مقایسهٔ روش‌ها

	محاسبه انتقال نوترون	حمل و نقل نوترون و ذرات به روش مونت کارلو
حوزه	فیزیک هسته‌ای	فیزیک هسته‌ای
خانواده	Process / pipeline	Process / pipeline
سال پیدایش	1942	1949
پدیدآور	Enrico Fermi, Leslie Szilard	Nicholas Metropolis, Stanislaw Ulam
نوع	computational simulation pipeline	probabilistic computational method
منبع بنیادین	Duderstadt, J. J., & Hamilton, L. J. (1976). Nuclear Reactor Analysis. John Wiley & Sons.	Metropolis, N., & Ulam, S. (1949). The Monte Carlo Method. Journal of the American Statistical Association, 44(247), 335–341.
نام‌های دیگر	neutron diffusion, neutron migration, transport equation solution	Monte Carlo simulation, stochastic transport, particle history method
مرتبط	5	5
خلاصه	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.	Monte Carlo neutron and particle transport is a stochastic simulation method that tracks individual particle histories through matter, developed by Metropolis and Ulam in 1949 during the Manhattan Project. By sampling random numbers to determine collision locations, energy transfers, and scattering angles, it produces unbiased estimates of reaction rates, flux distributions, and detector responses without discretizing angle or energy variables.