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Ali Denisov
Ali Denisov

MCNP5 Features and Capabilities: A Comprehensive Overview

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Point-wise cross section data are typically used, although group-wise data also are available. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(α,β) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields.

The gamma-ray shielding effectiveness of the lead bismuth germanoborate glasses has been studied. The mass attenuation coefficients (μ/ρ) of the selected glasses have been obtained through both XCOM program and MCNP5 simulation code. The (μ/ρ) values calculated in both methods are found to be in good agreement and these values are used to calculate effective atomic number, mean free path, half-value layer and energy exposure buildup factors. The shielding effectiveness of these samples has been compared with that of window glasses and some standard shielding concretes. The lower values of mean free path point to the fact that the selected glasses are efficient gamma shields.

count of the low-energy spectrum decreases with the increase of the target thickness [7] . This is because when the thickness of the target material is thinner, the probability of low-energy photons generated by Compton scattering through the target material increases, resulting in a higher peak at the low-energy end of the X-ray spectrum. When a certain thickness is reached, due to the thickness of the shielding material much greater than the mean free path of low-energy photons, most of these low-energy Compton scattered photons are absorbed.

Figure 14 shows the dose distribution and 2D-projected energy deposition density of 239,240Pu from alpha particles. Two approaches were considered in order to account for aerosol self-shielding: (a) point source approach with volume-weighted average exit energy (b) volume source approach which directly simulated alpha shielding in UO2 in the same calculation. In the first approach, the dose per decay was calculated by first running MCNP shielding calculations (Fig. 14a) to determine the exit energy as a function of sphere diameter per decay. These simulations considered uniformly distributed plutonium source in UO2 spheres with different sizes at room temperature. The average exit energy decreases as a function of the diameter of the carrier aerosol. After that, the volume-weighted average exit energy was calculated for the deposited aerosol particles (2-µm AMAD and 1.8 GSD for 239,240Pu), as shown in Fig. 14b. Figure 14c shows the organ dose per decay from 234,240Pu alpha particles by considering all deposited particles from CFPD predictions. As expected, elevated doses were noted in the trachea and lung only since the mean free path for alphas in tissue is less than 50 microns. In this approach, the sources, however, have been defined as point sources with shielding accounted for by weighting the position of each source particle by the product of the exit energy for its diameter and its volume. The volume-weighted average exit energy was used for these simulations.


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