• Economic and Environmental Geology
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• v.51 no.2
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• pp.77-85
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• 2018

There would be a possibility of uranium contamination around the nuclear power plants and the underground waste disposal sites, where the uranium could further migrate and diffuse to some distant places by groundwater. It is necessary to understand the biogeochemical behaviors of uranium in underground environments to effectively control the migration and diffusion of uranium. In general, various kinds of microbes are living in soils and geological media where the activity of microbes may be closely connected with the redox reaction of nuclides resulting in the changes of their solubility. We investigated the adsorption and precipitation behaviors of dissolved uranium on some solid materials using hydrogen gas as an electron donor instead of organic matters. Although the effect of hydrogen gas did not appear in a batch experiment that used granite as a solid material, there occurred a reduction of uranium concentration by 5~8% due to hydrogen in an experiment using bentonite. This result indicates that some indigenous bacteria in the bentonite that have utilized hydrogen as the electron donor affected the behavior (reduction) of uranium. In addition, the bentonite bacteria have showed their strong tolerance against a given high temperature and radioactivity of a specific waste environment, suggesting that the nuclear-biogeochemical reaction may be one of main mechanisms if the natural bentonite is used as a buffer material for the disposal site in the future.

• 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.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.5 no.1
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• pp.29-37
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• 2007

A conventional single compartment model cannot simulate reasonably the migration phenomenon of contaminants through unsaturated zone, due to the intrinsic unrealistic assumption of the compartment model that contaminants entering a compartment are immediately and uniformly mixed. Although, a multi-compartments model, in which even physically identical layer is divided into multiple compartments, may be used for explaining the retardation of contaminant mass flux along with increasing number of compartments, its numerical modeling is usually time-consuming and appropriate analytical solutions have not been reported yet. In order to improve the conventional compartment models on contaminant migration through unsaturated zone, a series of analytical solutions for multi-compartments model were derived and a generalized constraint under which the results from multi-compartments model can be simply approximated by single compartment model was proposed. The simplified approximation method was verified by a simple numerical analysis on the constraint under hypothetical conditions. It was also proved that the influent contaminant transfer rate from the bulk unsaturated zone can be generally represented into a time-dependent nominal transfer rate rather than a constant. In addition, the nominal transfer rate turned out to be very sensitive to the contaminant transfer rate between compartments in unsaturated zone, but to be almost insensitive to the transfer rate from contaminated zone. It is expected that the simplified approximation method developed in this study can be used for rapid and reasonable estimation of the migration phenomenon of contaminant through unsaturated zone, instead of time-consuming multi-compartments modeling.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.125-133
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• 2004

Fission products of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel, irradiated in HANARO research reactor with 61 ㎾/m of maximum linear power and 1,770 ㎿d/tU of average burn-up, was characterized by EPMA(Electron Probe Micro Analyzer). In order to find accurate characterization, the analysis results by EPMA of fresh simulated DUPIC fuel containing fission products as chemicals were compared with that of wet chemical analysis. The metallic precipitates observed at the center of the fresh simulated DUPIC fuel were about 1 $\mu\textrm{m}$ in size and their major components by EPMA were Mo-53.89 at.%, Ru-37.40 at.%, and Pd+Rh-8.71 at.%. Established procedure through the fresh simulated DUPIC fuel was applied to the irradiated DUPIC fuel. Observed size of metallic precipitates were 2∼2.5 $\mu\textrm{m}$ and their compositions were Mo-47.34 at.%, Ru-46 at.%, and Pd+Rh-6.65 at.%. What are uncommon things for this experiment, special treatment for improving the conductivity was attempted to the specimen and the conditions of exact irradiation of electron beam to small metallic precipitate were suggested.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.55-61
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• 2015

