• Journal of the Korean Geotechnical Society
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• v.35 no.7
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• pp.41-50
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• 2019

Gyeongju bentonite is a buffer material primarily considered in Korea and it is highly compacted as a part of an engineered barrier system (EBS) of high-level radioactive waste repository. The compacted bentonite undergoes swelling stress by groundwater penetration and thermal stress by decay heat from a canister. Therefore, the mechanical properties of the compacted bentonite buffer material is crucial for the performance assessment of EBS. This paper aims to evaluate deformation properties of Gyeongju compacted bentonite using seismic methods. Two sets of compacted bentonite specimens were prepared having dry densities of $1.59g/cm^3$ and $1.75g/cm^3$ with water contents of 10.6% and 8.7%. Free-free resonant column tests were performed to measure constrained and unconstrained compression wave velocities. With the measured wave velocities, Young's modulus ($E_{max}$) and constrained modulus ($M_{max}$), material damping ratio ($D_{min}$), and Poisson's ratio at small strain were determined. As results, this paper evaluates the deformation properties of Gyeongju compacted bentonite and compares them with the results of previous researches.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.23-32
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• 2011

An engineering scale test, which is called KENTEX, was carried out to understand and to analyze the coupled thermal, hydrological and mechanical phenomena in the engineered barrier system(EBS) of Korean reference disposal system. Using the experimental data obtained from KENTEX, the water saturation processes in bentonite could be analyzed. From the comparison between the model calculation using ABAQUS and the experimental results, the difference of the water content between them in the unsaturating part was large because the drying phenomena due to moisture redistribution by the temperature gradient could not be included in the model. In the saturating part, the difference of the water content between them was decreased gradually and showed to be small in the full saturation. And the time of about 95% saturation could be estimated about 500 days from the model calculation and experimental results. Also it could be known that the moisture redistribution in the unsaturated part could not be affected on the saturation time of bentonite in the repository. Therefore, it is considered that this model could be used to quantitatively predict the water saturation time in bentonite as EBS for the disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.107-119
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• 2014

The worldwide Status-of-Art survey for the in-situ experiments of the engineered barrier system for HLW underground disposal was performed as a preliminary action for the design of the in-situ demonstration at KURT. Some nations, which have executed or is ongoing the in-situ experiments at their underground research facilities, were summarized in this review. The demonstration projects reviewed were TBT/Sweden-France, LOT/Sweden, HE-E/Switzerland, PRACLAY/Belgium, FEBEX/Spain, HORONOBE/Japan, and BCE/Canada. The investigated items for the projects were mainly their purposes, constitutional structures, test conditions, monitoring parameters and the measuring tools, and test results. In this review, the hardware design and the assembling of the test system were more concentrated rather than their experimental results, because the purpose of this review is to achieve the necessary information for the practical design of the in-situ experiment to be installed at KURT. A mid scale in-situ demonstration of EBS at KURT, that is called IN-DBES, will be launched right after the completion the expanding project of KURT in 2015. It is hoped that the structural design, installing methods, hardware equipments required in the establishment of IN-DEBS will be referred on this review.

• 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%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.261-270
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• 2018

In this study, to replace the 'J-slot joint', a joint device between a disposal canister and an emplacement jig in Deep Borehole Disposal process, a novel joint device was designed and tested. The novel joint device was composed of a wedge on top of a disposal canister and a hook box at the end of a winch system. The designed joint device had merits in that it can recombine an emplaced canister freely without the replacement of the joint component. Moreover, it can be applied to various emplacement jigs such as drill pipes, wire-lines, and coiled tubing. To demonstrate the designed joint device, the joint device (${\Phi}110mm$, H 148 mm), a twin canister string (${\Phi}140mm$, H 1,105 mm), and a water tube (${\Phi}150mm$, H 1,500 mm) as a borehole model were manufactured at 1/3 scale. As deployment muds, Na-type bentonite (MX-80) and Ca-type (GJ II) bentonite muds were prepared at solid contents of 7wt% and 28wt%, respectively. The manufactured joint device showed good performance in pure water and viscous muds, with an operation speed of $10m{\cdot}min^{-1}$. It was concluded that the newly developed joint device can be used for the emplacement and retrieval of a deep disposal canister, below 3~5 km, in the future.

• Analytical Science and Technology
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• v.23 no.6
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• pp.566-571
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• 2010

This paper describes the sorption behaviors of aqueous uranium ions on the K-birnessite. K-birnessite was synthesized by adding a concentrated HCl to an aqueous solution of $KMnO_4$. Physicochemical characteristics of the K-birnessite, such as structure, specific surface area and surface charge, were investigated. K-birnessite is a layered material and the $K^+$ ions exist in the interlayer of layered K-birnessite. BET specific surface area of the K-birnessite was 38.30 m2/g. The surface charge of K-birnessite was $-1.65\;C/m^2$ at pH 5.00 and ionic strength of 0.010 M $NaClO_4$, at which the sorption experiments of uranium ions were carried out. Uranium ions were incorporated into the interlayer of the K-birnessite by cation-exchange reaction with $K^+$ ions, and the distribution coefficient is quite similar to those of common ion-exchange materials. The results might be applicable in the retardation of migration of radioactive materials from the underground disposal site of high-level radioactive waste.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.169-179
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• 2011

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over $100^{\circ}C$ were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

• Economic and Environmental Geology
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• v.40 no.4
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• pp.361-374
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• 2007

