• The Korean Journal of Nuclear Medicine
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• v.33 no.1
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• pp.100-109
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• 1999

Purpose: Preamplifier and amplifier are very important parts for developing a portable counting or imaging gamma probe. They can be used for analyzing pulses containing energy and position information for the emitted radiations. The commercial Nuclear Instrument Modules (NIMs) can be used for processing these pulses. However, it may be improper to use NIMs in developing a portable gamma probe, because of its size and high price. The purpose of this study was to develop both preamplifier and amplifier and measure their performance characteristics. Materials and Methods: The preamplifier and amplifier were designed as a charge sensitive device and a capacitor resistor-resistor capacitor (CR-RC) electronic circuit, respectively, and they were mounted on a print circuit board (PCB). We acquired and analyzed energy spectra for Tc-99m and Cs-137 using both PCB and NIMs. Multichannel analyzer (Accuspec/A, Canberra Industries Inc., Meriden Connecticut, U.S.A) and scintillation detectors (EP-047(Bicron Saint-Gobain/Norton Industrial EP-047 (Ceramics Co., Ohio, U.S.A) with $2"{\times}2"$ NaI(T1) crystal and R1535 (Hamamatsu Photonics K.K., Electron Tube Center, Shizuoka-ken, Japan) with $1"{\times}1"$ NaI(T1) crystal were used for acquiring the energy spectra. Results: Using PCB, energy resolutions of EP-047 detectors for Tc-99m and Cs-137 were 12.92% and 5.01%, respectively, whereas R1535 showed 13.75% and 5.19% of energy resolution. Using the NIM devices, energy resolutions of EP-047 detector for Tc-99m and Cs-137 were measured as 14.6% and 7.58%, respectively. However, reliable energy spectrum of R1535 detector could not be acquired, since its photomultiplier tube (PMT) requires a specific type of preamplifier. Conclusion: We developed a special preamplifier and amplifier suitable for a small sized gamma probe that showed good energy resolutions independent of PMT types. The results indicate that the PCB can be used in developing both counting and imaging gamma probe.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.73-78
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• 2015

In this study, a scintillation detector was fabricated using a silicon photomultiplier (SiPM) and a Ce:GAGG scintillator single crystal, and its spectroscopic properties were compared with those of commercially available LYSO and CsI:Tl scintillators using ${\gamma}$-ray spectroscopy. The energy resolutions of the self-produced scintillation detector composed of the scintillator single crystal (volume: $3{\times}3{\times}20mm^3$) and SiPM (Photosensitive area: $3{\times}3mm^2$) for standard ${\gamma}$-ray sources, such as $^{133}Ba$, $^{22}Na$, $^{137}Cs$ and $^{60}Co$ were measured and compared. As a result, the energy resolutions of the proposed Ce:GAGG scintillation detector for g-rays, as measured using its spectroscopic properties, were found to be 13.5% for $^{133}Ba$ 0.356 MeV, 6.9% for $^{22}Na$ 0.511 MeV, 5.8% for $^{137}Cs$ 0.662 MeV and 2.3% for $^{60}Co$ 1.33 MeV.

• Progress in Medical Physics
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• v.13 no.1
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• pp.44-50
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• 2002

CsI(T1) single crystal was grown in a Bridgman growing apparatus, which has the diameter of 11 mm and the mole ratio of 0.001 mol%. Radiation sensors were made with CsITl)crystal and two photodiodes, and measured spectroscopic characteristics and linearity for gamma-ray and X-ray. The energy resolution of CsI(T1) radiation sensor has been measured with $^{22}$ Na, $^{137}$ Cs and $^{60}$ Co gamma standard sources. Also output linearity of CsITl) sensor was measured for diagnostic radiation region. The energy resolutions of CsI(T1) radiation sensor for 0.511MeV gamma-ray from Na-22 source, 0.662MeV from Cs-137 source, and 1.17MeV and 1.332MeV from Co-60 source were 13.2%, 8.3%, 6.7%, and 5.1% respectively. Also the output linearity up to 80mAs current for 60kVp, 80kvp, 100kVp, 120kVp tube voltages has been studied.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.437-438
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• 2012

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.54-60
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• 2014

IAEA has employed various types of radiation detectors - HPGe, NaI, CZT - for accountancy of nuclear material. Among them, HPGe has been mainly used in verification activities required for high accuracy. Due to its essential cooling component(a liquid-nitrogen cooling or a mechanical cooling system), it is large and heavy and needs long cooling time before use. New hand-held portable HPGe has been developed to address such problems. This paper deals with results of performance evaluation test of the new hand-held portable HPGe prototype which was used during IAEA's inspection activities. Radioactive spectra obtained with the new portable HPGe showed different characteristics depending on types and enrichments of nuclear materials inspected. Also, Gamma-rays from daughter radioisotopes in the decay series of $^{235}U$ and $^{238}U$ and characteristic x-rays from uranium were able to be remarkably separated from other peaks in the spectra. A relative error of enrichment measured by the new portable HPGe was in the range of 9 to 27%. The enrichment measurement results didn't meet partially requirement of IAEA because of a small size of a radiation sensing material. This problem might be solved through a further study. This paper discusses how to determine enrichment of nuclear material as well as how to apply the new hand-held portable HPGe to safeguard inspection. There have been few papers to deal with IAEA inspection activity in Korea to verify accountancy of nuclear material in national nuclear facilities. This paper would contribute to analyzing results of safeguards inspection. Also, it is expected that things discussed about further improvement of a radiation detector would make contribution to development of a radiation detector in the related field.

• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.57-66
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• 1988

A series of experiments is performed to measure radon concentration in the near-surface soil gas at the four locations (Cheju-Si, Seoguipo-Si, Taejeong-eup, Seongsan-eup) in Cheju Island, using CR-39 detectors placed inside radon cups. Two types of radon cups are installed in shallow holes of about 15 cm in diameter and 50cm in depth. The optimum etching conditions, i.e., the concentration of NaOH solution, etchant temperature and etching time, are found to be 625N, $70^{\circ}C$ and 5.5 hours for CR-39 detectors. A typical conversion factor of radon cup is calculated as $$1track/mm^3{\cdot}30day=0.059Bq/{\ell}$$. Average radon concentrations over 30 days measured in Cheju Island from May 1, 1987 to April 23, 1988 are $3.1{\pm}0.3Bq/{\ell}$ for open radon cups and $1.7{\pm}0.2Bq/{\ell}$ for closed radon cups.

• Journal of radiological science and technology
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• v.44 no.5
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• pp.459-464
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• 2021

The detector performance is important role in acquiring the gamma rays from patients. Among parameters of detector performances, there is density, which relates to respond to gamma rays. Therefore, we confirm the detection efficiency according to various detector materials based on the density parameter using GATE (geant4 application for emission tomography) simulation tool. The NaI (density: 3.67 g/cm³), CZT (Cadimium Zinc Telluride) (density: 5.80 g/cm³), CdTe (Cadmium Telluride) (5.85 g/cm³), and GAGG (Gadoinium Aluminum Gallium Garnet) (density g/cm³) were used as detector materials. In addition, the point source and quadrant bar phantom, which is modeled for 0.5, 1.0, 1.5, and 2.0 mm thicknesses, were modeled to confirm the quatitative analysis using sensitivity (cps/MBq) and the full width at half maximum (FWHM, mm) at the 2.0 mm bar thickness containing visual evaluation. Based on the results, the sensitivity for NaI, CZT, CdTe, and GAGG detector materials were 0.12, 0.15, 0.16, and 0.18 cps/MBq. In addition, the FWHM for quadrant bar phantom in the 2.0 mm bar thickness is 3.72, 3.69, 3.70, and 3.73 mm for NaI, CZT, CdTe, and GAGG materials, respectively. Compared with performance of detector materials according to density, the high density can improve detection efficiency in terms of sensitivity and mean count. Among these detector materials, the GAGG material is efficient for detection of gamma rays.

• Analytical Science and Technology
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• v.18 no.1
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• pp.59-65
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• 2005

When the low level radioactivity sample is measured, it is required to have many samples. For increase of the sample volume, a scattering and absorbing probability of the emitted gamma-ray in the sample are to be increased. In order to correct the self-absorption effect, the counting efficiency must be calibrated according to a geometrical condition and sample density. But, it is impossible to determine efficiency for counting sample using standard source with the same geometrical condition and density. In this study, the measuring efficiencies were determined with various counting containers and densities. In order to compare the self-absorption effect with the sample density in the various sample container, the variation of the counting efficiency with the densities was investigated by adding NaI, which has high solubility and density. Also, they were compared with Monte Carlo simulation. The self-absorption effect was found to be significant in the low energy region below 0.5 MeV.

• Journal of radiological science and technology
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• v.45 no.1
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• pp.49-55
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• 2022

This study is to develop a mobile type environmental radiation measurement system for emergency response or environmental radiation monitoring of local governments near nuclear facilities. A mobile radiation measurement system can monitor radiation by field beyond the spatial constraints of a fixed environmental radiation monitor. If installed in local government infrastructure such as public transportation, environmental radiation can be monitored without additional manpower and measurement work. In addition, it is designed to enable monitoring and measurement of radiation from low to high doses as well as the environment in preparation for radioactive disasters such as nuclear power plant accidents. It is expected that this system will be utilized not only in normal times but also in the event of a radiation accident to improve the disaster prevention capabilities of local governments.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.526-531
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• 1996

국내의 방사성 동위원소 (Radioisotope; RI) 사용기관에서 발생, 수거되어 현재 원자력연구소 저장고 내에 보관중인 폐기물중 가연분을 대상으로 소각실증실험을 수행하였다. 폐기물 발생일을 기준하여 최소 2년이상이 경과한 RI 폐기물을 소각한 결과 폐기물의 부피와 무게를 각각 약 1/250, 1/30로 줄일 수 있었으며, 소각운전중 배출기체의 농도는 환경관련법상의 배출허용기준치 이하, 배출방사선량은 검출기의 측정한계 이하였다. 방사성동위원소 추적자로 I-131 ($Na^{131}$ I형태, 총 2mCi)을 사용한 모의 폐기물을 소각한 결과 대부분의 방사능은 소각재에 잔존하기 보다는 휘발되어 배기체와 함께 거동하며 주로 포대여과기의 여과층에서 잡히는 것으로 나타났다.