• Journal of Radiation Protection and Research
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• v.29 no.1
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• pp.17-23
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• 2004

In order to save time and money needed in the purchase commonly used gamma-ray standard sources, a new radioactive standard source was manufactured by the neutron activation of some regent in the research reactor HANARO. The source was manufactured with an aqueous solution by mixing and dissolving the irradiated reagents. The manufactured source was compared with a commercial standard source. It was confirmed that it could be used as an efficiency calibration source. Also, in order to compare the variation of efficiency due to the volume difference for various containers used in radioactivity assay, the efficiency variation as a function of sample volume was investigated.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.801-805
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• 1998

1000 MWe 국내 표준형 원전을 대상으로 노심이 손상되는 각종 중대사고 시나리오에 대하여 방사선원항 특성을 평가하기 위하여, 2단계 확률론적 안전성 평가 방법론에 따라 방사선원 방출군을 정의하고 원전 중대사고 발생시 격납건물 손상을 가정하여 각 방출군별로 격납건물 외부로 방출되는 방사능 방출율을 정량화하였다. 도출된 19개의 그룹중에서 방출률이 작거나 발생빈도가 낮은 7개를 제외하고 12가지 대표 사고경위에 대하여 계산을 수행하였으며, 분석결과는 격납건물 내에서 감쇄효과가 작은 증기발생기 세관 파단사고, 격납건물 격리 실패사고 및 조기 격납건물 파손사고 둥이 상대적으로 큰 방사능 방출량을 보여주었다

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.43-51
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• 2004

In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1＆2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.345-346
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• 2016

• Progress in Medical Physics
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• v.12 no.2
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• pp.141-146
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• 2001

It is well known that assurance of the radiation therapy needs for an accuracy of $\pm$ 5 % in the delivery of an absorbed dose to target volume. Therefore, the dose evaluation of brachytherapy source and/or linear accelerate beam must be a stability with accuracy. In an advanced country, they recommended to use the radioactive check source for reference air ionization chamber for a stable response of radiation field chamber. In this experiments, the radioactive source Sr-90 and PR-05 air ionization chamber were used for standard source and reference ion chamber. The response of reference ion chamber showed as an 1.000$\pm$ 0.010 uncertainty for 10 years long and the evaliuation f dose discrepancy of clinical field ion chamber showed as 0.997 $\pm$0.011 in a $^{60}$ Co brachytherapy soruce. In our experiments, we can assuarance the long halflife standard source is reliable to preserve the calibration factor of reference chamber in stability.

• Journal of the Korean Society of Radiology
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• v.15 no.3
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• pp.307-312
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• 2021

In nuclear medicine, radioactive isotope tracers are administered to the human body to obtain and evaluate disease morphological information and biological function information. Dose calibrator is a device used to measure the radioactivity of a single nuclide in medical institutions. Administration of the correct dose to the human body acts as an important factor in diagnosis and treatment, and measurement through a dose calibrator before administration is the most important factor. Dose calibrator performs daily quality control after installation in each medical institution. Quality control is a means of guaranteeing quality control after installation, and is essential for improving the quality of treatment and promoting patient safety. Therefore, accurate and standardized performance evaluation methods should be established. In this study, 3D printing was used for quantitative evaluation of quality control by increasing the accuracy and standardization of quality control. When the 3D printer was installed and reproducibility was tested, the error range of the expected value and reading value decreased by 0.302% in the F-18 nuclide and 0.09% in the ^99mTc-pertechnate nuclide than when the 3D printer was installed. The error rate for other nuclides was also found to have a low error rate for reproducibility tests when 3D printing was installed.

• 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.

• The Korean Journal of Nuclear Medicine Technology
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• v.24 no.1
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• pp.20-26
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• 2020

Purpose It is intended to figure out the errors derived from changes in depth and volume when measuring the Standard source and ^99mTc-pertechnetate by using a Dose calibrator. Then recommend appropriate measurement depth and volume. Materials and Methods As a Dose calibrator, CRC-15βeta and CRC-15R (Capintec, New Jersey, USA) was used, and the measurement sources were ⁵⁷Co, ¹³³Ba, ¹³⁷Cs and ^99mTc-pertechnetate was also adopted due to its high frequency of use. The Standard source was respectively measured the changes according to its depth without changing the volume, in a range of 0 cm to 15 cm from the bottom of the ion chamber. ^99mTc-pertechnetate was measured at each depth by changing the volume with 0.1 mL, 0.3 mL, 0.5 mL, 0.7 mL and 0.9 mL Respectively. And the depth range was from 0 cm to 15 cm at the bottom of the ion chamber. Results In the case of Standard source ⁵⁷Co, ¹³³Ba, ¹³⁷Cs and ^99mTc-pertechnetate, there were significant differences according to the measurement depth(p<0.05). ^99mTc-pertechnetate has a negative correlation coefficient according to the depth, and the error of the measured value was negligible at a depth from 0 cm to 7 cm at 0.3 mL and 0.5 mL, and the range of error increased as the volume increased. Conclusion In clinical practice, it is sometimes installed differently than the Standard depth recommended by the equipment company. If it's measured at the recommended depth and volume, it could be thought that unnecessary exposure of the operator and the patient will be reduced, and more accurate radiation exams will be possible in quantitative analysis.

• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.51-58
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• 2001

In the low level radioactivity measurement, such as environmental radioactivity, there were used commonly cylindrical and Marinelli type beakers by means of measurement container. If there are differences in the matrix density or sample height between standard source and sample, it must be determined full energy peak efficiency considering self absorption effect. In this paper, we compared measured efficiency with calculated full energy peak efficiencies in the HPGe detector using the Monte Carlo method. For cylindrical container, we calculated the variation of the efficiency with sample height. Also, we calculated the variation of the detection efficiency with apparent density in the cylindrical and Marinelli container. It was seen that it need to be corrected for self absorption in the energy range of below 1000keV. Also, in order to verify the validity of calculation, we compared the calculated value with reference value using NIST SRM 4353 reference soil.

• Journal of the Korean Society of Radiology
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• v.8 no.6
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• pp.293-299
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• 2014

Compton suppression device is a device by using the Compton scattering reaction and suppress the Compton continuum portion of the spectrum, so can be made more clear analysis of gamma ray peak in the Compton continuum region. Measurements above background occurs or, radioactivity counts of radioactivity concentration value of $^{40}K$ nuclides $^{137}Cs$ and natural radioactivity artificial radioactivity detected from the surface soil sample, unwanted non-target analysis and interference peak who dotted line you know the calibration of the measurement energy is allowed to apply the (Compton suppression) non-suppressed spectrum inhibition spectrum and (Compton Unsuppression) the background to the measured value of the activity concentration value of the standard-ray source is detected relative to the peak of By measuring according to the different distances cause $^{137}Cs$, and comparative analysis of the Monte Carlo simulation, in order to obtain a detection capability for efficient, looking at the Compton inhibitor, as the CSF value increases with increase in the distance, more It was found that the background due to Compton continuum of the measured spectrum suppression mode Compton unrestrained mode can know that the Compton suppression many were made, using a $^{137}Cs$ is reduced.