• 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.

• The Journal of Engineering Geology
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• v.21 no.4
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• pp.295-304
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• 2011

The first step in improving our understanding of uncertainties suclt as rock mass strength parameters and deformation modulus in rock masses around high-level radioactive waste disposal repositories, for improved safety, is to study the process of crack development in intact rock. Therefore, in this study, the fracture process and crack development were examined in samples of KURT granite taken from the KAERI Underground Research Tunnel (KURT), based on acoustic emission (AE) and moment tensor analysis. The results show that crack initiation, coalescence, and unstable crack occurred at rock uniaxial compressive strengths of 0.45, 0.73, and 0.84, respectively. In addition, moment tensor analysis indicated that during the early stage of loading, tensile cracks were predominant. With increasing applied stress, the number of shear cracks gradually increased. When the applied stress exceeded the stress level required for crack damage, unstable shear cracks which directly result in failure of the rock were generated along the failure plane.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.101-109
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• 2008

For the Kyungju bentonite which is considered as a candidate material for the buffer and backfill in the high-level waste repository, the thermal conductivities of compacted bentonite and a bentonite-sand mixture were measured. The thermal conductivities of the compacted bentonites with a dry density of 1.2 to $1.8\;Mg/m^3$ and the bentonite-sand mixture with a dry density of 1.6 and $1.8\;Mg/m^3$ were measured within the gravimetric water content range of 10wt% to 20wt% and the sand fraction range of 10 to 30wt%. The thermal conductivity of compacted bentonite and a bentonite-sand mixture increases with increasing dry density and sand weight fraction in the case of constant water weight fraction, and increases with increasing water weight fraction and sand weight fraction in the case of constant dry density. The empirical correlations to describe the thermal conductivity of compacted bentonite and a bentonite-sand mixture as a function of water fraction at each dry density were suggested. These correlations can predict the thermal conductivities of bentonite and a bentonite-sand mixture with a difference below 10%.

• Economic and Environmental Geology
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• v.56 no.5
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• pp.547-555
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• 2023

Due to enhanced precision in uranium isotope measurements with MC-ICP-MS, there has been a surge in studies concerning the naturally occurring uranium isotope ratio (²³⁸U/²³⁵U) and its associated fractionation processes. Several researchers have highlighted that the ²³⁸U/²³⁵U ratio, previously assumed to be constant, can vary by several per mil depending on different natural fractionation processes. This review paper outlines the uranium isotope values (δ²³⁸U) for major terrestrial reservoirs and their variations. It discusses the range of δ238U values and uranium isotope fractionation seen in uranium ore deposits, based on deposit type and ore-forming conditions. In conclusion, this paper emphasizes the importance of studies on uranium ore deposits. Such deposits serve as natural simulation models vital for designing high-level radioactive waste repository sites.

• Explosives and Blasting
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• v.26 no.2
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• pp.11-19
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• 2008

The development of an excavation damaged zone, EDZ, due to the blasting impact and stress redistribution after excavation, can influence on the long tenn stability, economy, and safety of the underground excavation. In this study, the size and characteristics of an EDZ around an underground research tunnel, which was excavated by controlled blasting, in KAERI were investigated. The results were implemented into the modelling for evaluating the influence of an EDZ on hydro-mechanical behavior of the tunnel. From in situ tests at KURT, it was possible to determine that the size of EDZ was about l.5rn. Goodman jack tests and laboratory tests showed that the rock properties in the EDZ were changed about 50% compared to the rock properties before blasting. The size and property change in the EDZ were implemented to a hydro-mechanical coupling analysis. In the modeling, rock strengths and elastic modulus were assumed to be decreased 50% and. the hydraulic conductivity was increased 1 order. From the analysis, it was possible to see that the displacement was increased while the stress was decreased because of an EDZ. When an EDZ was considered in the model, the tunnel inflow was increased about 20% compared to the case without an EDZ.

• Tunnel and Underground Space
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• v.29 no.5
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• pp.318-331
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• 2019

A three-dimensional(3D) equivalent continuum modeling was performed to analyze hydraulic behavior of rock mass considering discontinuities by using DFN model and smeared fracture model. DFN model was generated by FLAC3D and smeared fracture model was applied by using FISH functions, which is built-in functions in FLAC3D, for equivalent continuum model of fractured rock mass. Comparative analysis with 3DEC, which is for discontinuum analysis, was conducted to verify reliability of equivalent continuum analysis by using FLAC3D. Similar results of hydraulic analysis under the same conditions could be achieved. Equivalent continuum analysis of fractured rock mass by using DFN model was implemented to compare with existing analytical methods for inflow into the tunnel.

