• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.281-291
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• 2013

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.17-21
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• 2005

Shielded laser ablation system composed of laser system, image analyser, XYZ translator with motion controller, ablation chamber, manipulator and various optics was designed. Nd:YAG laser which can be tunable from 1064 nm to 266 m was selected as light source. CCD camera(< $\pm$200) was chosen to analyze a crater less than 50 un in diameter. XYZ translator was composed of three linear stage which can travel 50 w with a minimum movement of 1 um and motion controller. Before the performance test, each part of system was optically aligned. To perform the ablation test, the specimen was ablated by 50 um interval and observed by image analyser The shape of crater was almost round, indicating laser beam has homogeneous energy distribution. The resolution and magnification of image system were compatible with the design.

• Journal of the Korean Geotechnical Society
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• v.35 no.7
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• pp.41-50
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• 2019

Gyeongju bentonite is a buffer material primarily considered in Korea and it is highly compacted as a part of an engineered barrier system (EBS) of high-level radioactive waste repository. The compacted bentonite undergoes swelling stress by groundwater penetration and thermal stress by decay heat from a canister. Therefore, the mechanical properties of the compacted bentonite buffer material is crucial for the performance assessment of EBS. This paper aims to evaluate deformation properties of Gyeongju compacted bentonite using seismic methods. Two sets of compacted bentonite specimens were prepared having dry densities of $1.59g/cm^3$ and $1.75g/cm^3$ with water contents of 10.6% and 8.7%. Free-free resonant column tests were performed to measure constrained and unconstrained compression wave velocities. With the measured wave velocities, Young's modulus ($E_{max}$) and constrained modulus ($M_{max}$), material damping ratio ($D_{min}$), and Poisson's ratio at small strain were determined. As results, this paper evaluates the deformation properties of Gyeongju compacted bentonite and compares them with the results of previous researches.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.43-53
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• 2023

Sulfide concentrations critically affect worker safety and the integrities of underground facilities, such as deep geological repositories for spent nuclear fuel. Sulfide is highly sensitive to oxygen, which can oxidize sulfide to sulfate. This can hinder precise measurement of the sulfide concentration. Hence, a literature review was conducted, which revealed that two methods are commonly used: the methylene blue and sulfide ion-selective electrode (ISE) methods. Inductively coupled plasma optical emission spectroscopy (ICP-OES) was used for comparison with the two methods. The sulfide ISE method was found to be superior as it yielded results with a higher degree of accuracy and involved fewer procedures for quantification of the sulfide concentration in solution. ICP-OES results can be distorted significantly when sulfide is present in solution owing to the formation of H₂S gas in the ICP-OES nebulizer. Therefore, the ICP-OES must be used with caution when quantifying underground water to prevent any distortion in the measured results. The results also suggest important measures to avoid problems when using ICP-OES for site selection. Furthermore, the sulfide ISE method is useful in determining sulfide concentrations in the field to predict the lifetime of disposal canisters of spent nuclear fuel in deep geological repositories and other industries.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.149-164
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• 2023

As part of the safety case development for generic disposal sites in Korea, it is necessary to develop generic assessment models using various geosphere-biosphere interfaces (GBIs) and potentially exposed groups (PEGs) that reflect the natural environmental characteristics and the lifestyles of people in Korea. In this study, a unique modeling strategy was developed to systematically construct and select Korean generic biosphere assessment models. The strategy includes three process steps (combination, screening, and experts' scoring) for the biosphere system conditions. First, various conditions, such as climate, topography, GBIs, and PEGs, were combined in the biosphere system. Second, the combined calculation cases were configured into interrelation matrices to screen out some calculation cases that were highly unlikely or less significant in terms of the exposure dose. Finally, the selected calculation cases were prioritized based on expert judgment by scoring the knowledge, probability, and importance. The results of this study can be implemented in the development of biosphere assessment models for Korean generic sites. It is believed that this systematic methodology for selecting the candidate calculation cases can contribute to increasing the confidence of future site-specific biosphere assessment models.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.32-39
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• 2005

When PFC(Perfluorocarbonate) decontamination technology is applied to removal of radioactive contaminated particulate adhered at surface during the operation of nuclear research facilities, it is necessary to develop a filtration equipment to reuse of PFC solution due to high price, also to minimize the volume of second wastewater. Contaminated characteristics of hot particulate was investigated and a filtration process was presented to remove suspended radioactive particulate from PFC decontamination wastewater generated on PFC decontamination. The range of size of hot particulate adhered at the surface of research facilities measured by SEM was $0.1{\sim}10{\mu}m$. Hot particulate of more than $2{\mu}m$ in PFC contamination wastewater was removed by first filter and then hot particulate of more than $0.2{\mu}m$ was removed by second filter. Results of filter experiments showed that filtration efficiency of PVDF(Poly vinylidene fluoride), PP(Polypropylene), Ceramic filter was $95{\sim}97\%$. A ceramic filter showed a higher filtration efficiency with a little low permeate volume. Also, a ceramic of inorganic compound could be broken easily on experiment and has a high price but was highly stable at radioactivity in comparison of PVDF and PP of a macromolecule which generate $H_2$ gas in alpha radioactivity environment.

• Proceedings of the Korean Geotechical Society Conference
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• 2009.03a
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• pp.792-797
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• 2009

The stress-strain characteristics of compacted bentonite are investigated using experimental triaxial compression test by Hoek-cell. Special attention given to various dry density and water absorption ratio. Based on the test results, it is shown that the stress-strain relationship of compacted bentonite is highly influenced by dry density and water absorption ratio. Also, characteristics of Bentonite is similar to the clay rather than sand. Strength of compressed Bentonite increases with higher dry density. It shows maximum strength value, if in a same condition with dry density and constrain pressure. So we determine that value as the optimistic moisture contents for the maximun strength of compressed Bentonite.

• Geophysics and Geophysical Exploration
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• v.7 no.1
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• pp.70-79
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• 2004

We have examined the applicability of the triaxial drill-bit VSP method (TAD-VSP) to the geological survey of possible sites for a high-level radioactive waste disposal repository. The seismic energy generated by a drill bit is measured by a downhole multi-component detector, and the resulting signals are processed to image the geological structure deep underground. In order to apply the TAD-VSP method to civil-engineering-scale drilling, we have developed a small but highly sensitive and precise three-component downhole seismic measurement system, and recorded drill-bit signals at a granite quarry. We have successfully imaged discontinuities in the granite, possibly related to fractures, as highly reflective zones. The discontinuities imaged by the TAD-VSP method correlate well with the results of other borehole observations. In conclusion, the TAD-VSP method is usable in geological investigations for civil engineering because the equipment is compact and it is simple to acquire the drill-bit signal.

• Journal of Radiation Protection and Research
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• v.45 no.4
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• pp.178-186
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• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.