Neutronic Analysis For Nuclear Reactor Systems
Autor Bahman Zohurien Limba Engleză Paperback – 27 iun 2018
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Specificații
ISBN-13: 9783319827063
ISBN-10: 3319827065
Pagini: 551
Ilustrații: XXII, 551 p. 148 illus., 58 illus. in color.
Dimensiuni: 155 x 235 mm
Greutate: 0.79 kg
Ediția:Softcover reprint of the original 1st ed. 2017
Editura: Springer International Publishing
Colecția Springer
Locul publicării:Cham, Switzerland
ISBN-10: 3319827065
Pagini: 551
Ilustrații: XXII, 551 p. 148 illus., 58 illus. in color.
Dimensiuni: 155 x 235 mm
Greutate: 0.79 kg
Ediția:Softcover reprint of the original 1st ed. 2017
Editura: Springer International Publishing
Colecția Springer
Locul publicării:Cham, Switzerland
Cuprins
Table of Contents
About the Authors
Preface
Acknowledgment
Chapter One: Neutron Physics Background
1.0Nuclei – Sizes, Composition, and Binding Energies
1.1Decay of a Nucleus
1.2Distribution of Nuclides and Nuclear Fission/Nuclear Fusion
1.3Neutron-Nucleus Interaction
1.3.1Nuclear Reactions Rates and Neutron Cross Sections
1.3.2Effects of Temperature on Cross Section
1.3.3Nuclear Cross Section Processing Codes
1.3.4Energy Dependence of Neutron Cross Sections
1.3.5Types of Interactions
1.4Mean Free Path
1.5Nuclear Cross Section and Neutron Flux Summary
1.6Fission
1.7Fission Spectra
1.8The Nuclear Fuel
1.6.1Fertile Material
1.9Liquid Drop Model of a Nucleus
1.10Summary of Fission Process
1.11Reactor Power Calculation
1.12Relationship between Neutron Flux and Reactor Power
1.13References
1.14Problems
Chapter Two: Modeling Neutron Transport and Interactions
2.0Transport Equations
2.1Reaction Rates
2.2Reactor Power Calculation
2.3Relationship between Neutron Flux and Reactor Power
2.4Neutron Slowing Down and Thermalization
2.5Macroscopic Slowing Down Power
2.6Moderate Ratio
2.7Integro-Differential Equation (Maxwell-Boltzmann Equation)
2.8Integral Equation
2.9Multigroup Diffusion Theory
2.10The Multigroup Equations
2.11Generating the Coefficients
2.12Simplifications
2.13Nuclear Criticality Concepts
2.14Criticality Calculation
2.15The Multiplication Factor and a Formal Calculation of Criticality
2.16Fast Fission Factor Definition
2.17Resonance Escape Probability
2.18Group Collapsing
2.18.1Multigroup Collapsing to One Group
2.18.2Multigroup Collapsing to Two Group
2.18.3Two Group Criticality
2.19The Infinite Reactor
2.20Finite Reactor
2.21Time Dependence
2.22Thermal Utilization Factor
2.23References
2.24Problems
Chapter Three: Spatial Effects in Modeling Neutron Diffusion – One Group Models
3.0Nuclear Reactor Calculations
3.1.1Neutron Spectrum
3.2Control Rods in Reactors
3.2.1Lattice Calculation Analysis
3.3An Introduction to Neutron Transport Equation
3.4Neutron Current Density Concept in General
3.5Neutron Current Density and Fick’s Law
3.6Problem Classification and Neutron Distribution
3.7Neutron Slowing Down
3.8Neutron Diffusion Concept
3.9The One Group Model and One Dimensional Analysis
3.10.1Boundary Conditions for the Steady-State Diffusion Equation
3.10.2Boundary Conditions – Consistent and Approximate
3.10.3An Approximate Methods for Solving the Diffusion Equation
3.10.4The P1 Approximate Methods in Transport Theory
3.11Further Analysis Methods for One Group
<3.11.1Slab Geometry
3.11.2Cylindrical Geometry
