Neutronic Analysis For Nuclear Reactor Systems
Published on 8. December 2016
Book
Hardback
XXII, 551 pages
978-3-319-42962-5 (ISBN)
Description
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.
More details
Edition
1st ed. 2017
Language
English
Place of publication
Cham
Switzerland
Publishing group
Springer International Publishing
Target group
Professional and scholarly
Illustrations
90 s/w Abbildungen, 58 farbige Abbildungen, 137 farbige Tabellen
137 Tables, color; 58 Illustrations, color; 90 Illustrations, black and white; XXII, 551 p. 148 illus., 58 illus. in color.
Dimensions
Height: 23.5 cm
Width: 15.5 cm
Weight
9812 gr
ISBN-13
978-3-319-42962-5 (9783319429625)
DOI
10.1007/978-3-319-42964-9
Schweitzer Classification
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Bahman Zohuri
Neutronic Analysis For Nuclear Reactor Systems
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02/2019
2nd Edition
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Bahman Zohuri
Neutronic Analysis For Nuclear Reactor Systems
Book
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Bahman Zohuri
Neutronic Analysis For Nuclear Reactor Systems
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Person
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.
Content
Table of ContentsAbout the Authors PrefaceAcknowledgmentChapter One: Neutron Physics Background1.0 Nuclei - Sizes, Composition, and Binding Energies1.1 Decay of a Nucleus1.2 Distribution of Nuclides and Nuclear Fission/Nuclear Fusion1.3 Neutron-Nucleus Interaction1.3.1 Nuclear Reactions Rates and Neutron Cross Sections1.3.2 Effects of Temperature on Cross Section1.3.3 Nuclear Cross Section Processing Codes1.3.4 Energy Dependence of Neutron Cross Sections1.3.5 Types of Interactions1.4 Mean Free Path1.5 Nuclear Cross Section and Neutron Flux Summary1.6 Fission1.7 Fission Spectra1.8 The Nuclear Fuel1.6.1 Fertile Material1.9 Liquid Drop Model of a Nucleus1.10 Summary of Fission Process1.11 Reactor Power Calculation1.12 Relationship between Neutron Flux and Reactor Power1.13 References1.14 ProblemsChapter Two: Modeling Neutron Transport and Interactions2.0 Transport Equations2.1 Reaction Rates2.2 Reactor Power Calculation2.3 Relationship between Neutron Flux and Reactor Power2.4 Neutron Slowing Down and Thermalization2.5 Macroscopic Slowing Down Power2.6 Moderate Ratio2.7 Integro-Differential Equation (Maxwell-Boltzmann Equation)2.8 Integral Equation2.9 Multigroup Diffusion Theory2.10 The Multigroup Equations2.11 Generating the Coefficients2.12 Simplifications2.13 Nuclear Criticality Concepts2.14 Criticality Calculation2.15 The Multiplication Factor and a Formal Calculation of Criticality2.16 Fast Fission Factor Definition2.17 Resonance Escape Probability2.18 Group Collapsing2.18.1 Multigroup Collapsing to One Group2.18.2 Multigroup Collapsing to Two Group2.18.3 Two Group Criticality2.19 The Infinite Reactor2.20 Finite Reactor2.21 Time Dependence2.22 Thermal Utilization Factor2.23 References2.24 ProblemsChapter Three: Spatial Effects in Modeling Neutron Diffusion - One Group Models3.0 Nuclear Reactor Calculations3.1.1 Neutron Spectrum3.2 Control Rods in Reactors3.2.1 Lattice Calculation Analysis3.3 An Introduction to Neutron Transport Equation3.4 Neutron Current Density Concept in General3.5 Neutron Current Density and Fick's Law3.6 Problem