• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.127-131
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• 2022

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.219-230
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• 2004

Most of the low-level liquid radioactive wastes generated from PWR plants are classified into high or low total suspended solid(HTDS or LTDS), and into radiochemical and radioactive laundry waste. Although the evaporation process has a high decontami- nation ability, it has several problems such as corrosion, foam, and congestion. A new liquid waste disposal process using the ion-exchange demineralizer(IED), instead of the current evaporation process, has been introduced into the Yonggwang NPP #5 and 6. These two methods have been compared to understand the differences in this study. Aspects compared here were the released radioactivity amount of the liquid radioactive wastes, the dose of off-site residents, the decontamination factor, and the amount of the solid radioactive wastes. The IED system is designed to discharge higher radioactivity about 20% than the evaporating system, and the actual radioactivity released from the evaporating and IED system were 0.473mCi and 1.098mCi, respectively. The radioactivity released from the IED was 2.32 times higher than that of the evaporating system. The dose of off-site residents was $2.97{\times}10^{-6}$mSv for the evaporating system, and $6.47{\times}10^{-6}$mSv for IED. The decontamination factor(DF) of the evaporator is, in most cases, far lower than the lower limits of detection(LLD) with the Ge-Li detector. Due to the low concentration of the liquid wastes collected from the liquid waste system, the decontamination factor of IED is very low. Since there is not enough data on the amount of solid radioactive wastes generated by the evaporation system, the comparison on these two systems has been conducted on the basis of the design, and the comparison result was that the evaporating system generated more wastes about 40% than IED.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.

• Journal of Radiation Protection and Research
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• v.24 no.2
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• pp.101-108
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• 1999

The influence of the relative humidity, the temperature and the velocity of supply air on evaporation rate has been studied with non-boiling forced evaporation system in order to treat very low level radioactive liquid wastes produced from the decontamination and decommissioning activities. Experimental data on the evaporation rate have been obtained with the divers variables and experimental equation of air velocity was also obtained by the correlation of those data. The decontamination factor of this system was also obtained by the experimental data from a simulated liquid waste containing Cs-137 radio isotope ; $DF=10^4$. Since the commercial system will be operated for the treatment of the very low level radioactive liquid waste produced from decontamination & decommissioning of TRIGA Mark-II&III research reactor, the environmental assessment has been conducted to improve the operational safety. Exposure dose rate for an individual member of general public was assessed, and it showed that it was very lower than individual dose limits. The release of radioactivity of radioisotope material (Cs-137) to the environment was assessed, and result showed that it was $4.637{\times}10^{-14}\;{\mu}Ci/cc$.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1275-1276
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.69-69
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• 2019

The distillation of liquid cathode is necessary to separate cadmium from the actinide elements in the pyroprocessing since the actinide deposits are dissolved or precipitated in a liquid cathode. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. Several methods have been proposed to lower the splattering of cadmium during distillation. One of the important methods is an installation of crucible cover on the distillation crucible. A multi-layer porous round cover was proposed to avoid a cadmium splattering in our previous study. In this study, the effect of crucible cover on the cadmium distillation was examined to develop a splatter shield. Various surrogates were used for the actinides in the cadmium. The surrogates such as bismuth, zirconia, and tungsten don't evaporate at the operational temperature of the Cd distiller due to their low vapor pressures. The distillation experiments were carried out in a crucible equipped with cover and in a crucible without cover. About 40 grams of Cd was distilled at a reduced pressure for two hours at various temperatures. The mixture of the cadmium and the surrogate was heated at $470{\sim}620^{\circ}C$. Most of the bismuth remained in the crucible equipped with cover after distillation under $580^{\circ}C$ for two hours, whereas small amount of bismuth decreased in the crucible without cover above $580^{\circ}C$. The liquid bismuth escaped with liquid cadmium drop from the crucible without cover. It seems that the crucible cover played a role to prevent the splash of the liquid cadmium drop. The effect of the cover was not clear for the tungsten or zirconia surrogate since the surrogates remained as a solid powder at the experimental temperature. From the results of this work, it can be concluded that the crucible cover can be used to minimize the deposit loss by prevention of escape of liquid drop from the crucible during distillation of liquid cathode.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.103-103
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• 2018

