• TABLE OF CONTENTS
• 
  
• Foreword – Chancellor, HBNI
• Foreword – Chairman, Atomic Energy Commission, Secretary, Department of Atomic Energy
• Preface
• Acknowledgements
• Contributors
• 
  
• Chapter 1: Introduction to Nuclear Reactor Thermal Hydraulics
• 
  
• 1.1 Introduction
• 1.2 Fission Process
• 1.2.1 Fission Energy
• 1.3 Neutronics – Thermal-Hydraulics – Structural Analysis
• 1.4 Nuclear Reactor Concepts
• 1.5 Reactor Types
• 1.5.1 Pressurised Water Reactor (PWR)
• 1.5.2 Boiling Water Reactor (BWR)
• 1.5.3 Pressurised Heavy Water Reactor (PHWR)
• 1.5.4 Sodium Cooled Fast Reactor (SFR)
• 1.5.5 Gas Cooled Reactors (GCR)
• 1.6 Active and Passive Safety
• 1.7 Reactor Evolution
• 1.7.1 Molten Salt Reactor
• 1.7.2 Fusion
• 1.8 Thermal-Hydraulic Processes in Nuclear Reactors
• 1.9 Thermodynamics
• 1.9.1 Laws of Thermodynamics
• 1.9.2 Carnot Cycle
• 1.9.3 Rankine Cycle
• 1.9.4 Reheat Cycle
• 1.9.5 Regenerative Cycle (Feed Water Heating)
• 1.9.6 Brayton Cycle
• 1.10 Role of Thermal Hydraulics
• 1.10.1 Characteristics of a Reactor Coolant
• 1.10.2 Characteristics of a Reactor Fuel
• 1.10.3 Characteristics of Cladding Materials
• 1.11 Single and Two-Phase Flows
• 1.11.1 Critical Heat Flux
• 1.11.2 Critical/Choked Flow
• 1.12 Scope of this Monograph
• Bibliography
• Assignments
• 
  
• Chapter 2: Heat Generation and Reactor Kinetics
• 
  
• 2.1 Introduction
• 2.2 Heat Generation in Nuclear Fuel
• 2.3 Heat Generation in Moderator
• 2.4 Heat Generation in Reflectors and Shields
• 2.5 Heat Generation in Structures
• 2.6 Heterogeneous Core
• 2.7 Shutdown Energy Generation
• 2.8 Reactivity
• 2.8.1 Feedback Reactivity
• 2.9 Reactor Kinetics
• 2.10 Metallic Fuel vs. Ceramic Fuel
• 2.11 Fuel Design Principles
• 2.12 Simplified Treatment of Heat Generation Calculation
• 2.12.1 Volumetric Heat Generation Rate in Nuclear Reactor Fuel
• 2.12.2 Determination of Maximum Neutron Flux
• 2.12.3 Maximum and Average Linear Power of the Fuel Channel
• Nomenclature
• Bibliography
• Assignments
• 
  
• Chapter 3: Fundamentals of Nuclear Reactor Safety
• 
  
• 3.1 Introduction
• 3.2 Physics of Nuclear Safety
• 3.3 Designing for Safety of Nuclear Power Plants
• 3.4 Postulated Initiating Events (PIE)
• 3.5 Design Basis Events (DBE)
• 3.6 Redundancy, Diversity and Independence
• 3.6.1 Redundancy
• 3.6.2 Diversity
• 3.6.3 Independence
• 3.7 Safety Analysis
• 3.7.1 Deterministic Safety Analysis
• 3.7.2 Risk and Probabilistic Safety Analysis (PSA)
• 3.8 Safety Criteria for Nuclear Power Plants
• 3.9 Provisions of Safety Features in Design of Indian PHWRs
• 3.9.1 Shutdown System
• 3.9.2 Decay Heat Removal System
• 3.10 Design Provisions of Indian PHWRs Against Severe Plant Conditions
• 3.10.1 Assuring Shutdown
• 3.10.2 Assuring Decay Heat Removal
• 3.11 Provisions of Safety Features in Design of Indian FBRs
• 3.11.1 Shutdown Systems (SDS)
• 3.11.2 Decay Heat Removal Systems
• 3.12 Design Provisions in SFR Against Severe Plant Conditions
• 3.12.1 Assuring Shut Down
• 3.12.2 Sodium Leaks
• 3.12.3 Sodium Water Reaction in Steam Generator
• 3.12.4 Core Catcher
• 3.13 History of Some Accidents
• 3.13.1 Three Mile Island Accident (1979)
• 3.13.2 Chernobyl Accident (1986)
• 3.13.3 Fukushima Accident (2011)
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 4: Fundamentals of Single-Phase Fluid Flow and Heat Transfer
• 
  
