• Journal of the Korean Society of Radiology
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• v.10 no.3
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• pp.153-159
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• 2016

This study investigated relationship between secondary electrons produced from single gold nanoparticle as a result of its interaction with radiation and particle size and incidence energy, provided basic data related to the dose enhancement effect based on gold nanoparticles. Monte Carlo simulation was applied by using MCNPX MC code, 50, 100, 150 kV and 6, 15 MV x-ray energy was used. In a water phantom, single gold nanoparticles that are 30, 50, 70, 90, and 110 nm in diameter were placed and the tally volume was designated at every 10 nm. Difference in electrons produced from gold nanoparticles was normalized based on absence of nanoparticle. When the X ray energy decreased and the diameter of gold particles increased, more electrons were produced. When the energy was lower, in the linear formula related to nanoparticle size and electron production, the gradient was higher. And, in comparison to the MV X-ray, at kV X-ray, significantly more electrons were produced. This study can be used as data to understand the dose enhancement effect based on gold nanoparticles, and further research related to various materials that dose enhancement including gold nanoparticles needs to be conducted.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.293-302
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• 2019

Many researches have been done to develop and improve the performance of personal (individual) dosimeter response to cover a wide of neutron energy range (from thermal to fast). Depending on the individual category of the dosimeter, the semiconductor sensor has been used to simplify and lightweight. In this plan, it's very important to have a fairly accurate counting of doses rate in different energies. With a general design and single-sensor simulations, all optimal thicknesses have been extracted. The performance of the simulation scheme has been compared with the commercial and laboratory samples in the world. Due to the deviation of all dosimeters with a flat energy response, in this paper, has been used an idea of one semi-conductor sensor to have the flat energy-response in the entire neutron energy range. Finally, by analyzing of the sensors data as arrays for the first time, we have reached a nearly flat and acceptable energy-response. Also a comparison has been made between Lucite-PMMA ($H_5C_5O_2$) and polyethylene-PE ($CH_2$) as a radiator and $B_4C$ has been studied as absorbent. Moreover, in this paper, the effect of gamma dose in the dosimeter has been investigated and shown around the standard has not been exceeded.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1863-1870
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• 2023

Molybdenum-99 (⁹⁹Mo) is used for preparing Technetium-99 m (^99mTc), which is the most widely used isotope in nuclear medicine. In this work, a study for ⁹⁹Mo production based on a high-power electron accelerator has been performed as an alternative approach to produce ^99mTc. In this study, Monte Carlo MCNPX2.6 code has been used to examine a novel idea of simultaneous hybrid production of ⁹⁹Mo via both photoneutron and neutron capture reactions using an electron accelerator in heavy water tank. It is expected that this conceptual design including an arrangement of metallic plates of ¹⁰⁰Mo and ⁹⁸Mo produces total activity of 97.5 Ci at the end of 20-h continuous e-beam irradiation (30 MeV, 10 mA).

• Progress in Medical Physics
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• v.23 no.3
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• pp.162-168
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• 2012

In proton therapy, in vivo dose verification is one of the most important parts to fully utilize characteristics of proton dose distribution concentrating high dose with steep gradient and guarantee the patient safety. Currently, in order to image the proton dose distribution, a prompt gamma distribution detection system, which consists of an array of multiple CsI(Tl) scintillation detectors in the vertical direction, a collimator, and a multi-channel DAQ system is under development. In the present study, the optimal design of prompt gamma distribution detection system was studied by Monte Carlo simulations using the MCNPX code. For effective measurement of high-energy prompt gammas with enough imaging resolution, the dimensions of the CsI(Tl) scintillator was determined to be $6{\times}6{\times}50mm^3$. In order to maximize the detection efficiency for prompt gammas while minimizing the contribution of background gammas generated by neutron captures, the hole size and the length of the collimator were optimized as $6{\times}6mm^2$ and 150 mm, respectively. Finally, the performance of the detection system optimized in the present study was predicted by Monte Carlo simulations for a 150 MeV proton beam. Our result shows that the detection system in the optimal dimensions can effectively measure the 2D prompt gamma distribution and determine the beam range within 1 mm errors for 150 MeV proton beam.

• Journal of Radiation Protection and Research
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• v.37 no.1
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• pp.41-49
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• 2012

The objective of this study is for development of the reference Korean female phantom, HDRK-Woman. The phantom was constructed by adjusting a Korean woman voxel phantom to the Reference Korean data. The Korean woman phantom had been developed based on the high-resolution color slice images obtained from an adult Korean female cadaver. There were a total of 39 organs including the 27 organs specified in ICRP 103 for effective dose calculation. The voxel resolution of the phantom was $1.976{\times}1.976{\times}2.0619\;mm^3$ and the voxel array size is $261{\times}109{\times}825$ in the x, y and z directions. Then, the voxel resolution was changed to $2.0351{\times}2.0351{\times}2.0747\;mm^3$ for adjustment of the height and total bone mass of the phantom to the Reference Korean data. Finally, the internal organs and tissue were adjusted using in-house software program developed for 3D volume adjustment of the organs and tissue. The effective dose values of HDRK phantoms were calculated for broad parallel photon beams using MCNPX Monte Carlo code and compared with those of ICRP phantoms.

