• Journal of radiological science and technology
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• v.39 no.4
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• pp.643-649
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• 2016

When a cyclotron produces $^{18}F^-$, accelerated protons interact with metal parts of the cyclotron machine and induces radioactivity. Especially, the target window and chamber of the target assembly are the main parts where long-lived radionuclides are generated as they are incident by direct beams. It is of great importance to identify radionuclides induced in the target assembly for the safe operation and maintenance of a cyclotron facility. In this study, we analyzed major radionuclides generated in the target assembly by an operation of the Cyclotron 18/9 machine and measured dose rates after the operation to establish the radiation safety guideline for operators and maintenance personnel of the machine. Gamma spectroscopy with HPGe was performed on samples from the target chamber and Havar foil target window to identify the radionuclides generated during the operation for production of $^{18}F^-$- isotope and their specific activity. Also, the dose rates from the target were measured as a function of time after an operation. These data will help improve radiological safety of operating the cyclotron facilities.

• Journal of the Korean Recycled Construction Resources Institute
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• v.7 no.4
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• pp.316-322
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• 2019

A probabilistic approach considering uncertainties was employed to investigate the effects on the material characteristics and strength of nuclear bio-logical shield concrete walls, when exposed to long-term radiation during the entire service life. Time-dependent compressive and tensile strengths were estimated by conducting the neutron fluence analysis. For the contaminated concrete, individual compressive and tensile failure probabilities can be possibly evaluated by not only establishing limit-state function withthe predefined critical values but also performing Monte Carlo Simulation. Nuclear power plant types similar to the Kori Unit 1, which was shut off permanently in 2017 after the 40-year operation, were herein selected for an illustrative purpose. Consequently, the probability-based performance assessment and prediction of contaminated concrete walls were well demonstrated.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.77-78
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• 2016

사용후핵연료 및 조사핵연료의 안전성평가를 위한 산화시험과 소둔시험을 다양한 핵연료 개발과제에서 준비하고 있다. 한국원자력연구원의 PIEF에서 수행될 본 시험장치는 핫셀의 제한된 특성을 반영하여 효과적인 핵분열생성물의 핵종분석시험이 가능하도록 설계.제작되었다. 가열로의 소형화와 조작의 편이성을 반영하였으며, 핫셀 내 가열로 상부캡위에 Gamma Counter를 설치하여 핵분열생성물의 방출량 측정이 가능하도록 하였다. 시험운용에 있어 일정한 온도범위에서의 가열조건과 급격한 온도변화를 모사할 수 있는 기능이 포함되었다. 아울러, 핵분열생성물의 핵종분석은 계측 외에 세라믹 흡착제에 고착된 방사선량의 측정을 통해 교차평가가 가능하다.

• Fire Science and Engineering
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• v.30 no.6
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• pp.9-13
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• 2016

Anti-oxidation and flame resistant polyurethane nanofibers were prepared by electrospinning and aluminum hydroxide addition. Electrospinning was carried out under the following procedure conditions; applied voltage, 20 kV; polymer solution feeding rate, 1.2 ml/h; collector rolling speed, 120 rpm; and tip to collector distance, 15 cm. Aluminum hydroxide was added to the prepared polymer solution for electrospinning to enhance the oxidation and flame resistant properties. The thermal properties were investigated by thermogravimetric analysis to determine the polymer decomposition temperature, integral procedure decomposition temperature, final decomposition temperature, and remaining amount after thermal decomposition. The activated energy for polymer degradation was also investigated using the Horowitz-Metzger equation. The activation energy increased to more than 50%. The thermal properties of the polyurethane nanofibers were improved by a hydration reaction during the thermal decomposition of aluminum hydroxide around $300{\sim}500^{\circ}C$.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.596-600
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• 2003

The Advanced spent fuel Conditioning Process(ACP) is under development for the effective management of spent fuel which had been generated in nuclear plants. The ACP needs a hot cell where most operations will be peformed. To give priority to the environment safety, radiation doses evaluation for the radioactive nuclides were preliminarily peformed in both normal operation and accident case. The evaluation result shows a safe margin for regulation limits and SAR limit of IMEF where this facility will be constructed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.217-233
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• 2014