3.11.3Spherical Geometry
3.12Eigenfunction Expansion Methods and Eigenvalue Equations
3.12.1Eigenvalues and Eigenfunctions Problems
3.13Multi-Dimensional Models and Boundary Conditions
3.13.1The Unreflected Reactor Parallelepiped Core
3.13.2The Minimum Volume of the Critical Parallelepiped
3.13.3The Peak to Average Flux Ratio
3.13.4The Finite Height Cylindrical Core
3.14Relating k to the Criticality Condition
3.15Analytical Solution for the Transient Case for Reactor
3.16Criticality
3.17Bare Critical Reactor 1-Group Model
3.18Bare Critical Reactor 1-Group Model, Finite Geometries
3.19Reflected Critical Reactors- 1-Group Model
3.20Infinite Reflector Case
3.21Criticality for General Bare Geometries
3.22Reflected Reactor Geometries
3.23Reactor Criticality Calculations
3.24References
3.25Problems
Chapter Four: Energy Effects in Modeling Neutron Diffusion – Two Group Models
4.0One-Group Diffusion Theory
4.1Two-Group Diffusion Theory
4.2Few Group Analysis
4.2.12-Group Thermal Reactor Equations
4.2.22-Group Fast Reactor Equations
4.3Transverse Buckling Approximation
4.4Consistent Diffusion Theory Boundary Conditions
4.5Derivation of the One-Dimensional Multi-Group PN Equations
4.6Multi-Group Diffusion Equations - Solution Approach
4.6.1Infinite Medium for Group Collapse
4.6.2Zero-Dimensional Spectrum for Group Collapse
4.6.3Group Collapsing
4.6.4Group Collapse
4.7References
4.8Problems
Chapter Five: Numerical Methods in Modeling Neutron Diffusion
5.0Introduction
5.1Problem(s) Solved
5.1.1Transport Equation
5.1.2Angle Discretization
5.1.3Energy Discretization
5.1.4Spatial Discretization
5.1.5Matrix Formulation
5.2Solution Strategy
5.2.1Types of Outer Iterations
5.2.2Inhomogeneous Source (No Fission)
5.2.3Inhomogeneous Source (With Fission)
5.2.4Fission Eigenvalue Calculation
5.2.5Eigenvalue Search Calculation
5.3Middle Iterations
5.4Inner Iterations
5.5Upscatter Iterations
5.6Inhomogeneous Sources
5.7Background Concepts
5.7.1Mixing Tables
5.7.2Cross Section Collapsing
5.8Input Description5.9Output Description
5.10References
5.11Problems
Chapter Six: Slowing Down Theory
6.0Neutron Elastic and Inelastic Scattering for Slowing Down
6.1Derivation of the Energy and Transfer Cross Section
6.1.1Elastic Scattering
6.1.2Inelastic Scattering
6.2Derivation of the Isotropic Flux in an Infinite Hydrogen Moderator
6.3Derivation of the Isotropic Flux in a Moderator Other than Hydrogen A > 1
6.4Summary of Slowing Down Equations
6.5References
6.6Problems
Chapter Seven: Resonance Processing
7.0Difficulties Presented by Resonance Cross Sections
7.1What is Nuclear Resonance -- Compound Nucleus
7.1.1Breit-Wigner Resonance Reaction Cross Sections
7.1.2Resonance and Neutron Cross Section
7.2Doppler Effect and Doppler Broadening of Resonance
7.3Doppler Coefficient in Power Reactors
7.4Infinite Resonance Integrals and Group Cross Section
7.4.1The Flux Calculator Method
7.4.2The Bondarenko Method - The Bondarenko Factor
7.4.3The CENTRM Method
7.5Infinite Resonance Integrals and Group Cross Sections
7.6Dilution Cross Section - Dilution Factor
7.7Resonance Effects
7.8Homogeneous Narrow Resonance Approximation
7.9Homogeneous Wide Resonance Approximation
7.10Heterogeneous Narrow Resonance Approximation
7.11Heterogeneous Wide Resonance Approximation
7.12References