When a radiation detector is applied to the measurement of the radioactivity of high-level of radioactive materials or the rapid response to the nuclear accident, several collimators with the different inner radii should be prepared according to the level of dose rate. This makes the in-situ measurement impractical, because of the heavy weight of the collimator. In this study, an IRIS collimator was developed so as to have a function of controlling the inner radius, with the same method used in optical camera, to vary the attenuation ratio of radiation. The shutter was made to have the double tungsten layers with different phase angles to prevent the radiation from penetrating owing to the mechanical tolerance. The performance evaluation through the MCNP code was conducted by calculating the attenuation ratio according to the inner radius of the collimator. The attenuation ratio was marked on the outer scale ring of the collimator. It is expected that when a radiation detector with the IRIS collimator is used for the in-situ measurement, it can change the attenuation ratio of the incident photon to the detector without replacing the collimator.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.259-269
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• 2013

Applicability of ammonium molybdophosphate/iron oxides-polyacrylonitrile (AMP/IO-PAN) composites on the removal of radionuclides in the radioactive wastewater generated from nuclear power plants was investigated. The composites were characterized using the following analytical techniques: X-ray diffraction (XRD), Fourior transform-infrared (FT-IR) spectroscopy, scanning electron microscopy (SEM), particle size analyzer (PSA), nitrogen adsorption-desorption and magnetic property measurement system (MPMS). 10wt% of AMP/IO-PAN composite has a saturation magnetization of 2.038 emu/g. Single-solute sorptions of Co, Sr and Cs onto 10wt% of AMP/IO-PAN composite were investigated. The maximum sorption capacities ($Q^0$) predicted by the Langmuir model on 10wt% of AMP/IO-PAN composite were 0.097, 0.086 and 0.66 mmol/g for Co, Sr and Cs, respectively. The maximum sorption capacities ($Q^0$) of Cs predicted by Langmuir model on 0, 10, 20 and 30wt% of AMP/IO-PAN composites were 0.702, 0.655, 0.602 and 0.559 mmol/g, respectively. The maximum sorption capacities ($Q^0$) of Cs decreased with increasing the iron oxide content in the AMP/IO-PAN composites.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.193-200
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• 2005

[ $^{14}C$ ] and $^{3}H$ in the evaporator bottom (EB) discharged from the Nuclear power plant (NPP) were extracted simultaneously into a gaseous $^{14}CO_{2}$ and liquefied HTO by using the chemical oxidation, which is the method to oxidize samples completely using potassium persulfate and sulfuric acid. The extracted $^{14}C$ and $^{3}H$ were counted by the liquid scintillation counter (LSC) after the quench correction. To examine the recovery of $^{14}C$ using the radioactive standards, $Na_{2}^{14}CO_{3}$, $^{14}C-alcohol$, and $^{14}C-toluene$ as $^{14}C$, and HTO as $^{3}H$ were used. Also, the most suitable method for oxidizing $^{14}C-toluene$, which is difficult to be oxidized, was investigated through FT-IR spectra according to the concentration of sulfuric acid. With the identical method, $^{14}C$ and $^{3}H$ in the EB generated in the NPP were assayed in the range of $8.35{\sim}l.38{\times}10^3$ Bq/g and $2.46{\times}10^2{\sim}1.40{\times}10^4$ Bq/g, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.237-242
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• 2009

In the pyroprocessing of spent nuclear fuels, fuel materials are recovered by electrochemical reactions on the surface of electrodes as well as stirring the electrolyte in electrolytic cells such as electrorefiner, electroreducer and electrowinner. The system with this equipment should first be scaled-up in order to commercialize the pyroprocessing. So in this study, the scale-up for those electrolytic cells was studied to design a large-scale system which can be employed in a commercial process in the future. Basically the dimensions of both electrolytic cells and electrodes should be enlarged on the basis of the geometrical similarity. Then the criterion of constant power input per unit volume, characterizing the fluid behavior in the cells, was introduced in this study and a calculation process based on trial-and-error methode was derived, which makes it possible to seek a proper speed of agitation in the electrolytic cells. Consequently examples of scale-up for an arbitrary small scale system were shown when the criterion of constant power input per unit volume and another criterion of constant impeller tip speed were respectively applied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.229-236
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• 2009

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over $100^{\circ}C$ have been proposed. These new disposals have made it possible to introduce the concept of long tenn storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.