• Tunnel and Underground Space
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• v.31 no.6
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• pp.400-414
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• 2021

Under high-level radioactive waste repository conditions, bentonite as an engineered barrier material undergoes thermal, hydrological, mechanical, and chemical processes. We report the applications of X-ray Computed Tomography (CT) imaging technique on the characterization and analysis of bentonite over the past decade to provide a reference of the utilization of this technique and the recent research trends. This overview of the X-ray CT technique applications includes the characterization of the bentonite either in pellets or powder form. X-ray imaging has provided a means to extract grain information at the microscale and identify crack networks responsible for the pellets' heterogeneity. Regarding samples of pellets-powder mixtures under hydration, X-ray CT allowed the identification and monitoring of heterogeneous zones throughout the test. Some results showed how zones with pellets only swell faster compared to others composed of pellets and powder. Moreover, the behavior of fissures between grains and bentonite matrix was observed to change under drying and hydrating conditions, tending to close during the former and open during the latter. The development of specializing software has allowed obtaining strain fields from a sequence of images. In more recent works, X-ray CT technique has served to estimate the dry density, water content, and particle displacement at different testing times. Also, when temperature was added to the hydration process of a sample, CT technology offered a way to observe localized and global density changes over time.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.387-400
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• 1991

An ocean compartment model simulating transport of nuclides by advection due to ocean circulation and intertaction with suspended sediments is developed, by which concentration breakthrough curves of nuclides can be calculated as a function of time. Dividing ocean into arbitrary number of characteristic compartments and performing a balance of mass of nuclides in each ocean compartment, the governing equation for the concentration in the ocean is obtained and a solution by the numerical integration is obtained. The integration method is specially useful for general stiff systems. For transfer coefficients describing advective transport between adjacent compartments by ocean circulation, the ocean turnover time is calculated by a two-dimensional numerical ocean model. To exemplify the compartment model, a reference case calculation for breakthrough curves of three nuclides in low-level radioactive wastes, Tc-99, Cs-137, and Pu-238 released from hypothetical repository under the seabed is carried out with five ocean compartments. Sensitivity analysis studies for some parameters to the concentration breakthrough curves are also made, which indicates that parameters such as ocean turnover time and ocean water volume of compartments have an important effect on the breakthrough curves.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.87-96
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• 2004

Studies were conducted to select the candidate buffer material for a high-level waste (HLW) repository in Korea. This paper presents the hydraulic properties, the swelling properties, the thermal properties, and the mechanical properties as well as the radionuclide release-retarding capacity of Kyungju bentonite as part of those studies. Experimental results showed that the hydraulic conductivities of the compacted bentonite were very low and less than $10^{-11}$m/s. The values decreased with increasing the dry density of the compacted bentonite. The swelling pressures were in the range of 0.66 MPa to 14.4 ㎫ and they increased with increasing the dry density. The thermal conductivities were in the range of 0.80 ㎉/m $h^{\circ}C$ to 1.52 ㎉/m $h^{\circ}C$. The unconfined compressive strength, Young's modulus and Poison's ratio showed the range of 0.55 ㎫ to 8.83 ㎫, 59 ㎫ to 1275 ㎫, and 0.05 to 0.20, respectively, when the dry densities of the compacted bentonite were 1.4 Ms/㎥ to 1.8 Mg/㎥. The diffusion coefficients in the compacted bentonite were measured under an oxidizing condition. The values were $1.7{\times}10^{-10}$m^2$/s to 3.4{\times}10^{-10}$m^2$/s for electrically neutral tritium (H-3), 8.6{\times}10^{-14}$m^2$/s to 1.3{\times}10^{-12}$m^2$/s for cations (Cs, Sr, Ni), 1.2{\times}10^{-11}$m^2$/s to 9.5{\times}10^{-11}$m^2$/s for anions (I, Tc), and 3.0{\times}10^{-14} $m^2$/s to 1.8{\times}10^{-13}$m^2$/s $for actinides (U, Am), when tile dry densities were in the range of 1.2 Mg/㎥ to 1.8 Mg/㎥. The obtained results will be used in assessing the barrier properties of Kyungju bentonite as a buffer material of a repository in Korea.n Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.163-169
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• 2015

In many nuclear power plant sites in Korea, high density storage racks were installed in the spent fuel pool to expand the spent fuel storage capacity. Nevertheless, the capability of the Hanbit nuclear site will be saturated by 2024. Also, 10 NPPs will reach their design life expiration date by 2029. In the case of the US, SFPI (Spent Fuel Pool Island) operated temporarily as a spent fuel storage option before spent nuclear fuels were transported to an interim storage facility or a final disposal facility. As a spent fuel storage option after shutdown during decommissioning, the SFPI concept can be expected to have the following effects: reduced occupational exposure, lower cost of operation, strengthened safety, and so on. This paper presents a case study associated with the regulations, operating experiences, and systems of SFPI in the US. In conclusion, the following steps are recommended for applying SFPI during decommissioning in Korea: confirmation of design change scope of SFPI and expected final cost, the submission of a decommissioning plan which is reflected in SFPI improvement plans, safety assessment using PSR, application of an operating license change for design change, regulatory body review and approval, design change, inspection by the regulatory body, education and commissioning for SFPI, SFPI operation and periodic inspection, and dismantling of SFPI.