• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.87-96
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• 2004

Studies were conducted to select the candidate buffer material for a high-level waste (HLW) repository in Korea. This paper presents the hydraulic properties, the swelling properties, the thermal properties, and the mechanical properties as well as the radionuclide release-retarding capacity of Kyungju bentonite as part of those studies. Experimental results showed that the hydraulic conductivities of the compacted bentonite were very low and less than $10^{-11}$m/s. The values decreased with increasing the dry density of the compacted bentonite. The swelling pressures were in the range of 0.66 MPa to 14.4 ㎫ and they increased with increasing the dry density. The thermal conductivities were in the range of 0.80 ㎉/m $h^{\circ}C$ to 1.52 ㎉/m $h^{\circ}C$. The unconfined compressive strength, Young's modulus and Poison's ratio showed the range of 0.55 ㎫ to 8.83 ㎫, 59 ㎫ to 1275 ㎫, and 0.05 to 0.20, respectively, when the dry densities of the compacted bentonite were 1.4 Ms/㎥ to 1.8 Mg/㎥. The diffusion coefficients in the compacted bentonite were measured under an oxidizing condition. The values were $1.7{\times}10^{-10}$m^2$/s to 3.4{\times}10^{-10}$m^2$/s for electrically neutral tritium (H-3), 8.6{\times}10^{-14}$m^2$/s to 1.3{\times}10^{-12}$m^2$/s for cations (Cs, Sr, Ni), 1.2{\times}10^{-11}$m^2$/s to 9.5{\times}10^{-11}$m^2$/s for anions (I, Tc), and 3.0{\times}10^{-14} $m^2$/s to 1.8{\times}10^{-13}$m^2$/s $for actinides (U, Am), when tile dry densities were in the range of 1.2 Mg/㎥ to 1.8 Mg/㎥. The obtained results will be used in assessing the barrier properties of Kyungju bentonite as a buffer material of a repository in Korea.n Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.189-197
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• 2012

Geochemical composition of fracture filling minerals and groundwater was investigated to characterize geochemical characteristics of groundwater system at the KURT site. Minerals such as calcite, illite, laumontite, chlorite, epidote, montmorillonite, and kaolinite, as well as I/S mixed layer minerals were detected in the minerals extracted from the fracture surfaces of the core samples. The groundwater from the DB-1, YS-1 and YS-4 boreholes showed alkaline conditions with pH of higher than 8. The electrical conductivity (EC) values of the groundwater samples were around $200{\mu}S/cm$, except for the YS-1 borehole. Dissolved oxygen was almost zero in the DB-1 borehole indicating highly reduced conditions. The Cl- concentration was estimated around 5 mg/L and showed homogeneous distribution along depths at the KURT site. It might indicate the mixing between shallow groundwater and deep groundwater. The shallow groundwater from boreholes showed $Ca-HCO_3$ type, whereas deep groundwater below 300 m from the surface indicated $Na-HCO_3$ type. The isotopic values observed in the groundwater ranged from -10.4 to -8.2‰ for ${\delta}^{18}O$ and from -71.3 to -55.0‰for ${\delta}D$. In addition, the isotope-depleted water contained higher fluoride concentration. The oxygen and hydrogen isotopic values of deep groundwater were more depleted compared to the shallow groundwater. The results from age dating analysis using $^{14}C$ indicated relatively younger (2000~6000yr old) groundwater compared to other european granitic groundwaters such as Stripa (Sweden).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.45-53
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• 2012

The metal chloride wastes from a pyrochemical process to recover uranium and transuranic elements has been considered as a problematic waste difficult to apply to a conventional solidification method due to the high volatility and low compatibility with silicate glass. In this study, a dechlorination approach to treat LiCl-KCl waste for final disposal was adapted. In this study, a $SiO_2-Al_2O_3-P_2O_5$ (SAP) inorganic composite as a dechlorination agent was prepared by a conventional sol-gel process. By using a series of SAPs, the dechlorination behavior and consolidation of reaction products were investigated. Different from LiCl waste, the dechlorination reaction occurred mainly at two temperature ranges. The thermogravimetric test indicated that the first reaction range was about $400^{\circ}C$ for LiCl and the second was about $700^{\circ}C$ for KCl. The SAP 1071 (Si/Al/P=1/0.75/1 in molar) was found to be the most favorable SAP as a dechlorination agent under given conditions. The consolidation test revealed that the bulk shape and the densification of consolidated forms depended on the SAP/Salt ratios. The leaching test by PCT-A method was performed to evaluate the durability of consolidated forms. This study provided the basic information on the dechlorination approach. Based on the experimental results, the dechlorination method using a $SiO_2-Al_2O_3-P_2O_5$ (SAP) could be considered as one of alternatives for the immobilization of waste salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.27-36
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• 2012

