• Tunnel and Underground Space
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• v.19 no.6
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• pp.507-516
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• 2009

In cases of high-level radioactive waste repositories, heat load is apparent by radioactive waste decay. The safety of a waste repository would be influenced by changing circumstances caused by heat transfer through rock. Thus, a ventilation system is necessary to secure the waste repository. The first priority for building an appropriate ventilation system is completing a computer simulation research with thermal rock properties and a heat transfer coefficient. In this study, the heat transfer coefficient in KURT was calculated using the measurement of inner circumstance factors that include dry bulb and wet bulb temperature, rock surface temperature, and barometric pressure. The heater that is 2 m in length and 5 kw in capacity heats the inside of rock in the research module by $90^{\circ}C$. As a result of determining the heat transfer coefficient in the heating section, the changes of heat transfer coefficient were found to be a maximum of 7.9%. The average heat transfer coefficient is approximately 4.533 w/$m^2{\cdot}K$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.279-296
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• 2022

The fundamental characteristics of groundwater colloids, such as composition, concentration, size, and stability, were analyzed using granitic groundwater samples taken from the KAERI Underground Research Tunnel (KURT) site by such analytical methods as inductively coupled plasma-mass spectrometry, field emission-transmission electron microscopy, a liquid chromatography-organic carbon detector, and dynamic light scattering technique. The results show that the KURT groundwater colloids are mainly composed of clay minerals, calcite, metal (Fe) oxide, and organic matter. The size and concentration of the groundwater colloids were 10-250 nm and 33-64 ㎍·L^-1, respectively. These values are similar to those from other studies performed in granitic groundwater. The groundwater colloids were found to be moderately stable under the groundwater conditions of the KURT site. Consequently, the groundwater colloids in the fractured granite system of the KURT site can form stable radiocolloids and increase the mobility of radionuclides if they associate with radionuclides released from a radioactive waste repository. The results provide basic data for evaluating the effects of groundwater colloids on radionuclide migration in fractured granite rock, which is necessary for the safety assessment of a high-level radioactive waste repository.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.432-443
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• 1999

Using a two-dimensional numerical model, nuclide transport in the buffer between the canister and adjacent rock in a high-level radioactive waste repository is dealt with. Calculations are made for a typical case with a three-member decay chain, $^{234}$ U longrightarrow $^{230}$ Th longrightarrow $^{226}$ Ra. The solution method used here is based on a physically exact formulation by a control volume method directly integrating the governing equation over each control volume.

• Journal of the Korean Geotechnical Society
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• v.33 no.10
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• pp.25-31
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• 2017

The buffer is one of the major components of an engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW). As the buffer is located between a disposal canister and host rock, it is indispensable to assure the disposal safety of high-level radioactive waste. It can restrain the release of radionuclide and protect the canister from the inflow of groundwater. Since high quantity of heat from a disposal canister is released to the surrounding buffer, thermal properties of the buffer are very important parameters for the analysis of the entire disposal safety. Especially, temperature criteria of the compacted bentonite buffer can affect the design of HLW repository facility. Therefore, this paper investigated thermal properties for the Kyungju compacted bentonite buffer which is the only bentonite produced in South Korea. Hot wire method and dual probe method were used to measure thermal conductivity and specific heat capacity of the compacted bentonite buffer according to the temperature variation. Thermal conductivity and specific heat capacity were decreased dramatically when temperature variation was between $22^{\circ}C{\sim}110^{\circ}C$ as degree of saturation decreased according to the temperature variation. However, there was little variation under the high temperature condition at $110^{\circ}C{\sim}150^{\circ}C$.

