• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.1-12
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• 2022

Bentonite is the most probable candidate to be used as a buffer in a deep geological repository with high swelling properties, hydraulic conductivity, thermal conductivity, and radionuclide sorption ability. Among them, the radionuclide sorption ability prevents or delays the transport of radionuclides into the nearby environment when an accident occurs and the radionuclide leaks from the canister, so it needs to be strengthened in terms of long-term disposal safety. Here, we proposed a surface modification method in which some inorganic additives were added to form NaP zeolite on the surface of the bentonite yielded at Yeonil, South Korea. We confirmed that the NaP zeolite was well-formed on the bentonite surface, which also increased the sorption efficiency of Cs and Sr from groundwater conditions. Both NaP and NaX zeolite can be produced and we have demonstrated that the generation mechanism of NaX and NaP is due to the number of homogeneous/heterogeneous nucleation sites and the number of nutrients supplied from an aluminosilicate gel during the surface modification process. This study showed the potential of surface modification on bentonite to enhance the safety of deep geological radioactive waste repository by improving the radionuclide sorption ability of bentonite.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.283-294
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• 2023

APro, developed in KAERI for the process-based total system performance assessment (TSPA) of deep geological disposal systems, performs finite element method (FEM)-based multiphysics analysis. In the FEM-based analysis, the mesh element quality influences the numerical solution accuracy, memory requirement, and computation time. Therefore, an appropriate mesh structure should be constructed before the mesh stability analysis to achieve an accurate and efficient process-based TSPA. A generic reference case of DECOVALEX-2023 Task F, which has been proposed for simulating stationary groundwater flow and time-dependent conservative transport of two tracers, was used in this study for mesh stability analysis. The relative differences in tracer concentration varying mesh structures were determined by comparing with the results for the finest mesh structure. For calculation efficiency, the memory requirements and computation time were compared. Based on the mesh stability analysis, an approach based on adaptive mesh refinement was developed to resolve the error in the early stage of the simulation time-period. It was observed that the relative difference in the tracer concentration significantly decreased with high calculation efficiency.

• BMB Reports
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• v.56 no.9
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• pp.514-519
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• 2023

Methyltransferase-like 3 (METTL3), a key component of the m⁶A methyltransferase complex, regulates the splicing, nuclear transport, stability, and translation of its target genes. However, the mechanism underlying the regulation of METTL3 expression by alternative splicing (AS) remains unknown. We analyzed the expression pattern of METTL3 after AS in human tissues and confirmed the expression of an isoform retaining introns 8 and 9 (METTL3-IR). We confirmed the different intracellular localizations of METTL3-IR and METTL3 proteins using immunofluorescence microscopy. Furthermore, the endogenous expression of METTL3-IR at the protein level was different from that at the mRNA level. We found that 3'-UTR generation by intron retention (IR) inhibited the export of METTL3-IR mRNA to the cytoplasm, which in turn suppressed protein expression. To the best of our knowledge, this is the first study to confirm the regulation of METTL3 gene expression by AS, providing evidence that the suppression of METTL3 protein expression by IR is an integral part of the mechanism by which 3'-UTR generation regulates protein expression via inhibition of RNA export to the cytoplasm.

• Journal of Radiation Industry
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• v.17 no.4
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• pp.443-449
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• 2023

KAERI(Korea Atomic Energy Research Institute) is conducting research and development of large-scale radiation generators and the latest radiation measuring instruments. In particular, research and development of security screening equipment using an electron beam accelerator and a neutron generator is in progress recently. Globally, 20 ft containers are used to transport imports and exports, and electron beam accelerators are radiation sources to measure the shape of the material inside the container during customs inspections in each country. KAERI is developing a device that can use an electron beam accelerator and a neutron generator sequentially to grasp the shape of various materials as well as the location of the internal target material. In this study, when using the neutron generator, the radiation dose and the degree of activation by neutron for the facility and surrounding environment, facility equipment were simulated using MCNP and FISPACT code. As a result, the shielding structures inside and outside the radiation control area were satisfactory to the reference level established conservatively based on the Korean Nuclear Act.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.227-235
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• 2020

