Dynamic Simulation of Sodium Cooled Fast Reactors

Description

This book provides the basis of simulating a nuclear plant, in understanding the knowledge of how such simulations help in assuring the safety of the plants, thereby protecting the public from accidents. It provides the reader with an in-depth knowledge about modeling the thermal and flow processes in a fast reactor and gives an idea about the different numerical solution methods. The text highlights the application of the simulation to typical sodium-cooled fast reactor. The book • Discusses mathematical modeling of the heat transfer process in a fast reactor cooled by sodium. • Compares different numerical techniques and brings out the best one for the solution of the models. • Provides a methodology of validation based on experiments. • Examines modeling and simulation aspects necessary for the safe design of a fast reactor. • Emphasizes plant dynamics aspects, which is important for relating the interaction between the components in the heat transport systems. • Discusses the application of the models to the design of a sodium-cooled fast reactor It will serve as an ideal reference text for senior undergraduate, graduate students, and academic researchers in the fields of nuclear engineering, mechanical engineering, and power cycle engineering. Cover Half Title Title Page Copyright Page Table of Contents Preface About the Author Chapter 1: Introduction 1.1 General 1.2 Basics of Breeding 1.3 Uranium Utilization 1.4 Components of Fast Reactors 1.5 Overview of Fast Reactor Programs 1.6 Need for Dynamic Simulation 1.7 Design Basis 1.8 Plant Protection System 1.9 Sensors and Response Time 1.10 Scope of Dynamic Studies 1.11 Modeling Development Assignment References Chapter 2: Description of Fast Reactors 2.1 Introduction 2.2 Fast Breeder Test Reactor (FBTR) 2.2.1 Reactor Core 2.2.2 Reactor Assembly 2.2.3 Sodium Systems 2.2.4 Decay Heat Removal 2.2.5 Generating Plant 2.2.6 Instrumentation and Control 2.2.7 Safety 2.3 Prototype Fast Breeder Reactor 2.3.1 Reactor Core 2.3.2 Reactor Assembly 2.3.3 Main Heat Transport System 2.3.4 Steam Water System 2.3.5 Instrumentation and Control 2.3.6 Safety 2.4 Neutronic Characteristics of SFRs 2.5 Thermal-Hydraulic Characteristics of SFR Assignment References Chapter 3: Reactor Heat Transfer 3.1 Introduction 3.2 Reactor Core 3.2.1 Core Description 3.2.2 Fuel Pin 3.2.3 Subassembly 3.3 Coolant Selection 3.4 Control Material Selection 3.5 Structural Material Selection 3.6 Heat Generation 3.7 Reactivity Feedback 3.7.1 Doppler Effect 3.7.2 Sodium Density and Void Effects 3.7.3 Fuel Axial Expansion Effect 3.7.4 Structural Expansion 3.7.5 Bowing 3.8 Decay Heat 3.9 Solution Methods 3.9.1 Prompt Jump Approximation 3.9.2 Runge Kutta Method 3.9.3 Kaganove Method 3.9.4 Comparison of the Different Methods 3.9.5 Solution Methodology 3.10 Heat Transfer in Primary System 3.10.1 Core Thermal Model 3.10.2 Fuel Restructuring 3.10.3 Gap Conductance 3.10.4 Fuel Thermal Model 3.10.5 Solution Technique 3.11 Determination of Peak Temperatures: Hot Spot Analysis 3.12 Core Thermal Model Validation in FBTR and SUPER PHENIX 3.13 Mixing of Coolant Streams in Upper Plenum 3.13.1 Solution Technique 3.14 Lower Plenum/Cold Pool 3.15 Grid Plate 3.16 Heat Transfer Correlations for Fuel Rod Bundle Assignment References Chapter 4: IHX Thermal Model 4.1 Introduction 4.2 Experience in PHENIX 4.3 Thermal Model 4.4 Solution Techniques 4.4.1 Nodal Heat Balance Scheme 4.4.2 Finite Differencing Scheme 4.5 Choice of Numerical Scheme 4.5.1 Nodal Heat Balance for Unbalanced Flows 4.5.2 Modified Nodal Heat Balance Scheme (MNHB) 4.6 Heat Transfer Correlations 4.7 Validation Assignment References Chapter 5: Thermal Model of Piping 5.1 Introduction 5.2 Thermal Model 5.3 Solution Methods 5.4 Comparison of Piping Models Assignment References Chapter 6: Sodium Pump 6.1 Introduction 6.2 Electromagnetic Pumps 6.3 Centrifugal Pump 6.3.1 Pump Hydraulic Model 6.3.2 Pump Dynamic Model 6.3.3 Pump Thermal Model Assignment References Chapter 7: Transient Hydraulics