Neutronic Analysis For Nuclear Reactor Systems, Softcover reprint of the original 1st ed. 2017
Langue : Anglais
Auteur : Zohuri Bahman
This book covers the entire spectrum of the science and technology of nuclear reactor systems, from underlying physics, to next generation system applications and beyond. Beginning with neutron physics background and modeling of transport and diffusion, this self-contained learning tool progresses step-by-step to discussions of reactor kinetics, dynamics, and stability that will be invaluable to anyone with a college-level mathematics background wishing to develop an understanding of nuclear power. From fuels and reactions to full systems and plants, the author provides a clear picture of how nuclear energy works, how it can be optimized for safety and efficiency, and why it is important to the future.
Table of Contents
About the Authors
Preface
Acknowledgment
Chapter One: Neutron Physics Background
1.0 Nuclei – Sizes, Composition, and Binding Energies
1.1 Decay of a Nucleus
1.2 Distribution of Nuclides and Nuclear Fission/Nuclear Fusion
1.3 Neutron-Nucleus Interaction
1.3.1 Nuclear Reactions Rates and Neutron Cross Sections
1.3.2 Effects of Temperature on Cross Section
1.3.3 Nuclear Cross Section Processing Codes
1.3.4 Energy Dependence of Neutron Cross Sections
1.3.5 Types of Interactions
1.4 Mean Free Path
1.5 Nuclear Cross Section and Neutron Flux Summary
1.6 Fission
1.7 Fission Spectra
1.8 The Nuclear Fuel
1.6.1 Fertile Material
1.9 Liquid Drop Model of a Nucleus
1.10 Summary of Fission Process
1.11 Reactor Power Calculation
1.12 Relationship between Neutron Flux and Reactor Power
1.13 References
1.14 Problems
Chapter Two: Modeling Neutron Transport and Interactions
2.0 Transport Equations
2.1 Reaction Rates
2.2 Reactor Power Calculation
2.3 Relationship between Neutron Flux and Reactor Power
2.4 Neutron Slowing Down and Thermalization
2.5 Macroscopic Slowing Down Power
2.6 Moderate Ratio
2.7 Integro-Differential Equation (Maxwell-Boltzmann Equation)
2.8 Integral Equation
2.9 Multigroup Diffusion Theory
2.10 The Multigroup Equations
2.11 Generating the Coefficients
2.12 Simplifications
2.13 Nuclear Criticality Concepts
2.14 Criticality Calculation
2.15 The Multiplication Factor and a Formal Calculation of Criticality
2.16 Fast Fission Factor Definition
2.17 Resonance Escape Probability
2.18 Group Collapsing
2.18.1 Multigroup Collapsing to One Group
2.18.2 Multigroup Collapsing to Two Group
2.18.3 Two Group Criticality
2.19 The Infinite Reactor
2.20 Finite Reactor
2.21 Time Dependence
2.22 Thermal Utilization Factor
2.23 References
2.24 Problems
Chapter Three: Spatial Effects in Modeling Neutron Diffusion – One Group Models
3.0 Nuclear Reactor Calculations
3.1.1 Neutron Spectrum
3.2 Control Rods in Reactors
3.2.1 Lattice Calculation Analysis
3.3 An Introduction to Neutron Transport Equation
3.4 Neutron Current Density Concept in General
3.5 Neutron Current Density and Fick’s Law
3.6 Problem Classification and Neutron Distribution
3.7 Neutron Slowing Down
3.8 Neutron Diffusion Concept
3.9 The One Group Model and One Dimensional Analysis
3.10.1 Boundary Conditions for the Steady-State Diffusion Equation
3.10.2 Boundary Conditions – Consistent and Approximate
3.10.3 An Approximate Methods for Solving the Diffusion Equation
3.10.4 The P1 Approximate Methods in Transport Theory
3.11 Further Analysis Methods for One Group
<3.11.1 Slab Geometry
3.11.2 Cylindrical Geometry
3.11.3 Spherical Geometry
3.12 Eigenfunction Expansion Methods and Eigenvalue Equations
3.12.1 Eigenvalues and Eigenfunctions Problems
3.13 Multi-Dimensional Models and Boundary Conditions
3.13.1 The Unreflected Reactor Parallelepiped Core
3.13.2 The Minimum Volume of the Critical Parallelepiped
3.13.3 The Peak to Average Flux Ratio
3.13.4 The Finite Height Cylindrical Core
3.14 Relating k to the Criticality Condition
3.15 Analytical Solution for the Transient Case for Reactor
