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Neutron transportNeutron transport refers to determination of the neutron flux, observed from an efflux of neutrons from a neutron source. In relation to this, beams of free neutrons can be obtained via extraction from neutron sources. Additional recommended knowledge
BackgroundNeutron sourcesThe source can be considered as a volume filled with a very low pressure neutron gas (density up to 10^{24} n/m³), which has an energy distribution corresponding to the temperature of the volume. By inserting a beam tube into the volume, a current of neutrons can be observed flowing out through the tube, much like water flowing out from a tube that is stuck into a volume of wet sand. The current density (in n/m²s) depends on the solid angle formed by the aperture and the length of the tube, and also on the reflecting properties of the inner wall of the tube. Neutrons are indeed reflected at a wall, but usually at an extremely low rate, which is a function of the wall's nature and surface quality, the neutron energy, and the angle of incidence. Characteristics of neutron beamsThe moving neutron has a kinetic energy E = ½·m·v² (m = mass, v = velocity). The SI unit for energy is the joule (J), but for particles the energy unit of electronvolts (eV) is usually used (1 J = 6.2410^{18} eV). The energy can also be expressed in temperature (unit: kelvins) by the relation E = k·T (k = 1.38·10^{23} J/K). Furthermore, a moving particle can be considered a matter wave with the de Broglie wavelength λ = h/(2k·T·m)^{1/2} (h = 6.6·10^{34} J·s). The corresponding wavelength for the neutron is generally expressed in nanometers (nm) or sometimes in angstroms (Å), where 1 Å = 0.1 nm. A thermal neutron, for instance, with a temperature of 300 K (corresponding to E = 0.025 eV) travels with a speed of 2200 m/s, and has a wavelength of 0.17 nm. Neutron transport calculationsUsually the neutron transport calculations can be divided into two categories: shielding and criticality search. ShieldingIn the case of shielding calculation, there are three important things which must be modeled: neutron source, shield, and detector. Shielding calculation usually yields the neutron flux in the region of space occupied by the detector. Based on the flux in the detector, the shield thickness or source strength are optimized. Criticality searchIn criticality calculation, a fissile or fissionable material is part of the modeled geometry. Neutron source is the part of the modeled geometry which contains the fissile or fissionable material. The spacial dependence of neutron source intensity is proportional to the flux, which dictates the fission reaction density, hence spacial dependence of neutron source is proportional to the fissions induced by the neutrons. The parameter which is the result of criticality calculation is Keffective (K_{eff}). This is a parameter which reflects the time dependence on neutron density in a multilying medium.
In the case of a nuclear reactor, neutron density and power density are synonymous, hence during reactor startup K_{eff} > 1, during reactor operation K_{eff} = 1 and K_{eff} < 1 at reactor shutdown. MethodologyBoth shielding and criticality calculations can be done using deterministic methods such as diffusion theory, or using stochastic methods such as Monte Carlo. Deterministic methods usually involve multigroup approaches while Monte Carlo can work with multigroup and continuous energy crosssection libraries. Multigroup calculations are usually iterative, because the group constants are calculated using fluxenergy profiles, which are determined as the result of the neutron transport calculation.
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This article is licensed under the GNU Free Documentation License. It uses material from the Wikipedia article "Neutron_transport". A list of authors is available in Wikipedia. 