• Progress in Medical Physics
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• v.22 no.2
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• pp.79-84
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• 2011

In this study, Geant4 based Monte Carlo simulations were carried out for medical linear accelerator. Modified Medical Linac2 toolkit was used for calculation. The energy spectrum, most probable energy and the photon mean energy compared with the published results using the EGS4 code. The results well agreed with published results. The calculated results of photon fluence, energy fluence and mean energy according to the radius from the centre of the beam were analyzed. Monte Carlo simulation using Medical Linac2 code is considered to be useful for analysis of medical linear accelerator. Because the calculated results varies depending on Physics List model for same head structure. It it important to choose the right model for research purpose. Monte Carlo simulation using GEANT4 Medical Linac2 is a valuable for any novice to adopt this code to the study related to 6 MV photon fluence from medical linear accelerator.

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.2
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• pp.57-61
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• 2014

Purpose Iodine (I-131) is one of the most widely used radioactive isotopes for therapeutic in the field of nuclear medicine. Therapeutic I-131 capsule is made out of lead to shield high energy radiation. Accurate dosimetry is necessarily required to perform safe and effective work for relative workers. The Monte Carlo method is known as a method to predict the absorbed dose distribution most accurately in radiation therapy and many researchers constantly attempt to apply this method to the dose calculation of radiotherapy recently. This paper aims to calculate distance dependent and activity dependent therapeutic I-131 capsule using GEANT4. Materials and Methods Therapeutic capsules was implemented on the basis of the design drawings. The simulated dose was determined by generating of gamma rays of energy to more than 364 keV. The simulated dose from the capsule at the distance of 10 cm and 100 cm was measured and calculated in the model of water phantom. The simulated dose were separately calculated for each position of each detector. Results According to the domestic regulation on radiation safety, the dose at 10 cm and 100 cm away from the surface of therapeutic I-131 capsule should not exceed 2.0 mSv/h and 0.02 mSv/h, respectively. The simulated doses turned out to be less than the limit, satisfying the domestic regulation. Conclusion These simulation results may serve as useful data in the prediction of hands dose absorbed by I-131 capsule handling. GEANT4 is considered that it will be effectively used in order to check the radiation dose.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.302-303
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• 2011

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.191-196
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• 2012

A double-scattering type Compton camera which is appropriate to imaging a high-energy gamma source has been developed for nuclear material surveillance at Hanyang University. The double-scattering type Compton camera can provide high imaging resolution; however, it has disadvantage of relatively low imaging sensitivity than existing single-scattering type Compton camera. In this study, we introduce a novel concept of a dual-mode Compton camera which incorporates two different types of Compton camera, i.e., single- and double-scattering type. The dual-mode Compton camera can operate high-resolution mode and high-sensitivity mode in a single system. To maximize its performance, the geometrical configuration was optimized by using Geant4 Monte Carlo simulation toolkit. In terms of imaging sensitivity, high-sensitivity mode had higher sensitivity than high-resolution mode up to 100 times while high imaging resolution of the double-scattering Compton camera was maintained.

• Progress in Medical Physics
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• v.19 no.1
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• pp.21-34
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• 2008

The parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2006)] for calculating electron and photon beam doses has been developed based on the three dimensional geometry defined by computed tomography (CT) images and implemented on the Beowulf PC cluster. Understanding the limitations of Monte Carlo codes is useful in order to avoid systematic errors in simulations and to suggest further improvement of the codes. We evaluated the PMCEPT code by comparing its normalized depth doses for electron and photon beams with those of MCNP5, EGS4, DPM, and GEANT4 codes, and with measurements. The PMCEPT results agreed well with others in homogeneous and heterogeneous media within an error of $1{\sim}3%$ of the dose maximum. The computing time benchmark has also been performed for two cases, showing that the PMCEPT code was approximately twenty times faster than the MCNP5 for 20-MeV electron beams irradiated on the water phantom. For the 18-MV photon beams irradiated on the water phantom, the PMCEPT was three times faster than the GEANT4. Thus, the results suggest that the PMCEPT code is indeed appropriate for both fast and accurate simulations.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.647-653
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• 2020

Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3164-3170
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• 2021

The increasing interest that GEANT4 is gaining nowadays, because of its special capabilities, prompted us to address its reliability in neutronic calculation for the realistic and complex spallation target MEGAPIE of the Paul Scherrer Institute of Switzerland. In this paper we have specifically addressed the neutronic characterization of the active zone of this target. Three physical quantities are evaluated: neutron flux spectra and total neutron fluxes on target's z-axis, and the neutron yield as a function of the target's altitude and radius. Comparison of the obtained results with those of the MCNPX reference code and some experimental measurements have confirmed the impact of the geometrical and proton beam models on the neutron fluxes. It has also allowed to reveal the intrinsic influence of the code type. The resulting differences reach a factor of ~2 for the beam model and 4-18% for the other parameters cumulated. The analysis of the neutron yield has led us to conclude that: 1) Increasing the productivity of the MEGAPIE target cannot be achieved simply by increasing the thickness of the target, if the irradiation parameters are not modified. 2) The size of the spallation area needs to be redefined more precisely.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1456-1463
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• 2022

The aim of this work is to study the radiation shielding properties of cement samples with waste glass incortated into its composition. The mass attenuation coefficient (MAC) of the samples were experimentally determined to evaluate their radiation shielding ability. The experimental coefficient was evaluated using NaI detector for gamma energies between 59.53 keV and 1408.01 keV using different radioactive point sources Am-241, Eu-152, Co-60, and Cs-137, and the gamma transmission parameters half-value layer, mean free path, and transmission factor were calculated. The theoretical coefficient of the composites was determined using Geant4 and XCOM software. The results were also compared against Geant4 and XCOM simulations by calculating the relative deviation between the values to determine the accuracy of the results. In addition the mechanical properties (including Compressive and porosity) as well as the thermogravimetric analysis were tested for the present samples. Overall, it was concluded that the cement sample with 50% waste glass has the greatest shielding potential for radiation shielding applications and is a useful way to reuse waste glass.

• Journal of the Korean Physical Society
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• v.73 no.12
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• pp.1814-1820
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• 2018

With the publication of TRS-483 in late 2017 the IAEA has established an international code of practice for reference dosimetry in small and non-standard fields based on a formalism first suggested by Alfonso et al. in 2008. However, data on beam quality correction factors ($k^{f_{msr},f_{ref}}_{Q_{msr},Q_0}$) for the Leksell Gamma $Knife^{(R)}$ $Perfexion^{TM}$ is scarce and what little data is available was obtained under conditions not necessarily in accordance with the IAEA's recommendations. This study constitutes the first systematic attempt to calculate those correction factors by applying the new code of practice to Monte Carlo simulation using the GEANT4 toolkit. $k^{f_{msr},f_{ref}}_{Q_{msr},Q_0}$ values were determined for three common ionization chamber detectors and five different phantom materials, with results indicating that in most phantom materials, all chambers were well suited for reference dosimetry with the Gamma $Knife^{(R)}$. Similarities and differences between the results of this study and previous ones were also analyzed and it was found that the results obtained herein were generally in good agreement with earlier PENELOPE and EGSnrc studies.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2483-2488
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• 2023

Differential die-away analysis (DDAA) technology is a special nuclear material (SNM) active detection analysis technology. Be a nuclear material shielded or not, the technology can reveal the existence of nuclear materials by inducing fission from an external pulsed neutron source. In this paper, a detection model based on DDAA analysis technology was established by geant4 Monte Carlo simulation software, and the optimal sensitivity of the detection system is achieved by optimizing different configurations. After the geant4 simulation and optimization, a prototype was established, and experimental research was carried out. The result shows that the prototype can detect 200 g of ²³⁵U in a steel cylinder shield that's of 1.5 cm in inner diameter, 10 cm in thickness and 280 kg in weight.