3.16 Criticality
3.17 Bare Critical Reactor 1-Group Model
3.18 Bare Critical Reactor 1-Group Model, Finite Geometries
3.19 Reflected Critical Reactors- 1-Group Model
3.20 Infinite Reflector Case
3.21 Criticality for General Bare Geometries
3.22 Reflected Reactor Geometries
3.23 Reactor Criticality Calculations
3.24 References
3.25 Problems
Chapter Four: Energy Effects in Modeling Neutron Diffusion – Two Group Models
4.0 One-Group Diffusion Theory
4.1 Two-Group Diffusion Theory
4.2 Few Group Analysis
4.2.1 2-Group Thermal Reactor Equations
4.2.2 2-Group Fast Reactor Equations
4.3 Transverse Buckling Approximation
4.4 Consistent Diffusion Theory Boundary Conditions
4.5 Derivation of the One-Dimensional Multi-Group PN Equations
4.6 Multi-Group Diffusion Equations - Solution Approach
4.6.1 Infinite Medium for Group Collapse
4.6.2 Zero-Dimensional Spectrum for Group Collapse
4.6.3 Group Collapsing
4.6.4 Group Collapse
4.7 References
4.8 Problems
Chapter Five: Numerical Methods in Modeling Neutron Diffusion
5.0 Introduction
5.1 Problem(s) Solved
5.1.1 Transport Equation
5.1.2 Angle Discretization
5.1.3 Energy Discretization
5.1.4 Spatial Discretization
5.1.5 Matrix Formulation
5.2 Solution Strategy
5.2.1 Types of Outer Iterations
5.2.2 Inhomogeneous Source (No Fission)
5.2.3 Inhomogeneous Source (With Fission)
5.2.4 Fission Eigenvalue Calculation
5.2.5 Eigenvalue Search Calculation
5.3 Middle Iterations
5.4 Inner Iterations
5.5 Upscatter Iterations
5.6 Inhomogeneous Sources
5.7 Background Concepts
5.7.1 Mixing Tables
5.7.2 Cross Section Collapsing
5.8 Input Description5.9 Output Description
5.10 References
5.11 Problems
Chapter Six: Slowing Down Theory
6.0 Neutron Elastic and Inelastic Scattering for Slowing Down
6.1 Derivation of the Energy and Transfer Cross Section
6.1.1 Elastic Scattering
6.1.2 Inelastic Scattering
6.2 Derivation of the Isotropic Flux in an Infinite Hydrogen Moderator
6.3 Derivation of the Isotropic Flux in a Moderator Other than Hydrogen A > 1
6.4 Summary of Slowing Down Equations
6.5 References
6.6 Problems
Chapter Seven: Resonance Processing
7.0 Difficulties Presented by Resonance Cross Sections
7.1 What is Nuclear Resonance -- Compound Nucleus
7.1.1 Breit-Wigner Resonance Reaction Cross Sections
7.1.2 Resonance and Neutron Cross Section
7.2 Doppler Effect and Doppler Broadening of Resonance
7.3 Doppler Coefficient in Power Reactors
7.4 Infinite Resonance Integrals and Group Cross Section
7.4.1 The Flux Calculator Method
7.4.2 The Bondarenko Method - The Bondarenko Factor
7.4.3 The CENTRM Method
7.5 Infinite Resonance Integrals and Group Cross Sections
7.6 Dilution Cross Section - Dilution Factor
7.7 Resonance Effects
7.8 Homogeneous Narrow Resonance Approximation
7.9 Homogeneous Wide Resonance Approximation
7.10 Heterogeneous Narrow Resonance Approximation
7.11 Heterogeneous Wide Resonance Approximation
7.12 References
7.13 Problems
Chapter Eight: Heterogeneous Reactors and Wigner Seitz Cells
8.0 Homogeneous and Heterogeneous Reactors
8.1 Spectrum Calculation in Heterogeneous Reactors
8.2 Cross Section Self Shielding and Wigner-Seitz Cells
8.3 References
8.4 Problems
Chapter Nine: Thermal Spectra and Thermal Cross Sections
<9.0 Coupling to Higher Energy Sources9.1 Chemical Binding and Scattering Kernels
9.1.1 Scattering Materials
9.1.2 Thermal Cross Section Average
9.2 Derivation of the Maxwell-Boltzmann Spectrum
9.3 References
9.4 Problems
Chapter Ten: Perturbation Theory for Reactor Neutronics
10.0 Perturbation Theory
10.1 Zero Dimensional Methods
10.2 Spatial Method (1 Group)
10.3 References
10.4 Problems
Chapter Eleven: Reactor Kinetics and Point Kinetics
11.0 Time Dependent Diffusion Equation
11.1 Derivation of Exact Point Kinetics Equations (EPKE)
11.2 The Point Kinetics Equations
11.3 Dynamic versus Static Reactivity
11.4 Calculating the Time Dependent Shape Function
11.5 Point Kinetics Approximations
11.5.1 Level of Approximation to the Point Kinetics Equations
