• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.123-130
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• 2015

For the analysis of ⁹⁰Sr, which is a pure beta emitter, radiochemical separation from the main interfering elements such as Ca, Ba and Ra is required due to their similarity in chemical behavior to strontium. This study describes a radioanalytical procedure using extraction chromatography for separating Sr from interfering elements. The maximum capacity of the resin for Sr was approximately 6 mg per 1.5 mL of bed volume (BV). The recovery of Sr on the resin 1.5 mL (BV) was quantitative for the calcium level of 200 mg at the flow rate of 1 mL min^-1. However the chemical yield declined by increasing the flow rate by up to 5 mL min^-1 even at the calcium level of 200 mg. When using the same BV of Sr resin, the performance of the resin was enhanced as the cross-sectional area of the Sr resin column is small.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.421-425
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• 2003

The reductive extraction process is an important step to refine the TRU product from the electrorefining process for the preparation of transmutation reactor fuel. In this study, it was studied on the reductive extraction between the eutectic salt and Bi metal phases. The solutes were zirconium and the rare earth elements, where zirconium was used as a surrogate for the transuranic(TRU) elements. All the experiments were performed in a glove box filled with a argon gas. Li-Bi alloy was used as a reducing agent to reduce the high chemical activity of Li. The reductive extraction characteristics were examined using ICP, XRD and EPMA analysis. The reduction reaction was equilibrated within 3 hours after the Li addition. Three eutectic salt systems were compared and Zr was successfully separated from the rare earth elements in all the three salt systems.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.12
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• pp.1567-1572
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• 2013

To supply electric power in certain extreme environments such as a spacecraft or in military applications, a radioisotope thermoelectric generator has been highlighted as a useful energy source owing to its high energy density, long lifetime, and high reliability. A radioisotope thermoelectric generator generates electric power by using the heat energy converted from the radioactive energy of a radioisotope. In this study, FE analyses such as radiation shield analysis, heat transfer analysis, and power recovery rate analysis have been carried out to achieve an optimal design for a radioisotope thermoelectric generator using $SrTiO_2$.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.58-63
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• 1998

AMBIDEXTER(Advanced Molten-Salt Break even Inherently-Safe Dual-Mission EXperiment & TEst Reactor)는 토륨-우라늄 연료주기의 핵적자활성 요건을 설계하는 방법으로써 핵분열중간 생성물인 $^{233}$ Pa의 시간격리, 노내 방사성물질 농도저감, 잉여반응도 및 증식률향상을 위해 핵분열 생성물질의 온라인 정화.처리.재생 개념을 채택하고 있다. 본 연구에서는 AMBIDEXTER 로심의 핵분열성물질의 연소와 온라인 정화.처리에 따른 핵연료내 원소분포 변화를 기술하기 위해 핵분열생성물질의 평형포화농도에 대응하는 등가연소도(Equivalent Burnup)를 정의하고 이를 노심의 핵적자활성 요건에 대해 최적화하는 핵연료 정화공정의 시간상수 특성을 시뮬레이션 하였다. 핵분열생성물질농도의 동특성은 ORIGEN2 코드에 내장된 연속재처리 모델을 이용하여 해석하였으며 실용화가 입증된 후보정화공정들을 고려하여 모든 핵종을 5종의 핵종군으로 분류하여 평가하였다. 시뮬레이션 결과 유효정화주기를 0.1 (노심장전량/일)로 연속재처리 할 때 노심내 포화등 가연소도는 약 650 (MWD/TeH.E.)로 대응되며 이때 동일한 핵연료량으로부터 생성된 노내 핵분 열생성물질 평형농도는 최대연소도 33000MWD/TeU의 PWR 평형노심 BOC시의 대비해 약 1/10 에 해당하는 양이 잔유하는 것으로 나타났다.

• Journal of radiological science and technology
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• v.45 no.5
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• pp.433-438
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• 2022

Targeted radionuclides treatment (TRT) requires the establishment of treatment plans that consider various factors, such as the type of radionuclides, target organs, and administration methods. For this reason, in this study, the absorption dose of a single cell was analyzed according to the type of radioisotope used to treat target radionuclides. In this study, a simulation was performed on beta rays used in the treatment of target radionuclides at the cell level using MCNPX (ver. 2.5.0). First, according to the calculation formula, the beam path according to the type of radioisotope for treatment was calculated. Second, the amount of self-radiation by beta rays emitted from cell diameters of 5 ㎛ and 10 ㎛ cell nuclei was evaluated. As a result, it showed a high range proportional to the maximum energy of the beta-ray, and the highest self-dose distribution from 177 Lu radiation sources among therapeutic radioisotopes. This was analyzed as a result that is inversely proportional to the maximum energy of the beta-ray, and it suggests that the selection of a nuclide considering the range of the beta-ray is necessary in the treatment of target radionuclides in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.725-729
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• 1998

방사성 동위원소를 이용한 체외진단 측정법으로 RIA가 가장 널리 쓰여져 왔으나, 결합분획과 유리분획의 분리과정에서 오는 번거로운 단점을 개선하기 힘들었다. 1979년 Hart와 Greenwald에 의해 소개된 섬광근접측정법(Scintillation Proximity Assay, SPA)을 이용하여 RIA의 단점을 극복하고자 하는 연구가 시작되었다 주로 사용되는 방사성동위원소로 $^3$H, $^{125}$/I이 추천되는 핵종이며, 이 중에서 $^{125}$/I을 이용한 hCG(human chorionic gonadotropin) 체외진단법으로서 SPA의 기초연구를 수행하였다. SPA bead와 방사능 크기에 따라 측정되는 자연계수치의 증가와 hCG 표준물질을 농도별로 희석하여 유의성을 가지고 있는지 확인하였고, hCG 농도가 높은 것으로 판명된 환자의 혈청을 측정하여 임상적용의 가능성을 확인하였다.

