• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.207-215
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• 2020

The feasibility of rare earth (RE) removal process via oxidation reactions with UCl₃ was investigated using the HSC Chemistry code to reduce the concentrations of RE in transuranic (TRU) products. The composition and thermodynamic data of TRU and RE elements contained in the reference spent fuel were examined. The reactivity was evaluated by calculating equilibrium data considering oxidation reactions with UCl₃. Both RE removal rate and TRU recovery rate were evaluated for the two cases, wherein TRU products with different RE concentrations were used. When TRU products were reacted with UCl₃, selective oxidation was driven by the difference in the Gibbs free energy of each element. The calculation results imply that the TRU/RE ratio of the final product can be increased by removing RE elements while maintaining the maximum recovery rate of TRU, which is accomplished by controlling the amount of UCl₃ injected. Since the results of this study are based on thermodynamic equilibrium data, there are many limitations to apply to the actual process. However, it is expected to be used as an important data for the process design to supply the TRU product of pyroprocessing to SFR's fuel demanding low RE concentrations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.746-751
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• 2003

A method of measuring the length of defects on the wall of the spent nuclear fuel pool using the image processing and a laser slit beam is proposed. Since the defect monitoring camera is suspended by a crane and hinged to the crane hook, the camera viewing direction can not be adjusted to the orientation that is exactly perpendicular to the wall. Thus, the image taken by the camera, which is horizontally rotated along the axis of the camera supporting beam, is distorted and thus, the precise length can not be measured. In this paper, by using the LASER slit beam generator, the horizontally rotated angle of the camera is estimated. Once the angle is obtained, the distorted image can be easily reconstructed to the image normal to the wall. The estimation algorithm adopts a 3-dimensional coordinate transformation of the image plane where both the laser slit beam and the original image of the defects exist. The estimation equation is obtained by using the information of the beam projected on the wall and the parameters of this equation are experimentally obtained. With this algorithm, the original image of the defect taken at arbitrary rotated angle can be reconstructed to an image normal to the wall. From the result of a series of experiments, the accuracy of the defect is measured within 0.6 and 1.3 % error bound of real defect size in the air and underwater, respectively under 30 degree of the inclined angle of the laser slit beam generator. Also, the error increases as the inclined angle increases upto 60 degree. Over this angle, the defect length can not be measured since the defect image disappears. The proposed algorithm enables the accurate measurement of the defect length only by using a single camera and a laser slit beam.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.41-47
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• 2005

Preparation of ferrocyanide-anion exchange resin and adsorption test of the prepared resin on the Cs$^{+}$$ion were performed. Adsorption capability of the prepared resin on the Cs$^{+}$ion in the simulated citric acid based soil decontamination waste solution was 4 times greater than that of the commercial cation exchange resin. Adsorption equilibrium of the prepared resin on the Cs$^{+}$ion reached within 360 minutes. Adsorption capability on the Cs$^{+}$ion became to decrease above the necessary Co$^{2+}$ion concentration in the experimental range. Recycling test of the spent ion exchange resin by the successive application of hydrogen peroxide and hydrazine was also performed. It was found that desorption of Cs$^{+}$ion from the resin occurred to satisfy the electroneutrality condition without any degradation of the resin.

• Nuclear industry
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• v.4 no.6 s.22
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• pp.19-23
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• 1984

현재 우리나라에는 3기의 원전이 운전중에 있으며 6기가 건설중에 있다. 즉, 2기의 가압경수로(PWR)의 1기의 가압중수로(HWR)가 운전중에 있으며 건설중인 것은 모두 PWR이다. 현재 추진되고 있는 원전계획을 수행하면, 1990년에는 사용후핵연료가 년간 약200톤씩 방출될 것이며, 2000년에는 500톤 정도가 방출되게 된다. 이와 같은 핵사용후연료는 현재 소내저장하고 있으나 '90년대 중반 부터는 소내저장용량이 한계에 달하게 될 것으로 전망되고 있다. 따라서 본교에서의 이와 같은 소내저장 한계성에 대처할 수 있는 가능한 방안을 검토하고, 이를 경제성 측면에서 분석하고자 하였다. 사용후핵연료의 관리방안에 대한 경제성 분석을 위해서는 장래의 원전계획, 원자로형 및 핵연료주기방식 등에 대한 여러 가지 가정이 필요하게 된다. 원전계획은 정부에서 발표한 $\ulcorner$5차5개년 수정계획$\lrcorner$에 의거하여 원전시설용량은 현재의 2GWe에서 2000년에는 22GWe로, 2025년에는 44GWe로 늘어나는 것으로 보았다. 이와 같은 원전계획을 바탕으로 6가지 핵연료주기에 관한 시나리오를 설정하였다. 즉, 사용후핵연료를 비순환방식으로 운영하는 2가지 경우, 순환방식으로 운영되는 3가지 경우 그리고 FBR에 활용하는 1가지의 경우에 대하여 검토하였다. 사용후핵연료의 관리방식에 따른 장기적인 안목에서의 경제성 분석은 핵연료주기비용 뿐만아니라 원전의 투자비도 함께 분석하는 것이 합리적이며, 따라서 본교에서는 계획기간 동안의 6가지 시나리로에 따른 원전 및 핵연료주기에 관한 총 투자비를 비교하였고, 1982년 가격으로 현가화한 단가도 비교${\cdot}$검토 하였다. 이와 같은 6가지 시나리오에 대한 경제성을 비교해 본 결과, 핵연료주기선택의 경제성평가에 큰 영향을 주는 핵연료주기요소는 재처리비, 재처리시 방출되는 폐기물의 처리${\cdot}$처분비 그리고 사용후핵연료 저장방식으로 판명되었으며 6가지 시나리오에 대한 경제성 비교평가 결과, 다음과 같은 결론을 얻었다. 단기적인 안목에서는 소내저장용량을 확장하는 방안이 가장 바람직하며, 중기적인 관점에서는 소외집중저장설비가 활발히 수행되는 시점에서는 사용후핵연료를 재처리하여 재활용하는 방안도 강구되어야 할 것이다.

