• 방사선산업학회지
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• 제18권2호
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• pp.101-107
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• 2024

Rapid screening for internal contamination by alpha- and beta-emitting radionuclides is essential in situations involving radiation workers or radiation accidents. This study focused on the use of urine samples and liquid scintillation counting to quickly and accurately assess contamination. Calibration of the alpha and beta detection areas ensured precise measurement results. The major radionuclides recommended for surveillance during accidents were also considered. This study evaluated the effectiveness of the method by examining various parameters, including the limit of detection, linearity, sensitivity, selectivity, accuracy, ruggedness, and blind test sample analysis. The liquid scintillation counting method is an effective tool for screening urinary samples to detect alpha- and beta-emitting radionuclides, particularly during radiation emergencies, despite some limitations in precision.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1828-1834
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• 2019

The performance of a plastic scintillator for use in an in situ measurement system was analyzed using simulation and experimental methods. The experimental results of four major pure beta-emitting radionuclides, namely $^3H$, $^{14}C$, $^{32}P$, and $^{90}Sr/^{90}Y$, were compared with those obtained using a Monte Carlo N-particle (MCNP) code simulation. The MCNP simulation and experimental results demonstrated good agreement for $^{32}P$ and $^{90}Sr/^{90}Y$, with a relative difference of 1.95% and 0.43% between experimental and simulation efficiencies for $^{32}P$ and $^{90}Sr/^{90}Y$, respectively. However, owing to the short range of beta particles in water, the efficiency for $^{14}C$ was extremely low, and $^3H$ could not be detected. To directly measure the low-energy beta radionuclides considering their short range, a system where the source could flow directly to the scintillator was developed. The optimal thickness of the plastic scintillator was determined based on the suggested diameter. Results showed that the detection efficiency decreases with an increase in the depth of the water. The detection efficiency decreased drastically to approximately 10 cm, and the tendency was gradually constant.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1483-1488
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• 2017

In order to predict and control the environmental and health impacts of ionizing radiation in environmental sources such as groundwater, it is necessary to identify the radionuclides present. Beta-emitting radionuclides are frequently identified by measuring their characteristic energy spectra. The present work shows that self-attenuation effects from volume sources result in a geometry-dependent shift in the characteristic spectra, which needs to be taken into account in order to correctly identify the radionuclides present. These effects are shown to be compounded due to the subsequent shift in the photon spectra produced by the detector, in this case an inorganic solid scintillator ($CaF_2:Eu$) monitored using a silicon photomultiplier. Using tritiated water as an environmentally relevant, and notoriously difficult to monitor case study, analytical predictions for the shift in the energy spectra as a function of depth of source have been derived. These predictions have been validated using Geant4 simulations and experimental results measured using bespoke instrumentation.

• 한국응용약물학회:학술대회논문집
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• 한국응용약물학회 1999년도 춘계학술대회
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• pp.53-61
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• 1999

Radiation therapy has been used for the cancer treatment and radiation synovectomy$\^$1-3)/. There are two kinds of radiation therapy; the external radiation therapy and the internal radiation therapy. Hitherto, the external radiation therapy has been widely used, but for the lack of its selectivity it requires strong radiation dose and causes the irritation and damage of the normal tissue or organ. Therefore many researchers give their interests to the internal radiation therapy in which the radioactive materials are injected directly into the target organ or tissue. Many ${\beta}$-emitting radionuclides have been studied for the application of the internal radiation theraily. Among them, Holmium-166 has the many beneficial physical characteristics for the internal radiation therapy such as appropriate half life (26.8hr), high ${\beta}$ energy (max. 1.85 MeV(51%), 1.77 MeV (48%), mean 0.67MeV), and low ${\gamma}$ energy (0.081MeV) easily detected by ${\gamma}$-camera. In the internal radiation therapy, the administered radioactive materials should be retained in the target long enough to increase the therapeutic effects and avoid the damage in the normal tissue or organ. For this purpose, radionuclides are used as complex form with carriers. Carriers should have a high affinity with radionuclides in vivo and in vitro, so the complex can be evenly distributed in the lesion but can not be leaked out from the lesion.

• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.88-94
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• 2020

Background: The International Commission on Radiological Protection (ICRP) has recently published report series on the occupational intakes of radionuclides (OIR) for internal dosimetry of radiation workers. In this study, the optimized monitoring program including the monitoring interval and the minimum detectable activity (MDA) of major radionuclides was suggested to perform the routine individual monitoring of internal exposure based on the ICRP OIR. Materials and Methods: The derived recording levels and the critical monitoring quantities were reviewed from international standards or guidelines by the International Atomic Energy Agency (IAEA), the International Organization for Standardization (ISO), and the European Radiation Dosimetry Group (EURADOS). The OIR data viewer provided by ICRP was used to evaluate the monitoring intervals and the MDA, which are derived from the reference bioassay functions and the dose coefficients. Results and Discussion: The optimal monitoring intervals were determined taking account of two requirement conditions on the potential intake underestimation and the MDA values. The MDA requirement values of the selected radionuclides were calculated based on the committed effective dose from 0.1 mSv to 5 mSv. The optimized routine individual monitoring program was suggested including the optimal monitoring intervals and the MDA requirements. The optimal MDA values were evaluated based on the committed effective dose of 0.1 mSv. However, the MDA can be adjusted considering the practical operation of the routine individual monitoring program in the nuclear facilities. Conclusion: The monitoring intervals and the MDA as crucial factors for the routine monitoring were described to suggest the optimized routine individual monitoring program of the occupational intakes. Further study on the alpha/beta-emitting radionuclides as well as short lived gamma-emitting nuclides will be necessary in the future.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1169-1175
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• 2019

