Stimulated perturbation on the neutron flux distribution in the mutually-dependent source-to-absorber geometry

Abstract

The complexity of the neutron transport phenomenon throws its shadows on every physical system wherever neutron is produced or absorbed. The Monte Carlo N-Particle Transport Code (MCNP) was used to investigate the flux perturbations in the neutron field caused by an absorber. The geometry of the present experiment was designed to reach a simulation of an isotopic neutron field. The neutron source was a 241AmBe with the production physics of neutrons is dependent only on alpha- beryllium interaction and is independent of what happened to the neutron after it was generated. The geometries have been designed to get a volume of uniform neutron densities within a spherical volume of radius 15 cm in every neutron energy group up to 10 MeV. Absorbers of different dimensions were placed within the volume to investigate the field perturbation. Different neutron absorbers were used to correlate the phenomenon to the integral cross section of the absorber. Flux density inside and outside the absorber samples was determined, while the spatial neutron flux distribution produced by the AmBe source without absorber was taken as a reference. This study displayed that absorbers of various dimensions perturb the neutron field in a way that is dependent on the absorption and scattering cross-sections, particularly in the neutron resonance region. Unlike the simple picture of reducing the number density of neutrons, the perturbation was found to influence the moderation of neutrons in the medium, significantly above 1 MeV.

Mathematical formulae for neutron self-shielding properties of media in an isotropic neutron field

“…  the higher order terms were introduced by Blaauw [18] for extended neutron velocity distributions-denoted here . Remembering that the G (energy domain) is a neutron energy-specific parameter, any perturbation of the neutron energy distribution shall affect the experimental results as discussed in earlier work [15].…”

“…Due to symmetry operations in the diffusion equation, we shall take the condition that if there were a symmetry making one or more of these virtual lengths undeterminable, it should take an infinity value. The behavior of the absorption and scattering transport are strongly dependent on each other and depend on the neutron energy [15]. However, a single group that covers the entire energy range of neutrons (both thermal and epi-thermal) can be used to obtain a meaningful value of the average neutron-chord length.…”

“…Furthermore, the neutron's energy spectrum is varied, but ideally suited to research using the ideal Maxwellian distribution at room temperature, while the other distributions must be altered to match the reference nuclear reaction data [3,7,13]. Apart from that, neutrons are deeply employed in two significant geometries, including but not limited to: (1) beam geometry, where the neutron currents are assumed to travel in one direction, and (2) field geometry, where neutrons impact the sample from all directions, presuming the field is isotropic and has equal flux densities from all directions [14][15][16]. There is an important functional difference between these two geometries, i.e., the effect on the neutron flux itself.…”

“…In the case of field geometry, the net neutron current essentially disappears, while the fluence (or flux) becomes the observable quantity. There is an interplay between neutron absorption in the sample and the overall neutron flux [15].…”