• Journal of Radiation Protection and Research
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• v.46 no.3
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• pp.98-105
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• 2021

Background: In recent events of the coronavirus disease 2019 (COVID-19) pandemic, computed tomography (CT) scans are being globally used as a complement to the reverse-transcription polymerase chain reaction (RT-PCR) tests. It will be important to be aware of major organ dose levels, which are more relevant quantity to derive potential long-term adverse effect, for Korean pediatric and adult patients undergoing CT for COVID-19. Materials and Methods: We calculated organ dose conversion coefficients for Korean pediatric and adult CT patients directly from Korean pediatric and adult computational phantoms combined with Monte Carlo radiation transport techniques. We then estimated major organ doses delivered to the Korean child and adult patients undergoing CT for COVID-19 combining the dose conversion coefficients and the international survey data. We also compared our Korean dose conversion coefficients with those from Caucasian reference pediatric and adult phantoms. Results and Discussion: Based on the dose conversion coefficients we established in this study and the international survey data of COVID-19-related CT scans, we found that Korean 7-year-old child and adult males may receive about 4-32 mGy and 3-21 mGy of lung dose, respectively. We learned that the lung dose conversion coefficient for the Korean child phantom was up to 1.5-fold greater than that for the Korean adult phantom. We also found no substantial difference in dose conversion coefficients between Korean and Caucasian phantoms. Conclusion: We estimated radiation dose delivered to the Korean child and adult phantoms undergoing COVID-19-related CT examinations. The dose conversion coefficients derived for different CT scan types can be also used universally for other dosimetry studies concerning Korean CT scans. We also confirmed that the Caucasian-based CT organ dose calculation tools may be used for the Korean population with reasonable accuracy.

• Progress in Medical Physics
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• v.13 no.3
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• pp.149-155
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• 2002

Recent advances in radiation transport algorithms, computer hardware performance, and parallel computing make the clinical use of Monte Carlo based dose calculations possible. To compare the speed and accuracies of dose calculations between different developed codes, a benchmark tests were proposed at the XIIth ICCR (International Conference on the use of Computers in Radiation Therapy, Heidelberg, Germany 2000). A Monte Carlo treatment planning comprised of 28 various Intel Pentium CPUs was implemented for routine clinical use. The purpose of this study was to evaluate the performance of our system using the above benchmark tests. The benchmark procedures are comprised of three parts. a) speed of photon beams dose calculation inside a given phantom of 30.5 cm$\times$39.5 cm $\times$ 30 cm deep and filled with 5 ㎣ voxels within 2% statistical uncertainty. b) speed of electron beams dose calculation inside the same phantom as that of the photon beams. c) accuracy of photon and electron beam calculation inside heterogeneous slab phantom compared with the reference results of EGS4/PRESTA calculation. As results of the speed benchmark tests, it took 5.5 minutes to achieve less than 2% statistical uncertainty for 18 MV photon beams. Though the net calculation for electron beams was an order of faster than the photon beam, the overall calculation time was similar to that of photon beam case due to the overhead time to maintain parallel processing. Since our Monte Carlo code is EGSnrc, which is an improved version of EGS4, the accuracy tests of our system showed, as expected, very good agreement with the reference data. In conclusion, our Monte Carlo treatment planning system shows clinically meaningful results. Though other more efficient codes are developed such like MCDOSE and VMC＋＋, BEAMnrc based on EGSnrc code system may be used for routine clinical Monte Carlo treatment planning in conjunction with clustering technique.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.196-203
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• 1982

To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. k$_{eff}$ was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79.9.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.344-349
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• 2016

Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient's body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.44-51
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• 2023

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique. Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations. Results and Discussion: Damage yields of about 5.00×10^-4 were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive ⁶⁰Co by the aptamer is the same with the non-radioactive ⁵⁹Co prove the accuracy of the previous DNA damage calculation. Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste.

• Progress in Medical Physics
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• v.33 no.4
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• pp.129-135
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• 2022

Purpose: A full-energy-peak (FEP) efficiency correction is required through a Monte Carlo simulation for accurate radioactivity measurement, considering the geometrical characteristics of the detector and the sample. However, a relative deviation (RD) occurs between the measurement and calculation efficiencies when modeling using the data provided by the manufacturers due to the randomly generated dead layer. This study aims to optimize the structure of the detector by determining the dead layer thickness based on Monte Carlo simulation. Methods: The high-purity germanium (HPGe) detector used in this study was a coaxial p-type GC2518 model, and a certified reference material (CRM) was used to measure the FEP efficiency. Using the MC N-Particle Transport Code (MCNP) code, the FEP efficiency was calculated by increasing the thickness of the outer and inner dead layer in proportion to the thickness of the electrode. Results: As the thickness of the outer and inner dead layer increased by 0.1 mm and 0.1 ㎛, the efficiency difference decreased by 2.43% on average up to 1.0 mm and 1.0 ㎛ and increased by 1.86% thereafter. Therefore, the structure of the detector was optimized by determining 1.0 mm and 1.0 ㎛ as thickness of the dead layer. Conclusions: The effect of the dead layer on the FEP efficiency was evaluated, and an excellent agreement between the measured and calculated efficiencies was confirmed with RDs of less than 4%. It suggests that the optimized HPGe detector can be used to measure the accurate radioactivity using in dismantling and disposing medical linear accelerators.

• 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 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.

• Bulletin of the Korean Space Science Society
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• 2011.04a
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• pp.28.1-28.1
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• 2011

Radiation-induced displacement damage which has caused the increase of the dark current in the focal plane adopted in the Ozone Monitoring Instrument (OMI) was studied in regards of the primary protons and the secondaries generated by the protons in the orbit. By using the Monte Carlo N-Particle Transport Code System (MCNPX) version 2.4.0 along with the Stopping and Range of Ions in Matter version 2010 (SRIM2010), effects of the primary protons as well as secondary particles including neutron, electron, and photon were investigated. After their doses and fluxes that reached onto the charge-coupled device (CCD) were examined, displacement damage induced by major sources was presented.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.97-102
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• 1984

Criticality analyses of the KSC-1(Korean Shipping Cask-1) spent fuel shipping cask have been performed with the help of KENO-IV Monte Carlo computer code and 19-group CSLIB 19 cross section set which was generated from AMPX modular system. The analyses followed a benchmark calculation which has been made regard to the B & W CX-10 criticality facility in order to validate the Monte Carlo code cross section set described above. The KSC-1 shipping cask seems to be safe in the criticality point of view for the transport of one PWR spent fuel assembly under the normal conditions as well as the hypothetical accident conditions.