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP ($SiO_2-Al_2O_3-P_2O_5$) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of $Fe_2O_3$ into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP( Fe=0.1). The experimental results indicated that the addition of $Fe_2O_3$ increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing $B_2O_3$ into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP ($SiO_2-Al_2O_3-Fe_2O_3-P_2O_5-B_2O_3$) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3~4 g of wasteform for final disposal. The final volume would be about 3~4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.281-290
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• 2018

In order to understand the long-term behavior of radionuclides in granite environments, geochemical behavior characteristics of uranium in granitic host rock of KURT (KAERI Underground Research Tunnel) were investigated by dissolution experiment with different reaction time and solutions. In the dissolution experiment, significantly increased dissolution levels of uranium from granite powder samples were identified during the reaction time of 0~10 days for reaction solutions ($UD-CO_3$ and UD-Bg) containing a large amount of $CO_3{^{2-}}$. On the other hand, significantly increased dissolution levels of uranium were also identified for reaction solutions containing Na and Ca after 60 days. Dissolution of uranium continuously increased in reaction solutions of $UD-CO_3$ ($44.61{\mu}g{\cdot}L^{-1}$), UD-Bg ($41.01{\mu}g{\cdot}L^{-1}$), UD-Na ($26.87{\mu}g{\cdot}L^{-1}$), UD-Ca ($20.26{\mu}g{\cdot}L^{-1}$), UD-CaSi ($17.03{\mu}g{\cdot}L^{-1}$), and UD-Si ($10.47{\mu}g{\cdot}L^{-1}$) in the experimental period of ~270 days. However, after day 270, dissolution of uranium showed a decreasing tendency. This is thought to have occurred because existing uranium in granite samples reached the limit of dissolution by interaction with reaction solutions. Concentrations of dissolved uranium and points of maximum concentration value were found to differ depending on the $CO_3{^{2-}}$ presence in the mixed reaction solution and on the geochemical type of the water. It is estimated that differences in the reaction rate between the granite sample and the reaction solution are due to the influence of dissolved ions in the reaction solution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.117-131
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• 2006

A coupled T-H-M(Thermo-Hydro-Mechanical) analysis was carried out for KENTEX (KAERI Engineering-scale T-H-M Experiment for Engineered Barrier System), which is a facility for validating the coupled T-H-M behavior in the engineered barrier system of the Korean reference HLW(high-level waste) disposal system. The changes of temperature, water saturation, and stress were estimated based on the coupled T-H-M analysis, and the influence of the types of mechanical constitutive material laws was investigated by using elastic model, poroelastic model, and poroelastic-plastic model. The analysis was done using ABAQUS, which is a commercial finite element code for general purposes. From the analysis, it was observed that the temperature in the bentonite increased sharply for a couple of days after heating the heater and then slowly increased to a constant value. The temperatures at all locations were nearly at a steady state after about 37.5 days. In the steady state, the temperature was maintained at $90^{\circ}C$ at the interface between the heater and the bentonite and at about $70^{\circ}C$ at the interface between the bentonite and the confining cylinder. The variation of the water saturation with time in bentonite was almost same independent of the material laws used in the coupled T-H-M processes. By comparing the saturation change of T-H-M and that of H-M(Hydro-Mechanical) processes using elastic and poroelastic material mod31 respectively, it was found that the degree of saturation near the heater from T-H-M calculation was higher than that from the coupled H-M calculation mainly because of the thermal flux, which seemed to speed up the saturation. The stresses in three cases with different material laws were increased with time. By comparing the stress change in H-M calculation using poroelasetic and poroelasetic-plastic model, it was possible to conclude that the influence of saturation on the stress change is higher than the influence of temperature. It is, therefore, recommended to use a material law, which can model the elastic-plastic behavior of buffer, since the coupled T-H-M processes in buffer is affected by the variation of void ratio, thermal expansion, as well as swelling pressure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.207-217
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• 2011

