• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.51-62
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• 2020

This study was conducted to develop a method for depth assessment of embedded sources using gamma-spectrum ratio and for the evaluation of field applicability. To this end, Peak to Compton and Peak to valley ratio changes were evaluated according to ¹³⁷Cs, ⁶⁰Co, ¹⁵²Eu point source depth using HPGe detector and MCNP simulation. The effects of measurement distance of PTV and PTC methods were evaluated. Using the results, the source depth assessment equation using the PTC and PTV methods was derived based on the detection distance of 50 cm. In addition, the sensitivity of detection distance changes was assessed when using PTV and PTC methods, and error increased by 3 to 4 cm when detection distance decreased by 20 cm based on 50 cm. However, it was confirmed that if the detection distance was increased to 100 cm, the effects of detection distance were small. And PTV and PTC methods were compared with the two distance measurement method which evaluates the depth of source by the change of net peak counting rate according to the detection distance. As a result of source depth assessment, the PTV and PTC showed a maximum error of 1.87 cm and the two distance measurement method showed maximum error of 2.69 cm. The results of the experiment confirmed that the accuracy of the PTV and PTC methods was higher than two distance measurement. In addition, Sensitivity evaluation by horizontal position error of source has maximum error of less than 25.59 cm for the two distance measurement method. On the other hand, PTV and PTC method showed high accuracy with maximum error of less than 8.04 cm. In addition, the PTC method has lowest standard deviation for the same time measurement, which is expected to enable rapid measurement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.63-72
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• 2020

Dissolved hexavalent uranium can exist in the form of several different chemical species. Furthermore, species distributions depend on the pH value of the aqueous solution. Representatively, UO₂²⁺, UO₂OH⁺, (UO₂)₂(OH)₂²⁺, and (UO₂)₃(OH)₅⁺ species coexist in solutions at acidic and circumneutral pH values. When amorphous silica particles are suspended in an aqueous solution, the dissolved chemical species are easily adsorbed onto silica surfaces. In this study, it was examined whether the species distribution of the adsorbed U(VI) on a silica surface followed that of the dissolved U(VI) in an aqueous solution. Time-resolved luminescence spectra of three different dissolved species (UO₂²⁺, UO₂OH⁺, and (UO₂)₃(OH)₅⁺) and two different adsorbed species (≡SiO₂UO₂, ≡SiO₂(UO₂)OH^-, or ≡SiO₂(UO₂)₃(OH)^5-) were measured in the pH range 3.5-7.5. The spectral shapes of these chemical species were compared by changing the pH value; consequently, it was confirmed that the species distribution of the adsorbed U(VI) species was different from that of the dissolved U(VI) species.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.271-278
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• 2004

This paper shows our experiment is performed to understand the exposure tendency of $^{137}Cs$ according to the height of area and also, to supplement it by considering chemical characters of $^{137}Cs$ exposed to the soil. The samples we use for this experiment are from the general flat area of Yeonggwang county where it has NPPs, the high places of Keumjung & Bulgap mountains, and Naejan mountain where it is quite far from the NPPs. The data from this experiment show that the exposure of $^{137}Cs$ is not harmful since its range is around 252 Bq/kg-dry in most of sampled soils such as from the general flat area, the high place of Keumjung mountain where is 2 km away from the NPPs, the other high place of Bulgap mountain where is about 20 km away from the NPPs, and Naejan mountain where it is far from the NPPs. Not like the general flat area, however, the data show that the higher the area is the more $^{137}Cs$ is exposed. That is, at the top of mountains, the more $^{137}Cs$ is exposed compared to at the bottom area. It is almost $2{\~}6$ times more than the general flat area of Yeonggwang county where it has NPPs. The data also show that the spread of $^{137}Cs$ is deeply related to the geographical(the height of area, rainfall, etc..) factors and chemical factors of soils. As the geographical factors, there are far more chances to be exposed of $^{137}Cs$ at the high area of mountains through the air compared to at lower area and therefore, we can get more high-leveled readings of $^{137}Cs$ at the high area while it is low-leveled ones at the general flat area even if both of them have the same soil conditions. Regarding the chemical factors of soil, it is clarified that the CEC is the key factor. The CEC means the capability of sticking $^{137}Cs$ accumulated into the soil. Hence, the more CEC it has the more high-leveled readings of $^{137}Cs$ we get under the same geographical condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.9-17
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• 2010

The LCC (Liquid Cadmium Cathode) structure to be developed for inhibiting the formation and growth of the uranium dendrite has been known as a key part in the electrowinning process for the simultaneous recovering of uranium and TRU (TRans Uranium) elements from spent fuels. A zinc-gallium (Zn-Ga) experimental system which is able to be functional in aqueous condition and normal temperature has been set up to observe the formation and growth phenomena of the metal dendrites on liquid cathode. The growth of the zinc dendrites on the gallium cathode and the performance of the existing stirrer type and pounder type cathode structure were observed. Although the mechanical strength of the dendrites appeared to be weak in the electrolyte and easily crashed by the various cathode structures, it was difficult to effectively submerge the dendrite into the bottom of the liquid cathode. Based on the results of the aqueous phase experiments, a lab-scale electrowinning experimental apparatus which are applicable to the development of LCC srtucture for the electrowinning process was established and the performance tests of the different types of LCC structure were conducted to prohibit the uranium dendrite growth on LCC surface. The experimental results of the stirrer type LCC structures have shown that they could not effectively remove the uranium dendrites growing at the inner side of the LCC crucible and the performances of the paddle and harrow type LCC structure were similar. Therefore a mesh type LCC structure was developed to push down the uranium dendrites to the bottom of the LCC crucible growing on the LCC surface and at the inner side of the crucible. From the experimental results for the performance test of the mesh type LCC structure, the uranium was recovered over 5 wt% in cadmium without the growth of uranium dendrites. After completion of the experiments, solid precipitates of the bottom of the LCC crucible were identified as an intermetallic compound (UCd11) by the chemical analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.49-58
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• 2018

