• Journal of Nuclear Fuel Cycle and Waste Technology
• /
• v.1 no.1
• /
• pp.49-55
• /
• 2013

To characterize quantitatively the transport of $^{99}Tc$ and the global fallout ($^{137}Cs$, $^{90}Sr$, and $^{239+240}Pu$) for soils in Korea, the transport parameters of a convective-dispersion model, apparent migration velocity, and apparent dispersion coefficient were estimated from the vertical depth profiles of the radionuclides in soils. The vertical profiles of $^{99}Tc$ were measured from a pot experiment for paddy soil that had been sampled from a rice-field around the Gyeongju radioactive waste repository in Korea, and the vertical depth distributions of the global fallout $^{137}Cs$, $^{90}Sr$, and $^{239+240}Pu$ were measured from the soil samples that were taken from local areas in Korea. The front edge of the $^{99}Tc$ profiles reached a depth of about 12 cm in 138 days, indicating a faster movement than the fallout radionuclides. A weak adsorption of $^{99}Tc$ on the soil particles by the formation of Tc(VII) and a high water infiltration velocity seemed to have controlled the migration of $^{99}Tc$. The apparent migration velocity and dispersion coefficient of $^{99}Tc$ for the disturbed paddy soil were 2.88 cm/y and 6.3 $cm^2/y$, respectively. The majority of the global fallout $^{137}Cs$, $^{90}Sr$, and $^{239+240}Pu$ were found in the top 20 cm of the soils even after a transport of about 30 years. The transport parameters for the global fallout radionuclides were 0.01-0.1cm/y ($^{137}Cs$), 0.09-0.13cm/y ($^{90}Sr$), and 0.09-0.18cm/y ($^{239+240}Pu$) for the apparent migration velocity: 0.21-1.09 $cm^2/y$ ($^{137}Cs$), 0.12-0.7$cm^2/y$ ($^{90}Sr$), and 0.09-0.36$cm^2/y$ ($^{239+240}Pu$) for the apparent dispersion coefficient.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.3
• /
• pp.207-218
• /
• 2017

In this study, pitch-based activated carbon fibers (ACFs) were modified with pyrolysis fuel oil (PFO). Carbonized ACF samples were activated at $850^{\circ}C$, $880^{\circ}C$ and $900^{\circ}C$. A scanning electron microscope (SEM) and a BET surface area apparatus were employed to evaluate the indoor radon removal of each sample. Among three samples, the BET surface area and micropore area of ACF880 recorded the highest value with $1,420m^2{\cdot}g^{-1}$ and $1,270m^2{\cdot}g^{-1}$. Moreover, ACF880 had the lowest external surface area and BJH adsorption cumulative surface area of pores with $151m^2{\cdot}g^{-1}$ and $35.5m^2{\cdot}g^{-1}$. This indicates that satisfactory surface area depends on the appropriate temperature. With the above scope, ACF880 also achieved the highest radon absorption rate and speed in comparison to other samples. Therefore, we suggest that the optimum activation temperature for PFO containing ACFs is $880^{\circ}C$ for effective indoor radon adsorption.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.3
• /
• pp.181-188
• /
• 2010

Metal surface decontamination characteristics were investigated by using a laser ablation method. A second harmonic generation of a Q-switched Nd:YAG laser with a wave length of 532 nm, a pulse energy of 150 mJ and a pulse width of 5 ns was employed to assess the decontamination performance for metal surfaces contaminated with $CsNO_3$, $Co(NH_4)_2(SO_4)_2$, $Eu_2O_3$ and $CeO_2$. The ablation behavior was investigated for the decontamination variables such as a number of laser shots, laser fluence and an irradiation angle. Their optimum values were found to be 8, 13.3 J/$cm^2$ and $30^{\circ}$, respectively. The decontamination efficiency was different depending on the kinds of the contaminated ions, due to their different melting and boiling points and was in the order: $CsNO_3>Co(NH_4)_2(SO_4)_2>Eu_2O_3>CeO_2$. We also evaluated a correlation between the metal ablation thickness and the number of laser shots for the different laser fluences.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.2
• /
• pp.99-105
• /
• 2010

This paper investigated the characteristics of wasteform containing a spent zeolite used as a separating agent of FPs for recycling LiCl waste which would be generated from pyrochemical process of spent PWR fuel. In this study, a conventional borosilicate and Ca-rich glass were used as a consolidating agent for spent zeolite and it's mixing ratio was changed in the range, $25{\sim}35wt%$. The leach rates of Cs and Sr had about $0.1{\sim}0.01g/m^2day$ and $0.001{\sim}0.0001g/m^2day$, respectively. The leach resistance of Cs increased with amount of SAP and it showed about 10 times higher in the Ca-rich glass wasteform than in the conventional borosilciate glass wasteform. The compressive strength of wasteform was affected with the amount of glass. Thermal expansion rate of containing 30wt% glass has relatively lower than others. Also, the melting temperature was little changed in given mixing ratio of glass.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.9 no.2
• /
• pp.81-86
• /
• 2011

Polymer-modified cement is the composite material made by partially replacing and strengthening the cement hydrate binders of conventional mortar with polymeric modifiers such as polymer latexes and redispersible polymeric modifiers. It is known that the addition of polymer to cement mortar leads to improved quality, which would be expected to have a high chemical resistance. Therefore, the purpose of this study is to identify the improved chemical resistance, such as low permeability and low ion diffusivity, of the polymer-modified cement as a solidification agent for the radwaste. First, polymer-modified cement specimens by latex modification were prepared according to the polymer content from 0% to 30% to select the optimized polymer content. At those specimens, the water-to-cement (W/C) ratio was maintained to 33% and 50% respectively. After the much curing time, the structural integrity of specimens was evaluated through the compressive strength test and the porosity evaluation by the water immersion method. From the results, 10% of the polymer content at 33% of the W/C ratio was shown to have the most improved quality. Finally, the leaching test referredfrom ANS 16.1 for the specimens having the most improved quality was conducted. Dedicated specimens for the leaching test were then mixed with radioisotopes of $^{60}Co$ and $^{137}Cs$ at the specimen preparation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.2
• /
• pp.101-109
• /
• 2008

