• 대한방사선기술학회지:방사선기술과학
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• 제44권5호
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• pp.481-487
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• 2021

This study was conducted for the purpose of efficient radioactive waste disposal and management. Experiment was evaluated the decontamination efficiencies of the four types decontamination materials(Water, Alcohol, Decontamination Water, Decontamination Gel) with radioactive wastes generated during radio-pharmaceutical production process at Korea Institute Radiological and Medical Sciences(KIRAMS). The radioactive waste sample used in experiment is a lead plate of the fume hood that was disposed in April, 2019. In the experimental method, radioactive waste was measured before and after decontamination using a HPGe semiconductor detector and Gamma survey meter. The measured values before and after decontamination were evaluated for decontamination efficiency as a percentage. As a result, it was confirmed that a lot of specific activity and surface dose rate was removed from the radioactive wastes. In particular, when decontamination water was used, most of the radioactivity of radioactive wastes was removed. Considering these results, if decontamination water is used in decontamination of radioactive waste, decontamination efficiency equivalent to the disposition criteria can be expected with just one decontamination treatment. In addition, in the case of water and alcohol, only on decontamination was effective in approximately 75% and 95%. Otherwise, when decontamination gel was used, it was confirmed that the largest deviation occurred among all experimental results.

• 한국방사선학회논문지
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• 제16권5호
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• pp.519-526
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• 2022

능동형 전자식 개인피폭선량는 개인의 피폭 선량을 실시간으로 확인할 수 있는 장점을 가진 보조선량계이다. 하지만 국내에 사용되고 있는 다수의 능동형 개인피폭 선량계는 의료기관에서 사용하는 진단방사선 영역에서 큰 오차와 낮은 응답성을 가진다. 이에 본 연구에서는 Si 포토다이오드 검출기를 사용하는 능동형 전자식 개인선량계에서 저에너지 영역의 응답특성을 향상시키기 위한 에너지 보상 두께를 평가하였다. 40 kVp에서 80 kVp 영역에서는 Al 0.2 mm + Sn 1.0 mm 필터에서 우수한 응답특성을 보였고 80 kVp에서 120 kVp 영역에서는 Al 0.2 mm + Sn 1.6 mm 필터에서 우수한 응답특성을 보였다.

• 한국의학물리학회지:의학물리
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• 제32권4호
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• pp.145-152
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• 2021

Purpose: During the treatments of cancer patients with a linear accelerator (LINAC) using photon beams with energies ≥8 MV, the components inside the LINAC head get activated through the interaction of photonuclear reaction (γ, n) and neutron capture (n, γ). We used spectroscopy and measured the dose rate for the LINAC in operation after the treatment ended. Methods: We performed spectroscopy and dose rate measurements for three units of LINACs with a portable high-purity Germanium (HPGe) detector and a survey meter. The spectra were obtained after the beams were turned off. Spectroscopy was conducted for 3,600 seconds, and the dose rate was measured three times. We identified the radionuclides for each LINAC. Results: According to gamma spectroscopy results, most of the nuclides were short-lived radionuclides with half-lives of 100 days, except for ⁶⁰Co, ⁶⁵Zn, and ¹⁸¹W nuclides. The dose rate for three LINACs obtained immediately in front of the crosshair was in the range of 0.113 to 0.129 µSv/h. The maximum and minimum dose rates measured on weekends were 0.097 µSv/h and 0.092 µSv/h, respectively. Compared with the differences in weekday data, there was no significant difference between the data measured on Saturday and Sunday. Conclusions: Most of the detected radionuclides had half-lives <100 days, and the dose rate decreased rapidly. For equipment that primarily used energies ≤10 MV, when the equipment was transferred after at least 10 minutes after shutting it down, it is expected that there will be little effect on the workers' exposure.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4253-4264
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• 2022

