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	Neutrontransportberäkning	Reaktorkinetik
Ämnesområde	Kärnfysik	Kärnfysik
Familj	Process / pipeline	Process / pipeline
Ursprungsår	1942	1942
Upphovsperson	Enrico Fermi, Leslie Szilard	Enrico Fermi, George Westinghouse
Typ	computational simulation pipeline	dynamic systems analysis
Ursprungskälla	Duderstadt, J. J., & Hamilton, L. J. (1976). Nuclear Reactor Analysis. John Wiley & Sons.	Lamarsh, J. R. (1983). Introduction to Nuclear Engineering (2nd ed.). Addison-Wesley.
Alias	neutron diffusion, neutron migration, transport equation solution	neutron kinetics, power transient modeling, reactor control analysis
Närliggande	5	5
Sammanfattning	Neutron transport calculation is a computational method for determining the distribution and behavior of neutrons in a nuclear medium, developed during the Manhattan Project in the 1940s. It solves the Boltzmann transport equation to predict neutron flux, energy spectra, and reaction rates essential for reactor design and shielding analysis.	Reactor kinetics is the study of neutron population dynamics in a reactor core, originating from Fermi's first controlled chain reaction in 1942. It models power changes in response to control rod movements, temperature feedback, and accidental transients using coupled differential equations accounting for prompt and delayed neutrons, to ensure safe operation, predict transient behavior, and design control systems.