• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.137-144
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• 2005

Silica in primary system combines with an alkali grammatical particle metal and forms the zeolite layer which is hindering the heat transfer on the surface of the cladding. Zeolite layer becomes the cause of the damage in this way. The problems of the NPP's primary system have been issued steadily by EPRI. Through a series of experiments of the laboratory scale, we confirmed the applicability of NF membrane for silica removal, as silica rejection rate of NF membrane is about $60\;{\sim}\;70\%$ and boron rejection rate is about $10\;{\sim}\;20\%$. We accomplished a site experiment about four NF membranes manufactured by FilmTec and Osmonics Inc. In experiment using 400L of SFP water, when operation pressure is $10kg_{f}/cm^2$, we confirmed that the silica rejection rate of NF90-2540 manufactured by FilmTec Inc. is about $98\%$, boron rejection rate is about $43\%$. The silica rejection rate of NF270-2540 is about $38\%$, boron rejection rate is about $3.5\%$. Afterward, through additional experiments, such as long term characteristic experiments, we are going to design a optimum NF membrane system for silica removal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.179-189
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• 2014

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. An electrolytic reduction of the pyroprocessing is a process to reduce oxides into metals using LiCl as an electrolyte and requires a post-treatment process due to the inclusion of residual salt in porous metal products. A vacuum distillation has been adopted for various molten salt systems and could be applied to the post-treatment process of the electrolytic reduction. The residual salt in the metal products includes LiCl, alkali chlorides, and alkaline earth chlorides. In this paper, vapor pressures of chlorides have been estimated and the composition changes on the residual liquid during the vacuum distillation process have been calculated. A model combining a material balance and vapor-liquid equilibrium relations has been proposed under a constant vapor discharging flow rate and liquid composition changes have been calculated using the vapor pressures with respect to a dimensionless time. The behaviors have been compared with temperature and molten salt composition changes to simulate the process condition variation. The distillation of the residual salt has been dominated by LiCl which is the main component of the salt and CsCl of which vapor pressure is higher than that of LiCl would be readily removed. RbCl exhibits similar vapor pressure with LiCl and maintains its composition. However, $SrCl_2$ and $BaCl_2$ of which vapor pressures are much lower than that of LiCl are concentrated with time and expected to be possibly precipitated during the distillation when the initial compositions are increased.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.105-112
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• 2005

The electrolytic reduction of uranium oxide in a LiCl-Li$_{2}$O molten salt system has been studied in a 10 g U$_{3}$O$_{8}$ /batch-scale experimental apparatus with an integrated cathode assembly at 650$^{\circ}C$. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt$\%$ Li$_{2}$O system and the U$_{3}$O$_{8}$-LiCl-3 wt$\%$ Li$_{2}$O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.17-23
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• 2008

In this study, plasma processing of metal surface is experimentally investigated to enhance the surface decontamination efficiency and to find out the reaction mechanism. Cobalt, the major contaminant in the nuclear facilities, and three fluorine-containing gases, $CF_4/O_2$, $SF_6/O_2$, and $NF_3$ are chosen for the investigation. Thin metallic disk specimens are prepared and their surface etching reactions with the three plasma gases are examined. Results show that the maximum etching rate of $17.2\;{\mu}m/min.$ is obtained with NF3 gas at $420^{\circ}C$, while with $CF_4/O_2$, $SF_6/O_2$ gas plasmas those of $2.56\;{\mu}m/min.$ and $1.14\;{\mu}m/min.$ are obtained, respectively. Along with etching experiments, constituent elements of the reaction products are identified to be cobalt, oxygen, and fluorine by AES (Auger Electron Spectroscopy) analysis. It turns out that the oxygen atoms are physically adsorbed ones to the surface from the ambient not participation ones during the analysis after reaction, which supports that the surface reaction of cobalt is mainly to be a fluorination reaction.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.155-169
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• 2007

A transport container for a 500 kCi tritium storage vessel was developed, which could be used for the transport of metal tritide from Wolsong TRF facility to a disposal site. The structural, thermal, shielding, and confinement analyses were performed for the container in a view of Type B. As a result of structural analysis, the developed container sustained its integrity under normal and accidental conditions. The maximum temperature increase of the inner storage vessel by radiation was evaluated at $134.8^{\circ}C at room temperature. In $800^{\circ}C$ fire test, The thermal barrier of container sustained the inner vessel at $405^{\circ}C after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was $550^{\circ}C. In the evaluation of the shielding, the activity of radiation was nearly zero on the outer surface of inner vessel. Consequently the transport container for a 500 kCi tritium was evaluated to pass all the safety tests including accidental condition, so it was concluded that the designed transport container is proper to be used.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.329-333
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• 2005

To identify the effect of engineered barriers on the leach rate of cesium from spent PWR fuel under a synthetic granitic groundwater, the related leach tests with and without bentonite or metals have been performed up to about 6 years. The leach rates were decreased as a function of leaching time and then became a constant after a certain period. The period in a bare spent fuel was much longer than that with bentonite or metal sheets. The cumulative fraction of cesium released from the spent fuel with bentonite or with copper and stainless steel sheets was steadily increased, but the fraction from bare fuel was rapidly and then sluggishly increased. However, the values deducted its gap inventory from the cumulative fraction of cesium released from the bare fuel was almost very close to the others. These suggest that the initial release of cesium from bare fuel might be dependant on its gap inventory and the effect of engineered barriers on the long-term leach rate of cesium would be insignificant but the rate with engineered barriers could be reduced in the initial transient period due to their retardation effect. And the long-term leach rate of cesium from spent fuel in a repository would be approached to a constant rate of $2\times10^{-2}g/m^2-day$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.335-344
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• 2006