Since the implementation of the phosphate project in Bled El-Hadba (BEH) deposit, western region of Tébessa, no detailed study has been conducted to assess the natural radioactivity distribution and the associated radiological risk parameter for this open-pit mine. For the sake of determining a credible premining reference database for the region of interest, 21 samples were collected from different geological layers of the above-mentioned deposit. Gamma Spectrometry was applied for measuring radioactivity using a high resolution HPGe semiconductor detector. The obtained activity results have shown a significant broad variation in the radioactive contents for the different phosphate samples. The total average concentrations (in Bq·kg^-1) for ²²⁶Ra, ²³⁸U, ²³⁵U, ²³²Th and ⁴⁰K computed for the different type of phosphate layers were found to be 570 ± 169, 788 ± 280, 52 ± 18, 66 ± 6 and 81 ± 18 respectively. The mean activity concentrations of the measured radionuclides were compared to other regional and worldwide deposits. The ratios between the detected radioisotopes have been calculated for spatial distribution of natural radionuclides in the study area. Based on the aforementioned activity concentrations, the corresponding radiation hazard parameters were assessed. Correlations between the obtained parameters were drawn and a multivariate statistical analysis (Pearson Correlation, Cluster and Factor analysis) was carried out in order to identify the existing relationships.

• Journal of Radiation Protection and Research
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• 제47권4호
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• pp.214-225
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• 2022

Background: The conventional cerium-doped Gd₂Al₂Ga₃O₁₂ (GAGG(Ce)) scintillator-based gamma-ray imager has a bulky detector, which can lead to incorrect positioning of the gammaray source if the shielding against background radiation is not appropriately designed. In addition, portability is important in complex environments such as inside nuclear power plants, yet existing gamma-ray imager based on a tungsten mask tends to be weighty and therefore difficult to handle. Motivated by the need to develop a system that is not sensitive to background radiation and is portable, we changed the material of the scintillator and the coded aperture. Materials and Methods: The existing GAGG(Ce) was replaced with Bi₄Ge₃O₁₂ (BGO), a scintillator with high gamma-ray detection efficiency but low energy resolution, and replaced the tungsten (W) used in the existing coded aperture with lead (Pb). Each BGO scintillator is pixelated with 144 elements (12 × 12), and each pixel has an area of 4 mm × 4 mm and the scintillator thickness ranges from 5 to 20 mm (5, 10, and 20 mm). A coded aperture consisting of Pb with a thickness of 20 mm was applied to the BGO scintillators of all thicknesses. Results and Discussion: Spectroscopic characterization, imaging performance, and image quality evaluation revealed the 10 mm-thick BGO scintillators enabled the portable gamma-ray imager to deliver optimal performance. Although its performance is slightly inferior to that of existing GAGG(Ce)-based gamma-ray imager, the results confirmed that the manufacturing cost and the system's overall weight can be reduced. Conclusion: Despite the spectral characteristics, imaging system performance, and image quality is slightly lower than that of GAGG(Ce), the results show that BGO scintillators are preferable for gamma-ray imaging systems in terms of cost and ease of deployment, and the proposed design is well worth applying to systems intended for use in areas that do not require high precision.

• 한국방사선학회논문지
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• 제3권1호
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• pp.17-21
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• 2009

최근 의료진단 분야와 다른 적용분야를 위해 대면적 매트릭스 구조의 엑스선 영상이 활발하게 연구되어 오고 있다. 본 연구에서는, 의료진단을 위한 새로운 평판형 디지털 엑스선 가스 검출기를 제안하고 그에 따른 특성을 검증하고자 한다. 대기압에 반해 가스를 주입하는 어려움 때문에 챔버 형태의 구조로 만들어 질 뿐, 평판형 디지털 엑스선 가스 검출기는 아직 어디에서도 연구된 바 없다. 이에 본 연구에서는 디스플레이 패널 제작 기술을 이용하여 샘플제작을 성공하였다. 실험적인 측정을 위해 만들어진 샘플은 상판에는 유리기판위에 전극, 절연층, 산화마그네슘 보호막을 형성하였으며, 하판에는 엑스선 형광층과 전극을 형성하였다. 누설전류와 엑스선 민감도를 측정하였으며, 전기장에 대한 민감도의 선형성 측정 등의 전기적 특성평가를 실시하였다. 이에 대한 결과로 안정된 누설전류와 엑스선 민감도를 얻었다. 그리고 조사 선량에 따라 좋은 선형성을 보이는 등 넓은 진단 동적영역을 확인할 수 있었다. 이러한 결과로 평판형 엑스선 가스 검출기의 디지털 엑스선 영상 검출기로의 적용 가능성을 확인 할 수 있었다.

