• 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.

• Economic and Environmental Geology
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• v.56 no.5
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• pp.515-532
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• 2023

Bentonite, predominantly consists of expandable clay minerals, is considered to be the suitable buffering material in high-level radioactive waste disposal repository due to its large swelling property and low permeability. Additionally, the bentonite has large cation exchange capacity and specific surface area, and thus, it effectively retards the transport of leaked radionuclides to surrounding environments. This study aims to review the thermodynamic sorption models for four radionuclides (U, Am, Se, and Eu) and eight bentonites. Then, the thermodynamic sorption models and optimized sorption parameters were precisely analyzed by considering the experimental conditions in previous study. Here, the optimized sorption parameters showed that thermodynamic sorption models were related to experimental conditions such as types and concentrations of radionuclides, ionic strength, major competing cation, temperature, solid-to-liquid ratio, carbonate species, and mineralogical properties of bentonite. These results implied that the thermodynamic sorption models suggested by the optimization at specific experimental conditions had large uncertainty for application to various environmental conditions.

• 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.

• Tunnel and Underground Space
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• v.30 no.5
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• pp.446-461
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• 2020

Numerical simulations of CIEMAT column test in Spain are performed to investigate the coupled thermo-hydro-mechanical (THM) behavior of MX80 bentonite pellets using TOUGH2-FLAC3D. The heater power and injection pressure of water in the numerical simulations are identical to those in the laboratory test. To investigate the applicability of the thermo-hydraulic (TH) model used in TOUGH2 code to prediction of the coupled TH behavior, the simulation results are compared with the observations of temperature and relative humidity with time. The tendencies of the coupled behavior observed in the test are well represented by the numerical models and the simulator in terms of temperature and relative humidity evolutions. Moreover, the performance of the models for the reproduction and prediction of the coupled TH behavior is globally satisfactory compared with the observations. However, the calculated stress change is relatively small and slow due to the limitations of the simple elastic and swelling pressure model used in numerical simulations. It seems that the two models are insufficient to realistically reproduce the complex coupled THM behavior in the bentonite pellets.

• Economic and Environmental Geology
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• v.51 no.2
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• pp.77-85
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• 2018

There would be a possibility of uranium contamination around the nuclear power plants and the underground waste disposal sites, where the uranium could further migrate and diffuse to some distant places by groundwater. It is necessary to understand the biogeochemical behaviors of uranium in underground environments to effectively control the migration and diffusion of uranium. In general, various kinds of microbes are living in soils and geological media where the activity of microbes may be closely connected with the redox reaction of nuclides resulting in the changes of their solubility. We investigated the adsorption and precipitation behaviors of dissolved uranium on some solid materials using hydrogen gas as an electron donor instead of organic matters. Although the effect of hydrogen gas did not appear in a batch experiment that used granite as a solid material, there occurred a reduction of uranium concentration by 5~8% due to hydrogen in an experiment using bentonite. This result indicates that some indigenous bacteria in the bentonite that have utilized hydrogen as the electron donor affected the behavior (reduction) of uranium. In addition, the bentonite bacteria have showed their strong tolerance against a given high temperature and radioactivity of a specific waste environment, suggesting that the nuclear-biogeochemical reaction may be one of main mechanisms if the natural bentonite is used as a buffer material for the disposal site in the future.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.382-393
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• 2020

