• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.25-32
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• 2006

This study examined the possibility of using dark-striped field mice as a biological indicator for the environmental radio-surveillance. For this study, dark-striped field mice were caught from five areas of Kyonggi, Kyongsang, Chungchong and Cholla provinces. The external morphological characteristics and isoenzymic types of dark-striped field mice were studied after they were captured. Among the external morphological characteristics, the dark-brown coat, dark back stripe, head-to-tail length, tail length, and ear length matched the taxonomical characteristics of dark-striped field mice. The analyses on L-lactate dehydrogenase, aspartate aminotransferase, and malate dehydrogenese revealed that one species of dark-striped field mice, called Apodemus agrarius, was inhabitated throughout a wide range of Korea. On the other hand, the frequency of micronuclei in peripheral polychromatic erythrocytes to survived mice after irradiation also analyzed. The LD50/30 of A. cgrarius and ICR mice were approximately 5 Gy and 7.9Gy, respectively. The results of the study reveal that wild A. asrarius have a high potential as a biological monitoring system to determine the impact of radiation in areas such as those within the vicinity of nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.49-62
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• 2011

Recently, International Atomic Energy Agency and major leading countries in radioactive waste management tend to subdivide the categories of radioactive waste based upon risk-graded approach. In this context, the category of very low level waste has been newly introduced, or optimized management options for this kind of waste have been pursued in many countries. The application of engineered surface landfill type facilities dedicated to dispose of very low level waste has been gradually expanded, and it was analyzed that their design concept of isolation has been much advanced than those of the old fashioned surface trench-type disposal facilities for low and intermediate level waste, which were usually constructed in 1960's. In addition, the management options for very low level waste in major leading countries are varied depending upon and interfaced with the affecting factors such as: national framework for clearance, legal and practical availability of low and intermediate level waste repository and/or non-nuclear waste landfill, public acceptance toward alternative waste management options, and so forth. In this regard, it was concluded that optimized long-term management options for very low level waste in Korea should be also established in a timely manner through comprehensive review and discussions, in preparation of decommissioning of large nuclear facilities in the future, and be implemented in a systematic manner under the framework of national policy and management plan for radioactive waste management.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.33-40
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• 2011

In the compacted bentonite containing $Ag_2O$, the transport of iodide ion was investigated by Through-diffusion method. It is confirmed that Iodide ion is transported by diffusion process in the compacted bentonite containing $Ag_2O$ as well as in the compacted bentonite without $Ag_2O$. However, the lag-time of iodide ion in the compacted bentonite containing $Ag_2O$ is larger than that in the compacted bentonite without $Ag_2O$. The increase of the lag-time was observed in pure iodide ion solution and also in 0.1M NaCl-iodide ion solution. The apparent diffusion coefficient of iodide ion in the compacted bentonite containing $Ag_2O$ has lower value than that in the compacted bentonite without $Ag_2O$. The effect of $Ag_2O$ on the effective diffusion coefficient was not clearly investigated in the compacted bentonite containing $Ag_2O$ while the values of effective diffusion coefficient of iodide ion in the compacted bentonite without $Ag_2O$ obtained in this study were similar to those in the compacted bentonite reported in the literature.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.23-32
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• 2011

An engineering scale test, which is called KENTEX, was carried out to understand and to analyze the coupled thermal, hydrological and mechanical phenomena in the engineered barrier system(EBS) of Korean reference disposal system. Using the experimental data obtained from KENTEX, the water saturation processes in bentonite could be analyzed. From the comparison between the model calculation using ABAQUS and the experimental results, the difference of the water content between them in the unsaturating part was large because the drying phenomena due to moisture redistribution by the temperature gradient could not be included in the model. In the saturating part, the difference of the water content between them was decreased gradually and showed to be small in the full saturation. And the time of about 95% saturation could be estimated about 500 days from the model calculation and experimental results. Also it could be known that the moisture redistribution in the unsaturated part could not be affected on the saturation time of bentonite in the repository. Therefore, it is considered that this model could be used to quantitatively predict the water saturation time in bentonite as EBS for the disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.141-148
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• 2011

It is clarified in the radioactive waste transfer regulation that the concentration of radioactive waste for the major radio nuclide has to be examined when radioactive waste is guided to the radioactive waste stores. In case of the low level radioactive waste sample, the analytical results of radioactive waste concentration frequently show a value lower than minimum detectable activity (MDA). Since the MDA value basically depends on the amount of a sample, background value, measurement time, counting efficiency, and etc, it would be necessary to increase a sample amount with a intention of minimizing MDA. In order to measure a concentration of $^{129}I$ in low and medium level radioactive waste, $^{129}I$ was collected by using a distillation technique after leaching the simulated radioactive waste sample with a non-volatile acid. The recovery of $^{129}I$ measured was compared with that measured with column elution technique which is a conventional method using an anion-exchange resin. The recovery of inactive iodide by using the distillation method and column elution were found as $86.5{\pm}0.9%$ and $87.3{\pm}2.7%$, respectively. The recovery and MDA value calculated for distillation technique when 100 g of extracted solution of $^{129}I$ was taken, were found to be $84.6{\pm}1.6%$ and $1.2{\times}10^{-4}Bq/g$, respectively. Consequently, the proposed technique with simplified process lowered the MDA value more than 10 times compared to the column elution technique that has a disadvantage of limited sampling amount.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.13-21
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• 2011

