• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2404-2411
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• 2024

The GEOUNED code is specifically designed to convert CAD models, defined using the B-rep approach, into MC radiation transport models, defined using the CSG approach, and vice versa from MC to CAD. This code incorporates standard features commonly found in conversion tools, including decomposition, conversion, and automatic void generation. Additionally, it introduces innovative features, mainly in the automatic void generation part, which are described in this article. GEOUNED has demonstrated successful application in highly detailed 3D models used in fusion neutronics, which are known for their complex geometries, particularly those utilized in ITER. The article includes examples showcasing GEOUNED's performance in these challenging models, as well as custom applications that highlight its flexibility in addressing non-standard problems. The code is open-source and utilizes Open CASCADE as the geometry engine, with FreeCAD serving as the Python API.

• 방사선산업학회지
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• 제4권1호
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• pp.33-38
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• 2010

The MCNP5 neutron, electron, photon Monte Carlo transport program was installed on the KISTI's SUN Tachyon computer using the parallel programming. Electron beam accelerators were modeled and shielding calculations were performed in order to investigate the reduction of computation time in the supercomputer environment. It was observed that a speedup of 40 to 80 of computation time can be obtained using 64 CPUs compared to an IBM PC.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.310-314
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• 2016

Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

• Journal of Radiation Protection and Research
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• 제45권2호
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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.

• 대한방사선기술학회지:방사선기술과학
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• 제47권3호
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• pp.213-218
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• 2024

The myocardial nuclear medicine examination is widely performed to diagnose myocardium disease using various radionuclides. Although image quality according to radionuclides has improved, the radiation exposure for target organ as well as peripheral organs should be considered. Here, the aim of this study was to evaluate absorbed dose (Gy) for peripheral organs in myocardial nuclear medicine scan from myocardium according to various scan environments based on Monte Carlo simulation. The simulation environment was modeled 5 cases, which were considered by radionuclides, number of injections, and radiodosage. In addition, the each radionuclide simulation such as distribution fraction was considered by recommended standard protocol, and the mesh computational female phantom, which is provided by International Commission on Radiological Protection (ICRP) 145, was used using the particle and heavy ion transport code system (PHITS) version 3.33. Based on the results, the closer to the myocardium, the higher the absorbed dose values. In addition, application for dual injection for radionuclides leaded to high absorbed dose compared with single injection for radionuclide. Consequently, there is difference for absorbed dose according to radionuclides, number of injections, and radiodosage. To detect the accurate diseased area, acquisition for improved image quality is crucial process by injecting radionuclides, however, we need to consider absorbed dose both target and peripheral inner organs from radionuclides in terms radiation protection for patient.

• Journal of Radiation Protection and Research
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• 제10권1호
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• pp.41-49
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• 1985

1,300 MWe 가압경수로 공동내에서 중성자의 흐름해석이 수행되었다. 중성자의 흐름을 해석하는데는 1차원 수송코드인 ANISN, 2차원 수송코드인 DOT3.5, 3차원 Monte Carlo 코드인 TRIPOLI-02와 이들을 접속시켜주는 DOTTRI 등의 전산코드가 이용되었고, 본 계산에 사용된 전산기는 IBM 3033형이었다. 계산된 선속 및 선량율은 900 MW 가압경수로의 공동내에서 측정한 측정치와 비교검토 되었고, 그 결과 중성자 군별로 약간의 오차는 있었으나 전체적으로 큰 오차는 없었다. 이 결과는 대용량의 원자로 차폐설계, 원자로보수시, 기타 원자로 공동내에 출입할 경우에 방사선방어상 필요한 방어수단을 제공하는데 기여하였다.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1885-1891
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• 2023

The present work aims to fabricate Na_1+xZr₂Si_xP_3-xO₁₂ compound at various calcination temperatures based on the zircon mineral. The fabricated compound was calcinated at 250, 500, and 1000℃. The effect of calcination temperature on the structure, crystal phase, and radiation shielding properties was studied for the fabricated compound. The X-ray diffraction diffractometer demonstrates that, the monoclinic crystal phase appeared at a calcination temperature of 250℃ and 500℃ is totally transformed to a high-symmetry hexagonal crystal phase under a calcination temperature of 1000℃. The radiation shielding capacity was also qualified for the fabricated compounds using the Monte Carlo N-Particle transport code in the g-photons energy interval between 15keV and 122keV. The impacts of calcination temperature on the g-ray shielding behavior were clarified in the present study, where the linear attenuation coefficient was enhanced by 218% at energy of 122keV, when the calcination temperature increased from 250 to 1000℃, respectively.

• Journal of Radiation Protection and Research
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• 제18권2호
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• pp.17-25
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• 1993

액체섬광 계수기에 대한 $^3H,\;^{14}C,\;^{36}Cl$의 베타선 스펙트럼의 시뮬레이션을 Monte Carlo 법으로 수행하였다. 베타선의 손실에너지는 'continuous slowing down approximation(CSDA)' 모형으로부터 구하였다. 순 베타방출 핵종이 균일하게 용해되어 있는 섬광용액은 $1.25cm{\times}3.5cm$의 원통형 geome try로 가정하였다. $^{14}C,\;^{36}Cl$에 대한 계산결과는 계수효율의 경우 실험 값과 2%의 범위 내에서 잘 일치하였으나 $^3H$에 대해서는 상당한 편차를 나타내었다. 이는 저 에너지 베타선에 대한 섬광용액 자체의 quenching 효과에 기인하는 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2361-2369
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• 2020

Compton cameras using large scintillators have been developed for high imaging sensitivity. These scintillator-based Compton cameras, however, mainly due to relatively low energy resolution, suffer from undesired background-radiation signals, especially when radioactive materials' activity is very low or their location is far from the Compton camera. To alleviate this problem for a large-size Compton camera, in the present study, a hybrid-type shielding system was designed that combines an active shield with a veto detector and a passive shield that surrounds the active shield. Then, the performance of the hybrid shielding system was predicted, by Monte Carlo radiation transport simulation using Geant4, in terms of minimum detectable activity (MDA), signal-to-noise ratio (SNR), and image resolution. Our simulation results show that, for the most cases, the hybrid shielding system significantly improves the performance of the large-size Compton camera. For the cases investigated in the present study, the use of the shielding system decreased the MDA by about 1.4, 1.6, and 1.3 times, increased the SNR by 1.2-1.9, 1.1-1.7, and 1.3-2.1 times, and improved the image resolution (i.e., reduced the FWHM) by 7-8, 1-6, and 3-5% for ¹³⁷Cs, ⁶⁰Co, and ¹³¹I point source located at 1-5 m from the imaging system, respectively.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.