• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.145-150
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• 2010

The paper is describing work being developed in the frame of a 5-year IAEA Coordinated Research Programme (CRP) started in late 2005. Participants gained knowledge of modelling methodologies and experience in the development and use of rather sophisticated simulation tools in support of site characterization and performance assessment calculations. These goals were achieved by a coordinated effort, in which the advantages and limitations of numerical models are examined and demonstrated through a comparative analysis of simplified, illustrative test cases. This knowledge and experience should help them address these issues in their own country's nuclear waste program. Coordination efforts during the first three years of the project aimed at enabling this transfer of expertise and maximizing the learning experience of the participants as a group. This was accomplished by identifying common interests of the participants (i.e., Process Modelling and Total System Performance Assessment methodology), and by defining complementary tasks that are solved by the members. Synthesis of all available results by comparative assessments is planned in the coming months. The project will be completed end of 2010. This paper is summarizing activities up to November 2009.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.135-148
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• 2005

In the stage of conceptual design for the construction and operation of the geologic repository for radioactive wastes, it is important to consider a repository ventilation system which serves the repository working environment, hygiene & safety of the public at large, and will allow safe maintenance like moisture content elimination in repository for the duration of the repositories life, construction/operation/closure, also allowing safe waste transportation and emplacement. This paper describes the possible ventilation system design criteria and requirements for the prospective Korean radioactive waste repositories with emphasis on the underground rock cavity disposal method in the both cases of low & medium-level and high-level wastes. It was found that the most important concept is separate ventilation systems for the construction (development) and waste emplacement (storage) activities. In addition, ventilation network system modeling, natural ventilation, ventilation monitoring systems & real time ventilation simulation, and fire simulation & emergency system in the repository are briefly discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.145-154
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• 2007

In hydrothermal reaction tests, smectite-to-illite conversion was identified using a domestic bentonite which is favorably considered as a buffer material, and its dependency on various hydrothermal conditions was investigated. The analysis results of the XRD and Si concentration indicated that the smectite-to-illite conversion was a major process of bentonite alteration under the hydrothermal conditions. The temperature, potassium concentration in solution, and pH were observed to significantly affect the smectite-to illite conversion. A model of conversion reaction rate was suggested to evaluate the long-term stability of smectite composing a major constituent of bentonitic buffer. It was expected from the evaluation results that the smectite would keep its integrity for very long disposal time under a normal condition, whitens it might be converted to illite by 50 percent after over $5{\times}10^4$ year of disposal time under a conservative condition and consequently lose its swelling capacity as a buffer material of a repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.183-190
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• 2009

The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.135-147
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• 2016

The buffer in geological disposal system is one of the major elements to restrain the release of radionuclide and to protect the container from the inflow of groundwater. The buffer material requires long-term stability, low hydraulic conductivity, low organic content, high retardation of radionuclide, high swelling pressure, and high thermal conductivity. These requirements could be determined by the quantitative analysis results. In case of South Korea, the bentonites produced in Gyeongju area have been regarded as candidate buffer/backfill materials at KAERI (Korea Atomic Energy Research Institute) since 1997. According to the study on several physical and chemical characteristics of domestic bentonite in the same district, this is the Ca-type bentonite with about 65% of montmorillonite content. Through this study, we present the criteria for the performance evaluation items and methods when collecting new buffer/backfill materials.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.4
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• pp.459-467
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• 2021

In the majority of countries, the upper limit of buffer temperature in a repository is set to below 100℃ due to the possible illitization. This smectite-to-illite transformation is expected to be detrimental to the swelling functions of the buffer. However, if the upper limit is increased while preventing illitization, the disposal density and cost-effectiveness for the repository will dramatically increase. Thus, understanding the characteristics and creating a database related to the buffer under the elevated temperature conditions is crucial. In this study, a strategy to investigate the bentonite found in Korea under the elevated temperatures from a mineral transformation and radionuclides retardation perspective was proposed. Certain long-term hydrothermal reactions generated the bentonite samples that were utilized for the investigation of their mineral transformation and radionuclide retardation characteristics. The bentonite samples are expected to be studied using in-situ synchrotron-based X-Ray Diffraction (XRD) technique to determine the smectite-to-illite transformation. Simultaneously, the 'high-temperature and high-pressure mineral alteration measurement system' based on the Diamond Anvil Cell (DAC) will control and provide the elevated temperature and pressure conditions during the measurements. The kinetic models, including the Huang and Cuadros model, are expected to predict the time and manner in which the illitization will become detrimental to the performance and safety of the repository. The sorption reactions planned for the bentonite samples to evaluate the effects on retardation will provide the information required to expand the current knowledge of repository optimization.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.143-149
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• 2012

