• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.133-141
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• 2020

A geological repository comprises a natural barrier and an engineered barrier system. Its design components consist of canisters, buffers, backfill, and near-field rock. Among the engineered barrier system components, bentonite buffers minimize the groundwater flow from near-field rock and prevent the release of nuclide. Investigation of the hydraulic conductivity of the buffer to groundwater flow is an important factor in the performance evaluation of the stability and integrity of the engineered barrier of the repository. In this study, saturated hydraulic conductivity tests were performed using Gyeongju bentonite at various dry densities and temperatures, and a hydraulic conductivity prediction model was developed through multiple regression analysis using the 120 result sets of hydraulic conductivity. The test results showed that the hydraulic conductivity tends to decrease as the dry density increases. In addition, the hydraulic conductivity increased with increasing temperature. The multiple regression analysis results showed that the coefficient of determination (R²) of the hydraulic conductivity prediction equation was as high as 0.93. The hydraulic conductivity prediction equation presented in this study could be used for the design of engineered barrier systems.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4173-4180
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• 2023

The stability of engineered barriers in high-level radioactive waste disposal systems can be influenced by the decay heat generated by the waste. This study focuses on the thermal analysis of various canister designs to effectively lower the maximum temperature of the engineered barrier. A numerical model was developed and employed to investigate the heat dissipation potential of copper rings placed across the buffer. Various canister designs incorporating copper rings were presented, and numerical analysis was performed to identify the design with the most significant temperature reduction effect. The results confirmed that the temperature of the buffer material was effectively lowered with an increase in the number of copper rings penetrating the buffer. Parametric studies were also conducted to analyze the impact of technical gaps, copper thickness, and collar height on the temperature reduction. The numerical model revealed that the presence of gaps between the components of the engineered barrier significantly increased the buffer temperature. Furthermore, the reduction in buffer temperature varied depending on the location of the gap and collar. The methods proposed in this study for reducing the buffer temperature hold promise for contributing to cost reduction in radioactive waste disposal.

• Journal of Mechanical Science and Technology
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• v.15 no.3
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• pp.327-338
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• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell, and lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and high swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear structural analysis. Canister types studied hear are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• 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 Korean Tunnelling and Underground Space Association
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• v.21 no.5
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• pp.587-609
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• 2019

Short-and long-term stabilities of bentonite, favored material as buffer in geological repositories for high-level waste were reviewed in this paper in addition to alternative design concepts of buffer to mitigate the thermal load from decay heat of SF (Spent Fuel) and further increase the disposal efficiency. It is generally reported that the irreversible changes in structure, hydraulic behavior, and swelling capacity are produced due to temperature increase and vapor flow between $150{\sim}250^{\circ}C$. Provided that the maximum temperature of bentonite is less than $150^{\circ}C$, however, the effects of temperature on the material, structural, and mineralogical stability seems to be minor. The maximum temperature in disposal system will constrain and determine the amount of waste to be disposed per unit area and be regarded as an important design parameter influencing the availability of disposal site. Thus, it is necessary to identify the effects of high temperature on the performance of buffer and allow for the thermal constraint greater than $100^{\circ}C$. In addition, the development of high-performance EBS (Engineered Barrier System) such as composite bentonite buffer mixed with graphite or silica and multi-layered buffer (i.e., highly thermal-conductive layer or insulating layer) should be taken into account to enhance the disposal efficiency in parallel with the development of multilayer repository. This will contribute to increase of reliability and securing the acceptance of the people with regard to a high-level waste disposal.

• Tunnel and Underground Space
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• v.31 no.4
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• pp.289-308
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• 2021

This study conducts coupled thermo-hydro-mechanical numerical modeling to investigate the maximum temperature and conditions for securing mechanical stability of the high-level radioactive waste repository when temperature criteria of bentonite buffer are 100℃ and 125℃, respectively. In case of temperature criterion of buffer as 100℃, the maximum temperatures at the interface between canister and buffer are calculated to be 99.4℃ and 99.8℃, respectively for a case with disposal tunnel spacing of 40 m and deposition hole spacing of 5.5 m and for the other case with disposal tunnel spacing of 30 m and deposition hole spacing of 6.5 m. In case of temperature criterion of buffer as 125℃, spacings of disposal tunnel and deposition hole could be decreased to 30 m and 4.5 m, respectively, which reduces the disposal area up to 55% compared to the disposal area of KRS⁺. According to analysis of mechanical stability for various disposal spacings, RMR of rock mass for KRS⁺ should be larger than 72.4 which belongs to good rock in RMR classification to prevent failure of rock mass. As disposal spacing is decreased, required RMR of rock mass is increased. In order to prevent failure of rock mass for a case with disposal tunnel spacing of 30 m and deposition hole spacing of 4.5 m, RMR larger than 87.3 is needed. However, mechanical stability of the repository is secured for all cases with RMR over 75 considering the enhancement of rock strength due to confining stress induced by swelling of the bentonite buffer and backfill.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.11-21
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• 2013