• 한국방사선학회논문지
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• 제8권1호
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• pp.1-10
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• 2014

양방향 뇌혈관촬영기는 하나의 선원에서 방사선이 나오는 것이 아니라 정방향과 측방향에서 방사선 피폭이 이루어지기 때문에 방사선 관계종사자의 피폭이 더욱 많아질 수밖에 없다. 따라서 환자가 받는 선량도 중요하겠지만 시술을 시행하고 있는 방사선 관계종사자 역시 피폭선량을 줄일 수 있는 방법에 대해 많은 관심을 보이고 있다. 본 연구의 목적은 양방향 뇌혈관촬영기의 혈관검사 및 중재적방사선시술에 있어 X-선 관구에서 직접 조사되는 1차 방사선, 정방향관구와 측방향 검출기사이에서 발생하는 1차 산란방사선, 상대적으로 적지만 촬영실 벽이나 바닥에서 반사되는 2차 산란방사선 발생을 간과하지 않을 수 없기에 기존의 일반적인 차폐방법인 천정형 차폐, 테이블형 차폐방법에서 보다 더 적극적인 차폐방법인 측방향 차폐체의 방어용구를 설치하여 시술자가 받는 직접방사선 및 산란방사선에 의한 피폭선량을 최대한 줄이고자 노력하였다. 그 결과 투시측정에서는 생식샘 약 3.64배, 갑상샘 약 3.13배, 눈 약 4.42배 정도 더 감소하였고, 디지털 감산 혈관조영측정에서는 생식샘 약 4.98배, 갑상샘 약 3.00배, 눈 약1.67배의 피폭선량 감소효과를 얻어내었다. 결론적으로 양방향 뇌혈관촬영기의 혈관검사 및 중재적방사선시술시 설치하였던 측방향 차폐체의 방어용구는 일반적인 차폐방법보다 시술자의 피폭선량을 감소시키는데 많은 효과를 주었다고 사료된다.

• 대한방사선기술학회지:방사선기술과학
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• 제30권2호
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• pp.153-159
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• 2007

PACS의 도입에 발맞추어 시작된 digital 의료영상은 현재 방사선 진단 및 치료 영역에서 일반화되었으며 특히 진단영역에서는 눈부신 발전을 거듭하고 있다. 치료영역에서의 digital image 구현은 상대적으로 느린 성장과정을 거쳐 오늘에 이르고 있다. 본 논문은 이러한 analog에서 digital로 변화하는 흐름을 인식하여 정도관리(Quality Assurance) 업무 중 mechanical check 부분을 digital image base 업무로 대체하여 수행하고, 기존의 육면체 또는 사각형틀의 개념을 탈피한 구형(球形)의 Spherical mechanical check device(SMCD) 고안하여 그 실용성을 실험하였다. Source(target)에서 image detecter 간의 거리가 항상 일정하고, Spherical mechanical check device(SMCD)의 중심까지 거리가 일정하다면 어느 방향에서 SMCD를 exposure하더라도 그 크기는 항상 일정하게 영상으로 표현 될 것이다. 이를 위해 정확한 반구(半球)를 2개 정밀 제작하여 그것을 합쳤을 때 정원(正圓)의 구(球)가 되도록 하였다. 이를 이용하여 radiation field와 light field의 일치도, radiation field 크기의 정확도($(5{\times}5,\;10{\times}10,\;15{\times}15\;cm)$), Collimator field 크기의 정확도($5{\times}5,\;10{\times}10,\;15{\times}15\;cm$), Gantry rotation isocenter check, Collimator rotation isocenter check, Room laser accuracy check, Collimator rotation angle check와 Couch rotation angle check 등 기존의 mechanical check를 digital image를 이용하여 실행할 수 있었으며, 기존의 Flat 또는 정육면체 형태의 mechanical check device로는 쉽게 하기 힘든 non-coplanar field에 적용되는 Gantry와 Couch가 동시에 rotation되었을 때 그 isocenter의 일치도를 real time으로 확인할 수 있었다.

