• Progress in Medical Physics
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• v.27 no.4
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• pp.232-235
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• 2016

The purpose of this study is to develop the treatment planning system (TPS) based on Monte-Carlo simulation for BNCT. In this paper, we will propose a method for dose estimation by Monte-Carlo simulation using the CT image, and will evaluate the accuracy of dose estimation of this TPS. The complicated geometry like a human body allows defining using the lattice function in MCNPX. The results of simulation such as flux or energy deposition averaged over a cell, can be obtained using the features of the tally provided by MCNPX. To assess the dose distribution and therapeutic effect, dose distribution was displayed on the CT image, and dose volume histogram (DVH) was employed in our developed system. The therapeutic effect can be efficiently evaluated by these evaluation tool. Our developed TPS could be effectively performed creating the voxel model from CT image, the estimation of each dose component, and evaluation of the BNCT plan.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.

• Journal of the Korean Society of Radiology
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• v.13 no.4
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• pp.565-572
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• 2019

In this study, shielding analysis of material and thickness of 3D printer filaments was performed for the manufacture of custom shielding by radiation workers during outdoor radiographic test. The shielding was attached to the ICRU Slab Phantom after selecting the voxel source $^{192}Ir$ and $^{75}Se$ through simulation using MCNPX, and the distance between the source and the slab Phantom was set at 100 cm. The 12 shielding materials were divided into 5 mm units up to 200 mm from the absence of shielding materials to evaluate the energy absorbed per unit mass of each shielding material. The results showed that the shielding effect was high in the order of ABS + Tungsten, ABS + Bismuth, PLA + Copper, PLA + Iron from all sources of radiographic test. However, compared to lead, the shielding effect was somewhat lower. Based on this study in the future, further study of the atomic number and the high density filament material is necessary.

• 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.

• Progress in Medical Physics
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• v.22 no.4
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• pp.190-197
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• 2011

Currently, the dose distribution calculation used by commercial treatment planning systems (TPSs) for high-dose rate (HDR) brachytherapy is derived from point and line source approximation method recommended by AAPM Task Group 43 (TG-43). However, the study of Monte Carlo (MC) simulation is required in order to assess the accuracy of dose calculation around three-dimensional Ir-192 source. In this study, geometry factor was calculated using segmented sources integration method by dividing microSelectron HDR Ir-192 source into smaller parts. The Monte Carlo code (MCNPX 2.5.0) was used to calculate the dose rate $\dot{D}(r,\theta)$ at a point ($r,\theta$) away from a HDR Ir-192 source in spherical water phantom with 30 cm diameter. Finally, anisotropy function and radial dose function were calculated from obtained results. The obtained geometry factor was compared with that calculated from line source approximation. Similarly, obtained anisotropy function and radial dose function were compared with those derived from MCPT results by Williamson. The geometry factor calculated from segmented sources integration method and line source approximation was within 0.2% for $r{\geq}0.5$ cm and 1.33% for r=0.1 cm, respectively. The relative-root mean square error (R-RMSE) of anisotropy function obtained by this study and Williamson was 2.33% for r=0.25 cm and within 1% for r>0.5 cm, respectively. The R-RMSE of radial dose function was 0.46% at radial distance from 0.1 to 14.0 cm. The geometry factor acquired from segmented sources integration method and line source approximation was in good agreement for $r{\geq}0.1$ cm. However, application of segmented sources integration method seems to be valid, since this method using three-dimensional Ir-192 source provides more realistic geometry factor. The anisotropy function and radial dose function estimated from MCNPX in this study and MCPT by Williamson are in good agreement within uncertainty of Monte Carlo codes except at radial distance of r=0.25 cm. It is expected that Monte Carlo code used in this study could be applied to other sources utilized for brachytherapy.

• Journal of the Korean Society of Radiology
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• v.10 no.5
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• pp.327-335
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• 2016

The study is enforce to study image quality evaluation of condition provide the IEC and combination of clinical conditions each quality of radiation that image quality to assess the conditions provided to IEC in the clinical environment to conduct image quality assessment of the digital radiography system in the detector have environmental limits. First, image quality evaluation was evaluated by measuring the MTF, NPS using four quality of radiation and Using MCNPX simulation lastly DQE make a image quality evaluation after calculating the particle fluence to analyze spectrum quality of radiation. Second, Using MCNPX simulation of four quality of radiation was evaluated absorbed dose rate about electronic 1 per unit air, water, muscle, bone by using Radiation flux density and energy, mass-energy absorption coefficient of matter. Results of evaluation of image quality, MTF of four quality of radiation was satisfied diagnosis frequency domain 1.0 ~ 3.0 lp/mm of general X-ray that indicated 1.13 ~ 2.91 lp/mm spatial frequency. The NPS has added filter, spatial frequency 0.5 lp/mm at standard NPS showed a tendency to decrease after increase. Unused added filter, spatial frequency 0.5 lp/mm at standard NPS showed a certain NPS result value after decrease. DQE in 70 kVp / unuesd added filter(21 mm Al) / SID 150 cm that patial frequency 1.5 lp/mm at standard showed a tendency to decrease after certain value showed. Patial frequency in the rest quality of radiation was showed a tendency to decrease after increase. Results of evaluation of absorbed dose, air < water < muscle < bone in the order showed a tendency to increase. Based on the results of this study provide to basic data that present for the image quality evaluation method of a digital radiation imaging system in various the clinical condition.

• Progress in Medical Physics
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• v.24 no.4
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• pp.253-258
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• 2013

The purpose of this study is to evaluate and analyze the relationship between the external radiation dose reconstruction which is transmitted from the patient who receives radiation treatment through electronic portal imaging device (EPID) and the internal dose derived from the Monte Carlo simulation. As a comparative analysis of the two cases, it is performed to provide a basic indicator for similar studies. The geometric information of the experiment and that of the radiation source were entered into Monte Carlo n-particle (MCNPX) which is the computer simulation tool and to derive the EPID images, a tally card in MCNPX was used for visualizing and the imaging of the dose information. We set to source to surface distance (SSD) 100 cm for internal measurement and EPID. And the water phantom was set to be 100 cm of the source to surface distance (SSD) for the internal measurement and EPID was set to 90 cm of SSD which is 10 cm below. The internal dose was collected from the water phantom by using mesh tally function in MCNPX, accumulated dose data was acquired by four-portal beam exposures. At the same time, after getting the dose which had been passed through water phantom, dose reconstruction was performed using back-projection method. In order to analyze about two cases, we compared the penetrated dose by calibration of itself with the absorbed one. We also evaluated the reconstructed dose using EPID and partially accumulated (overlapped) dose in water phantom by four-portal beam exposures. The sum dose data of two cases were calculated as each 3.4580 MeV/g (absorbed dose in water) and 3.4354 MeV/g (EPID reconstruction). The result of sum dose match from two cases shows good agreement with 0.6536% dose error.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.21-34
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• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

• 대한방사선방어학회:학술대회논문집
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• 2009.11a
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• pp.176-177
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• 2009

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.182-183
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• 2011