• Progress in Medical Physics
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• v.15 no.4
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• pp.210-214
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• 2004

Since production of radioactive isotope for using PET, a lot of neutrons were produced. The produced neutrons were mainly shielded by concrete facility. Secondary photons are generated and emitted from the concrete shielding wall of the PET cyclotron since the proton-generated neutrons are thermalized and absorbed in the concrete wall and emit secondary radiations, i.e., photons. This study calculated neutron dose and photon dose at outside of the accelerator facility using MCNPX code. As results of the calculation, total dose were calculated less than limited dose by law.

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

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.375-376
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• 2010

• 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.

• Progress in Medical Physics
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• v.17 no.4
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• pp.224-231
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• 2006

The electron energy spectra for 6 MeV electron beam were calculated using a MCNPX code. The head of the linear accelerator (ML6M; Mitsubishi, Japan) was modelled for this study. The energy spectrum of the initial electron beam was assumed to be Gaussian and the mean energy was determined by evaluating the measured and calculated values of $R_{50}$ and dose profiles in air. The energy distributions for electrons and photons at the interested points in the head of the linear accelerator were calculated by appling the Initial beam parameters. The effect of contaminant photons on depth dose curves were estimated by the photon energy spectra at the end of the applicator.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.350-351
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• 2005

• 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.

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

• 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.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2014.05a
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• pp.816-818
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• 2014

Gas target chamber has been developed for producing $^{123}I$ which is radiopharmaceuticals for diagnosis of thyroid cancer, and modeled how to occur nuclear reaction between chamber and $^{124}Xe$ with energy 30MeV inside the gas target chamber by using the MCNPX. The beam energy was lost as the beam spread when beam hit inside the gas target chamber. The cooling water was used not to change the gas target chamber as loss of energy transfer to the thermal energy. Spiral cooling line was designed for cooling the target chamber efficiently. By using the c30 cyclotron, $^{124}Xe(p,2n)$, $^{124}Xe(p,n)$, $^{124}Xe(p,pn)$ nuclear reactions were studied. In this study, we predict the production yield.