• 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.

• Journal of the Korean Geotechnical Society
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• v.40 no.2
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• pp.29-40
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• 2024

This study aimed to evaluate the swelling behavior characteristics of KJ-II buffer blocks by performing numerical analysis of swelling pressure measurement experiments using the nonlinear elasticity model of COMSOL Multiphysics. The analysis was conducted under boundary conditions that included isotropic constraints and water injection pressure, mirroring the experimental settings. Validation of the numerical model was achieved by comparing its outputs with experimental results. The validated model was then used to simulate swelling deformations under unconfined conditions and to analyze swelling pressure as influenced by dry density and the geometric shape of the buffer material. The results accurately represented the swelling deformation observed during the saturation process and demonstrated that swelling pressure increases with higher dry density. Moreover, simulations concerning the geometric shape of the buffer material indicated a markedly faster rate of pressure increase in U-shaped samples compared to cylindrical ones. Analysis suggested that stress manifested preemptively near the internal edges of U-shaped samples during saturation. To enhance the simulation's fidelity to actual buffer material behavior, further refinement of the analysis model using a nonlinear elasticity model is recommended.

• Geomechanics and Engineering
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• v.34 no.4
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• pp.449-459
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• 2023

The design and development of underground nuclear waste repositories should cover the performance evaluation of the different components such as the construction materials because the long term stability will depend on their response to the surrounding conditions. In South Korea, Gyeonju bentonite has been proposed as a candidate to be used as buffer and backfilling material, especially in the form of blocks to speed up the construction process. In this study, various cylindrical samples were prepared with different dry density and water content, and their physical and mechanical properties were analyzed and correlated with X-ray CT observations. The main objective was to characterize the samples and establish correlations for non-destructive estimation of physical and mechanical properties through the utilization of X-ray CT images. The results showed that the Uniaxial Compression Strength and the P-wave velocity have an increasing relationship with the dry density. Also, a higher water content increased the values of the measure parameters, especially for the P-wave velocity. The X-ray CT analysis indicated a clear relation between the mean CT value and the dry density, Uniaxial Compression Strength, and P-wave velocity. The effect of the higher water content was also captured by the mean CT value. Also, the relationship between the mean CT value and the dry density was used to plot CT dry densities using CT images only. Moreover, the histograms also provided information about the samples heterogeneity through the histograms' full width at half maximum values. Finally, the particle size and heterogeneity were also analyzed using the Madogram function. This function identified small particles in uniform samples and large particles in some samples as a result of poor mixing during preparation. Also, the μ_max value correlated with the heterogeneity, and higher values represented samples with larger ranges of CT values or particle densities. These image-based tools have been shown to be useful on the non-destructive characterization of bentonite samples, and the establishment of correlations to obtain physical and mechanical parameters solely from CT images.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.941-948
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• 2024

The effects of radiation on the corrosion of canister materials were investigated for the reliable disposal of high-level radioactive waste. The test specimens were gamma-irradiated at a very low dose rate of approximately 0.1 Gy/h for six and twelve months. The copper and cast iron species were less corroded when irradiated. It is hypothesized that gamma rays suppress the formation of lower-enthalpy species like metal oxides and activate reductive reactions. In contrast, it was difficult to evaluate the effect of radiation on the corrosion of titanium and stainless steel.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.14 no.4
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• pp.515-523
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• 2001

In this paper results of a linear structural analysis for design and dimensioning of disposal containers for spent pressurized water reactor nuclear fuel and spent Canadian deuterium and uranium reactor nuclear fuel are presented. The container structure studied here is a solid structure with a cast insert and a corrosion resistant outer shell, which is designed for the spent nuclear fuel disposal in a deep repository. An evenly distributed load of hydrostatic pressure from the groundwater and large swelling pressure from the bentonite buffer are applied on the container. Hence, the container must be designed to endure these large pressure loads. In this study, the array type of inner baskets and thicknesses of outer shell and lid/bottom are attempted to be determined through a linear static structural analysis.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.20 no.2
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• pp.157-164
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• 2007

The prediction of the temperature distribution change of the spent nuclear fuel disposal canister and its surrounding structures (bentonite buffer, granitic rock etc.) due to the spent fuel heat is very important for the design of the 500m deep granitic repository for the spent nuclear fuel disposal canister (about 10,000 years long) deposition. In this study, the temperature distribution change of the composite structure which comprises the canister, the bentonite buffer, the deposition tunnel due to the spent fuel heat is computed using the numerical analysis method. Specially, the temperature distribution change of the composite structure is analysed as the deposition time elapses up to m years. The analysis result shows that the temperature of each part of the repository increases slowly in different way but the latest part temperature increases slowly up to 150 years and thereafter decreases slowly.