• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.315-334
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• 2014

Wolsong Low- and Intermediate-level radioactive waste (LILW) disposal center has two different types of disposal facilities and interacts with the neighboring Wolsong nuclear power plant. These situations impose a high level of complexity which requires in-depth understanding of phenomena in the safety assessment of the disposal facility. In this context, multidimensional radionuclide transport model and hydraulic performance assessment model should be developed to identify more realistic performance of the complex system and reduce unnecessary conservatism in the conventional performance assessment models developed for the $1^{st}$ stage underground disposal. In addition, the advanced performance assessment model is required to calculate many cases to treat uncertainties or study parameter importance. To fulfill the requirements, this study introduces the development of two-dimensional integrated near-field performance assessment model combining near-field hydraulic performance assessment model and radionuclide transport model for the $2^{nd}$ stage near-surface disposal. The hydraulic and radionuclide transport behaviors were evaluated by PORFLOW and GoldSim. GoldSim radionuclide transport model was verified through benchmark calculations with PORFLOW radionuclide transport model. GoldSim model was shown to be computationally efficient and provided the better understanding of the radionuclide transport behavior than conventional model.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.2
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• pp.95-102
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• 2013

There was a study for biological characteristics, except for physico-chemical and mineralogical properties, on the natural bentonite that is considered as a buffer material for the high-level radioactive waste disposal site. A bentonite slurry that was prepared from a local 'Gyeongju bentonite' in Korea was incubated in a serum bottle with nutrient media over 1 week and its stepwise change was observed with time. From the activated bentonite in the nutrient media, we can find a certain change of both solid and liquid phases. Some dark and fine sulfides began to be generated from dissolved sulfate solution, and 4 species of sulfate-reducing bacteria (SRB) were identified as living cells in samples that were periodically taken and incubated. These results show that sulfate-reducing (or metal-reducing) bacteria are adhering and existing in the powder of bentonite, suggesting that there may be a potential occurrence of longterm biogeochemical effects in and around the bentonite buffer in underground anoxic environmental conditions.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.45-59
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• 2023

The transport calculation for a wide variety of radionuclides contained in high-level radioactive waste, especially spent nuclear fuel, is computationally difficult, and input data collection for this also take a considerable amount of time. Accordingly, considering limited resources, it is possible to reduce the calculation time while minimizing impact on accuracy by including only radionuclides important to calculation result through applying some criteria among potential radiation source terms that may release into environment. In this paper, therefore, we reviewed and analyzed the screening process performed to select radionuclides to be considered in the safety assessment for the KBS-3 type repository in Sweden and Finland. In both countries, it was confirmed that a list of radionuclides was selected by comprehensively considering screening criteria such as radioactivity inventory, half-life, radiotoxicity, risk quotient, and transport properties, and etc. A comparison of radionuclides included in the radiological safety assessment in both countries suggests that most of nuclides are considered in common, and a few nuclides considered only in one country are due to differences in decay chain treatment or spent fuel types. As of now, since most of information on the disposal facility in Korea has not been determined, it is necessary to comprehensively model release and transport of all radionuclides considered in Sweden and Finland when performing the radiological safety assessment. Based on these results, we derived the screening concept of selecting a list of radionuclides to be considered in the radiological safety assessment for the domestic KBS-3 type geological disposal facility, and this result is expected to be used as technical basis for confirming conformity with the safety objective. In a more detailed evaluation reflecting domestic characteristics in the future, it would be desirable to consider only radionuclides selected in accordance with the screening procedure. However, further research should be conducted to determine the quantitative limit for each criteria.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.207-213
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• 2010

The effect of temperature on swelling pressure was observed with a Korean domestic Ca-bentonite which has been considered as a potential buffer material in the engineering barrier of a high level radioactive waste (HLW) disposal system. The Ca-bentonite was compacted to a dry density of 1.6 g/$cm^3$, and then de-ionized water was supplied into it with a constant pressure of 0.69 MPa. The equilibrium swelling pressures were measured with different temperatures of $25^{\circ}C$, $30^{\circ}C$, $40^{\circ}C$, $50^{\circ}C$, $60^{\circ}C$, $70^{\circ}C$, respectively. The Ca-bentonite showed a sufficiently high swelling pressure of 5.3 MPa at room temperatures. Then it was clearly showed that the equilibrium swelling pressure was decreased with an increase of temperature. Interestingly, there were some differences in temperature effect on the equilibrium swelling pressure when the environmental temperature is increasing or decreasing. For further clarifying the swelling behaviour of a Korea domestic Ca-bentonite, the change of a compaction level, and the composition variation of a supplied water would be needed to use in conceptual design of HLW disposal system.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2174-2181
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• 2024