For the immobilization of high-radioactive nuclides such as Cs and Sr by high-temperature thermal decomposition, this study was carried out to investigate the phase transformation with calcined temperature by using TGA (thermogravimetric analysis) and XRD (X-ray diffraction) in the Cs-adsorbed CHA (chabazite zeolite of K type)-Cs and CHA-PCFC (potassium cobalt ferrocyanide)-Cs zeolite system, and Sr-adsorbed 4A-Sr and BaA-Sr zeolite system, respectively. In the case of CHA-Cs zeolite system, the structure of CHA-Cs remained at up to $900^{\circ}C$ and recrystallized to pollucite ($CsAlSi_2O_6$) at $1,100^{\circ}C$ after undergoing amorphous phase at $1,000^{\circ}C$. However, the CHA-CFC-Cs zeolite system retained the CHA-PCFC-Cs structure up to $700^{\circ}C$, but its structure collapsed in $900{\sim}1,000^{\circ}C$, and then transformed to amorphous phase, and recrystallized to pollucite at $1,100^{\circ}C$. In the case of 4A-Sr zeolite system, on the other hand, the structure of 4A-Sr maintained up to $700^{\circ}C$ and its phase transformed to amorphous at $800^{\circ}C$, and recrystallized to Sr-feldspar ($SrAl_2Si_2O_8$, hexagonal) at $900^{\circ}C$ and to $SrAl_2Si_2O_8$ (triclinic) at $1,100^{\circ}C$. However, the BaA-Sr zeolite system structure began to break down at below $500^{\circ}C$, and then transformed to amorphous phase in $500{\sim}900^{\circ}C$ and recrystallized to Ba/Sr-feldspar (coexistence of $Ba_{0.9}Sr_{0.1}Al_2Si_2O_8$ and $Ba_{0.5}Sr_{0.5}Al_2Si_2O_8$) at $1,100^{\circ}C$. All of the above zeolite systems recrystallized to mineral phase through the dehydration/(decomposition) ${\rightarrow}$ amorphous ${\rightarrow}$ recrystallization with increasing temperature. Although further study of the volatility and leachability of Cs and Sr in the high-temperature thermal decomposition process is required, Cs and Sr adsorbed in each zeolite system are mineralized as pollucite, Sr-feldspar and Ba/Sr-feldspar. Therefore, Cs and Sr seen to be able to completely immobilize in the calcining wasteform/(solidified wasteform).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.93-100
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• 2009

This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a $(NH_4)_2CO_3$ carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that $H_2O_2$ is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M $(NH_4)_2CO_3$-0.5 M $H_2O_2$ solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98${\pm}$2% for Re and Te, 94${\pm}$2% for Cs, and 29${\pm}$2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10${\sim}$20 minutes) due to their high solubility in the $(NH_4)_2CO_3$ solution regardless of the addition of $H_2O_2$, and independent of the concentrations of $Na_2CO_3$ and $H_2O_2$. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the $(NH_4)_2CO_3$ concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9${\sim}$10 in order to minimize the dissolution of FP.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.223-234
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• 2016

This study aimed to the adsorption-removal of high- radioactive iodide (I) contained in the initially generated high-radioactive seawater waste (HSW), with the use of AgX (Ag-impregnated X zeolite). Adsorption of I by AgX (hereafter denoted as AgX-I adsorption) was increased by increasing the Ag-impregnated concentration in AgX, and its concentration was suitable at about 30 wt%. Because of AgCl precipitation by chloride ions contained in seawater waste, the leaching yields of Ag from AgX (Ag-impregnated concentration : about 30~35 wt%) was less than those in distilled water (< 1 mg/L). AgX-I adsorption was above 99% in the initial iodide concentration ($C_i$) of 0.01~10 mg/L at m/V (ratio of weight of adsorbent to solution volume)=2.5 g/L. This shows that efficient removal of I is possible. AgX-I adsorption was found to be more effective in distilled water than in seawater waste, and the influence of solution temperature was insignificant. Ag-I adsorption was better described by a Freundlich isotherm rather than a Langmuir isotherm. AgX-I adsorption kinetics can be expressed by a pseudo-second order rate equation. The adsorption rate constants ($k_2$) decreased by increasing $C_i$, and conversely increased by increasing the ratio of m/V and the solution temperature. This time, the activation energy of AgX-I adsorption was about 6.3 kJ/mol. This suggests that AgX-I adsorption is dominated by physical adsorption with weaker bonds. The evaluation of thermodynamic parameters (a negative Gibbs free energy and a positive Enthalpy) indicates that AgX-I adsorption is a spontaneous reaction (forward reaction), and an endothermic reaction indicating that higher temperatures are favored.