mechanism of Coupled Neutronic Thermal-Hydraulic Instability 333
5.5.2 Coupled Neutronic Thermal-Hydraulic Stability Analysis Method 335
5.5.2.1 Neutron Kinetics Model 335
5.5.2.2 Fuel Rod Heat Transfer Model 337
5.5.2.3 Water Rod Heat Transfer Model 340
5.5.2.4 Excore Circulation Model 340
5.5.2.5 Stability Criteria 341
5.5.3 Coupled Neutronic Thermal-Hydraulic Stability Analyses 341
5.5.3.1 Coupled Neutronic Thermal-Hydraulic Stability at Full Power Normal Operation 344
5.5.3.2 Coupled Neutronic Thermal-Hydraulic Stability at Partial Power Operations at 25MPa 344
5.5.3.3 Coupled Neutronic Thermal-Hydraulic Stability at Pressurization Phase 347
5.5.3.4 Parametric Studies of Coupled Neutronic Thermal-Hydraulic Stability 348
5.6 Design of Startup Procedures with Both Thermal and Stability Considerations 352
5.7 Design and Analysis of Procedures for System Pressurization and Line Switching in Sliding Pressure Startup Scheme 355
5.7.1 Motivation and Purpose 355
5.7.2 Redesign of Sliding Pressure Startup System 356
5.7.3 Redesign of Sliding Pressure Startup Procedures 357
5.7.3.1 Procedures Before Nuclear Heating 359
5.7.3.2 Start of Nuclear Heating and Feedwater Warming 359
5.7.3.3 System Pressurization to the Operating Point 360
5.7.3.4 Switch to Once-Through Mode 360
5.7.4 System Transient Analysis 360
5.8 Summary 362
References 364
Chapter 6: Safety 365
6.1 Introduction 365
6.2 Safety Principle 365
6.3 Safety System Design 366
6.3.1 Equipment 366
6.3.1.1 Reactor Shutdown System 367
6.3.1.2 Coolant Supply System 368
6.3.1.3 Valves for Coolant Discharge and Isolation 369
6.3.2 Actuation Conditions of the Safety System 371
6.4 Selection and Classification of Abnormal Events 373
6.4.1 Reactor Coolant Flow Abnormality 374
6.4.2 Other Abnormalities 376
6.4.3 Event Selection for Safety Analysis 377
6.4.4 Uniqueness in the LOCA of the Super LWR 378
6.5 Safety Criteria 379
6.5.1 Criteria for Fuel Rod Integrity 380
6.5.2 Criteria for Pressure Boundary Integrity 381
6.5.3 Criteria for ATWS 381
6.6 Safety Analysis Methods 382
6.6.1 Safety Analysis Code for Supercritical Pressure Condition 382
6.6.2 Safety Analysis Code for Subcritical Pressure Condition 387
6.6.3 Blowdown Analysis Code 388
6.6.4 Reflooding Analysis Code 393
6.7 Safety Analyses 396
6.7.1 Abnormal Transient Analyses at Supercritical Pressure 398
6.7.1.1 Partial Loss of Reactor Coolant Flow 398
6.7.1.2 Loss of Offsite Power 399
6.7.1.3 Loss of Turbine Load 400
6.7.1.4 Isolation of Main Steam Line 402
6.7.1.5 Pressure Control System Failure 402
6.7.1.6 Loss of Feedwater Heating 402
6.7.1.7 Inadvertent Startup of AFS 403
6.7.1.8 Reactor Coolant Flow Control System Failure 404
6.7.1.9 Uncontrolled CR Withdrawals 404
6.7.1.10 Summary 406
6.7.2 Accident Analyses at Supercritical Pressure 407
6.7.2.1 Total Loss of Reactor Coolant Flow 407
6.7.2.2 Reactor Coolant Pump Seizure 409
6.7.2.3 CR Ejections 409
6.7.2.4 Summary 411
6.7.3 Loss of Coolant Accident Analyses 411
6.7.3.1 Large LOCA 411
6.7.3.2 Small LOCA 416
6.7.3.3 Summary 416
6.7.4 ATWS Analysis 417
6.7.4.1 ATWS Analysis with Alternative Action 418
6.7.4.2 ATWS Analysis Without Alternative Action 420
6.7.4.3 Sensitivity Analyses in ATWS Events 423
6.7.4.4 Summary 426
6.7.5 Abnormal Transient and Accident Analyses at Subcritical Pressure 428
6.8 Development of a Transient Subchannel Analysis Code and Application to Flow Decreasing Events 431
6.8.1 A Transient Subchannel Analysis Code 431
