• Journal of Korean Society of Environmental Engineers
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• v.37 no.8
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• pp.465-471
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• 2015

Adsorption experiments for radionuclides such as $^3H$, $^{90}Sr$ and $^{99}Tc$ were conducted using fractured rock collected in unsaturated zone. The released radionuclide through artificial barrier from the near surface repository can be transported by the flow of rainfall or pore water through fractures in unsaturated zone and reach to groundwater flow. Therefore, it is important to investigate transport behavior (retardation) of radionuclides through fractured rock for the safety assessment and long-term performance of repository. Fractured rock samples were collected and characterized by X-ray microtomography (XMT) analysis, which can be used to develop a more robust unsaturated fracture transport model. When fracture-filling materials are exist, distribution coefficient of $^{90}Sr$ is higher than without fracture-filling materials. In this study, batch sorption distribution coefficient ($K_d$) of radionuclide was determined and used to increase our understanding of radionuclide retardtion through fracture-filling materials.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.95-96
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• 2009

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.401-402
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• 2012

• Resources Recycling
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• v.33 no.2
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• pp.54-61
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• 2024

When steel contaminated with Pb, produced by the decay of natural radionuclides, is remelted, Pb distributes among the metal, slag, and gas phases. In this study, 5 wt%Pb was added to Fe and melted with CaO-SiO₂-Al₂O₃-MgO slag to investigate Pb's distribution in the metal/slag/gas. As slag basicity ((wt%CaO)/(wt%SiO₂)) increased, Pb solubility in Fe slightly increased, while Pb in the slag tended to decrease. Consequently, the slag/metal distribution ratio of Pb decreased with increasing basicity. Thermodynamic calculations revealed that the slag/Fe phase distribution ratio of Pb remained very low irrespective of the activity coefficient of PbO in the slag, consistent with the experimental results. The calculated evaporation rate of Pb in Fe-Pb was approximately 22 times that of Fe; hence, most of the Pb evaporated into the gas phase.

• Korean Chemical Engineering Research
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• v.45 no.6
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• pp.627-632
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• 2007

A fundamental study on the melt decontamination of metal wastes generated by dismantling the nuclear facility, the melting of metal wastes such as stainless steel and carbon steel have been carried out to investigate the distribution phenomena of the radioisotopes such as $^{60}Co$ and $^{137}Cs$ into the ingot, slag and dust phases by using the various slag types, slag concentration and basicity in an arc furnace. The $^{60}Co$ remained homogeneously in the ingot phase above 90 % and it was barely present in the slag below 10 %. The effect of the slag composition on the distribution for Co-60 was not considerable, but a basic slag former with high fluidity showed effective. $^{137}Cs$ was completely eliminated from the melt of the stainless steel as well as the carbon steel and distributed to the slag and the dust phase.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.95-104
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• 2005

Effects of the aluminum melting temperature, melting time and a kind of flux agents on the distribution of surrogate nuclide were investigated in the electric furnace at the aluminum melting including surrogate radionuclides(Co, Cs, Sr) in order to establish the fundamental research of the melting technology for the metallic wastes from the decommissioning of the TRIGA research reactor. It was verified that the fluidity of aluminum melt was increased by adding flux agent but it was slightly varied according to the sort of flux agents. The results of the XRD analysis showed that the surrogate nuclides move into the slag phase and then they were combined with aluminum oxide to form more stable compound. The weight of the slag generated from aluminum melting test increased with increasing melting temperature and melting time and the increase rate of the slag depended on the kind of flux agents added in the aluminum waste. The concentration of the cobalt in the ingot phase decreased with increasing reaction temperature but it increased in the slag phase up to 90$\%$according to the experimental conditions. The volatile nuclides such as Cs and Sr considerably transferred from the ingot phase to the slag and dust phase.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.281-291
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• 2013

A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.129-139
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• 2008

In this study, a sorption experiment of multivalent nuclides such as Eu(III) and Th(IV) relatively stable for redox reactions was carried out for bentonite colloids which had been prepared from the domestic Gyeongju bentonite. The amounts of the nuclides lost by an attachment to bottle walls, by a precipitation, and by a colloid formation were estimated by performing blank tests for the sorption experiments. Sorption coefficients, $K_d's$, reflecting the mass losses were obtained and investigated for the sorption of Eu(III) and Th(IV) onto the bentonite colloids. The net sorption coefficients $K_d's$ considering all the three mass losses were measured as about $10^6-10^7\;mL/g$ and $7{\times}10^6-10^7\;mL/g$ for Eu(III) and Th(IV), respectively, depending on pH. In particular, a precipitation occurred mainly at a pH greater than 5 for Eu(III) and a precipitation and colloid formation significantly occurred at a pH greater than 3 for Th(IV). The precipitation and colloid formation of the multivalent nuclides of Eu(III) and Th(IV) therefore should be considered when $K_d's$ are rightly obtained over the pH range where their precipitation and colloid formation become significant at a given concentration.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.507-515
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• 1994

Fly ash and betonite samples were selected and characteristics of them were investigated. Fly ash was found to be similar to bentonite in particle size distribution but quite different in microstructure. The most special aspect of fly ash was high alkalinity of its solution. Distribution coefficients of Cs and Co on the samples were measured to survey the effects of mixing. Fly ash showed higher distribution coefficient of Co than that of Cs. Through various experiments, factors affecting the distribution coefficients of Co and Cs on mixture of bentonite and fly ash were identified. Comparison of the distribution coefficients of Cs on fly ash and bentonite mixture with those on sand and bentonite mixture suggests that fly ash would be useful as an efficient additive of backfill material if pertinent mixing ratio was chosen.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.79-86
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• 2009

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.