In this paper a computer program was developed, which simulates the Li reduction process of PWR spent fuel, and the amount of a produced metal or chloride compound was calculated at the various amount of Li with the program. It establishes a database, which is composed of some characteristics related to a chemical reaction equation and thermodynamic data, and it calculates the transformed rate of PWR spent fuel oxide at the certain amount of Li by using the database as input data. As the results of the performance test of the program, it was validated that the transformed values of oxides, except for $Eu_2O_3$ and $Sm_2O_3$, were almost the same to within about a 6 % error with those calculated by the previous code and that the calculated amount of Li was also exactly consistent with the theoretical one, which is used for a complete reaction of each oxide in a single chemical reaction. A relationship between Li and the transformed metal of each oxide was analyzed on the basis of the quantities calculated with the verified development program. Of the results, when the amount of Li was given to be 250 mole, the 83.73 percentage of $UO_2$ was transformed into U while the remainder was still to be $UO_2$. In addition, it was appeared that the 297 mole of Li was needed to completely convert $UO_2$ into U.

• 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.

• Economic and Environmental Geology
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• v.56 no.6
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• pp.847-870
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• 2023

The safe disposal of high-level radioactive waste requires accurate predictions of the long-term geochemical behavior of radionuclides. To achieve this, the present study was conducted to model geochemical behaviors of uranium (U), plutonium (Pu), and palladium (Pd) under different hydrogeochemical conditions that represent deep groundwater in Korea. Geochemical modeling was performed for five types of South Korean deep groundwater environment: high-TDS saline groundwater (G1), low-pH CO₂-rich groundwater (G2), high-pH alkaline groundwater (G3), sulfate-rich groundwater (G4), and dilute (fresh) groundwater (G5). Under the pH and Eh (redox potential) ranges of 3 to 12 and ±0.2 V, respectively, the solubility and speciation of U, Pu, and Pd in deep groundwater were predicted. The result reveals that U(IV) exhibits high solubility within the neutral to alkaline pH range, even in reducing environment with Eh down to -0.2 V. Such high solubility of U is primarily attributed to the formation of Ca-U-CO₃ complexes, which is important in both G2 located along fault zones and G3 occurring in granitic bedrocks. On the other hand, the solubility of Pu is found to be highly dependent on pH, with the lowest solubility in neutral to alkaline conditions. The predominant species are Pu(IV) and Pu(III) and their removal is predicted to occur by sorption. Considering the migration by colloids, however, the role of colloid formation and migration are expected to promote the Pu mobility, especially in deep groundwater of G3 and G5 which have low ionic strengths. Palladium (Pd) exhibits the low solubility due to the precipitation as sulfides in reducing conditions. In oxidizing condition, anionic complexes such as Pd(OH)₃^-, PdCl₃(OH)₂^-, PdCl₄^2-, and Pd(CO₃)₂^2- would be removed by sorption onto metal (hydro)oxides. This study will improve the understanding of the fate and transport of radionuclides in deep groundwater conditions of South Korea and therefore contributes to develop strategies for safe high-level radioactive waste disposal.

• Journal of Radiation Industry
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• v.18 no.1
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• pp.43-52
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• 2024

Currently, there are no interim storage facilities and permanent disposal facilities in Korea, so all spent nuclear fuels are temporarily stored. However, the temporary storage facility is approaching saturation, and as a measure to this, the 2nd Basic Plan for the Management of High-Level Radioactive Waste presented an operation plan for dry interim storage facilities and dry temporary storage facilities on the NPP on-site. The dry storage can be operated in various ways, and to select the optimal dry storage method, the reduction of exposure for workers must be considered. Accordingly, it is necessary to develop a worker exposure scenario according to the dry storage method and evaluate and compare the radiological impact for each method. The purpose of this study is to develop an exposure scenario for workers transporting spent nuclear fuel by dry storage method. To this end, first, the operation procedure of the foreign commercial spent nuclear fuel dry storage system was analyzed based on the Final Safety Analysis Report (FSAR). 1) the concrete overpack-based system, 2) the metal overpack-based system, and 3) the vertical storage module-based system were selected for analysis. Factors were assumed that could affect the type of work (working distance, working hours, number of workers, etc.) during transportation work. Finally, the work type of the processes involved in transporting spent nuclear fuel by dry storage method was set, and an exposure scenario was developed accordingly. The concrete overpack method, the metal overpack method, and the vertical storage module method were classified into a total of 31, 9, and 23 processes, respectively. The work distance, work time, and number of workers for each process were set. The product of working hours and number of workers (Man-hour) was set high in the order of concrete overpack method, vertical storage module method, and metal overpack method, and short-range work (10 cm) was most often applied to the concrete overpack method. The results of this study are expected to be used as basic data for performing radiological comparisons of transport workers by dry storage method of spent nuclear fuel.