• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.331-341
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• 2016

To evaluate the physicochemical and radiological properties of raw materials and by-products in domestic distribution, about 220 samples with 16 species were prepared. We measured the energy spectrum and the chemical content, such as U, Th, and K, using a $LaBr_3$ scintillation detector and ED-XRF. In addition, HPGe detector was used to analyze the radioac-tivity of $^{234}Th$, $^{234}mPa$, and $^{214}Bi$ in uranium decay series and $^{228}Ac$, $^{212}Pb$, and $^{208}Tl$ in thorium decay series, and $^{40}K$. The correlation between characteristic variables, such as the count rate in several ROIs, chemical content, and radioactivity, was assessed to infer the radioactivity of natural radionuclides through a rapid screening method. Based on the results, a characteristic database for raw material and by-product in domestic distribution was established and it will provide useful information in the analysis procedure and improve the accuracy and reproducibility in the analysis of natural radionuclides.

• Korean Journal of Microbiology
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• v.39 no.3
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• pp.175-180
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• 2003

Microbial Fe(III) reduction is an important factor for biogeochemical cycle in anaerobic environments, especially sediment of freshwater such as lakes, ponds and rivers. In addition, the Fe(III) reduction serves as a model for potential mechanisms for the oxidation of organic compounds and the reduction of toxic heavy metals, such as chrome or uranium. Shewanella putrefaciens DK-1 was a gram-negative, facultative anaerobic Fe(III) reducer and used ferric ion as a terminal electron acceptor for the oxidation of organic compounds to $CO_{2}$ or other oxidized metabolites. The ability of reducing activity and utilization of various electron acceptors and donors for S. putrefaciens DK-1 were investigated. S. putrefaciens DK-1 was capable of using a wide variety of electron acceptor, including $NO_{3}^{-}$, Fe(III), AQDS, and Mn(IV). However, its ability to utilize electron donors was limited. Lactate and formate were used as electron donors but acetate and toluene were not used. Fe(III) reduction of S. putrefaciens DK-l was inhibited by the presence of either $NO_{3}^{-}$ or $NO_{2}^{-}$. Further S. putrefaciens DK-1 used humic acid as an electron acceptor and humic acid was re-oxidized by nitrate. Environmental samples showing the Fe(III)-reducing activity were used to investigate effects of the limiting factors such as carbon, nitrogen and phosphorus on the Fe(III) reducing bacteria. The highest Fe (III) reducing activity was measured, when lactate as a carbon source and S. putrefaciens DK-1 as an Fe(III) reducer added in untreated sediment samples of Cheon-ho and Dae-ho reservoirs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.59-64
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• 2018

In case of high contents of rare earths in the LiCl-KCl salt, it is not easy to recover U and TRU metals as a usable resource form from LiCl-KCl eutectic salts generated from the pyroprocessing of spent nuclear fuel. In this study, a conversion of $UCl_3$ into an oxide form using $K_2CO_3$ and an electrodeposition of $NdCl_3$ into a metal form in $LiCl-KCl-UCl_3-NdCl_3$ system were conducted to resolve the problem. Before conducting the conversion, experimental conditions for the conversion were determined by performing a thermodynamic equilibrium calculation. In this study, almost all of $UCl_3$ disappeared in the LiCl-KCl salt when the injection of $K_2CO_3$ reached theoretical equivalent for the conversion, and then $NdCl_3$ was effectively electrodeposited as a metal form using liquid zinc cathode. After that, the LiCl-KCl salt became transparent, and uranium oxides were precipitated to the bottom of the LiCl-KCl salt. These results will be utilized in designing a process to separate U and rare earths in LiCl-KCl salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.83-91
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• 2018