11.6 Adiabatic Approximation
11.7 Adiabatic Approximation with Pre-Computed Shape Functions
11.8 Quasi-Static Approximation
11.9 Zero Dimensional Reactors
11.10 References
11.11 Problems
Chapter Twelve: Reactor Dynamics
12.0 Background on Nuclear Reactor
12.1 Neutron Multiplication
12.2 Simple Feedbacks
12.3 Multiple Time Constant Feedbacks
12.4 Fuchs-Nordheim models
12.5 References
12.6 Problems
Chapter Thirteen: Reactor Stability
13.0 Frequency Response
13.1 Nyquist Plots
13.2 Non-Linear Stability
13.3 References
13.4 Problems
Chapter Fourteen: Numerical Modeling for Time Dependent Problems
14.0 Fast Breeder Reactor History and Status
14.1 The Concept of Stiffness
14.2 The Quasi-Static Method
14.3 Bethe-Tait Models
14.4 References
14.5 Problems
Chapter Fifteen: Fission Product Buildup and Decay
15.0 Background Introduction
15.1 Nuclear Fission and the Fission Process
15.2 Radioactivity and Decay of Fission Product
15.3 Poisons Produced by Fission
15.4 References
15.5 Problems
Chapter Sixteen: Fuel Burnup and Fuel Management
16.0 The World’s Energy Resources
16.1 Today’s Global Energy Market
16.2 Fuel Utilization and Fuel Burnup
16.3 Fuel Reprocessing
16.3.1 PUREX Process
16.3.2 Transuranium Elements
16.3.3 Vitrification
16.4 Fuel Management for Nuclear Reactors
16.5 Nuclear Fuel Cycle
16.6 Store and Transport High Burnup Fuel
16.7 Nuclear Reactors for Power Production
16.8 Future Nuclear Power Plants Systems
16.9 Next Generation of Nuclear Power Reactors for Power Production
16.10 References
16.11 ProblemsAppendix A: Laplace Transforms
A-1 Definition of Laplace Transform
A-2 Basic Transforms
A-3 Fundamental Properties
A-4 Inversion by Complex Variable Residue Theorem
Appendix B: Transfer Functions and Bode Plots
B-1 Transfer Functions
B-2 Sample Transforms
B-3 Fourier Transforms
B-4 Transfer Functions
B-4 Feedback and Control
B-5 Graphical Representation (Bode and Nyquist Diagram)
B-6 Root Locus Construction Rules
B-7 References
INDEX
Dr. Bahman Zohuri is founder of Galaxy Advanced Engineering, Inc. a consulting company that he formed upon leaving the semiconductor and defense industries after many years as a Senior Process Engineer for corporations including Westinghouse and Intel, and then as Senior Chief Scientist at Lockheed Missile and Aerospace Corporation. During his time with Westinghouse Electric Corporation, he performed thermal hydraulic analysis and natural circulation for Inherent Shutdown Heat Removal System (ISHRS) in the core of a Liquid Metal Fast Breeder Reactor (LMFBR). While at Lockheed, he was responsible for the study of vulnerability, survivability and component radiation and laser hardening for Defense Support Program (DSP), Boost Surveillance and Tracking Satellites (BSTS) and Space Surveillance and Tracking Satellites (SSTS). He also performed analysis of characteristics of laser beam and nuclear radiation interaction with materials, Transient Radiation Effects in Electronics (TREE), Electromagnetic Pulse (EMP), System Generated Electromagnetic Pulse (SGEMP), Single-Event Upset (SEU), Blast and, Thermo-mechanical, hardness assurance, maintenance, and device technology. His consultancy clients have included Sandia National Laboratories, and he holds patents in areas such as the design of diffusion furnaces, and Laser Activated Radioactive Decay. He is the author of several books on nuclear engineering heat transfer.
Covers the fundamentals of neutronic analysis for nuclear reactor systems in order to help readers understand nuclear reactor theory
Applies the described principles to nuclear reactor power systems
Assumes no previous knowledge of nuclear physics or engineering
Date de parution : 06-2018
Ouvrage de 551 p.
15.5x23.5 cm
Disponible chez l'éditeur (délai d'approvisionnement : 15 jours).
Prix indicatif 158,24 €
Ajouter au panier
© 2024 LAVOISIER S.A.S.
Ces ouvrages sont susceptibles de vous intéresser
Nuclear Reactor Physics 190,72 €