• 4.1 Introduction
• 4.2 Physical Principles Governing Fluid Dynamics
• 4.3 Conservation of Mass
• 4.3.1 Mass Conservation Equation Using System Approach
• 4.3.2 Mass Conservation Equation Using Control Volume Approach
• 4.3.3 Condition for Incompressible Flow
• 4.4 Conservation of Momentum
• 4.4.1 Deformation of Flowing Fluid Elements
• 4.4.2 Navier–Stokes Equation
• 4.4.3 Euler Equations
• 4.5 Conservation of Energy
• 4.5.1 Mechanical Energy Balance Equation
• 4.5.2 Thermal Energy Balance Equation
• 4.6 Boundary Layer Theory
• 4.6.1 Velocity Boundary Layer
• 4.6.2 Thermal Boundary Layer
• 4.6.3 Heat Transfer Coefficient
• 4.6.4 Prandtl Number and Boundary Layer Thickness
• 4.7 Turbulent Flow
• 4.7.1 Characteristics of Turbulent Flow
• 4.8 Approaches to Turbulence Modelling
• 4.8.1 Direct Numerical Simulation
• 4.8.2 Large Eddy Simulation
• 4.8.3 Reynolds Decomposition for Turbulence
• 4.9 Convective Heat Transfer
• 4.9.1 Natural Convection
• 4.9.2 Natural Convection Over a Heated Vertical Plate
• 4.9.3 Non-Dimensional Parameters in Natural Convection
• Summary
• Nomenclature
• Subscripts
• Greek Symbols
• Bibliography
• Assignments
• 
  
• Chapter 5: Fundamentals of Two-Phase Flow and Heat Transfer
• 
  
• 5.1 Introduction
• 5.2 Definition of the Basic Two-Phase Flow Parameters
• 5.3 Flow Regimes and Analysis Methodologies
• 5.3.1 Flow Regimes
• 5.3.2 Analysis Methodology
• 5.3.3 Drift Flux Model
• 5.3.4 Separated Flow Model
• 5.4 Pressure Drop in Two-Phase Flow
• 5.4.1 Pressure Drop Relationship
• 5.4.2 Flow Pattern Specific Pressure Drop Relationships
• 5.5 Void Fraction Correlations
• 5.5.1 Slip Ratio Models
• 5.5.2 Kβ Models
• 5.5.3 Correlations Based on the Drift Flux Model
• 5.6 Boiling Heat Transfer
• 5.6.1 Pool Boiling
• 5.6.2 Flow Boiling
• 5.7 Critical Heat Flux
• 5.7.1 CHF by Departure from Nucleate Boiling (DNB)
• 5.7.2 Conditions Leading to DNB
• 5.7.3 CHF by Film Dryout
• 5.8 Important Correlation for CHF in Nuclear Reactors
• Summary
• Nomenclature
• Subscripts
• Bibliography
• Assignments
• 
  
• Chapter 6: Fuel Element Heat Transfer
• 
  
• 6.1 Introduction
• 6.2 Principal Fuel and Clad Materials
• 6.3 Heat Conduction in Fuel Elements
• 6.4 Thermal Properties of UO₂
• 6.4.1 Thermal Conductivity
• 6.4.2 Fission Gas Release
• 6.4.3 Melting Point
• 6.4.4 Specific Heat
• 6.5 Steady State Temperature Distribution in Plate Type Fuel Elements
• 6.6 Steady State Temperature Distribution in Cylindrical Fuel Pins
• 6.7 Temperature Distribution in Restructured Fuel Elements
• 6.7.1 Zone-3
• 6.7.2 Zone-2
• 6.7.3 Zone-1
• 6.7.4 Implications of Fuel Restructuring in Core Design
• 6.8 Pellet-Clad Gap Conductance
• 6.9 Overall Resistance
• 6.10 Thermal Contact Conductance in PHWR
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 7: Heat and Fluid Flow in Reactor Components
• 
  