The Geyoungju Ca-bentonite with dry density of 1.6 g/$cm^3$ has been considered as a standard buffer material for the disposal of high level waste in KAERI disposal system design. But it had relatively lower thermal conductivity compared with other surrounding materials, that was one of key parameters to limit the increase of the disposal density in the disposal system. In this study, various additives were selected and mixed with the Ca-bentonite in different mixing methods in order to increase the thermal conductivity from 0.8 W/mK to 1.0 W/mK. As an additive, CNT (Cabon Nano Tube), graphite, alumina, CuO, and $Fe_2O_3$ were selected, which are chemically stable and have good thermal conductivity. As mixing methods, dry hand-mixer mixing, wet milling and dry ball mill mixing were applied for the mixing. Above all, the ball mill mixing was proved to be most effective since the produced mixture was most homogeneous and showed higher increase in the thermal conductivity. From this study, it was confirmed that the thermal conductivity for the Geyoungju Ca-bentonite could be improved by adding small amount of highly thermal conductive material to 1.0 W/mk. In conclusion, it was believed that the experimental results will be valuable in the disposal system design if the additive effects on the swelling and permeability on the compact bentonite are also approved in further studies.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.239-247
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• 2005

In this study, an experimental study on the sorption properties of uranium(VI) onto bentonite colloids generated from a domestic calcium bentonite (called as Gyeongju bentonite). Gyeongju bentonite has been considered as a potential candidate buffer material in the Korean disposal concept for high-level radioactive wastes. The size and concentration of the bentonite colloids used in the sorption experiment were measured by a filtration method. The result showed that the concentration of the synthesized bentonite colloid suspension was 5100ppm and the size of the most of bentonite colloids(over $98\%$) was in the range of 200-450nm in diameter. The amount of uranium lost by the sorption onto bottle walls, by precipitation, and by ultrafiltration or colloid formation was analyzed by carrying out some blank tests. The loss of uranium by the ultrafiltration was significant in the lower ionic strength(i.e., in the case of 0.001M $NaClO_4$) due to the cationic sorption effect onto the ultrafilter by a surface charge reversion. The distribution coefficient (or pseudo-colloid formation constant) for the sorption of uranium(VI) onto bentonite colloids was $10^4^{\sim}10^6$ mL/g depending upon pH and the distribution coefficient was highest in the neutral pH around 6.5.

• Tunnel and Underground Space
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• v.34 no.2
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• pp.104-126
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• 2024

Buffer and backfill play an essential role in isolating high-level radioactive waste and retard the migration of leaked radionuclides in deep geological disposal system. A bentonite mixture, which exhibits a swelling property, is considered for buffer and backfill materials, and excessive groundwater inflow from surrounding rock mass may affect stability and efficiency of their role as an engineered barrier. Therefore, stringent quality control as well as in-situ installation management and inflow water constrol for buffer and backfill are required to ensure the safety of deep disposal facilities. In this study, we analyzed the design requirements of buffer and backfill by examining various laboratory tests and a field study of the Steel Tunnel Test at the Äspö Hard Rock Laboratory in Sweden. We introduced how to control the quality of buffer and backfill construction in-field, and also presented how to handle excessive groundwater inflow into disposal caverns, validating the groundwater retention capacity of bentonite pellets and the effectiveness of geotexile use.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.83-100
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• 1999

As disposal packaging concepts of spent fuels generated from the domestic NPP, two types, one is to package PWR and CANDU spent fuels in different containers and the other is to package them together, were proposed. The configuration of the containers and the layout of underground repository, such as the container spacing and the deposition tunnel spacing, were developed. The layout of underground repository satisfies the thermal constraint of the bentonite buffer surrounding disposal container, which should be lower than $100^{\circ}C$ in order to keep the physical and chemical properties of bentonite From the spent fuel packaging concepts and container emplacement methods, seven options were developed. With a typical pair-wise comparison methods, AHP, the most promising disposal concept was selected based on the technology Point of view.