Simulation 7.1 Introduction 7.2 Momentum Equations 7.3 Free Level Equations 7.4 Core Coolant Flow Distribution 7.5 IHX Pressure Drop Correlations 7.5.1 Resistance Coefficient for Cross-Flow 7.5.2 Resistance Coefficient for Axial Flow 7.6 Pump Characteristics 7.7 Computational Model 7.8 Validation Studies 7.9 Secondary Circuit Hydraulics 7.9.1 Secondary Hydraulics Model 7.9.2 Natural Convection Flow in Sodium Validation Studies Assignment References Chapter 8: Steam Generator 8.1 Introduction 8.2 Heat Transfer Mechanisms 8.3 Steam Generator Designs 8.3.1 Conventional Fossil-Fueled Boilers 8.3.1.1 Drum Type 8.3.1.2 Once Through Steam Generators (OTSG) 8.3.2 Sodium-Heated Steam Generators 8.4 Thermodynamic Models 8.5 Mathematical Model 8.6 Heat Transfer Correlations 8.6.1 Single-Phase Liquid Region 8.6.2 Nucleate Boiling 8.6.3 Dry-Out 8.6.4 Post Dry-Out 8.6.5 Superheated Region 8.6.6 Sodium Side Heat Transfer 8.7 Pressure Drop 8.8 Computational Model 8.8.1 Solution of Water/Steam Side Equations 8.8.2 Solution of Sodium, Shell, and Tube Wall Equations 8.9 Steam Generator Model Validation Assignment References Chapter 9: Computer Code Development 9.1 Introduction 9.2 Organization of DYNAM 9.3 Axisymmetric Code STITH-2D 9.4 One-Dimensional CFD-Coupled Dynamics Tool 9.5 Comparison of Predictions of DYANA-P and DYANA-HM References Chapter 10: Specifying Sodium Pumps Coast-Down Time 10.1 Introduction 10.2 Impact of Coast-Down Time in Loop-Type SFR 10.3 Impact of Coast-Down Time in Pool-Type SFR 10.4 Considerations for Determining Flow Coast-Down Time 10.5 SCRAM Threshold versus Coast-Down Time 10.5.1 FHT Effect on Maximum Temperatures 10.5.2 FHT to Avoid SCRAM for Short Power Failure 10.6 Secondary Pump FHT 10.7 Primary FHT for Unprotected Loss of Flow Assignment References Chapter 11: Plant Protection System 11.1 Introduction 11.2 Limiting Safety System Settings (LSSS) for FBTR 11.2.1 Safety Signals and Settings 11.2.2 Limiting Safety System Settings (LSSS) Adequacy 11.3 Limiting Safety System Settings for PFBR 11.3.1 Design Basis Events 11.3.2 Core Design Safety Limits 11.3.3 Selection of SCRAM Parameters 11.4 Shutdown System 11.5 Event Analysis Assignment References Chapter 12: Decay Heat Removal System 12.1 Introduction 12.2 Natural Convection Basics 12.3 DHR System Options in SFR 12.3.1 DHR in Primary Sodium 12.3.2 DHR in Secondary Sodium 12.3.3 Steam Generator Auxiliary Cooling System 12.3.4 DHR Through Steam Water System 12.3.5 Reactor Vessel Auxiliary Cooling System 12.4 DHR in FBTR 12.4.1 Heat Removal by Air in SG Casing 12.4.2 Loss of Off-site and On-site Power with SG Air Cooling 12.4.3 Loss of Off-site and On-site Power without Reactor Trip (ULOF) 12.5 DHR in PFBR 12.5.1 Thermal Model 12.5.2 Decay Heat Exchanger (DHX) Thermal Model 12.5.3 Hot Pool Thermal Model 12.5.4 Air Heat Exchanger (AHX) Thermal Model 12.5.5 Piping 12.5.6 Expansion Tank 12.5.7 Air Stack/Chimney 12.5.8 Hydraulic Model of SGDHR 12.5.9 DHDYN Validation on SADHANA Loop 12.6 Role of Inter-Wrapper Flow 12.7 Role of Secondary Thermal Capacity Assignment References Chapter 13: Modeling of Large Sodium–Water Reaction 13.1 Introduction 13.2 Leak Rate 13.2.1 Water Leak Rate Model 13.2.2 Steam Leak Rate Model 13.3 Reaction Site Dynamics 13.3.1 Spherical Bubble Model 13.3.2 Columnar Bubble Model 13.3.3 Solution Technique 13.3.4 Validation of Reaction Site Model 13.4 Sodium Side System Transient 13.5 Discharge Circuit System Transient 13.6 Analysis of Pressure Transients for PFBR 13.7 Failure of a Greater Number of Tubes Than Design Basis Leak Nomenclature Assignment References Appendix A: Brief Description of Codes A.1 SAS4A/SASSYS-1 code (ANL) A.2 CATHARE code and specific model for SFR calculations (CEA and IRSN) A.3 STAR-CD and DYANA-P System Codes (IGCAR) A.4 GRIF Code A.5 NETFLOW++ code (University of Fukui) A.6 MARS-LMR Code (KAERI) A.7 TRACE Code (PSI) A.8 SAC-CFR (NCEPU) A.9 THACS (XJTU) References Appendix B: Primary Pump Discharge Pipe Break Modeling B.1 Introduction B.2 Model Equations B.3 Numerical Solution B.4 Results References Index