• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.483-490
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• 2012

PWR spent fuels produced in the Republic of Korea are expected to be recycled by pyroprocess in the long term future. Even though pyroprocess waste amounts can be smaller than that of PWR spent fuel assembly in case of direct disposal, this process essentially will produce various and unique radioactive wastes. The goals of this article are to characterize these wastes, calculate the amount of wastes, design disposal systems for each waste and evaluate the radiation safety of each system by dose assessment. The absorbed dose results of the metal and ceramic waste for the engineering barrier system (EBS) showed $2.21{\times}10^{-2}$ Gy/h and $1.15{\times}10^{-2}$ Gy/h, which are lower than the recommended value of 1 Gy/h. These results confirmed that the newly proposed disposal systems have a safety margin for the radiation produced from each waste.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.233-241
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• 2021

A series of three bench-scale experiments was performed to investigate the conversion of sodium metal to sodium chloride via reactions with non-metal and metal chlorides. Specifically, batches of molten sodium metal were separately contacted with ammonium chloride and ferrous chloride to form sodium chloride in both cases along with iron in the latter case. Additional ferrous chloride was added to two of the three batches to form low melting point consolidated mixtures of sodium chloride and ferrous chloride, whereas consolidation of a sodium-chloride product was performed in a separate batch. Samples of the products were characterized via X-ray diffraction to identify attendant compounds. The reaction of sodium metal with metered ammonium chloride particulate feeds proceeded without reaction excursions and produced pure colorless sodium chloride. The reaction of sodium metal with ferrous chloride yielded occasional reaction excursions as evidenced by temperature spikes and fuming ferrous chloride, producing a dark salt-metal mixture. This investigation into a method for controlled conversion of sodium metal to sodium chloride is particularly applicable to sodium containing elevated levels of radioactivity-including bond sodium from nuclear fuels-in remote-handled inert-atmosphere environments.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.219-228
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• 2012

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.228-233
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• 2013

It has been continuously generated the requirement for the replacement of the main components such as a steam generator due to the deterioration of the nuclear power plant all around the world. Also, a large amount of radioactive metal was generated during the decommissioning in a short period. It is required to make an accurate measurement of the residual radioactivity for recycling the metal waste for releasing from regulatory control. In planning the measurement procedures, the influence of geometry, self-absorption, density and other relevant factors on the representativeness of the measurements should be considered for the decommissioning metal waste. In this study, the method for measurement procedures, the source term evaluation, the ways to secure representative samples, the measurement device for wide area and the self-absorption correction factors for different density were evaluated. The metal samples for measurement were prepared for securing the simple geometry and representative by melting process. The developed correction method for measuring the radioactivity a variety density of metal waste could improve the reliability of the evaluation results for clearance.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.32-37
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• 2005

The characteristics of the metal melting and radionuclide distribution of the radioactive has been investigated in a lab-scale arc furnace. The slag former based on the constituents of silica, calcium oxide, aluminum oxide, borate and calcium fluoride additions was used for melting of the stainless and carbon steel. In the melting of the stainless steel, the amount of slag formation increased with an increase of the concentration of the slag former. But the effects of the slag basicity on the amount of stag formation showed a local maximum value of the slag formation with an increase of the basicity index in the melting of the stainless steel as well as in the melting of the carbon steel. With an increase of the amount of slag former addition, the trends of the cobalt distribution into the ingot and the stag depended on the kind of slag former used in the melting of the stainless steel while the effect of the slag basicity on the distribution of the cobalt was not clarified in the melting of carbon steel. Tn the melting of the carbon steel, the strontium was captured at up to $50\%$ into the slag phase. Cesium was completely eliminated from the melt of the stainless steel as well as the carbon steel and distributed to the dust phase.

• Journal of computational fluids engineering
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• v.21 no.2
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• pp.54-61
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• 2016

When spent fuel assemblies from the reactor of nuclear power plants(NPPs) are transported, the assemblies are exposed to short-term operations that can affect the peak cladding temperature of spent fuel assemblies. Therefore, it needs to perform the analysis of heat transfer on spent fuel dry cask during the operation. For 3 dimensional computational fluid dynamnics(CFD) simulation, it is proposed that the short-term operation is divided into three processes: Wet, dry, and vacuum drying condition. The three processes have different heat transfer mode and medium. Metal transportation cask, which is Korea Radioactive Waste Agency(KORAD)'s developing cask, is evaluated by the methods proposed in this work. During working hours, the boiling at wet process does not occur in the cask and the peak cladding temperatures of all processes remain below $400^{\circ}C$. The maximum peak cladding temperature is $173.8^{\circ}C$ at vacuum drying process and the temperature rise of dry, and vacuum drying process occurs steeply.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.19 no.1
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• pp.121-133
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• 2020