The liquid cathode processing is necessary to separate cadmium from the actinide elements in the pyroprocessing since the actinide deposits are dissolved or precipitated in a liquid cathode. Distillation process was employed for the cathode processing owing to the compactness. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. Several methods have been proposed to lower the splattering of cadmium during distillation. A multi-layer porous round cover was proposed to avoid a cadmium splattering in our previous study. In this study, distillation behavior of $Cd-ZrO_2$ and Cd - Bi systems were investigated to examine a multi-layer porous round cover for the development of the cadmium splatter shield of distillation crucible. It was designed that the cadmium vapor can be released through the holes of the shield, whereas liquid drops can be collected in the multiple hemisphere. The cover was made with three stainless steel round plates with a diameter of 33.50 mm. The distance between the hemispheres and the diameter of the holes are 10 and 1 mm, respectively. Bismuth or zirconium oxide powder was used as a surrogate for the actinide elements. About 40 grams of Cd was distilled at a reduced pressure for two hours at various temperatures. The mixture of the cadmium and the surrogate was distilled at 470, 570 and $620^{\circ}C$ in the crucible with the cover. Most of the bismuth or zirconia remained in the crucible after distillation at 470 and $570^{\circ}C$ for two hours. It was considered that the crucible cover hindered the splattering of the liquid cadmium from the distillation crucible. A considerable amount of the surrogate material reduced after distillation at $620^{\circ}C$ due to the splattering of the liquid cadmium. The low temperature is favorable to avoid a liquid cadmium splattering during distillation. However, the optimum temperature for the cadmium distillation should be decided further, since the evaporation rate decreases with a decreasing temperature.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.1
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• pp.85-96
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• 2021

Radioactive mixed waste (RMW) means waste mixed with radioactive substances and hazardous substances. In Korea, there are definitions and disposal restrictions on RMW in the Nuclear Safety Management Act, but it is difficult to apply because the contents are insufficient, so this paper proposed applicable management standards. The main RMW generated from nuclear power plants is waste oil, waste asbestos, PCB, and waste fluorescent liquid, and their radiation characteristics are mostly at very low levels and some are estimated at low levels. In addition to nuclear power plants, RMW also occurs in research institutes, industries, and hospitals. The acceptance criteria of all disposal facilities in the world basically prohibit disposal of RMW unless the hazardous substances of RMW are removed or mitigated below the standard value. Cases in Korea, the United States, Japan and Europe were reviewed to propose the RMW management standards in Korea. With reference to the results of the above review, this paper clearly defined RMW and proposed detailed management standards for the separation, storage, treatment and disposal of hazardous substances by applying the Waste Control Act. It also mentioned legislation of management standards, regulatory methods, and acceptance criteria of disposal facility operator.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.73-73
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• 2017

Pyroprocessing at KAERI (Korea Atomic Energy Research Institute) consists of pretreatment, electroreduction, electrorefining and electrowinning. SFR (Sodium Fast Reactor) fuel is prepared from the electrowinning process which is composed of LCC (Liquid Cadmium Process) and Cd distillation et al. LCC is an electrochemical process to obtain actinides from spent fuel. In order to recover actinides inert anodes such as carbon material are used, where chlorine gas ($Cl_2$) evolves on the surface of the carbon material. And, stainless steel (SUS) crucible should be installed in large-scale electrowinning system. Therefore, the effect of chlorine on the SUS material needs to be studied. LiCl-KCl-$UCl_3$-$NdCl_3$-$CeCl_3$-$LaCl_3$-$YCl_3$ salt was contained in 2 kinds of electrolytic crucible having an inner diameter of 5cm, made of an insulated alumina and an SUS, respectively. And, three kinds of electrodes such as cathode, anode, reference were used for the electrochemical experiments. Both solid tungsten (W) and LCC were used as cathodes. Cd of 45 g as the cathode material was contained in alumina crucibles for the deposition experiments, where the crucible has an inner diameter of 3 cm. Glassy carbon rod with the diameter of 0.3 cm was employed as an anode, where shroud was not used for the anode. A pyrex tube containing LiCl-KCl-1mol% AgCl and silver (Ag) wire having a diameter of 0.1cm was used as a reference electrode. Electrodeposition experiments were conducted at $500^{\circ}C$ at the current densities of $50{\sim}100mA/cm^2$. In conclusion, Fe ions were produced in the salt during the electrodeposition by the reaction of chlorine evolved from the anode and Fe of the SUS crucible and thereby LCC system using SUS crucible showed very low current efficiencies compared with the system using the insulated alumina crucible. Anode shroud needs to be installed around the glassy carbon not to influence surrounding SUS material.