• Nuclear Engineering and Technology
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• 제31권6호
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• pp.83-100
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• 1999

As disposal packaging concepts of spent fuels generated from the domestic NPP, two types, one is to package PWR and CANDU spent fuels in different containers and the other is to package them together, were proposed. The configuration of the containers and the layout of underground repository, such as the container spacing and the deposition tunnel spacing, were developed. The layout of underground repository satisfies the thermal constraint of the bentonite buffer surrounding disposal container, which should be lower than $100^{\circ}C$ in order to keep the physical and chemical properties of bentonite From the spent fuel packaging concepts and container emplacement methods, seven options were developed. With a typical pair-wise comparison methods, AHP, the most promising disposal concept was selected based on the technology Point of view.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.432-443
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• 1999

Using a two-dimensional numerical model, nuclide transport in the buffer between the canister and adjacent rock in a high-level radioactive waste repository is dealt with. Calculations are made for a typical case with a three-member decay chain, $^{234}$ U longrightarrow $^{230}$ Th longrightarrow $^{226}$ Ra. The solution method used here is based on a physically exact formulation by a control volume method directly integrating the governing equation over each control volume.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.588-599
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• 2019

This research presents a structural design of high-level waste (HLW) container using ultra-high performance fiber reinforced concrete (UHP-FRC) material. The proposed design aims to overcome the drawbacks of the existing concrete containers which are heavy, difficult to fabricate, and expensive. In this study, the dry storage container (DSC) that commonly used at Canadian Nuclear facilities is selected to present the proposed design. The design has been performed such that the new UHP-FRC alternative has a structural stiffness equivalent to the existing steel-concrete-steel container under various loading scenarios. Size optimization technique is used with the aim of maximizing stiffness, and minimizing the cost while satisfying both the design stresses and construction requirements. Then, the integrity of the new design has been evaluated against accidental drop-impact events based on realistic drop scenarios. The optimization results showed: the stiffness of the UHP-FRC container (300 mm wall thick) is being in the range of 1.35-1.75 times the stiffness of existing DSC (550 mm wall thick). The use of UHP-FRC leads to decrease the container weight by more than 60%. The UHP-FRC container showed a significant enhancement in performance in comparison to the existing DSC design under considered accidental drop impact scenarios.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 춘계학술대회 논문집
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• pp.1636-1639
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• 2005

The process equipment and remote handling for the deep geological disposal of high-level radioactive waste(HLW) should be checked prior to the operation in view of reliability and operability. In this study, the concept of virtual environment workcell is implemented to analyze and define the feasible disposal process instead of real mock-up, which is very expensive and time consuming. To do this, the parts of process equipment for the disposal and maintenance will be modeled in 3-D graphics, assembled, and kinematics will be assigned. Also, the virtual workcell for the encapsulation and disposal process of spent fuel will be implemented in the graphical environment, which is the same as the real environment. This virtual workcell will have the several functions for verification such as analyses for the equipment's work space, the collision detection, the path planning and graphic simulation of the processes etc. This graphic virtual workcell of the HLW disposal process can be effectively used in designing of the processes for the hot cell equipment and enhance the reliability of the spent fuel management.

• 방사성폐기물학회지
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• 제19권3호
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• pp.331-338
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• 2021

Extensive studies have been conducted on thermal conductivity of bentonite buffer materials, as it affects the safety performance of barriers engineered to contain high-level radioactive waste. Bentonite is composed of several minerals, and studies have shown that the difference in the thermal conductivity of bentonites is due to the variation in their mineral composition. However, the specific reasons contributing to the difference, especially with regard to the thermal conductivity of bentonites with similar mineral composition, have not been elucidated. Therefore, in this study, bentonites with significantly different thermal conductivities, but of similar mineral compositions, are investigated. Most bentonites contain more than 60% of montmorillonite. Therefore, it is believed that the exchangeable cations of montmorillonite could affect the thermal conductivity of bentonites. The effect of bentonite type was comparatively analyzed and was verified through the effective medium model for thermal conductivity. Our results show that Ca-type bentonites have a higher thermal conductivity than Na-type bentonites.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• 제1권1호
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• pp.9-27
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• 2013

Nagra was established in 1972 by the Swiss nuclear power plant operators and the Federal Government to implement permanent and safe disposal of all types of radioactive waste generated in Switzerland. The Swiss Nuclear Energy Act specifies that these shall be disposed of in deep geological repositories. A number of different geological formations and sites have been investigated to date and an extended database of geological characteristics as well as data and state-of-the-art methodologies required for the evaluation of the long-term safety of repository systems have been developed. The research, development, and demonstration activities are further supported by the two underground research facilities operating in Switzerland, the Grimsel Test Site and the Mont Terri Project, along with very active collaboration of Nagra with national and international partners. A new site selection process was approved by the Federal Government in 2008 and is ongoing. This process is driven by the long-term safety and feasibility of the geological repositories and is based on a step-wise decision-making approach with a strong participatory component from the affected communities and regions. In this paper a brief history and the current status of the Swiss radioactive waste management program are presented and special characteristics that may be useful beyond the Swiss program are highlighted and discussed.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.1041-1048
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• 2021

An engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW) is composed of a disposal canister with spent fuel, a buffer material, a gap-filling material, and a backfill material. As the buffer is located in the empty space between the disposal canisters and the surrounding rock mass, it prevents the inflow of groundwater and retards the spill of radionuclides from the disposal canister. Due to the fact that the buffer gradually becomes saturated over a long time period, it is especially important to investigate its thermal-hydro-mechanical-chemical (THMC) properties considering variations of saturated condition. Therefore, this paper suggests a new method of measuring thermal conductivity and water suction for single compacted bentonite at various levels of saturation. This paper also highlights a convenient method of saturating compacted bentonite. The proposed method was verified with a previous method by comparing thermal conductivity and water suction with respect to water content. The relative error between the thermal conductivity and water suction values obtained through the proposed method and the previous method was determined as within 5% for compacted bentonite with a given water content.

