• 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.

• Journal of Radiation Protection and Research
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• v.35 no.4
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• pp.142-150
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• 2010

The purpose of this article is to evaluate the radiation safety of CANDU spent fuel disposal system by using MCNPX which was revised in order to improve disposal efficiency. This research analyzed every system components's configuration, dimension and material. Geometric modeling and dose assessment for each system components showed that dose results for inner components had high values, but final disposal system had enough margin for radiation safety.

• 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.51 no.3
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• pp.647-653
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• 2019

A comprehensive study of photon interaction features has been made for some alloys containing Pd and Ag content to evaluate its possible use as alternative gamma radiations shielding material. The mass attenuation coefficient (${\mu}/{\rho}$) of the present alloys was measured at various photon energies between 81 keV - 1333 keV utilizing HPGe detector. The measured ${\mu}/{\rho}$ values were compared to those of theoretical and computational (MCNPX code) results. The results exhibited that the ${\mu}/{\rho}$ values of the studied alloys are in same line with results of WinXCOM software and MCNPX code results at all energies. Moreover, Pd75/Ag25 alloy sample has the maximum radiation protection efficiency (about 53% at 81 keV) and lowest half value layer, which shows that Pd75/Ag25 has superior gamma radiation shielding performance among the compared other alloys.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.853-859
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• 2019

A comprehensive study of photon interaction features has been made for some alloys containing Pd and Ag content to evaluate its possible use as alternative gamma radiations shielding material. The mass attenuation coefficient (${\mu}/{\rho}$) of the present alloys was measured at various photon energies between 81 keV-1333 keV utilizing HPGe detector. The measured ${\mu}/{\rho}$ values were compared to those of theoretical and computational (MCNPX code) results. The results exhibited that the ${\mu}/{\rho}$ values of the studied alloys are in the same line with results of WinXCOM software and MCNPX code results at all energies. Moreover, Pd75/Ag25 alloy sample has the maximum radiation protection efficiency (about 53% at 81 keV) and lowest half value layer, which shows that Pd75/Ag25 has superior gamma radiation shielding performance among the other compared alloys.

• 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.

• Journal of the Korean Society of Radiology
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• v.9 no.5
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• pp.295-300
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• 2015

There are various type of particle accelerators such as Kyoungju 100-MeV proton beam accelerator in Korea. And Korea plans to build large particle accelerator such as heavy ion accelerator and 4th generation light source facility. The accelerated high energy particles of these facility produce 2nd neutron after nuclear reaction with target materials. And then these 2nd neutron activate structural materials and surrounding environment. Accordingly, it is important to consider the activation and shielding calculation on design of facility for safety operation. In this study, we tried to calculate and compare the neutron flux from the interaction $^{la}150$ beam with target material(Cu) according to thickness of iron and concrete shielding material by MCNPX 2.7 with nuclear library JENDL/HE 07and la150. To verify the properties of nuclear library, we compared computational results with experimental value. These results can be used for dose evaluation technology in planning of the shielding of large particle accelerator.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.172-178
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• 2013

In this study, neutron detection efficiency of the UNDA system, which has been developed for study on nuclear material accountancy in a future pyro-process facility, was evaluated by using the MCNPX code. The detection efficiency was evaluated as a function of (1) positions of $^{252}Cf$ neutron source in the axial and radial directions, and (2) thicknesses and locations of the container filled with the depleted uranium materials for two different designs of the UNDA. In the case of $^{252}Cf$ source positions, detection efficiency was distributed from 6.83% to 13.35%. As $^{252}Cf$ source was positioned at upper part in the axial direction, detection efficiency was decreased after a slight increase. On the other hands, as $^{252}Cf$ source was positioned at outer part in the radial direction, detection efficiency was increased. In the case of container thickness, there was a slight decline when the thickness was increased. As the container was located at upper part, detection efficiency was decreased and as the container was located at outer part, detection efficiency was increased. Detection efficiency was varied from 10.31% to 13.61%. These values were higher than that of $^{252}Cf$ source case. The UNDA with polyethylene cover has about 2% higher detection efficiency than the UNDA without the cover.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.252-253
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• 2011

• The Journal of the Korea Contents Association
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• v.15 no.11
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• pp.298-306
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• 2015

This study is the material of the additional filter(Cu, Ni, CaWO4, Gd+Ba) being used when the diagnosis X-ray was varied to evaluate the spatial dose distribution accordingly. And it suggest to find a suitable material. Experiments using MCNPX program based on the Monte Carlo simulation method was carried out by selecting the chest and abdomen taken. As a result, each material per dose, the average scatter dose is approximately 62%, 100 cm radius of the point of the simulated body surface exposure dose and 50 cm radius centered on the point average about 47%. It is determined that an Al material is currently available in accordance with the result to be replaced by Cu, Ni material is sufficient. With just the thickness due to the difference in the atomic number and density adjusted to be about one-tenth of the Al it will be suitable.