Neutronic Analysis For Nuclear Reactor Systems


169,68 €*

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Bahman Zohuri
861 g
235x155x31 mm

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.
Covers the fundamentals of neutronic analysis for nuclear reactor systems in order to help readers understand nuclear reactor theory
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

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