• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.181-190
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• 2012

Naturally occurring radioactive materials (NORM) in building materials are main sources of external radiation exposure to the general public. The objective of this study was to assess external radiation dose in Korean dwellings due to NORM in concrete walls. Reference room model for dose assessment was made by analyzing room structure and housing scale of Korean dwellings. In addition, dose assessments were made for varying room sizes. Absorbed doses to air and effective dose rates were calculated using radiation transport code MCNPX. Assuming a reference room of $3{\times}4{\times}2.8m^3$, absorbed dose rates in air were 0.80, 0.97, 0.08 nGy $h^{-1}$ per Bq $kg^{-1}$ for uranium series, thorium series, and $^{40}K$, respectively. Effective dose rates were 0.57, 0.69, 0.058 nSv $h^{-1}$ per Bq $kg^{-1}$, respectively. Radiation dose resulting from concrete of ceiling and floor increased with room area while radiation dose from concrete of walls decreased with room area. Therefore, total radiation doses were almost the same for the varying room area from 5 to $30m^2$. Effective dose in Korean dwellings was calculated based on measurement data of NORM concentration in concrete and occupancy fraction of Korean population by location. Annual effective dose was 0.59 mSv assuming that indoor occupancy fraction was 0.89 and concentrations of uranium series, thorium series and $^{40}K$ were 26, 39, 596 Bq $kg^{-1}$, respectively. Finally, annual effective dose in Korean dwellings can be calculated by the following equation: Effective dose=indoor occupancy fraction${\times}8760\;h\;y^{-1}{\times}(0.57C_U+0.69C_{Th}+0.058C_K)$.

• Journal of Radiation Protection and Research
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• v.41 no.1
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• pp.71-79
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• 2016

Background: It is necessary to analyze radioactivity of naturally occurring radioactive materials (NORM) in products to ensure radiological safety required by Natural Radiation Safety Management Act. The pretreatments for the existing analysis methods require high technology and time. Such destructive pretreatments including grinding and dissolution of samples make impossible to reuse products. We developed a rapid and simple procedure of radioactivity analysis for thorium series in the products containing NORM. Materials and Methods: The developed method requires non-destructive or minimized pretreatment. Radioactivity of the product without pretreatment is initially measured using gamma spectroscopy and then the measured radioactivity is adjusted by considering material composition, mass density, and geometrical shape of the product. The radioactivity adjustment can be made using scaling factors, which is derived by radiation transport Monte Carlo simulation. Necklace, bracelet, male health care product, and tile for health mat were selected as representative products for this study. The products are commonly used by the public and directly contacted with human body and thus resulting in high radiation exposure to the user. Results and Discussion: The scaling factors were derived using MCNPX code and the values ranged from 0.31 to 0.47. If radioactivity of the products is measured without pretreatment, the thorium series may be overestimated by up to 2.8 times. If scaling factors are applied, the difference in radioactivity estimates are reduced to 3-24%. Conclusion : The developed procedure in this study can be used for other products with various materials and shapes and thus ensuring radiological safety.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.2
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• pp.145-151
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• 2013

Purpose: In this study, we analyzed how the dose change by field size effects on atomic number of shielding materials while using 6 MeV election beam. Materials and Methods: The parallel plate chamber is mounted in $25{\times}25cm^2$ the phantom such that the entrance window of the detector is flush with the phantom surface. phantom was covered laterally with aluminum, copper and lead which thickness have 5% of allowable transmission and then the doses were measured in field size $6{\times}6$, $10{\times}10$ and $20{\times}20cm^2$ respectively. 100 cGy was irradiated using 6 MeV electron beam and SSD (Source Surface Distance) was 100 cm with $10{\times}10cm^2$ field size. To calculate the photon flux, electron flux and Energy deposition produced after pass materals respectively, MCNPX code was used. Results: The results according to the various shielding materials which have 5% of allowable transmission are as in the following. Thickness change rate with field size of $6{\times}6cm^2$ and $20{\times}20cm^2$ that compared to the field size of $10{\times}10cm^2$ found to be +0.06% and -0.06% with aluminum, +0.13% and -0.1% with copper, -1.53% and +1.92% with lead respectively. Compare to the field size $10{\times}10cm^2$, energy deposition for $6{\times}6cm^2$ and $20{\times}20cm^2$ had -4.3% and +4.85% respectively without shielding material. With aluminum it had -0.87% and +6.93% respectively and with lead it had -4.16% and +5.57% respectively. When it comes to photon flux with $6{\times}6cm^2$ and $20{\times}20cm^2$ of field sizes the chance -8.95% and +15.92% without shielding material respectively, with aluminum the number -15.56% and +16.06% respectively and with copper the chance -12.27% and +15.53% respectively, with lead the number +12.36% and -19.81% respectively. In case of electron flux in the same condition, the number -3.92% and +4.55% respectively without shielding material respectively, with aluminum the number +0.59% and +6.87% respectively, with copper the number -1.59% and +3.86% respectively, with lead the chance -5.15% and +4.00% respectively. Conclusion: In this study, we found that the required thickness of the shielding materials got thinner with low atomic number substance as the irradiation field is increasing. On the other hand, with high atomic number substance the required thickness had increased. In addition, bremsstrahlung radiation have an influence on low atomic number materials and high atomic number materials are effected by scattered electrons.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.69-75
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• 2020