• 자원환경지질
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• 제55권6호
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• pp.737-760
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• 2022

고준위 방사성 폐기물(High-level radioactive waste; HLW)의 지질처분을 위해서는 심부 지하 환경에 대한 이해가 선행되어야 하며, 이는 지질학적, 수리지질학적, 지구화학적, 지질공학적 조사를 통해 가능하다. 우리나라는 HLW의 지질처분을 계획하고 있으나, 심부 지하 환경의 지구화학적 특성에 관한 연구가 부족한 편이다. 이에 본 논문에서는 지질처분 부지 선정을 위한 지구화학적 조사를 중심으로 선진국의 심부 지하수 연구 동향을 살펴봄으로써 앞으로 국내 수리지구화학 분야의 연구 과제를 도출하는데 참고하고자 하였다. 해외 8개 국가(미국, 캐나다, 핀란드, 스웨덴, 프랑스, 독일, 일본, 스위스)의 심부 지하 환경 조사 방법 및 결과와 함께 지질처분 부지 결정 과정과 향후 연구 계획을 살펴본 결과, 해외 선진국에서는 심부 지하 환경의 지구화학적 특성화를 위해 지하수 및 난대수층 내 간극수의 수화학과 동위원소(예: SO₄^2-의 ³⁴S, ¹⁸O, DIC의 ¹³C, ¹⁴C, H₂O의 ²H, ¹⁸O), 균열 충전광물(fracture-filling minerals), 유기물, 콜로이드, 산화-환원 지시자(예: Eh, Fe²⁺/Fe³⁺, H₂S/SO₄^2-, NH₄⁺/NO₃^-) 등을 조사하고 있으며, 이들 지구화학 자료의 통합 해석을 통해 해당 심부 환경이 지질처분에 적합한지를 평가하였다. 국내의 경우, 인공신경망을 이용한 Self-Organizing Map(자기조직화 지도), 다변량 통계 기반 M3 모델링(지하수 혼합 모델), 반응-경로 모델(reaction path model) 등을 이용하여 심부 지하수의 수화학적 유형 분류 및 진화 패턴 규명, 천부 지하수 혼합 영향, 균열 충전광물과 지하수화학 사이의 관계를 규명한 바 있다. 그러나 지질처분 부지를 선정하는데 있어 과학적 근거를 확보하기 위해 중요한 기타 지구화학 자료(예: 동위원소, 산화-환원 지시자, 용존유기물)가 매우 부족한 현실이며, 따라서 최적의 지질 처분지를 찾기 위해서는 지역별/유형별 심부 지하수에 대한 지구화학적 자료 구축이 요구된다.

• 터널과지하공간
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• 제14권6호
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• pp.429-439
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• 2004

국내 원전에서 발생되는 36,000톤의 사용후핵연료를 처분하기 위해서는 약 $4km^2$의 지하 처분장이 필요하다. 본 연구에서는 굴착량과 처분장 면적을 최소화하기 위한 지하 심부 처분장 배치의 최적화를 실시하였다. 열 해석 결과를 토대로 처분 터널과 처분공 간격이 처분장 배치에 미치는 영향을 고려한 결과, 처분장 면적과 굴착량은 처분 터널의 길이가 길어짐에 따라 감소하였다. 주어진 열적 기준을 만족하면서 처분장 면적을 줄이기 위해서는 처분 터널의 간격을 줄이고 처분공 간격을 늘리는 것이 유리하였으며, 반면에 굴착량을 최소화하는 경우 처분공 간격을 줄이고 처분 터널 간격을 늘려주는 것이 효과적인 것으로 나타났다.

• 방사성폐기물학회지
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• 제8권3호
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• pp.207-213
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• 2010

국산 칼슘 벤토나이트를 대상으로 온도가 팽윤압에 미치는 영향을 관찰하였다. 벤토나이트를 건조밀도 1.6 g/$cm^3$으로 압축하고, 0.69 MPa의 일정한 수압으로 증류수를 공급하여 팽윤압을 측정하였다. 온도 영향 실험은 $25^{\circ}C$, $30^{\circ}C$, $40^{\circ}C$, $50^{\circ}C$, $60^{\circ}C$, $70^{\circ}C$, respectively. The Ca-bentonite showed a sufficiently high swelling pressure of 5.3 MPa에서 승온조건과 감온조건으로 수행하였다. 압축 벤토나이트가 물과 접촉하여 상온에서 5.3 MPa의 충분히 높은 팽윤압이 작용하는 것을 실험적으로 확인하였다. 팽윤압은 온도가 높을수록 감소하는 것으로 나타났다. 승온조건과 감온조건에서의 온도에 따른 팽윤압 거동에 차이를 보이며, 승온조건에서 온도에 따른 변화가 심하게 나타났다. 향후 온도 조건 외에 벤토나이트의 압축밀도 변화, 지하수 조성에 따라 팽윤압 특성이 어떻게 변화하는지에 대해 평가한다면, 앞으로 국내 고준위 폐기물 처분장의 개념 설계에 유용하게 활용될 수 있을 것으로 본다.