For the Kyungju bentonite which is considered as a candidate material for the buffer and backfill in the high-level waste repository, the thermal conductivities of compacted bentonite and a bentonite-sand mixture were measured. The thermal conductivities of the compacted bentonites with a dry density of 1.2 to $1.8\;Mg/m^3$ and the bentonite-sand mixture with a dry density of 1.6 and $1.8\;Mg/m^3$ were measured within the gravimetric water content range of 10wt% to 20wt% and the sand fraction range of 10 to 30wt%. The thermal conductivity of compacted bentonite and a bentonite-sand mixture increases with increasing dry density and sand weight fraction in the case of constant water weight fraction, and increases with increasing water weight fraction and sand weight fraction in the case of constant dry density. The empirical correlations to describe the thermal conductivity of compacted bentonite and a bentonite-sand mixture as a function of water fraction at each dry density were suggested. These correlations can predict the thermal conductivities of bentonite and a bentonite-sand mixture with a difference below 10%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.3
• /
• pp.253-266
• /
• 2016

In accordance with the classification system of radioactive waste in Korea, all the disused sealed radioactive sources (DSRSs) fall under the category of EW, VLLW or LILW, and should be managed in compliance with the restrictions for the disposal method. In this study, the management and disposal method are drawn in consideration of half-life of radionuclides contained in the source and A/D value (i.e. the activity A of the source dividing by the D value for the relevant radionuclide, which is used to provide an initial ranking of relative risk for sources) in addition to the domestic classification scheme and disposal method, based on the characteristic analysis and review results of the management practices in IAEA and foreign countries. For all the DSRSs that are being stored (as of March 2015) in the centralized temporary disposal facility for radioisotope wastes, applicability of the derivation result is confirmed through performing the characteristic analysis and case studies for assessing quantity and volume of DSRSs to be managed by each method. However, the methodology derived from this study is not applicable to the following sources; i) DSRSs without information on the radioactivity, ii) DSRSs that are not possible to calculate the specific activity and/or the source-specific A/D value. Accordingly, it is essential to identify the inherent characteristics for each of DSRSs prior to implementation of this management and disposal method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.3
• /
• pp.229-243
• /
• 2013

An in situ solute migration system was designed and installed in KAERI Underground Research Tunnel (KURT) constructed in the site of Korea Atomic Energy Research Institute (KAERI) in order to investigate the migration and retardation of non-sorbing and sorbing tracers through a rock fracture. The system is composed of three main parts including injection, extraction, and data treatment. For the selection of a water-conducting fracture, boreholes were drilled. The fractures in the drilled boreholes were investigated using borehole image analysis using borehole image processing system (BIPS). The results of BIPS analysis showed that borehole YH 3-1 and YH 3-2 were connected each other. Moreover, hydraulic tests were carried out to determine the test section with connectivity for the in situ experiments. The in situ solute migration experiments were accomplished to understand the migration of solutes through fractures in KURT using non-sorbing tracers which were fluorescein sodium, eosin-B, bromide and sorbing tracers which were rubidium, nickel, zirconium, and samarium.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.4
• /
• pp.421-429
• /
• 2018

Hydrazine based reductive dissolution applied on magnetite oxide was investigated. Dissolution of Fe(II) and Fe(III) from magnetite takes place either by protonation, surface complexation, or reduction. Solution containing hydrazine and sulfuric acid provides hydrogen to break bonds between Fe and oxygen by protonation and electrons for the reduction of insoluble Fe(III) to soluble Fe(II) in acidic solution of pH 3. In terms of dissolution rate, numerous transition metal ions were examined and Cu(II) ion was found to be the most effective to speed up the dissolution. During the cycle of Cu(I) ions to Cu(II) ions, the released electron promoted the reduction of Fe(III) and Cu(II) ions returned to Cu(I) ion due to the oxidation of hydrazine. In the experimental results, the addition of a very low amount of cupric ion (about 0.5 mM) to the solution increased the dissolution rate about 40% on average and up to 70% for certain specific conditions. It is confirmed that even though the coordination structure of copper ions with hydrazine is not clear, the $Cu(II)/H^+/N_2H_4$ system is acceptable regarding the dissolution performance as a decontamination reagent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.3
• /
• pp.329-338
• /
• 2019

Suction of unsaturated bentonite buffers is a very important input parameter for hydro-mechanical performance assessment and design of an engineered barrier system. This study analyzed suction measurement techniques and constitutive models of unsaturated porous media reported in the literature, and suggested suction measurement techniques and constitutive models suitable for bentonite buffer in an HLW repository. The literature review showed the suction of bentonite buffer to be much higher than that of soil, as measured by total suction including matric suction and osmotic suction. The measurement methods (RH-Cell, RH-Cell/Sensor) using a relative humidity sensor were suitable for suction measurement of the bentonite buffer; the RH-Cell /Sensor method was more preferred in consideration of the temperature change due to radioactive decay heat and measurement time. Various water retention models of bentonite buffers have been proposed through experiments, but the van Genuchten model is mainly used as a constitutive model of hydro-mechanical performance assessment of unsaturated buffers. The water characteristic curve of bentonite buffers showed different tendencies according to bentonite type, dry density, temperature, salinity, sample state and hysteresis. Selection of water retention models and determination of model input parameters should consider the effects of these controlling factors so as to improve overall reliability.