• 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.2 no.4
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• pp.253-262
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• 2004

Construction of tunnels in a deep crystalline host rock for a potential High-Level Radioactive Waste(HLW) repository inevitably generates an excavation disturbed zone (EDZ). There have been a series of debates on whether a permeability in an EDZ increases or not and what would be the maximum depth of an EDZ. Recent studies show mixed opinions on permeability. However, there has been an international consensus on the thickness of an EDZ; 30 cm for TBM and 1 meter for controlled blast. One of the impacts of an EDZ is on determining the distance between adjacent deposition holes. The void gap by the excavation hinders relaxation of temperature profiles so that the current Korean reference designing distance between holes should be stretched out more to keep the maximum temperature in a buffer region below 100 degrees Celsius. The other impact of an EDZ is on the long-term post closure radiological safety. To estimate the impact, the reference scenario, the well scenario, is chosen. Released nuclides diffuse through a bentonite buffer region experiencing strong sorption and reach a fracture surrounded by a porous medium. Inside a fractured porous region, radionuclides migrate by advection and dispersion with matrix diffusion into a porous medium. Finally, they reach a well assumed to be a source of potable water for local residents. The annual individual dose is assessed on this well scenario to find out the significance of an EDZ. A profound sensitivity study was performed, but all results show that the impact is negligible. Even though the role of an EDZ turns out to be limited on overall safety assessment, still it is worthwhile to study the chemical role of an EDZ, such as a potential source for natural colloids, potential sealing of an open fracture by fine clay particles generated by the process of an EDZ, and alteration of a sorption mechanism by an EDZ in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.65-93
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• 2023

Several countries have been operating radioactive waste disposal (RWD) programs to construct their own repositories and have used natural analogues (NA) studies directly or indirectly to ensure the reliability of the long-term safety of deep geological disposal (DGD) systems. A DGD system in Korea has been under development, and for this purpose a generic NA study is necessary. The Korea Atomic Energy Research Institute has just launched the first national NA R&D program in Korea to identify the role of NA studies and to support the safety case in the RWD program. In this article, we review some cases of NA studies carried out in advanced countries considering crystalline rocks as candidate host rocks for high-level radioactive waste disposal. We examine the differences among these case studies and their roles in reflecting each country's disposal repository design. The legal basis and roadmap for NA studies in each country are also described. However because the results of this analysis depend upon different environmental conditions, they can be only used as important data for establishing various research strategies to strengthen the NA study environment for domestic disposal system research in Korea.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.218-231
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• 2002

Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high- level radioactive waste repository by adopting a finite difference method utilizing the control- volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical harriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three- dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of $^{234}$ Ulongrightarrow$^{230}$ Thlongrightarrow$^{226}$ Ra, which is the most important chain as far as the human environment is concerned.

• Tunnel and Underground Space
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• v.17 no.6
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• pp.464-474
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• 2007

An enhancement in the performance and safety of a high-level waste repository requires a validation of its engineered barrier. An engineering-scale test (named "KENTEX") has been conducted to investigate the thermal-hydro-mechanical behaviors in the engineered barrier of the Korean reference disposal system The validation test started on May 31, 2005 and is still under operation. The experimental data obtained allowed a preliminary and qualitative interpretation of the thermal-hydro-mechanical behaviors in the bentonite blocks. The temperature was higher as it became closer to the heater, while it became lower as it was farther away from the heater. The water content had a higher value in the part close to the hydration surface than that in the heater part. The relative humidity data suggested that a hydration of the bentonite blocks might occur by different drying-wetting processes, depending on their position. The total pressure was continuously increased by the evolution of the saturation front in the bentonite blocks and thereby the swelling pressure. Near the heater region, there was also a significant contribution of the thermal expansion of bentonite and the vapor pressure in the pores of the bentonite blocks.

• Tunnel and Underground Space
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• v.23 no.5
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• pp.372-382
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• 2013

Thermal-hydro-mechanical (THM) modeling is a critical R&D issue in the performance and safety assessment of a high-level waste repository. With an $\ddot{A}$sp$\ddot{o}$ prototype repository, its thermal behavior was analyzed and then compared with in-situ experimental data for its validation. A model simulation was used to calculate the temperature distributions in the deposition holes, deposition tunnel, and surrounding host rock. A comparison of the simulation results with the experimental data was made for deposition hole DH-6, which showed that there was a temperature difference of $2{\sim}5^{\circ}C$ depending on the location of the measuring points, but there was a similar trend in the evolution curves of temperature as a function of time. It was expected that the coupled modeling of the thermal behavior with the hydro-mechanical behavior in the buffer and backfill of the $\ddot{A}$sp$\ddot{o}$ prototype repository would give a better agreement between the experimental and model calculation results.