In this study, we discussed the limitations of gross alpha measurements for the characterization of radioactive wastes produced in nuclear facilities through experimental tests and Monte Carlo N-particle transport simulations. The determination of gross alpha is essential for the disposal of radioactive waste produced in nuclear facilities in Korea. The measurements of gross alpha are easy to perform and yield rapid analytical results, but it cannot be used for quantitative analysis. The error of counting efficiency for gross alpha with various masses of the deposit on planchets using KCl and ²⁴¹Am was determined. The relative deviation of the counting efficiency in samples having the same mass was 20%. Uranium was extracted from the soil through acid leaching and extraction chromatography, and the concentration of U determined by inductively coupled plasma-mass spectrometry (ICP-MS) was compared with the results for gross alpha. The gross alpha was underestimated by 50% compared to the U concentration by ICP-MS. The counting efficiency depended on the energy from the alpha emitters, which differed by up to three times in determination of the counting efficiency depending on the kinds of alpha radionuclides of interest. Therefore, the gross alpha is not compatible with the sum of radioactivity for each alpha emitter and is suitable as a screening method.

• Microbiology and Biotechnology Letters
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• v.37 no.3
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• pp.287-292
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• 2009

The nucleocapsid (N) protein of porcine reproductive and respiratory syndrome virus (PRRSV) is a basic multifunctional protein which has been reported to be a serine phosphoprotein with yet-identified functions. As a first step towards understanding the general role of N protein phosphorylation during virus replication, the non-phosphorylated mutant N gene was constructed by mutating all serine residues to alanine. This recombinant N protein was identified to be unphosphorylated, confirming that serine residues truly function as core amino acids responsible for N protein phosphorylation. The PRRSV N protein has been shown to possess the biological features of nuclear localization and N-N homodimerization which individually play critical roles in virus infection. In the present study, therefore, it was attempted to investigate whether these two properties of the N protein are modulated by its phosphorylation status. However, experimental results showed that the non-phosphorylated N protein was still present in the nucleus and nucleolus, and was able to associate with itself by non-covalent interactions. Taken together, the data suggest phosphorylation-independent regulation of N protein nuclear transport or oligomerization, thereby implying the potential involvement of phosphorylation in regulating the activities of the N protein at other levels including RNA-binding capacity.

• Journal of the Korean Electrochemical Society
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• v.4 no.3
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• pp.87-93
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• 2001

Pressure-balanced external Ag/AgCl electrode has been extensively used for both Pressurized Water Reactor (PWR) and Boiling Water Reactor (PWR) environments. The use of KCI-based buffer solution often becomes the source of electrode potential drift due to slow leakage through its porous plug, typically made of zirconia. It is reported that results of our effort to improve the stability of electrode potential by using high purity water as the filling solution in which $Cl^-$ ion activity can be established and maintained at the solubility of AgCl even with the sustained leakage for a long period. Stability tests have been made in boron and lithium mixture solution at $288^{\circ}C$. The electrode potential remained stable within 10 mV over one week period. And after a thermal cycle between 288 to $240^{\circ}C$ the potential shift of Ag/AgCl electrodes did not exceed 15 mV By using the limiting equivalent ionic conductances and Agar's hydrodynamic theory, the thermal liquid junction potential (TLJP) of the electrode has been predicted. The calculated values for the water-fiued Ag/AgCl electrode potential, in which the chlorine concentration in the filling solution was derived from the measured data at ambient temperature, had a good agreement with the experimental values.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.9-15
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• 2006

For the accurate safety assessment of potential radioactive waste disposal site which is located in the crystalline rock it is important to simulate the mass transportation through engineered and natural barrier system precisely, characterized by porous and fractured media respectively. In this work the methods to construct discrete fracture network for the analysis of flow and mass transport through fractured-porous medium are described. The probability density function is adopted in generating fracture properties for the realistic representation of real fractured rock. In order to investigate the intersection between a porous and a fractured medium described by a 2 dimensional rectangular and a cuboid grid respectively, an additional imaginary fracture is adopted at the face of a porous medium intersected by a fracture. In order to construct large scale flow paths an effective method to find interconnected fractures and algorithms of swift detecting connectivities between fractures or porous medium and fractures are proposed. These methods are expected to contribute to the development of numerical program for the simulation of radioactive nuclide transport through fractured-porous medium from radioactive waste disposal site.

• 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.