• Proceedings of the KIEE Conference
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• 1997.07e
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• pp.1660-1662
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• 1997

High temperature arc plasma technologies are recently being developed in Europe, Japan and United States as one or the treatment schemes of municipal wastes, industrial wastes and vitrification of low level radioactive wastes. An experimental plasma melting furnace, a transferred type plasma torch and 100kW class power supply have been made. Operation of this system and some basic experimental results for solid wastes treatment are reported.

• 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.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.335-342
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• 2011

As part of the Department of Energy's Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.341-344
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• 1997

A vitrification process is under development in KEPRI for the treatment of low-and medium-level radioactive waste. Although the project is for developing and building Vitrification Pilot Plant in Korea, one of KEPRI's concerns is the quality control of the vitrified glass. This paper discusses a methodology for the estimation of glass composition by on-line measurement of molten glass properties, which could be applied to the plant for real-time quality control of the glass product. By remotely measuring viscosity and density of the molten glass, the glass characteristics such as composition can be estimated and eventually controlled. For this purpose, using the database of glass composition vs. physical properties in isothermal three-component system of SiO$_2$-Na$_2$O-B$_2$O$_3$, a software TERNARY has been developed which determines the glass composition by using two known physical properties(e.g. density and viscosity).

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1191-1198
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• 2023

Bentonite buffer material is a critical component in an engineered barrier system (EBS) for disposing high-level radioactive waste (HLW). The bentonite buffer material protects the disposal canister from groundwater penetration and releases decay heat to the surrounding rock mass; thus, it should possess high thermal conductivity, low hydraulic conductivity, and moderate swelling pressure to safely dispose the HLWs. Bentonite clay is a suitable buffer material because it satisfies the safety criteria. Several additives have been suggested as mixtures with bentonite to increase the thermal-hydraulic-mechanical-chemical (THMC) properties of bentonite buffer materials. Therefore, this study investigated the geotechnical, mineralogical, and THMC properties of several candidate additives such as sand, graphite, granite, and SiC powders. Datasets obtained in this study can be used to select adequate additives to improve the THMC properties of the buffer material.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.205-215
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• 2008

SITES(Site Information and Total Environmental Data Management System) has been developed for the purpose of systematically managing site characteristics and environmental data produced during the pre-operational, operational, and post-closure phases of a radioactive waste disposal facility. SITES is an integration system, which consists of 4 modules, to be available for maintenance of site characteristics data, for safety assessment, and for site/environment monitoring; site environmental data management module(SECURE), integrated safety assessment module(SAINT), site/environment monitoring module(SUDAL) and geological information module for geological data management(SITES-GIS). Each module has its database with the functions of browsing, storing, and reporting data and information. Data from SECURE and SUDAL are interconnected to be utilized as inputs to SAINT. SAINT has the functions that multi-user can access simultaneously via client-server system, and the safety assessment results can be managed with its embedded Quality Assurance feature. Comparison between assessment results and environmental monitoring data can be made and visualized in SUDAL and SITES-GIS. Also, SUDAL is designed that the periodic monitoring data and information could be opened to the public via internet homepage. SITES has applied to the Wolsong low- and intermediate-level radioactive waste disposal center in Korea, and is expected to enhance the function of site/environment monitoring in other nuclear-related facilities and also in industrial facilities handling hazardous materials.

• Journal of the Korean Ceramic Society
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• v.41 no.3
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• pp.210-215
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• 2004

Porous ceramics for water-permeable paving brick was prepared by the sintering of mixed materials comprising of sewage sludge ash, waste porcelain fragment, waste glaze and low-grade clay at 1,000$^{\circ}C$ for 2 h, and the physical $.$mechanical properties, the permeability and the freeze-thaw resistance of specimens with preparation parameters were investigated. The physical mechanical properties were increased in specimens while porosity and permeability were decreased with increasing sewage sludge ash content and sintering temperature on the properties of specimens showed the opposite results. The bulk density, porosity, compressive strength and permeability (passed charge) of 30A60F specimens with 30 wt% of sewage sludge ash content, waste porcelain fragment size with 1∼2 mm and sintered at 1,000$^{\circ}C$ for 2 h were 2.17, 46.2%, 221 kgf/$\textrm{cm}^2$ and 3,150 coulombs, respectively. The permeability was increased with increasing waste porcelain fragment size, however compressive strength was decreased. The freeze-thaw resistance of 30A60F specimen with 1∼2 mm of fragment size was superior to that of the other specimens. The 30A60F specimens can be used for the water-permeable paving brick with the high permeability and adequate strength. The heavy metals included in the all specimens showed lower than the standard level.

• Journal of the Korean Recycled Construction Resources Institute
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• v.6 no.1
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• pp.88-94
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• 2011

To make recycling aggregate, quantity of fine particles increase due to multi-crushing. Though this particles were mixed with recycling aggregate, those have to be disparted from aggregate in the high quality recycling aggregate, because of the cause of low quality. Considering reactivity, fine particles is better than coarse one. Therefore, it needs to develop suitable usage. We try to make cement extruding material by using the fine particles from concrete recycling, as a silicious replacement. Test results are as follows ; 1) Waste concrete powder has major ingredients such as $SiO_2$ and CaO, its density is $2.45g/cm^3$ being similar to silica powder, its diameter is range 13 to $141{\mu}m$. 2) Considering to strength properties according to particle size, specimen was made using small particles is higher strength than large one. 3) Despite of exception in the autoclaved curing, when the replacement of waste fine particle increase, strength of extruding panel shows almost same level.

• Tunnel and Underground Space
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• v.29 no.4
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• pp.215-229
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• 2019

This technical note describes the current status of Finnish radioactive waste disposal project which started to construct the repository for spent nuclear waste for the first time in the world. Finland started operating nuclear power plant in 1977 and is currently operating four nuclear power plants. After detailed site surveys started in 1993, Olkiluoto was finally selected by the parliament of Finland as the site for geological disposal in 2001 followed by a construction license in 2015. If the operating license is approved by the government in the 2020s, it would be the world's first case of geological disposal. In ONKALO, a site-specific underground research facility at the site of Olkiluoto, various studies were conducted to verify the safety of the repository. Finland uses the KBS-3 disposal concept, and Korea considers a similar disposal concept because of similar rock formations. The entire process in Finland including the operation status of intermediate and low-level waste disposal, site investigation and selection stages, and the latest rock mechanics and hydrogeological studies in ONKALO are presented. Suggestions for the radioactive waste disposal in Korea is given based on the Finnish case.

• Journal of Korean Tunnelling and Underground Space Association
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• v.26 no.3
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• pp.191-208
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• 2024

In this study, a hydro-mechanical-damage coupled analysis model was developed to evaluate the structural safety of radioactive waste disposal structures. The Mazars damage model, widely used to model the fracture behavior of brittle materials such as rocks or concrete, was coupled with conventional hydro-mechanical analysis and the developed model was verified via theoretical solutions from literature. To derive the numerical input values for damage-coupled analysis, uniaxial compressive strength and Brazilian tensile strength tests were performed on concrete samples made using the mix ratio of the disposal concrete silo cured under dry and saturated conditions. The input factors derived from the laboratory-scale experiments were applied to a two-dimensional finite element model of the concrete silos at the Wolseong Nuclear Environmental Management Center in Gyeongju and numerical analysis was conducted to analyze the effects of damage consideration, analysis technique, and waste loading conditions. The hydro-mechanical-damage coupled model developed in this study will be applied to the long-term behavior and stability analysis of deep geological repositories for high-level radioactive waste disposal.