• Journal of the Korea Concrete Institute
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• v.25 no.3
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• pp.321-329
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• 2013

Recently, the medium-low level radioactive waste from nuclear power plant must be transported from temporary storage to the final repository. Medium-low level radioactive waste, which is composed mainly of the liquid ion exchange resin, has been consolidated with cementitious material in the plastic or iron container. Since cementitious material is brittle, it would generate cracks by impact load during transportation, signifying leakage of radioactive ray. In order to design the safety transporting equipment, there is a need to check the compressive strength of the current waste. However, because it is impossible to measure strength by direct method due to leakage of radioactive ray, we will estimate the strength indirectly by the dynamic modulus of elasticity. Therefore, it must be identified the relationship between of strength and dynamic modulus of elasticity. According to the waste acceptance criteria, the compressive strength of cement based solid is defined as more than 3.44 MPa (500 psi). Compressive strength of the present solid is likely to be significantly higher than this baseline because of continuous hydration of cement during long period. On this background, we have tried to produce the specimens of the 28 day's compressive strength of 3 to 30 MPa having the same material composition as the solid product for the medium-low level radioactive waste, and analyze the relationship between the strength and the dynamic modulus of elasticity. By controling the addition rates of AE agent, we made the mixture containing the ion exchange resin and showing the target compressive strength (3~30 MPa). The dynamic modulus of elasticity of this mixtures is 4.1~10.2 GPa, about 20 GPa lower in the equivalent compressive strength level than that of ordinary concrete, and increasing the discrepancy according to increase strength. The compressive strength and the dynamic modulus of elasticity show the liner relationship.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.229-238
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• 2007

In this paper, structural design requirements and safety evaluation criteria of the spent nuclear fuel disposal canister are studied for deep geological deposition. Since the spent nuclear fuel disposal canister emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for the spent nuclear fuel disposal canister should be secured. Usually this repository is expected to locate at a depth of 500m underground. The canister which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock is a solid structure with cast iron insert, corrosion resistant overpack and lid and bottom, and entails an evenly distributed load of hydrostatic pressure from underground water and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. If the canister is not designed for all possible external loads combinations, structural defects such as plastic deformations, cracks, and buckling etc. may occur in the canister during depositing it in the deep repository. Therefore, various structural analyses must be performed to predict these structural problems like plastic deformations, cracks, and buckling. Structural safety evaluation criteria of the canister are studied and defined for the validity of the canister design prior to the structural analysis of the canister. And structural design requirements(variables) which affect the structural safety evaluation criteria should be discussed and defined clearly. Hence this paper presents the structural design requirements(variables) and safety evaluation criteria of the spent nuclear fuel disposal canister.

• Tunnel and Underground Space
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• v.11 no.2
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• pp.132-145
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• 2001

The purpose and need of the study is to quantify the advantage or disadvantage of the environmental friendliness of the partitioning of nuclear fuel cycle. To this end, a preliminary study on the quantitative effect of the partition on the permanent disposal of spent PWR and CANDU fuel (HLW) was carried out. Before any analysis, the so-called reference radionuclide release scenario from a potential repository embedded into a crystalline rock was developed. Firstly, the feature, event and processes (FEPs) which lead to the release of nuclides from waste disposed of in a repository and the transport to and through the biosphere were identified. Based on the selected FEPs, the ‘Well Scenario’which might be the worst case scenario was set up. For the given scenario, annual individual doses to a local resident exposed to radioactive hazard were estimated and compared to that from direct disposal. Even though partitioning and transmutation could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and also minimize the repository area through the proper handling of nuclides, it should overcome major disadvantages such as echnical issues on the partitioning and transmutation system, cost, and public acceptance, and environment friendly issues. In this regard, some relevant issues are also discussed to show the direction for further studies.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.357-368
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• 2008

On July 31, 2008, the Government issued the construction and operation permit for the first low and intermediate level radioactive waste disposal facility in the Republic of Korea. In this paper, the fundamental regulatory framework, regulatory requirements and technical standards of the disposal facility are introduced, and the phased review process adopted for evaluation of the safety of the facility is briefly described. The Atomic Energy Act sets forth a stepwise regulatory framework for the whole life-cycle of the disposal facility such as siting, design, construction, operation, closure and institutional control. More detailed regulatory requirements and technical standards are stipulated in the subsequent regulations of the Atomic Energy Act and a series of Notices issued by the Ministry of Eduction, Science and Technology. The Korea Institute of Nuclear Safety, as entrusted by the Ministry under the Atomic Energy Act, conducted safety review on the disposal facility, and evaluated the compliance with relevant criteria in all technical elements(i.e. siting and structural safety, radiological environmental impact, operational safety, systems and components, quality assurance, and total systematic performance assessment, etc.). The overall safety review process can be phased into inception phase, initial review phase, main review phase and completion phase. The review results were reported to and deliberated by the five Sub-committees of the Special Committee on Nuclear Safety, and then reported to the Ministry. The Ministry issued the construction and operation permit of the disposal facility through the deliberation of the review results by the Nuclear Safety Commission. Hereafter, the safety of the repository will be reassured by a series of subsequent regulatory inspections and reviews under the Atomic Energy Act. In addition, the licensee's continuous implementation of the "Safety Promotion Plan" may also enhance the long-term safety of the repository and contribute to build-up the confidence of the safety case.

