• IEIE Transactions on Smart Processing and Computing
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• 제6권1호
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• pp.66-70
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• 2017

Radioactive materials are used in medicine, non-destructive testing, and nuclear plants. Source localization is especially important during nuclear decommissioning and decontamination because the actual location of the radioactive source within nuclear waste is often unknown. The coded-aperture imaging technique started with space exploration and moved into X-ray and gamma ray imaging, which have imaging process characteristics similar to each other. In this study, we simulated $21{\times}21$ and $37{\times}37$ coded aperture collimators based on a modified uniformly redundant array (MURA) pattern to make a gamma imaging system that can localize a gamma-ray source. We designed a $21{\times}21$ coded aperture collimator that matches our gamma imaging detector and did feasibility experiments with the coded aperture imaging system. We evaluated the performance of each collimator, from 2 mm to 10 mm thicknesses (at 2 mm intervals) using root mean square error (RMSE) and sensitivity in a simulation. In experimental results, the full width half maximum (FWHM) of the point source was $5.09^{\circ}$ at the center and $4.82^{\circ}$ at the location of the source was $9^{\circ}$. We will continue to improve the decoding algorithm and optimize the collimator for high-energy gamma rays emitted from a nuclear power plant.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4142-4149
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• 2021

In the radiation protection application, the metal-polymer composites have been developed for their radiation shielding properties. In this research, the elastomer composites doped by 10 ㎛ and 100nm size of lead, bismuth and tungsten particles as filler with 30 and 60 wt percentages were prepared. To survey the shielding properties of the polymer composites using gamma-ray emitted from 152Eu and 137Cs sources, the gamma flux was measured by using NaI(Tl) detector, then the linear attenuation coefficient was calculated. Also, the Monte Carlo simulation (MCs) method was used. The results showed a direct relationship between the linear attenuation coefficients of the absorbent and filler ratio. Also, the decrease in the particle size of the shielding material in each weight percentage improved the radiation shielding features. When the dimension of the particles was in the order of nano-size, more attenuation was achieved. At low energies used for medical diagnostic X-ray applications due to the predominance of the photoelectric effect, bismuth and lead were suitable selection as filler.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.957-962
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• 2018

The gamma radiation shielding capability (GRSC) of two clay-materials (Ball clay and Kaolin)of Southwestern Nigeria ($7.49^{\circ}N$, $4.55^{\circ}E$) have been investigated by determine theoretically and experimentally the mass attenuation coefficient, ${\mu}/{\rho}(cm^2g^{-1})$ of the clay materials at photon energies of 609.31, 1120.29, 1173.20, 1238.11, 1332.50 and 1764.49 keV emitted from $^{214}Bi$ ore and $^{60}Co$ point source. The mass attenuation coefficients were theoretically evaluated using the elemental compositions of the clay-materials obtained by Particle-Induced X-ray Emission (PIXE) elemental analysis technique as input data for WinXCom software. While gamma ray transmission experiment using Hyper Pure Germanium (HPGe) spectrometer detector to experimentally determine the mass attenuation coefficients, ${\mu}/{\rho}(cm^2g^{-1})$ of the samples. The experimental results are in good agreement with the theoretical calculations of WinXCom software. Linear attenuation coefficient (${\mu}$), half value layer (HVL) and mean free path (MFP) were also evaluated using the obtained ${\mu}/{\rho}$ values for the investigated samples. The GRSC of the selected clay-materials have been compared with other studied shielding materials. The cognizance of various factors such as availability, thermo-chemical stability and water retaining ability by the clay-samples can be analyzed for efficacy of the material for their GRSC.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2797-2801
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• 2023

CdZnTeSe (CZTS) has attracted attention for applications in X- and gamma-ray detectors owing to its improved properties compared to those of CdZnTe (CZT). In this study, we grew and processed single crystals of CZT and CZTS using the Bridgeman method to confirm the feasibility of using a dosimeter for high-energy X-rays in radiotherapy. We evaluated their linearity and precision using the coefficient of determination (R²) and relative standard deviation (RSD). CZTS showed sufficient RSD values lower than 1.5% of the standard for X-ray dosimetry, whereas CZT's RSD values increased dramatically under some conditions. CZTS exhibited an R² value of 0.9968 at 500 V/cm, whereas CZT has an R² value of 0.9373 under the same conditions. The X-ray response of CZTS maintains its pulse shape at various dose rates, and its properties are improved by adding selenium to the CdTe matrix to lower the defect density and sub-grain boundaries. Thus, we validated that CZTS shows a better response than CZT to high-energy X-rays used for radiotherapy. Further, the applicability of an onboard imager, a high-energy X-ray (>6 MV) image, is presented. The proposed methodology and results can guide future advances in X-ray dose detection.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.275-279
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• 2001