Classification and Neutron Distribution3.7 Neutron Slowing Down3.8 Neutron Diffusion Concept3.9 The One Group Model and One Dimensional Analysis3.10.1 Boundary Conditions for the Steady-State Diffusion Equation3.10.2 Boundary Conditions - Consistent and Approximate3.10.3 An Approximate Methods for Solving the Diffusion Equation3.10.4 The P1 Approximate Methods in Transport Theory3.11 Further Analysis Methods for One Group<3.11.1 Slab Geometry3.11.2 Cylindrical Geometry3.11.3 Spherical Geometry3.12 Eigenfunction Expansion Methods and Eigenvalue Equations3.12.1 Eigenvalues and Eigenfunctions Problems3.13 Multi-Dimensional Models and Boundary Conditions3.13.1 The Unreflected Reactor Parallelepiped Core3.13.2 The Minimum Volume of the Critical Parallelepiped3.13.3 The Peak to Average Flux Ratio3.13.4 The Finite Height Cylindrical Core3.14 Relating k to the Criticality Condition3.15 Analytical Solution for the Transient Case for Reactor3.16 Criticality3.17 Bare Critical Reactor 1-Group Model3.18 Bare Critical Reactor 1-Group Model, Finite Geometries3.19 Reflected Critical Reactors- 1-Group Model3.20 Infinite Reflector Case3.21 Criticality for General Bare Geometries3.22 Reflected Reactor Geometries3.23 Reactor Criticality Calculations3.24 References3.25 ProblemsChapter Four: Energy Effects in Modeling Neutron Diffusion - Two Group Models4.0 One-Group Diffusion Theory4.1 Two-Group Diffusion Theory4.2 Few Group Analysis4.2.1 2-Group Thermal Reactor Equations4.2.2 2-Group Fast Reactor Equations4.3 Transverse Buckling Approximation4.4 Consistent Diffusion Theory Boundary Conditions4.5 Derivation of the One-Dimensional Multi-Group PN Equations4.6 Multi-Group Diffusion Equations - Solution Approach4.6.1 Infinite Medium for Group Collapse4.6.2 Zero-Dimensional Spectrum for Group Collapse4.6.3 Group Collapsing4.6.4 Group Collapse4.7 References4.8 ProblemsChapter Five: Numerical Methods in Modeling Neutron Diffusion5.0 Introduction5.1 Problem(s) Solved5.1.1 Transport Equation5.1.2 Angle Discretization5.1.3 Energy Discretization5.1.4 Spatial Discretization5.1.5 Matrix Formulation5.2 Solution Strategy5.2.1 Types of Outer Iterations5.2.2 Inhomogeneous Source (No Fission)5.2.3 Inhomogeneous Source (With Fission)5.2.4 Fission Eigenvalue Calculation5.2.5 Eigenvalue Search Calculation5.3 Middle Iterations5.4 Inner Iterations5.5 Upscatter Iterations5.6 Inhomogeneous Sources5.7 Background Concepts5.7.1 Mixing Tables5.7.2 Cross Section Collapsing5.8 Input Description5.9 Output Description5.10 References5.11 ProblemsChapter Six: Slowing Down Theory6.0 Neutron Elastic and Inelastic Scattering for Slowing Down6.1 Derivation of the Energy and Transfer Cross Section6.1.1 Elastic Scattering6.1.2 Inelastic Scattering6.2 Derivation of the Isotropic Flux in an Infinite Hydrogen Moderator6.3 Derivation of the Isotropic Flux in a Moderator Other than Hydrogen A > 16.4 Summary of Slowing Down Equations6.5 References6.6 ProblemsChapter Seven: Resonance Processing7.0 Difficulties Presented by Resonance Cross Sections7.1 What is Nuclear Resonance -- Compound Nucleus7.1.1 Breit-Wigner Resonance Reaction Cross Sections7.1.2 Resonance and Neutron Cross Section7.2 Doppler Effect and Doppler Broadening of Resonance7.3 Doppler Coefficient in Power Reactors7.4 Infinite Resonance Integrals and Group Cross Section7.4.1 The Flux Calculator