The minimum detectable activity (MDA) value was derived according to the flow rate of the sample and degree of amplification of the device by sending the sample directly from the collection site to the detection part through a pump. This method can lead to reduction in time and cost compared to the existing measurement method that uses a pre-treatment process. In this study, experiments were conducted on $^3H$ and $^{90}Sr$, which are the major pure beta-emitting radionuclides, by setting the sample flow rate and the amplification gain as factors. The MDA values were derived according to the flow rates, considering that the flow rate can affect the MDA values. There were no change in the MDA under different flow rates of 0, 600, 800, and 1000 mL/min. Therefore, it was confirmed that the flow rate may not be considered when collecting samples for monitoring in actual field. As the degree of amplification of the amplifier increased, the time required to reach the target MDA decreased. When the amplification was quadrupled, the detection efficiency increased by approximately 23.4 times, and the time to reach the MDA decreased to approximately 1/550 times. This method offers the advantage of real-time on-site monitoring.

• 센서학회지
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• 제21권5호
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• pp.367-373
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• 2012

We fabricated a fiber-optic alpha/beta detector, which is composed of a sensing probe, a plastic optical fiber, a photomultiplier tube, and a multichannel analyzer, to obtain the energy spectra of radioactive isotopes. As inorganic scintillators of a sensing probe, a ZnS(Ag) film was coupled with a $CaF_2$(Eu) crystal for alpha and beta spectroscopy. In this study, $^{210}Po$ and $^{90}Sr$ were used as alpha and beta sources, respectively, and we measured the radiation energy spectra using a fiber-optic alpha/beta detector to identify alpha and beta emitting radionuclides for nuclear medicine application. Also, the variations of energy spectrum were obtained according to the length of plastic optical fiber.

• Journal of Radiation Protection and Research
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• 제18권2호
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• pp.17-25
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• 1993

액체섬광 계수기에 대한 $^3H,\;^{14}C,\;^{36}Cl$의 베타선 스펙트럼의 시뮬레이션을 Monte Carlo 법으로 수행하였다. 베타선의 손실에너지는 'continuous slowing down approximation(CSDA)' 모형으로부터 구하였다. 순 베타방출 핵종이 균일하게 용해되어 있는 섬광용액은 $1.25cm{\times}3.5cm$의 원통형 geome try로 가정하였다. $^{14}C,\;^{36}Cl$에 대한 계산결과는 계수효율의 경우 실험 값과 2%의 범위 내에서 잘 일치하였으나 $^3H$에 대해서는 상당한 편차를 나타내었다. 이는 저 에너지 베타선에 대한 섬광용액 자체의 quenching 효과에 기인하는 것으로 판단된다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 춘계 학술대회
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• pp.284-289
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• 2005

중저준위 방사성폐기물의 핵종 관리를 포함한 안전 규제 정책에 따라 핵종 유출을 방지하고 제조된 고화체로부터 C-14의 회수율을 알고자 국내 원자력 발전소에서 발생되는 중 저준위 방사성 폐기물 중 농축폐액을 시멘트로 일정시간 고화시켜 시료를 준비하였다. 이들 시멘트 고화체는 화학적 산화법을 적용하여 C-14를 분리하였으며, 화학적 분리된 C-14는 액체섬광계수기를 이용하여 방출되는 베타선을 계측하였고, 소광보정 곡선을 이용하여 방사능을 측정하였다. 또한 원전에서 제조된 시멘트 고화체를 동일한 방법으로 C-14를 정량하였으며, 그 방사능 결과는 $2.7E+00\;{\sim}\;3.1E+02$ Bq/g의 범위로 분리 검출되었다.

• 대한방사성의약품학회지
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• 제5권1호
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• pp.26-35
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• 2019

$^{188}Re$ is one of the most readily available generator derived and useful radionuclides for therapy emitting ${\beta}^-$ particles (2.12 MeV, 71.1% and 1.965 MeV, 25.6%) and imageable gammas (155 keV, 15.1%). The $^{188}W/^{188}Re$ generator is an ideal source for the long term (4-6 months) continuous availability of no carrier added (NCA) $^{188}Re$ suitable for the preparation of radiopharmaceuticals for radionuclide therapy. Rhenium-188 has been used for the preparation of therapeutic radiopharmaceuticals for the management of diseases such as bone metastasis, rheumatoid arthritis and primary cancers. Several early phase clinical studies using radiopharmaceuticals based on $^{188}Re$ -labeled phosphonates, antibodies, peptides, lipiodol and particulates have been reported. In this review, we addressed the current development status of $^{188}Re$ radiopharmaceuticals for liver cancer therapy and their applications.