In this work, Sr_0.5Zr₂(PO₄)₃-SmPO₄ dual-phase ceramics were prepared via in-situ synthesis process, which is a potential novel nuclear waste form for immobilizing the fission product ⁹⁰Sr and the trivalent actinide radionuclides in high-level waste (HLW). And the preparation technology, microstructure and chemical durability of Sr_0.5Zr₂(PO₄)₃-SmPO₄ dual-phase ceramics were systematically investigated. It was confirmed that the optimum microwave-sintering temperature (1050 ℃) and heat preservation time (1.5 h) is estimated by Archimedes method. Besides, the as-prepared samples that were consisted of strontium zirconium phosphate (SrZP) and monazite showed the remarkable densification, in which the two crystalline phases were intermixed well with each other. Meanwhile, the formation and evolution of microstructure was also consistent with the variational rule of Sr_0.5Zr₂(PO₄)₃/SmPO₄, indicating that there was not mutual reaction during the in-situ synthesis process. The PCT and MCC-1 experimental results demonstrated that the elemental normalized leaching rates of tested samples are all at a low level (LR_Sr ~10^-4 g·m^-2·d^-1, LR_Zr ~10^-8-10^-6 g·m^-2·d^-1, LR_Sm ~10^-7-10^-5 g·m^-2·d^-1 and LR_P ~10^-4 g·m^-2·d^-1). It is indicated that Sr_0.5Zr₂(PO₄)₃-SmPO₄ dual-phase ceramics possesses excellent chemical durability for HLW disposal.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.130-139
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• 2004

The cladding materials remaining after reprocessing process of the nuclear fuel, generally called as hulls, are classified as a high-level radioactive waste. They are usually packaged in the container for disposal after being compacted, melted, or solidified into the matrix. The efforts to fabricate a better ingot for a more favorable disposal to the environment have failed due to the technical difficulties encountered in the chemical decontamination method. In the early 1990s, the accumulation of radio-chemical data on hulls and the advent of new technology such as a laser or plasma have made the pre-treatment of the hulls more efficient. This paper summarizes the information regarding the radio-chemical analysis of the hull through a literature survey and determines the characteristics of the hull and depth profile of the radio-nuclides within the hull thickness. The feasibility study was carried out to evaluate the reduction of the radioactivity by peeling off the surface of the hull with the application of laser technology.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.243-253
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• 2009

In this study, research status on radionuclide and colloid migration in underground research facilities including KURT (KAERI Underground Research Tunnel) was investigated. Some foreign underground research facilities constructed in crystalline rock formations such as granite were briefly introduced and compared. International joint researches concerned with the radionuclide and colloid migration were investigated particularly for the Grimsel Test Site (GTS) and $\ddot{A}$sp$\ddot{o}$ Hard Rock Laboratory by analyzing major research items, on-going research projects, and future plans.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.1
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• pp.9-16
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• 2024

The bentonite buffer material is a crucial component in an engineered barrier system used for the disposal of high-level radioactive waste. Because a large amount of heat from the disposal canister is released into the bentonite buffer material, the thermal conductivity of the bentonite buffer is a crucial parameter that determines the design temperature. At the Korea Atomic Energy Research Institute (KAERI), a new standard bentonite (Bentonil-WRK) has been used since 2022 because Gyeongju (KJ) bentonite is no longer produced. However, the currently available data are insufficient, making it essential to investigate both the basic and complex properties of Bentonil-WRK. Thus, this study evaluated its geotechnical and thermal properties and developed a thermal conductivity empirical model that considers its dry density, water content, and temperature variations from room temperature to 90℃. The coefficient of determination (R²) for the model was found to be 0.986. The thermal conductivity values of Bentonil-WRK were 1-10% lower than those of KJ bentonite and 10-40% higher than those of MX-80 bentonites, which were attributable to mineral-composition differences. The thermal conductivity of Bentonil-WRK ranged between 0.504 and 1.149 W·(m^-1·K^-1), while the specific heat capacity varied from 0.826 to 1.138 (kJ·(kg^-1·K^-1)).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.121-133
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• 2014

If the spent fuels or the high-level radioactive wastes can be disposed of in the depth of 3~5 km and more stable rock formation, it has several advantages. For example, (1)significant fluid flow through basement rock is prevented, in part, by low permeability, poorly connected transport pathways, and (2)overburden self-sealing. (3)Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose-critical radionuclides at the depth. Finally, (4) high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept to the deep geological disposal concept(DGD), very deep borehole disposal(DBD) technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, for the preliminary applicability analyses of the DBD system for the spent fuels or high level wastes, the DBD concepts which have been developed by some countries according to the rapid advance in the development of drilling technology were reviewed. To do this, the general concept of DBD system was checked and the study cases of foreign countries were described and analyzed. These results will be used as an input for the analyses of applicability for DBD in Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.