• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.393-400
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• 2006

In design of a deep geological repository for the high level wastes, it is very important that the temperature of the bentonite block should not be over $100^{\circ}C$ to maintain the integrity of the bentonite buffer block from the decay heat. In this study, for the layout of the repository to meet the requirement, the analysis of the disposal tunnel and disposal pit spacing was carried out. To do this, based on the reference repository concept, several cases of cooling times and disposal tunnel and disposal pit spacing were compared. The thermal stabilities of the disposal systems were analyzed in terms of the cooling time and spacing. The results showed that it was more desirable to determine the layout of the repository in terms of disposal pit spacing than the disposal tunnel spacing. The results of these analyses can be used in the deep geological repository design. The detailed analyses with the exact site characteristics data will reduce the uncertainty of the results.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.107-119
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• 2014

The worldwide Status-of-Art survey for the in-situ experiments of the engineered barrier system for HLW underground disposal was performed as a preliminary action for the design of the in-situ demonstration at KURT. Some nations, which have executed or is ongoing the in-situ experiments at their underground research facilities, were summarized in this review. The demonstration projects reviewed were TBT/Sweden-France, LOT/Sweden, HE-E/Switzerland, PRACLAY/Belgium, FEBEX/Spain, HORONOBE/Japan, and BCE/Canada. The investigated items for the projects were mainly their purposes, constitutional structures, test conditions, monitoring parameters and the measuring tools, and test results. In this review, the hardware design and the assembling of the test system were more concentrated rather than their experimental results, because the purpose of this review is to achieve the necessary information for the practical design of the in-situ experiment to be installed at KURT. A mid scale in-situ demonstration of EBS at KURT, that is called IN-DBES, will be launched right after the completion the expanding project of KURT in 2015. It is hoped that the structural design, installing methods, hardware equipments required in the establishment of IN-DEBS will be referred on this review.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.247-260
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• 2010

The developement of a HLW disposal canister is under way in KAERI using Cold Spray Coating technique. To estimate corrosion behavior of a cold sprayed copper, a creivice corrosion test was conducted at Southwest Research Institute(SWRI) in the United State. For the measurement of repassivation potential needed for crevice corrosion, three methods such as (1) ASTM G61-86 : Cyclic Potentiodynamic Polarization Measurements, (2) Potentiodynamic Polarization plus intermediate Potentiostatic Hold method, and (3) ASTM G192-08 (THE method) : Potentiodynamic- Galvanostatic-Potentiostatic Method, were introduced in this report. In the crevice corrosion test, the occurrence of corrosion at crevice area was optically determined and the repassivation potentials were checked for three kind of copper specimens in a simulated KURT underground water, using a crevice former dictated in ASTM G61-86. The applied electrochemical test techniques were cyclic polarization, potentiostatic polarization, and electrochemical impedance spectroscopy. As a result of crevice corrosion tests, every copper specimens including cold sprayed one did not show any corrosion figure on crevice areas. And the open-cell voltage, at which corrosion reaction initiates, was influenced by the purity of copper, but not their manufacturing method in this experiment. Therefore, it was convinced that there is no crevice corrosion for the cold sprayed copper in KURT underground environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.169-179
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• 2011

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over $100^{\circ}C$ were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.311-318
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• 2005

Transparency on the Total System Performance Assessment (TSPA) is the key issue to enhance the public acceptance for a permanent high level radioactive repository. To approve it, all performances on TSPA through Quality Assurance is necessary. The integrated Cyber R&D Platform is developed by KAERI using the T2R3 principles applicable for five major steps in R&D's. The proposed system is implemented in the web-based system so that all participants in TSPA are able to access the system. It is composed of FEAS (FEp to Assessment through Scenario development) showing systematic approach from the FEPs to Assessment methods flow chart, PAID (Performance Assessment Input Databases) showing PA(Performance Assessment) input data set in web based system and QA system receding those data. All information is integrated into Cyber R&D Platform so that every data in the system can be checked whenever necessary. For more user-friendly system, system upgrade included input data & documentation package is under development. Throughout the next phase R&D, Cyber R&D Platform will be connected with the assessment tool for TSPA so that it will be expected to search the whole information in one unified system.

• Tunnel and Underground Space
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• v.33 no.1
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• pp.42-56
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• 2023

In deep geological disposal of high-level radioactive waste (HLW), the natural barrier must physically support the disposal facility and delay the movement of radionuclides for at least hundreds of thousands of years. To evaluate the long-term geological evolution of the natural barriers, it is essential to analyze the long-term behavior of rock joints, including the frictional healing behavior. This study aimed to experimentally analyze the frictional healing behavior of rock joints under hydro-mechanical (H-M) conditions through the slide-hold-slide (SHS) test. The SHS tests were performed under mechanical and H-M conditions for joint specimens of different roughness. In the H-M conditions, the frictional healing rate tended to increase, which was more evident in the specimens with large roughness. In addition, it was confirmed that the effect of the hydro-mechanical conditions was more significant when the effective normal stress acting on the joint surface was small. These results are expected to be used as fundamental data to understand the frictional healing behavior of rock joints in the natural barriers.