Underwater cutting has a different mechanism than dry cutting, and there are more restrictions than benefits. Due to these constraints, research and development of underwater cutting has been very limited. At present, reactor dismantling is emerging as an important task worldwide, and reactor pressure containers, a key part of the reactor, are decommissioned based on underwater cutting. Reactor pressure containers are high-level radioactive waste, which is one of the main goals of today, such as to bridge the gap between environmental, safety, and cutting performance; hence, a process suitable for cutting should be applied. Therefore, many studies are being conducted on underwater cutting in connection with the dismantling of nuclear reactors in various areas in order to find appropriate processes. This paper first introduces the core technology of underwater cutting processes and discusses various processes. The emphasis is then placed on the adequacy of the reactor dismantling application. More specifically, we examine the suitability for the mechanical and thermal cutting processes, respectively, to find a solution suitable for dismantling a reactor. We discuss how each solution can sufficiently perform the specified functions at each stage of reactor dismantling and suggest that these processes can perform all of the work of underwater cutting.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.379-379
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• 2018

MXenes are a new family of 2D transition metal carbide nanosheets analogous to graphene (Lv et al., 2017; Sun et al., 2018). Due to the easy availability, hydrophilic behavior, and tunable chemistry of MXenes, their use in applications for environmental pollution remediation such as heavy metal adsorption has recently been explored (Li et al., 2017). In this study, three-dimensional (3D) MXene ($Ti_3C_2T_x$) films with high adsorption capacity, good mechanical strength, and high selectivity for specific radionuclide from aquose solution were successfully fabricated by a polymeric precursor method using vacuum-assisted filtration. The highest removal efficiency on the films was 99.54%, 95.61%, and 82.79% for $Sr^{2+}$, $Co^{2+}$, and $Cs^+$, respectively, using a film dosage of 0.06 g/ L in the initial radionuclide solution (each radionuclide concentration = 1 mg/L and pH = 7.0). Especially, the adsorption process reached an equilibrium within 30 min. The expanded interlayer spacing of $Ti_3C_2T_x$ sheets in MXene films showed excellent radionuclide selectivity ($Cs^+$ and/or $Sr^{2+}/Co^{2+}$) (Simon, 2017). Besides, the MXene films was not only able to be easily retrieved from an aqueous solution by filtration after decontamination processes, but also to selectively separate desired target radionuclides in the solutions. Therefore, the newly developed MXene ($Ti_3C_2T_x$) films has a great potential for radionuclide removal from aqueous solution.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.165-169
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• 2004

It was studied on the reductive extraction between the eutectic salt and Bi metal phases. The solutes were zirconium and the rare earth elements, where zirconium was used as the surrogate for the transuranic(TRU) elements. All the experiments were performed in a glove box filled with argon gas. Two types of experimental conditions were used -high and low initial solute concentrations in salt. Li-Bi alloy was used as a reducing agent to reduce the high chemical activity of Li. The reductive extraction characteristics were examined using ICP, XRD and EPMA analysis. Zirconium was successfully separated from the rare earth elements by the reductive extraction method. The LiF-NaF-KF system was favorable among the fluoride salt systems, whereas the LiCl-KCl system was favorable among the chloride salt systems. When the solute concentrations were high, intermetallic compounds were found near the salt-metal interface.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.300-315
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• 2005

Recently plasma surface-cleaning or surface-etching techniques have been focused in the respect of decontamination of spent or used nuclear parts and equipment. In this study decontamination rate of metallic cobalt surface was experimentally investigated via its surface etching rate with a $CF_4-O_2$ mixed gas plasma and metallic surface wastes of cobalt oxides were simulated and decontaminated with $NF_3$ - Ar mixed gas plasma. Experimental results revealed that a mixed etchant gas with about $80{\%}\;CF_4-20{\%}\;O_2$ gives the highest reaction rate of cobalt disk and the rate reaches with a negative 300 DC bias voltage up to $0.43\;{\mu}m$/min at $380^{\circ}C$ and $20{\%}\;NF_3-80\%$ Ar mixed gas gives $0.2\;{\mu}m$/min of reaction rate of cobalt oxide film.