In uranium enrichment plants and nuclear fuel fabrication facilities, exact measurement of fissile isotope enrichment of uranium is required for material accounting in international safeguards inspection as well as process quality control. The purpose of this study was to develop a simple measurement system which can portably be used at nuclear fuel fabrication plants especially dealing with low enriched uranium. For this purpose, a small size CZT (CdZnTe) detector was used, and the detector performance in low uranium gamma/X -rays energy range was investigated by use of various enriched uranium oxide samples. New enrichment measurement technique and analysis method for low enriched uranium oxide, so-called, 'semi-peak ratio technique' was developed. The newly developed method was considered as an alternative technique for the low enrichment and would be useful to account nuclear material in safeguarding activity at nuclear fuel fabrication facility.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.929-936
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• 2018

A basic characterization of uranium samples was performed using gamma- and X-ray spectroscopy. The studied uranium samples were eight types of certified reference materials with $^{235}U$ enrichments in the range of 1-97%, and the measurements were performed over 24 h using a high-resolution and high-purity planar germanium detector. A general peak analysis of the spectrum and the $XK_{\alpha}$ region of the uranium spectra was carried out by using HyperGam and HyperGam-U, respectively. The standard reference sources were used to calibrate the spectroscopy system. To obtain the absolute detection efficiency, an effective solid angle code, EXVol, was run for each sample. Hence, the peak activities and isotopic activities were determined, and then, the total U content and $^{234}U$, $^{235}U$, and $^{238}U$ isotopic contents were determined and compared with those of the certified reference values. A new method to determine the model age based on the ratio of the activities of $^{223}Ra$ and $^{235}U$ in the sample was studied, and the model age was compared with the known true age. In summary, the present study developed a method for basic characterization of uranium samples by nondestructive gamma-ray spectrometry in 24 h and to obtain information on the sample age.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2489-2497
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• 2024

The current study aims to develop new clay bricks doped with metallic waste for radiation shielding applications. The aforementioned new bricks were fabricated with various metallic waste concentrations under a pressure rate reaching ≈114 MPa and firing temperature of 1100 ℃. The impacts of the metallic waste and the firing temperature on the developed brick samples' physical, radiation shielding, and structural properties were studied. In order to identify the fabricated bricks' mineral content, the X-ray diffraction pattern was used. Additionally, the fabricated bricks' porosity and density were experimentally determined, where the porosity was reduced by 28.03%, while their densities increased by ≈ 10.5% by raising the concentration of metallic waste. The linear attenuation coefficient (LAC) for the developed brick was investigated experimentally using a NaI (Tl) scintillation detector over the 0.033-1.408 MeV energy interval. The measured LAC values were enhanced by increasing the concentrations of metallic waste within the fabricated bricks over the examined energy interval. The fabricated brick's LAC enhancement improves the gamma-ray shielding characteristics. Therefore, the fabricated bricks are a cheap and suitable choice for radiation protection applications.

• Journal of Radiation Protection and Research
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• 제39권3호
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• pp.150-158
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• 2014

각 의료기관 내 방사선 관련 종사자나 방사선 치료환자들은 진단 및 치료 시 필연적으로 의료상 피폭을 수반하게 된다. 국제방사선방호위원회(ICRP) 권고나 국제원자력기구(IAEA)의 기준에 따라 기준선량 제약치를 적용 및 권고 받고 있지만 1차 피폭대상자인 종사자나 환자들의 피폭최적관리를 위해서는 잠재적인 피폭대상자들에게 기존의 피폭관리 시스템보다 직접적이고 가용성이 높은 측정 및 분석 방법이 필요하다. 따라서 본 연구에서는 기존에 구비된 휴대용 단말과 연동하여 원거리에서 실시간으로 방사선 모니터링이 가능한 시스템을 개발하였다. 모니터링 시스템은 검출부, 영상부, 통신부 세 부분으로 구성되었다. 검출부는 시스템의 소형화를 위해 실리콘 광증배소자(silicon photomultiplier) 기반 섬광검출기를 설계하였으며, 영상부는 무선 CCD (charge-coupled device)카메라 모듈을 사용하여 검출부와 함께 Bluetooth 통신모듈을 통해 휴대용 단말로 측정된 방사선 정보와 영상이 전송된다. 제작된 시스템은 성능 평가를 위해 진단용 X-ray 발생장치와 $^{137}Cs$, $^{22}Na$, $^{60}Co$, $^{204}Tl$, $^{90}Sr$ 선원을 사용하였다. 측정결과를 통해 개발된 시스템은 gamma, beta, X-ray에 대해서 검출 반응성을 확인하였고, 방사선 세기에 따른 응답 선형성과 MCNPX 전산코드를 이용한 측정 거리에 따른 시스템의 검출 정확도 평가 시 3% 내외의 오차범위를 확인하였다. 본 연구의 결과는 방사선 검출 시스템 구성의 비용절감 효과와 개인피폭정도관리에 기여할 것으로 기대한다.