6.8.2 Analyses of Flow Decreasing Events 433
6.8.2.1 Partial Loss of Reactor Coolant Flow 434
6.8.2.2 Total Loss of Reactor Coolant Flow 436
6.8.3 Summary 439
6.9 Simplified Level-1 Probabilistic Safety Assessment 439
6.9.1 Preparation of Event Trees 439
6.9.2 Initiating Event Frequency and Mitigation System Unavailability 447
6.9.3 Results and Considerations 448
6.9.4 Summary 451
6.10 Summary 452
References 453
Chapter 7: Fast Reactor Design 456
7.1 Introduction 456
7.2 Design Goals, Criteria, and Overall Procedure 456
7.2.1 Design Goals and Criteria 456
7.2.2 Overall Design Procedure 458
7.3 Concept of Blanket Assembly with Zirconium Hydride Layer 460
7.3.1 Effect of Zirconium Hydride Layer on Void Reactivity 460
7.3.2 Effect of Zirconium Hydride Layer on Breeding Capability 465
7.3.3 Effect of Hydrogen Loss from Zirconium Hydride Layers on Void Reactivity 466
7.4 Fuel Rod Design 468
7.4.1 Introduction 468
7.4.2 Failure Modes of Fuel Cladding 469
7.4.2.1 Melting of Fuel Pellets 469
7.4.2.2 Overheating of Cladding 470
7.4.2.3 Cladding Collapse 470
7.4.2.4 Rod Overpressure 470
7.4.2.5 Pellet Cladding Interaction 471
7.4.2.6 Other Failure Modes 471
7.4.3 Fuel Rod Design Criteria 471
7.4.3.1 Thermal Design Criteria 471
7.4.3.2 Hydrodynamic Design Criterion 472
7.4.3.3 Thermo-Mechanical Design Criteria 473
7.4.4 Fuel Rod Design Method 474
7.4.5 Fuel Rod Design and Analysis 477
7.4.5.1 Admissible Design Area 478
7.4.5.2 Nuclear Performance of Candidate Fuel Rod Designs 479
7.4.6 Summary of Fuel Rod Design 480
7.5 Core Design Method and 1,000MWe Class Core Design 482
7.5.1 Discussion of Neutronic Calculation Methods 482
7.5.2 Core Design Method 483
7.5.2.1 Nuclear Design Method 485
7.5.2.2 Thermal-Hydraulic Design Method 491
7.5.2.3 Neutronic Thermal-Hydraulic Coupled Equilibrium Core Calculation Method 492
7.5.3 Materials Used in Core Design 494
7.5.4 Fuel Assembly Design 495
7.5.5 Core Arrangement 496
7.5.5.1 In-Vessel Flow Path 496
7.5.5.2 Fuel Loading Pattern for Negative Void Reactivity 497
7.5.6 Design of 1,000MWe Class Core 498
7.6 Subchannel Analysis 506
7.6.1 Introduction 506
7.6.2 Temperature Difference Arising from Subchannel Heterogeneity 508
7.6.2.1 Introduction of Fuel Assembly and Subchannel Parameters 508
7.6.2.2 Rise of Maximum Cladding Surface Temperature by Subchannel Heterogeneity 508
7.6.3 Evaluation of MCST over Equilibrium Cycle 510
7.7 Evaluation of Maximum Cladding Surface Temperature with Engineering Uncertainties 514
7.7.1 Treatment of Downward Flow 514
7.7.2 Nominal Conditions and Uncertainties 516
7.7.3 Statistical Thermal Design of the Super FR 520
7.7.4 Comprehensive Evaluation of Maximum Cladding Surface Temperature at Normal Operation 521
7.8 Design and Improvements of 700MWe Class Core 523
7.8.1 Design of Reference Fuel Rod and Core 524
7.8.2 Core Design Improvement for Negative Local Void Reactivity 524
7.8.2.1 Principles for Reducing Local Void Reactivity 525
7.8.2.2 Sensitivity Analyses for Negative Local Void Reactivity 529
7.8.2.3 Example of Improved Core with Negative Local Void Reactivity 532
7.8.3 Core Design Improvement for Higher Power Density 533
7.8.3.1 Principle of Improving Power Density 533
7.8.3.2 Sensitivity Analyses for Higher Power Density 534
7.8.3.3 Example of Improved Core with Higher Power Density 536
7.9 Plant Control 537
7.9.1 Plant Transient Analysis Code for the Super FR 538
7.9.2 Basic Plant Dynamics of the Super FR 538
7.9.3 Design of Reference Control System 540
