• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.83-91
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• 2018

While using an electrokinetic method to analyze the characteristics of cement solidification of radioactive wastes from decontaminated uranium soil and concrete, the compressive strength, pH, electrical conductivity, irradiation effects, and volume expansion were measured for the solidified cement specimens. The workability of cement solidified from radioactive waste was about 170-190%. After the solidified cement was irradiated, the compressive strength decreased by about 15%, but met the criteria ($34kgf{\cdot}cm^{-2}$) of KORAD (Korea Radioactive Waste Agent). According to the results of SEM-EDS for solidified cement, the aluminum phase was well combined with cement, while the calcium phase was separated from cement. The volume of solidified cement in radioactive wastes was dependent on the waste-to-cement ratio and the amount of water, and increased by about 30% under the conditions used in this study. Therefore, it was concluded that permanent disposal of electrokinetically decontaminated radioactive wastes is appropriate.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.489-504
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• 2006

A review has been made for the previous studies on safety of a geologic repository for high-level radioactive wastes (HLW) related to autocatalytic criticality phenomena with positive reactivity feedback. Neutronic studies on geometric and materials configuration consisting of rock, water and thermally fissile materials and the radionuclide migration and accumulation studies were performed previously for the Yucca Mountain Repository and a hypothetical water-saturated repository for vitrified HLW. In either case, it was concluded that it would be highly unlikely for an autocatalytic criticality event to happen at a geologic repository. Remaining scenarios can be avoided by careful selection of a repository site, engineered-barrier design and conditioning of solidified HLW. Thus, criticality safety should be properly addressed in regulations and site selection criteria. The models developed for radiological safety assessment to obtain conservatively overestimated exposure dose rates to the public may not be used directly for the criticality safety assessment, where accumulated fissile materials mass needs to be conservatively overestimated. The models for criticality safety also require more careful treatment of geometry and heterogeneity in transport paths because a minimum critical mass is sensitive to geometry of fissile materials accumulation.

• Journal of the Korean Ceramic Society
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• v.25 no.2
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• pp.101-110
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• 1988

In order to improve the physical properties of concrete used for treatment and disposal container of low-and intermediate-level radioactive wastes, OPC (ordinary portland cement), ACPC (asphalt coated portland cement) and EPC(epoxy-portland cement) concrete specimens were prepared, and the physical properties of each concrete specimen were tested. According to the experimental results, EPC concrete showed better physical properties than ACPC and OPC concrete, however, ACPC concrete proved to be a best material for treatment and disposal container of radwastes in view of economic aspect and physical properties.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.194-213
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• 1980

The paper describes and quantifies the wastes arising in the nuclear fuel cycle for Light Water Reactors, Heavy Water Reactors and Fast Breeder Reactors. The management and disposal technologies are indicated, together with their environmental impacts. Both once-through and uranium-plutonium recycle systems are evaluated, and comparisons are made on the basis of tingle reference technologies for waste management, and for one gigawatt/year of electricity generation. Environmental impacts are assessed, particularly that of health and safety, and a reference costing system is applied purely as a basis for comparing the fuel cycles. From this study it call be concluded generally that the relative differences of the impacts of waste management and disposal between the selected fuel cycles are not decisive factors in choosing a fuel cycle. Employing the technologies assumed, the radioactive wastes from any of the fuel cycles studied can be managed and disposed of with a high degree of safety and without undue risk to man or the environment. The cost of waste management and disposal is only a few percent of the value of the electricity generated and does not vary greatly between fuel cycles.

• 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.

• Journal of the Korean Ceramic Society
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• v.52 no.2
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• pp.92-102
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• 2015

Current plans for legacy nuclear wastes stored in underground tanks at the U.S. Department of Energy's Hanford Site in Washington are that they will be separated into high-level waste and low-activity waste fractions that will be vitrified separately. Formulating optimized glass compositions that maximize the waste loading in glass is critical for successful and economical treatment and immobilization of these nuclear wastes. Glass property-composition models have been developed and applied to formulate glass compositions for various objectives for the past several decades. Property models with associated uncertainties combined with composition and property constraints have been used to develop preliminary glass formulation algorithms designed for vitrification process control and waste-form qualification at the planned waste vitrification plant. This paper provides an overview of the current status of glass property-composition models, constraints applicable to Hanford waste vitrification, and glass formulation approaches that have been developed for vitrification of hazardous and highly radioactive wastes stored at the Hanford Site.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.293.1-293.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.283.2-283
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.282.1-282
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.297-297
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• 2002