• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.168-171
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• 2003

The IMEF(Irradiated Materials Examination Facility), located in KAERI site, is a hot cell facility to test and evaluate the irradiation defects or embrittlement through post-irradiation examination(PIEs) of irradiated nuclear fuels and structural materials which are come from HANARO research reactor and commercial nuclear power plants. Therefore, to carry out its own function, the high level solid radioactive wastes, produced through PIEs, are periodically carried out and managed from hot cell to monolith. So far, approximately 30 drums which contains 50 liters are transported to monolith, and it is shown that the quantity is slowly increasing, In this paper, the procedures and work contents of the high level solid radwaste carrying out and management for IMEF are described in detail.

• The Journal of Engineering Geology
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• v.16 no.1 s.47
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• pp.69-83
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• 2006

This study aims to evaluate a complex groundwater flow system around the underground oil storage caverns using the concept of hydraulic compartment. For the hydrogeological analysis, the hydraulic testing data, the evolution of groundwater levels in 28 surface monitoring boreholes and pressure variation of 95 horizontal and 63 vertical water curtain holes in the caverns were utilized. At the cavern level, the Hydraulic Conductor Domains(fracture zones) are characterized one local major fracture zone(NE-1)and two local fracture zones between the FZ-1 and FZ-2 fracture zones. The Hydraulic Rock Domain(rock mass) is divided into four compartments by the above local fracture zones. Two Hydraulic Rock Domains(A, B) around the FZ-2 zone have a relatively high initial groundwater pressures up to $15kg/cm^2$ and the differences between the upper and lower groundwater levels, measured from the monitoring holes equipped with double completion, are in the range of 10 and 40 m throughout the construction stage, indicating relatively good hydraulic connection between the near surface and bedrock groundwater systems. On the other hand, two Hydraulic Rock Domains(C, D) adjacent to the FZ-1, the groundwater levels in the upper and lower zones are shown a great difference in the maximum of 120 m and the high water levels in the upper groundwater system were not varied during the construction stage. This might be resulted from the very low hydraulic conductivity$(7.2X10^{-10}m/sec)$ in the zone, six times lower than that of Domain C, D. Groundwater recharge rates obtained from the numerical modeling are 2% of the annual mean precipitation(1,356mm/year) for 20 years.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1511-1518
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• 2021

In most countries, the thermal criteria for the engineered barrier system (EBS) is set to below 100 ℃ due to the possible illitization in the buffer, which will likely be detrimental to the performance and safety of the repository. On the other hand, if the thermal criteria for the EBS increases, the disposal density and the cost-effectiveness for the high-level radioactive wastes will dramatically increase. Thus, fundamentals on the thermal behavior of the buffer under the elevated temperatures is of crucial importance. Yet, the behaviors at the elevated temperatures of the bentonite under groundwater-saturated conditions have not been reported to-date. Here, we have developed an in-situ synchrotron-based method for the thermal behavior study of the buffer under the elevated temperatures (25-250 ℃), investigated dspacings of the montmorillonite in the Korean bentonite (i.e., Ca-type) at dry and KURT (KAERI Underground Research Tunnel) groundwater-saturated conditions (KJ-ii-dry and KJ-ii-wet), and compared the behaviors with that of MX-80 (i.e., Na-type, MX-80-wet). The hydration states analyzed show tri-, bi-, and mono-hydrated at 25, 120, and 250 ℃, respectively for KJ-ii-wet, whereas tri-, mono-, and de-hydrated at 25, 150, and 250 ℃, respectively for MX-80-wet. The Korean bentonite starts losing the interlayered water at lower temperatures; however, holds them better at higher temperatures as compared with MX-80.

• Tunnel and Underground Space
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• v.10 no.2
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• pp.237-248
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• 2000

Based on the preliminary results from the therm analysis, which is currently carrying, three-dimensional computer simulations using a finite element code, ABAQUS Ver. 5.8, were designed to determine the mechanically stable cavern and deposition hole spacing. Linear elastic modeling for the cases with different cavern and deposition hole spacing were carried out under three different in situ stress conditions. From the simulations, the response of the rock to the stress redistribution after the excavation of the openings could be investigated. Also the optimum cavern and deposition hole spacing could be estimated based on the factor of safety. When the in situ stress determined from the actual stress measurements in Korea were used, the case with cavern spacing of 40m and deposition hole spacing of 3m was in very stable condition, because the factor of safety was calculated as 3.42., When the in situ stress conditions for Sweden and Canada were used, the previous case, they seem to be in stable condition, since the factors of safety are still higher than 1.0. From these results, it was concluded that the rock will not fail even after the stress redistribution.