Method7.4.2 The Bondarenko Method - The Bondarenko Factor7.4.3 The CENTRM Method7.5 Infinite Resonance Integrals and Group Cross Sections7.6 Dilution Cross Section - Dilution Factor7.7 Resonance Effects7.8 Homogeneous Narrow Resonance Approximation7.9 Homogeneous Wide Resonance Approximation7.10 Heterogeneous Narrow Resonance Approximation7.11 Heterogeneous Wide Resonance Approximation7.12 References7.13 ProblemsChapter Eight: Heterogeneous Reactors and Wigner Seitz Cells8.0 Homogeneous and Heterogeneous Reactors8.1 Spectrum Calculation in Heterogeneous Reactors8.2 Cross Section Self Shielding and Wigner-Seitz Cells8.3 References8.4 ProblemsChapter Nine: Thermal Spectra and Thermal Cross Sections<9.0 Coupling to Higher Energy Sources9.1 Chemical Binding and Scattering Kernels9.1.1 Scattering Materials9.1.2 Thermal Cross Section Average9.2 Derivation of the Maxwell-Boltzmann Spectrum9.3 References9.4 ProblemsChapter Ten: Perturbation Theory for Reactor Neutronics10.0 Perturbation Theory10.1 Zero Dimensional Methods10.2 Spatial Method (1 Group)10.3 References10.4 ProblemsChapter Eleven: Reactor Kinetics and Point Kinetics11.0 Time Dependent Diffusion Equation11.1 Derivation of Exact Point Kinetics Equations (EPKE)11.2 The Point Kinetics Equations11.3 Dynamic versus Static Reactivity11.4 Calculating the Time Dependent Shape Function11.5 Point Kinetics Approximations11.5.1 Level of Approximation to the Point Kinetics Equations11.6 Adiabatic Approximation11.7 Adiabatic Approximation with Pre-Computed Shape Functions11.8 Quasi-Static Approximation11.9 Zero Dimensional Reactors11.10 References11.11 ProblemsChapter Twelve: Reactor Dynamics12.0 Background on Nuclear Reactor12.1 Neutron Multiplication12.2 Simple Feedbacks12.3 Multiple Time Constant Feedbacks12.4 Fuchs-Nordheim models12.5 References12.6 ProblemsChapter Thirteen: Reactor Stability13.0 Frequency Response13.1 Nyquist Plots13.2 Non-Linear Stability13.3 References13.4 ProblemsChapter Fourteen: Numerical Modeling for Time Dependent Problems14.0 Fast Breeder Reactor History and Status14.1 The Concept of Stiffness14.2 The Quasi-Static Method14.3 Bethe-Tait Models14.4 References14.5 ProblemsChapter Fifteen: Fission Product Buildup and Decay15.0 Background Introduction15.1 Nuclear Fission and the Fission Process15.2 Radioactivity and Decay of Fission Product15.3 Poisons Produced by Fission15.4 References15.5 ProblemsChapter Sixteen: Fuel Burnup and Fuel Management16.0 The World's Energy Resources16.1 Today's Global Energy Market16.2 Fuel Utilization and Fuel Burnup16.3 Fuel Reprocessing16.3.1 PUREX Process16.3.2 Transuranium Elements16.3.3 Vitrification16.4 Fuel Management for Nuclear Reactors16.5 Nuclear Fuel Cycle16.6 Store and Transport High Burnup Fuel16.7 Nuclear Reactors for Power Production16.8 Future Nuclear Power Plants Systems16.9 Next Generation of Nuclear Power Reactors for Power Production16.10 References16.11 ProblemsAppendix A: Laplace TransformsA-1 Definition of Laplace TransformA-2 Basic TransformsA-3 Fundamental PropertiesA-4 Inversion by Complex Variable Residue TheoremAppendix B: Transfer Functions and Bode PlotsB-1 Transfer FunctionsB-2 Sample TransformsB-3 Fourier TransformsB-4 Transfer FunctionsB-4 Feedback and ControlB-5 Graphical Representation (Bode and Nyquist Diagram)B-6 Root Locus Construction RulesB-7 ReferencesINDEX