• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.302-307
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• 1989

Nuclide release through penetrations in radioactive waste container is analyzed. Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant nuclide release, but not with the way these penetrations occur. Numerical illustrations show that mass transport from multiple holes can be significant and may approach the mass transfer rate calculated from bare waste forms. Although partially-failed containers may present an important long-term barrier to release of radionuclides, numerous small holes on a container surface have the potential of bypassing the effectiveness of these barriers.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.485-495
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• 2005

This study was conducted to investigate sorption and desorption behaviors of radionuclides (Cobalt and Strontium) in natural soil. Sorption kinetics and isotherms were analyzed to predict sorption behaviors of radionuclides in natural soil and the experimental data were fitted to several sorption models. Desorption experiments were also performed with or without CMCD at constant pH and ion strength conditions. The results showed that $Sr^{2+}$ was more strongly sorbed than $Co^{2+}$ in natural soil. Both $Co^{2+}$ and $Sr^{2+}$ followed a pseudo-second order kinetics and Sips model. The desorption-resistance of $Co^{2+}$ and $Sr^{2+}$ was estimated using a natural surfactant Carboxymethyl-${\beta}$-cyclodextrin(CMCD) or non-desorbing fraction. Desorption of radionuclides was partially irreversible and $Sr^{2+}$ was more resistant than $Co^{2+}$ Addition of CMCD facilitated desorption of $Co^{2+}$ and $Sr^{2+}$ from soil.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.54-59
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• 2003

Several radionuclides are considered as an object of the assesment to develop a scaling factor and a periodical verification method which are needed for the evaluation of radionuclide inventory of various radioactive wastes from nuclear power plants in Korea. A selective separation of $^{55}Fe$, $^{90}FSr$ 및 $^{94}Nb$ which should be recovered individually for the radiochemical analysis was described in detail. Sorption and desorption behaviours of ion exchange and extraction chromatographic resins for Fe, Sr, Nb and co-existing metal ions were Investigated using an artificial waste solution simulated of chemical composition of real radioactive wastes. Separation conditions available for the sequential recovery of these metal ions from a single sample were optimized to minimize a discharge of radioactive wastes produced through the analytical process and a radiation exposure to analysts. Their recovery yields were measured with reliability.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.79-86
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• 2009

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.582-590
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• 1995

우리 나라에서 건설될 중저준위폐기물 처분장의 기본설계개념에 따르면, 중준위폐기물 동굴에는 점토질 뒷채움재가 설치될 예정이다. 이 뒷채움재의 후보물질로는 원료 입수의 용이성 및 경제성 측면에서 천연점토와 분쇄암석의 혼합물이 고려되고 있다. 그러나 이 혼합물이 폐기물처분장의 뒷채움재로 사용되기 위해서는 주요특성들이 뒷채움재의 요구사항을 만족시켜야한다. 이 보고서에서는 경상북도 연일지역에서 산출되는 천연점토와 분쇄암석 혼합물에 대해 핵종유출저지 관점에서 가장 중요한 특성인 수리특성과 방사성핵종 확산특성 및 역학적 특성을 측정하여 뒷채움재료로서의 사용 가능성을 평가하였다. 혼합물의 수리전도도는 동일 밀도에서 점토함량이 감소함에 따라 증가되었으나, 점토함량이 적은 경우에도 상당히 낮은 값을 유지하여 뒷채움재를 통한 핵종이동의 주요 메커니즘은 확산이 될 것이라는 것을 보여주었다. 압축점토에 대한 방사성핵종 복산계수 측정결과, 동일밀도의 나트륨벤토나이트에서 측정된 값의 범위 내에 있었으며, 점토-분쇄암석 혼합물의 액소성한계, 다짐특성, 압축강도및 압밀계수등의 역학적 특성도 나트륨벤토나이트-모래 혼합물에서 얻어진 값과 유사하였다. 이러한 결과들로부터 연일산 천연점토-분채암석 혼합물을 고가의 나트륨벤토나이트 대용으로 처분장 뒷채움재로 사용할 수 있다는 것을 알 수 있었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.281-291
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• 2013