• The Journal of Engineering Geology
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• v.33 no.2
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• pp.323-334
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• 2023

The vast volumes of data that are generated during site characterization and associated research for the disposal of high-level radioactive waste require effective data management to properly chronicle and archive this information. The Swedish Nuclear Fuel and Waste Management Company, SKB, established the SICADA database for site selection, evaluation, analysis, and modeling. The German Federal Company for Radioactive Waste Disposal, BGE, established ArbeitsDB, a database and document management system, and the ELO data system to manage data collected according to the Repository Site Selection Act. The U.K. Nuclear Waste Services established the Data Management System to manage any research and survey data pertaining to nuclear waste storage and disposal. The U.S. Department of Energy and Office of Civilian Radioactive Waste Management established the Technical Data Management System for data management and subsequent licensing procedures during site characterization surveys. The presented cases undertaken by these national agencies highlight the importance of data quality management and the scalability of data utilization to ensure effective data management. Korea should also pursue the establishment of both a data management concept for radioactive waste disposal that considers data quality management and scalability from a long-term perspective and an associated data management system.

• The Journal of Engineering Geology
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• v.33 no.2
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• pp.335-353
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• 2023

Site selection processes for high-level radioactive waste disposal facilities in different countries differ in terms of local geology and degree of public engagement. There seem to be three alternative processes for site selection: (1) selection with community consent after government choice; (2) selection with continuous community engagement after exclusion of unsuitable areas based on existing survey data; or (3) site selection where communities have expressed a willingness to participate. The Yucca Mountain site in Nevada, USA, was selected as the final disposal site by process (1) through six stages, but its development was suspended owing to opposition from the local governor and environmental groups. In Sweden, Switzerland, and Germany, process (2) is used and sites are selected through three stages. Sweden and Switzerland have completed site selection, and Germany is currently engaged in the process. The UK adopted process (3) with six stages, although the process has been suspended owing to poor community participation. In Korea, temporary storage facilities for spent nuclear fuel will reach saturation from 2030, so site selection must be promoted through various laws and systems, with continuous communication with local communities based on transparent and scientifically undertaken procedures.

• The Journal of Engineering Geology
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• v.13 no.2
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• pp.171-191
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• 2003

The objective of this present study is to understand long term(500 years) thermohydromechanical interaction behavior on joints adjacent to a repository cavern, when high level radioactive wastes are disposed of within discontinuous granitic rock masses, and then, to contribute this understanding to the development of a disposal concept. The model includes a saturated discontinuous granitic rock mass, PWR spent nuclear fuels in a disposal canister surrounded with compacted bentonite inside a deposition hole, and mixed bentonite backfilled in the rest of the space within a repository cavern. It is assumed that two joint sets exist within a model. Joint set 1 includes joints of $56^{\circ}$ dip angle, spaced 20m apart, and joint set 2 is in the perpendicular direction to joint set 1 and includes joints of $34^{\circ}$ dip angle, spaced 20m apart. The two dimensional distinct element code, UDEC is used for the analysis. To understand the joint behavior adjacent to the repository cavern, Barton-Bandis joint model is used. Effect of the decay heat from PWR spent fuels on the repository model has been analyzed, and a steady state flow algorithm is used for the hydraulic analysis.

• Geomechanics and Engineering
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• v.27 no.6
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• pp.583-594
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• 2021

The gas permeability behavior of unsaturated bentonite-based materials is of major importance for ensuring effective sealing of high-level radwaste repositories. This study investigated this by taking a sample of Granular Bentonite Material (GBM) at the end of the Engineered Barrier Emplacement (EB) experiment in the Opalinus Clay, placing it under different humidity conditions until it achieved equilibration, and testing the change in the gas permeability under loading and unloading. Environmental humidity is shown to have a significant effect on the water content, saturation, porosity and dry density of GBM and to affect its gas permeability. Higher sensitivity to confining pressure is exhibited by samples equilibrated at higher relative humidity (RH). It should be noted that for the sample at RH=98%, when the confining pressure is raised from 1 MPa to 6 MPa, gas permeability can be reduced from 10^-16 m² to 10^-19 m², which is close to the requirements of gas tightness. Due to higher water content and easier compressibility, samples equilibrated under higher RH show greater irreversibility during the loading and unloading process. The effective gas permeability of highly saturated samples can be increased by 2-3 orders of magnitude after 105℃ drying. In addition, cracks possibly occurred during the dehydration and drying process will become the main channel for gas migration, which will greatly affect the sealing performance of GBM.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.123-136
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• 2018

The objective of this study is to propose a new methodology to fabricate a reliable engineering-scale buffer block, which shows homogeneous and uniform distribution in buffer block density, for in-situ experiments. In this study, for the first time in Korea, floating die press and CIP (Cold Isostatic Press) are applied for the manufacture of an engineering-scale bentonite buffer. The optimized condition and field applicability are also evaluated with respect to the method of manufacturing the buffer blocks. It is found that the standard deviation of the densities obtained decreases noticeably and that the average dry density increases slightly. In addition, buffer size is reduced by about 5% at the same time. Through the test production, it is indicated that the stress release phenomenon decreases after the application of the CIP method, which leads to a reduction in crack generation on the surface of the buffer blocks over time. Therefore, it is confirmed that the production of homogeneous buffer blocks on industrial scale is possible using the method suggested in this study, and that the produced blocks also meet the design conditions for dry density of buffer blocks in the AKRS (Advanced Korea Reference Disposal System of HLW).