7.13Problems
Chapter Eight: Heterogeneous Reactors and Wigner Seitz Cells
8.0Homogeneous and Heterogeneous Reactors
8.1Spectrum Calculation in Heterogeneous Reactors
8.2Cross Section Self Shielding and Wigner-Seitz Cells
8.3References
8.4Problems
Chapter Nine: Thermal Spectra and Thermal Cross Sections
<9.0Coupling to Higher Energy Sources9.1Chemical Binding and Scattering Kernels
9.1.1Scattering Materials
9.1.2Thermal Cross Section Average
9.2Derivation of the Maxwell-Boltzmann Spectrum
9.3References
9.4Problems
Chapter Ten: Perturbation Theory for Reactor Neutronics
10.0Perturbation Theory
10.1Zero Dimensional Methods
10.2Spatial Method (1 Group)
10.3References
10.4Problems
Chapter Eleven: Reactor Kinetics and Point Kinetics
11.0Time Dependent Diffusion Equation
11.1Derivation of Exact Point Kinetics Equations (EPKE)
11.2The Point Kinetics Equations
11.3Dynamic versus Static Reactivity
11.4Calculating the Time Dependent Shape Function
11.5Point Kinetics Approximations
11.5.1Level of Approximation to the Point Kinetics Equations
11.6Adiabatic Approximation
11.7Adiabatic Approximation with Pre-Computed Shape Functions
11.8Quasi-Static Approximation
11.9Zero Dimensional Reactors
11.10References
11.11Problems
Chapter Twelve: Reactor Dynamics
12.0Background on Nuclear Reactor
12.1Neutron Multiplication
12.2Simple Feedbacks
12.3Multiple Time Constant Feedbacks
12.4Fuchs-Nordheim models
12.5References
12.6Problems
Chapter Thirteen: Reactor Stability
13.0Frequency Response
13.1Nyquist Plots
13.2Non-Linear Stability
13.3References
13.4Problems
Chapter Fourteen: Numerical Modeling for Time Dependent Problems
14.0Fast Breeder Reactor History and Status
14.1The Concept of Stiffness
14.2The Quasi-Static Method
14.3Bethe-Tait Models
14.4References
14.5Problems
Chapter Fifteen: Fission Product Buildup and Decay
15.0Background Introduction
15.1Nuclear Fission and the Fission Process
15.2Radioactivity and Decay of Fission Product
15.3Poisons Produced by Fission
15.4References
15.5Problems
Chapter Sixteen: Fuel Burnup and Fuel Management
16.0The World’s Energy Resources
16.1Today’s Global Energy Market
16.2Fuel Utilization and Fuel Burnup
16.3Fuel Reprocessing
16.3.1PUREX Process
16.3.2Transuranium Elements
16.3.3Vitrification
16.4Fuel Management for Nuclear Reactors
16.5Nuclear Fuel Cycle
16.6Store and Transport High Burnup Fuel
16.7Nuclear Reactors for Power Production
16.8Future Nuclear Power Plants Systems
16.9Next Generation of Nuclear Power Reactors for Power Production
16.10References
16.11ProblemsAppendix A: Laplace Transforms
A-1Definition of Laplace Transform
A-2Basic Transforms
A-3Fundamental Properties
A-4Inversion by Complex Variable Residue Theorem
Appendix B: Transfer Functions and Bode Plots
B-1Transfer Functions
B-2Sample Transforms
B-3Fourier Transforms
B-4Transfer Functions
B-4Feedback and Control
B-5Graphical Representation (Bode and Nyquist Diagram)
B-6Root Locus Construction Rules
B-7References
INDEX
Notă biografică
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.
Textul de pe ultima copertă
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.
Caracteristici
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
Applies the described principles to nuclear reactor power systems
Assumes no previous knowledge of nuclear physics or engineering