• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.495-503
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• 2000

The hydraulic conductivities in the bentonite-sand mixtures with high density were measured, and the effects of sand content and dry density on the hydraulic conductivity were investigated. The hydraulic conductivities of the bentonite-sand mixtures with a dry density of 1.6 Mg/㎥ and 1.8 Mg/㎥ are less than 10$^{-11}$ m/s when the sand content is not higher than 70 wt%. However at the sand content of 90 wt%, the hydraulic conductivity increases rapidly At the same dry density, the logarithm of hydraulic conductivity increases linearly with increasing sand content. The hydraulic conductivity of the bentonite-sand mixture can be explained by the concept of effective clay dry density, and using this concept, the hydraulic conductivities for the mixtures with various sand contents and dry densities can be estimated.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.171-188
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• 2012

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called "A-KRS," in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been deterministically assessed with 5 various normal and abnormal scenarios associated with nuclide release and transport in and around the repository. Dose exposure rates to the farming exposure group have been evaluated in accordance with all the scenarios and then compared among other.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.263-272
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• 2012

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.276-276
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.334-334
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.290.1-290.1
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• 2003

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.187-187
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• 2009

Opening of fractures induced by shear dilation or normal deformation can be a significant source of fracture permeability change in fractured rock, which is important for the performance assessment of geological repositories for spent nuclear fuel. As the repository generates heat and later cools the fluid-carrying ability of the rocks becomes a dynamic variable during the lifespan of the repository. Heating causes expansion of the rock close to the repository and, at the same time, contraction close to the surface. During the cooling phase of the repository, the opposite takes place. Heating and cooling together with the, virgin stress can induce shear dilation of fractures and deformation zones and change the flow field around the repository. The objectives of this work are to examine the contribution of thermal stress to the shear slip of fracture in mid- and far-field around a KBS-3 type of repository and to investigate the effect of evolution of stress on the rock mass permeability. In the first part of this study, zones of fracture shear slip were examined by conducting a three-dimensional, thermo-mechanical analysis of a spent fuel repository model in the size of 2 km $\times$ 2 km $\times$ 800 m. Stress evolutions of importance for fracture shear slip are: (1) comparatively high horizontal compressive thermal stress at the repository level, (2) generation of vertical tensile thermal stress right above the repository, (3) horizontal tensile stress near the surface, which can induce tensile failure, and generation of shear stresses at the comers of the repository. In the second part of the study, fracture data from Forsmark, Sweden is used to establish fracture network models (DFN). Stress paths obtained from the thermo-mechanical analysis were used as boundary conditions in DFN-DEM (Discrete Element Method) analysis of six DFN models at the repository level. Increases of permeability up to a factor of four were observed during thermal loading history and shear dilation of fractures was not recovered after cooling of the repository. An understanding of the stress path and potential areas of slip induced shear dilation and related permeability changes during the lifetime of a repository for spent nuclear fuel is of utmost importance for analysing long-term safety. The result of this study will assist in identifying critical areas around a repository where fracture shear slip is likely to develop. The presentation also includes a brief introduction to the ongoing site investigation on two candidate sites for geological repository in Sweden.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.41-52
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• 2013

Adequate design of engineered barriers, including canister, buffer and backfill, is important for the safe disposal of high-level radioactive waste. Three-dimensional computer simulations were carried out under different condition to examine the thermal and mechanical behavior of engineered barriers and rock mass. The research looked at five areas of importance, the effect of the swelling pressure, water content of buffer, density of compacted bentonite, emplacement type and the selection of failure criteria. The results highlighted the need to consider tensile stress in the outer shell of a canister due to thermal expansion of the canister and the swelling pressure from the buffer for a more reliable design of an underground repository system. In addition, an adequate failure criterion should be used for the buffer and backfill.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.90-103
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• 2002

The thermal and mechanical properties of compacted bentonite and bentonite-sand mixture were collected from the literatures and compiled. The thermal conductivity of bentonite is found to increase almost linearly with increasing dry density and water content of the bentonite. The specific heat can also be expressed as a function of water ontent, and the coefficient of thermal expansion is almost independent on the dry density. The logarithm of unconfined compressive strength and Young’s modulus of elasticity increase linearly with increasing dry density, and in the case of constant dry density, it can be fitted to a second order polynomial of water content. Also the unconfined compressive strength and Young’s modulus of elasticity of the bentonite-sand mixture decreases with increasing sand content. The Poisson’s ratio remains constant at the dry density higher than 1.6 Mg/m$_3$, and the shear strength increases with increasing dry density.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.105-112
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• 2004

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.