• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.225-235
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• 2005

This paper outlined the status of the development of Korean Reference Disposal (KRS­1) system for high-level radioactive wastes. The repository concept was based on the engineering barrier system which KAERI has developed through a long-term research and development program. The design requirements were prepared for the conceptual design of the repository. The amount of PWR and CANDU spent fuels were projected with the current nuclear power plan. The disposal rates of PWR and CANDU spent fuels were analyzed. The reference geologic characteristics including classification of fracture zones were set for the KRS. The disposal concepts and the layout of the repository were described.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.281-294
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• 2012

The current status of the computer codes for the analysis of coupled thermal-hydrological-mechanical behavior occurred in a high-level waste repository was investigated. Based on the reported results on the comparison between the predictions using the computer codes and the experimental data from the in-situ tests, the reliability of the existing computer codes was analyzed. The presented codes simulated considerably well the coupled thermal-hydrological-mechanical behavior in the near-field rock of the repository without buffer, but the predictions for the engineered barrier system of the repository located at saturated hard rock were not satisfactory. To apply the current thermal-hydrological-mechanical models to the assessment of the performance of engineered barrier system, a major improvement on the mathematical models which analyze the distribution of water content and total pressure in the buffer is required.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.279-292
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• 2004

The underground research tunnel is essential to validate the integrity of a reference high-level waste disposal system, and the safety of geological disposal. In this study, a basic design of an underground research tunnel (URT) was tried to be developed. The candidate site for URT was described briefly, and it was intended to suggest the basic concept of the underground research tunnel. In order to develop the design of URT based on the basic concept, design requirements were established. Based on the basic concept and the design requirements, the basic design of URT was performed. Research items to be studied in the URT were also derived in this study.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.105-115
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• 2024

In this study, we employed a small-scale experiment to demonstrate the introduction of a thin copper heat dissipation plate into a bentonite buffer layer of an engineered barrier system. This experiment designed for spent nuclear fuel disposal can effectively reduce the maximum temperature of the bentonite buffer layer, and ultimately, make it possible to reduce the area of the disposal site. For the experiment, a small-scale engineered barrier system with a copper heat dissipation plate was designed and manufactured. the thickness of the cylindrical buffer was about 2 cm, which was about 1/20 of KAERI Repository System (KRS). At a power supply of 250 W, the maximum buffer temperature reduced to a mere 1.8℃ when the thin copper plate was introduced. However, the maximum surface temperature reduced to a remarkable 9.1℃, when a U-collar copper plate was introduced, which had a good contact with the other barrier layers. Consequently, we conclude that the introduction of the thin copper plate into the engineered barrier system for spent nuclear fuel disposal can effectively reduce the maximum buffer temperature in high-level radioactive waste disposal repositories.

• Analytical Science and Technology
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• v.17 no.3
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• pp.251-262
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• 2004

It is for the purpose of investigating the tritium distribution in the leachates, the raw and treated leachates and the condensates of the methane gas, which have occurred from domestic solid waste landfills. Also it aims to measure the tritium distribution level on the colloid size of the leachates, the raw and treated leachates. It was found that the major inorganic contaminants of the leachates were Na, K, Ca, Mg, $NH{_4}^+$-N and $Cl^-$. The mean tritium level of the raw leachates of the investigated 13 landfill sites for 6 months was 17 ~ 1196 TU. It corresponded to a several scores or hundreds of magnitude higher value than that of the normal environmental sample level except for two landfill sites. Also such a high concentration of the tritium was found in the treated leachates and methane gas condensates as well. Nevertheless it is important to emphasize that the tritium level which was found in this research is about 100 times lower than the tritium limit for the drinking water quality. And most of the tritium existed in the dissolved colloid of the leachate of which the colloid size is below $0.45{\mu}m$. Also, according to the tritium analysis results of the leachates after filtration with $0.45{\mu}m$ membrane filter for some landfills, it is likely that some tritium of the leachate would be distributed in a colloid size over $0.45{\mu}m$. In general the relationship between the tritium and other contaminants in the raw leachate was low, but it was relatively high between the tritium and TOC. However, the tritium content in the leachate had no meaningful relationship with the scale, hydrological characteristics and age of the landfill.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.245-251
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• 2004

Soil washing equipment was developed for decontamination of radioactively contaminated soil generated during normal operation or decommissioning and verification experiments were performed. Decontamination effciency above $80{\%}$ was achieved. In case of low radiation level and large particle size, decontamination efficiency was higher. According to the ratio of volume of water to soil quantity, decontamination efficiency was higher in case of high radiation level. Re-decontamination using decontaminated soil was effective in case of small particles. Using soil washing equipment, radioactivity of contaminated soil generated in nuclear power plant can be decreased and volume of soil for disposal can be decreased. And this equipment can be used in decommissioning.

• Journal of Korean Tunnelling and Underground Space Association
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• v.25 no.6
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• pp.509-523
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• 2023

The high-level nuclear waste (HLW) repository is exposed to complex environmental conditions consisting of high temperature, high humidity, and radiation, resulting in structural deterioration. Therefore, structural health monitoring is essential, and piezo sensors are used to detect cracks and estimate strength. However, since the monitoring sensors installed in the disposal tunnel and disposal container cannot be replaced or removed, the quantitative life of the monitoring sensor and its suitability must be assessed. In this study, the life of a piezo sensor for monitoring was assessed using an accelerated life test (ALT). The failure mode and mechanism of the piezo sensor under high temperature conditions were determined, and temperature stress's influence on the piezo sensor's life was analyzed. ALT was conducted on temperature stress and the relationship between temperature stress and piezo sensor life was suggested. The life of the piezo sensor was assessed using the Weibull probability distribution and the Arrhenius acceleration model. The suggested relationship can be used in multiple stress ALT designs for more precise life assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.447-457
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• 2019

Japan enacted the "Designated Radioactive Waste Final Disposal Act" for the geological disposal of high-level radioactive waste in 2000 and began the site selection process. However, no local government wanted to participate in the siting process. Therefore, in 2015, the Japanese government developed a new site selection process during the literature survey step, and on June 28, 2017 they published a "Nationwide Map of Scientific Features for Geological Disposal" created with the aim of promoting public participation from local governments. This map illustrated the requirements and criteria to be considered in the early or conceptual stages of securing a geological repository and was useful for improving public understanding and exchanging opinions with local governments by analyzing the suitability of different geological disposal sites.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.179-188
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• 2016

Spent fuels from nuclear power plants, as well as high-level radioactive waste from the recycling of spent fuels, should be safely isolated from human environment for an extremely long time. Recently, meaningful studies on the development of deep borehole radioactive waste disposal system in 3-5 km depth have been carried out in USA and some countries in Europe, due to great advance in deep borehole drilling technology. In this paper, domestic deep geoenvironmental characteristics are preliminarily investigated to analyze the applicability of deep borehole disposal technology in Korea. To do this, state-of-the art technologies in USA and some countries in Europe are reviewed, and geological and geothermal data from the deep boreholes for geothermal usage are analyzed. Based on the results on the crystalline rock depth, the geothermal gradient and the spent fuel types generated in Korea, a preliminary deep borehole concept including disposal canister and sealing system, is suggested.