• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Journal of Radiation Protection and Research
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• v.42 no.3
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• pp.154-157
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• 2017

Background: Shielding properties of compound or mixture is presented in terms of mass attenuation coefficients using Monte Carlo simulation. Mass attenuation coefficients of cement, gypsum and the mixture of gypsum and $PbCO_3$ has been investigated using monte carlo MCNPX. Materials and Methods: The mass attenuation coefficients of cement, gypsum and the mixture of gypsum and $PbCO_3$ were calculated for photon energies 365.5, 661.6, 1,173.2, and 1,332.5 keV energies. Results and Discussion: The simulated values of mass attenuation coefficients were compared avaialable experimental results, theoretical values by XCOM and found good comparability of the results. Conclusion: Standard simulation geometry used in the present investigation would be very useful for various types of sample for shielding and dosimetry applications.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4231-4235
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• 2022

Considering the importance of deuterium in nuclear science including medical and industrial researches such as (BNCT) and nuclear reactors respectively, it is important to study various possible ways in addition to common methods for measuring its concentration. This study is an effort to measure deuterium concentration using PGNAA. The main idea is to calculate the area under 2.23 MeV gamma-rays photo peak resulting from neutron collision with Hydrogen atoms which are in mix with deuterium in samples. The study carried out by both simulation and experiment. Monte Carlo MCNPX2.6 code has been used for simulation and based on its acceptable results an experimental setup has been arranged. The coordination of results was in the range of R = 0.99 and R = 0.98 in simulation and experiment respectively. The accuracy of the study has been investigated by measuring the concentration of an unknown sample by both PGNAA and Fourier transform infrared spectroscopy (FT-IR) methods in which there were acceptable correlation between these two methods.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.269-274
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• 2019

$^{64}Cu$ is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, ${\beta}^-$ and ${\beta}^+$) and 12.7 h half-life. Production of $^{64}Cu$ by irradiation $^{nat}Cu$ and $^{nat}CuNPs$ in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of $^{63}Cu(n,{\gamma})^{64}Cu$ reaction was done with TALYS-1.8 code. The activity value of $^{64}Cu$ was calculated with theoretical approach and MCNPX-2.6 code. The results were compared with related experimental results which showed good adaptations between them.

• Journal of Astronomy and Space Sciences
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• v.41 no.1
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• pp.17-23
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• 2024

The safety of electronic components used in aerospace systems against cosmic rays is one of the most important requirements in their design and construction (especially satellites). In this work, by calculating the dose caused by proton beams in geostationary Earth orbit (GEO) orbit using the MCNPX Monte Carlo code and the MULLASSIS code, the effect of different structures in the protection of cosmic rays has been evaluated. A multi-layer radiation shield composed of aluminum, water and polyethylene was designed and its performance was compared with shielding made of aluminum alone. The results show that the absorbed dose by the simulated protective layers has increased by 35.3% and 44.1% for two-layer (aluminum, polyethylene) and three-layer (aluminum, water, polyethylene) protection respectively, and it is effective in the protection of electronic components. In addition to that, by replacing the multi-layer shield instead of the conventional aluminum shield, the mass reduction percentage will be 38.88 and 39.69, respectively, for the two-layer and three-layer shield compared to the aluminum shield.

• Journal of IKEEE
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• v.19 no.3
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• pp.282-287
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• 2015

In this study, an optical fiber detector was constructed by using a Ce:GAGG scintillator, optical fiber, and photomultiplier. The single crystal size of the scintillator was set to $3{\times}3{\times}20mm^3$ after simulating the counting efficiency of gamma rays in the scintillator by using the MCNPX code. The constructed detector used the standard gamma ray sources $^{137}Cs$ and $^{133}Ba$ to measure radiation and analyze the spectral characteristics of gamma rays. The resulting trend curve showed excellent linearity with an R-squared value of 0.99741, and the detector characteristics were found to vary 2% or less with distance based on comparison with the MCNPX value. Furthermore, the spectroscopic analysis of the gamma ray energy from the single-ray and mixed-ray sources showed that $^{137}Cs$ had its peak energy at 662 keV, and $^{133}Ba$ had at 356 keV. It seems that if the fiber-optics detector is used, working hours and exposure of worker can be reduced.