• Proceedings of the Korean System Dynamics Society
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• 2006.04a
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• pp.123-149
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• 2006

본 연구는 방사성폐기물 처분시설(radioactive waste repository)의 입지를 가정하여, 처분시설이 경상북도 경주시에 발생시키는 경제적, 사회적 효과를 분석하는데 목적이 있다. 정부는 처분장 유치의 유인책(incentives)으로서 경주 지역주민들을 위하여 다양한 정책적 수단을 마련하였다. 처분시설 입지에 따른 특별지원금 3,000 억원 지원, 수거물 반입 수수료 지원(년당 50-100억원), 한국수력원자력(주) 본사이전, 양성자가속기 사업 추진 등이 그것이다. 이들 지원사업들이 가져오는 변화를 시스템 다이내믹스(System Dynamics) 기법을 적용하여 지역사회의 인구, 산업, 토지, SOC, 지방재정 등이 어떻게 변화하는지 추적해 봄으로써 도시체제의 동태성(urban system dynamics)을 이해하고 처분장 시설이 지역에 입지했을때, 미래에 발생 가능한 문제점이 없는지 밝혀내고자한다. 이를 위하여 시뮬레이션 모델링에 입지 지역의 특성과 현황을 반영하여 처분장입지에 따른 지역의 동태적인 변화과정과 경향을 추정해 보고, 현재 예정되어 있는 지원사업이 충분한지, 이외에 다른 정책적 지원이 필요한지를 알아본다. 본 연구의 의미는 이처럼 경주지역 주민들이 처분장의 지역입지를 만족스럽게 행각하고 소외감 없이 생활을 영위할 수 있도록 정책적 지원 프로그램을 작성하는데 기초가 되는 연구라는 점에 있다고 하겠다.

• Proceedings of the Korean Geotechical Society Conference
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• 2009.03a
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• pp.792-797
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• 2009

The stress-strain characteristics of compacted bentonite are investigated using experimental triaxial compression test by Hoek-cell. Special attention given to various dry density and water absorption ratio. Based on the test results, it is shown that the stress-strain relationship of compacted bentonite is highly influenced by dry density and water absorption ratio. Also, characteristics of Bentonite is similar to the clay rather than sand. Strength of compressed Bentonite increases with higher dry density. It shows maximum strength value, if in a same condition with dry density and constrain pressure. So we determine that value as the optimistic moisture contents for the maximun strength of compressed Bentonite.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.260-268
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• 1997

The effects of three mechanisms, calcium depletion, sulphate and carbonate penetration, on the concrete degradation have been studied. The shrinking core model (SCM) and the HYDROGEOC. HEM (HGC) model have been applied to evaluate how fast the mechanisms proceed. The SCM is an analytical approximation model and the HGC is a numerical mass transport model coupled with chemical reaction. The SCM leads to more conservative results than the HGC, and turns out to be very useful in the viewpoint of simplicity and conservatism. During 300 years, calcium has been depleted within 10 cm from the concrete outer surface, and sulphate has penetrated less than 13.5 cm into the concrete. Carbonate has not penetrated own 7 cm into the concrete in contact with the bentonite, and, furthermore, its penetration into the concrete with the groundwater is negligible. Conclusively, the concrete is expected to maintain its integrity for at least 300 years that are regarded as institutional control period of intermediate and low-level radioactive waste repository.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.101-112
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• 1990

For the performance assessment of the radioactive waste disposal system (repository), a biosphere model is suggested. This biosphere model is intended to calculate the annual doses to man caused by the contaminated river water for eight pathways and four radionuclides. This model can also be applied to assess the radiological effects of contaminated well water. To account for the uncertainties on the model parameter values, parameter distributions are assigned to these model parameters. Then, Monte Carlo simulation method with Latin Hypercube sampling technique is used. Also, sensitivity analysis is performed by using the Spearman rank correlation coefficients. It is found that these methods are a very useful tool to treat uncertainties and sensitivities on the model parameter values and to analyze the biosphere model. A conversion factor is proposed to calculate the annual dose rate to humans arising from a unit radionuclide concentration in river water. This conversion factor allows for the substitution of the biosphere model in a probabilistic performance assessment computer code by one single variable.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.37-50
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• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• Geomechanics and Engineering
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• v.2 no.3
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• pp.213-227
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• 2010

Compacted layers of sand-bentonite mixtures have been proposed and used in a variety of geotechnical projects as engineered barriers for the enhancement of impervious landfill liners, cores of zoned earth dams and radioactive waste repository systems. This paper presents a study on the valorization of local materiel such as dune sand from Laghouat region and mine bentonite intended for the realization of liner base layers in the conception of insulation barriers for hazardous waste centers. In the practice we try to get an economical mixture that satisfies the hydraulic and mechanical properties specified by regulation rules. The effect of the bentonite additions on the mixture is reflected by its capability of clogging the matrix pores upon swelling. In order to get an adequate dune sand-bentonite mixture, an investigation on hydraulic and mechanical behaviours is carried out in this study for different mixtures. Using oedometer test, the adequate bentonite addition to the mixture, which satisfies the conditions on permeability, is found to be around 12% to 15%. These results are also confirmed by direct measurement using triaxial cell.