• Korean Journal of Materials Research
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• v.26 no.12
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• pp.681-687
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• 2016

Zirconium(Zr) nuclear fuel cladding tubes are made using a three-time pilgering and annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zr is dissolved in HF and $HNO_3$ mixed acid during the process and pickling waste acid, including dissolved Zr, is totally discarded after being neutralized. In this study, the waste acid was recycled by adding $BaF_2$, which reacted with the Zr ion involved in the waste acid; $Ba_2ZrF_8$ was subsequently precipitated due to its low solubility in water. It is very difficult to extract zirconium from the as-recovered $Ba_2ZrF_8$ because its melting temperature is $1031^{\circ}C$. Hence, we tried to recover Zr using an electrowinning process with a low temperature molten salt compound that was fabricated by adding $ZrF_4$ to $Ba_2ZrF_8$ to decrease the melting point. Change of the Zr redox potential was observed using cyclic voltammetry; the voltage change of the cell was observed by polarization and chronopotentiometry. The structure of the electrodeposited Zr was analyzed and the electrodeposition characteristics were also evaluated.

• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.117-121
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• 2016

Background: Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Materials and Methods: Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. Results and Discussion: The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. Conclusion: The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.388-394
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• 2003

To handle the high level radioactive materials such a spent fuel, the master-slave manipulators (MSM) are widely used as a remote handling device in nuclear facilities such as the hot cell with sealed and shielded space. In this paper, the Digital Mockup which simulates the remote operation of the Advanced Conditioning Process(ACP) is developed. Also, the workspace and the motion of the slave manipulator, as well as, the remote operation task should be analyzed. The process equipment of ACP and Maintenance/Handling Device are drawn in 3D CAD models using IGRIP. Modeling device of manipulator is assigned with various mobile attributes such as a relative position, kinematics constraints, and a range of mobility, The 3D graphic simulator using the external input device of space ball displays the movement of manipulator. To connect the external input device to the graphic simulator, the interface program of external input device with 6 DOF is deigned using the Low Level Tele-operation Interface(LLTI). The experimental result shows that the developed simulation system gives much-improved human interface characteristics and shows satisfactory response characteristics in terms of synchronization speed. This should be useful for the development of work's education system in the virtual environment.

• Journal of Digital Contents Society
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• v.5 no.1
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• pp.22-27
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• 2004

To handle the high level radioactive materials such a spent fuel, the master-slave manipulaters (MSM) are wide1y used as a remote handling device in nuclear facilities such as the hot cell with sealed and shielded space. In this paper, the Digital Mockup which simulates the remote operation of the Advanced Conditioning Process(ACP) is developed. Also, the workspace and the motion of the slave manipulator, as well as, the remote operation task should be analyzed. The process equipment of ACP and Maintenance/Handling Device are drawn in 3D CAD model using IGRIP. Modeling device of manipulator is assigned with various mobile attributes such as a relative position, kinematics constraints, and a range of mobility. The 3D graphic simulator using the extermal input device of spare ball displays the movement of manipulator. To connect the exterral input device to the graphic simulator, the interface program of external input device with 6 DOF is deigned using the Low Level Tele-operation Interface(LLTI). The experimental result show that the developed simulation system gives much-improved human interface characteristics and shows satisfactory reponse characteristics in terms of synchronization speed. This should be useful for the development of work`s education system in the environment.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.83-83
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• 2017

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. Distillation process was employed for the cathode processing. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. In this study, a mesh-covered crucible was investigated for the sat distillation of electrorefiner uranium deposits. A liquid salt separation step and a vacuum distillation step were combined for salt separation. The adhered salt in uranium deposits was efficiently removed in the mesh-covered crucible. The salt distiller was operated simply since repeated cooling - heating step was not necessary for the change of the crucible. The operation time could be reduced by the use of the mesh-covered crucible and the combined operation of the two steps. A method to preserve a vacuum level was proposed by double O-rings during the operation of the distiller with the mesh-covered crucible. After the salt distillation, the salt content was measured and was below 0.1wt% after the salt distillation. The residual salt after the salt distillation can be removed further during melting of uranium metal.