• Proceedings of the KIEE Conference
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• 2006.04a
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• pp.324-326
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• 2006

CCD형 영상소자는 방사선 피폭 시 표면과 격자내부에 모두 손상을 받게 되며, 감마방사선이나 X선과 같은 고에너지의 이온화 방사선에 노출될 경우 격자 실리콘 내부에 전자-전공쌍(Electron-hole pair, EHP)이 발생된다. 이러한 EHP는 CCD의 순간 출력 광전류로 변환되어 백색 화소 형태의 영상잡음으로 가시화되며, 이 화소 수는 피폭 방사선량에 비례하여 증가하는 특성을 지니고 있다. 따라서 출력 영상정보를 분석하면 조사된 방사선의 양과 특성을 측정할 수 있다. 본 연구에서는 CCD를 이용하여 가상의 방사능 물질 누출 공간에서 방사선원의 방향과 거리정보를 고속으로 탐지하기 위한 장치와 고속 측정 알고리즘을 구현하고 실제 방사선장에서 실증시험을 수행하였다. 방사선 탐지기는 콘형 납 콜리메이터(Collimator)와 가시광 변환용 신틸레이터(CsITl) 및 차폐체로 구성된 센서부와 제어 및 방사광 신호처리를 수행하는 PC부로 구성된다. 감마방사선($^{60}Co$) 방사선장 실증시험에서 방사선원간 거리 83cm에서 측정된 거리 탐지는 5.3%의 오차로 확인되었다. 이 방사선 탐지기는 임의의 고방사선 누출사고에 대한 초기대응 작업을 수행하기 위한 무인 이동로봇용 방사선 탐지기로 활용이 가능하다.

• Journal of Digital Convergence
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• v.10 no.11
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• pp.449-457
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• 2012

The purpose of this study is to investigate the relationship between radiation origin and health professionals, and to reduce exposed dose of radiation through efficient management. Increasing exposed dose of radiation to health professionals are caused by the increase of PET/CT use and a radioactive isotope. Hence, in this study, space dose from each origin of radiation generating was analyzed and the use of personnel protective clothing and shields was compared. As a result of this study, we confirmed that the exposed dose of radiation was much higher in case of wearing personnel protective clothing(0.5 mm pb) than no wearing personnel protective clothing under high energy gamma radiation(511 keV) of the position emitter($^{18}F$).

• The Korean Journal of Nuclear Medicine Technology
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• v.22 no.1
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• pp.46-50
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• 2018

Purpose Generally, a collimator that installed in front of detector set a direction of gamma ray and remove a scatter ray. By the way, a lateral or oblique scatter ray is detected into crystal through collimator. At this study, we will evaluate a mount of count and spectrums of lateral scatter ray. Materials and Methods We used the SKY LITE (philips, netherlands) as a gamma camera, and $^{99m}Tc$, 1.11 GBq point source as a phantom. we put this point source at backside 50 cm of detector. After acquiring this for 1 min, we turned a detector next 10 degrees. Likely this, we acquired images at every 10 degrees from $0^{\circ}$ to $360^{\circ}$, analyzed images and spectrums. In case of patient study, we choose a 3 phase bone scan patient who had a hand disease, because scatter rays from body would detect on crystal. After acquiring blood flow and blood pool images, we analyzed images and spectrums. Additional, we put a lead gown on patient's hand, body. And then we compared and evaluated 3 type blood pool images (non lead gown, lead gown on a hand and on body). Results In case of phantom study, scatter ray counts at backside ($270^{\circ}-90^{\circ}$) are same with a background count. By the way, counts of scatter ray of oblique side ($0^{\circ}-50^{\circ}$, $220^{\circ}-270^{\circ}$) are 100-600 cps, furthermore, counts at frontside are over 4 Mcps. In case of patient study, a counts of hand blood pool scan are 1510 cps. But counts of hand with lead gown on hands and on body are each 1554 cps, 1299 cps. Conclusion Therefore, even though there is a collimator in front of detector, lateral scatter rays detect on crystal and affect to images and spectrums. Especially, if there is a high activity source at outside of detector when we examine low activity organs like hands or foot, we have to shield and remove the source at outside for a good image.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.309-317
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• 1997