Securing the radiological safety is a prerequisite for the safe management of the naturally occurring radioactive materials (NORM) which cannot be reused. This becomes a crucial focus of our R&D efforts upon the implementation of the Act on Protective Action Guidelines against Radiation in the Natural Environment. To secure the safety, the establishment of technical bases and procedures for securing radiological safety related to the disposal of NORM is required. Thus, it is necessary to analyze the characteristics, to collect the data, to have the radiological safety assessment methodologies and tools, to investigate disposal methods and facilities, and to study the effects of the input data on the safety for the NORM wastes. Here, we assess the environmental impact of the NORM waste disposal with respect to the major domestic and foreign NORM characteristics. The data associated with major industries are collected/analyzed and the status of disposal facilities and methodologies relevant to the NORM wastes is investigated. We also suggest the conceptual design concept of a landfill disposal facility and the management plan with respect to the major NORM wastes characteristics. The radionuclide pathways are identified for the atmospheric transport and leachate release and the environmental impact assessment methodology for the NORM waste disposal is established using a relevant code. The assessment and analysis on the exposure doses and excessive cancer risks for the NORM waste disposal are performed using the characteristics of the representative domestic NORM wastes including flying ash, phosphor gypsum, and redmud. The results show that the exposure dose and the excessive cancer risks are very low to consider any radiation effects. This study will contribute to development in the areas of the regulatory technology for securing radiological safety relevant to NORM waste disposal and to the implementation technology for the Act.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.73-80
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• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.316-317
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• 2004

사용후핵연료를 이용한 건식 핵연료 원격 제조 공정중 OREOX공정으로부터 핵분열기체 방출특성 평가를 위한 실험을 수행하였다. 사용후핵연료 분말화 공정인 1차 산화 및 OREOX공정에서 방출되는 핵분열기체를 실시간으로 측정할 수 있는 장치를 제작$\cdot$설치하였으며, 측정 대상 핵분열기체는 Kr-85를 포함하여 C-14$^{14}CO_2$ 형태), I-129, 기체상 트리튬 등이다. 그림 1은 방출되는 핵분열기체를 포집 또는 연속측정하기 위한 개념도이며, 그림 2는 핫셀구역에 설치된 장치 사진이다.(중략)

• The Korean Journal of Rheology
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• v.5 no.1
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• pp.23-33
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• 1993

저밀도폴리에틸렌과 나일론6의 블렌드가 용융압출기의 사용에 의해 제조되었다. 제 조된 비상용 블랜드의 유변학적, 형태학적 분석을 통하여 이들사이의 상관관계를 조사하였 다. 전단속도, 계면장력, 용융점도, 분산상의 농도들에 대한 분산상의 크기의 의존성을 평가 하였다. 한편 신장유동에 의한 모폴로지 변화를 조사하기 위하여 기제조된 블렌드를 사용하 여 용융방사를 하였다. 이러한 방식실험으로부터 본연구에 사용된 저밀도 폴리에틸렌과 나 일론6는 분산상과 연속상의 신장점도의 상대적 크기에 무관하게 피브릴화가 되어 방사 연신 비의 증가에 의해 매우 가는 분산된 피브릴을 얻을수 있었다.

• Proceedings of the Korean Information Science Society Conference
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• 2002.10d
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• pp.556-558
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• 2002

항공사진에 포함된 기점 마크의 방사 및 기하 특성을 이용하여 마크의 중심 위치를 자동으로 인식하기 위한 방안을 제안한다. 마크를 포함하는 배경 영역의 방사 특성에 기반을 푼 전략에 근거하여 입력된 영상을 이치화한 다음 형태 연산자를 적용시켜 기전 마크가 있는 후보 영역을 추출한다. 기하 특성에 기반을 둔 전략에 근거하여 ▽$^2$G 필터링과 대칭성 강조 필터링을 적용시킨 후, 대칭이 가장 강하게 나타나는 위치인 마크의 중심 위치를 구한다. 66매의 기점 마크 영상에 대한 평가 결과 중심 위치가 1 화소의 정확도까지 얻어질 수 있다는 것을 확인할 수 있었다.