Background: Dose conversion coefficients (DCCs) have been commonly used to estimate radiation-dose absorption by human organs based on physical measurements of fluence or kerma. The International Commission on Radiological Protection (ICRP) has reported a library of DCCs, but few studies have been conducted on their applicability to non-Caucasian populations. In the present study, we collected a total of 8 Korean pediatric and adult voxel phantoms to calculate the organ DCCs for idealized external photon-irradiation geometries. Materials and Methods: We adopted one pediatric female phantom (ETRI Child), two adult female phantoms (KORWOMAN and HDRK Female), and five adult male phantoms (KORMAN, ETRI Man, KTMAN1, KTMAN2, and HDRK Man). A general-purpose Monte Carlo radiation transport code, MCNPX2.7 (Monte Carlo N-Particle Transport extended version 2.7), was employed to calculate the DCCs for 13 major radiosensitive organs in six irradiation geometries (anteroposterior, posteroanterior, right lateral, left lateral, rotational, and isotropic) and 33 photon energy bins (0.01-20 MeV). Results and Discussion: The DCCs for major radiosensitive organs (e.g., lungs and colon) in anteroposterior geometry agreed reasonably well across the 8 Korean phantoms, whereas those for deep-seated organs (e.g., gonads) varied significantly. The DCCs of the child phantom were greater than those of the adult phantoms. A comparison with the ICRP Publication 116 data showed reasonable agreements with the Korean phantom-based data. The variations in organ DCCs were well explained using the distribution of organ depths from the phantom surface. Conclusion: A library of dose conversion coefficients for major radiosensitive organs in a series of pediatric and adult Korean voxel phantoms was established and compared with the reference data from the ICRP. This comparison showed that our Korean phantom-based data agrees reasonably with the ICRP reference data.

• Journal of radiological science and technology
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• v.40 no.1
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• pp.93-99
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• 2017

Dose enhancement effects at megavoltage (MV) X and ${\gamma}-ray$ energies, and the effects of different energy levels on incident energy, dose enhancement agents, and concentrations were analyzed using Monte Carlo simulations. Gold, gadolinium, Iodine, and iron oxide ($Fe_2O_3$) were compared as dose enhancement agents. For incident energy, 4, 6, 10 and 15 MV X-ray spectra produced by a linear accelerator and a Co-60 ${\gamma}-ray$ were used. The dose enhancement factor (DEF) was calculated using an ICRU Slab phantom for concentrations of 7, 18, and 30 mg/g. The DEF was higher at higher concentrations of dose enhancement agents and at lower incident energies. The calculated DEF ranged from 1.035 to 1.079, and dose enhancement effects were highest for iron oxide, followed by iodine, gadolinium, and gold. Thus, this study contributes to improving the therapeutic ratio by delivering larger doses of radiation to tumor volume, and provides data to support further in vivo and in vitro studies.

• 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.

• Journal of the Korean Society of Radiology
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• v.7 no.4
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• pp.259-264
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• 2013

Total body irradiation in the treatment of childhood leukemia, which is one of the pre-treatment with stem cell transplantation is being used, the current organization using compensators are treated. However, under the terms of the compensator organization long-term impact on the human body, it is difficult to assess directly. In this study, we use the mathematical simulation of radiation exposures body energy and the distance to the crew and the patient (source surface distance, SSD), and patients with tissue compensators change of the distance along the body of the organ doses were evaluated. As a result, the surface dose of energy 4 MV, SSD 280 cm, tissue compensators and the patient when the distance 30 cm 5.84 G / min showed the highest levels. In addition, patients with tissue compensators and the distance apart when 30 cm TBI represents the ideal dose distribution was found.

• Journal of Radiation Protection and Research
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• v.48 no.4
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• pp.197-203
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• 2023

Background: The thermoluminescent dosimeter (TLD) and Monte Carlo (MC) dosimetry are carried out to determine the occupational dose for personnel in the handling of ¹²⁵I seed sources. Materials and Methods: TLDs were placed in different layers of the Alderson-Rando phantom in the thyroid, lung and also eyes and skin surface. An ¹²⁵I seed source was prepared and its activity was measured using a dose calibrator and was placed at two distances of 20 and 50 cm from the Alderson-Rando phantom. In addition, the Monte Carlo N-Particle Extended (MCNPX 2.6.0) code and a computational phantom with a lattice-based geometry were used for organ dose calculations. Results and Discussion: The comparison of TLD and MC results in the thyroid and lung is consistent. Although the relative difference of MC dosimetry to TLD for the eyes was between 4% and 13% and for the skin between 19% and 23%, because of the existence of a higher uncertainty regarding TLD positioning in the eye and skin, these inaccuracies can also be acceptable. The isodose distribution was calculated in the cross-section of the head phantom when the ¹²⁵I seed was at two distances of 20 and 50 cm and it showed that the greatest dose reduction was observed for the eyes, skin, thyroid, and lungs, respectively. The results of MC dosimetry indicated that for near the head positions (distance of 20 cm) the absorbed dose rates for the eye lens, eye and skin were 78.1±2.3, 59.0±1.8, and 10.7±0.7 µGy/mCi/hr, respectively. Furthermore, we found that a 30 cm displacement for the ¹²⁵I seed reduced the eye and skin doses by at least 3- and 2-fold, respectively. Conclusion: Using a computational phantom to monitor the dose to the sensitive organs (eye and skin) for personnel involved in the handling of ¹²⁵I seed sources can be an accurate and inexpensive method.