• 7.1 Introduction
• 7.2 Single-Phase / Boiling Pressure Drop
• 7.3 Pressure Drop in Fuel Rod Bundle
• 7.4 Effect of Creep
• 7.5 Effect of Bundle Management
• 7.6 Pressure Drop in a Vertical Boiling Pipe
• 7.7 Orificing Requirement
• 7.8 Hot Spot Factors in Nuclear Reactor Design
• 7.9 Subchannel Analysis
• 7.10 Subchannel Analysis Model
• 7.11 Liquid Film Flow Analysis
• 7.12 Flow Loops
• 7.13 Loss of Flow and Pump Coast Down
• 7.14 Flow Instabilities in Thermosyphon System
• 7.15 Heat Transfer During LOCA
• Summary
• Nomenclature
• Subscripts
• Bibliography
• Assignments
• 
  
• Chapter 8: Numerical Methods in Fluid Flow and Heat Transfer
• 
  
• 8.1 Introduction
• 8.2 Overview of CFD Simulation
• 8.3 Deciding Objective of the CFD Simulation
• 8.4 Deciding the Computational Domain
• 8.5 Identifying Type of Flow
• 8.6 Governing Equations
• 8.7 Discretization
• 8.8 Mesh Generation
• 8.9 Discretization of Governing Equations
• 8.10 Finite Difference Method
• 8.11 Finite Volume Method
• 8.12 Discretization of 2D Steady State Conduction Equation using Finite Volume Method
• 8.13 Discretization of 3D Scalar Transport Equation
• 8.14 Navier–Stokes Equations
• 8.15 Boundary Conditions
• 8.16 Numerical Solution Methods for Discretized Equations
• 8.17 Numerical Solution of Multi-Phase Flows
• 8.18 Post-Processing of CFD Simulation Results
• Bibliography
• Exercises
• 
  
• Chapter 9: Thermal Hydraulic Analysis of PHWRs
• 
  
• 9.1 Introduction
• 9.2 Primary Heat Transport System
• 9.3 Secondary Coolant System
• 9.4 Reactor Shutdown Systems (SDS)
• 9.5 Emergency Core Cooling System (ECCS)
• 9.6 Containment Spray System (CSS)
• 9.7 Passive Decay Heat Removal System (PDHRS)
• 9.8 Deterministic Safety Analysis (DSA)
• 9.9 Reactor Kinetics
• 9.10 Transient Analysis for Loss of Regulation Accident (LORA)
• 9.11 Accident Analysis (DBA, DEC): Modelling and Results
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 10: Transient Analysis of Fast Reactors
• 
  
• 10.1 Introduction
• 10.2 Fast Reactor
• 10.3 Features of Transient Models
• 10.4 Modelling of Primary Sodium System
• 10.5 Secondary Sodium Circuit Thermal Model
• 10.6 Integration of the Models and Calculation Sequence
• 10.7 Design Basis Events
• 10.8 Core Design Safety Limits
• 10.9 Reactor SCRAM Parameters
• 10.10 PFBR Transient Studies
• 10.11 Safety Grade Decay Heat Removal System
• Summary
• Nomenclature
• Subscripts
• Bibliography
• Assignments
• 
  
• Chapter 11: Thermal Hydraulics of Advanced Reactors
• 
  
• 11.1 Introduction
• 11.2 Drivers for Introduction of Passive Safety Features in Advanced Nuclear Power Plants
• 11.3 Impact of Introduction of Passive Safety Features on Safety
• 11.4 Trend in Design of Advanced Reactors with Passive Safety Features
• 11.5 Relevance of Passive Safety Features for Large Scale Deployment of Nuclear Power
• 11.6 Thermal Hydraulics Phenomena with Passive Systems
• 11.7 Modelling Challenges
• Summary
• Bibliography
• Assignments
• 
  
• Annexure A
• Annexure B
• Annexure C
• 
  
• Index

Dr. G. Vaidyanathan is a retired Outstanding Scientist and Director, Fast Reactor Technology from the Indira Gandhi Centre for Atomic Research, India. His experience of 38 years includes Design, Analysis, Experimentation and Project Management in the area of nuclear energy. He has been involved in the Sodium Cooled Fast Reactors and Pressurised Heavy Water Reactors, which form the main stay of Indian Nuclear Power Programme. He was a visiting professor at SRM University for nearly 8 years, where he taught the subjects on Nuclear Engineering, safety, and Alternative systems of Energy. He was also a guest faculty in Nuclear Engineering at the Indian Institute of Technology, Madras, where he taught post graduate students. He has also taught at Amity University and National Institute of Technologies.