7.9.4 Improvement of Feedwater Controller 542
7.9.4.1 Feedback from Power to Flow Rate Ratio 542
7.9.4.2 Feedback from Reactor Power 543
7.9.4.3 Feedback from Derivative of Power 545
7.9.5 Plant Stability Analyses 545
7.9.5.1 10% Decrease in Power Setpoint 546
7.9.5.2 1% Increase in Pressure Setpoint 546
7.9.5.3 4C increase in Steam Temperature Setpoint 547
7.9.5.4 Impulsive Decrease in Feedwater Flow Rate by 5% 547
7.9.5.5 10C Decrease in Feedwater Temperature 548
7.9.6 Comparison of Improved Feedwater Controllers 549
7.9.7 Summary of Improvement of Feedwater Controller 550
7.10 Thermal and Stability Considerations During Power Raising Phase of Plant Startup 551
7.10.1 Introduction 551
7.10.2 Calculation of Flow Distribution 552
7.10.3 Thermal and Thermal-Hydraulic Stability Considerations 554
7.10.4 Sensitivity Analyses 562
7.11 Safety 565
7.11.1 Introduction 565
7.11.2 Analyses of Abnormal Transients and Accidents at Supercritical Pressure 566
7.11.3 Analyses of Loss of Coolant Accidents 571
7.11.4 Analyses of Anticipated Transient Without Scram Events 578
7.12 Summary 579
References 582
Chapter 8: Research and Development 585
8.1 Japan 585
8.1.1 Concept Development 585
8.1.2 Thermal Hydraulics 589
8.1.3 Materials and Water Chemistry 591
8.2 Other Countries 595
8.2.1 Europe 595
8.2.1.1 HPLWR-I Project 595
8.2.1.2 HPLWR-II Project 596
8.2.2 Canada 597
8.2.3 Korea 598
8.2.4 China 598
8.2.5 USA 599
8.3 International Activities 601
8.3.1 Generation-IV International Forum 601
8.3.2 IAEA-Coordinated Research Program 601
8.3.3 International Symposiums 602
References 604
Appendix A: Supercritical Fossil Fired Power Plants - Design and Developments 612
Introduction 612
Improvement of Steam Conditions 612
Boiler Design Features 614
Natural Circulation Boilers 614
Once-Through Boilers (UP: Universal Pressure Boiler for Constant Pressure Operation) 614
Once-Through Boilers (Benson Boilers for Sliding Pressure Operation) 617
Sliding Pressure Operation 617
Typical Arrangement of a Benson Boiler 618
Water Chemistry Guidelines 619
Characteristics of Water Chemistry in Boilers 619
Application of Low pH Coordinated Phosphate Treatment for Natural Circulation Boilers 620
Water Treatment Methods in Actual Circumstances 620
Chemical Analysis Results of the Scale 620
Zinc Compounds in Ammonia Water 622
Reaction of Zinc Compounds in Sodium Phosphate Solution 622
Research Conclusions 623
Doing CWT on Once-Through Type Boilers 623
Observation of Pressure Drop in Boiler 623
Pressure Parts Materials 623
Materials for Conventional Super Critical Boilers 623
Materials for the Advanced Super Critical Boiler 626
Summary 631
References 631
Appendix B: Review of High Temperature Water and Steam Cooled Reactor Concepts 632
Introduction 632
Supercritical Pressure Reactors 632
Water Moderated, Supercritical Steam Cooled Reactor (WH, 1957) 633
Heavy Water Moderated, Light Water Cooled, Once-Through Pressure-Tube Type Reactor (GE Hanford, 1959) 636
SCOTT-R, Once-Through, Graphite Moderated, Light Water Cooled Tube Reactor (WH, 1962) 639
SC-PWR: Indirect-Cycle, Supercritical-Pressure PWR (WH) 640
SCLWR and SCFR: Light Water Cooled (Moderated) Once-Through Reactor with RPV (the University of Tokyo, 1992) 641
B500SKDI, Natural Circulation Integrated SCPWR (Kurchatov, Institute 1992) 645
CANDU-X, Supercritical-Pressure CANDU (AECL, 1998) 648
Nuclear Superheaters (GE, 1950s-1960s) 649
Steam Cooled Fast Breeder Reactors 651
Summary 655
References 631
Index 658