The engineered barrier system of high-level radioactive waste disposal must maintain its performance in the long term, because it must play a role in slowing the rate of leakage to the surrounding rock mass even if a radionuclide leak occurs from the canister. In particular, it is very important to clarify gas dilation flow phenomenon clearly, that occurs only in a medium containing a large amount of clay material such as a bentonite buffer, which can affect the long-term performance of the bentonite buffer. Accordingly, DECOVALEX-2019 Task A was conducted to identify the hydraulic-mechanical mechanism for the dilation flow, and to develop and verify a new numerical analysis technique for quantitative evaluation of gas migration phenomena. In this study, based on the conventional two-phase flow and mechanical behavior with effective stresses in the porous medium, the hydraulic-mechanical model was developed considering the concept of damage to simulate the formation of micro-cracks and expansion of the medium and the corresponding change in the hydraulic properties. Model verification and validation were conducted through comparison with the results of 1D and 3D gas injection tests. As a result of the numerical analysis, it was possible to model the sudden increase in pore water pressure, stress, gas inflow and outflow rate due to the dilation flow induced by gas pressure, however, the influence of the hydraulic-mechanical interaction was underestimated. Nevertheless, this study can provide a preliminary model for the dilation flow and a basis for developing an advanced model. It is believed that it can be used not only for analyzing data from laboratory and field tests, but also for long-term performance evaluation of the high-level radioactive waste disposal system.

• Tunnel and Underground Space
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• v.29 no.4
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• pp.262-279
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• 2019

In the engineering barriers of high-level radioactive waste disposal, gases could be generated through a number of processes. If the gas production rate exceeds the gas diffusion rate, the pressure of the gas increases and gases could migrate through the bentonite buffer. Because people and the environment can be exposed to radioactivity, it is very important to clarify gas migration in terms of long-term integrity of the engineered barrier system. In particular, it is necessary to identify the hydro-mechanical mechanism for the dilation flow, which is a very important gas flow phenomenon only in medium containing large amounts of clay materials such as bentonite buffer, and to develop and validate new numerical approach for the quantitative evaluation of the gas migration phenomenon. Therefore, in this study, we developed a two-phase flow model considering the mechanical damage model in order to simulate the gas migration in the engineered barrier system, and validated with 1D gas flow modelling through saturated bentonite under constant volume boundary conditions. As a result of numerical analysis, the rapid increase in pore water pressure, stress, and gas outflow could be simulated when the dilation flow was occurred.

• Tunnel and Underground Space
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• v.32 no.6
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• pp.491-501
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• 2022

Bentonite has been proposed as a buffer and backfill material for high-level radioactive waste repository. Under such repository environment conditions, bentonite is subjected to combined thermal, hydrological, mechanical, and chemical processes. This study evaluates the feasibility of applying X-ray CT technology on the characterization of bentonite under hydration conditions using a newly developed testing cell. The cylindrical cell is made of platic material, with a removable cap to place the sample, enabling to apply vertical pressure on the sample and to measure swelling pressure. The hydration test was carried out with a sample made of Gyeonju bentonite, with a dry density of 1.4 g/cm³, and a water content of 20%. The sample had a diameter of 27.5 mm and a height of 34 mm. During the test, water was injected at a constant pressure of 0.207 MPa, and lasted for 7 days. After one day of hydration, bentonite swelled and filled out the space inside the cell. Moreover, CT histograms showed how the hydration process induced an initial increase and later progressive decrease on the density of the sample. Detailed profiles of the mean CT value, CT standard deviation, and CT gradient provided more details on the hydration process of the sample and showed how the bottom and top regions exhibited a decrease on density while the middle region showed an increase, especially during the first two days of hydration. Later, the differences in CT values with respect to the initial state decreased, and were small at the end of testing. The formation and later reduction of cracks was also characterized through CT scanning.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.270-275
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• 2003

The heat transfer analysis was conducted for the conceptual design of high level waste canisters. The temperature distribution due to the heat generation from four PWR spent fuel bundles which were contained in a canister located in a borehole 500 m below the surface was obtained. NISA computer program based upon FEM was used for the numerical solution. The temperature distribution in the composite system of $\ulcorner$canister ＋ buffer ＋ tunnel ＋ rock$\lrcorner$ due to heat generation from the spent fuel was obtained. In the case of 40m tunnel spacing and 6m borehole spacing the temperature showed the maximum value of $87.5^{\circ}C$around 15-16 years after disposal and decreased.

• 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.