• 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.24 no.6
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• pp.405-417
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• 2014

In a high-level waste repository, the gap fill of the engineered barrier is an important component that influences the performance of the buffer and backfill. This paper reviewed the overseas status of R&D on the gap fill used engineered barriers, through which the concept of the gap fill, manufacturing techniques, pellet-molding characteristics, and emplacement techniques were summarized. The concept of a gap fill differs for each country depending on its disposal type and concept. Bentonite has been considered a major material of a gap fill, and clay as an inert filler. Gap fill was used in the form of pellets, granules, or a pellet-granule blend. Pellets are manufactured through one of the following techniques: static compaction, roller compression, or extrusion-cutting. Among these techniques, countries have focused on developing advanced technologies of roller compression and extrusion-cutting techniques for industrial pellet production. The dry density and integrity of the pellet are sensitive to water content, constituent material, manufacturing technique, and pellet size, and are less sensitive to the pressure applied during the manufacturing. For the emplacement of the gap fill, pouring, pouring and tamping, and pouring with vibration techniques were used in the buffer gap of the vertical deposition hole; blowing through the use of shotcrete technology and auger placement and compaction techniques have been used in the gap of horizontal deposition hole and tunnel. However, these emplacement techniques are still technically at the beginning stage, and thus additional research and development are expected to be needed.

• Journal of the Korean Geotechnical Society
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• v.35 no.11
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• pp.7-14
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• 2019

The compacted bentonite buffer is one of the major components of an engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW), and it is considered the best candidate for the buffer material. The buffer is located between disposal canisters and near-field rock mass, and it interrupts the release of radionuclide from disposal canisters and protect them from the penetration of groundwater. At initial disposal condition, degree of saturation of the compacted bentonite buffer decreases because of high thermal quantities released from the disposal canisters. However, the degree of saturation of the compacted bentonite buffer gradually increases caused by inflow of groundwater. The saturated and unsaturated behavior of the buffer is a very important input data since it can determine the safety performance of EBS. Therefore, this paper investigated water retention capacity (WRC) for the Korean compacted bentonite buffer. The WRC of the compacted bentonite buffer was derived by measuring volumetric water content and water suction when temperature variation was between 24℃~125℃ considering decrease of degree of saturation with respect to temperature increase. The WRC was also derived with the same volumetric water content under the room temperature condition, and it showed 1~15% larger water suction than high temperature condition.

• Tunnel and Underground Space
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• v.19 no.3
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• pp.203-212
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• 2009

The prediction of near field behavior around an underground high-level radioactive waste repository is important for the repository design as well as the safety assessment. In this study, a sensitivity analysis for seven parameters consisted of design parameters and material properties was carried out using a three-dimensional finite difference code. From the sensitivity analysis, it was found that the effects of borehole spacing, tunnel spacing, cooling time and rock thermal conductivity were more significant than the other parameters. For getting a statistical distribution of buffer and rock temperatures around the repository, an artificial neural network, backpropagation, was applied. The reliability of the trained neural network was tested with the cases with randomly chosen input parameters. When the parameter variation is within ${\pm}10%$, the prediction from the network was found to be reliable with about a 1% error. It was possible to calculate the temperature distribution for many cases quickly with the trained neural network. The buffer and rock temperatures showed a normal distribution with means of $98^{\circ}C$ and $83.9^{\circ}C$ standard deviations of $3.82^{\circ}C$ and $3.67^{\circ}C$, respectively. Using the neural network, it was also possible to estimate the required change in design parameters for reducing the buffer and rock temperatures for $1^{\circ}C$.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.219-228
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• 2012

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.381-390
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• 2017

In Korea, commercial nuclear power plants and research reactors have on-site storage systems for the spent nuclear fuel, but it is difficult to expand the facilities used for the storage systems. If decommissioning of nuclear power plants starts, an amount of high level radioactive waste will be generated. In this study, a radiological impact assessment of the railroad transport of high level radioactive waste was carried out considering radiation workers and the public, using the developed transport container as the transport package. The dose rates for workers and the public during the transport period were estimated, considering anticipated transport scenarios, and the results compared with the regulatory limit. A sensitivity analysis was also carried out by considering the different release ratios of the radioactive materials in the high level radioactive waste, and different distances between the transport container and workers during loading and unloading phases and while attaching another freight car. For all the anticipated transport scenarios, the radiological impacts for workers and the public met the regulatory limits.

• Tunnel and Underground Space
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• v.23 no.5
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• pp.372-382
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• 2013

Thermal-hydro-mechanical (THM) modeling is a critical R&D issue in the performance and safety assessment of a high-level waste repository. With an $\ddot{A}$sp$\ddot{o}$ prototype repository, its thermal behavior was analyzed and then compared with in-situ experimental data for its validation. A model simulation was used to calculate the temperature distributions in the deposition holes, deposition tunnel, and surrounding host rock. A comparison of the simulation results with the experimental data was made for deposition hole DH-6, which showed that there was a temperature difference of $2{\sim}5^{\circ}C$ depending on the location of the measuring points, but there was a similar trend in the evolution curves of temperature as a function of time. It was expected that the coupled modeling of the thermal behavior with the hydro-mechanical behavior in the buffer and backfill of the $\ddot{A}$sp$\ddot{o}$ prototype repository would give a better agreement between the experimental and model calculation results.

• Tunnel and Underground Space
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• v.21 no.6
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• pp.518-525
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• 2011

In this study, the natural ventilation pressure resulting from the large altitude difference which is a characteristic of high radioactive waste repository and the caloric value of the heat emitted by wastes was calculated and based on the results, natural ventilation quantities were calculated. A high radioactive waste repository can be considered as being operated through closed cycle thermodynamic processes similar to those of thermal engines. The heat produced by the heating of high radioactive wastes in the underground repository is added to the surrounding air, and the air goes up through the upcast vertical shaft due to the added heat while working on its surroundings. Part of the heat added by the work done by the air can be temporarily changed into mechanical energy to promote the air flow. Therefore, if a sustained and powerful heat source exists in the repository, the heat source will naturally enable continued cyclic flows of air. Based on this assumption, the quantity of natural ventilation made during the disposal of high radioactive wastes in a deep geological layer was mathematically calculated and based on the results, natural ventilation pressure of $74{\sim}183$Pa made by the stack effect was identified along with the resultant natural ventilation quantity of $92.5{\sim}147.7m^3/s$. The result of an analysis by CFD was $82{\sim}143m^3/s$ which was very similar to the results obtained by the mathematical method.

• Tunnel and Underground Space
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• v.22 no.6
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• pp.429-437
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• 2012

This study predicted temperature in the disposal tunnels using computational fluid dynamics based on natural ventilation quantity that comes from high altitude and temperature differences that are the characteristics of high level waste repository. The result of the previous study that evaluated quantitatively natural ventilation quantity using a hydrostatic method and CFD shows that significant natural ventilation quantity is generated. From the result, this study performed the prediction of temperature in disposal tunnels by natural ventilation quantity by the caloric values of the wastes, at both deep geological repository and surface repository. The result of analysis shows that deep geological repository is effective for thermal control in the disposal tunnels due to heat transfer to rock and the generation of sufficient natural ventilation quantity, while surface repository was detrimental to thermal control, because surface repository was strongly affected by external temperature, and could not generate sufficient natural ventilation quantity. Moreover, this study found that in the case of deep geological repository with a depth of 200 m, the heatof about $10^{\circ}C$ was transferred to the depth of 500 m. Thus, it is considered that if the high level waste repository scheduled to be built in the country is designed placing an emphasis on thermal control, deep geological repository rather than surface repository is more appropriate.