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NE410/510 - Lecture 8: The P1 Approximation and the Neutron Diffusion Equation
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In this lecture we introduce a series of approximations to convert the moments of the Boltzmann Transport Equation into the Neutron Diffusion Equation.
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NE410/510 - Lecture 8: The P1 Approximation and the Neutron Diffusion Equation
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NE410/510 - Lecture 9: The Critical Condition and Vacuum Boundary Conditions
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NE410/510 - Lecture 18: Nuclear Reactor Kinetics
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NE410/510 - Lecture 11: The Diffusion Equation for Reflected Geometries
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NE410/510 - Lecture 5: The Four- and Six-Factor Formulas
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NE410/510 - Lecture 10: The Diffusion Equation in Multiple Dimensions
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NE410/510 - Lecture 7: The Moments of the Boltzmann Transport Equation
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NE410/510 - Lecture 14: Generating Multigroup Cross Sections and Neutron Flux Spectra
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NE410/510 - Lecture 19: A Reactor Physics Explanation of the Chernobyl Disaster
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NE410/510 - Lecture 6: The Boltzmann Transport Equation
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NE410/510 - Lecture 2: Neutron Cross Sections
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NE410/510 - Lecture 12: Finite Difference Diffusion Methods
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NE410/510 - Lecture 3: The Physics of Nuclear Fission
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NE410/510 - Lecture 13: Multigroup Diffusion Equations
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NE410/510 - Lecture 15: Lethargy and Introduction to Slowing-Down Theory
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NE410/510 - Lecture 4: Elastic Scattering Kinematics
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Nuclear Reactor Kinetics
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Lecture 8 - Neutron flux; microscopic/macroscopic cross sections; reaction rate equation
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NE499/515 - Lecture 1: Introduction to Nuclear Criticality Safety
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Reactor Physics 8
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Lecture 7 Non-Leakage Probability
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Lecture 8 Four-Factor Formular
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NE499/515 - Lecture 8: Nuclear Reactor Kinetics and the Tokai-Mura Criticality Accident (CA-6)
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REACTOR PHYSICS - How to Control a Nuclear Reactor
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