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

• Journal of the Korean Society of Radiology
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• v.16 no.7
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• pp.1007-1014
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• 2022

Targeted radionuclide therapy (TRT) is a method of treating tumor cells using radiopharmaceuticals. Cells and nuclei constituting tissues of the human body are composed of spherical and oval shapes, but cancer cells are composed of various cell types. Therefore, this study analyzed the absorbed dose for each organelle according to the change in the size of the cell nucleus for beta-emitting nuclides during targeted radionuclide therapy through the Monte Carlo method. Cells were set in two sphere shapes, 5 ㎛ and 10 ㎛, and the internal structure was divided into cell nucleus, cytoplasm, and cell surface. Next, the absorbed dose according to the increase in the size of the cell nucleus was evaluated. As a result, ¹⁷⁷Lu among the target radionuclides showed the highest dose in all cell compartments. As the ratio of the nucleus in the cell increased, the absorbed dose on the cell surface increased, but the absorbed dose in the cytoplasm and nucleus tended to decrease. Accordingly, it is judged that it is important to select a radionuclide considering the size of cancer cells and determine an appropriate amount of radioactivity during targeted radionuclide treatment.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.571-576
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• 1996

지하암반층에 처분한 방사성핵종들의 이동방식을 이해하기 위한 방법의 일환으로 다차원 불균일장에대한 모사기법에 대해 다루었다. 이차원 암반균열평면을 설정하고 균열틈 값이 공간적으로 다양한 형태를 가질 때 지하수가 흐르는 유형과 핵종들이 이동하는 특성을 이해하기 위해, 공간적으로 다양하게 변해 가는 균열틈을 가진 암반층을 모사한 다음, 여기서 지하수와 방사성핵종들의 흐름특성을 파악하고 물질이동에 영향을 미치는 중요한 수리학적 매개변수들에 대해 논의하였다. 암반 균열폭 값은 추계적 방식으로 설정한 확률밀도분포와 공간상호간섭길이를 함수로하여 통계적방식으로 계산하였다. 각 매개변수들의 값의 차이에 따라 달라지는 분포상태변화를 그림으로 표현하여 설명하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.821-826
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• 1995

방사성 폐기물 드럼 내에 함유된 핵종의 종류 및 농도를 비파괴적으로 평가하기 위하여 드럼 이 상하 또는 회전 운동을 하게 한 후 감마선을 검출하는 감마 스케닝 자동화 장치 및 관련평가 기술을 개발하였다. 자동화 장치를 컴퓨터로 제어하기 위한 컴퓨터 프로그램은 사용자가 사용하기에 편리하도록 컴퓨터 화면에 나타난 여러가지 기능 키들을 마우스를 사용하여 간단하게 조작할 수 있게 하였다. 또한 본 연구에서 개발한 몬테칼로 전산 코드의 검증 작업도 수행하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.123-133
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• 2010

A statistical evaluation methodology was developed to determine the compliance of candidate waste stream with clearance criteria based upon distribution of radionuclide in a waste stream at a certain confidence level. For the cases where any information on the radionuclide distribution is not available, the relation between arithmetic mean of radioactivity concentration and its acceptable maximum standard deviation was demonstrated by applying widely-known Markov Inequality and One-side Chebyshev Inequality. The relations between arithmetic mean and its acceptable maximum standard deviation were newly derived for normally or lognormally distributed radionuclide in a waste stream, using probability density function, cumulative density function, and other statistical relations. The evaluation methodology was tested for a representative case at 95% of confidence level and 100 Bq/g of clearance level of radioactivity concentration, and then the acceptable range of standard deviation at a given arithmetic mean was quantitatively shown and compared, by varying the type of radionuclide distribution. Furthermore, it was statistically demonstrated that the allowable range of clearance can be expanded, even at the same confidence level, if information on the radionuclide distribution is available.