• Journal of Radiation Protection and Research
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• 제48권4호
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• pp.167-174
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• 2023

Background: A nondestructive test is commonly used to inspect the surface defects and internal structure of an object without any physical damage. X-rays generated from an electron accelerator or a tube are one of the methods used for nondestructive testing. The high penetration of X-rays through materials with low atomic numbers makes it difficult to discriminate between these materials using X-ray imaging. The interaction characteristics of neutrons with materials can supplement the limitations of X-ray imaging in material discrimination. Materials and Methods: The radiation image acquisition process for air-cargo security inspection equipment using X-rays and neutrons was simulated using a GEometry ANd Tracking (Geant4) simulation toolkit. Radiation images of phantoms composed of 13 materials were obtained, and the R-value, representing the attenuation ratio of neutrons and gamma rays in a material, was calculated from these images. Results and Discussion: The R-values were calculated from the simulated X-ray and neutron images for each phantom and compared with those obtained in the experiments. The R-values obtained from the experiments were higher than those obtained from the simulations. The difference can be due to the following two causes. The first reason is that there are various facilities or equipment in the experimental environment that scatter neutrons, unlike the simulation. The other is the difference in the neutron signal processing. In the simulation, the neutron signal is the sum of the number of neutrons entering the detector. However, in the experiment, the neutron signal was obtained by superimposing the intensities of the neutron signals. Neutron detectors also detect gamma rays, and the neutron signal cannot be clearly distinguished in the process of separating the two types of radiation. Despite these differences, the two results showed similar trends and the viability of using simulation-based radiation images, particularly in the field of security screening. With further research, the simulation-based radiation images can replace ones from experiments and be used in the related fields. Conclusion: The Korea Atomic Energy Research Institute has developed air-cargo security inspection equipment using neutrons and X-rays. Using this equipment, radiation images and R-values for various materials were obtained. The equipment was reconstructed, and the R-values were obtained for 13 materials using the Geant4 simulation toolkit. The R-values calculated by experiment and simulation show similar trends. Therefore, we confirmed the feasibility of using the simulation-based radiation image.

• Radiation Oncology Journal
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• 제3권2호
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• pp.145-151
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• 1985

Wedge filter를 사용할 때 $^{60}Co\;\gamma$선과 10MV X-선의 주변선량분포를 반자동식 물팬톰장치에 의해 제어되는 고체 방사선검출기로 측정하였다. wedge filter의 날과 등의 방향을 잇는 주단면상에서 주변선량을 측정하였다. 방사선의 투과력과 wedge 각, wedge에 대한 방향, 조사면의 변화에 대하여 주변선량의 면화를 고찰하였다. 선축에 수직인 방향의 측정은 깊이 dm, 5cm, 10cm, 15cm에서 선측과 평행인 방향에서는 조사면 경계에서 3cm, 5cm, 10cm 떨어진 위치에서 측정이 이루어졌다. 측정으로부터 wedge가 사용되는 조사면에 대한 주변선량분포가 아래와 같음을 볼 수 있었다. 1. wedge filter의 날의 방향의 주변선량이 등의 방향의 주변선량보다 높았다. 2. 표면근처에서는 깊이가 깊어짐에 따라 선량이 감소하였다. 3. $^{60}Co\;\gamma$선의 경우 어떤 깊이에서는 조사면의 경계로부터 거리가 떨어짐에 따라 주변선량이 단조감소하지 않고 다시 증가하는 영역이 있었다. 4. wedge를 사용하는 경우 주변선량은 방사선의 투과력이나 조사면의 크기뿐만 아니라 wedge의 각과 방향에도 영향을 받는다.