• Analytical Science and Technology
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• v.18 no.4
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• pp.263-270
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• 2005

Nuclear track is an useful tool for elemental analysis of radionuclides, such as uranium, plutonium and thorium, etc., and for elements undergoing nuclear reactions with thermal neutrons such as lithium and boron. This method has various application fields such as detecting fissionable radionuelides, measuring the fission rate in nuclear technology, analyzing cosmic radiation from meteorite, calculating the age of minerals as well as their history, etc. Track registration method has been applied to the microscopic analysis of boron and fissionable element such as uranium in KAERI. This report reviews the theoretical background of the nuclear track formation, practical procedures to obtain etched tracks and a perspective of the future.

• Progress in Medical Physics
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• v.25 no.1
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• pp.8-14
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• 2014

The amounts of radioactive wastes to be disposed in the medical institute have been increased due to development of radiation diagnosis and therapy rapidly. They are produced mostly by the very short lived radioisotopes such as $^{18}F$ used in PET/CT, $^{99m}Tc$, $^{123}I$, $^{125}I$ and $^{201}Tl$, etc. IAEA proposed a criteria for the clearance level of waste which depends on the individual ($10{\mu}Sv/y$) and collective dose (1 man-Sv/y), and concentration of each nuclide (IAEA Safety Series No 111-P-1.1, 1992 and IAEA RS-G-1.7, 2004). Radioactive wastes of $^{18}F$, $^{99m}Tc$, $^{123}I$, $^{125}I$ and $^{201}TI$ in the several types of container like Marinelli beaker, vial and plastic, were collected to measure the concentration of the waste of each nuclide in accordance with IAEA criteria. The measurement method and procedure of determining specific activity of the wastes using gamma emitters like MCA, gamma counter and beta emitters were developed. For the efficiency calibration of the detectors, CRM (certified reference material) which has the same dimension and shape was provided by Korea Research Institute of Standards and Science (KRISS). Correction factor of the radioactivity decay was calculated based on the measurement results, and the consideration of mutual relation with theoretical equation. The result of this study will be proposed as ISO standard.

• The Korean Journal of Nuclear Medicine Technology
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• v.13 no.3
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• pp.123-131
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• 2009

Purpose: We tested a sample of nuclear medicine workers at Korean healthcare institutions for internal contamination with radioactive isotopes, measuring concentrations and evaluating doses of individual exposure. Materials and Methods: The detection and measurement was performed on urine samples collected from 25 nuclear medicine workers at three large hospitals located in Seoul. Urine samples were collected once a week, 100~200 mL samples were gathered up to 6~10 times weekly. A high-purity germanium detector was used to measure gamma radiations in urine samples for the presence of radioactive isotopes. Based on the detection results, we estimated the amounts of intake and committed effective doses using IMBA software. In cases where committed effective doses could not be adequately evaluated with IMBA software, we estimated individual committed effective doses for radionuclides with a very short half life such as $^{99m}Tc$ and $^{123}I$, using the methods recommended by International Atomic Energy Agency. Results: Radionuclides detected through the analysis of urine samples included $^{99m}Tc$, $^{123}I$, $^{131}I$ and $^{201}Tl$, as well as $^{18}F$, a nuclide used in Positron Emission Tomography examinations. The committed effective doses, calculated based on the radionuclide concentrations in urine samples, ranged from 0 to 5 mSv, but were, in the majority of cases, less than 1 mSv. The committed effective dose exceeded 1 mSv in three of the samples, and all three were workers directly handling radioactive sources. No nurses were found to have a committed effective dose in excess of 1 mSv. Conclusions: To improve the accuracy of results, it may be necessary to conduct a long-term study, performed over a time span wide enough to allow the clear determination of the influence of seasonal factors. A larger sample should also help increase the reliability of results. However, as most Korean nuclear medicine workers are currently not necessary to monitored routinely for internal contamination with radionuclides. Notwithstanding, a continuous effort is recommended to reduce any unnecessary exposure to radioactive substances, even if in inconsequential amounts, by regularly surveying workplace environments and frequently monitoring atmospheric concentrations of radionuclides.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.671-676
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• 1998

방사성 $\beta$핵종 중의 하나인 $^{14}$ C은 중수로에서 상대적으로 많이 방출된다. 본 연구에서는 국내 유일의 중수로 원전인 월성 원자력발전소에서 배기구를 통해 환경으로 방출된 $^{14}$ C을 감시하기 위한 두 가지 타입의 대기시료 포집기술을 개발하였다. 하나는 전원을 이용하는 능동시료포집법(active air sampling)이며, 다른 하나는 전원을 사용하지 않는 수동시료포집법(passive air sampling)이다. 시료분석의 재현성 측면에서는 능동시료포집법이 보다 나은 것으로 알려져 있으나 시료채취 장소의 제한성 등으로 인해 최근에 들어와 수동시료포집법의 사용을 선호하고 있다. 본 연구를 통해 두가지 시료 채취법에 대한 측정오차의 신뢰성을 검증하기 위해 동위원소 분별효과를 비교 평가해 본 결과, 측정상의 오차가 약 2% 정도로 나타나 시료채취가 간단한 수동시료포집법의 사용타당성 및 활용 가능성을 입증하였다.