The measured radioactivity of gamma-emitting radionuclides in each radioactive waste drum using the non-destructive waste assay method is underestimated than real radioactivity in radioactive waste drum because the gamma-rays are attenuated within the medium. Therefore, the measured radioactivity should be corrected for the attenuation of gamma-rays. For the correction of the attenuation of gamma-rays, the attenuation correction method should be applied differently by considering the distribution and density of medium in radioactive wastes drum generated from nuclear power plants. In this study, the model drums were fabricated for simulating five types of radioactive waste drums generated from nuclear power plant and the optimum methods of the attenuation correction were experimentally determined to analyze the activity of radionuclides in the waste drum accurately using the segmented gamma scanning system. With the determination of the attenuation correction methods from the experimental results the transmission method and the average density method for the miscellaneous waste drum, the transmission method and the differential peak absorption method for the shielded miscellaneous waste drum were used to measure the density of medium in waste drums. Also, the average density method and the differential peak absorption method for the spent resin drum, the paraffin solidified drum, and the spent filter drum were used.

• Journal of radiological science and technology
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• v.38 no.1
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• pp.23-30
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• 2015

Positron emission tomography scan has been growing diagnostic equipment in the development of medical imaging system. Compare to 99mTc emitting 140 keV, Positron emission radionuclide emits 511 keV gamma rays. Because of this high energy, it needs to reduce radioactive emitting from patients for radio technologist. We searched the external dose rates by changing distance from patients and measure the external dose rates when we used shielder investigate change external dose rates. In this study, the external dose distribution were analyzed in order to help managing radiation protection of radio technologists. Ten patients were searched (mean age: $47.7{\pm}6.6$, mean height: $165.5{\pm}3.8cm$, mean weight: $65.9{\pm}1.4kg$). Radiation was measured on the location of head, chest, abdomen, knees and toes at the distance of 10, 50, 100, 150, and 200 cm, respectively. Then, all the procedure was given with a portable radiation shielding on the location of head, chest, and abdomen at the distance of 100, 150, and 200 cm and transmittance was calculated. In 10 cm, head ($105.40{\mu}Sv/h$) was the highest and foot($15.85{\mu}Sv/h$) was the lowest. In 200 cm, head, chest, and abdomen showed similar. On head, the measured dose rates were $9.56{\mu}Sv/h$, $5.23{\mu}Sv/h$, and $3.40{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.24{\mu}Sv/h$, $1.67{\mu}Sv/h$, and $1.27{\mu}Sv/h$ in 100, 150, and 200 cm on head. On chest, the measured dose rates were $8.54{\mu}Sv/h$, $4.90{\mu}Sv/h$, $3.44{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.27{\mu}Sv/h$, $1.34{\mu}Sv/h$, and $1.13{\mu}Sv/h$ in 100, 150, and 200 cm on chest. On abdomen, the measured dose rates were $9.83{\mu}Sv/h$, $5.15{\mu}Sv/h$, and $3.18{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.60{\mu}Sv/h$, $1.75{\mu}Sv/h$, and $1.23{\mu}Sv/h$ in 100, 150, and 200 cm on abdomen. Transmittance was increased as the distance was expanded. As the distance was further, the radiation dose were reduced. When using shielder, the dose were reduced as one-forth of without shielder. The Radio technologists are exposed of radioactivity and there were limitations on reducing the distance with Therefore, the proper shielding will be able to decrease radiation dose to the technologists.