Rare earth elements(REEs) have been used as an useful tool in understanding the various geological processes such as evolution and differentiation in the crust. The REEs also have been used as an analog of actinides for radioactive wastes at the water-rock interactions. Using physicochemical properties of the REEs and actinides, we have shown that Eu is an optimum analogue for understanding the behavior of Am in subsurface environments. Factors affecting sorption behavior of radioactive nuclides in groundwater were investigated by batch experiments. Four nuclides such as $^{241}Am,\;^{152}Eu,\;^{160}Tb\;and\;^{60}Co$ were selected to test our hypothesis, and $^{160}Tb$ and $^{60}Co$ were specifically used to compare to the sorption behavior between $^{241}Am-^{152}Eu$ and other radioactive nuclides. Four different rock samples and one groundwater were used in the batch experiments where solution pH for all experiments was fixed at 5.5. Our results demonstrate that $^{241}Am,\;^{152}Eu,\;and\;^{160}Tb$ show similar sorption behavior whereas $^{60}Co$ is different in sorption behavior at the mineral-water interface, suggesting that the sorption behavior of $^{60}Co$ is affected by different rock types. Our results also show that 1) Eu in REEs is optimum analogue of fate and transport of Am in subsurface environments, and 2) mineral compositions such as $SiO_2,\;TiO_2,\;P_2O_5$ and distribution of REEs such as Eu anomaly play key roles in affecting sorption behavior of radioactive nuclides even though physicochemical properties of geological materials such as specific surface area and cation exchange capacity can not be ruled out.

• Economic and Environmental Geology
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• v.56 no.4
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• pp.421-433
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• 2023

The final disposal of spent nuclear fuel(SNF) from nuclear power plants takes place in a deep geological repository. The metal canister encasing the SNF is made of cast iron and copper, and is engineered to effectively isolate radioactive isotopes for a long period of time. The SNF is further shielded by a multi-barrier disposal system comprising both engineering and natural barriers. The deep disposal environment gradually changes to an anaerobic reducing environment. In this environment, sulfide is one of the most probable substances to induce corrosion of copper canister. Stress-corrosion cracking(SCC) triggered by sulfide can carry substantial implications for the integrity of the copper canister, potentially posing a significant threat to the long-term safety of the deep disposal repository. Sulfate can exist in various forms within the deep disposal environment or be introduced from the geosphere. Sulfate has the potential to be transformed into sulfide by sulfate-reducing bacteria(SRB), and this converted sulfide can contribute to the corrosion of the copper canister. Bentonite, which is considered as a potential material for buffering and backfilling, contains oxidized sulfate minerals such as gypsum(CaSO4). If there is sufficient space for microorganisms to thrive in the deep disposal environment and if electron donors such as organic carbon are adequately supplied, sulfate can be converted to sulfide through microbial activity. However, the majority of the sulfides generated in the deep disposal system or introduced from the geosphere will be intercepted by the buffer, with only a small amount reaching the metal canister. Pyrite, one of the potential sulfide minerals present in the deep disposal environment, can generate sulfates during the dissolution process, thereby contributing to the corrosion of the copper canister. However, the quantity of oxidation byproducts from pyrite is anticipated to be minimal due to its extremely low solubility. Moreover, the migration of these oxidized byproducts to the metal canister will be restricted by the low hydraulic conductivity of saturated bentonite. We have comprehensively analyzed and summarized key research cases related to the presence of sulfates, reduction processes, and the formation and behavior characteristics of sulfides and pyrite in the deep disposal environment. Our objective was to gain an understanding of the impact of sulfates and sulfides on the long-term safety of high-level radioactive waste disposal repository.

• Economic and Environmental Geology
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• v.50 no.2
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• pp.117-128
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• 2017

Occurrence characteristics and existing forms of U-Th containing minerals in KURT (KAERI Underground Research Tunnel) granite are investigated to understand long-term behavior of radionuclides in granite considered as a candidate rock for the geological disposal of high-level radioactive waste. KURT granite primarily consists of quartz, feldspar and mica. zircon, REE(Rare Earth Element)-containing monazite and bastnaesite are also identified. Besides, secondary minerals such as sericite, microcline and chlorite including quartz vein and calcite vein are observed. These minerals are presumed to be accompanied by a post-hydrothermal process. U-Th containing minerals are mainly observed at the boundaries of quartz, feldspar and mica, mostly less than $30{\mu}m$ in size. Quantitative analysis results using EPMA (Electron Probe Micro-Analyzer) show that 74.2 ~ 96.5% of the U-Th containing minerals consist of $UO_2$ (3.39 ~ 33.19 wt.%), $ThO_2$ (41.61 ~ 50.24 wt.%) and $SiO_2$ (15.43 ~ 18.60 wt.%). Chemical structure of the minerals calculated using EPMA quantitative analysis shows that the U-Th minerals are silicate minerals determined as thorite and uranothorite. The U-Th containing silicate minerals are formed by a magmatic and hydrothermal process. Therefore, KURT granite formed by a magmatic differentiation is accompanied by an alteration and replacement owing to a hydrothermal process. U-Th containing silicate minerals in KURT granite are estimated to be recrystallized by geochemical factors and parameters such as temperature, pressure and pH owing to the hydrothermal process. By repeated dissolution/precipitation during the recrystallization process, U-Th containing silicate minerals such as thorite and uranothorite are formed according to the variation in the concentrated amount of U and Th.