A Comparison of Conventional Empirical Formula and MCNPX Code in the Estimations of Photon and Neutron Skyshine Rates for an 18MV Radiotherapy Bunker

Purpose: A physical phenomenon, scattering the radiation by the atmosphere above the room to the points at ground level around the linac treatment room is known as skyshine radiation. This study aimed to estimate photon and neutron skyshine from a linac in a high-energy radiation therapy facility. Materials and Methods: The empirical method of NCRP report 151 and MC simulations were employed to estimate skyshine radiation dose from the 18MV linac photon beam. A linac and its bunker were modeled and skyshine dose equivalent from photons and secondary neutrons were derived and compared in the control room, corridor, sidewalk and, parking. Results: The photon skyshine dose rates calculations by the MC method varied from 0.43 µSv/h at the sidewalk to 6.2 µSv/h at the control room. The ratios of NCRP to MCNP calculations varied from 3.58 for the corridor to 16.14 for the control room. For the neutron skyshine dose rate at distances shorter than 20m, it was found to be 10.4 nSv/h and the ratios of the NCRP to MCNP were 1.26 at the control room and 3.34 at the sidewalk. Conclusion: It was concluded that the empirical method overestimates photon and neutron skyshine dose rates in comparison to the MCNPX code. The refinement of the proposed empirical method of NCRP 151 and application of MC methods are strongly suggested for more reliable calculations of skyshine radiations.

Calculation of response functions for cylindrical nested neutron spectrometer

In a recent work, a new neutron spectrometer, namely Cylindrical Nested Neutron Spectrometer (CNNS). It works under the same principles as a Bonner Sphere Spectrometer (BSS), except that different amounts of moderator around a thermal neutron detector are configured by adding or removing cylindrical shells. The CNNS consists of a 4mm x 4mm 6LiI(Eu) scintillator crystal and nested cylindrical polyethylenemoderators. The objective of this paper is describing the use of MCNPX code for determining a optimal ratio between height and diameter of the moderators in order to remain isotropic angular response to neutrons like BSS and determining of response functions for moderators of different diameters at 104 energy points from 0.001 eV to 19.95 MeV.

Gamma Shielding Experiment Simulation utilizing MCNPX Code

Experimental investigation requires materials, radiation sources, and test arrangements with a high monetary financial plan. Furthermore, radiation exposure involves people during the experiment. On the contrary, the simulation technique for examining radiation interactions is radio-logically safer, less timeconsuming, cost-effective, and applicable for all desired radiation sources. Through 48.86 mCi 662 keV Caesium-137 gamma-ray source; shielding experiment as well as simulation of it with MCNPX were performed for three shielding materials Lead, Copper, and Aluminum. These materials were placed in front of the gamma source and the emergent radiation was counted in a Geiger- Muller detector to understand the attenuation quality of these materials to each other. These courses of action were simulated utilizing the MCNPX code version 2.7.0 and the results likewise gave and looked at that of the experiment. There are huge similarities of shielding behavior between MCNPX simulation and experiments for the three absorbing materials. The modeled geometry of this MCNPX simulation could be used for future approaches of new designs and structures of radiation shielding, especially where no analogous experimental data exist Journal of Engineering Science 12(2), 2021, 11-21

Improvement of the Point Spread Function by Collimators’ New Configuration in Nuclear Medicine

Objective: The collimators in which the various geometrical configurations have been suggested to optimize the sensitivity and resolution have a key role in acquiring the qualified images in nuclear medicine towards a better recognition of some diseases. Methods: In this study, a new configuration as a geometrical combination of the conical, cylindrical and spherical (CCS) volumes for parallel hole collimators which is assessed by using the volumetric-parametric method has been introduced to improve point spread function (PSF) being the collimators response on the radioactive point source. It has been simulated by the MCNPX code at the various energies values of the point source along with the traditional collimator in which included the cylindrical volume only. Results: The PSF will transmogrify from a delta function to a distribution which can correlate with a Gaussian distribution, while the scattered gamma rays were increased. The simulation results have indicated that the PSF in the CCS configuration is narrower than that of the cylindrical one at all the energies, leading the improvement of the resolution. Also, the theoretical results are agreement with the simulated ones. The more the energy value of the source, the more broaden the PSF will be due the more penetration strength. The narrower the PSF, the better the qualified image will be. Conclusion: This method may be employed to determine the accurate attenuation coefficient of absorbers as well.

The effect of control rods on the reactivity and flux distribution of BWR 4 bundle using MCNPX Code

This paper has three main objectives related to the neutronic and burnup analysis of the BWR (Boiling Water Reactor) Four-Lattice. The first objective is to provide partial validation of the MCNPX code for this lattice by comparing its results with Scale-5.1 results. Validation of the MCNPX to calculate effective multiplication factor and reactivity rod worth for the F-Lattice is provided. This is carried out in case of instantly removing the control blade and replacing it with a graphite moderator. Moreover, spatial neutron flux distributions using F-mesh card over the bundle and the control blade are investigated at inserting and withdrawing the B4C. The second objective is to perform parametric design studies of the F-Lattice. Areas of particular interest are the effect of increased or decreased blade width on the neutron flux throughout the bundle. It is found that the presence of carbon in the control blade at withdrawing the B4C makes the reactor supercritical, (K-eff = 1.22206). On the other hand, the use of B4C blade presents (K-eff = 0.93521). Consequently, the reactivity of 10% B4C thinner case is higher that of 10% B4C thicker. The simulation also showed that the B4C blade had an effective role in decreasing the thermal flux at the periphery of the bundle. This is contrast to the effect of carbon that moderates fast to thermal neutrons. The third part of this work aims at studying the burnup calculations using MCNPX code for 30 days burn with 1 day time step then for 20 months burn with 2 week time steps for the lattice. At the end of the work, it is very important to determine the most proper bundle model that achieves a prolonged fuel burn and flatting thermal flux distribution. For reaching this goal, three cases (B4C, 10% thinner of B4C and 10% thicker) are simulated by MCNPX code till 70 GWd/ton. It is found that the B4C and 10% thicker are the appropriate models that can satisfy the safety considerations of the Compact Modular Boiling Water Reactor.

In Silicon Monte Carlo Simulation Trials for Investigation of V2O5 Reinforcement Effect on Ternary Zinc Borate Glasses: Nuclear Radiation Shielding Dynamics

In the current study, promising glass composites based on vanadium pentoxide (V2O5)-doped zinc borate (ZnB) were investigated in terms of their nuclear-radiation-shielding dynamics. The mass and linear attenuation coefficient, half-value layer, mean free path, tenth-value layer, effective atomic number, exposure-buildup factor, and energy-absorption-buildup factor were deeply simulated by using MCNPX code, Phy-X PSD code, and WinXcom to study the validation of ZBV1, ZBV2, ZBV3, and ZBV4 based on (100−x)(0.6ZnO-0.4B2O3)(x)(V2O5) (x = 1, 2, 3, 4 mol%) samples against ionizing radiation. The results showed that attenuation competencies of the studied glasses slightly changed while increasing the V2O5 content from 1 mol% to 4 mol%. The domination of ZnO concentration in the composition compared to B2O3 makes ZnO substitution with V2O5 more dominant, leading to a decrease in density. Since density has a significant role in the attenuation of gamma rays, a negative effect was observed. It can be concluded that the aforementioned substitution can negatively affect the shielding competencies of studied glasses.