• 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.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2016.05a
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• pp.778-779
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• 2016

In This paper, We designed the scintillation detector for measuring radiation signals in units of pixels for a radiation source that is distributed in the space. And we carried out a study to design a radiation imaging by the module for obtaining the detection signal. For measuring radiation distribution we configure a radiation detector combining CsI(Tl) scintillator and a photodiode. In addition, its performance was verified via gamma irradiation test.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.1
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• pp.80-85
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• 2012

Purpose: The whole body bone scan is an examination that visualizing physiological change of bones and using bone-congenial radiopharmaceutical. The patients are intravenous injected radiopharmaceutical which labeled with radioactive isotope ($^{99m}Tc$) emitting 140 keV gammarays and scanned after injection. The 3 principles of radiation protection from external exposureare time, distance and shielding. On the 3 principles of radiation protection basis, radiopharmaceutical might just as well be injected rapidly for reducing radiation because it might be the unopened radiation source. However the radiopharmaceuticals are injected into patient directly and there is a limitation of distance control. This study confirmed the change of radiation exposure as change of distance from radiopharmaceutical and observed the change of radiation exposure afte rsetting a shelter for help to control radio-technician's exposure. Materials & methods: For calculate the average of injection time, the trained injector measured the injection time for 50 times and calculated the average (2 minutes). We made a source as filled the 99mTc-HDP 925 MBq 0.2 mL in a 1 mL syringe and measured the radiation exposure from 50 cm,100 cm,150 cm and 200 cm by using Geiger-Mueller counter (FH-40, Thermo Scientific, USA). Then we settled a lead shielding (lead equivalent 6 mm) from the source 25 cm distance and measured the radiation exposure from 50 cm distance. For verify the reproducibility, the measurement was done among 20 times. The correlation between before and after shielding was verified by using SPSS (ver. 18) as paired t-test. Results: The radiation doses according to distance during 2 minutes from the source without shielding were $1.986{\pm}0.052{\mu}$ Sv in 50 cm, $0.515{\pm}0.022{\mu}$ Sv in 100 cm, $0.251{\pm}0.012{\mu}$ Sv in 150 cm, $0.148{\pm}0.006{\mu}$ Sv in 200 cm. After setting the shielding, the radiation dose was $0.035{\pm}0.003{\mu}$ Sv. Therefore, there was a statistical significant difference between the radiation doses with shielding and without shielding ($p$<0.001). Conclusion: Because the great importance of whole body bone scan in the nuclear medicine, we should make an effort to reduce radiation exposure during radiopharmaceutical injections by referring the principles of radiation protection from external exposure. However there is a limitation of distance for direct injection and time for patients having attenuated tubules. We confirmed the reduction of radiation exposure by increasing distance. In case of setting shield from source 25 cm away, we confirmed reducing of radiation exposure. Therefore it would be better for reducing of radiation exposure to using shield during radiopharmaceutical injection.

• Progress in Medical Physics
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• v.26 no.4
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• pp.250-257
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• 2015

Multi-leaf collimator (MLC) systems are frequently used to deliver photon-based radiation, and allow conformal shaping of treatment beams. Many proton beam centers currently make use of aperture and snout systems, which involve use of a snout to shape and focus the proton beam, a brass aperture to modify field shape, and an acrylic compensator to modulate depth. However, it needs a lot of time and cost of preparing treatment, therefore, we developed the manual MLC for solving this problem. This study was carried out with the intent of designing an MLC system as an alternative to an aperture block system. Radio-activation and dose due to primary proton beam leakage and the presence of secondary neutrons were taken into account during these iterations. Analytical calculations were used to study the effects of leaf material on activation. We have fabricated tray model for adoption with a wobbling snout ($30{\times}40cm^2$) system which used uniform scanning beam. We designed the manual MLC and tray and can reduce the cost and time for treatment. After leakage test of new tray, we upgrade the tray with brass and made the safety tool. First, we have tested the radio-activation with usually brass and new brass for new manual MLC. It shows similar behavior and decay trend. In addition, we have measured the leakage test of a gantry with new tray and MLC tray, while we exposed the high energy with full modulation process on film dosimetry. The radiation leakage is less than 1%. From these results, we have developed the design of the tray and upgrade for safety. Through the radio-activation behavior, we figure out the proton beam leakage level of safety, where there detects the secondary particle, including neutron. After developing new design of the tray, it will be able to reduce the time and cost of proton treatment. Finally, we have applied in clinic test with original brass aperture and manual MLC and calculated the gamma index, 99.74% between them.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.503-510
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• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.