• Explosives and Blasting
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• v.27 no.1
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• pp.21-31
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• 2009

An excavation damaged zone (EDZ) is created by fracturing, excavation or stress redistribution of tunnels. In this zone the mechanical and hydraulic properties of rock are changed, which makes additional cracks and serves as a dominant pathway of groundwater flow. In this study, an assessment on an EDZ size was practiced by the measurement of the deformation modulus at the KAERI underground research tunnel (KURT), and the information was applied to the modelling analysis using FLAC2D software. The EDZ at KURT fell into the range of 0.6~1.8m and the deformation moduli of the EDZ generally correspond to about 40% of intact rock mass. With a consideration of the EDZ in numerical analysis, tunnel displacements increased by about 65% and the maximum principal stress decreased to 58% from the case without EDZ. The plastic zone of the tunnel was enlarged to the crown and invert zones of the tunnel within the range of the length of rock bolts. About 2% of the total tunnel displacement with EDZ was suppressed by the KURT support system. It is anticipated that the investigation of an EDZ can be used as an important and fundamental research for validating the overall performance of a high level waste disposal system.

• Journal of the Korean Recycled Construction Resources Institute
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• v.9 no.4
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• pp.451-459
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• 2021

The purpose of this study is to evaluate the performance of a solidifying agent for recycling the fine powder separated from the nuclear power plant decommissioned concrete as a solidifying agent(SA) for radioactive waste. In order to evaluate the performance of the solidifying agent, a powder simulating the fine powder of waste concrete separated from the dismantled concrete of a nuclear power plant was produced, and the main variables were the type of binder and the replacement ratio of zeolite. The solidifying agent was evaluated for fluidity performance, compressive strength, and leaching resistance to non-radioactive cesium. The compressive strength of SA increased as the zeolite replacement ratio increased, and the SA containing 5% or more of zeolite showed a compressive strength that was 1.4 to 1.7 times higher than the acceptance criteria. The cesium leaching index of all specimens was 6 or higher, satisfying the acceptance criteria, and the leaching index of SA was 1.47~1.63 times higher than that of OPC. In particular, the average leaching index after 28 days of the 5% zeolite-substituted solidifying agent was 9.15, which was improved by about 6.4% compared to OPC, and it was confirmed that the zeolite was effective in improving the leaching resistance to cesium ions by showing stable performance over the entire period.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.285-299
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• 2006

At the end of 2005, 443 nuclear reactors were operating in 32 countries worldwide. They had provided about 3,000 TWh, which was just over 16 percent of global electricity supply. With the generating capacity of 368 GWe in 2004, the spent fuel generation rate worldwide, now becomes at about 11,000 tHM/y. Projections indicate that cumulative amounts to be generated by the year 2020, the time when most of the existing NPP will be closed to the end of their licensed lifetime, may be close to 445,000 tHM. In this regard, spent fuel management is a common issue in all countries with nuclear reactors. Whatever their national policy and/or strategy is selected for the backend of the nuclear fuel cycle, the management of spent fuel will contribute an impending and imminent issues to be resolved in the foreseeable future. The 2nd Review Meeting of the Contracting Parties to the Joint Convention was held in Vienna from 15 to 24 May 2006. The meeting gave an opportunity to exchange information on the national policy and strategy of spent fuel management of the Contracting Parties, to discuss their situations, prospects and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should be taken. In this paper, an overview of national and global trends of spent fuel management is discussed. In addition, some directions are identified and recent activities of each Member States in the subject area are summarized.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.235-243
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• 2006

In this study, an experimental study on the sorption properties of uranium(VI) onto a bentonite colloid generated from Gyeongju bentonite which is a potential buffer material in a high-level radioactive waste repository was performed as a function of the pH and the ionic strength. The bentonite colloid prepared by separating a colloidal fraction was mainly composed of montmorillonite. The concentration and the size fraction of the prepared bentonite colloid measured using a gravitational filtration method was about 5100 ppm and 200-450 nm in diameter, respectively. The amount of uranium removed by the sorption reaction bottle walls, by precipitation, and by ultrafiltration was analyzed by carrying out some blank tests. The removed amount of uranium was found not to be significant except the case of ultrafiltration at 0.001 M $NaClO_4$. The ultrafiltration was significant in the lower ionic strength of 0.001 M $NaClO_4$ due to the cationic sorption onto the ultrafilter by a surface charge reversion. The distribution coefficient $K_d$ (or pseudo-colloid formation constant) of uranium(VI) for the bentonite colloid was about $10^4{\sim}10^7mL/g$ depending upon pH and ionic strength of $NaClO_4$ and the $K_d$ was highest in the neutral pH around 6.5. It is noted that the sorption of uranium(VI) onto the bentonite colloid is closely related with aqueous species of uranium depending upon geochemical parameters such as pH, ionic strength, and carbonate concentration. As a consequence, the bentonite colloids generated from a bentonite buffer can mobilize the uranium(VI) as a colloidal form through geological media due to their high sorption capacity.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.13-19
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• 2012

The long-lived fission products $^{79}Se$ and $^{99}Tc$ have been considered as the major concern nuclides for the disposal of radioactive waste because of their high solubilities and the existence of anionic species in natural water. In this study, the solubilities of $FeSe_2(s)$ and $TcO_2(s)$, known as respective Solubility Limiting Solid Phase (SLSP) of selenium and technetium, were measured in the KURT (KAERI Underground Research Tunnel) groundwater under various pH and redox conditions. And their solubilities and major species were also calculated using geochemical codes under conditions similar to experimental solutions. Experimental results and calculation for $FeSe_2$ show that the solubility of selenium was found to be below $1{\times}10^{-6}mol/L$ under the condition of pH 8~9.5 and Eh=-0.3~-0.4 V while the dominant species was identified as $HSe^-$. For $TcO_2$, the solubility of technetium was found to be $5{\times}10^{-8}{\sim}1{\times}10^{-9}mol/L$ in the solutions of pH 6~9.5 and Eh<-0.1 V, while the dominant species was $TcO(OH)_2$. However, when the Eh of the solution is -0.35 V, $TcO(OH)_3^-$ and $TcO_4^-$ are calculated as the dominant species at pH 10.5~12 and pH>12, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.159-166
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• 2010