• Journal of the Korean Society of Radiology
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• v.8 no.5
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• pp.249-254
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• 2014

Medical linear accelerator is used for radiotherapy since it was developed in 1952, the utilization rate is further increased. It is used high energy radiotherapy using the energy of the photon of 6 MeV or more is universal at present, but the creation of the neutron by photonuclear reaction cause a problem that is radiation exposure of patients and operators. Therefore, in this study, to analyze the spectrum of the photon beam of 6 to 24 MV that occurred in the medical linear accelerator using the Monte Carlo code MCNPX, the number of photons of 7.41 MeV or more, which is a neutron production threshold energy of tungsten and average energy. The result of 24 MV in the beginning and the 8 MV was 0.59% of the total number of detected photons and it was founded that the number of photons are increased which are possible to cause the photonuclear reaction.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4671-4678
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• 2022

The purpose of this comprehensive research is to observe the impact of scintillator crystal type on entire detection process. For this aim, MCNPX (version 2.6.0) is used for designing of a physical plastic scintillation detector available in our laboratory. The modelled detector structure is validated using previous studies in the literature. Next, different types of plastic scintillation crystals were assessed in the same geometry. Several fundamental detector properties are determined for six different plastic scintillation crystals. Additionally, the deposited energy quantities were computed using the MCNPX code. Although six scintillation crystals have comparable compositions, the findings clearly indicate that the crystal composed of PVT 80% + PPO 20% has superior counting and detecting characteristics when compared to the other crystals investigated. Moreover, it is observed that the highest deposited energy amount, which is a result of the highest collision number in the crystal volume, corresponds to a PVT 80% + PPO 20% crystal. Despite the fact that plastic detector crystals have similar chemical structures, this study found that performing advanced Monte Carlo simulations on the detection discrepancies within the structures can aid in the development of the most effective spectroscopy procedures by ensuring maximum efficiency prior to and during use.

• The Journal of the Korea Contents Association
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• v.14 no.2
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• pp.467-474
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• 2014

The relative dose calculated by MCNPX and the relative dose measured by ionization chamber and solid phantoms evaluated the accuracy comparing with Monte Carlo simulation. In order to apply Monte Carlo simulation the intraluminal brachytherapy of extrahepatic bile duct cancer, 192Ir sealed radioactive source replicate, Bile duct and surrounding organs were made using KMIRD phantom based on a South Korea standard man. To check the absorbed dose of normal organs around bile duct, we set the specific effective energy and initial radioactivity to 1 Ci using MCNPX. Evaluation of the accuracy of the Monte Carlo simulation, the difference of the relative dose is the most 1.96% that satisfy the criteria that is the relative error less than 2% suggested by MCNPX code. In addition, The specific effective energy and absorbed dose of normal organs that were relatively adjacent to bile duct such as right side of kidney, liver, pancreas, transverse colon, spinal cord, stomach and small intestine were relatively high. on the contrary, the organs that were relatively distant to bile duct such as left side of kidney, spleen, ascending colon, descending colon and sigmoid colon were relatively low.

• Journal of the Korean Society of Safety
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• v.33 no.2
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• pp.28-31
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• 2018

In display and semi-conductor manufacturing process, there are numerous unstable factors such as particle concentration, minimal vibration, changes in magnetic field, or electrostatic that becomes an issue to be managed and controlled. In the recent, X-ray ionization is widely used that is neutralized by separating air or gas molecules in the area where the static must be resolved. The mono-type of X-ray ionizer was not capable to be used in $8^{th}$ generation panels manufacturing plant due to its insufficient ionizing coverage since the panel itself is approximately in $2m{\times}3m$. To resolve the current problem, the development of new type called, "Multi-type X-ray ionizer" has resulted in covering enough ionizing space in $8^{th}$ generation panels industry. Comparing mono and multi types with MCNPX code simulation, the multi one indicates more X-ray flux, efficiency, and ionization performance in comparison with either a mono-type or multi-type in array format. In addition, the ionizing efficiency of overlapping area with multi-type showed 30% higher effectiveness rate as to the ordinary mono-type.