While using an electrokinetic method to analyze the characteristics of cement solidification of radioactive wastes from decontaminated uranium soil and concrete, the compressive strength, pH, electrical conductivity, irradiation effects, and volume expansion were measured for the solidified cement specimens. The workability of cement solidified from radioactive waste was about 170-190%. After the solidified cement was irradiated, the compressive strength decreased by about 15%, but met the criteria ($34kgf{\cdot}cm^{-2}$) of KORAD (Korea Radioactive Waste Agent). According to the results of SEM-EDS for solidified cement, the aluminum phase was well combined with cement, while the calcium phase was separated from cement. The volume of solidified cement in radioactive wastes was dependent on the waste-to-cement ratio and the amount of water, and increased by about 30% under the conditions used in this study. Therefore, it was concluded that permanent disposal of electrokinetically decontaminated radioactive wastes is appropriate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.125-131
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• 2009

An electrolytic system (zinc anode-gallium cathode) was setup to evaluate the performance of several stirrers prepared for this study, where stirrers have been used to prevent uranium from forming dendrite on the cathode in pyrochemical process. In the case of no-stirring condition, zinc dendrites began to grow on the gallium surface in 1 hour and some dendrite grew out of the cathode crucible around 6 hours. When a rectangular stirrer or a tilt stirrer was rotated, at 40${\sim}$150 rpm, to mix the liquid gallium cathode, dendritic growth of zinc metal was prevented irrespective of revolution speed, but some of the deposits overflowed out of the cathode crucible owing to the large centrifugal forces at 150 rpm. The harrow stirrer did not nearly retard the dendrite growth at 40 rpm, but the dendrite growth was retarded at higher than 100 rpm and the zinc deposits also did not overflow at 150 rpm. Pounder could also prevent the dendrite growth to some extent but it had some difficulties in operation compared with other types of stirrers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.109-114
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• 2009

Absorption spectra for a quantitative analysis of actinide elements such as U(VI) and Pu(V) were measured by using a liquid waveguide capillary cell (LWCC) which has an optical path length of 1.0 meter. In order to investigate radioactive elements, a LWCC is installed in a glove box and is coupled to a spectrophotometer with optical fibers. Limits of detection (LOD) for the system were determined as 0.74 and 0.35 M with molar absorption coefficients of 8.14${\pm}$0.07 (414 nm) and 17.00${\pm}$0.16 (569 nm) $M^{-1}cm^{-1}$ for U(VI) and Pu(V) ions, respectively. The measured LOD values are about 30 times more sensitive when compared to those achievable by using a conventional quartz cell with an optical path length of 1.0 cm. As an application with an enhanced sensitivity, a quantitative analysis for micromolar concentrations of Pu(V) has been performed to decrease the uncertainty in the formation constant of the Pu(VI)-OH complex.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.211-221
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• 2016

Waste salt generated from a pyro-processing for the recovery of uranium and transuranic elements has high volatility at vitrification temperature and low compatibility in conventional waste glasses. For this reason, KAERI (Korea Atomic Energy Research Institute) suggested a new method to de-chlorinate waste salt by using an inorganic composite named SAP ($SiO_2-Al_2O_3-P_2O_5$). In this study, the de-chlorination behavior of waste salt and the microstructure of consolidated form were examined by adding $B_2O_3$ and $Fe_2O_3$ to the original SAP composition. De-chlorination behavior of metal chloride waste was slightly changed with given compositions, compared with that of original SAP. In the consolidated forms, the phase separation between Si-rich phase and P-rich phase decreases with the amount of $Al_2O_3$ or $B_2O_3$ as a connecting agent between Si and P-rich phase. The results of PCT (Product Consistency Test) indicated that the leach-resistance of consolidated forms out of reference composition was lowered, even though the leach-resistance was higher than that of EA (Environmental Assessment) glass. From these results, it could be inferred that the change in the content of Al or B in U-SAP affected the microstructure and leach-resistance of consolidated form. Further studies related with correlation between composition and characteristics of wasteform are required for a better understanding.

• 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.8 no.4
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• pp.347-353
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• 2010

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be $1.40{\times}10^{15}$ Bequerels, 236 Watts, $4.34{\times}10^9m^3$-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20~45 % and 30~45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

• 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.