The block-scale groundwater flow system at Olkiluoto site in Finland was simulated. The heterogeneous and anisotropic hydraulic conductivity field for the domain was constructed from the discrete fracture network, which considered only the fractured zones identified in the deep boreholes installed in the study site. The groundwater flow model was calibrated by adjusting the recharge rate and the transmissivities of the fractured zones to fit the calculated hydraulic heads and into- and out-flow rates in the observation intervals of the boreholes with the observed ones. In the calibrated model, the calculated flow rates at some intervals were not in accordance with the observed ones although the calculated hydraulic heads fit well with the observed ones, which revealed that the number of the conduits for groundwater flow is insufficient in the conceptual model for groundwater flow modeling. Therefore, it was recommended that the potential local conduits such as background fractures should be added to the present conceptual model.

• Economic and Environmental Geology
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• v.52 no.2
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• pp.189-197
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• 2019

HLW (High Level Radioactive Waste) is one of the problems that must be solved in the countries that implement nuclear power generation. Most countries that are concerned about HLW treatment are considering complete isolation from human society by disposing them deep underground. For perfect isolation, understanding the characteristics of underground rocks is very important. In particular, understanding the characteristics of discontinuity as a path way is one of the first things in order to predict the movement of exposed nuclear species to the surface. In this study, we used 500m underground core samples obtained from granite and gneiss area. The purpose of this study is to understand the characteristics of the discontinuities in each rock type and to analyze the properties of the joints in the underground relative to the surrounding environment. For this purpose, the types of discontinuities were classified and the distribution of each discontinuity were analyzed through visual analysis of the each sample obtained at 500m underground. This study can be used as a basic data for understanding the properties of discontinuities in the rock of the survey area and it can be also used as an important data for understanding the distribution characteristics of discontinuities according to the rock types.

• Korean Journal of Mineralogy and Petrology
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• v.36 no.2
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• pp.107-115
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• 2023

Cobalt can be released into the natural environment as industrial waste from the alloying industry and as acid mine drainage, and it is also a radionuclide (⁶⁰Co) that constitutes high-level radioactive waste. Smectite is a mineral that can be useful for adsorption and isolation of this element. In this study, Cheto-type montmorillonite (Cheto-MM), which is the source clays of The Clay Mineral Society (CMS) and already well-characterized, was used. The effect of the adsorption site affected by the presence of interlayer water on the adsorption of cobalt before and after dehydration by heating was evaluated and the adsorption mechanism of cobalt on Cheto-MM was studied by applying adsorption kinetics and adsorption isotherm models. The results showed that the adsorption characteristics changed with dehydration and subsequent shrinkage, and cobalt was found to be adsorbed at the edge of Cheto-MM for about 38% and adsorbed at the interlayer site for about 62%, suggesting that the cobalt adsorption of Cheto-MM is significantly influenced by the interlayer. By applying the adsorption kinetic models, the cobalt adsorption kinetics of Cheto-MM is explained by a pseudo-second-order model, and the concentration-dependent adsorption was best described by the Langmuir isotherm adsorption model. This study provides basic knowledge on the adsorption characteristic of cobalt on montmorillonite with different adsorption sites and is expected to be useful in predicting the adsorption behavior of smectite in high-level radioactive waste disposal sites in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.249-256
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• 2005

Thermal decomposition and stabilization characteristics of the solid cake after the dissolution of nitrate of the lagoon sludge was investigated. Most of the nitrates were dissolved in the water and removed to the filtrate, but small amount of nitrates, calcium carbonate and uranium were remained in the solid cake. The solid cake was thermally decomposed in the muffle furnace at $900^{\circ}C$ for 5 hours. Uranium, which is in the lagoon 1, was stabilized with $NaNO_3$ decomposition to $Na_{2}O{\cdot}2UO_3$ form. For the lagoon 2, it is confirmed that CaO, which was created by thermal decomposition of the $Ca(NO_3)_2$ and $CaCO_3$, was transferred to $Ca(OH)_2$ in the air with water. Because it is known that $Ca(OH)_2$ is stable material, further additives did not need to the stabilization of the thermal decomposition of the lagoons.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.329-333
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• 2005

To identify the effect of engineered barriers on the leach rate of cesium from spent PWR fuel under a synthetic granitic groundwater, the related leach tests with and without bentonite or metals have been performed up to about 6 years. The leach rates were decreased as a function of leaching time and then became a constant after a certain period. The period in a bare spent fuel was much longer than that with bentonite or metal sheets. The cumulative fraction of cesium released from the spent fuel with bentonite or with copper and stainless steel sheets was steadily increased, but the fraction from bare fuel was rapidly and then sluggishly increased. However, the values deducted its gap inventory from the cumulative fraction of cesium released from the bare fuel was almost very close to the others. These suggest that the initial release of cesium from bare fuel might be dependant on its gap inventory and the effect of engineered barriers on the long-term leach rate of cesium would be insignificant but the rate with engineered barriers could be reduced in the initial transient period due to their retardation effect. And the long-term leach rate of cesium from spent fuel in a repository would be approached to a constant rate of $2\times10^{-2}g/m^2-day$.