Groundwater flow simulations were performed to obtain data of groundwater flow used in a safety assessment for a hypothetical geological disposal facility assumed to be located in the KURT (KAERI Underground Research Tunnel) site. A regional scale modeling of the groundwater flow system was carried out to make boundary conditions for a local scale modeling. And, fracture zones identified at the study site were involved in the local scale groundwater flow model. From the results of the local scale modeling, a hydraulic head distribution was indicated and it was used in a particle tracking simulation for searching pathway of groundwater from the location of the hypothetical disposal facility to the surface where the groundwater reached. The flow distance and discharge rate of the groundwater in the KURT site were calculated. It was thought that the modeling methods used in this study was available to prepare the data of groundwater flow in a safety assessment for a geological disposal facility of radioactive wastes.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.271-305
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• 2020

An effect of coupled thermo-hydro-mechanical and chemical (THMC) behavior is an essential part of the performance and safety assessment of geological disposal systems for high-level radioactive waste and spent nuclear fuel. Furthermore, numerical models and modeling techniques are necessary to analyze and predict the coupled THMC behavior in the disposal systems. However, phenomena associated with the coupled THMC behavior are nonlinear, and the constitutive relationships between them are not well known. Therefore, it is challenging to develop numerical models and modeling techniques to analyze and predict the coupled THMC behavior in the geological disposal systems. It is also difficult to verify and validate the development of the models and techniques because it requires expensive laboratory tests and in-situ experiments that need to be performed for a long time. DECOVALEX was initiated in 1992 to efficiently develop numerical models and modeling techniques and validate the developed models and techniques against the lab and in-situ experiments. In Korea, Korea Atomic Energy Research Institute has participated in DECOVALEX-2011, DECOVALEX-2015, and DECOVALEX-2019 since 2008. In this study, all tasks in the three DECOVALEX projects were introduced to the researcher in the field of rock mechanics and geotechnical engineering in Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.183-193
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• 2018

Currently, the Korea Atomic Energy Research Institute (KAERI) is planning to build the Ki-Jang Research Reactor (KJRR) in Ki-Jang, Busan. It is important to safely dispose of low-level radioactive waste from the operation of the reactor. The most efficient way to treat radioactive waste is cement solidification. For a radioactive waste disposal facility, cement solidification is performed based on specific waste acceptance criteria such as compressive strength, free-standing water, immersion and leaching tests. Above all, the leaching test is important to final disposal. The leakage of radioactive waste such as $^{137}Cs$ causes not only regional problems but also serious global ones. The cement solidification method is simple, and cheaper than other solidification methods, but has a lower leaching resistance. Thus, this study was focused on the development of cement solidification for an enhancement of cesium leaching resistance. We used Zeolite and Loess to improve the cesium leaching resistance of KJRR cement solidification containing simulated KJRR liquid waste. Based on an SEM-EDS spectrum analysis, we confirmed that Zeolite and Loess successfully isolated KJRR cement solidification. A leaching test was carried out according to the ANS 16.1 test method. The ANS 16.1 test is performed to analyze cesium ion concentration in leachate of KJRR cement for 90 days. Thus, a leaching test was carried out using simulated KJRR liquid waste containing $3000mg{\cdot}L^{-1}$ of cesium for 90 days. KJRR cement solidification with Zeolite and Loess led to cesium leaching resistance values that were 27.90% and 21.08% higher than the control values. In addition, in several tests such as free-standing water, compressive strength, immersion, and leaching tests, all KJRR cement solidification met the waste acceptance or satisfied the waste acceptance criteria for final disposal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.247-254
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• 2012

Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of $SiO_2$ and $B_2O_3$ for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.