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Introduction to Nuclear Power Plants | Introducing fundamentals of nuclear reactor phyiscs & engineering (4.26) | ![]() |
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Two-body collision mechanics | Inducing the relations between Lab frame & CM (5.2) | ![]() |
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Differential cross section, neutron number density | Explainations on differential, integral cross section, and scattering krnel (5.6) | ![]() |
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Angular neutron flux, scalar flux, net current, partial current | Interpreting definitions of neutron flux and current (5.9) | ![]() |
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Reaction rate, effective cross sections | Interpreting means of reaction rate & effective cross section (5.11) | ![]() |
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Numerical integration for Prob. 2.1, 4-factor formula, diffusion equation | 1. Inducing the flux for neutrons obeying the Maxwell-Boltxman distribution 2. Introducing the neutron diffusion equation. (5.13) | ![]() |
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Diffusion coefficient, mean cosine scattering angle, 1-group diff. equation | 1. providing an interpretation of the diffusion coefficient for the current. 2. Discussing limitations and applicability of diffusion theory. 3. Derivation of the one-group diffusion equation.(5.18) | ![]() |
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Boundary conditions for diffusion equation, plane source problem | 1. disscusion of the boundary conditions necessary for the solution of Eq. 4.1. 2. providing source conditons for solutions of the steady-state diffusion equation (5.20) | ![]() |
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Review on neutron flux, mean square distance, finite slab problem | providing solutions of the steady-state diffusion equation according to each source type (5.23) | ![]() |
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Neutron albedo | providing the solution of the steady-state diffusion equation incase of two slabs of finite thickness (5.3) | ![]() |
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P.S. #3, distributed source problem, kernel technique | providing the solution fo the diffusion equation with distributed sources (6.1) | ![]() |
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P.S. #3, multiplying medium, geometric and material buckling | introducing the concept of material and geometric bucklings together with the one-group form of the effective mutiplication factor (6.3) | ![]() |