To characterize the site specific properties of a study area for high-level radioactive waste disposal research in KAERI, the several geological investigations such as surface geological surveys and borehole drillings were carried out since 1997. Especially, KURT (KAERI Underground Research Tunnel) was constructed to understand the further study of geological environments in 2006. As a result, the first geological model of a study area was constructed by using the results of geological investigation. The objective of this research is to construct a hydrogeological model around KURT area on the basis of the geological model. Hydrogeological data which were obtained from in-situ hydraulic tests in the 9 boreholes were estimated to accomplish the objective. And, the hydrogeological properties of the 4 geological elements in the geological model, which were the subsurface weathering zone, the log angle fracture zone, the fracture zones and the bedrock were suggested. The hydrogeological model suggested in this study will be used as input parameters to carry out the groundwater flow modeling as a next step of the site characterization around KURT area.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.121-129
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• 2011

The effects of the applied current density, the $AgNO_3$ concentration, the scrubbing liquid flow rate and the NO-air mixed gas flow rate on the NO removal efficiency were investigated by using $Ag^{2+}$ mediated electrochemical oxidation (MEO). Results showed that the NO removal efficiency increased with increasing the applied current density. The effect of the $AgNO_3$ concentration on the NO removal efficiency was negligibly small in the concentration of $AgNO_3$ above 0.1 M. When the scrubbing liquid flow rate increased, the NO removal efficiency was gradually increased. On the other hands, the NO removal efficiency decreased with increasing the NO-air mixed gas flow rate. As a result of the treatment of NO-air mixed gas by using the MEO process with the optimum operating condition and the chemical absorption process using 3 M $HNO_3$ solution as a scrubbing liquid, the removal efficiency of NO and $NO_x$ was achieved as 95% and 63%, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.17-23
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• 2008

In this study, plasma processing of metal surface is experimentally investigated to enhance the surface decontamination efficiency and to find out the reaction mechanism. Cobalt, the major contaminant in the nuclear facilities, and three fluorine-containing gases, $CF_4/O_2$, $SF_6/O_2$, and $NF_3$ are chosen for the investigation. Thin metallic disk specimens are prepared and their surface etching reactions with the three plasma gases are examined. Results show that the maximum etching rate of $17.2\;{\mu}m/min.$ is obtained with NF3 gas at $420^{\circ}C$, while with $CF_4/O_2$, $SF_6/O_2$ gas plasmas those of $2.56\;{\mu}m/min.$ and $1.14\;{\mu}m/min.$ are obtained, respectively. Along with etching experiments, constituent elements of the reaction products are identified to be cobalt, oxygen, and fluorine by AES (Auger Electron Spectroscopy) analysis. It turns out that the oxygen atoms are physically adsorbed ones to the surface from the ambient not participation ones during the analysis after reaction, which supports that the surface reaction of cobalt is mainly to be a fluorination reaction.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.1-9
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• 2008

This study examined the effect of an electrokinetic-flushing remediation for a soil of a high permeability. The soil was sampled from the site around a research atomic reactor which had high hydro-conductivities due to a high content of sand in the soil. The flow rate of the washing reagent was fast at the beginning but it was reduced as time lapsed. In the case of using citric acid as a washing reagent, the flow rate was fastest, 78.7 ml/day. The removal efficiencies of $Co^{2+}$ and $Cs^+$ from a soil cell with acetic acid were the highest, which were 95.2% and 84.2% respectively. The soil waste-solution volume generated from the electrokinetic remediation was reduced to about 1/20 of that from the soil washing remediation. Meanwhile, the electrokinetic-flushing method enhanced the removal efficiencies of $Co^{2+}$ and $Cs^+$ from the soil by about 6% and 2% respectively, compared to those by the electrokinetic method. Consequently, it was found that the electrokinetic-flushing method was more effective for the remediation of a soil with a high permeability.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.19-27
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• 2007

Combined method of extraction chromatography and liquid scintillation analysis was established for determinating $^{55}Fe\;and\;^{63}Ni$ radioactivity in solid samples. Activated concrete wastes generated from dismantling KRR-2 were analysed. The sequential separation including precipitation and extraction chromatography resulted in the above 90% chemical recoveries of Fe. Above 62% recoveries of Ni were obtained by this procedure exception to 43.6 and 46.5% recoveries. The seperation and counting procedure was also confirmed with spiked samples of known quantity. The measured and spiked quantity were agreed with the 3.7% and 0.7% variations in the $^{55}Fe\;and\;^{63}Ni$ experiments, respectively. The radioactivities of $^{55}Fe$ in the dismantled concretes are shown from below MDA to maximum 362 Bq/g. The radioactivities of $^{63}Ni$ in all concrete samples are below MDA. The $^{63}Ni$ doesn't exist in dismantled concretes from KRR-2. The radioactivity of $^{55}Fe$ is decreased rapidly as the sampling depth is increased from the concrete surface.