• 한국의학물리학회지:의학물리
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• 제29권3호
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• pp.92-100
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• 2018

The manufacturer of a linear accelerator (LINAC) has reported that the target melting phenomenon could be caused by a non-recommended output setting and the excessive use of monitor unit (MU) with intensity-modulated radiation therapy (IMRT). Due to these reasons, we observed an unexpected beam interruption during the treatment of a patient in our institution. The target status was inspected and a replacement of the target was determined. After the target replacement, the beam profile was adjusted to the machine commissioning beam data, and the absolute doses-to-water for 6 MV and 10 MV photon beams were calibrated according to American Association of Physicists in Medicine (AAPM) Task Group (TG)-51 protocol. To verify the beam data after target replacement, the beam flatness, symmetry, output factor, and percent depth dose (PDD) were measured and compared with the commissioning data. The difference between the referenced and measured data for flatness and symmetry exhibited a coincidence within 0.3% for both 6 MV and 10 MV, and the difference of the PDD at 10 cm depth ($PDD_{10}$) was also within 0.3% for both photon energies. Also, patient-specific quality assurances (QAs) were performed with gamma analysis using a 2-D diode and ion chamber array detector for eight patients. The average gamma passing rates for all patients for the relative dose distribution was $99.1%{\pm}1.0%$, and those for absolute dose distribution was $97.2%{\pm}2.7%$, which means the gamma analysis results were all clinically acceptable. In this study, we recommend that the beam characteristics, such as beam profile, depth dose, and output factors, should be examined. Further, patient-specific QAs should be performed to verify the changes in the overall beam delivery system when a target replacement is inevitable; although it is more important to check the beam output in a daily routine.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2002년도 제3회 최신 분말제품 응용기술 Workshop
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• pp.25-37
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• 2002

The most important industrial application of gamma radiation in characterizing green compacts is the determination of the density. Examples are given where this method is applied in manufacturing technical components in powder metallurgy. The requirements imposed by modern quality management systems and operation by the workforce in industrial production are described. The accuracy of measurement achieved with this method is demonstrated and a comparison is given with other test methods to measure the density. The advantages and limitations of gamma ray densitometry are outlined. The gamma ray densitometer measures the attenuation of gamma radiation penetrating the test parts (Fig. 1). As the capability of compacts to absorb this type of radiation depends on their density, the attenuation of gamma radiation can serve as a measure of the density. The volume of the part being tested is defined by the size of the aperture screeniing out the radiation. It is a channel with the cross section of the aperture whose length is the height of the test part. The intensity of the radiation identified by the detector is the quantity used to determine the material density. Gamma ray densitometry can equally be performed on green compacts as well as on sintered components. Neither special preparation of test parts nor skilled personnel is required to perform the measurement; neither liquids nor other harmful substances are involved. When parts are exhibiting local density variations, which is normally the case in powder compaction, sectional densities can be determined in different parts of the sample without cutting it into pieces. The test is non-destructive, i.e. the parts can still be used after the measurement and do not have to be scrapped. The measurement is controlled by a special PC based software. All results are available for further processing by in-house quality documentation and supervision of measurements. Tool setting for multi-level components can be much improved by using this test method. When a densitometer is installed on the press shop floor, it can be operated by the tool setter himself. Then he can return to the press and immediately implement the corrections. Transfer of sample parts to the lab for density testing can be eliminated and results for the correction of tool settings are more readily available. This helps to reduce the time required for tool setting and clearly improves the productivity of powder presses. The range of materials where this method can be successfully applied covers almost the entire periodic system of the elements. It reaches from the light elements such as graphite via light metals (AI, Mg, Li, Ti) and their alloys, ceramics ($AI_20_3$, SiC, Si_3N_4, $Zr0_2$, ...), magnetic materials (hard and soft ferrites, AlNiCo, Nd-Fe-B, ...), metals including iron and alloy steels, Cu, Ni and Co based alloys to refractory and heavy metals (W, Mo, ...) as well as hardmetals. The gamma radiation required for the measurement is generated by radioactive sources which are produced by nuclear technology. These nuclear materials are safely encapsulated in stainless steel capsules so that no radioactive material can escape from the protective shielding container. The gamma ray densitometer is subject to the strict regulations for the use of radioactive materials. The radiation shield is so effective that there is no elevation of the natural radiation level outside the instrument. Personal dosimetry by the operating personnel is not required. Even in case of malfunction, loss of power and incorrect operation, the escape of gamma radiation from the instrument is positively prevented.