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Preface; About This Document; Acknowledgments; Contents; About the Author; Chapter 1: Neutron Physics Background; 1.1 Nuclei: Sizes, Composition, and Binding Energies; Pion; Atom Density; 1.2 Decay of a Nucleus; 1.3 Distribution of Nuclides and Nuclear Fission/Nuclear Fusion; Gamma Decay; 1.4 Neutron-Nucleus Interaction; 1.4.1 Nuclear Reaction Rates and Neutron Cross Sections; 1.4.2 Effects of Temperature on Cross Section; 1.4.3 Nuclear Cross-Sectional Processing Codes; 1.4.4 Energy Dependence of Neutron Cross Sections; 1.4.5 Types of Interactions; 1.5 Mean Free Path
1.6 Nuclear Cross Section and Neutron Flux Summary1.7 Fission; 1.8 Fission Spectra; 1.9 The Nuclear Fuel; 1.9.1 Fertile Material; 1.10 Liquid Drop Model of a Nucleus; 1.11 Summary of Fission Process; 1.12 Reactor Power Calculation; 1.13 Relationship Between Neutron Flux and Reactor Power; Problems; References; Chapter 2: Modeling Neutron Transport and Interactions; 2.1 Transport Equations; 2.2 Reaction Rates; 2.3 Reactor Power Calculation; 2.4 Relationship Between Neutron Flux and Reactor Power; 2.5 Neutron Slowing Down and Thermalization; 2.6 Macroscopic Slowing Down Power
2.7 Moderate Ratio2.8 Integrodifferential Equation (Maxwell-Boltzmann Equation); 2.9 Integral Equation; 2.10 Multigroup Diffusion Theory; 2.11 The Multigroup Equations; 2.12 Generating the Coefficients; 2.13 Simplifications; 2.14 Nuclear Criticality Concepts; 2.15 Criticality Calculation; 2.16 The Multiplication Factor and a Formal Calculation of Criticality; 2.17 Fast Fission Factor epsi Definition; 2.18 Resonance Escape Probability p; 2.19 Group Collapsing; 2.19.1 Multigroup Collapsing to One Group; 2.19.2 Multigroup Collapsing to Two Group; 2.19.3 Two-Group Criticality
2.20 The Infinite Reactor2.21 Finite Reactor; 2.22 Time Dependence; 2.23 Thermal Utilization Factor f; Problems; References; Chapter 3: Spatial Effects in Modeling Neutron Diffusion: One-Group Models; 3.1 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 Fickś Law; 3.6 Problem Classification and Neutron Distribution; 3.7 Neutron Slowing Down; Isotropic Source and Scattering [4]
3.8 Neutron Diffusion Concept3.9 The One-Group Model and One-Dimensional Analysis; 3.9.1 Boundary Conditions for the Steady-State Diffusion Equation; 3.9.2 Boundary Conditions: Consistent and Approximate; 3.9.3 An Approximate Method for Solving the Diffusion Equation; 3.9.4 The P1 Approximate Methods in Transport Theory; 3.10 Further Analysis Methods for One Group; 3.10.1 Slab Geometry; 3.10.2 Cylindrical Geometry; 3.10.3 Spherical Geometry; 3.11 Eigenfunction Expansion Methods and Eigenvalue Equations; 3.11.1 Eigenvalue and Eigenfunction Problems
1.6 Nuclear Cross Section and Neutron Flux Summary1.7 Fission; 1.8 Fission Spectra; 1.9 The Nuclear Fuel; 1.9.1 Fertile Material; 1.10 Liquid Drop Model of a Nucleus; 1.11 Summary of Fission Process; 1.12 Reactor Power Calculation; 1.13 Relationship Between Neutron Flux and Reactor Power; Problems; References; Chapter 2: Modeling Neutron Transport and Interactions; 2.1 Transport Equations; 2.2 Reaction Rates; 2.3 Reactor Power Calculation; 2.4 Relationship Between Neutron Flux and Reactor Power; 2.5 Neutron Slowing Down and Thermalization; 2.6 Macroscopic Slowing Down Power
2.7 Moderate Ratio2.8 Integrodifferential Equation (Maxwell-Boltzmann Equation); 2.9 Integral Equation; 2.10 Multigroup Diffusion Theory; 2.11 The Multigroup Equations; 2.12 Generating the Coefficients; 2.13 Simplifications; 2.14 Nuclear Criticality Concepts; 2.15 Criticality Calculation; 2.16 The Multiplication Factor and a Formal Calculation of Criticality; 2.17 Fast Fission Factor epsi Definition; 2.18 Resonance Escape Probability p; 2.19 Group Collapsing; 2.19.1 Multigroup Collapsing to One Group; 2.19.2 Multigroup Collapsing to Two Group; 2.19.3 Two-Group Criticality
2.20 The Infinite Reactor2.21 Finite Reactor; 2.22 Time Dependence; 2.23 Thermal Utilization Factor f; Problems; References; Chapter 3: Spatial Effects in Modeling Neutron Diffusion: One-Group Models; 3.1 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 Fickś Law; 3.6 Problem Classification and Neutron Distribution; 3.7 Neutron Slowing Down; Isotropic Source and Scattering [4]
3.8 Neutron Diffusion Concept3.9 The One-Group Model and One-Dimensional Analysis; 3.9.1 Boundary Conditions for the Steady-State Diffusion Equation; 3.9.2 Boundary Conditions: Consistent and Approximate; 3.9.3 An Approximate Method for Solving the Diffusion Equation; 3.9.4 The P1 Approximate Methods in Transport Theory; 3.10 Further Analysis Methods for One Group; 3.10.1 Slab Geometry; 3.10.2 Cylindrical Geometry; 3.10.3 Spherical Geometry; 3.11 Eigenfunction Expansion Methods and Eigenvalue Equations; 3.11.1 Eigenvalue and Eigenfunction Problems