• Journal of the Korean Society of Safety
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• v.32 no.2
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• pp.34-37
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• 2017

In recent emerging industry, Display field becomes bigger and bigger, and also semiconductor technology becomes high density integration. In Flat Panel Display, there is an issue that electrostatic phenomenon results in fine dust adsorption as electrostatic capacity increases due to bigger size. Destruction of high integrated circuit and pattern deterioration occur in semiconductor and this causes the problem of weakening of thermal resistance. In order to solve this sort of electrostatic failure in this process, Soft X-ray ionizer is mainly used. Soft X-ray Ionizer does not only generate electrical noise and minute particle but also is efficient to remove electrostatic as it has a wide range of ionization. X-ray Generating efficiency has an effect on soft X-ray Ionizer affects neutralizing performance. There exist variable factors such as type of anode, thickness, tube voltage etc., and it takes a lot of time and financial resource to find optimal performance by manufacturing with actual X-ray tube source. MCNPX (Monte Carlo N-Particle Extended) is used for simulation to solve this kind of problem, and optimum efficiency of X-ray generation is anticipated. In this study, X-ray generation efficiency was measured according to target material thickness using MCNPX under the conditions that tube voltage is 5 keV, 10 keV, 15 keV and the target Material is Tungsten(W), Gold(Au), Silver(Ag). At the result, Gold(Au) shows optimum efficiency. In Tube voltage 5 keV, optimal target thickness is $0.05{\mu}m$ and Largest energy of Light flux appears $2.22{\times}10^8$ x-ray flux. In Tube voltage 10 keV, optimal target Thickness is $0.18{\mu}m$ and Largest energy of Light flux appears $1.97{\times}10^9$ x-ray flux. In Tube voltage 15 keV, optimal target Thickness is $0.29{\mu}m$ and Largest energy of Light flux appears $4.59{\times}10^9$ x-ray flux.

• Journal of Radiation Protection and Research
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• v.38 no.1
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• pp.37-43
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• 2013

In case of neutron dose measurement using TLDs (thermo-luminescence dosimeters), because the neutron energy dependence of the TLD is very high, the calibration of the energy response according to the characteristics of the neutron spectrum of workplace is required. In the present study, the ambient dose equivalent rates inside and around the Long-Counter (neutron detector) with narrow and complex inside in the neutron field of $^{252}Cf$ were evaluated. The calibration factors to account for the neutron energy dependence of TLDs were established for both the bare and $D_2O$ modulated $^{252}Cf$ neutron beams, respectively. The values of the TLD's measurement were compared with the computational results of the MCNPX (Monte Carlo N-Particles transport code). When using the two calibration factors of the TLD than a single calibration factor, the measured and the calculated values at the point of verification outside and inside the Long-Counter were in more good agreement. This results show that TLD should be calibrated in the reference neutron field similar to workplace situation.

• Journal of Radiation Protection and Research
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• v.29 no.2
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• pp.97-104
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• 2004

MIRD-type Korean adult male phantom, 'KMIRD' was constructed to calculate Korean-specific dosimetric quantities for radiation protection consideration. The external shape of KMIRD was based on national physical standard data of Korean. KMIRD has thicket trunk than MIRD5 and arm models divided from trunk. The height and weight of the KMIRD are 171 cm and 63.8 kg. ICRP23 data were referred to constitute organs and tissues of KMIRD. However nine organs were constructed based on Korean reference data provided by Radiation Health Research Institute. In the present study, the MCNPX2.3 Monte Carlo transport code was combined with KMIRD to calculate dose conversion coefficients for photon in the energy range from 0.05 to 10 MeV. The simulated irradiation geometries are broad parallel photon beams in AP, PA, LLAT and RLAT direction. Absorbed dose conversion coefficients were compared with data calculated with MIRD5, MIRD-type phantom based on ICRP23 reference man. In some organs, the discrepancies between two phantoms amount up to nearly 30%. The effective doses conversion coefficients of KMIRD are lower than those of MIRD5. The dose discrepancies between two MIRD-type phantoms ate because of physical differences between Korean and Western, also geometric differences between two phantoms. KMIRD should be revised using the full set of Korean reference data of all organs. The developed MIRD-type Korean adult male phantom can be applied to dose assessment of internal exposure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.201-209
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• 2016

The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except $5{\times}5{\times}2m\;size$) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a $50{\times}50{\times}2m\;size$ facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2007.06a
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• pp.328-328
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• 2007

Large area matrix-addressed image detectors are a recent technology for x-ray imaging with medical diagnostic and other applications. The imaging properties of x-ray pixel detectors depend on the quantum efficiency of x-rays, the generated signal of each x-ray photon and the distribution of the generated signal between pixels. In a phosphor coated detector the light signal is generated by electrons captured in the phosphor screen. In our study we simulated the lateral spread distributions for phosphor coupled detector by Monte Carlo simulations. Most simulations of such detectors simplify the setup by only taking the conversion layer into account neglecting behind. The Monte Carlo code MCNPX has been used to simulate the complete interaction and subsequent charge transport of x-ray radiation. This has allowed the analysis of charge sharing between pixel elements as an important limited factor of digital x-ray imaging system. The parameters are determined by lateral distribution of x-ray photons and x-ray induced electrons. The primary purpose of this study was to develop a design tool for the evaluation of geometry factor in the phosphor coupled optical imaging detector. In order to evaluate the spatial resolution for different phosphor material, phosphor geometry we have developed a simulation code. The developed code calculates the energy absorption and spatial distribution based on both the signal from the scintillating layer and the signal from direct detection of x-ray in the detector. We show that internal scattering contributes to the so-called spatial resolution drop of the image detector. Results from the simulation of spatial distribution in a phosphor pixel detector are presented. The spatial resolution can be increased by optimizing pixel size and phosphor thickness.