In this study the complex formation reactions between uranium(VI) and 2,6-dihydroxybenzoate (DHB) as a model ligand of humic acid were investigated by using UV-Vis spectrophotometry and time-resolved laser-induced fluorescence spectroscopy (TRLFS). The analysis of the spectrophotometric data, i.e., absorbance changes at the characteristic charge-transfer bands of the U(VI)-DHB complex, indicates that both 1:1 and 1:2 (U(VI):DHB) complexes occur as a result of dual equilibria and their distribution varies in a pH-dependent manner. The stepwise stability constants determined (log $K_1$ and log $K_2$) are $12.4{\pm}0.1$ and $11.4{\pm}0.1$. Further, the TRLFS study shows that DHB plays a role as a fluorescence quencher of U(VI) species. The presence of both a dynamic and static quenching process was identified for all U(VI) species examined, i.e., ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$, and $(UO_2)_3{(OH)_5}^+$. The fluorescence intensity and lifetimes of each species were measured from the time-resolved spectra at various ligand concentrations, and then analyzed based on Stern-Volmer equations. The static quenching constants (log $K_s$) obtained are $4.2{\pm}0.1$, $4.3{\pm}0.1$, and $4.34{\pm}0.08$ for ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$, and $(UO_2)_3{(OH)_5}^+$, respectively. The results of Stern-Volmer analysis suggest that both mono- and bi-dentate U(VI)-DHB complexes serve as groundstate complexes inducing static quenching.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.357-365
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• 2016

We observed the ${\gamma}-ray$ shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of $0.371cm^{-1}$ from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a ${\gamma}-ray$ of $^{137}Cs$, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of $3,175kg{\cdot}m^{-3}$. Although the unit weight of the concrete with OSS and OSG was $3,052kg{\cdot}m^{-3}$, which was lower than the maximum unit weight condition by $123kg{\cdot}m^{-3}$, its attenuation coefficient was improved by $0.012cm^{-1}$. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced ${\gamma}-ray$ shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.101-112
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• 2016

This study investigated the removal of Sr, which was one of the high radioactive nuclides, by adsorption with Barium (Ba) impregnated 4A zeolite (BaA) from high-radioactive seawater waste (HSW). Adsorption of Sr by BaA (BaA-Sr), in the impregnated Ba concentration of above 20.2wt%, was decreased by increasing the impregnated Ba concentration, and the impregnated Ba concentration was suitable at 20.2wt%. The BaA-Sr adsorption was added to the co-precipitation of Sr with $BaSO_4$ precipitation in the adsorption of Sr by 4A (4A-Sr) within BaA. Thus, it was possible to remove Sr more than 99% at m/V (adsorbent weight/solution volume)=5 g/L for BaA and m/V >20 g/L for 4A, respectively, in the Sr concentration of less than 0.2 mg/L (actual concentration level of Sr in HSW). It shows that BaA-Sr adsorption is better than 4A-Sr adsorption in for the removal capacity of Sr per unit gram of adsorbent, and the reduction of the secondary solid waste generation (spent adsorbent etc.). Also, BaA-Sr adsorption was more excellent removal capacity of Sr in the seawater waste than distilled water. Therefore, it seems to be effective for the direct removal of Sr from HSW. On the other hand, the adsorption of Cs by BaA (BaA-Cs) was mainly performed by 4A within BaA. Accordingly, it seems to be little effect of impregnated Ba into BaA. Meanwhile, BaA-Sr adsorption kinetics could be expressed the pseudo-second order rate equation. By increasing the initial Sr concentrations and the ratios of V/m, the adsorption rate constants ($k_2$) were decreased, but the equilibrium adsorption capacities ($q_e$) were increasing. However, with increasing the temperature of solution, $k_2$ was conversely increased, and $q_e$ was decreased. The activation energy of BaA-Sr adsorption was 38 kJ/mol. Thus, the chemical adsorption seems to be dominant rather than physical adsorption, although it is not a chemisorption with strong bonding form.